Mr. Bryan C. Hanson 

Senior Vice President 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

May 19, 2016 

Exelon Generation Company, LLC 

President and Chief Nuclear Officer (CNO) 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

SUBJECT: BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT 

NOS. 1 AND 2, CLOSEOUT OF GENERIC LETTER 2004-02, "POTENTIAL 

IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION 

DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED-WATER 

REACTORS" (CAC NOS. MC4667, MC4668, MC4669, AND MC4670) 

Dear Mr. Hanson: 

The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 2004-02, "Potential 

Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at 

Pressurized-Water Reactors" (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML042360586) dated September 13, 2004, requesting that licensees 

address the issues raised by Generic Safety Issue 191, "Assessment of Debris Accumulation on 

Pressurized Water Reactor Sump Performance." 

The stated purpose of GL 2004-02 was focused on demonstrating compliance with Title 10 of 

the Code of Federal Regulations (10 CFR) Part 50 Section 50.46, "Acceptance Criteria for 

Emergency Core Cooling Systems (ECCSs) for Light Water Reactors." Specifically, the GL 

requested addressees to perform an evaluation of the ECCS and containment spray system 

(CSS) recirculation and, if necessary take additional action to ensure system function, in light of 

the potential susceptibility of pressurized-water reactor sump screens to blockage during a 

design-basis accident (OBA) requiring recirculation operation of the ECCS or CSS, and on the 

potential for additional adverse effects due to debris blockage of flowpaths necessary for ECCS 

and CSS recirculation and containment drainage. 

By letter dated March 7, 2005 (ADAMS Accession No. ML050670026), as supplemented by 

letters dated July 27, 2005 (ADAMS Accession No. ML052140271), September 1, 2005 

(ADAMS Accession No. ML052450029), May 31, 2006 (ADAMS Accession No. ML061520299) 

December 31, 2007 (ADAMS Accession No. ML080280562). July 15, 2008 (ADAMS Accession 

No. ML081980241), September 19, 2008 (ADAMS Accession No. ML082660245), May 29, 

2009 (ADAMS Accession No. ML091520737), October 30, 2015 (ADAMS Accession No. 

ML 15303A408), and April 15, 2016 (ADAMS Accession No. ML 16106A219), Exelon Generation 

Company, LLC (the licensee) provided responses to GL 2004-02 for Braidwood Station, Units 1 

and 2 (Braidwood), and Byron Station, Unit Nos. 1 and 2 (Byron). 

The NRC staff has completed its review of the licensee responses and supplements associated 

with the GL 2004-02 for Braidwood and Byron. Based on its review, the NRC staff finds that the 

information provided demonstrates that debris will not inhibit the ECCS) or the CSS 
B. Hanson - 2 -

performance of its intended safety function to assure adequate long term core cooling following 

a OBA in accordance with 10 CFR 50.46, "Acceptance criteria for ECCSs for light-water nuclear 

power reactors." Therefore, the NRC staff concludes that the licensee's responses to GL 2004-

02 are adequate and considers GL 2004-02 closed for Braidwood and Byron. 

Enclosed is the summary of the NRC staff's review. If you have any questions, please call me 

at 301-415-6606 or via e-mail at Joel.Wiebe@nrc.gov. 

Docket Nos. STN 50-456, STN 50-457, 

STN 50-454 and STN 50-455 

Enclosure: 

NRC review for Braidwood and Byron 

cc w/enclosure: Distribution via Listserv 

Sincerely, 

~oel S. Wiebe, Senior Project Manager 

Plant Licensing Branch 111-2 

Division of Operating Reactor Licensing 

Office of Nuclear Reactor Regulation 
U.S. NUCLEAR REGULATORY COMMISSION (NRC) STAFF REVIEW 

