ML23193A938
| ML23193A938 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 07/18/2023 |
| From: | John Klos Plant Licensing Branch II |
| To: | Carr E Virginia Electric & Power Co (VEPCO) |
| Klos, J | |
| References | |
| EPID L-2017-LRC-0000 GL 2004-02 | |
| Download: ML23193A938 (1) | |
Text
July 18, 2023 Mr. Eric S. Carr Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
SURRY POWER STATION, 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 (EPID L-2017-LRC-0000)
Dear Mr. Stoddard:
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 PWR [Pressurized Water Reactor] Sump Performance.
By letter dated May 14, 2013 (ML13140A095), Virginia Electric and Power Company (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of GSI-191 and GL 2004-02 for Surry Power Station Units 1 and 2 (Surry).
On July 23, 2019 (ML19203A303), GSI-191 was closed. It was determined that the technical issues identified in GSI-191 were now well understood and therefore GSI-191 could be closed. Prior to and in support of closing GSI-191, the NRC staff issued a technical evaluation report on in-vessel downstream effects (ML19178A252 and ML19073A044 (not publicly available, proprietary information)). Following the closure of GSI-191, the NRC staff also issued review guidance for in-vessel downstream effects, NRC Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of GL 2004-02 Responses (ML19228A011), to support review of the GL 2004-02 responses.
The NRC staff has reviewed the licensees responses and request for additional information supplements associated with GL 2004-02. Based on the evaluations, 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 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46. Specifically, GL 2004-02 requested addressees to perform an evaluation of the emergency core cooling system and containment spray system 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
E. Carr the information provided by the licensee demonstrates that debris will not inhibit the emergency core cooling system or containment spray system performance following a postulated loss-of coolant accident. Therefore, the ability of the systems to perform their safety functions, to assure adequate long term core cooling following a design-basis accident, as required by 10 CFR 50.46, has been demonstrated.
Therefore, the NRC staff finds the licensees responses to GL 2004-02 are adequate and considers GL 2004-02 closed for Surry.
Enclosed is the summary of the NRC staffs review. If you have any questions, please contact me at (301) 415-5136 or by email at John.Klos@nrc.gov.
Sincerely,
/RA/
John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281
Enclosure:
NRC Staff Review of GL 2004-02 for Surry Units 1 and 2 cc: Listserv
U.S. NUCLEAR REGULATORY COMMISSION STAFF REVIEW OF THE DOCUMENTATION PROVIDED BY VIRGINIA ELECTRIC AND POWER COMPANY FOR SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 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 plants 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 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 and 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 pass through (termed 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 quantity 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
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 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.
The GSI-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 GSI-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 GSI-191 in 2002 and documented the results in a parametric study that concluded that sump clogging at PWRs was a credible concern.
The GSI-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 GSI-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 GSI-191, the NRC issued Bulletin 03-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors (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) (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 (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 GSI-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 (ML040840034).
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 GL 2004-02 to the NRC. 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 ECCS or CSS and on the potential for additional adverse effects due to debris blockage of flow paths necessary for ECCS and CSS recirculation and containment drainage. GL 2004-02 required addressees to provide the NRC a written response in accordance with 10 CFR 50.54(f).
By letter dated May 28, 2004 (ML041550661), the Nuclear Energy Institute (NEI) submitted a report (NEI 04-07) describing a methodology for use by PWR licensees in the evaluation of containment sump performance. This is also called the Guidance Report (GR). NEI requested that the NRC review the methodology. The methodology was intended to allow licensees to address and resolve GSI-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 GSI-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 (ML043280641), the NRC issued an evaluation of the NEI methodology. The NRC staff concluded that the methodology, as approved in accordance with the NRC staff safety evaluation (SE), provides an acceptable overall guidance methodology for the plant-specific evaluation of the ECCS or CSS sump performance following postulated DBAs.
Taken together NEI 04-07 and the associated NRC staff SE are often referred to as the GR/SE.
In response to the NRC staff SE conclusions on NEI 04-07 Pressurized Water Reactor Sump Performance Evaluation Methodology (ML050550138 and ML050550156), the Pressurized Water Reactor Owners Group 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 GSI-191, Revision 1 (not publicly available), to address the effects of debris on piping systems and components (NRC Final SE at ML073520295).
TR-WCAP-16530-NP-A, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191, issued March 2008 (ML081150379), to provide a consistent approach for plants to evaluate the chemical effects that may occur post-accident in containment sump fluids (NRC Final SE at ML073521072).
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 (ML13239A114), to address the effects of debris on the reactor core (NRC Final SE at ML13084A154).
The NRC staff reviewed the TRs and found them acceptable to use (as qualified by the limitations and conditions stated in the respective SEs). A more detailed evaluation of how the TRs were used by the licensee is contained in the evaluations below.
After the NRC staff evaluated 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 (RAIs) concerning GL 2004-02. By letter dated August 15, 2007 (ML071060091), the NRC issued the content guide describing the necessary information to be submitted to allow the NRC staff to verify that each licensees 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 DBA. By letter dated November 21, 2007 (ML073110389), the NRC issued a Revised Content Guide (hereafter referred to as the 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 GSI-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 (ML121320270). 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 GSI-191. The Commission issued a Staff Requirement Memorandum on December 14, 2012 - SRM-SECY-12-0093 Closure Options for Generic Safety Issue Pressurized-Water Reactor Sump Performance, (ML12349A378), approving all three options for closure of GSI-191.
By letter dated May 14, 2013 (ML13140A095), Virginia Electric and Power Company (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of GSI-191 and GL 2004-02 for Surry Power Station Units 1 and 2 (Surry).
On July 23, 2019 (ML19203A303), GSI-191 was closed. It was determined that the technical issues identified in GSI-191 were now well understood and therefore GSI-191 could be closed.
Prior to and in support of closing the GSI, the NRR staff issued a technical evaluation report on in-vessel downstream effects (IVDEs) (ML19178A252 and ML19073A044 (non-public version)).
Following the closure of the GSI, the NRR staff also issued review guidance for IVDEs to support review of the GL 2004-02 responses, NRC Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (ML19228A011).
The following is a list of documentation provided by the licensee in response to GL 2004-02:
Table 1 - GL 2004-02 CORRESPONDENCE DOCUMENT DATE ACCESSION NUMBER DOCUMENT March 4, 2005 ML050630559 Millstone, Units 2 and 3, North Anna, Units 1 and 2 and Surry, Units 1 and 2, NRC Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors 90 Day Response.
September 1, 2005 ML052500378 Kewaunee, Millstone, North Anna, and Surry Stations - Response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors."
DOCUMENT DATE ACCESSION NUMBER DOCUMENT February 9, 2006 ML060380017 Surry, Units 1 and 2, RAI, Response to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized-Water Reactors."
February 29, 2008 ML080650561 Millstone Power Station Units 2 and 3 - Submittal of Supplemental Information of Corrective Actions, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors.
February 29, 2008 ML080650562 Surry, Units 1 and 2, Supplemental Response to NRC GL-2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.
August 4, 2008 ML082170207 (Package) North Anna Power Station, Unit 2, ML072740400 (Content) Audit of Corrective Actions for North Anna Power Generic Letter 2004-02 (TAC NO.
Station Corrective MC4697) [referred to as the North Actions for Generic Anna Audit]
Letter 2004-02 February 27, 2009 ML090641018 Surry, Units 1 & 2 - Updated Supplemental Response to NRC Generic Letter 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.
June 18, 2009 ML091540954 Surry, Units 1 & 2-Request for Additional Information Regarding Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors:
Extension Approval for Surry 1 & 2.
DOCUMENT DATE ACCESSION NUMBER DOCUMENT December 17, 2009 ML093521426 Surry, Units 1 & 2, Response to Request for Additional Information NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.
April 13, 2010 ML101040082 Surry, Units 1 & 2 - NRC Generic Letter 2004-02 Response to Request for Additional Information, Results of Finite Element Analysis of Pipe Insulation Jacketing and Band Spacing.
May 14, 2013 ML13140A095 Surry Power Station, Units 1 and 2, NRC Generic Letter 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Generic Safety Issue (GSI)-191 Closure Option.
August 13, 2015 ML15232A026 Millstone, North Anna, and Surry -
NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, Commitment Changes.
February 25, 2021 ML21056A541 Surry Power Station Units 1 And 2
- NRC Generic Letter 2004-02, "Potential Impact Of Debris Blockage On Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors" Final Supplemental
Response
September 9, 2022 ML22251A129 Millstone Power Station, Units 2 and 3, North Anna Power Station, Units 1 and 2, and Surry Power Station, Units 1 and 2 - Request for Additional Information Related to Response to Generic Letter 2004-04 (EPID L-2017-LRC-0000)
DOCUMENT DATE ACCESSION NUMBER DOCUMENT November 7, 2022 ML22312A442 Millstone, Units 1 and 2, North Anna, Units 1 and 2, Surry, Units 1 and 2, NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" Fleet Response to RAI May 8, 2023 ML23128A162 Millstone, Surry and North Anna Stations - NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" Fleet Response to Request for Additional Information The NRC staff reviewed the information provided by the licensee in response to GL 2004-02 and all RAIs. The following is a summary of the NRC staff review.
