Mr. Kevin K. Davison 

Site Vice President 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

April 10, 2015 

Prairie Island Nuclear Generating Plant 

Northern States Power Company - Minnesota 

1717 Wakonade Drive East 

Welch, MN 55089 

SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 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" 

(TAC NOS. MC4707 AND MC4708) 

Dear Mr. Davison: 

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 (GSI) - 191, "Assessment of Debris 

Accumulation on Pressurized Water Reactor [PWR] 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 For Light Water Reactors." Specifically, the GL requested 

addressees to perform an evaluation of the emergency core cooling system (ECCS) and 

containment spray system (CSS) recirculation and, if necessary, take additional action to ensure 

system function. This action was requested in light of the potential susceptibility of PWR sump 

screens to blockage during a design-basis accident requiring recirculation operation of the 

ECCS or CSS, and based 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. ML050670014), Nuclear Management 

Company, LLC, a predecessor license holder to Northern States Power Company, a Minnesota 

corporation (NSPM, the licensee), provided responses to GL 2004-02 for the Prairie Island 

Nuclear Generating Plant (PINGP), Units 1 and 2. In order to complete its review, the NRC staff 

issued requests for additional information (RAls) in a letter dated June 2, 2005 (ADAMS 

Accession No. ML051520370). 

By letter dated July 11, 2005 (ADAMS Accession No. ML051930570), the licensee provided 

responses to the RAls. By letter dated August 31, 2005 (ADAMS Accession 

No. ML052440054), the licensee provided supplemental information. The NRC staff reviewed 

the licensee's RAI responses and supplemental information, and determined that further 

information was necessary to complete its review. The staff issued additional RAls by letter 

dated February 9, 2006 (ADAMS Accession No. ML060370394). 
K. Davison - 2 -

By letters dated February 28, 2008 (ADAMS Accession No. ML080590629), and March 31, 

2008 (ADAMS Accession No. ML080920532), the licensee responded to the RAls and provided 

supplemental information. 

The NRC staff reviewed the RAI responses and supplemental information. In a letter dated 

June 1, 2009 (ADAMS Accession No. ML091330014), the NRC staff concluded that it had no 

further questions regarding the licensee's completion of corrective actions for GL 2004-02, with 

the exception of the resolution of the issue of in-vessel downstream effects. The staff's 

conclusion was based, in part, on the very low debris loading at PINGP as discussed in the 

licensee's RAI responses and supplemental information. 

By letter dated March 9, 2015 (ADAMS Accession No. ML 15068A337), the licensee addressed 

in-vessel downstream effects, the remaining open issue for PINGP associated with GL 2004-02. 

The NRC staff performed a review of the licensee's response and documented its conclusions 

in the enclosed summary (ADAMS Accession No. ML 15084A023). 

The NRC staff reviewed all of NSPM's responses and RAI supplements regarding GL 2004-02. 

Based on its review, the staff finds that NSPM provided adequate information as requested by 

GL 2004-02. The staff finds the information provided demonstrates that debris will not inhibit 

ECCS or CSS performance and the intended system functions in accordance with 10 CFR 

50.46 to assure adequate long-term core cooling following a design basis accident. 

Based on the above, NRC staff finds the licensee's responses to GL 2004-02 are adequate and 

considers GL 2004-02 closed for the Prairie Island Nuclear Generating Plant, Units 1 and 2. No 

further information or action is requested of the licensee. 

If you have any questions, please call me at 301-415-3049. 

Docket Nos. 50-282 and 50-306 

Enclosure: 

Closeout Summary 

cc w/encl: Distribution via Listserv 

Sincerely, 

Terry A. Beltz, Senior Project Manager 

Plant Licensing Branch 111-1 

Division of Operating Reactor Licensing 

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

OF THE DOCUMENTATION PROVIDED BY 

NORTHERN STATES POWER COMPANY- MINNESOTA 

FOR THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 

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 

pressurized-water reactors (PWRs) once the primary water source has been depleted. 

If a 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 cooling 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 

strainers 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 

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 

Enclosure 
- 2 -

debris alone. These research findings prompted the U.S. Nuclear Regulatory Commission 

(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 rule, which is deterministic, 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 a 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 review and endorsement of the GL from the CRGR 

(ADAMS Accession No. ML040840034). 
- 3 -

The GL 2004-02 requested that addressees perform an evaluation of the ECCS 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 this letter to the NRG. This request is based on the identified 

potential susceptibility of PWR recirculation sump screens to debris blockage during 

design-basis accidents requiring recirculation operation of 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. The 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. ML041550279), 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 Package 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 staff's safety evaluation (SE), provides an acceptable 

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

performance following postulated design basis accidents. 

In response to the NRG staff SE conclusions on NEI 04-07, 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 (ADAMS Accession 

No. ML 13239A 114 ), to address the effects of debris on the reactor core. 
- 4 -

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

qualified by the limitations and conditions (L&Cs) stated in the respective SEs). 

