ML15246A129

Attachment 2 - Requests for Exemptions for STP Piloted Risk-Informed Approach to Closure for GS1-191 Through 3-3 Technical Specifications Bases Page Markups (Information Only)

ML15246A129
Person / Time
Site:	South Texas
Issue date:	08/20/2015
From:	South Texas
To:	Office of Nuclear Reactor Regulation
References
GL-04-002, GSI-191, NOC-AE-15003241, TAC MF2400, TAC MF2401, TAC MF2402, TAC MF2403, TAC MF2404, TAC MF2405, TAC MF2406, TAC MF2407, TAC MF2408, TAC MF2409
Download: ML15246A129 (84)

Text

NOC-AE-1 5003241 Attachment 2 Requests for Exemptions for STP Piloted Risk-Informed Approach to Closure for GS1-1 91 2-1 General 2-2 Request for Exemption from 10CFR50.46(d) 2-3 Request for Exemption from GDC-35 2-4 Request for Exemption from GDC-38 2-5 Request for Exemption from GDC-41 NOC-AE-1 5003241 Attachment 2-1 Page 1 of 11 2-1 General Introduction In support of the South Texas Project (STP) risk-informed approach to addressing Generic Safety Issue (GSI)-191 and response to GL 2004-02, Attachments 2-2 through 2-5 provide STP Nuclear Operating Company (STPNOC) requests for exemptions under 10CFR5O.12 from certain requirements in 1OCFR50.46 and 10CFR50 Appendix A General Design Criteria (GDC). The exemption requests complement a proposed license amendment request (LAR) provided in Attachment 3 to this letter, proposing methodology changes that will be incorporated in the South Texas Project (STP) Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) and Technical Specifications based on NRC acceptance of the risk-informed method and results.Specific exemption requests, pertaining to requirements that concern Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) system functions for core cooling, and containment heat removal and atmosphere cleanup following a postulated loss of cooling accident (LOCA), are provided as follows: Attachment 2-2, Request for from 10OCFR50.46(d)

Attachment 2-3, Request for Exemp)tion from GDC 35 Attachment 2-4, Request for from GDC 38 Attachment 2-5, Request for from GDC 41 Approval of the exemptions will allow use of a risk-informed method to account for the probabilities and uncertainties associated with mitigation of the effects of debris following postulated LOCAs. The method evaluates the effects on strainer blockage and core blockage resulting from debris concerns raised by GS1-191. In order to confirm acceptable sump design, the risk associated with GS1-191 is evaluated to include the failure mechanisms associated with loss of core cooling and strainer blockage.With respect to other requirements for ECCS, Attachment 2-2 addresses I0CFR50 Appendix K. Attachment 2-5 addresses 10CFR50.67 and GDC 19. Based on those evaluations, STPNOC concluded that no exemptions were needed for Appendix K, I0CFR50.67 or GDC 19.Each separate Attachment 2-2 through 2-5 identifies the applicable rule from which exemption is requested, the regulatory requirements involved, the purpose of the request, and the technical basis and justification for the exemption request, including the presence of special circumstances pursuant to 10CFR5O.12(a).

The requested exemptions are part of a risk-informed approach to resolve GS1-191 issues. The risk-informed approach is designed to be consistent with the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."

NOC-AE-1 50032441 Attachment 2-1 Page 2 of 11 The scope of the exemptions applies for all debris effects addressed in the risk-informed element of the STP RoverD methodology described in Attachment 1 that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine fiber debris that exceeds the quantity bounded by STP plant-specific testing described in Attachment

1. That scope is generally described as breaks larger than approximately 12.8" ID in locations where a sufficient amount of fiber debris can be generated and transported to the sump to exceed the amount of fine fiber debris in, the STP plant-specific testing described in Attachment

1. Forty-five weld locations have currently been identified on the pressurizer surge line and RCS main loop piping. To minimize the potential that a later analysis could cause the specific locations to change, the requested exemptions are based on the breaks' ability to generate sufficient transportable debris, as described in RoverD. The key elements of each of the exemption requests are: 1. It applies only to the effects of debris as described in Attachment 1.2. It applies only for LOCA breaks that can generate and transport fiber debris that is not bounded by STP plant-specific testing.3. It applies to any LOCA break that can generate and transport fiber debris that is not bounded by STP plant-specific testing and is not limited to the 45 specific break locations noted in this application, provided that the ACDF and ALERF associated with the break size remains in Region Ill of RG 1.174.The exemptions are requested for the scope of breaks that can generate fiber debris that exceeds the amount of fiber debris bounded by the plant-specific testing. In Attachment 1-2 and 1-3, STPNOC determined that only large breaks were in this scope and listed 45 examples.

STPNOC is requesting exemption for this scope of breaks to allow evaluation of the debris effects using a risk-informed methodology because there is no practical deterministic methodology currently available.

STPNOC uses the plant-specific core fiber loading test analysis described in Attachment 1-2 together with a thermal-hydraulic screening analysis to show there is no risk contribution from downstream effects above the risk assessed for ECCS sump strainer fiber loading. In Section 5 of Attachment 1-3, STPNOC used RELAP5 to perform the thermal-hydraulic evaluation of down-stream effects necessary to support the risk assessment and confirmed adequate core cooling for the entire spectrum of breaks.STPNOC is not requesting exemption for the thermal-hydraulic analysis because it is a calculation used in a risk-informed screening (see Section 2 of Attachment 1-3). The use of the thermal-hydraulic analysis is addressed in the LAR for the methodology change (Attachment 3).The STP risk-informed approach addresses the five key principles in RG 1.174 for risk-informed decision-making.

The resulting risk metrics, i.e. changes in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), associated with GSI-191 concerns are used to determine whether plant modifications are warranted to ensure acceptable sump performance.

The requested exemptions support this pproach.A generic methodology for the STP approach is provided in Attachment I to this letter.

NOC-AE-1 5003241 Attachment 2-1 Page 3 of 11 The approach is intended to be a pilot for other licensees that are pursuing a risk-informed approach to addressing GSI-l191.

The STP approach is the risk-informed part of an overall graded approach that is based on the amount of fiber insulation in the plant, as discussed in SECY-1 2-0093, "Closure Options for Generic Safety Issue -191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance".

STP Units I and 2 contain large amounts of fiber-debris material such as insulation and coatings in the containment buildings and are expected to have higher risk of containment sump strainer blockage and in-vessel core blockage as a result of potential debris-generating postulated loss of coolant accidents (LOCAs) than plants with relatively less fiber loading.Based on the results for STP Units I and 2 showing that the risk for the effects of debris is less than the threshold for Region Ill, "Very Small Changes," of RG 1.174, no additional physical changes to the facility or changes to the operation of the facility are proposed.Backgqround and Overview GS1-191 concerns the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized-water reactors (PWRs) and result in loss of net positive suction head (NPSH) for the ECCS and CSS pumps, impeding the flow of water from the sump. Generic Letter (GL) 2004-02 requested licensees to address GSI-191 issues focused on demonstrating compliance with the 10CFR5O.46 ECCS acceptance criteria.

GL 2004-02 requested licensees to perform new, more realistic analyses using an NRC-approved methodology and to confirm the functionality of the ECCS and CSS during design basis accidents that require containment sump recirculation.

As stated in GL 2004-02: Although not traditionally considered as a component of the I OCFR5O. 46 ECCS evaluation model, the calculation of sump performance is necessary to determine if the sump and the ECCS are predicted to provide enough flow to ensure long-term cooling.Based on the new information identified during the efforts to resolve GS1-191, the staff has determined that the previous guidance used to develop current licensing basis analyses does not adequately and completely model sump screen debris blockage and related effects. As a result, due to the deficiencies in the previous guidance, an analytical error could be introduced which results in ECCS and CSS performance that does not conform with the existing applicable regulatory requirements outlined in this generic letter. Therefore, the staff is revising the guidance for determining the susceptibility of PWR recirculation sump screens to the adverse effects of debris blockage during design basis accidents requiring recirculation operation of the ECCS or CSS. In light of this revised staff guidance, it is appropriate to request that addressees perform new, more realistic analyses and submit information to confirm the functionality of the ECCS and CSS during design basis accidents requiring recirculation operations.

NOC-AE-1 5003241 Attachment 2-1 Page 4 of 11 Also, in its section on Applicable Regulatory Requirements, GL 2004-02 identifies requirements from the Code of Federal Regulations, as excerpted below.NRC regulations in Title 10, of the Code of Federal Regulations Section 50.46,10 CFR 50.46, require that the ECCS have the capability to provide long-term cooling of the reactor core following a LOCA. That is, the ECCS must be able to remove decay heat, so that the core temperature is maintained at an acceptably low value for the extended period of time required by the long-lived radioactivity remaining in the core.Similarly, for PWRs licensed to the General Design Criteria (GDCs) in Ap~pendix A to IOCFR50, GDC 38 provides requirements for containment heat removal systems, and GDC 41 provides requirements for containment atmosphere cleanup. Many PWR licensees credit a CSS, at least in part, with performing the safety functions to satisfy these requirements, and PWRs that are not licensed to the GDCs may similarly credit a CSS to satisfy licensing basis requirements.

In addition, PWR licensees may credit a CSS with reducing the accident source term to meet the limits of 10 CFR Part 100 or 10 CFR 50.67. GDC 35 is listed in 10 CFR 50.4 6(d) and specifies additional ECCS requirements.

PWRs that are not licensed to the GDCs typicaliy have similar requirements in their licensing basis.STP Units 1 and 2 have implemented compensatory and mitigative measures in response to Bulletin 2003-01 and GL 2004-02 to address the potential for sump strainer clogging and other concerns associated with GS1-191. Larger containment sump strainers have been installed that greatly reduce the potential for loss of net positive suction head (NPSH). Defense in Depth measures such as operating procedures and instrumentation to monitor core level and temperature, and actions taken by operators if core blockage is indicated, are described in Attachment 1-4.The Commission issued Staff Requirements Memorandum (SRM)-SECY-10-01 13 directing the staff to consider alternative options for resolving GS1-191 that are innovative and creative, as well as risk-informed and safety conscious.

Subsequently, STPNOC, through interactions with the staff, developed a risk-informed approach to address GS1-191 using the methods described in RG 1.174 and in a letter dated December 14, 2011 (ML1 1354A386), notified the NRC of the intent to seek exemption from certain requirements of 10CFR50.46.

SECY-12-0093 described the staff plans to use STP as a pilot for other licensees choosing to use this approach, and the STP approach referred to as risk-informed Option 2. This approach requires an exemption request in accordance with 10CFR50.12 from certain requirements of 10CFR50.46, based in part on meeting the guidance in RG 1.174. Because the residual risk of GSI-1 91 concerns meets RG 1.174 acceptance guidelines, the approach allows fiber insulation and other contributors to GSI-191 concerns to remain in containment.

The STP risk-informed methodology is described in Attachments 1-2 and 1-3.Attachments 2-2, 2-3, 2-4 and 2-5 address the deterministic requirements in I0CFR50.46, GDC 35, 38 and 41 with proposed exemptions and justify that no exemption to I0CFR50.67 or GDC I9 is needed.