2.0 GENERAL DESCRIPTION OF CORRECTIVE ACTIONS FOR THE RESOLUTION OF GL-2004-02 GL 2004-02 Requested Information Item 2(b) requested a general description of and implementation schedule for all corrective actions. The following is a list of corrective actions completed by the licensee at Surry in support of the resolution of GL 2004-02:
Performed evaluation using the guidance of NEI 04-07.
Performed downstream effects evaluation using the TR-WCAP-16406-P-A, Revision 1 methodology.
Performed containment walkdowns using the guidance of NEI 02-01, Condition Assessment Guidelines: Debris Sources Inside PWR Containments, April 19, 2002 (ML021490212).
Enhanced the modification and maintenance processes relative to GL 2004-02 controls to insure operability of the containment sumps.
Installed new ECCS recirculation spray strainers 5,750 square feet (ft²) for Unit 1 and 5,800 ft2 for Unit 2 and low-head safety injection (LHSI) strainers (~2,200 ft² for each unit).
Installed engineered safeguards features circuitry to start sequentially recirculation spray (RS) pumps (inside containment recirculation spray (IRS) pumps first, then outside containment recirculation spray (ORS)) to ensure adequate strainer submergence and NPSH margins.
Removed insulation from ZOI and any area in containment that could result in debris generation.
Installed drain in the primary shield wall of the incore sump room to reduce holdup volume and increase total volume of water available for strainer submergence and recirculation.
Confirmed ECCS sump strainer performance by performing a prototype chemical precipitates head loss test.
Modified the containment sump level transmitters to protect them from clogging due to debris.
- Modified level transmitters within the sump by drilling holes through stilling wells at various places to prevent the element from clogging.
- Provided level transmitters located above the containment floor with debris shields to protect them.
Installed air ejectors on LHSI pump cans.
Based on the information provided by the licensee, the NRC staff considers this item closed for GL 2004-02.
3.0 BREAK SELECTION The objective of the break selection process is to identify the break sizes and locations that present the greatest challenge to post accident sump performance. The term ZOI used in this section refers to the zone representing the volume of space affected by the ruptured piping.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through December 17, 2009.
The licensee generally provided the content guide specified information and made reference to the August 4, 2008, North Anna audit as being largely applicable. The submittal described using the GR/SE discrete approach for selecting potential break locations for evaluation. The locations selected were chosen to maximize the debris source term and transport potential. Breaks in any of the large diameter RCS loop piping would generate the largest amounts and types of debris.
The licensee postulated three breaks as being potentially limiting.
The licensee did not postulate a potential break in an RCS loop at the reactor vessel nozzle.
The referenced audit report for North Anna states that because the reactor vessel (in those units) is insulated entirely with reflective metal insulation (RMI) insulation, a reactor vessel nozzle break would not generate a worst-case debris load and was determined not to warrant further evaluation.
The NRC staff requested that the licensee provide additional information regarding the amount of aluminum that could be generated as debris if a reactor vessel nozzle break were to occur.
The licensee stated that the reactor vessel insulation is stainless-steel RMI and does not
contain any aluminum. The licensee stated that although the Surry Updated Final Safety Analysis Report (UFSAR) states that the RMI contains aluminum, the RMI is actually manufactured completely of stainless-steel. The licensee reviewed design specifications and design drawings to verify the materials of construction of the RMI. The licensee also stated that the UFSAR will be revised to reflect the correct materials of construction for the RMI. The staff finds the licensee response to its request acceptable because the licensee provided adequate references to show that the RMI is constructed of stainless-steel and does not contain aluminum. Therefore, there is no concern with aluminum debris generation from the RMI.
NRC STAFF CONCLUSION:
For this review area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the break selection evaluation for Surry is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02.
4.0 DEBRIS GENERATION/ZONE OF INFLUENCE (EXCLUDING COATINGS)
The objective of the debris generation/ZOI evaluation is to determine the limiting amounts and combinations of debris that can occur from the postulated breaks in the RCS.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through April 13, 2010.
The licensee generally provided the content guide specified information and referenced the August 4, 2008, North Anna audit as being largely applicable to Surry. The analysis for Unit 1 was applied to Unit 2 due to similar design.
The licensee assumed the approved methodology (GR/SE) default ZOI values of 2.0D for their Transco RMI and a ZOI of 17.0D for TempMat with fiberglass cloth covering for its generic fiberglass, Thermal Insulating Wool (TIW) and Transco Thermal Wrap were based on a similarity to Nukon insulation. The licensee used a ZOI of 11.7D for TempMat with stainless-steel mesh covering and TempMat with silicone impregnated cloth covering. The licensee used a 5.4D ZOI for Paroc (mineral wool) similar to the treatment of the same material at North Anna.
A maximum GR/SE default ZOI of 28.6D was used for missile barrier silicone foam penetration seals. A 5.45D ZOI was used for asbestos and/or asbestos/ calcium-silicate. The licensee did not describe the construction of the asbestos or asbestos/calcium-silicate or calcium silicate (cal-sil) insulation systems.
The NRC staff requested that the licensee provide construction details for cal-sil and asbestos insulation systems and provide an evaluation to demonstrate that the systems are as structurally robust as the systems tested for ZOI determination approved in the GR/SE. The licensee provided details of construction for the insulation systems. The licensee stated that the cal-sil and asbestos were originally installed with an epoxy based jacketing system. Since that system had not been shown to be DBA qualified, the licensee installed stainless-steel jacketing over the epoxy jacketing. The licensee stated that the ZOIs for the plant systems were determined by reviewing industry testing performed on similar insulation systems. The industry testing was conducted with bands placed closer together than the plant spacing used at Surry.
The licensee therefore committed to perform a finite element analysis (FEA) to validate that the installed system would not be overstressed by LOCA jet impingement. The licensee provided the FEA results to the staff for review. The FEA concluded that the installation of jacketed cal-sil and asbestos at Surry was bounded by the testing conducted to determine the ZOI for the GR/SE.
The NRC staff reviewed the licensees FEA and determined that it did not adequately model the insulation system at Surry because of a lack of conservatism in material properties and some of the modeling methods.
While not included in the FEA model developed by the licensee, the NRC staff noted that an epoxy based jacketing system is included in the actual, physical piping insulation in Surry. The underlying epoxy jacketing is likely to provide substantial protection for the insulation material.
However, the protection afforded by the epoxy jacket has not been defined. The NRC staff believes that it is likely that the combination of the epoxy and stainless-steel jacketing systems would likely afford protection equal to or better than the tested insulation systems, but had no basis, other than engineering judgment, for this position.
An Integrated Review Team (IRT) comprised of senior NRC technical staff reviewed the issue regarding comparison of the Surry insulation system to the tested system to determine whether the licensees evaluation included adequate conservatism to offset the unknowns regarding the issue. The IRT determined that it is reasonable to expect that the installed system ZOI is comparable to the tested system ZOI. The IRT considered the technical information from the licensee and from the staffs review. The IRT also considered the impact that could occur if the installed insulation system is not as robust as the tested system. For this assessment, the IRT noted that there are considerable conservatisms in other review areas that were included in the head loss testing to offset additional debris generation from a somewhat larger ZOI. In particular, the IRT noted that in developing the debris loads for strainer head loss testing the licensee used debris transport assumptions resulting in conservatively high transport fractions for all debris types. The licensee also conservatively assumed that asbestos, cal-sil, mineral wool, and coatings all fail as small fines (that lead to higher head losses). The head loss testing protocol was also established to maximize strainer head loss. Furthermore, conservative chemical effects assumptions (such as pH values) were used to maximize effects of potential precipitates. Realistically, the NRC staff would expect lower transport fractions for the debris, a distribution of debris sizes rather than all fines, more random debris introduction and phenomena timing (less severe from a head loss perspective), and a lower chemical load.
The NRC staff concluded that it is reasonable to expect that the ZOI for the installed insulation system is comparable to the tested system. The staff also concluded that even if the realistic ZOI for the installed insulation system is somewhat larger than the tested system, it is reasonable to expect that conservatisms in the transport, debris characteristics, and head loss testing elements of the licensees response are sufficient to offset the impact of a larger ZOI.
NRC STAFF CONCLUSIONS:
For the debris generation/ZOI review area, the licensee provided information such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. Specifically, the licensee committed to use staff approved guidance to address this issue or provided additional justification to demonstrate that the evaluation is adequate.
Therefore, the NRC staff concludes that the debris generation/ZOI evaluation for Surry is acceptable. The NRC staff considers this item closed for GL 2004-02.
5.0 DEBRIS CHARACTERISTICS The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to strainer head loss.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through February 27, 2009.
The licensee provided size distribution information for the different debris types (i.e., insulation, coatings, and latent debris). The licensee also provided bulk and material densities for fibrous and particulate debris, and the assumed specific surface areas for fibrous and particulate debris constituents.
The licensee used industry standard debris sizing, densities, and destruction properties. For head loss testing adequate surrogates were chosen to represent debris that was not available or safe (asbestos). The licensee assumed that insulation was rendered into debris sizes based on the staff SE on the NEI guidance. Asbestos, cal-sil, Paroc Mineral Wool, silicone foam, and coatings were all assumed to fail as small fines.