After the NRC staff's evaluation of licensee responses to GL 2004-02, the 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 ECCS and CSS will perform their intended functions following 

any design-basis accident. 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 

• Net Positive Suction Head 

• 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 

• Licensing Basis 

Resolution of GSl-191 has been more difficult than anticipated. Based on the interactions with 

stakeholders and the results of the industry testing, the NRC staff in 2012 developed three 

options that will be effective ways 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 

Performance," dated July 9, 2012 (ADAMS Accession No. ML 121310648). The options are 

summarized as follows: 

• Option 1 would require licensees to demonstrate compliance with 1 O 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. 
- 5 -

• Option 2 requires implementation of additional mitigative 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 

TR-WCAP 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 options are 

innovative and creative, as well as risk informed and safety-focused. The Commission issued a 

Staff Requirements Memorandum on December 14, 2012 (ADAMS Accession 

No. ML 12349A378), approving all three options for closure of GSl-191. 

By letter dated May 9, 2013 (ADAMS Accession No. ML 13129A375), Northern States Power 

Company - Minnesota (the licensee) stated that they will pursue Option 1 for the closure of 

GSl-191 and GL 2004-02 for the Prairie Island Nuclear Generating Plant, Units 1 and 2 (Prairie 

Island). 

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 

March 7, 2005 ML050670014 

1st NRC RAI letter dated June 2, 2005 (Accession No. ML051520370) 

Licensee Responses to RAls 

July 11, 2005 ML051930570 

August 31, 2005 ML052440054 Supplemental Information 

2nd NRC RAI letter dated February 9, 2006 (Accession No. ML060370394) 

Licensee Responses to RAls 

February 28, 2008 ML080590629 

March 31, 2008 ML080920532 
- 6 -

NRC's No Further Question letter dated June 1, 2009 (Accession No. ML091330014) 

Licensee Response to In-Vessel Downstream Effects 

March 9, 2015 ML 15068A337 

The NRC staff reviewed the licensee's responses through March 31, 2008, and by letter dated 

June 1, 2009 (ADAMS Accession No. ML091330014). The staff concluded that it had no further 

questions regarding the licensee's completion of corrective actions for GL 2004-02 in 

establishing compliance with 10 CFR 50.46, with the exception of demonstrating that in-vessel 

downstream effects issues are resolved. The staff's in-vessel downstream effects review is 

discussed below. Additionally, the NRC staff notes that the debris loading at Prairie Island is 

less than 15 grams of fiber per fuel assembly, as discussed in the licensee's RAls responses 

and supplemental information. 

2.0 IN-VESSEL 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 

March 31, 2008. 

By letters dated February 28, 2008, and March 31, 2008, the licensee submitted supplemental 

responses to GL 2004-02. The licensee stated in these letters that it performed an evaluation of 

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

Prairie Island 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. 

Therefore, in its letter dated May 9, 2013, the licensee committed to submit its final response 

prior to the end of the second Unit 1 refueling outage occurring after January 1, 2013, 

documenting how Prairie Island meets the L&Cs in the final NRC SE. 

By letter dated June 1, 2009, the NRC staff expressed reasonable assurance that the likelihood 

of unacceptable in-vessel debris impact at Prairie Island was very low because of the low debris 

loading. However, because the GL 2004-02 response referred to and relied on a topical report 

for which the NRC had not yet issued an SE, the NRC staff deferred issuance of a closure letter 

to Prairie Island for GL 2004-02 until uncertainties regarding the issues with WCAP-16793-NP 

were reduced. Further, the staff 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 by demonstrating, without reference to WCAP-16793 or the NRC staff's 

SE, that in-vessel downstream effects were addressed for Prairie Island. 
- 7 -

FINAL NRC STAFF REVIEW 

By letter dated March 9, 2015, the licensee submitted a revised GL 2004-02 in-vessel 

downstream effects resolution for Prairie Island. The final NRC staff's review is based on the 

licensee's March 9, 2015, letter. 

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 a 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 LOCA Deposition Analysis Model (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; 

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

that would pass through the ECCS sump strainer and reach the core inlet after a LOCA; 

and, 

f) A description of any deviations from, or exceptions to the WCAP or the NRC SE for the 

WCAP. 

In its letter dated March 9, 2015, the licensee stated that Prairie Island meets the requirements 

of 10 CFR 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 WCAP-16793-NP, Revision 2, 
- 8 -

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

downstream effects analysis in accordance with the WCAP and associated SE indicates 

compliance with 10 CFR 50.46 as it relates to in-vessel downstream effects, and resolves this 

final outstanding item for Prairie Island. 

The licensee determined the amount of fiber that could bypass its ECCS strainer using the 

methods allowed in the NRC Staff SE forWCAP-16793, Revision 2. The SE allows a licensee 

to determine the quantity of strainer bypass for its strainer by using the results of strainer testing 

conducted on a strainer of the same manufacture and perforation size as that in the plant. If 

necessary, the results were to be prorated to the plant's strainer area, approach velocity, debris 

types, and debris quantities. 