5003241 Attachment 2-1 Page 5 of 11 Special Circumstances Common to Proposed Exemptions to I0CFR50.46(d), GDC 35, GDC 38, and GDC 41 10CFR50.12(a)(2)(ii) applies: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.An objective of each of the regulations (10CFR50.46(d), GOC 35, GDC 38 and GDC 41) for which an exemption is proposed is to maintain low risk to the public health and safety through functions (ECCS and/or CSS) that are supported by the containment sump. By regulatory precedent, licensees are required to demonstrate this capability by the use of a bounding calculation or other deterministic method. The supporting analysis demonstrates that a risk-informed approach to sump performance is consistent with the Commission's Safety Goals for nuclear power plants and supports operation of those functions with a high degree of reliability.

Consequently, the special circumstances described in 10CFR50.12(a)(2)(ii) apply to each of the exemptions proposed by STPNOC.10OCFR50 .1 2(a)(2)(iii) applies: Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, ,o that are significantly in excess of those incurred by others similarly situated.In order to meet a deterministic threshold value for containment debris loading, the amount of debris generating contributors in the STP plant design would need to be significantly reduced. Estimates of radiological exposure for insulation modifications are significant and on the order of hundreds of person-Rem, depending on the scope of the modifications.

With respect to the presence of such special circumstances, dose estimates for removal of insulation from STP Units 1 and 2 are described in some detail in Reference 1 to the cover letter. These dose estimates are for additional modifications to insulation in containment that would be required to achieve full resolution of GS1-1 91 using the previous deterministic methods. The residual risk associated with GS1-191 concerns bounds the expected improvement to overall plant risk that could be achieved by implementing these modifications.

STPNOC estimated the occupational dose for STP Units t and 2 that would be expected to be expended if plant modifications were undertaken for GS1-1 91, including insulation replacement and other modifications.

The scope of the estimate included the replacement of fiberglass insulation with reflective metal insulation (RMI) for reactor coolant pump (RCP) insulation and a portion of the steam generator (SG) insulation, and the banding of existing fiberglass insulation on piping in containment.

SG insulation replacement considered whether locations were within the I17D zone of influence (ZO I).The total dose expected to be expended for STP Units 1 and 2 in support of insulation replacement for GS1-191 is estimated to be 158 to 176 remn (79 -88 rem per unit).

~NOC-AE-1 5003241 Attachment 2-1 Page 6 of 11 These values significantly exceed the industry ALARA standard of 55 rem per fuel cycle for collective radiation exposure.For the above estimates, the highest dose contributor is personnel work hours in close proximity to high dose sources such as steam generators and primary coolant piping.The estimates did not include any replacement of reactor pressure vessel (RPV)insulation, which is RMI as originally designed for STP, therefore while the estimates may be indicative of a plant with high fiber loading, they do not necessarily account for activities that may be required for similar plants assuming 100-percent replacement of fiber insulation in all areas that could be affected by a postulated LOCA. The dose estimates for STP Units 1 and 2, in addition to the actual insulation replacement, considered man-hours required to erect and remove scaffolding to support the insulation modifications and the dose associated with removal of insulation.

The estimates did not consider dose associated with disposal of removed insulation or dose associated with potential hanger modifications for small bore piping insulation change to RMI.The dose considerations discussed above demonstrate that compliance would result in substantial personnel exposure due to insulation modifications in the containment which is not commensurate with the expected safety benefit based on the results showing that the risk associated with GS1-191 concerns is less than the threshold for Region Ill in RG 1.174. Consequently, the special circumstances described in I0CFR50.12(a)(2)(iii) apply to each of the exemptions proposed by STPNOC.Environmental Consideration Pursuant to the requirements of 10CFR51.41 and 10CFR51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments," the following information is provided.

As demonstrated below, this exemption qualifies for a categorical exclusion in 10CFR51.22.

However, if the NRC determines that an environmental assessment is necessary, this information will support a finding of no significant impact. The assessment applies to all of the proposed exemptions.

Identification of the Proposed Action The proposed exemption is to allow for use of a risk-informed approach to evaluate the residual risk associated with GS1-191, i.e. those concerns that have not been fully addressed using deterministic methods, for the purpose of amending the design basis for acceptable mitigation of the effects of debris during recirculation mode following postulated LOCAs. Approval of the proposed exemption would complement approval of the methodology change to be incorporated in the UFSAR and TS, as provided in Attachment 3 to this letter, for implementation of the risk-informed method for STP Units 1 and 2.Need for the Proposed Action In the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities", the Commission stated that "the use of PRA technology in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NOC-AE-1 5003241 Attachment 2-1 Page 7 of 11 NRC's deterministic approach" and that is consistent with traditional defense-in-depth concepts.The intent of the Commission's Policy Statement intent is to use the PRA to further understand the risk associated with a proposed change for the purpose of removing unnecessary conservatism associated with regulatory requirements in order to focus attention and allocation of resources to areas of true safety significance.

To implement the Commission Policy Statement, the NRC issued RG 1.174 to provide guidance on an acceptable approach to risk-informed decision-making, based on a set of five key principles.

The proposed action is needed to allow STPNOC to use a risk-informed method to address the issues associated with GSI-191 concerns regarding the potential for insulation and other debris generated in the event of a postulated LOCA within the containment impacting acceptable recirculation operation for ECCS, and challenge the ability of ECCS to provide adequate long-term core cooling. This proposed exemption is consistent with the key principle in RG 1.1 74 which requires the proposed change to meet current regulations unless explicated related to a requested exemption.

Environmental Impacts Consideration The proposed exemption has been evaluated and determined to result in no significant radiological environmental impacts associated with the implementation of the change.This conclusion is based on the following.

The proposed exemption is to allow a risk-informed method for demonstrating the design and licensing bases for the ECCS are not significantly affected by debris effects identified in GS1-1 91. No physical modifications or changes to operating requirements are proposed for the site or facility, including any systems, structures and components relied upon to mitigate the consequences of a LOCA. The intent of the proposed change is to quantify the risk associated with GS1-191 concerns.

This quantification, provided in the form of risk metrics using the guidance in RG 1.174, demonstrates that the risk is less than the threshold for Region Ill, "Very Small Changes," in RG 1.174. Therefore, the proposed exemption supports a change that represents a very small change in Large Early Release Frequency (LERF) that corresponds to an insignificant impact on the environment.

Based on the results of the risk-informed method demonstrating that the increases in risk are very small, the proposed exemption has a negligible effect on accident probability, and adequate assurance of public health and safety is maintained.

The proposed exemption does not involve any changes to the facility or facility operations that could create a new or significantly affect a previously analyzed accident or release path, and therefore would result in no significant changes in the types or quantities of radiological effluents that may be released offsite. The proposed change does not affect the generation of any radioactive effluents, and does not affect any of the permitted effluent release paths.The proposed exemption has no impact on requirements related to the integrity of the reactor coolant system piping or any other aspect related to the initiation of a LOCA. No physical modifications or changes to operating requirements are proposed for the facility, including any systems, structures and components relied upon to mitigate the NOC-AE-1 5003241 Attachment 2-1 Page 8 of 11 consequences of a LOCA. Therefore, the proposed exemption does not affect the probability of an accident initiator.

The proposed exemption does not significantly impact a release of radiological effluents during and following a postulated LOCA. The design-basis LOCA radiological consequence analysis in the current licensing basis is a deterministic evaluation based on the assumption of a major rupture of the reactor coolant system piping and a significant amount of core damage as specified in RG 1.183. The current licensing basis analysis shows the resulting doses to the public and to control room and technical support center personnel are acceptable.

The proposed change validates and does not change the input parameter value used in the radiological analysis.

Therefore, the proposed exemption does not affect the amount of radiation exposure resulting from a postulated LOCA.The proposed exemption does not involve any changes to the site property, physical changes to the facility, or changes to the operation of the facility.

Therefore there are no irreversible and irretrievable commitments of resources which would be involved in the proposed action should it be implemented.

The risk-informed method requires a determination that the risk associated with the proposed change meets the Commission's safety goals. Therefore the proposed action would not result in a significant increase in any radiological hazard beyond those events previously analyzed in the UFSAR. There will be no change to radioactive effluents that affect radiation exposures to plant workers and members of the public. Therefore, no significant changes or different types of radiological impacts are expected as a result of the proposed action. The proposed exemption does not change the input parameter value used in the radiological analysis.

Therefore, the proposed change would not significantly increase the probability or consequences of an accident, and there will be no significant offsite impact to the public from approval of the proposed exemption.

No additional physical modifications or changes to operating requirements are proposed for the facility, including any systems, structures and components relied upon to mitigate the consequences of a LOCA. Therefore, the proposed exemption does not result in a significant increase in individual or cumulative occupational radiation exposure, and will not cause radiological exposure in excess of the dose criteria for restricted and unrestricted access specified in 10 CER Part 20.The proposed exemption does not involve any changes to non-radiological plant effluents or any activities that would adversely affect the environment.

The proposed exemption does not affect any procedures used to operate the facility, or any physical characteristics of the facility, system, structure and components.

The proposed change only pertains to the licensing basis for components located within the restricted area of the facility, to which access is limited to authorized personnel.

Therefore the proposed exemption would not create any significant non-radiological impacts on the environment in the vicinity of the plant.Since implementation of the exemption request, if approved, would result in no physical changes to the facility, there is no possibility of irreversible or irretrievable commitments of resources.

Similarly, the proposed exemption does not involve the use of any resources not previously considered by the NRC in its past environmental statements

:.NOC-AE-1 5003241 Attachment 2-1 Page 9 of 11 for issuance of the facility operating licenses or other licensing actions for the facility.As a result, the proposed exemption does not involve any unresolved conflicts concerning alternative uses of available resources.

Alternatives The alternative to this exemption is compliance with the existing provisions in 10CFR50.46(d) and the relevant GDC. Compliance with 10CFR50.46(d) and the relevant GDC would entail removal and disposal of significant amounts of insulation and installation of new insulation less likely to impact sump performance in the event of a LOCA. As discussed below, the alternative would not be environmentally preferable or cost justified.

The exemption entails a very small risk and, correspondingly, an environmental impact, which is so small that it is remote and speculative for environmental purposes.Removal and reinstallation of insulation would entail appreciable radiation exposures to workers (estimated from 158 to 176 rem). This option results in extensive modifications to the facility and significant occupational dose. As such, the alternative is not environmentally preferable.

Additionally, the cost of the installation replacement would be approximately

$55 million. This cost is not justified in light of, the very small risk associated with the risk-informed exemption., Categorical Exclusion Consideration STPNOC has evaluated the proposed exemption against the criteria for identification of licensing and regulatory actions requiring environmental assessments in accordance with I0CFR5I1.21 and determined that the proposed exemption meets the criteria and is eligible for categorical exclusion as set forth in 10OCFR5I1.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9).This determination is based on the fact that this exemption request is from requirements under 10CFR50 with respect to the installation or use of a facility component located within the restricted area, as defined in 10CFR20, specifically to authorize a change to the licensing basis for ECCS as it relates to acceptable containment sump performance in recirculation mode following a postulated LOCA. The proposed amendment has been evaluated to meet the following criteria under 10OCFR51 .22(c)(9).(i) The exemption involves no significant hazards consideration.