NRC STAFF CONCLUSIONS:
For the debris characteristics review area the licensee provided information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the debris characteristics evaluation for Surry is acceptable. The NRC staff considers this item closed for GL 2004-02.
6.0 LATENT DEBRIS The objective of the latent debris evaluation process is to provide a reasonable approximation of the amount and types of latent debris (e.g., miscellaneous fiber, dust, dirt) existing within the containment and its potential impact on sump screen head loss. The guidance documents used for the review include the Revised Content Guide dated November 2007, the GR/SE, and NEI 02-01.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through February 27, 2009.
The licensees latent debris methodology and approach are consistent with the NRC SE guidance. The licensee provided assurance that it has performed an appropriate evaluation of latent debris in containment.
The licensee presented the results of its latent debris analyses. A total of 46 debris samples were taken from 12 surface types during a walk down of Unit 1, and 44 samples were taken from 11 surface types during a walk down of Unit 2; these surfaces included a variety of
horizontal and vertical surfaces and equipment. Unit 1 results were applied to Unit 2 because Unit 1 had a larger amount of latent debris.
The masses of the latent debris samples were measured to an accuracy of 0.01 gram and were statistically analyzed to determine a 90 percent confidence limit of the mean values. The bounding mass densities, when extrapolated to the areas of each containment building, resulted in total latent debris masses of 121 pound-mass (lbm) for Unit 1 and 51 lbm for Unit 2. The latent debris properties and characteristics were assumed using the guidance in the GR/SE (i.e., 15 percent fiber and 85 percent particulate). A bounding value for the latent debris source term of 121 lbm was used to size the strainers for both units.
The sacrificial screen surface area due to miscellaneous debris (tags, tape, labels, placards, and glass) was 150 ft2, which is larger than the recommended 75 percent of total foreign material (164.85 ft² x .75 = 123.64 ft²).
The amount of latent debris and foreign material considered for Surry are 127.05 lbm of latent debris and 164.85 ft2 of foreign material, including glass.
Based on the above the NRC staff finds that the licensee used conservative values and approved methodologies in the evaluation of latent debris at Surry.
NRC STAFF CONCLUSION:
For the latent debris review area, the licensee has provided information such that there is reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. The NRC staff considers the latent debris area to be adequately evaluated by the licensee. Therefore, the NRC staff considers this item closed for GL 2004-02 for Surry.
7.0 DEBRIS TRANSPORT The objective of the debris transport evaluation process is to estimate the fraction of debris that would be transported from debris sources within containment to the sump suction strainers.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through December 17, 2009.
The NRC staff reviewed the licensees supplemental response in the transport area and concluded that it was generally consistent with the approved guidance and that it further incorporated significant conservatism.
The licensee stated that the transport calculation followed guidance from the GR/SE. The licensee stated that four major debris transport modes were considered: blowdown, washdown, pool fill-up, and recirculation.
For the recirculation phase, a computational fluid dynamic (CFD) calculation was performed using Fluent. The licensees analysis was conservative, using the maximum continuous velocity between the break location and the strainer to determine debris transport. CFD results were only used to determine transport for large pieces of RMI and non-fine fibrous debris. Other materials were assumed to be small fines with 100 percent transport. Lift over curb velocity
metrics were used to eliminate transport for some large pieces of debris because the strainer is raised 6 inches (in. or ) off the ground. The licensee considered that silicone foam will float and did not model transport of this type of debris. The basis for non-transport of silicone foam was from testing done in NUREG/CR-6772, GSI-191: Separate-Effects Characterization of Debris Transport in Water, dated August 2002 (ML022410104). The licensees transport fraction of 0.5 for broken light bulb glass was based on engineering judgment that the staff considers conservative. Insulation jacketing was considered non-transportable and unable to lift onto the raised strainers. No transport to inactive containment pool volumes was modeled.
The NRC staff requested additional information regarding how transported debris was assumed to be apportioned between the RS and LHSI strainers, and that the licensee provide the basis for considering dual-train operation of the LHSI system to be bounded by single-train operation.
The licensee stated that the RS strainer could be exposed to 100 percent of the plant debris, and thus, the strainer is qualified for 100 percent of the plant debris loading. The licensee stated that the LHSI strainer will receive debris after the LHSI pumps switch over to recirculation mode.
During recirculation, the licensee stated that the debris split between the RS and LHSI strainers is determined by the flow split between strainers. The licensee stated that the 40 percent debris loading on the LHSI strainer would bound the limiting LHSI NPSH case of a single LHSI pump operating at maximum flow and two RS pumps operating at minimum flow. In this case, the LHSI strainer would draw 37 percent of the flow during cold leg recirculation and 39 percent during hot leg recirculation. If two LHSI pumps were operating, the flow fraction to the LHSI strainer would be 42 percent. Although not strictly bounded, the licensee addressed this case by pointing out that the tested strainer area was actually modeling a lower effective strainer area than installed in the plant. As such, the quantity of particulate debris used in the testing was considered to be greater than the 42 percent scaled loading. The licensee stated that the quantity of fiber in the test was not a contributor for this issue, since it was a thin bed test, and sufficient fiber had been added to achieve a thin bed. Although the licensee acknowledged that a higher flow rate would lead to greater debris bed compression, the licensee stated that this effect would be more than offset by the required NPSH (NPSHR) decrease that would be associated from lower per pump flow rates associated with two-train LHSI operation. The licensee stated that the plant emergency operating procedures (EOPs) contain direction to maintain at least two RS pumps operating during long-term recovery following a LOCA. This step is important to ensure that the debris accumulation on the LHSI strainer does not exceed the assumed value.
The NRC staff found the licensees response acceptable because the RS strainers are conservatively designed to accept 100 percent of the plant debris loading. Also, the staff considered the licensees treatment of them to be acceptable based primarily on the following points: (1) the licensee analyzed a flow split, that although not the most limiting, was only slightly lower than the limiting flow split, (2) based on conservative margins in the licensees testing, the debris quantities tested were representative of the worst case debris split, (3) the only case where a more limiting flow split (two LHSI pumps operating) could occur would have significantly greater NPSH margins ( 8 foot difference due to lower per pump flow rate) that would exceed the expected increase in debris bed head loss due to the higher total sump flow (licensee estimated about 0.25 feet), (4) the licensee conservatively neglected any debris transport to the RS strainer prior to LHSI strainer operation in determining the debris splits, and (5) the licensee has established EOPs to ensure that post-LOCA pump operation will be consistent with accident analysis assumptions.
NRC STAFF CONCLUSION:
For the transport area, the licensee has provided adequate information so that the NRC staff has reasonable assurance that the area was addressed conservatively. Therefore, the NRC staff considers this area closed for GL 2004-02.
8.0 HEAD LOSS AND VORTEXING The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainer and to evaluate the susceptibility of the strainer to vortex formation.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through December 17, 2009.
The Surry supplemental response cited the August 4, 2008, North Anna audit and Millstone 2 (ML072290550) since the Surry testing was conducted by the same vendor in a similar manner by Atomic Energy of Canada, Limited (AECL) (ML062020596). The staff found both audits by AECL approach to head loss and vortex testing and evaluation to be generally favorable.
Chemical effects were not addressed in the first Surry supplemental response but were included in the updated supplemental response. The staff later followed up the original audits with an additional chemical effects audit at North Anna. The chemical effects audit was also generally favorable.
The licensees approach to reducing strainer head loss was to install two arrays of AECL strainer modules in place of the original sump strainer. One strainer supplies the RS pumps, and one supplies the LHSI pumps. Scaled testing of the strainer modules was conducted at AECL.
The new strainer modules increase the area of the strainer significantly. The RS strainer area is about 6,220 ft2 for Unit 1 and 6,260 ft2 for Unit 2, and the LHSI strainer is about 2,180 ft2 on Unit 1 and 2,230 ft2 on Unit 2. The bottom of the RS strainer is about 6 inches off the floor. The LHSI strainer is installed on top of the RS strainer. This configuration was developed because the RS strainer is placed in service prior to the LHSI strainer, and the pumps have different suction paths. Spray from the refueling water storage tank continues to add volume to the containment sump after the RS strainer begins recirculation. This allows the LHSI strainer to be fully submerged prior to placing it in service. The design head losses for the various strainer and pump combinations are provided in a table below and compared to allowable head losses.
The AECL strainer design incorporates internal orifices to force uniform flow through all strainer perforated surfaces.
Integrated chemical effects testing results were used to determine the debris head losses for the various cases in the table below. The internal strainer losses were calculated separately and added to the debris head loss for each case to determine the overall head loss. Based on the NPSH section of the submittal there is additional margin beyond that indicated by this table.
The Surry testing was conducted by AECL at their labs in Chalk River, Ontario, Canada. The Surry testing was conducted using various test facilities. AECL has performed large-scale testing, reduced scale testing (Rig 33), and chemical testing (Rig 89). In general, the staff has found the testing procedures used by AECL for non-chemical debris to result in prototypical or conservative head loss predictions for the strainer. Testing of AECL strainers have found that, in general, thin beds (vs. full loads) present the most challenging head losses.