OF THE DOCUMENTATION PROVIDED BY 

EXELON GENERATION COMPANY, LLC 

FOR THE BRAIDWOOD STATION UNITS 1 AND 2 AND 

FOR BYRON STATION UNITS 1 AND 2 

DOCKET NOS. 50-456. 50-457. 50-454, 50-455 

CONCERNING RESOLUTION OF GENERIC LETTER 2004-02 

POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING 

DESIGN BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS 

1.0 INTRODUCTION 

A fundamental function of the emergency core cooling system (ECCS) is to recirculate water 

that has collected at the bottom of the containment through the reactor core following a break in 

the reactor coolant system (RCS) piping to ensure long-term removal of decay heat from the 

reactor fuel. Leaks from the RCS, hypothetical scenarios known as loss-of-coolant accidents 

(LOCAs), are part of every plant's design-basis. Hence, nuclear plants are designed and 

licensed with the expectation that they are able to remove reactor decay heat following a LOCA 

to prevent core damage. Long-term cooling following a LOCA is a basic safety function for 

nuclear reactors. The recirculation sump provides a water source to the ECCS in a 

pressurized-water reactor (PWR) once the primary water source has been depleted. 

If an LOCA occurs, piping thermal insulation and other materials may be dislodged by the 

two-phase coolant jet emanating from the broken RCS pipe. This debris may transport, via 

flows coming from the RCS break or from the containment spray system (CSS), to the pool of 

water that collects at the bottom of containment following a LOCA. Once transported to the 

sump pool, the debris could be drawn towards the ECCS sump strainers, which are designed to 

prevent debris from entering the ECCS and the reactor core. If this debris were to clog the 

strainers and prevent coolant from entering the reactor core, containment cooling could be lost 

and result in core damage and containment failure. 

It is also possible that some debris would bypass the sump strainer and lodge in the reactor 

core. This could result in reduced core cooling and potential core damage. If the ECCS strainer 

were to remain functional, even with core cooling reduced, containment cooling would be 

maintained and the containment function would not be adversely affected. 

Findings from research and industry operating experience raised questions concerning the 

adequacy of PWR sump designs. Research findings demonstrated that, compared to other 

LOCAs, the amount of debris generated by a high-energy line break (HELB) could be greater. 

The debris from a HELB could also be finer (and thus more easily transportable) and could be 

Enclosure 
- 2 -

comprised of certain combinations of debris (i.e., fibrous material plus particulate material) that 

could result in a substantially greater flow restriction than an equivalent amount of either type of 

debris alone. These research findings prompted the NRC to open Generic Safety Issue (GSI) -

191, "Assessment of Debris Accumulation on PWR Sump Performance," in 1996. This resulted 

in new research for PWRs in the late 1990s. 

GSl-191 focuses on reasonable assurance that the provisions of Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.46(b)(5), are met. This deterministic rule requires 

maintaining long-term core cooling after initiation of the ECCS. The objective of GSl-191 is to 

ensure that post-accident debris blockage will not impede or prevent the operation of the ECCS 

and CSS in recirculation mode at PWRs during LOCAs or other HELB accidents for which sump 

recirculation is required. The NRC completed its review of GSl-191 in 2002 and documented 

the results in a parametric study that concluded that sump clogging at PWRs was a credible 

concern. 

GSl-191 concluded that debris clogging of sump strainers could lead to recirculation system 

ineffectiveness as a result of a loss of net positive suction head (NPSH) for the ECCS and CSS 

recirculation pumps. Resolution of GSl-191 involves two distinct but related safety concerns: 

(1) potential clogging of the sump strainers that results in ECCS and/or CSS pump failure; and 

(2) potential clogging of flow channels within the reactor vessel because of debris bypass of the 

sump strainer (in-vessel effects). Clogging at either the strainer or in-vessel channels can result 

in loss of the long-term cooling safety function. 

After completing the technical assessment of GSl-191, the NRC issued Bulletin 03-01, 

"Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water 

Reactors" (Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML031600259), on June 9, 2003. The Office of Nuclear Reactor Regulation (NRR) 

requested and obtained the review and endorsement of the bulletin from the Committee to 

Review Generic Requirements (CRGR) (ADAMS Accession No. ML031210035). As a result of 

the emergent issues discussed in Bulletin 03-01, the NRC staff requested an expedited 

response from PWR licensees on the status of their compliance of regulatory requirements 

concerning the ECCS and CSS recirculation functions based on a mechanistic analysis. The 

NRC staff asked licensees, who chose not to confirm regulatory compliance, to describe any 

interim compensatory measures that they had implemented or will implement to reduce risk until 

the analysis could be completed. All PWR licensees responded to Bulletin 03-01. The NRC 

staff reviewed all licensees' Bulletin 03-01 responses and found them acceptable. 