An evaluation of the three criteria set forth in 10CFR50.92(c) as applied to the exemption is provided below. The evaluation is consistent with the no significant hazards consideration determination provided in Attachment 3 in support of the LAR.(1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Approval of the proposed exemption and accompanying license amendment request would allow the results of a risk-informed evaluation to be included in the UFSAR that concludes the ECCS and CSS systems will operate with a high probability following a

NOC-AE-1 5003241 Attachment 2-1 Page 10 of 11 LOCA when considering the impacts and effects of debris accumulation on containment emergency sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss of coolant accidents (LOCAs).The proposed change does not implement any physical changes to the facility or any structures, systems and components (SSCs), and does not implement any changes in plant operation.

The proposed change confirms that required SSCs supported by the containment sumps will perform their safety functions with a high probability, and does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits. The safety analysis acceptance criteria in the UFSAR continue to be met for the proposed change. The proposed change does not affect initiating events because it addresses existing initiating events; i.e., loss of coolant accidents.

The proposed change does not significantly affect the operation of the containment systems needs to ensure that there is a large margin between the temperature and pressure conditions reached in the containment and those that would lead to failure so that there is a high degree of confidence that damage of the containment cannot occur.The calculated risk associated with the proposed change is very small and less than the threshold for Region Ill as defined by RG 1.174, for both CDF and LERF. In accordance with the guidance of RG 1.174, there is substantial safety margin and defense in depth that provide additional confidence that the design basis functions are maintained.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any the accident previously evaluated in the UFSAR.(2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change is a risk-informed analysis of debris effects from accidents that are already evaluated in the STP UFSAR; no new or different kind of accident is being evaluated.

The change neither installs nor removes any plant equipment, nor alters the design, physical configuration, or mode of operation of any plant structure, system or component.

The proposed change does not introduce any new failure mechanisms or malfunctions that can initiate an accident.

The proposed change does not introduce failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident.Therefore, the proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated.

(3) The proposed change does not involve a significant reduction in a margin of safety.The proposed change does not involve a change in any functional requirements, the configuration, or method of performing functions of plant SSCs. The effects from a full spectrum of LOCAs, including double-ended guillotine breaks for all piping sizes up to and including the largest pipe in the reactor coolant system, are analyzed.

Appropriate redundancy and consideration of loss of offsite power and worst case single failure are retained, such that defense-in-depth is maintained.

' :NOC-AE-1 5003241 Attachment 2-1I-J ~Pagel11 ofi11 Application of the risk-informed methodology showed that the increase in risk from the contribution of debris effects is very small as defined by RG 1.174 and that there is adequate defense in depth and safety margin. Consequently, STP determined that the containment sumps will continue to support the ability of safety related components to perform their design functions when the effects of debris are considered.

The proposed change does not alter the manner in which safety limits are determined or acceptance criteria associated with a safety limit. The proposed change does not implement any changes to plant operation, and does not significantly affect SSCs that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

The proposed change does not significantly affect the existing safety margins in the barriers for the release of radioactivity.

There are no changes to any of the safety analyses in the UFSAR. Therefore, the proposed change does not involve a significant reduction in a margin of safety.(ii) The exemption involves no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.No physical modifications or changes to operating requirements are proposed for the facility, including any systems, structures and components relied upon to mitigate the consequences of a LOCA. Approval of the exemption requires the calculated risk associated with GS1-191 to meet the acceptance guidelines in RG 1 .174, thereby maintaining public health and safety. Therefore there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.(iii) The proposed exemption involves no significant increase in individual or cumulative occupational radiation exposure.No physical modifications or changes to operating requirements are proposed for the facility, including any sYstems, structures and components relied upon to mitigate the consequences of a LOCA. Therefore, with respect to installation or use of a facility component located within the restricted area there is no significant increase in individual or cumulative occupational radiation exposure as a result of granting the exemption request.Based on the above, STPNOC concludes that the proposed exemption meets the eligibility criteria for categorical exclusion set forth in 10CFR51 .22(c)(9).

Additional technical justification for this conclusion is provided on the basis that the guidance and acceptance criteria provided in RG 1.174 are satisfied as described in Attachments 1-2, 1-3 and 1-4.

Attachment 2-2 Request for Exemption from 10OCFR50.46(d)

NOC-AE-1 5003241 Attachment 2-2 Page 1 of 8 Request for Exemption from Certain Requirements of 10CFR50.46(d)

1. Exemption Request Pursuant to 10CFR50.12, STP Nuclear Operating Company (STPNOC) is submitting, this request for exemption from certain requirements of 10OCFR50.46(d), "..ot her requirements," as specified in 10OCFR50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors." 10OCFR50.46(d) states: The requirements of this section are in addition to any other requirements applicable to ECCS set forth in this part. The criteria set forth in paragraph (b), with cooling performance calculated in accordance with an acceptable evaluation model, are in implementation of the general requirements with respect to ECCS cooling performance design set forth in this part, including in particular Criterion 35 of appendix A.The STP risk-informed approach to addressing GS1-191 and responding to GL 2004-02 is consistent with the NRC staff safety evaluation of NEI 04-07 that discussed the modeling of sump performance as follows: While not a component of the 10CFR5Q.46 ECCS evaluation model, the calculation of sump performance is necessary to determine if the sump and the residual heat removal system are configured properly to provide enough flow to ensure long-term cooling, which is an acceptance criterion of I0CFR50.46.

Therefore, the staff considers the modeling of sump performance as the validation of assumptions made in the ECCS evaluation model. Since the modeling of sump performance is a boundary calculation for the ECCS evaluation model, and acceptable sump performance is necessary for demonstrating long-term core cooling capability (10CFR50.46(b)

(5)), the requirements of 10CFR50.46 are applicable.

The proposed 10CFR50.46c rule change RIN 3150-AH4 (ML12283A174) clarifies that the specific part of 10CFR50.46 that is affected is 10CFR50.46(d).

That section of 1OCFR50.46 establishes the relation between 10CFR50.46 and the General Design Criteria.

The proposed rule would specifically allow the application of a risk-informed methodology to meet the requirements for assessing the effects of debris.By the reference to the General Design Criteria, 10OCFR50.46(d) incorporates their requirements into 10OCFR50.46.

For consistency with the proposed exemption to GDC 35, STP proposes exemption to 50.46(d).

The result will be that the risk-informed methodology will be allowed rather than the currently required demonstration of mitigation capability by use of a bounding calculation or other deterministic method to model LOCA debris effects, as discussed in Generic Letter 2004-02 and associated guidance documents such as NEI 04-07 and its associated NRC Safety Evaluation.

STPNOC requests an exemption from that requirement in order to enable the use of a risk-informed method to demonstrate acceptable sump performance and LOCA debris mitigation and to validate assumptions in the Emergency Core Cooling System (ECCS) evaluation model.

NOC-AE-1 5003241 Attachment 2-2 Page 2 of 8 The scope of the exemption applies for all debris effects addressed in the STP RoverD methodology described in Attachment 1 that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine fiber debris that exceeds the quantity bounded by STP plant-specific testing described in Attachment

1. That scope is generally described as breaks larger than approximately 12.8" ID in locations where a sufficient amount of fiber debris can be generated and transported to the sump to exceed the amount of fine fiber debris in the STP plant-specific testing described in Attachment

1. Forty-five weld locations have currently been identified on the pressurizer surge line and RCS main loop piping. To minimize the potential that a later analysis could cause the specific locations to change, the requested exemption is based on the breaks' ability to generate sufficient transportable debris, as described in RoverD.The key elements of the exemption request are: 1. It applies only to the effects of debris as described in Attachment 1.2. It applies only for LOCA breaks that can generate and transport fiber debris that is not bounded by STP plant-specific testing.3. It applies to any LOCA break that can generate and transport fiber debris that is not bounded by STP plant-specific testing and is not limited to the 45 specific break locations noted in this application, provided that the ACDF and ALERF associated with the break size remain in Region Ill of RG 1.174.This exemption request is complemented by the' accompanying License Amendment Request (LAR) (Attachment

3) seeking NRC approval of the changes to the South Texas Project Electric Generating Station (STPEGS) Updated Final Safety Analysis Report (UFSAR) and Technical Specifications, to amend the licensing basis based on acceptable design of the containment sump. The risk-informed method provides assurance, with high probability, for acceptable sump performance and debris mitigation during ECCS operation in recirculation mode as calculated by the ECCS evaluation model.2. Regulatory Requirements Involved By regulatory precedent, licensees are required to demonstrate this compliance with the relevant regulations by the use of a bounding calculation or other deterministic method.STPNOC seeks exemption to the extent that 10CFR50.46(d) requires deterministic calculations or other analyses to address the concerns raised by GS1-191 related to acceptable plant performance during the recirculation mode in containment following a LOCA. The proposed changes to the licensing basis and Technical Specifications, submitted for NRC approval with Attachment 3, address GS1-191 and provide closure to GL 2004-02 for STP Units 1 and 2 on the basis that the associated risk is shown to meet the acceptance guidelines in Regulatory Guide (RG) 1.174 and that, in conjunction with the existing licensing basis, adequate safety is demonstrated.

This exemption request is for the purpose of allowing the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated loss of coolant accidents (LOCAs). The effects of LOCA debris have been evaluated, using deterministic methods, to meet the current licensing basis assumptions for analyzing the effects of post-LOCA debris blockage in the sump and in-vessel; however, these NOC-AE-1 5003241 Attachment 2-2 Page 3 of 8 evaluations have not been shown to fully address debris effects for the as-built, as-operated plant. The risk-informed approach evaluates the debris effects as part of the assessment of the residual risk associated with GSI-1 91 concerns.

Based on confirmation of acceptable EGOS design as determined by the resulting risk meeting the acceptance guidelines in RG 1.174, the licensing basis for ECCS compliance with 10CFR50.46(d) is amended.2.1 Evaluation of Impacts on the Balance of 10CFR50.46 and Appendix K to I0CFR50 The exemption request to support closure for GL 2004-02 for STP is intended to address EGOS cooling performance design as presented in 10CFR50.46(d) as it relates to imposing the requirements of the General Design Criteria.For the purposes of demonstrating the balance of the acceptance criteria of 10OCFR50.46, the design and licensing basis descriptions of accidents requiring ECGS operation, including analysis methods, assumptions, and results, which are provided in South Texas Project Electric Generating Station (STPEGS) Updated Final Safety Analysis Report (UFSAR) Chapters 6 and 15 remain unchanged.

The performance evaluations for accidents requiring ECCS operation described in UFSAR Chapters 6 and 15, based on the Appendix K Large-Break Loss-of-Coolant Accident (LBLOCA) analysis, demonstrate that for breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in 10CFR50.46 and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

The requirements of 10CFR50.46(a)(1) remain applicable to the model of record that meets the required features of Appendix K. Approval of the requested exemption does not impact the current EGGS evaluation.

This evaluation model remains the licensing basis for demonstrating that the EGGS calculated cooling performance following postulated LOCAs does not exceed the acceptance criteria.The STP risk-informed approach uses the break frequencies from NUREG 1829 to quantify the residual risk associated with GS1-191 for those LOCAs which have not been resolved using deterministic methods, and shows that it meets the acceptance guidelines defined in RG 1.174. The exemption request is specific to the requirement for demonstrating ECCS cooling performance design as required by 10CFR50.46(d) as it pertains to the application of the General Design Criteria, and to provide regulatory consistency between the requirements of 10CFR5O.46(d) and the GDC. It is not intended to be applicable to other requirements provided in 10GFR50.46 or Appendix K to 10CFR50.As noted above, the NRC staff considers the modeling of sump performance as an input to the EGGS evaluation model, and therefore the requirements of 10CFR50.46 are applicable.