One item from the North Anna audit that also applies to Surry, but is accepted based on the North Anna response, is described here. The licensee assumed that at the beginning of LHSI operation in the recirculation mode there would be no debris accumulation on the LHSI strainer, and that the strainer head loss due to debris would reach the peak thin bed head loss after a period of time. The licensee stated that the maximum expected rate of head loss increase starting from a clean strainer value at the time of LHSI pump start, including significant conservatism, would not result in a negative NPSH margin throughout the required operational period for the LHSI pump. This argument is acceptable to the NRC staff and applicable to Surry.
By letter dated June 18, 2009, the staff requested that the licensee address multiple areas of additional information regarding head loss and vortexing. The licensee addressed the staffs concerns by letter dated December 17, 2009. A summary of the RAIs from the staff and the subsequent responses from the licensee are listed below.
The NRC staff requested that the licensee justify the application of a viscosity-based temperature correction to the head loss test results. The licensee stated that this question was based on a review of Rig-33 testing while Rig-89 testing is considered to provide the design-basis head loss for the site. The licensee provided additional time dependent head loss curves for the Rig-33 and Rig-89 testing. The response stated that the reductions in head loss occurred slowly and were not indicative of the formation of bore holes or other bed disruptions. The licensee also stated that the bed was inspected following the testing and that no signs of cracking, bore holes, or other degradation were observed. The NRC staff reviewed the
licensees response and agreed that there were no indications of abrupt pressure changes in the head loss traces and the bed had no indication of bore holes or other degradation.
The NRC staff requested that the licensee provide a plant-specific evaluation of the results of the Rig-33 and Rig-89 tests, in conjunction with plant-specific conservatisms that show that the strainers will function under design conditions. The staff was concerned with application of only the Rig-89 test results to the Surry strainers because the non-chemical head losses for the Rig-89 testing were significantly lower than the Rig-33 head losses which were conducted with only non-chemical debris. The licensee stated that an evaluation similar to that conducted for North Anna, and accepted by the staff, is also applicable to the Surry condition, but that a modification to the Surry LHSI pumps would be required to ensure adequate margin is maintained when considering the results of both the Rig-89 and Rig-33 tests. The licensee listed several conservatisms that were used during the testing and evaluation of the Surry strainers.
The licensee stated that the short term, non-chemical head loss requirements are bounded by both the Rig-33 and Rig-89 tests. The response stated that there is significant head loss margin even after the head losses attained during head loss testing are factored into the margin calculation. The NPSH margin for the LHSI pumps is limited by the ability to maintain the pump casing full of water. The licensee calculated the maximum allowable strainer head loss to ensure the LHSI pump casings remain full to be 2.2 feet. North Anna has similar pumps installed, but do not have the same limitation because the pumps have air ejectors to remove any air from the pump casings. The licensee stated that they will install air ejectors on the Surry LHSI pumps to gain margin for strainer head loss. The licensee calculated that once the air ejectors are installed the maximum allowable strainer head loss will increase from 2.2 ft to over 10.7 feet. The licensee stated that the modification of the pumps, along with other conservatisms in the head loss test and evaluation, will provide adequate long-term NPSH margin to offset uncertainties between the Rig-33 and Rig -89 testing. The licensee also stated that the RS pumps have significant margin above the maximum design allowable strainer head loss of 5 feet. The ORS pumps have an additional 10.29 ft of allowable head loss, and the IRS pumps have an additional 12.5 ft of allowable head loss above the allowable strainer design head loss. The NRC staff found the response acceptable because the evaluation of head loss margin, including consideration of the modification to the LHSI pumps, shows that there will be adequate NPSH margin throughout the post-LOCA mission time for the RS and LHSI pumps.
The NRC staff requested that the licensee provide additional information on strainer submergence levels during small break LOCAs (SBLOCAs) and large break LOCAs (LBLOCAs) because the licensee assumed that the submergence was the same for both SBLOCAs and LBLOCAs. The licensee provided significant additional information regarding sump water levels under various LOCA conditions. The licensee stated that the submergence values used in the head loss evaluation were conservatively low. These values were based on LBLOCA conditions, but bounded SBLOCA conditions because during a SBLOCA there are additional water sources, or fewer hold up volumes that offset the potential for the accumulators to retain their volume during a SBLOCA. The NRC staff reviewed the information provided by the licensee regarding additional sources and holdup volumes and found the response to be acceptable. The licensee identified that under very small break conditions that the LHSI strainer could be placed into service with the RCS fully refilled with the strainer submerged by about 1.1 inches instead of the 8.2-inch LBLOCA submergence. However, the licensee stated that LHSI recirculation would not be required under these conditions because normal charging and residual heat removal would be used for managing RCS inventory and heat removal. The NRC staff found that the licensee demonstrated that the submergence values used for the strainer evaluation were conservative for most potential post-LOCA scenarios. In addition, the NRC staff noted that the licensee tested the strainer for vortex formation under conditions more limiting
than those possible following any size LOCA with no vortex formation observed. Therefore, the staff found the response acceptable.
The NRC staff requested that the licensee provide information that justified that head loss fluctuations observed during the LHSI test would not affect viscosity-based temperature corrections to the test data due to bore hole formation or similar bed degradation. The licensee stated that during the time that the fluctuation occurred, the debris bed was relatively fragile because the head loss across it was low, and it was not compact. During the time period in question, the licensee concluded that the debris bed cracked and repaired itself. However, after all of the chemicals were added, head loss increased, and the bed became stable, and no additional pressure driven bed degradation occurred. In addition, the licensee stated that no viscosity correction was applied to the result of the head loss test because the test temperature and long-term sump temperature are the same. The staff finds the response to the question to be acceptable because the licensees description of the bed formation is reasonable. The licensee stated that because no viscosity correction was applied to the test result, the issue of bed degradation has no effect on the application of the test result to the plant condition.
NRC STAFF CONCLUSION:
Based on the test results provided by the licensee, the NRC staff concluded that the head loss portion of the analysis has been completed adequately. Testing and analysis were conducted using approved guidance in the Revised Content Guide dated November 2007 or alternate methods determined acceptable by the NRC staff. The other information provided by the licensee provide adequate documentation that the strainer will perform its function during any required recirculation operation at Surry. Therefore, the NRC staff concludes that the licensees evaluation of this area is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02.
9.0 NET POSITIVE SUCTION HEAD The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a LOCA considering a spectrum of break sizes.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through December 17, 2009.
The licensee presented a summary of their revised available NPSH (NPSHA) results in response to NRC GSI-191 Audit Open Item 3.7-1 (August 4, 2008, North Anna audit) to correct an error that occurred in the Surry calculation as well. Revised calculations were performed with a GOTHIC model after correcting the containment water level vs. volume input for the model.
The result was a reduction in RS pump and LHSI pump available NPSH. The NRC staff asked multiple questions related to Surry NPSH because the licensee relied upon similarity to North Anna as a basis to demonstrate adequacy of the evaluation. The staff requested that the licensee provide a plant-specific NPSH evaluation for Surry because Surry and North Anna have some design differences which require the NPSH calculations to be independent.
By letter dated December 17, 2009, the licensee responded to the staffs requests and provided a plant-specific NPSH evaluation for Surry. The staff determined that the licensee provided an
appropriate level of detail in responding to the staffs concerns and adequately addressed all areas of the GL 2004-02 content guide.
NRC STAFF CONCLUSION:
For the NPSH area, the licensee has provided information such that the NRC staff has reasonable assurance that it has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the NPSH evaluation for Surry is acceptable. The NRC staff considers this area closed for GL 2004-02.
10.0 COATINGS EVALUATION The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristics for coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.
NRC STAFF REVIEW:
The staff reviewed the coatings in containment, condition assessment program, debris generation assumptions, debris characteristics, assumptions of coatings debris transport, and head loss testing. A 10D ZOI was used for all qualified coatings. All qualified coatings within the ZOI and all unqualified coatings in containment were assumed to fail as fine particulate. The licensee assumed 100 percent of the coating debris particulate would be transported to the sump. This assumption was carried over into the head loss testing inputs. In addition, the licensee added a 7 percent margin to the coatings debris source term. The licensee observed a continuous debris bed during testing and treated all the generated coatings debris as fine particulate for testing. Walnut shell flour was used as the surrogate material for coatings debris during testing. The licensees coating assessment program meets staff guidance.
The staff has performed a detailed review and determined that the licensees analyses and testing with respect to coatings were performed in accordance with the NRC staff review guidance. The staff finds all aspects of the coatings review area acceptable.
NRC STAFF CONCLUSION:
For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the coatings evaluation for Surry is acceptable. The NRC staff considers this item closed for GL 2004-02.
11.0 DEBRIS SOURCE TERM The objective of the debris source term section is to identify any significant design and operational measures taken to control or reduce the plant debris source term to prevent potential adverse effects on the ECCS and CSS recirculation functions.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through December 17, 2009.
The licensee provided a summary of the containment housekeeping programmatic controls in place to control or reduce debris.
The licensee stated that insulation inside containment that could contribute to spray or submergence generated debris that was found to be damaged, degraded, or covered with an unqualified coating system, was removed or jacketed with a jacketing system qualified for a DBA.
The licensee has housekeeping and foreign materials exclusion programs in place to maintain cleanliness of containment and to protect plant equipment by preventing entry of foreign material. Tags and stickers are controlled by procedure and the strainer is designed for a bounding amount of such material.