In developing Bulletin 03-01, the NRC staff recognized that it might be necessary for licensees 

to undertake complex evaluations to determine whether regulatory compliance exists in light of 

the concerns identified in the bulletin and that the methodology needed to perform these 

evaluations was not currently available. As a result, that information was not requested in 

Bulletin 03-01, but licensees were informed that the NRC staff was preparing a Generic Letter 

(GL) that would request this information. GL 2004-02, "Potential Impact of Debris Blockage on 

Emergency Recirculation During Design-basis Accidents at Pressurized-Water Reactors," dated 

September 13, 2004 (ADAMS Accession No. ML042360586), was the follow-on information 

request referenced in Bulletin 03-01. This document set the expectations for resolution of PWR 

sump performance issues identified in GSl-191, to ensure the reliability of the ECCS and CSS 

at PWRs. NRR requested and obtained the review and endorsement of the GL from the CRGR 

(ADAMS Accession No. ML040840034). 
- 3 -

GL 2004-02 requested that addressees perform an evaluation of the EGGS and CSS 

recirculation functions in light of the information provided in the letter and, if appropriate, take 

additional actions to ensure system function. Additionally, addressees were requested to 

submit the information specified in GL 2004-02 to the NRG. The request was based on the 

identified potential susceptibility of PWR recirculation sump screens to debris blockage during 

design-basis accidents (DBAs) requiring recirculation operation of EGGS or CSS and on the 

potential for additional adverse effects due to debris blockage of flow paths necessary for EGGS 

and CSS recirculation and containment drainage. GL 2004-02 required addressees to provide 

the NRG a written response in accordance with 10 CFR 50.54(f). 

By letter dated May 28, 2004 (ADAMS Accession No. ML041550661), the Nuclear Energy 

Institute (NEI) submitted a report describing a methodology for use by PWRs in the evaluation 

of containment sump performance. NEI requested that the NRG review the methodology. The 

methodology was intended to allow licensees to address and resolve GSl-191 issues in an 

expeditious manner through a process that starts with a conservative baseline evaluation. The 

baseline evaluation serves to guide the analyst and provide a method for quick identification and 

evaluation of design features and processes that significantly affect the potential for adverse 

containment sump blockage for a given plant design. The baseline evaluation also facilitates 

the evaluation of potential modifications that can enhance the capability of the design to address 

sump debris blockage concerns and uncertainties and supports resolution of GSl-191. The 

report offers additional guidance that can be used to modify the conservative baseline 

evaluation results through revision to analytical methods or through modification to the plant 

design or operation. 

By letter dated December 6, 2004 (ADAMS Accession No. ML043280641), the NRG issued an 

evaluation of the NEI methodology. The NRG staff concluded that the methodology, as 

approved in accordance with the NRG staff safety evaluation (SE), provides an acceptable 

overall guidance methodology for the plant-specific evaluation of the EGGS or CSS sump 

performance following postulated DBAs. 

In response to the NRG staff SE conclusions on NEI 04-07 "Pressurized Water Reactor 

Sump Performance Evaluation Methodology" (ADAMS Accession Nos. ML050550138 and 

ML050550156), the Pressurized Water Reactor Owners Group (PWROG) sponsored the 

development of the following Westinghouse Commercial Atomic Power (WCAP) Topical 

Reports (TRs): 

• TR-WCAP-16406-P-A, "Evaluation of Downstream Sump Debris Effects in Support of 

GSl-191," Revision 1 (not publicly available), to address the effects of debris on piping 

systems and components. 

• TR-WCAP-16530-NP-A, "Evaluation of Post-accident Chemical Effects in Containment 

Sump Fluids to Support GSl-191," issued March 2008 (ADAMS Accession No. 