Consistent with this, the requirements and attributes for the proposed STP risk-informed method include a full spectrum of postulated, double-ended guillotine breaks is evaluated, up to and including the largest piping in containment.

NOC-AE-1 5003241 Attachment 2-2 Page 4 of 8 Engineering analyses and evaluations used to perform plant-specific testing to address the deterministic scope of the STP analysis consider a wide range of effects, including those addressed in NEI 04-07 and related NRC guidance for evaluation of sump performance.

The proposed exemption does not affect any of the 10CFR50.46 (a)(1) or Appendix K requirements for an acceptable ECCS evaluation model, and does not change the EGGS acceptance criteria in 50.46(b) as it applies to the calculated results. Application of the exemption request allows use of a risk-informed approach to evaluate the effects of LOCA debris associated with GS1-191 that may be present in an acceptable evaluation model. The results of the risk-informed method demonstrate that the risk associated with GS1-191 meets the acceptance guidelines of RG 1.174. The current licensing basis for addressing the adequacy of EGGS to meet the criteria of 100FR50.46, including the Appendix K Large-Break LOCA analysis and the associated Chapter 15 accident analysis for LOCA, remain in place.2.1 Evaluation of Impacts on other Regulatory Requirements

-Conclusion The proposed exemption dioes not result in any physical changes to the facility or changes to the operation of the plant, and does not change any of the programmatic requirements.

Based on demonstrating acceptable LOCA debris mitigation and containment emergency sump and EGGS design for amending the current licensing basis for 10CFR50.46(d) as described above, compliance with other regulatory requirements that rely on acceptable design for these systems and components continue to be met in the current licensing basis.3. Basis for the Exemption Request Under 10CFR50.12, a licensee may request and the NRC may grant exemptions from the requirements of 10CFR5O which are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and when special circumstances are present.The exemption request meets a key principle of RG 1.174, which states "The proposed change meets the current regulations unless it is explicitly related to a requested exemption." This exemption request is provided in conjunction with the proposed License Amendment Request in Attachment

3.3.1 Justification

for the Exemption Request As required by 1OCFR50.12(a)(2), the Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR50.12(a)(2) are applicable.

STPNOC has evaluated the proposed exemption against the conditions specified in I0CFR50.12(a) and determined that this proposed exemption meets the requirements for granting an exemption from the regulation, and that special circumstances are present.The information supporting the determination is provided below.Pursuant to 10OCFR50.

12, "Specific exemptions," the NRC may grant exemptions from the requirements of this part provided the following three conditions are met as required by 10OCFR50.1I2(a)(1):

NOC-AE-1 5003241 Attachment 2-2 Page 5 of 8 The exemption is authorized by law.The NRC has authority under the Atomic Energy Act of 1954, as amended, to grant exemptions from its regulations if doing so would not violate the requirements of law. This exemption is authorized by law as is provided by 10CFR50.12 which provides the NRC authority to grant exemptions from 10CFR50 requirements with provision of proper justification.

Approval of the exemption from 10CFR50.46(d) would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, any of the Commission's regulations, or any other law.The exemption does not present an undue risk to the public health and safety.The purpose of 10CFR50.46 is to establish acceptance criteria for ECCS performance, and together with GDC 35, to provide a high confidence that the systems will perform th'e required functions.

The proposed exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated initiating events is not increased.

The PRA and engineering analysis demonstrate that the calculated risk is small and consistent with the intent of the Commission's Safety Goal Policy Statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.As discussed in previous 10CFR50.46 rulemaking, the probability of a large break LOCA is sufficiently low that the application of a risk-informed approach to evaluate the ability of the ECCS to meet 10CFR50.46 and relevant GDC with high probability and with low uncertainty, rather than using a calculational model using deterministic methods to achieve similar understanding, would have little effect on public risk. This is applicable to evaluating acceptable containment sump design in support of ECCS and CSS recirculation modes.The proposed change is to apply a risk-informed method rather than a traditional deterministic method to quantify the risk associated with GS1-191 and to establish a high probability of success for performance of ECCS in accordance with the ECCS cooling performance design addressed in 10CFR50.46(d).

The risk-informed approach involves a complete evaluation of the spectrum of LOCA breaks, including double-ended guillotine breaks, up to and including the largest pipe in the reactor coolant system. The risk-informed approach analyzes LOCAs, regardless of break size, using the same methods, assumptions, and criteria in order to quantify the uncertainties and overall risk metrics.This ensures that large break LOCAs with relatively small contribution to CDF due to the low probability of such a break as well as smaller break LOCAs with higher probabilities of occurrence are considered in the results. Since the design basis requirement for consideration of a double-ended guillotine breaks of the largest pipe in the reactor coolant system is retained and since no physical changes to the facility or changes to the operation of the facility are being made, the existing defense-in-depth and safety margin established for the design of the facility is not reduced.

NOC-AE-1 5003241 Attachment 2-2 Page 6 of 8 This exemption only affects 10CFR50.46(d), and does not impact the adequacy of the acceptance criteria for cladding performance that is important to maintain adequate safety margins.The exemption is consistent with the common defense and security.The exemption involves a change to the licensing basis for the plant that has no relation to the control of licensed material or any security requirements that apply to STP Units 1 and 2. Therefore the exemption is consistent with the common defense and security.3.2 Special Circumstances This section discusses the presence of special circumstances as related to 10CFR50.12(a).

10CFR50.12(a)(2) states that NRC will not consider granting an exemption to the regulations unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR50.12(a)(2) are applicable.

Such special circumstances are present in this instance to warrant exemption from the implicit requirement in 10CFR50.46(d) to reference GDC which use a deterministic calculational method as the design basis for acceptable sump performance to validate the results of the ECCS evaluation model demonstrating long-term cooling criterion is met.Approval of this exemption request would allow the use a risk-informed method to amend the design basis for acceptable performance of the containment emergency sump, as a validation of inputs in the ECCS evaluation model, and in support of the existing licensing bases for compliance with 10CFR50.46.

Specifically, 10OCFR50.1I2(a)(2)(ii) applies: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.The intent of 10OCFR50.46(d) is to ensure ECCS cooling performance design requirements imposed by 10CFR50.46 are complemented by the requirements of the relevant GDC.This exemption request is consistent with that purpose in that use of the proposed risk-informed approach accounts for the effect of debris on the ECCS cooling performance and supports a high probability of successful ECCS performance, based on the risk results meeting the acceptance guidelines of RG 1.174.As discussed in the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities", NRC regulations and their implementation are generally based on deterministic approaches that consider a set of challenges to safety and determine how those challenges should be mitigated.

The need is. based for this exemption is based on the requirements in the regulations for using deterministic methods to demonstrate acceptable design.Regulatory requirements are largely based on a deterministic framework, and NOC-AE-1 5003241 Attachment 2-2 Page 7 of 8 are established for design basis accidents, such as the LOCA, with specific acceptance criteria that must be satisfied.

Licensed facilities must be provided with safety systems capable of preventing and mitigating the consequences of design basis accidents to protect public health and safety. The deterministic regulatory requirements were designed to ensure that these systems are highly reliable.

The LOCA analysis and the General Design Criteria (GDC) were established as part of this deterministic regulatory framework.

In comparison, the risk-informed approach considers nuclear safety in a more comprehensive way by examining the likelihood of a broad spectrum of initiating events and potential challenges, considering a wide range of credible events and assessing the risk based on mitigatin~g system reliability.

An objective of 10CFR50.46 is to maintain low risk to the public health and safety through a reliable ECCS, as supported by the containment sump. The supporting analysis demonstrates that a risk-informed approach to sump performance is consistent with the Commission's Safety Goals for nuclear power plants and supports ECCS operation with a high degree of reliability.

Consequently, the special circumstances described in 10OCFR50.1I2(a)(2)(ii) apply.Specifically, 1OCFR50.1I2(a)(2)(iii) applies: Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.The specific hardship is the excessive 158 -.176 remn occupational radiological dose that are estimated to be incurred for plant modifications to remove insulation.

This was described in Attachment 2-1, above.In conclusion, special circumstances in 10OCFR50.1I2(a)(2)(ii) and 10OCFR50.1I2(a)(2)(iii) are present as required by 10CFR50.12(a)(2) for consideration of the request for exemption.

4. Technical Justification for the Exemption Technical justification for the risk-informed method is provided in Attachment 1 and Attachment 3.The proposed risk-informed approach meets the key principles in RG 1.174 in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in small increase in risk, and is monitored using performance measurement strategies.

This proposed exemption to allow use of the risk-informed method is consistent with the key principle in RG 1.174 that requires the proposed change to meet current regulations unless explicated related to a requested exemption.

NOC-AE-1 5003241 Attachment 2-2 Page 8 of 8 The results show that the risk associated with GS1-1 91 concerns is less than the threshold in Region Ill, "Very Small Changes," of RG 1.174, and therefore are consistent with the Commission's Safety Goals for public health and safety.5. Conclusion Approval of an exemption to allow the use of the risk-informed approach is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security as required by 10CFR50.12(a)(1).

Furthermore, special circumstances required by 10CFR50.12(a)(2) are present for item 10CFR50.12(a)(2)(ii) in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.Based on the determination that the risk of the exemption meets the acceptance guidelines of RG 1.174, the results demonstrate there is reasonable assurance that the ECCS will function in the recirculation mode and that the public health and safety will be protected.

6. Implementation STPNOC requests that this exemption request be approved for implementation by November 2015.

NOC-AE-1 5003241 Attachment 2-3 Request for Exemption from General Design Criterion 35 NOC-AE-1 5003241 Attachment 2-3 Page 1 of 7 Request for Exemption from Certain Requirements of General Design Criterion 35 1. Exemption Request Pursuant to 10CFR5O.12, STPNOC is submitting this request for exemption from certain requirements of 10CFR50 Appendix A, General Design Criterion (GDC) 35, which states: Criterion 35-N Emergency core cooling. A system to provide abundant emergency core cooling shall be provided.

The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.By regulatory precedent, licensees are required to demonstrate this capability by the use of a bounding calculation or other deterministic method. STPNOC requests an exemption from those deterministic requirements in order to enable the use of a risk-informed method to demonstrate acceptable sump design and ECCS performance with regard to the effects of LOCA debris.Approval of this exemption will allow use of a risk-informed method to account for the probabilities and uncertainties associated with mitigation of the effects of debris following postulated LOCAs. The method evaluates the effects on strainer blockage and core blockage resulting from debris concerns raised by GS1-191. In order to confirm acceptable sump design, the risk associated with GSI-191 is evaluated to include the failure mechanisms associated with loss of core cooling and strainer blockage.The scope of the exemption applies for all debris effects addressed in the STP RoverD methodology described in Attachment 1 that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine fiber debris that exceeds the quantity bounded by STP plant-specific testing described in Attachment

1. That scope is generally described as breaks larger than approximately 12.8" ID in locations where a sufficient amount of fiber debris can be generated and transported to the sump to exceed the amount of fine fiber debris in the STP plant-specific testing described in Attachment

1. Forty-five weld locations have currently been identified on the pressurizer surge line and RCS main loop piping. To minimize the potential that a later analysis could cause the specific locations to change, the requested exemption is based on the breaks' ability to generate sufficient transportable debris, as described in RoverD.The key elements of the exemption request are: 1. It applies only to the effects of debris as described in Attachment

1.