The NRC staff identified that labels within containment are metal and are attached to the components with stainless-steel banding. The metal labels will remain in place when subjected to the effects of containment spray. The licensee assumed that if these labels are attached to components located within the ZOI, they will be dislodged by the LOCA jet. However, since they are significantly denser than the sump water, they will not readily transport to the sump, and are therefore not considered a debris source. Labels and stickers were identified on cable trays, equipment, pipe, and concrete walls. These labels and stickers are assumed to become a source of debris under the effects of spray or when located within the ZOI. Rather than counting each type of label, the licensee determined a typical distribution of this material in various areas and these quantities were extrapolated to the entire containment. Plant drawings were also used to assist in the determination of the quantity of labels.
The licensee has a program in place for the control of protective coatings, and coating condition assessment inside containment. Program specifications, standards, and procedures have been reviewed and updated to address GSI-191. Inside containment, coating of structures and components is controlled by coating permits.
The NRC staff requested that the licensee provide additional information with regard to its debris source term evaluation. A summary of the RAIs and the subsequent responses from the licensee are provided below.
The staff requested that the licensee provide a description of how permanent plant changes inside containment are programmatically controlled to not change the analytical assumptions and numerical inputs of the licensees analyses supporting the conclusion that the assumed debris source term within the containment is not adversely affected. The licensee stated that it has implemented a fleet GSI-191 Program which is controlled by procedures. The fleet program designates a GSI-191 Fleet Lead and Site Program Owners and delineates staff and management responsibilities to ensure the GSI-191 design and licensing bases and technical documents established for each site are properly maintained. The staff found the licensees response acceptable because these changes will enable the identification of changes, which could affect the GSI-191 design-basis.
The staff requested that the licensee provide a description of how maintenance activities, including associated temporary changes, which could affect the licensees analytical assumptions and numerical inputs of the licensees analyses relating to its resolution of sump performance issues, are assessed, and managed in accordance with the Maintenance Rule, 10 CFR 50.65. The licensee stated that the Maintenance Rule Program provides interface arrangements with other programs, such as the plant modification and preventive maintenance
programs. The staff determined that the licensees response was acceptable because the licensee has provided reasonable assurance that the work activities that may have had an impact on the GSI-191 design-basis are covered under the plant modification process.
The staff considers that the programmatic controls put in place by the licensee ensure that the GSI-191 design-basis is being maintained effectively with respect to the plant modification process. In addition, the licensees procedural controls provide processes that allow the licensee to assess and manage maintenance activities inside containment that could impact the plant's GSI-191 design-basis in accordance with the Maintenance Rule.
NRC STAFF CONCLUSION:
For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. The licensee has provided information necessary for the NRC staff to conclude that the debris source term is controlled to an acceptable level such that the recirculation function will not be adversely affected. Therefore, the NRC staff concludes that the debris source term evaluation for Surry is acceptable. The NRC staff considers this item closed for GL 2004-02.
12.0 SCREEN MODIFICATION PACKAGE The objective of the screen modification package section is to provide a basic description of the sump screen modification.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through February 27, 2009.
The licensee provided the information requested in the content guide.
The licensees responses provided a basic description of the major features of the new sump strainers. The strainers were designed by AECL. Each strainer assembly is separate and consists of a train of individual strainer modules and a pump suction assembly. Each strainer assembly is connected to the pumps for which it is intended to supply recirculating water. The strainers are finned modular strainer assemblies. The RS strainer header assembly is mounted on the containment floor in and around the containment sump. The LHSI strainer header assembly is mounted on top of the RS strainer assembly. The LHSI and RS modules strainers are manufactured together, as single modules, but the LHSI and RS flow streams are separated using solid divider plates internal to the modules. Each strainer module has a number of fins attached to the body of the module that allows the containment sump water to flow into the header assembly to the pump suction module, then to the respective pumps.
The RS strainer assembly has a surface area of approximately 6,220 ft2 for Unit 1, and approximately 6,260 ft2 for Unit 2. The LHSI strainer assembly has a surface area of approximately 2,180 ft2 for the Unit 1, and approximately 2,230 ft2 for Unit 2. The strainer assemblies are fully submerged at the start of recirculation. The pump suction header connects to the respective strainer header modules. There is one pump suction header for the RS pumps and a separate pump suction header for the LHSI pumps.
The strainer header modules consist of a rectangular header that has perpendicular fins on both sides of the header. The fins are perforated corrugated stainless-steel. The maximum opening size in the fins is a 0.0625-inch diameter hole. Fins are nominally 4 inches apart (center to center distance). The RS fins are approximately 6 inches off the containment floor, which permits water to flow under the strainer and prevents large debris from building up around the fins thus blocking the effective flow area. Debris collects on and between the fins and the water passes through the fins and down the headers to the RS and LHSI suction pipes. The strainer assemblies are designed to prevent particles larger than 0.0625 inches from entering the RS and LHSI systems.
A line from the CSS is connected to the suction header for each ORS pump. A bleed line from the RS system downstream of the RS heat exchangers is connected to the suction header for each IRS pump. Cold water supplied by these lines increases the NPSHA for the pumps by reducing the temperature of the water at the pump suction. The point at which the lines inject into the suction header was changed to ensure proper mixing of the containment sump water with the injected flow. The new arrangement ensures that the cold water is injected into the suction header far enough upstream from the pump casing to allow mixing prior to entering the pump casing.
NRC STAFF CONCLUSION:
The licensee has provided information necessary for the NRC staff review. Based on its review the NRC staff finds the licensee has provided sufficient information for Surry as required by GL 2004-02 and considers this item closed for GL 2004-02.
13.0 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 staffs review is based on Section 3k, Sump Structural Analysis, of the licensees February 29, 2008, submittal (ML080650562), as well as the RAI response submitted December 17, 2009. The guidance documents used for the review include the Revised Content Guide from November 2007 and Regulatory Guide 1.82, Revision 3 Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident (ML033140347).
The licensee provided a summary of their strainer structural analysis. The discussion of the methodology, and the assumptions and parameters in the structural analysis were clear. Each of the technical issues specified in the GL content guidance document were addressed by the licensee. Information concerning the design inputs, codes, loads, and load combinations used for the strainer structural analysis was presented. The methods used were standard industry practice for structural analysis and the assumptions were conservative and realistic.
The licensee stated that a FEA using the Analysis System (ANSYS) computer program was performed on models of the replacement strainer modules, including all associated welded and bolted connections, to demonstrate structural integrity. The models were subjected to bounding loading combinations consisting of deadweight, seismic loading, differential pressure, live loads, and hydrodynamic loading. Consideration was also given for the thermal effects on the
replacement strainer modules under LOCA conditions. The maximum induced stresses for each component were then evaluated against the applicable allowable stress values from the licensees design code of record.
The licensee stated that all the components which were analyzed for structural qualification purposes under the various loading combinations described above were found to be within the American Society of Mechanical Engineers (ASME) code allowable values. A summary table of the component margins of safety (i.e., M.S. = [Allowable/Actual] - 1) was provided and all the margins of safety remained above the acceptable limit of zero. The information provided by the licensee shows that the sump structural evaluation contains inherent conservatism by complying with accepted design standards (e.g., ASME Code).
The licensee stated that, based on an engineering evaluation, dynamic effects, such as pipe whip, jet impingement, or missile impacts, were not a factor in evaluating the structural integrity of the replacement modules since they were sufficiently isolated and protected from HELBs. A license amendment was issued for the Surry Power Station on October 15, 2007, regarding technical specification requirements for the replacement sump strainer surveillance requirements. In the NRC staffs SE regarding this license amendment (ML072690396), the staff concluded that the licensee had provided sufficient justification regarding the possibility of dynamic effect loadings on the replacement strainers such that these effects could be neglected due to the physical location of the strainers and the presence of missile barriers. Therefore, the staff finds it acceptable that dynamic effects were not evaluated for the replacement strainers.
The licensee stated that backflushing is not credited in the strainer design and analysis.
The licensees original supplement response lacked the appropriate level of detail regarding the design margins associated with the structural qualification of the strainer components. To address this, the NRC staff requested that the licensee provide additional detail in this area. The licensees response summarized the margin of safety for the strainer components, showing that all the safety margins remain above zero, with the lowest safety margin being 0.19 for the bolts associated with the saddle vertical plate.
NRC STAFF CONCLUSION:
For the sump structural analysis review area, the licensee provided information such that the NRC staff has reasonable assurance 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.
14.0 UPSTREAM EFFECTS The objective of the upstream effects assessment is to evaluate the flow paths upstream of the containment sump for holdup of inventory, which could reduce flow to the sump.
NRC STAFF REVIEW:
The NRC staff review is based upon documentation provided by the licensee through February 27, 2009.
The NRC staff determined that the licensee has appropriately accounted for potential holdup volumes, choke points, and other physical obstructions that could prevent water from draining to
the basement. The licensee applied the same methodology that was accepted by the staff in the August 4, 2008, North Anna audit to Surry and stated that any changes required to account for the few dissimilarities between containment configurations were included in the evaluation.
Drawings, plant procedures, and engineering calculations were reviewed to identify potential water holdup locations during the evaluation of containment water level.