ML081150379), to provide a consistent approach for plants to evaluate the chemical 

effects that may occur post-accident in containment sump fluids. 

• TR-WCAP-16793-NP-A, "Evaluation of Long-Term Cooling Considering Particulate, 

Fibrous and Chemical Debris in the Recirculating Fluid," Revision 2, issued July 2013 

(ADAMS Accession No. ML 13239A 114), to address the effects of debris on the reactor 

core. 
- 4 -

The NRC staff reviewed the TRs and found them acceptable to use (as qualified by the 

limitations and conditions stated in the respective SEs}. 1 A more detailed evaluation of how the 

TRs were used by the licensee is contained in the evaluations below. 

After the NRC staff evaluation of licensee responses to GL 2004-02, the NRC staff found that 

there was a misunderstanding between the industry and the NRC on the level of detail 

necessary to respond to GL 2004-02. The NRC staff in concert with stakeholders developed a 

content guide for responding to requests for additional information (RAls) concerning 

GL 2004-02. By letter dated August 15, 2007 (ADAMS Accession No. ML071060091 }, the NRC 

issued the content guide describing the necessary information to be submitted to allow the NRC 

staff to verify that each licensee's analyses, testing, and corrective actions associated with 

GL 2004-02 are adequate to demonstrate that the ECCS and CSS will perform their intended 

function following any OBA. By letter dated November 21, 2007 (ADAMS Accession No. 

ML073110389), the NRC issued a revised content guide. 

The content guide described the following information needed to be submitted to the NRC: 

• corrective actions for GL 2004-02, 

• break selection, 

• debris generation/zone of influence (ZOI) (excluding coatings}, 

• debris characteristics, 

• latent debris, 

• debris transport, 

• head loss and vortexing, 

• NPSH, 

• coatings evaluation, 

• debris source term, 

• screen modification package, 

• sump structural analysis, 

• upstream effects, 

• downstream effects - components and systems, 

• downstream effects -fuel and vessel, 

• chemical effects, and 

• licensing basis 

Based on the interactions with stakeholders and the results of the industry testing, the NRC staff 

in 2012 developed three options to resolve GSl-191. These options were documented and 

proposed to the Commission in SECY-12-0093, "Closure Options for Generic Safety 

Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump 

1 Safety Evaluation by the Office of Nuclear Reactor Regulation, Topical Report (TR) WCAP-16406-P, Revision 1, "Evaluation of 

Downstream Sump Debris Effects in Support of GSl-191," Pressurized Water Reactors Owners Group Project No. 694 (ADAMS 

Accession No. ML073520295); Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report 

(TR), WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191" (ADAMS 

Accession No. ML073521072) ; Final Safety Evaluation by the Office of Nuclear Reactor Regulation, Topical Report WCAP-16793-

NP, Revision 2, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid," 

Pressurized Water Reactor Owners Group Project No. 694 (ADAMS Accession No. ML 13084A 154). 
- 5 -

Performance," dated July 9, 2012 (ADAMS Accession No. ML 121320270). The options are 

summarized as follows: 

• Option 1 would require licensees to demonstrate compliance with 10 CFR 50.46, 

"Acceptance criteria for emergency core cooling systems for light-water nuclear power 

reactors," through approved models and test methods. These will be low fiber plants 

with less than 15 grams of fiber per fuel assembly. 

• Option 2 requires implementation of additional mitigating measures and allows additional 

time for licensees to resolve issues through further industry testing or use of a risk 

informed approach. 

o Option 2 Deterministic: Industry to perform more testing and analysis and submit 

the results for NRC review and approval (in-vessel only). 

o Option 2 Risk Informed: Use the South Texas Project pilot approach currently 

under review with NRR staff. 

• Option 3 involves separating the regulatory treatment of the sump strainer and in-vessel 

effects. 

The options allowed industry alternative approaches for resolving GSl-191. The Commission 

issued a Staff Requirement Memorandum on December 14, 2012 (ADAMS Accession No. 

ML 12349A378), approving all three options for closure of GSl-191. 