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

deposit thickness for Prairie Island using WCAP-16793-NP, Revision 2, and the associated 

NRC SE for that document. The evaluation results are as follows: 

1. The maximum calculated cladding temperature is 420°F. This is less than the 

WCAP-recommended maximum cladding temperature of 800°F. 

2. The total deposition thickness is 0.0036 inches (3.6 mils). This is less than the 

WCAP-recommended total debris deposition thickness of 0.050 inches. 

Also, in its letter dated March 9, 2015, the licensee satisfactorily demonstrated compliance with 

the 14 L&Cs of the NRC's SE for WCAP-16793-NP-A, Revision 2. 

Based on the above information, the licensee documented that Prairie Island, Units 1 and 2, 

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

limitations, and conditions listed in the associated NRC SE. 

NRC STAFF FINAL 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). The 

staff finds that the licensee's response in addressing in-vessel downstream effects for the 

Prairie Island Nuclear Generating Plant, Units 1 and 2, satisfies the requirements stated in TR 

WCAP-16793-NP-A, Revision 2, and the NRC's SE for that document. As such, the staff 

concludes that the licensee adequately addressed the potential effects of ECCS sump strainer 

bypassed debris on core cooling at Prairie Island. Therefore, the NRC staff considers this item 

closed for GL 2004-02. 

3.0 CONCLUSION 

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

responses associated with GL 2004-02. The staff's conclusion associated with in-vessel 

downstream effects is documented above. Based on the above evaluation and its June 1, 

2009, letter, the NRC staff finds that the licensee provided adequate information as requested in 

GL 2004-02. 
- 9 -

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

10 CFR 50.46. Specifically, the GL requested addressees to perform an evaluation of ECCS 

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

of 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 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. 

Based on the above, the NRC staff finds the licensee's responses to GL 2004-02 to be 

adequate, and considers GL 2004-02 to be closed for the Prairie Island Nuclear Generating 

Plant, Units 1 and 2. 
K. Davison - 2 -

By letters dated February 28, 2008 (ADAMS Accession No. ML080590629), and March 31, 

2008 (ADAMS Accession No. ML080920532), the licensee responded to the RAls and provided 

supplemental information. 

The NRC staff reviewed the RAI responses and supplemental information. In a letter dated 

June 1, 2009 (ADAMS Accession No. ML091330014), the NRC staff concluded that it had no 

further questions regarding the licensee's completion of corrective actions for GL 2004-02, with 

the exception of the resolution of the issue of in-vessel downstream effects. The staffs 

conclusion was based, in part, on the very low debris loading at PINGP as discussed in the 

licensee's RAI responses and supplemental information. 

By letter dated March 9, 2015 (ADAMS Accession No. ML 15068A337), the licensee addressed 

in-vessel downstream effects, the remaining open issue for PINGP associated with GL 2004-02. 

The NRC staff performed a review of the licensee's response and documented its conclusions 

in the enclosed summary (ADAMS Accession No. ML 15084A023). 

The NRC staff reviewed all of NSPM's responses and RAI supplements regarding GL 2004-02. 

Based on its review, the staff finds that NSPM provided adequate information as requested by 

GL 2004-02. The staff finds the information provided demonstrates that debris will not inhibit 

ECCS or CSS performance and the intended system functions in accordance with 10 CFR 

50.46 to assure adequate long-term core cooling following a design basis accident. 

Based on the above, NRC staff finds the licensee's responses to GL 2004-02 are adequate and 

considers GL 2004-02 closed for the Prairie Island Nuclear Generating Plant, Units 1 and 2. No 

further information or action is requested of the licensee. 

If you have any questions, please call me at 301-415-3049. 

Docket Nos. 50-282 and 50-306 

Enclosure: 

Closeout Summary 

cc w/encl: Distribution via Listserv 

DISTRIBUTION: 

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RidsNrrLAMHenderson Resource 

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A D AM SA ccess1on N os.: p ac k aae: 

OFFICE NRR/LPL3-1/PM 

NAME TBeltz 

DATE 03/25/2015 

OFFICE NRR/SSIB/BC 

NAME VCusumano 

DATE 04/07/2015 

Sincerely, 

IRA/ 

Terry A. Beltz, Senior Project Manager 

Plant Licensing Branch 111-1 

Division of Operating Reactor Licensing 

Office of Nuclear Reactor Regulation 

RidsNrrSsib Resource 

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SSanders, NRR 

JStang, NRR 

ML 15084A026 L e tt er: ML 15062A301 s ummary: ML 15084A023 

NRR/LPL3-1/LA NRR/PGCB/BC 

MHenderson SStuchell 

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DPelton TBeltz 

04/08/2015 04/10/2015 

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