NOC-AE-1 5003241 Attachment 2-3 Page 2 of 7 2. It applies only for LOCA breaks that can generate and transport fiber debris that is not bounded by STP plant-specific testing.3. It applies to any LOCA break that can generate and transport fiber debris that is not bounded by STP plant-specific testing and is not limited to the 45 specific break locations noted in this application, provided that the ACDF and ALERF associated with the break size remains in Region IlI of RG 1.174.This exemption request is complemented by the accompanying License Amendment Request ([AR) (Attachment

3) seeking NRC approval of the methodology changes that will be incorporated in the South Texas Project Electric Generating Station (STPEGS)Updated Final Safety Analysis Report (UFSAR) and Technical Specifications, to amend the licensing basis based on acceptable design of the containment sump. The risk-informed method provides assurance, with high probability, for acceptable sump performance and debris mitigation during ECCS operation in recirculation mode as calculated by the ECCS evaluation model. /2. Regulatory Requirements Involved STPNOC seeks exemption to the extent that GDC 35 requires deterministic calculations or other analyses to address the concerns raised by GS1-191 related to acceptable plant performance during the recirculation mode in containment following a LOCA. The proposed changes to the licensing basis and Technical Specifications, submitted for NRC approval with the LAR in Attachment 3, address GS1-191 and provide closure to GL 2004-02 for STP Units 1 and 2 on the basis that the associated risk is shown to meet the acceptance guidelines in Regulatory Guide (RG) 1.174 and that, in conjunction with the existing licensing basis, adequate safety is demonstrated.

This exemption request is for the purpose of allowing the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated loss of coolant accidents (LOCAs). The effects of LOCA debris have been evaluated, using deterministic methods, to meet the current licensing basis assumptions for analyzing the effects of post-LOCA debris blockage in the sump and in-vessel; however, these evaluations have not been shown to fully address debris effects for the as-built, as-operated plant. The risk-informed approach evaluates the debris effects as part of the assessment of the residual risk associated with GS1-191 concerns.

Based on confirmation of acceptable ECCS design as determined by the resulting risk meeting the acceptance guidelines in RG 1.174, the licensing basis for ECCS compliance with GDC 35 is amended.2.1 Evaluation of Impacts on other Regulatory Requirements

-Conclusion The proposed exemption does not result in any physical changes to the facility or changes to the operation of the plant, and does not change any of the programmatic requirements.

Based on demonstrating acceptable containment emergency sump and ECCS design for amending the current licensing basis for compliance with GDC 35 as described above, compliance with other regulatory requirements that rely on acceptable design for these systems and components continue to be met in the current licensing basis.

NOC-AE-1 5003241 Attachment 2-3 Page 3 of 7 3. Basis for the Exemption Request Under 10CFR50. 12, a licensee may request and the NRC may grant exemptions from the requirements of 10CFR50 which are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and when special circumstances are present.The exemption request meets a key principle of RG 1.174, which states "The proposed change meets the current regulations unless it is explicitly related to a requested exemption." This exemption request is provided in support of the proposed change provided in the License Amendment Request provide in Attachment

3.3.1 Justification

for the Exemption Request As required by 10CFR50.12(a)(2), the Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR50.12(a)(2) are applicable.

STPNOC has evaluated the proposed exemption against the conditions specified in 10CFR50.12(a) and determined that this proposed exemption meets the requirements for granting an exemption from the regulation, and that special circumstances are present.The' information supporting the determination is provided below.Pursuant to 1OCFR50.12, "Specific exemptions," the NRC may grant exemptions from the requirements of this part provided the following three conditions are met as r'equired by 10OCER50.1I2(a)(1):

The exemption is authorized by law.The NRC has authority under the Atomic Energy Act of 1954, as amended, to grant exemptions from its regulations if doing so would not violate the requirements of law. This exemption is authorized by law as is provided by 10CFR50.12 which provides the NRC authority to grant exemptions from 10CER50 requirements with provision of proper justification.

Approval of the exemption would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, the Commission's regulations, or any other law The exemption does not present an undue risk to the public health and safety.The proposed change is to apply a risk-informed method rather than a traditional deterministic method in order to quantify the residual risk associated with GS1-191 and to establish a high confidence of acceptable ECCS design. The purpose of GDC 35 is to establish acceptable design for ECCS, and together with the acceptance criteria of 10CFR50.46, to provide a high probability that the systems will perform the required functions.

The proposed exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated initiating events is not increased.

The PRA and engineering analysis demonstrate that the calculated risk is very small and consistent with the intent of the Commission's Safety Goal Policy Statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.

NOC-AE-1 5003241 Attachment 2-3 Page 4 of 7 As discussed in previous I0CFR50.46 rulemaking, the probability of a large break LOCA is sufficiently low that the application of a risk-informed approach to evaluate the ability of the ECCS to meet its design requirements with high probability and with low uncertainty, rather than using a calculational model using deterministic methods to achieve similar understanding, would have little effect on public risk. This is applicable to evaluating the effects of debris on acceptable ECCS design during the recirculation modes.The risk-informed approach involves a complete evaluation of the spectrum of LOCA breaks, including double-ended guillotine break, up to and including the largest pipe in the reactor coolant system. The risk-informed approach analyzes LOCAs, regardless of break size, using the same methods, assumptions, and criteria in order to quantify the uncertainties and overall risk metrics. This ensures that large break LOCAs with relatively small contribution to CDF due to the low probability of such a break as well as smaller break LOCAs with higher probabilities of occurrence are considered in the results. Since the design basis requirement for consideration of a double-ended guillotine break of the largest pipe in the reactor coolant system is retained and since no physical changes to the facility or changes to the operation of the facility are being made, the existing defense-in-depth and safety margin established for the design of the facility is not reduced.The exemption is consistent with the common defense and security.The exemption involves a change to the licensing basis for the plant that has no relation to the possession of licensed material or any security requirements that apply to STP Units 1 and 2. Therefore the exemption is consistent with the common defense and security.3.2 Special Circumstances This section discusses the presence of special circumstances as related to 10CFR50.12(a).

100FR5O.12(a)(2) states that NRC will not consider granting an exemption to the regulations unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR5O.12(a)(2) are applicable.

Such special circumstances are present in this instance to warrant exemption from the implicit requirement in GDC 35 to use a deterministic method to evaluate for acceptable containment emergency sump design. Approval of the exemption request would allow use of a risk-informed method to amend the design basis for acceptable containment sump design in support of ECCS design for compliance with GDC 35. Specifically, 10OCFR50.1I2(a)(2)(ii) applies: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.The intent of GDC 35 is to ensure ECCS design provides abundant core cooling to mitigate fuel and clad damage and clad metal-water reaction following any loss of reactor coolant.GDC 35 sets forth the general ECCS cooling performance design requirements, which are NOC-AE-1 5003241 Attachment 2-3 Page 5 of 7 in addition to the requirements of I0CFR50.46.

This exemption request is consistent with that purpose in that use of the proposed risk-informed approach demonstrates a high probability of successful ECGS performance, which includes realistically available long term cooling, based on the risk results meeting the acceptance guidelines of RG 1.174.The risk-informed approach assesses EGGS design for a full spectrum of breaks, and assesses equipment failures that include loss of offsite power and worst case single failure, consistent with the GDC 35 requirements.

Since the proposed exemption does not involve any physical changes to the plant, there is no effect on the GDC 35 requirements for EGOS design for redundancy in components and features, interconnections, leak detection, isolation, and containment capabilities.

The current licensing basis evaluations for EGGS compliance with GDC 35 for these aspects continue to be met.As discussed in the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities", NRC regulations and their implementation are generally based on deterministic approaches that consider a set of challenges to safety and determine how those challenges should be mitigated.

This request does not seek exemption from any explicit language in the regulatory requirements.

Rather, the need is based on the implicit requirements in the regulations for using deterministic methods to demonstrate acceptable design. Regulatory requirements are largely based on a deterministic framework, and are established for design basis accidents, such as the LOGA, with. specific acceptance criteria that must be satisfied.

Licensed facilities must be provided with safety systems capable of preventing and mitigating the consequences of design basis accidents to protect public health and safety. The deterministic regulatory requirements were designed to ensure that these systems are highly reliable.

The LOGA analysis and the General Design Griteria (GDG) were established as part of this deterministic regulatory framework.

In comparison, the probabilistic approach considers nuclear safety in a more comprehensive way by examining the likelihood of a broad spectrum of initiating events and potential challenges, considering a wide range of credible events and assessing the risk based on mitigating system reliability.

An objective of GDC 35 is to maintain low risk to the public health .and safety through a reliable EGGS, as supported by the containment sump. The supporting analysis demonstrates that a risk-informed approach to sump performance is consistent with the Gommission's Safety Goals for nuclear power plants, and supports EGGS operation with a high degree of reliability.

Gonsequently, the special circumstances described in I OCFR50.1I2(a)(2)(ii) apply.

NOC-AE-1 5003241 Attachment 2-3 Page 6 of 7 Specifically, I10CFR50.12(a)(2)(iii) applies: Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.The specific hardship is the excessive 158 -176 rem occupational radiological dose that are estimated to be incurred for plant modifications to remove insulation.

This was described in Attachment 2-1.In conclusion, special circumstances in 10CFR50.12(a)(2)(ii) and 10CFR50.12(a)(2)(iii) are present as required by 100FR50.12(a)(2) for consideration of the request for exemption.

4. Technical Justification for the Exemption Technical

]ustification for the risk-informed method is provided in Attachment 1 and in the LAR (Attachment 3).The proposed risk-informed approach meets the key principles in RG 1.174 in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in small increase in risk, and is monitored using performance measurement strategies.

This proposed exemption to allow use of the risk-informed method is consistent with the key principle in RG 1.174 that requires the proposed change to meet current regulations unless explicated related to a requested exemption.

The results show that the risk associated with GS1-1 91 concerns is less than the threshold in Region Ill, "Very Small Changes," of RG 1.174, and therefore are consistent with the Commission's Safety Goals for public health and safety.5. Conclusion Approval of an exemption to allow the use of the risk-informed approach is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security as required by 10CFR50.12(a)(1).

Furthermore, special circumstances required by 10CFR50.12(a)(2) are present for item 10CFR50.12(a)(2)(ii) in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.Based on the determination that the risk of the exemption meets the acceptance guidelines of RG 1.174, the results demonstrate there is reasonable assurance that the ECCS will function in the recirculation mode and that the public health and safety will be protected.

NOC-AE-1 5003241 Attachment 2-3 Page 7 of 7 6. Implementation STPNOC requests that this exemption request be approved for implementation by November 2015.

NOC-AE-1 5003241 Attachment 2-4 Request for Exemption from General Design Criterion 38 NOC-AE-1 5003241 Attachment 2-4 Page 1 of 6 Request for Exemption from Certain Requirements of General Design Criterion 38 1. Exemption Request Pursuant to 10CFR50.12, STPNOC is submitting this request for exemption from certain requirements of 10CFR50 Appendix A, General Design Criteria (GDC) 38.Criterion 38-- Containment heat removal. A system to remove heat from the reactor containment shall be provided.