By letter dated May 14, 2013 (ADAMS Accession No. ML 13135A069), Exelon Generation 

Company, LLC (the licensee) stated that they will pursue Option 1 for the closure of GSl-191 

and GL 2004-02 for the Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 

(Braidwood and Byron). 

The following is a list of documentation provided by the licensee in response to GL 2004-02: 

RESPONSES TO GL 2004-02 

DOCUMENT DATE ACCESSION NUMBER DOCUMENT 

March 7, 2005 ML050670026 Initial Response to GL 

June 3, 2005 ML051520515 1st NRC RAI 

Julv 27, 2005 ML052140271 Licensee Response to RAI 

September 1, 2005 ML052450029 Supplemental Information 

Februarv 9, 2006 ML060390378 2nd NRC RAI 

May 31, 2006 ML061520299 Supplemental Information 

December 31, 2007 ML080280562 Supplemental Information 

Julv 15, 2008 ML081980241 Completion of Activities 

July 24, 2008 ML081930604 1st NRC Sump Structural 

Analysis RAI and No Further 

Questions Letter 

September 19, 2008 ML082660245 Licensee Response to RAI 

April 15, 2009 ML090960135 2nd NRC Sump Structural 

Analysis RAI 
- 6 -

May 29, 2009 ML091520131 Licensee Response to RAI 

May 14, 2013 ML 13135A069 Path Forward for Resolution 

October 30, 2015 ML 15303A408 Licensee Response to 

In-vessel Down Stream 

Effects 

April 15, 2016 ML 16106A219 Supplemental Information 

The NRC staff reviewed the licensee's responses through December 31, 2007, and by letter 

dated July 24, 2008 (ADAMS Accession No. ML081930604), concluded that the NRC staff had 

no further questions regarding the licensee's completion of corrective actions for GL 2004-02, 

with the exception of structural issues concerning the new sump strainer, and the demonstration 

that in-vessel downstream effects issues are resolved. The NRC staff's conclusion is based on 

the very low debris loading at Byron and Braidwood, as discussed in the licensee's RAI 

responses, and supplemental information. 

2.0 SUMP STRUCTURAL ANALYSIS 

The objective of the sump structural analysis section is to verify the structural adequacy of the 

sump strainer including seismic loads and loads due to differential pressure, missiles, and jet 

forces. 

NRC Staff Review: 

The NRC staff's original review of Section 3k, Sump Structural Analysis, of the licensee's 

December 31, 2007, submittal, led to the conclusion that the licensee did not adequately 

address the information requested by the Revised Content Guide for GL 2004-02 Item 2(d)(vii). 

In its submittal, the licensee stated that an analysis was performed for the ECCS sump strainer 

components using ANSYS finite element program software. The licensee stated the design 

requirements for the ECCS sump strainer ensured structural integrity for the following loads: the 

dead weight of the strainer, full debris loading, differential pressure, and seismic load for 

operating basis earthquake and safe-shutdown earthquake events, and the load combinations 

were provided. The licensee provided the stress ratios for the main sump components, which is 

a comparison of the maximum design stress to the allowable stress limits. All of the results 

were less than one, with the exception of 'Strut Head 1,' for which the licensee provided an 

acceptable explanation. The submittal further noted that the trash rack is outside the zone of 

influence for all applicable design basis high-energy line breaks (HELB) and; therefore, dynamic 

loads were not considered in the structural analysis. The licensee noted that backflushing was 

not incorporated into the strainer design. 

The NRC staff review of the licensee's submittal identified several issues. Specifically, the 

licensee's submittal failed to identify which edition of the American Society of Mechanical 