The licensee provided the GOTHIC analysis that calculates the sump water level and pump NPSH for the staff to review. The analysis makes corrections for water holdup in the refueling canal (the drain is assumed to plug and pass no water), in the reactor cavity and instrument tunnel, on condensed films and heat structures, as films on platforms and structures, and in insulation. The same analysis methodology was reviewed by the staff during the August 4, 2008, North Anna audit. The methodology was found acceptable. However, some of the plant-specific holdup volumes are different due to plant geometry differences. For Surry, the holdup in the refueling canal is 1,720 cubic feet (ft3) (1,850 ft3 for North Anna) and the holdup in the reactor cavity and instrument tunnel below the incore sump room drain is 2,485 ft3 (2,830 ft3 at North Anna).
The design change process directs the evaluation of plant changes for effects on the flow of water to the containment strainers. In addition, containment closeout procedures are being processed to direct a review for potential flow chokepoints.
NRC STAFF CONCLUSIONS:
For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. The NRC staff considers this item closed for GL 2004-02.
15.0 DOWNSTREAM EFFECTS - COMPONENTS AND SYSTEMS The objective of the downstream effects, components and systems section is to evaluate the effects of debris carried downstream of the containment sump screen on the function of the ECCS and CSS in terms of potential wear of components and blockage of flow streams.
NRC STAFF REVIEW:
The NRC staff review is based on documentation provided by the licensee through February 27, 2009.
The licensee provided a detailed description of the evaluations performed to assess the effects of recirculated sump fluid on the components and systems located downstream of the ECCS sump strainer. The licensee stated that the evaluation was performed following the guidelines of WCAP-16406-P, Revision 1, and the NRC SE on that document, without exception.
The licensee addressed the following in its downstream effects analysis:
- 1. Extent of wear of the high head safety injection (HHSI) pumps (charging pumps), ORS pumps, IRS pumps, LHSI pumps, manually throttled valves, motor operated valves, orifices, flow venturis, RS nozzles, and heat exchangers,
- 2. Effects of wear on the performance of the component listed in (1) above,
- 3. Effects of debris on pressure relief valves which could potentially open during recirculation and piston check valves which open during recirculation, and
- 4. Potential for blockage of downstream components, including instrumentation, by debris.
The evaluation results show that all downstream components are acceptable per the criteria set forth in WCAP-16406-P, Rev. 1.
The licensee stated that the evaluation of pump hydraulic performance and mechanical dynamic performance was based on pump design performance characteristics supported by approximately ten years of in-service testing that demonstrated that there had been no statistically significant degradation of the performance of the Surry HHSI, LHSI, IRS, and ORS pumps over that period. The abrasive wear of the ECCS and recirculation spray system (RSS) pumps was conservatively calculated, and the worn-condition pump hydraulic performance was evaluated for its effect on system minimum performance requirements. The licensees overall system performance evaluation concluded that the ECCS and RSS pumps would meet their hydraulic performance requirements at the end of their 30-day mission time.
The impact of abrasive debris on the performance of pump mechanical shaft seals was evaluated for the LHSI, HHSI, and ORS pumps (IRS pumps do not utilize a mechanical seal).
The licensee concluded that the debris-laden recirculation fluid would not adversely impact the performance of the mechanical seals during the mission time.
The licensee performed an evaluation of the effect of the increased flow clearances resulting from the abrasive and erosive wear of pump components to determine if ECCS and RSS pumps would be capable of operating satisfactorily, without excessive vibration, over the required post-LOCA mission time. The LHSI, HHSI, IRS, and ORS pumps were found to satisfy the WCAP-16406-P, Rev. 1, dynamic performance requirements.
The licensee stated that the heat exchangers in the recirculation flow-paths were evaluated for wear effects due to debris-laden fluid flow. The evaluation concluded that the actual wall thickness of the heat exchangers tubes minus the tube wall thickness lost due to erosion during a 30-day period is greater than the minimum wall thickness required to withstand both the internal tube design pressure and the external shell design pressure. Therefore, the heat exchanger tubes were determined to have sufficient wall thickness to withstand the erosive effect of the debris-laden post-LOCA recirculated sump water for a period of 30 days. Further, tube blockage is not expected to occur because the tube internal diameter is greater than the maximum debris size and the flow velocity is greater than the debris settling velocity.
The licensee stated that the manually throttled valves, motor operated valves, flow venturis, orifices, and RS nozzles in the ECCS/RSS recirculation flow path were evaluated for the effects of wear due to debris-laden fluid flow. These components were evaluated individually and, with the exception of the plate orifices in the safety injection (SI) system, were found to meet the criteria set forth in WCAP-16406-P, Rev. 1. The wear of the plate orifices in the SI system flow path was included in the evaluation of system flow effects and found to have an insignificant effect. Relief valves in the recirculation flow path were evaluated for the ability to reseat in the event of opening considering the debris-laden fluid. None of the relief valves have the potential to lift during the recirculation phase; therefore, the potential for debris blockage in the open position does not exist. Piston check valves were evaluated for the potential to malfunction due to debris, and it was determined that failure of the piston check valves to close would have no effect on system functions required for the recirculation phase.
Instrumentation, except for the reactor vessel level instrumentation system (RVLIS), in the recirculation flow path that would be required to function after a LOCA was verified by the
licensee to be mounted such that debris will not affect the operation of the instruments.
Therefore, instrumentation will not be adversely affected by debris in the recirculation flow path.
The licensee stated that the RVLIS instrument tubing has no flow through it so debris would not be drawn into the sensing lines.
The licensee stated that the Surry sump strainers could have minor fit-up gaps up to 0.125-inch-wide for a total of 1 percent of strainer total flow area, and a limited number of 0.1875-inch wide by 1-inch long gaps. The licensee stated that the following five areas of the downstream effects analysis that could be affected by increased debris resulting from increased gap size were evaluated: (1) bypass fraction and debris size, (2) downstream component wear, (3) downstream component blockage, (4) fuels blockage, and (5) strainer hydraulics. The evaluation concluded that the presence of 0.125-inch-wide gaps for 1 percent of strainer flow area, and 0.1875-inch-wide by 1-inch-long gaps limited to four on the LHSI strainer and eight on the RS strainer would have no significant effect on the results of the downstream effects analyses for systems and components.
NRC STAFF CONCLUSION:
Because the licensee demonstrated that the ECCS equipment downstream of the ECCS sump strainers can perform their safety-related functions to mitigate the consequences of a HELB or LOCA using approved analytical methods prescribed in TR WCAP-16406-P-A, Rev 1 and the associated NRC SE (including limitations and conditions), the NRC staff concludes that the downstream effects of debris laden recirculated sump fluid on ex-vessel downstream components and systems have been adequately addressed at Surry. The NRC staff considers this item closed for GL 2004-02.
16.0 DOWNSTREAM EFFECTS - FUEL AND VESSEL 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 long-term core cooling.
NRC STAFF REVIEW The NRC staff review is based on documentation provided by the licensee through May 8, 2023.
In a February 29, 2008, supplemental response, the licensee stated that the evaluations completed to date followed the guidance in WCAP-16793-NP, Rev 0. Because later strainer penetration evaluations found that the fiber limits associated with WCAP-16793-NP could be exceeded, the licensee reperformed its in-vessel evaluation based on WCAP-17788 and the associated NRC staff guidance.
In a May 14, 2013, letter, the licensee stated that testing had been conducted to determine the amount of fiber that could penetrate the strainer. Testing was conducted for North Anna, Surry, and Millstone 2. The licensee stated that 99.91 percent of the fiber is captured by the Surry strainer on its first pass. The testing used grab samples to determine the penetration amounts.
The staff has recommended that penetration testing use full flow filtering to capture fiber that bypasses the strainer.
By letter dated August 13, 2015, the licensee informed the NRC that Surry would demonstrate compliance with the WCAP-17788-P reactor vessel debris acceptance criteria, instead of relying on WCAP-16793-NP to evaluate reactor vessel debris.
In its February 25, 2021, response the licensee stated that fiber from asbestos, Thermal Wrap, TempMat, Paroc/mineral wool, fiberglass, TIW, and latent fiber could potentially reach the sump strainers and that this amount was bounded by the tested amount of 1,230.7 lbm. The licensee added additional fiber for conservatism for the downstream in-vessel effect analysis. The licensee assumed that 526 pounds of fiber arrived at the ECCS strainer based on the transport split between the spray and ECCS strainers. This amount included some excess source term for conservatism. The licensee stated that because the strainer fiber bypass testing performed by AECL for the strainer design installed at Surry used a grab sample method, there is no data for the quantity of bypassed fiber as the debris bed is forming and therefore, cumulative fiber bypass fractions could not be determined. In addition, the mix of fibrous insulation types significantly changed, which impacts the theoretical debris bed thickness for determination of fiber fraction. The licensee stated it used fiber bypass data from other plants to apply to the AECL strainer at Surry. The licensee noted that it has two hydraulically independent strainers that serve the LHSI system and RS system pumps and since only the LHSI strainer delivers sump water to the reactor vessel, the bypass fraction was only determined for the LHSI strainer.