The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system~safety function can be accomplished, assuming a single failure.By regulatory precedent, licensees are required to demonstrate this capability by the use of a bounding calculation or other deterministic method. STPNOC requests an exemption from those deterministic requirements in order to enable the use of a risk-informed method to demonstrate acceptable sump design and ECCS and CSS performance with regard to the effects of LOCA debris.Approval of this exemption will allow use of a risk-informed method to account for the probabilities and uncertainties associated with mitigation of the effects of debris following postulated LOCAs. The method evaluates the effects on strainer blockage and ~core blockage resulting from debris concerns raised by GS1-1 91. In order to confirm acceptable sump design, the risk associated with GS1-191 is evaluated to include the failure mechanisms associated with loss of core cooling and strainer blockage.The scope of the exemption applies for all debris effects addressed in the STP RoverD methodology described in Attachment 1 that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine fiber debris that exceeds the quantity bounded by STP plant-specific testing described in Attachment

1. That scope is generally described as breaks larger than approximately 12.8" ID in locations where a sufficient amount of fiber debris can be generated and transported to the sump to exceed the amount of fine fiber debris in the STP plant-specific testing described in Attachment

1. Forty-five weld locations have currently been identified on the pressurizer surge line and RCS main loop piping. To minimize the potential that a later analysis could cause the specific locations to change, the requested exemption is based on the breaks' ability to generate sufficient transportable debris, as described in RoverD.The key elements of the exemption request are: 1. It applies only to the effects of debris as described in Attachment

1.

NOC-AE-1 5003241 Attachment 2-4 Page 2 of 6 2. It applies only for LOCA breaks that can generate and transport fiber debris that is not bounded by STP plant-specific testing.3. It applies to any LOCA break that can generate and transport fiber debris that is not bounded by STP plant-specific testing and is not limited to the 45 specific break locations noted in this application, provided that the ACDF and ALERF associated with the break size remains in Region Ill of RG 1.174.This exemption request is complemented by the accompanying License Amendment Request (LAR) (Attachment

3) seeking NRC approval of the changes to the South Texas Project Electric Generating Station (STPEGS) Updated Final Safety Analysis Report (UFSAR) and Technical Specifications, to amend the licensing basis based on acceptable design of the containment sump. ,The risk-informed method provides assurance, with high probability, for acceptable sump performance and debris mitigation during ECCS operation in recirculation mode as calculated by the ECCS evaluation model.2. Regulatory Requirements Involved STPNOC seeks exemption to the extent that GDC 38 requires deterministic calculations or other analyses to address the concerns raised by GS1-191 related to acceptable plant performance during the recirculation mode in containment following a LOCA. The proposed changes to the licensing basis and Technical Specifications, submitted for NRC approval with the LAR in Attachment 3, address GS1-1 91 and provide closure to GL 2004-02 for STP Units 1 and 2 on the basis that the associated risk is shown to meet the acceptance guidelines in Regulatory Guide (RG) 1.174 and that, in conjunction with the existing licensing basis, adequate safety is demonstrated.

This exemption request is for the purpose of allowing the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated loss of coolant accidents (LOCAs). The effects of LOCA debris have been evaluated, using deterministic methods, to meet the current licensing basis assumptions for analyzing the effects of post-LOCA debris blockage in the sump and in-vessel; however, these evaluations have not been shown to fully address debris effects for the as-built, as-operated plant. The risk-informed approach evaluates the debris effects as part of the assessment of the residual risk associated with GS1-191 concerns.

Based on confirmation of acceptable ECCS design as determined by the resulting risk meeting the acceptance guidelines in RG 1.174, the licensing basis for ECCS compliance with GDC 38 is amended.2.1 Evaluation of Impacts on other Regulatory Requirements

-Conclusion The proposed exemption does not result in any physical changes to the facility or changes to the operation of the plant, and does not change any of the programmatic requirements.

Based on demonstrating acceptable containment heat removal design for amending the current licensing basis for compliance with GDC 38 as described above, compliance with other regulatory requirements that rely on acceptable design for these systems and components continue to be met in the current licensing basis.

NOC-AE-1 5003241 Attachment 2-4 Page 3 of 6 3. Basis for the Exemption Request Under 10CFR50.12, a licensee may request and the NRC may grant exemptions from the requirements of 10CFR5O which are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and when special circumstances are present.3.1 Justification for the Exemption Request As required by 10CFR50.12(a)(2), the Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR50.12(a)(2) are applicable.

STPNOC has evaluated the proposed exemption against the conditions specified in 10CER50. 12(a) and determined that this proposed exemption meets the requirements for granting an exemption from the regulation, and that special circumstances are present.The information supporting the determination is provided below.Pursuant to 10CFR50.12, "Specific exemptions," the NRC may grant exemptions from the requirements of this part provided the following three conditions are met as required by 10OCER50.

12(a)(1): The exemption is authorized by law.The NRC has authority under the Atomic Energy Act of 1954, as amended, to grant exemptions from its regulations if doing so would not violate the requirements of law. This exemption is authorized by law as is provided by 10CFR50.12 which provides the NRC authority to grant exemptions from 10CFR50 requirements with provision of proper justification.

Approval of the exemption would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, the Commission's regulations, or any other law The exemption does not present an undue risk to the public health and safety.The proposed change is to apply a risk-informed method rather than a traditional deterministic method in order to quantify the residual risk associated with GS1-191 and to establish a high confidence of acceptable containment sump design. The purpose of GDC 38 is to establish acceptable design for the containment heat removal system, which includes the CSS and functions in conjunction with ECCS, to provide a high probability that the systems will perform the required functions.

The proposed exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated initiating events is not increased.

The PRA and engineering analysis demonstrate that the calculated risk is small and consistent with the intent of the Commission's Safety Goal Policy Statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.As discussed in previous 10CFR50.46 rulemaking, the probability of a large break LOCA is sufficiently low that the application of a risk-informed approach to evaluate the ability of the ECCS to meet 10CFR50.46(d) with high probability and with low uncertainty, rather than using a calculational model using deterministic methods to achieve similar NOC-AE-1 5003241 Attachment 2-4 Page 4 of 6 understanding, would have little effect on public risk. This is applicable to evaluating acceptable containment sump design in support of ECCS and CSS recirculation modes.The risk-informed approach involves a complete evaluation of the spectrum of LOCA breaks, including double-ended guillotine break, up to and including the largest pipe in the reactor coolant system. The risk-informed approach analyzes LOCAs, regardless of break size, using the same methods, assumptions, and criteria in order to quantify the uncertainties and overall risk metrics. This ensures that large break LOCAs with relatively small contribution to CDF due to the low probability of such a break as well as smaller break LOCAs with higher probabilities of occurrence are considered in the results. Since the design basis requirement for consideration of a double-ended guillotine break of the largest pipe in the reactor coolant system is retained and since no physical changes to the facility or changes to the operation of the facility are being made, the existing defense-in-depth and safety margin established for the design of the facility is not reduced.The exemption is consistent with the common defense and security.The exemption involves a change to the licensing basis for the plant that has no relation to the possession of licensed material or any security requirements that apply to STP Units 1 and 2. Therefore the exemption is consistent with the common defense and security.3.2 Special Circumstances This section discusses the presence of special circumstances as related to 10CFR50.12(a).

10CFR50.12(a)(2) states that NRC will not consider granting an exemption to the regulations unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 100FR50.12(a)(2) are applicable.

Such special circumstances are present in this instance to warrant exemption from the implicit requirement in GDC 38 to use a deterministic method to evaluate for acceptable containment emergency sump design. Approval of the exemption request would allow use of a risk-informed method to amend the design basis for acceptable containment sump design in support of CSS design for compliance with GDC 38. Specifically, 10OCFR50.

12(a)(2)(ii) applies: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.The intent of GDC 38 is to ensure a containment heat removal system is provided to rapidly reduce containment pressure and temperature following any LOCA and maintain them at acceptably low levels. This exemption request is consistent with that purpose in that use of the proposed risk-informed approach demonstrates a high probability of successful ECCS and CSS performance, which includes realistically available recirculation flow, based on the risk results meeting the acceptance guidelines of RG 1.174. The risk-informed approach assesses the design for a full spectrum of breaks, and NOC-AE-1 5003241 Attachment 2-4 Page 5 of 6 assesses equipment failures that include loss of offsite power and worst case single failure, consistent with the GDC 38 requirements.

Since the proposed exemption does not involve any physical changes to the plant, there is no affect on the GDC 38 design requirements for redundancy in components and features, interconnections, leak detection, isolation, and containment capabilities.

The current licensing basis evaluations for ECCS and CSS compliance with GDC 38 for these aspects continue to be met.As discussed in the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities", NRC regulations and their implementation are generally based on deterministic approaches that consider a set of challenges to safety and determine how those challenges should be mitigated.

This request does not seek exemption from any explicit language in the regulatory requirements.

Rather, the need is based on the implicit requirements in the regulations, for using deterministic methods to demonstrate acceptable design. Regulatory requirements are largely based on a deterministic framework, and are established for design basis accidents, such as the LOCA, with specific acceptance criteria that must be satisfied.

Licensed facilities must be provided with safety systems capable of preventing and mitigating the consequences of design basis accidents to protect public health and safety. The deterministic regulatory requirements were designed to ensure that these systems are highly reliable.

The LOCA analysis and the General Design Criteria (GDC) were established as part of this deterministic regulatory framework.

In comparison, the risk-informed approach considers nuclear safety in a more comprehensive way by examining the likelihood of a broad spectrum of initiating events and potential challenges, considering a wide range of credible events and assessing the risk based on mitigating system reliability.

An objective of GDC 38 is to maintain low risk to the public health and safety through a reliable CSS, as supported by the containment sump. The supporting analysis demonstrates that a risk-informed approach to sump performance is consistent with the Commission's Safety Goals for nuclear power plants, and supports 0S5 operation with a high degree of reliability.

Consequently, the special circumstances described in 10OCFR50.

12(a)(2)(ii) apply.Specifically, 10OCFR50.1I2(a)(2)(iii) applies: Compliance would result in undue hardship or other costs that are significantly in excess* of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

NOC-AE-1 5003241 Attachment 2-4 Page 6 of 6 The specific hardship is the excessive 158 -176 rem occupational radiological dose that are estimated to be incurred for plant modifications to remove insulation.

This was described in Attachment 2-1.In conclusion, special circumstances in 10CFR50.12(a)(2)(ii) and 10CFR50.12(a)(2)(iii) are present as required by 10CFR50.12(a)(2) for consideration of the request for exemption.

4. Technical Justification for the Exemption Technical justification for the risk-informed method is provided in Attachment 1 and in the LAR (Attachment 3).The proposed risk-informed approach meets the key principles in RG 1.174 in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in small increase in risk, and is monitored using performance measurement strategies.

This proposed exemption to allow use of the risk-informed method is consistent with the key principle in RG 1.174 that requires the proposed change to meet current regulations unless explicated related to a requested exemption.

The results show that the risk associated with GS1-1 91 concerns is less than the threshold in Region Ill, "Very Small Changes," of RG 1.174, and therefore are consistent with the Commission's Safety Goals for public health and safety.5. Conclusion Approval of an exemption to allow the use of the risk-informed approach is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security as required by 10CFR50.12(a)(1).

Furthermore, special circumstances required by 10CFR50.12(a)(2) are present for item 10CFR50.12(a)(2)(ii) in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.Based on the determination that the risk of the exemption meets the acceptance guidelines of RG 1.174, the results demonstrate there is reasonable assurance that the ECCS and CSS will function in the recirculation mode and that the public health and safety will be protected.