Engineers Boiler and Pressure Vessel Code (ASME Code) was used for the strainer structural 

analysis, and no results were provided for the structural qualification of the trash rack. There 

were also discrepancies in the written evaluation as to which loadings apply to which structures, 

and contradictory statements were made regarding the inclusion of hydrodynamic loads or 

masses. Additionally, the licensee did not properly evaluate the dynamic effects due to a highenergy line break within a 10-diameter radius of the existing plant piping. To address these 

issues, by letter dated July 24, 2008, the NRC issued five RAls. 
- 7 -

In its response, dated September 19, 2008, the licensee identified the edition of the ASME Code 

used to support their structural evaluations and provided a tabulated summary of the stress 

ratios for the trash rack structures. The licensee also clarified its comments regarding the 

structural integrity of the trash rack and identified which loads are applicable to the strainer and 

the trash rack. The licensee analyzed the jet force impacts on the trash rack using simplified 

methods described in the updated final safety analysis report and demonstrated that no load 

would be imposed on the trash racks. The licensee also noted that the interaction between the 

strainer and the water inside the sump pit was not relevant due to the small volume of the sump. 

Therefore, hydrodynamic loads did not need to be considered. 

The NRC staff reviewed the RAI responses and found them acceptable, except for the response 

regarding the hydrodynamic loads. Several components (notably Strut Head 1 ), had such high 

stress ratios that even a small hydrodynamic loading value could have an impact on the 

acceptability of particular strainer components. To address this, the NRC issued a follow-up 

RAI by letter dated April 15, 2009, to which the licensee responded by letter dated May 29, 

2009. In its response, the licensee explained that the loads used in the analysis of the 

components in question were very conservatively estimated. Based on the conservative 

methods used to analyze the components, adequate margin exists to account for any increase 

in loading due to hydrodynamic forces. Based on the RAI response, the NRC staff determined 

that the licensee had provided sufficient information to justify the omission of the hydrodynamic 

loading in the licensee's structural analyses. 

NRC Staff Conclusion: 

Based on the above, the NRC staff concludes that the licensee's structural analysis of the 

Braidwood and Byron replacement sump strainers is adequate and reasonable assurance exists 

that the sump strainer assemblies will remain structurally adequate under normal and abnormal 

loading conditions such that the assemblies will be able to perform their intended design 

functions. The NRC staff considers this item closed for GL 2004-02. 

3.0 DOWNSTREAM EFFECTS - FUEL AND VESSEL 

The objective of the downstream effects review, 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 long-term core cooling. 

Initial NRC Staff Review: 

The initial NRC staff review is based on documentation provided by the licensee through 

December 31, 2007. 

In the December 31, 2007, submittal, the licensee stated that it performed an evaluation of the 

effects of ECCS sump strainer bypassed debris on post-LOCA long-term core cooling using the 

guidance in WCAP-16793-NP, Revision 0. The evaluation showed that long-term core cooling 

can be achieved and the maximum fuel clad temperature, after the initial core quench, can be 

maintained below 800 degrees Fahrenheit (°F). However, the licensee acknowledged that the 

NRC staff had not issued a final SE on WCAP-16793-NP, Revision 0, and, therefore, committed 

to submitting a final response in November 2015, following of receipt of the final NRC SE. 
- 8 -

By letter July 24, 2008, the NRC staff expressed reasonable assurance that the likelihood of 

unacceptable in-vessel debris impact for Byron and Braidwood is very low because of the low 

debris loading at all four units. However, because the GL 2004-02 response referred to and 

relied on a TR for which the NRC had not yet issued an SE, the NRC deferred issuance of a 

closure letter to Byron and Braidwood for GL 2004-02 until uncertainties regarding the issues 

with WCAP-16793-NP were reduced. Further, the NRC stated that the licensee could wait for 

the issues to be resolved through the WCAP process or could demonstrate that in-vessel 

downstream effects issues were resolved for Byron and Braidwood by demonstrating, without 

reference to WCAP-16793 or the NRC staff SE, that in-vessel downstream effects were 

addressed for Byron and Braidwood. 

Final NRC Staff Review: 

By letter dated October 30, 2015, the licensee submitted a revised GL 2004-02 in-vessel 

downstream effects resolution for Byron and Braidwood. By letter dated April 15, 2016, the 

licensee submitted a revision to the October 30, 2015, submittal. The final NRC staff review is 

based on the licensee's October 30, 2015, and April 15, 2016, submittals. 