The licensee applied Point Beach test results to the Surry strainer. The licensee stated that because Surry has a higher strainer approach velocity than Point Beach, it was necessary to apply a correction factor to scale the Point Beach data to the higher velocity. The licensee derived the correction factor from the Vogtle plant tests that recorded bypass fractions at various velocities. The licensee determined that bypass mass, normalized by flow rate, was linearly related to approach velocity, which supported the calculation of cumulative bypass fractions for the Vogtle strainer at flow rates comparable to Surry and Point Beach. The licensee then was able to determine a cumulative bypass correction factor at a given debris bed thickness by scaling the Vogtle data at the Surry velocity to the Point Beach test velocity.
The geometry of the Point Beach disk strainer was compared to the Surry strainer and assessed to be equivalent in its hydraulic performance characteristics. The licensee stated that the design of the strainer ensures uniform debris disposition. The licensee assumed that all the sacrificial area would be available for formation of the fibrous debris bed to minimize the thickness of the calculated theoretical debris bed, which would result in a larger cumulative bypass fraction for the maximum debris load. The licensee considered the slightly larger strainer perforation size for Point Beach (0.066) to Surry (0.0625) and determined it has a conservative influence on cumulative bypass fractions when applying the Point Beach test results to Surry.
The licensee discussed conservatisms applied when determining the cumulative bypass fraction for the Surry strainer and provided a table listing the critical parameter comparison for sump strainer bypass testing. The cumulative bypass fraction at the theoretical debris bed thickness of 0.382 was calculated as 8.4 percent and was corrected with a velocity factor of 1.442, resulting in a cumulative fiber bypass fraction of 12.1 percent. The NRC staff drafted a question regarding the methodology used to calculate the fiber amount that would penetrate the strainer. The staff did not understand how the bed thickness correction factor was implemented to scale Point Beach test results to the Surry plant conditions. The licensee responded to the NRC in a November 7, 2022, RAI response (). Based on the response, the NRC staff developed an additional clarification question on the issue that requested the licensee to justify the method used. The licensee responded in letter dated May 8, 2023(). The response recalculated the amount of debris penetrating the strainer and the amounts of debris that can arrive at the core.
The response also required the licensee to revise the assumptions and methods used to evaluate the in-vessel debris load acceptability because the in-vessel debris amounts increased. The following discussion reflects the final debris values.
Surry is a Westinghouse 3-loop design that uses Westinghouse 15x15 Upgrade Fuel assemblies with Optimized ZIRLO cladding. Unit 1 is currently irradiating Framatome AGORA-5A-I fuel lead test assemblies. Therefore, the in-vessel debris load calculations were performed for both Westinghouse 15x15 Upgrade Fuel and Framatome AGORA-5A-I fuel assemblies. The licensee stated the Surry core contains 157 fuel assemblies.
Surry is planning to convert from a Westinghouse downflow barrel/baffle reactor vessel design to a Westinghouse upflow design to minimize baffle jetting issues. Because the potential design changes in-vessel flow along with two different fuel product types and result in four potential configurations and associated fiber limits, the licensee evaluated the following four scenarios against the proprietary total in-vessel (core inlet and heated core) fibrous debris limit contained in WCAP-17788-P, Volume 1, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (ML20010F181), for Westinghouse fuel and against Framatome calculations contained in FS1-0046625, Rev. 1, GSI-191 In-Vessel Debris Limits for Framatome Fuel, March 2020 (Framatome Proprietary).
- 1. Downflow and Westinghouse fuel - (WCAP-17788-P, Volume 1, Rev. 1, Table 6-3)
- 2. Downflow and Framatome fuel - (FS1-0046625, Table 7-2)
- 3. Upflow and Westinghouse fuel - (WCAP-17788-P, Volume 1, Rev. 1, Table 6-3)
- 4. Upflow and Framatome fuel - (FS1-0046625, Table 7-2)
The licensee calculated the fibrous debris amounts using the methodology from WCAP-17788-P.
The licensee calculated that the maximum amount of fiber to potentially reach the reactor vessel is 19.72 grams/fuel assembly (g/FA) for all four fuel configurations and fuel types, which is less than the proprietary in-vessel fibrous debris limit in Section 6.5 of WCAP-17788, Volume 1 for the Westinghouse fuel and reactor vessel flow configurations. However, the calculated Framatome AGORA-5A-I fuel core inlet fiber values are not bounded by the limits provided in FS1-0046625, Table 7-2, for Framatome downflow and upflow plant configurations. Based on the NRC staff questions (cited above) regarding the bed thickness correction factor the licensee recalculated the amounts of fiber reaching the vessel. The updated amount of debris was calculated to be 74.1 g/FA. This is greater than the core inlet debris limit, but less than the total core debris limit.
Per the 2019 NRC staff guidance, licensees may justify that a non-uniform debris bed will form at the core inlet allowing adequate flow to assure [long-term core cooling] LTCC, even though the average debris load per FA metric is exceeded.
The licensee stated that non-uniform debris accumulation is consistent with the discussion in the NRC staff Review Guidance and IVDEs technical evaluation report and supports the buildup of debris to the core inlet fiber thresholds. The addition of debris beyond this threshold will tend to push the additional debris towards the assemblies with lower amounts of debris. At some point, enough debris will be added to the RCS that the resistance at the core inlet could be high enough to reverse the flow in the baffle region such that debris can bypass the core inlet and reach the heated core. As described in WCAP-17788-P, provided the total amount of fiber to the RCS remains less than or equal to the value provided in WCAP-17788-P, Section 6.5, LTCC will
be assured. As shown in Tables 6 and 7 of the licensees submittal, the total in-vessel fiber load is less than the value provided in WCAP-17788-P, Section 6.5, which assures LTCC. Therefore, the licensee stated that use of Framatome AGORA-5A-I fuel in both upflow and downflow configurations is acceptable for Surry.
The licensee stated that the earliest SSO time for Surry is 30.3 minutes and chemical effects timing (tchem) is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 30.3 minutes is greater than the analyzed time of 20 minutes, and therefore conservative. The licensee stated that Test Group 16 from WCAP-17788-P was representative of the Surry plant conditions and based its tchem (precipitation time) on those test results. The NRC staff reviewed the WCAP-17788-P, Volume 5 test data and concluded that the Test Group 16 data is reasonably representative of the projected Surry post-LOCA environment.
The data supports the plants use of a 24-hour tchem since no precipitation was detected during the 24-hour test duration. The licensee, however, submitted a license amendment request to switch the post-LOCA pH buffering chemical from sodium hydroxide to sodium tetraborate. This buffer switch will reduce pH during the early phases of containment spray and therefore result in a lower long-term dissolved aluminum concentration in the post-LOCA sump pool. This will increase the time to chemical precipitation relative to the WCAP-17788-P autoclave test conditions and is therefore acceptable. The licensee stated that Hot Leg Switchover (HLSO) occurs no later than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after event initiation to mitigate the potential for boric acid precipitation, which is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee stated that tblock is 260 minutes for the downflow vessel design, (WCAP-17788, Volume 1, Table 6-1), which is the earliest time that complete fuel inlet blockage can occur while not compromising LTCC (WCAP-17788-P) nor inhibiting LTCC. Tblock would be 143 minutes should Surry convert to an upflow vessel design.
The licensee confirmed that tchem of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is greater than either tblock of 260 minutes or 143 minutes. Since tchem is longer than the analyzed tblock time for the plant and also longer than the maximum 9-hour HLSO time for the plant, alternate flowpaths (AFPs) are available for coolant to reach the core should the core inlet be blocked by chemical precipitates, combined with other debris.
The licensee stated that Surry core thermal power (2,597 megawatts thermal (MWt) at which most safety analyses are conducted or 100.387 percent of licensed core thermal power) is less than the applicable analyzed thermal powers in the WCAP (3,658 MWt for upflow plants and 2,951 for downflow plants) and is therefore bounded by the WCAP AFP analysis.
The licensee stated that the downflow AFP resistance for Surry is not bounded by the WCAP-17788-P analysis. However, the WCAP allows the AFP to be corrected by the ratio of plant thermal power to analyzed core thermal power. The licensee stated that the adjusted Surry specific downflow AFP resistance is bounded by the WCAP-17788 analyzed AFP resistance (Table 6-2 of WCAP-17788, Volume 4). Given that Surry is planning to convert to an upflow barrel/baffle reactor vessel design, the licensee determined the AFP resistance for an upflow design. The Surry specific AFP resistance for the barrel/baffle region is similar to other Westinghouse 3-loop converted upflow plants provided in Table RAI-4.2-24 of WCAP-17788, Volume 4 and is less than the analyzed value, and therefore, is bounded by the resistance applied to the 3-loop converted upflow AFP analysis. Both the upflow and downflow AFP resistances for Surry are bounded by the analyzed values.
The licensee stated that the minimum plant-specific ECCS recirculation flow rate is 18.47 gallons per minute (gpm) per FA and that the ECCS recirculation flow rate corresponding to the most limiting fiber injection hot leg break scenario is 26.11 gpm/FA. The licensee stated that these flow rates are within the range of ECCS recirculation flow rates considered in the AFP analysis.
Because the updated fiber amounts for the core increased to a value greater than the core inlet debris limit (for some of the potential Surry configurations), the licensee cited NRC staff guidance that the debris bed is realistically expected to collect non-uniformly. As a result, the licensee concluded that the amount of debris required to completely block the core inlet would be greater than that assumed in the analyses and the updated amount of debris would not result in blockage at the core inlet, and the current long-term core cooling (LTCC) analyses remain applicable.