6. Implementation STPNOC requests that this exemption request be approved for implementation by November 2015.

NOC-AE-1 5003241 Attachment 2-5 Request for Exemption from General Design Criterion 41 NOC-AE-1 5003241 Attachment 2-5 Page 1 of 7 Request for Exemption from Certain Requirements of General Design Criterion 41 1. Exemption Request Pursuant to 10CFR50.12, STPNOC is submitting this request for exemption from certain requirements of 10OCFR50 Appendix A, General Design Criteria (GDC) 41.Criterion 41- Containment atmosphere cleanup. Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.By regulatory precedent, licensees are required to demonstrate this capability by the use of a bounding calculation or other deterministic method. STPNOC requests an exemption from those deterministic requirements in order to enable the use of a risk-informed method to demonstrate acceptable sump design and Containment Cleanup performance with regard to the effects of LOCA debris.Approval of this exemption will allow use of a risk-informed method to account for the probabilities and uncertainties associated with mitigation of the effects of debris following postulated LOCAs. The method evaluates the effects on strainer blockage and core blockage resulting from debris concerns raised by GS1-1 91. In order to confirm acceptable sump design, the risk associated with GS1-191 is evaluated to include the failure mechanisms associated with loss of core cooling and strainer blockage.The scope of the exemption applies for all debris effects addressed in the STP RoverD methodology described in Attachment I that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine fiber debris that exceeds the quantity bounded by STP plant-specific testing described in Attachment

1. That scope is generally described as breaks larger than approximately 12.8" ID in locations where a sufficient amount of fiber debris can be generated and transported to the sump to exceed the amount of fine fiber debris in the STP plant-specific testing described in Attachment

1. Forty-five weld locations have currently been identified on the pressurizer surge line and RCS main loop piping. To minimize the potential that a later analysis could cause the specific locations to change, the requested exemption is based on the breaks' ability to generate sufficient transportable debris, as described in RoverD.The key elements of the exemption request are:

NOC-AE-1 5003241 Attachment 2-5 Page 1 of 7 Request for Exemption from Certain Requirements of General Design Criterion 41 1. Exemption Request P~ursuant to 10OCFR50.

12, STPNOC is submitting this request for exemption from certain requirements of 10OCFR50 Appendix A, General Design Criteria (GDC) 41.S Criterion 41- Containment atmosphere cleanup. Systems to control fission* products, hydrogen, oxygen, and other substances which may be released into the: reacltor containment shall be provided as necessary to reduce, consistent with the functionizqg of other associated systems, the concentration and quality of fission prod"ucts to the environment following postulated accidents, and to control the co~ce-ntration of hydrogen or oxygen and other substances in the containment

.... ::atmosphere following postulated accidents to assure that containment integrity is~~maintain~ed. "i.*: Each *-system shall have suitable, redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to~assure that for onsite electric power system operation (assuming offsite power is not available) and for .offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. By regulatory precedent, licensees are required to demonstrate this capability by the use of a bounding cialculation or other deterministic method. STPNOC requests an exemption from those deterministic requirements in order to enable the use of a risk-informed method to demonstrate acceptable sump design and Containment Cleanup performance with regard to the effects of LOCA debris.*Approval of this exemption-will allow use of a risk-informed method to account for the*probabilitieS uncertainties associated with mitigation of the effects of debris following....postulated:

L'OCAs. The method evaluates the effects on strainer blockage and core'*blockage resulting from debris 6oncerns raised by GS1-1 91. In order to confirm acceptable

• "sump deSigrn,'-the risk: associated with GS1-191 is evaluated to include the failure'""mechanisms aSsociated wffithloss of core cooling and strainer blockage.::*" The scope of th'e exemption applies for all debris effects addressed in the STP RoverD..i ,methodology.idescribed in Att;achment 1 that was used to respond to GL 2004-02, and which are associated with LOCA break sizes and locations that potentially generate fine* fiber debris that ?xceeds the quantity bounded by STP plant-specific testing described in.:AttachmentA 1,That Scope isgenerally described as breaks larger than approximately 12.8"*ID in Iocatio~ns~wjpere a sufficient amount of fiber debris can be generated and transported

• to the sump tp exceed the amount of fine fiber debris in the STP plant-specific testing* described in. Attachment

1. F.o~rty-five weld locations have currently been identified on the:pressurizersurge line and RCS main loop piping. To minimize the potential that a later--*analysis could the specific locations to change, the requested exemption is based-on the b reaks' ability to generate sufficient transportable debris, as described in RoverD.The key elements of the exemption request are:

NOC-AE-1 5003241 Attachment 2-5 Page 2 of 7 1. It applies only to the effects of debris as described in Attachment 1.2. It applies only for LOCA breaks that can generate and transport fiber debris that is not bounded by STP plant-specific testing.3. It applies to any LOCA break that can generate and transport fiber debris that is not bounded by STP plant-specific testing and is not limited to the 45 specific break locations noted in this application, provided that the ACDF and ALERF associated with the break size remains in Region Ill of RG 1.174.This exemption request is complemented by the accompanying License Amendment Request (LAR) (Attachment

3) seeking NRC approval of the changes to the South Texas Project Electric Generating Station (STPEGS) Updated Final Safety Analysis Report (UFSAR) and Technical Specifications, to amend the licensing basis based on acceptable design of the containment sump. The risk-informed method provides assurance, with high probability, for acceptable sump performance and debris mitigation during ECCS operation in recirculation mode as calculated by the ECCS evaluation model.2. Regulatory Requirements Involved STPNOC seeks exemption to the extent that GDC 41 requires deterministic calculations or other analyses to address the concerns raised by GS1-191 related to acceptable plant performance during the recirculation mode in containment following a LOCA. The proposed changes to the licensing basis and Technical Specifications, submitted for NRC approval with the LAR in Attachment 3, address GSI-191 and provide closure to GL 2004-02 for STP Units 1 and 2 on the basis that the associated risk is shown to meet the acceptance guidelines in Regulatory Guide (RG) 1.174 and that, in conjunction with the existing licensing basis, adequate safety is demonstrated.

This exemption request is for the purpose of allowing the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated loss of coolant accidents

'(LOCAs).

The effects of LOCA debris have been evaluated, using deterministic methods, to meet the current licensing basis assumptions for analyzing the effects of post-LOCA debris blockage in the sump and in-vessel; however, these evaluations have not been shown to fully address debris effects for the as-built, as-operated plant. The risk-informed approach evaluates the debris effects as part of the assessment of the residual risk associated with GS1-191 concerns.

Based on confirmation of acceptable ECCS design as determined by the resulting risk meeting the acceptance guidelines in RG 1.174, the licensing basis for CSS compliance with GDC 41 is amended.2.1 Evaluation of Impacts on I0CFR50,67 and GOC 19 The impact of the proposed exemption on the licensing basis analysis for demonstrating radiological consequences of the design basis LOCA meet the radiological dose guidelines specified in 10CFR50.67 and the dose limits specified in GDC 19 was evaluated.

The risk-informed method provides confirmation of reliable ECCS and CSS performance as required for the licensing basis analyses that demonstrate the requirements of 10CFR50.67 and GDC 19. The method demonstrates that sump performance continues to support reliable plant design and operation and does not entail any exemption from 10CFR50.67 or GDC 19.

NOC-AE-1 5003241 Attachment 2-5 Page 3 of 7 I10CFR50.67 Accident Source Term For STP Units I and 2, which have implemented the Alternative Source Term (AST), the design-basis LOCA radiological consequence LOCA analysis is a deterministic evaluation based on the assumption of a major rupture of the reactor coolant system piping and the assumption of the deterministic failure of the ECCS to provide adequate core cooling. This scenario results in a significant amount of core damage as specified in RG 1.183 and does not represent any specific accident sequence, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 source term characteristics.

Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the likely incidents evaluated for design-basis transient analyses.

Since deterministic failure of ECCS is assumed at the onset of the accident by the analysis, the reliability of the containment emergency sumps with respect to ECCS operation does not affect the analysis for dose consequences.

The regulation itself requires "reasonable assurance" and STPNOC believes the risk-informed method confirms reliable CSS operation as an input to the AST analysis and meets the"reasonable assurance" standard.

Therefore, for the purposes of this exemption request, the current licensing basis analyses for I0CFR50.67 are considered to be met.GDC 19 Control Room With respect to accident dose, GDC 19 states: "Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident." The accident source term governed by 10OCFR50.67 applies a 5-rem limit for control room dose, consistent with GDC 19. Therefore, STPNOC believes the discussion above with regard to I0CFR50.67 also applies to GDC 19. STPNOC also notes that the deterministic language that appears in GDC 35 and GDC 38 does not appear in GDC 19.STPNOC has determined that control room dose will be less than 5 rem if CSS operates for at least 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> post-LOCA 1.Based on the safety margins in the evaluation of the debris effects described in Attachment 1-4 and the conservatism in the dose calculation, STPNOC has confidence CSS will operate long enough to assure that the dose will be less than 5 rem.Based on the review above, STPNOC believes no exemption to GDC 19 is needed.2.2 Evaluation of Impacts on other Regulatory Requirements

-Conclusion The proposed exemption does not result in any physical changes to the facility or changes to the operation of the plant, and does not change any of the programmatic requirements.

Based on demonstrating acceptable containment emergency sump, ECCS and CSS design for amending the current licensing basis for compliance with GDC 41 as described above, compliance with other regulatory requirements that rely on acceptable design for these systems and components continue to be met in the current licensing basis.RAI response letter dated March Il, 2014, Att. I, p.4 (ML I4086A386)

NOC-AE-1 5003241 Attachment 2-5 Page 4 of 7 3. Basis for the Exemption Request Under 10CFR50.12, a licensee may request and the NRC may grant exemptions from the requirements of 10CFR50 which are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and when special circumstances are present.3.1 Justification for the Exemption Request As required by 10CFR5O.12(a)(2), the Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR50.12(a)(2) are applicable.

STPNOC has evaluated the proposed exemption against the conditions specified in 10CFR50.12(a) and determined that this proposed exemption meets the requirements for granting an exemption from the regulation, and that special circumstances are present.The information supporting the determination is provided below.Pursuant to 10CFR50.12, "Specific exemptions," the NRC may grant exemptions from the requirements of this part provided the following three conditions are met as required by 10OCFRS0.

12(a)(1): The exemption is authorized by law.The NRC has authority under the Atomic Energy Act of 1954, as amended, to grant exemptions from its regulations if doing so would not violate the requirements of law. This exemption is authorized by law as is provided by 10CFR50.12 which provides the NRC authority to grantt exemptions from 10CFR50 requirements with provision of proper justification.

Approval of the exemption would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, the Commission's regulations, or any other law The exemption does not present an undue risk to the public health and safety.The proposed change is to apply a risk-informed method rather than a traditional deterministic method in order to quantify the residual risk associated with GS1-191 and to establish a high confidence of acceptable containment sump design. The purpose of GDC 41 is to establish acceptable design for the containment atmospheric cleanup system, which includes the CSS, to provide a high probability that the systems will perform the required functions.

The proposed exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated initiating events is not increased.