Evaluation Criteria 

On April 8, 2013, the NRC staff issued an SE (ADAMS Accession No. ML 13084A 154) on TR 

WCAP-16793-NP, Revision 2, finding the TR an acceptable model for assessing the effects of 

sump strainer bypassed fibrous, particulate, and chemical debris on core cooling in PWRs. The 

TR guidance and acceptance bases were developed through analyses and flow testing using 

representative fuel assemblies and ECCS flow rates. In order to demonstrate adequate core 

cooling capability, the TR, the limitations and conditions section of the NRC SE of the TR, and 

the GL 2004-02 response to the content guide (ADAMS Accession No. ML073110278) require 

certain actions of the licensee. These requirements and the licensee's actions for meeting 

these requirements are described herein. 

The GL 2004-02 response content guide required the response to item (n), "Downstream 

Effects - Fuel and Vessel" to confirm that the licensee's evaluation is consistent with, or 

bounded by, the industry generic guidance contained in TR WCAP-16793-NP, as modified by 

the NRC staff's conditions and limitations stated in the NRC SE on that document. Also, the 

response shall briefly summarize the application of the WCAP evaluation methods and include 

the following information: 

a) The available driving head and ECCS flow rate used in the evaluation of the hot-leg 

break LOCA scenario, 

b) The type(s) of fuel and inlet filters installed in the plant, 

c) The results of the LOCADM calculation, including the predicted peak clad temperature, 

d) The amount of fiber (in grams per fuel assembly) that is assumed to reach the core inlet 

after a LOCA, 
- 9 -

e) The method(s) used to estimate the quantity and size distribution of the fibrous debris 

that would pass through the EGGS sump strainer and reach the core inlet after a LOGA 

and, 

f) A description of any deviations from, or exceptions to the WGAP or the NRG SE for the 

WGAP. 

By letter dated October 30, 2015, the licensee stated that all units at Byron and Braidwood meet 

the requirements of 10 GFR 50.46, "Acceptance criteria for emergency core cooling systems for 

light-water nuclear power reactors," based on approved models for analyses, strainer head loss 

testing, and its analysis of in-vessel downstream effects. As the WGAP-16793-NP, Revision 2, 

methodology represents an NRG-approved model, successful completion of the in-vessel 

downstream effects analysis in accordance with the WGAP and associated SE shows 

compliance with 10 GFR 50.46 as it relates to in-vessel downstream effects. The licensee 

updated its in-vessel evaluation by letter dated April 15, 2016. This letter confirmed that the 

Byron and Braidwood units meet the NRG-approved acceptance criteria for in-vessel debris. 

Some of the values provided in the October 2015, letter, were changed slightly. This section 

reflects the updated values. 

The licensee determined the amount of fiber that could bypass their EGGS strainer using the 

methods allowed in the NRG staff SE for WGAP-16793, Revision 2. The SE allows licensees to 

determine the quantity of strainer bypass for their strainer by using the results of strainer testing 

conducted on a strainer of the same manufacture and same perforation size as the plant 

strainer. If necessary, the results are to be prorated to the plant's strainer area, approach 

velocity, debris types, and debris quantities. The licensee compared bypass results obtained 

specifically for Byron and Braidwood, but collected using techniques not fully acceptable to the 

NRG, to results obtained for plants with similar strainers and post-accident conditions. The 

licensee corrected the results to the results previously accepted by the NRG for the other plants 

and included conservatism in the evaluation. The licensee's calculations found that the 

in-vessel debris loading for the plants is less than the limit defined in the staff SE for 

WGAP-16793, Rev. 2 and that all other limitation and conditions of the staff SE are met. 

Therefore, the debris amount is acceptable. 

The licensee performed a plant-specific evaluation of the maximum fuel clad temperature and 

deposit thickness for Byron and Braidwood using WGAP-16793-NP, Revision 2, and the 

associated NRG SE for that document. The evaluation results are: 

1. The maximum calculated cladding temperature is less than 620 °F. This is less than the 

WGAP-recommended maximum cladding temperature of 800 °F. 

2. The total deposition thickness is less than 0.018 inches (18 mils). This is less than the 

WGAP recommended total debris deposition thickness of 0.050 inch (50 mils). 