The NRC staff reviewed the licensees information and found that it had generally followed staff guidance in the in-vessel evaluation. All the key parameters were bounded by the WCAP-17788-P analyses. The only issues identified by the staff was the potential for fiber amounts to be higher than calculated by the licensee and the lack of information regarding the ability of the new fuel design to accommodate the debris amounts specified in WCAP-17788-P.
Both Surry units are converting from downflow to upflow barrel/baffle design. They also have different fuel designs and are changing fuel designs. Therefore, the licensee evaluated four barrel/baffle and fuel combinations. Two are for the current conditions and two are the planned final configurations. Prior to the recalculation of core fiber loads, two of the configurations allowed the units to meet the fiber acceptance limits established in WCAP-17788-P for the core inlet. The configuration for upflow and downflow barrel baffle designs using Framatome AGORA-5A-1 fuel design have core inlet limits that are less than the initial fiber amounts calculated by the licensee to arrive at the inlet. The licensee justified the fiber amount for the AGORA-5A-1 fuel configurations by crediting non-uniform debris buildup at the core inlet as allowed by NRC staff guidance. After the debris amounts were recalculated, the licensee credited non-uniform debris bed deposition at the core inlet for all the configurations.
Although the downflow AFP resistance is not bounded by the WCAP-17788-P analysis, it is bounded by the power corrected AFP resistance. The correction for thermal power has been accepted by the NRC staff as an appropriate method.
During its review of the February 25, 2021, submittal, the NRC staff identified additional information required to ensure that the in-vessel evaluation was performed acceptably. In a letter dated September 9, 2022, the NRC staff requested additional information regarding the licensees method for correcting the penetration values for fiber bed thickness, clarifications on the references used, clarifications regarding fuel design, and confirmation that the RSS would start and run at the flow rate assumed for the period during which penetration is calculated.
The licensee provided responses to the NRC staff questions in its letter dated November 7, 2022. The NRC staff found that the references used by the licensee were correct. The NRC staff found that the new fuel design that did not have specific headloss test data was adequately evaluated by the licensee based on information that demonstrated that the inlet filter for the new fuel design will be similar in filtering efficiency to the two inlet filters to which it was compared.
Additionally, the NRC guidance recognizes the deposition of debris at the core inlet will be non-uniform due to flow variations across the core. The non-uniform deposition results in conservatism in the fiber values that were developed assuming a uniform deposition of debris.
The licensee also provided assurance that the spray system will remain in service during the times assumed in the analysis. The NRC staff found that these responses are consistent with staff guidance and the licensees evaluation. The issues regarding the fiber penetration methodology are discussed above.
The NRC staff recognizes that the licensee is citing NRC staff guidance that allows crediting non-uniform debris bed distribution at the core inlet combined with maintaining the total debris amount reaching the reactor less than the total core fiber limit as the primary criteria for maintaining adequate LTCC. The licensee also provided information that indicates that the AFPs will be available to provide cooling to the core should the core inlet become blocked. This is an acceptable evaluation methodology to ensure LTCC will not be compromised by debris entering the core.
NRC STAFF CONCLUSION For the in-vessel downstream effects review area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the licensees evaluation of this area is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02.
17.0 CHEMICAL EFFECTS:
The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on sump strainer head loss and core cooling. Chemical effects within the reactor vessel were evaluated in Section 16, Downstream - Fuel and Vessel.
NRC STAFF REVIEW:
The licensees chemical effects evaluation uses a methodology similar to the North Anna Chemical Effects evaluation that was audited by the NRC staff in February 2009 (ML090410626).
Bench top tests were used to assess the solubility, precipitate behavior, and the applicability of industry data on plant-specific dissolution and precipitation test conditions and materials.
Reduced scale testing evaluated the influence of chemical products in head loss across the strainer surfaces by simulating the post-LOCA plant-specific chemical environment. These tests verify that adequate NPSH is available to support the operation of the LHSI and RS pumps during the post-LOCA recirculation mode. The methodology to evaluate strainer chemical effects consisted of three elements:
- 1. An assessment of potential precipitates, including determination of reactive material amounts present in the containment sump pool, pH and temperature profiles in containment, and a review of existing test and scientific literature data.
- 2. Bench-top testing to demonstrate that the solubility behavior of potential precipitates determined from literature is reproducible under plant conditions and to confirm that precipitates can be produced, if required, for reduced scale testing.
- 3. Reduced scale testing to determine the influence of chemical test products present in the containment sump pool on the head loss across the test strainer.
From the debris generation and transportation analyses performed in the containment of both Surry Units 1 and 2, the licensee determined that the fibrous sources considered in these analyses includes asbestos, Thermal wrap, TempMat, fiberglass, PAROC/mineral wool, TIW, and latent fiber. The licensee used AECL to perform chemical effects head loss testing and its
evaluation for Surry. Chemical effects testing consisted of bench-top testing and the reduced scale test loop (Test Rig 89).
Following the initial review of the chemical effects area, the staff required additional information in order to complete its review. The licensee stated that chemical effects would be insignificant during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a LOCA. The staff requested that the licensee provide the technical justification, including all related analysis, for this assertion. The licensee provided analysis results which showed that the short -term aluminum concentration in the sump would not approach the concentration needed to precipitate for both a bounding maximum temperature profile (the greatest aluminum corrosion) and a bounding maximum temperature minus 25 degrees Fahrenheit. The licensee demonstrated significant margin to precipitation and the NRC staff found the response acceptable for demonstrating that aluminum will not precipitate in the short term. The licensee also incorporated the following conservatisms into their chemical effects evaluation:
The licensee performed bench tests and long-term (>30 days) chemical loop tests (Rig 89) at AECL to understand the plant-specific chemical source term and the head loss resulting from chemical precipitates. The NRC staff visited AECL to observe the Rig 89 testing and evaluate the test methodology. It was determined to be acceptable by the NRC staff.
The licensee estimates that the containment pool equilibrium pH will be approximately 8 following a LOCA. To provide a conservative analysis, the licensee calculated the aluminum release at a pH of 9 but performed testing at a pH of 7. The higher pH assumption provides for a conservative calculation for aluminum release since aluminum corrosion increases with pH. The Rig 89 tests at pH 7 result in a much lower aluminum solubility during the test compared to the expected plant pH, which promoted aluminum precipitation during the test.
The NRC staff concludes that the licensee has provided sufficient information to address all RAIs related to chemical effects. The licensee has submitted a license amendment request to switch from sodium hydroxide to sodium tetraborate to buffer the post-LOCA pool pH. The licensee has performed analysis that demonstrates the post-LOCA pool aluminum concentration will be reduced and the margin to precipitation is increased by switching to sodium tetraborate.
Therefore, there are no remaining concerns related to chemical effects.
NRC STAFF CONCLUSION:
For the chemical effects area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that chemical effects have been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the chemical effects evaluation for Surry is acceptable. The NRC staff considers this area closed for GL 2004-02.
18.0 LICENSING BASIS The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the changes associated with GL 2004-02.
The licensee committed to change the UFSAR in accordance with 10 CFR 50.71(e) to reflect the changes to the plant in support of the resolution to GL 2004-02. In addition, the licensee stated that changes would be made to the UFSAR describing the new licensing basis to reflect
the revised debris loading as it affects ECCS sump strainer performance and in-vessel effects, including the following:
Break Selection Debris Generation Latent Debris Debris Transport Head Loss Additional Design Considerations NRC STAFF CONCLUSION:
For this review area the licensee has provided information, such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. Based on the licensees commitment, the NRC staff has confidence that the licensee will affect the appropriate changes to the Surry UFSAR, in accordance with 10 CFR 50.71(e), that will reflect the changes to the licensing basis as a result of corrective actions made to address GL 2004-02. Therefore, the NRC considers this item closed for GL 2004-02.
19.0 CONCLUSION
The NRC staff performed a thorough review of the licensees responses and RAI supplements to GL 2004-02. The NRC staff conclusions are documented above. Based on the above evaluations 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 10 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 of the potential for debris to adversely affect LTCC. The NRC staff finds the information provided by the licensee demonstrates that debris will not inhibit the ECCS or CSS performance following a postulated LOCA. Therefore, the ability of the systems to perform their safety functions, to assure adequate LTCC following a DBA, as required by 10 CFR 50.46, has been demonstrated.
Therefore, the NRC staff finds that the licensees responses to GL 2004-02 are adequate and considers GL 2004-02 closed for the Surry Power Station, Units 1 and 2.
Principal contributors: S. Smith, NRR A. Russell, NRR P. Klein, NRR M. Yoder, NRR B. Lehman, NRR Date: July 18, 2023
ML23193A938 *by e-mail OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA* NRR/DSS/STSB/BC*
NAME JKlos KGoldstein VCusumano DATE 07/12/2023 07/18/2023 06/16/2023 OFFICE NRR/DNRL/NCSG/BC* DORL/LPL2-1/(A)BC* DORL/LPL2-1/PM NAME SBloom MMarkley (SWilliams for) JKlos DATE 07/12/2023 07/18/2023 07/18/2023