Also, in the letter dated October 30, 2015, the licensee satisfactorily demonstrated compliance 

with the 14 limitations and conditions (L&Gs) of the NRG SE for WGAP-16793-NP-A, Revision 

2. The April 15, 2016 letter confirmed that the L&Gs continue to be met. 
- 10 -

Based on the above information, the licensee has documented that all the Byron and Braidwood 

units meet the requirements specified in WCAP-16793-NP, Revision 2, and the specifications, 

limitations, and conditions listed in the associated NRC SE. 

NRC Staff Conclusion: 

The NRC staff reviewed the description of the analyses, strainer bypass testing, and compliance 

with the L&Cs of the SE, as described in the licensee's GL 2004-02 response to Item (n) and 

finds that the licensee's response addressing in-vessel downstream effects for all units at Byron 

and Braidwood satisfies the requirements stated in TR WCAP-16793-NP-A, Revision 2 and the 

NRC SE for that document. Therefore, the NRC staff concludes that the licensee has 

adequately addressed the potential effects of ECCS sump strainer bypassed debris on core 

cooling at Byron and Braidwood. The NRC staff considers this item closed for GL 2004-02. 

4.0 CONCLUSION 

The NRC staff has performed a thorough review of all of the licensee's responses and RAI 

supplements to GL 2004-02. The NRC staff conclusions associated with in-vessel downstream 

effects is documented above. Based on the above evaluation, and the NRC June 1, 2009, letter 

the NRC staff finds the licensee has provided adequate information as requested by 

GL 2004-02. 

The stated purpose of GL 2004-02 was focused on demonstrating compliance with 

1 O CFR 50.46. Specifically, the GL requested addressees to perform an evaluation of the 

ECCS and CSS recirculation and, if necessary, take additional action to ensure system function, 

in light the potential for debris to adversely affect long-term core cooling. The NRC staff finds 

the information provided by the licensee demonstrates that debris will not inhibit the ECCS or 

CSS from performing their intended functions in accordance 10 CFR 50.46 to assure adequate 

long-term core cooling following a design basis accident. 

Therefore, the NRC staff concludes that the licensee's responses to GL 2004-02 are adequate 

and considers GL 2004-02 closed for the Byron Unit Nos. 1 and 2, and Braidwood, Units 1 and 

2. 

Principle Contributor: S. Smith 

Date of issuance: May 19, 2016 
B. Hanson - 2 -

performance of its intended safety function to assure adequate long term core cooling following 

a OBA in accordance with 1 O CFR 50.46, "Acceptance criteria for ECCSs for light-water nuclear 

power reactors." Therefore, the NRC staff concludes that the licensee's responses to GL 2004-

02 are adequate and considers GL 2004-02 closed for Braidwood and Byron. 

Enclosed is the summary of the NRC staff's review. If you have any questions, please call me 

at 301-415-6606 or via e-mail at Joel.Wiebe@nrc.gov. 

Docket Nos. STN 50-456, STN 50-457, 

STN 50-454 and STN 50-455 

Enclosure: 

NRC review for Braidwood and Byron 

cc w/enclosure: Listserv 

DISTRIBUTION: 

PUBLIC 

RidsNrrPMByron 

RidsRgn3MailCenter 

Accession No.: ML 15296A358 

LPL3-2 R/F 

RidsNrrPMBraidwood 

RidsNrrDorlDpr 

Sincerely, 

IRA/ 

Joel S. Wiebe, Senior Project Manager 

Plant Licensing Branch 111-2 

Division of Operating Reactor Licensing 

Office of Nuclear Reactor Regulation 

RidsN rrDorl Lpl3-2 

RidsNrrLASRohrer 

RidsAcrs_MailCTR 

*via Memo 

OFFICE N RR/LPL3-2/PM NRR/LPL3-2/LA NRR/SSIB/BC NRR/LPL3-2/BC(A) N RR/LPL3-2/PM 

NAME JWiebe SRohrer VCusumano* GEMiller JWiebe 

DATE 05/11/16 05/11/16 04/28/16 05/13/16 05/19/16 

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