The PRA and engineering analysis demonstrate that the calculated risk is small and consistent with the intent of the .Commission's Safety Goal Policy Statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.The probability of a large break LOCA is sufficiently low that the application of a risk-informed approach to evaluate the ability of the CSS to meet GDC 41 with high probability and with low uncertainty, rather than using a calculational model using deterministic methods to achieve similar understanding, would have little effect on public risk. This is applicable to evaluating acceptable containment sump design in support of ECCS and CSS recirculation modes.

NOC-AE-1 5003241 Attachment 2-5 Page 5 of 7 The risk-informed approach involves a complete evaluation of the spectrum of LOCA breaks, including double-ended guillotine break, up to and including the largest pipe in the reactor coolant system. The risk-informed approach analyzes LOCAs, regardless of break size, using the same methods, assumptions, and criteria in order to quantify the uncertainties and overall risk metrics. This ensures that large break LOCAs with relatively small contribution to CDF due to the low probability of such a break as well as smaller break LOCAs with higher probabilities of occurrence are considered in the results. Since the design basis requirement for consideration of a double-ended guillotine break of the largest pipe in the reactor coolant system is retained and since no physical changes to the facility or changes to the operation of the facility are being made, the existing defense-in-depth and safety margin established for the design of the facility is not reduced.The exemption is consistent with the common defense and security.The exemption involves a change to the licensing basis for the plant that has no relation to the possession of licensed material or any security requirements that apply to STP Units I and 2. Therefore the exemption is consistent with the common defense and security.3.2 Special Circumstances This section discusses the presence of special circumstances as related to 10CFR50.12(a).

10CFR50.12(a)(2) states that NRC will not consider granting an exemption to the regulations unless special circumstances are present. Special circumstances are present whenever one of the listed items (i) through (vi) under 10CFR5O.12(a)(2) are applicable.

Such special circumstances are present in this instance to warrant exemption from the implicit requirement in GDC 41 to use a deterministic method to evaluate for acceptable containment emergency sump design. Approval of the exemption request would allow use of a risk-informed method to amend the design basis for acceptable containment sump design in support of CSS design for compliance with GDC 41.Specifically, I10CFR50.12(a)(2)(ii) applies: Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.The intent of GDC 41 is to ensure systems in the plant design to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

This exemption request is consistent with that purpose in that use of the proposed risk-informed approach demonstrates a high probability of successful containment emergency sump and OSS performance, which includes realistically available recirculation flow, based on the risk S NOc-AE-1 5003241 Attachment 2-5 Page 6 of 7 results meeting the acceptance guidelines of RG 1.174. The risk-informed approach assesses the design for a full spectrum of breaks, and assesses equipment failures that include loss of offsite power and worst case single failure, consistent with the GDC 41 requirements.

Since the proposed exemption does not involve any physical changes to the plant, there is no effect on the GDC 41 design requirements for redundancy in components and features, interconnections, leak detection, isolation, and containment capabilities.

The current licensing basis evaluations for CSS compliance with GDC 41 for these aspects continue to be met.As discussed in the Commission's Policy Statement on "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities", NRC regulations and their implementation are generally based on deterministic approaches that consider a set of challenges to safety and determine how those challenges should be mitigated.

This request does not seek exemption from any explicit language in, the regulatory requirements.

Rather, the need is based on the implicit requirements in the regulations for using deterministic methods to demonstrate acceptable design. Regulatory requirements are largely based on a deterministic framework, and are established for design basis accidents, such as the LOCA, with specific acceptance criteria that must be satisfied.

Licensed facilities must be provided with safety systems capable'of preventing and mitigating the consequences of design basis accidents to protect public health and safety. The deterministic regulatory requirements were designed to ensure that these systems are highly reliable.

The LOCA analysis and the General Design Criteria (GDC) were established as part of this deterministic regulatory framework.

In comparison, the risk-informed approach considers nuclear safety in a more comprehensive way by examining the likelihood of a broad spectrum of initiating events and potential challenges, considering a wide range of credible events and assessing the risk based on mitigating system reliability.

An objective of GDC 41 is to maintain low risk to the public health and safety through a reliable CSS, as supported by the containment sump. The supporting analysis demonstrates that a risk-informed approach to sump performance is consistent with the Commission's Safety Goals for nuclear power plants, and supports CSS operation with a high degree of reliability.

Consequently, the special circumstances described in I10CFR50.1I2(a)(2)(ii) apply.Specifically, 1 0CFR50.1 2(a)(2)(iii) applies: Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

NOC-AE-1 5003241 Attachment 2-5 Page 7 of 7 The specific hardship is the excessive 158 -176 rem occupational radiological dose that are estimated to be incurred for plant modifications to remove insulation.

This was described in Attachment 2-1, above.In conclusion, special circumstances in 10CFR50.12(a)(2)(ii) and 10CFR50.12(a)(2)(iii) are present as required by 10CFR50.12(a)(2) for consideration of the request for exemption.

4. Technical Justification for the Exemption Technical justification for the risk-informed method is provided in Attachment 1 and in the LAR (Attachment 3).The proposed risk-informed approach meets the~ key principles in RG 1.174 in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in small increase in risk, and is monitored using performance measurement strategies.

This proposed exemption to allow use of the risk-informed method is consistent with the key principle in RG 1.174 that requires the proposed change to meet current regulations unless explicated related to a requested exemption.

The results show that the risk associated with GS1-1 91 concerns is less than the threshold in Region Ill, "Very Small Changes," of RG 1.174, and therefore are consistent with the Commission's Safety Goals for public health and safety.5. Conclusion Approval of an exemption to allow the use of the risk-informed approach is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security as required by I0CFR50.12(a)(1).

Furthermore, special circumstances required by 10OCFR50.1I2(a)(2) are present for item 10OCFR5O.

12(a)(2)(ii) in that application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.Based on the determination that the risk of the exemption meets the acceptance guidelines of RG 1.174, the results demonstrate there is reasonable assurance that the CSS will function in the recirculation mode and that the public health and safety will be protected.

6. Implementation STPNOC requests that this exemption request be approved for implementation by N~ovember 2015.

NOC-.AE-1 5003241 Attachment 3 License Amendment Request for STP Piloted Risk-Informed Approach to Closure for GS1-191

: NOC-AE-1 5003241 Attachment 3 Page 1 of 22 License Amendment Request for STP Piloted Risk-Informed Approach to Closure for GSI-191 1. Summary Description

-Methodology Change and Technical Specification Change In accordance with 10CFRS0.59(c)(2)(viii), STP Nuclear Operating Company requests an amendment to Operating Licenses NPF-79 and NPF-80 for South Texas Project Units 1 and 2 pursuant to 10CFR50.90.

The proposed amendment will revise the licensing basis as described in the South Texas Project Electric Generating Station Updated Final Safety Analysis Report to allow the use of a risk-informed approach to address safety issues discussed in Generic Safety Issue -191, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance".

The risk-informed approach is consistent with the guidance of NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis".In addition to the overall methodology change to allow a risk-informed approach, the thermal-hydraulic analysis described in Sections 5Sand 6 of Attachment 1-3 is included as part of the proposed methodology change as a method of core cooling evaluation not described in the STP current licensing basis (CLB). The T-H analysis is part of the screening process for the risk-informed approach and shows that, even assuming full blockage of all flow into the core during HLB, there will be adequate cooling regardless of break size.In addition, STPNOC proposes to amend the STP Unit 1 and Unit 2 Operating Licenses to revise the Technical Specifications (TS) for the Emergency Core Cooling System (ECCS) and the Containment Spray System (CSS). The changes proposed for these TS would add a required action and completion time specific to the effects of debris to TS 3/4.5.2, "ECCS Subsystems

-Tavg Greater Than or Equal to 35QO and TS 3/4.6.2, "Depressurization and Cooling Systems -Containment Spray System". The proposed TS changes will align the TS with the risk-informed methodology change.The proposed changes will apply only for the effects of debris as described in GS1-191 and GL 2004-02.The proposed change associated with the change in methodology is to use a risk-informed approach to determine the design requirements to address the effects of LOCA debris instead of a traditional deterministic approach.

The details of the approach are provided in Attachment

1. The debris analysis covers a full spectrum of postulated LOCAs, including double-ended guillotine breaks, for all pipe sizes up to and including the design basis accident LOCA, to provide assurance that the most severe postulated loss-of-coolant accidents are evaluated.

The deterministic current licensing basis (CLB) will continue to apply to LOCA break sizes that generate fine fiber debris that is bounded by STP plant-specific testing. The proposed methodology change will apply for LOCAs that can generate and transport fine fiber debris that is not bounded by the plant-specific testing. STP conservatively relegates to failure the LOCA break sizes that can generate and transport fine fiber debris that is not bounded by the STP plant-specific testing. STP applies NUREG 1829 to determine the break frequency for the smallest of those breaks to obtain the highest frequency, and uses that frequency as the ACDF for comparison to the criteria in RG 1.174. The results of the evaluation show that the risk from the proposed change is "very small" in that it is in Region Ill of RG 1.174. The NOC-AE-1 5003241 Attachment 3 Page 2 of 22 methodology includes~conservatisms in the plant-specific testing and in the assumption that all the unbounded breaks are relegated to failure.The proposed TS change associated with the change in methodology will create actions and completion times in the ECCS and CSS TS. ECCS and CSS are the only TS systems that depend on the containment emergency sumps as a support system and are therefore the only systems that are directly subject to the effects of debris. The purpose of the proposed changes is to establish actions that are associated with conditions that can potentially affect the effects of debris and providing a required action time that is commensurate with the very low risk associated with debris effects. The proposed action is based on the amount of debris tested in the STP plant-specific testing so that the determination of operability is performed without needing a risk assessment, which makes the process consistent with NRC guidance on operability determinations.

The evaluation of the proposed change to the TS is performed in accordance with applicable guidance from RG 1.177 and compares ACDF and ALERF for extending required completion time to restore ECCS and CSS made inoperable from debris effects. The change in completion time is assumed to be 90 days.The proposed change to the licensing basis implements a risk-informed rather than a deterministic method to demonstrate compliance.

In conjunction with the proposed LAR, STPNOC is also requesting exemptions from 10CFR50.46(d), GDC 35, GDC 38 and GDC 41, as provided in Attachments 2-1 through 2-4 to this letter. The UFSAR markups are attached for the staff's information.

2.1 Detailed

Description

-Methodology Change Upon approval of the licensing basis changes, STPNOC will make the following changes to the STP UFSAR:*Add Appendix 6A, "Risk-Informed Approach to Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents".

This appendix will describe the evaluations performed using a risk-informed approach to address GS1-191 concerns including the effects on long-term cooling due to debris accumulation on containment sump strainers for ECCS and CSS in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss of coolant accidents (LOCAs).* Make conforming changes to UFSAR Chapter 3 descriptions of evaluations against GDC 35, GDC 38 and GDC 41* Make conforming changes to UFSAR Chapter 6* Make conforming changes to UFSAR Chapter 15.Modifications Previously Implemented to Address GS1-191 The current licensing basis for the sumps is based on a deterministic methodology that was used to analyze susceptibility of the ECCS and CSS recirculation functions for adverse effects of post-accident debris. The methodology was largely in accordance with Nuclear Energy Institute report NEI 04-07 and the associated NRC Safety Evaluation Report (SER). In addition, evaluations were performed in accordance with WCAP-16406-P-A, "Evaluation of Downstream Sump Debris Effects in Support of GSI- 191 ," Revision 1 and WCAP-1 6793-NP," Evaluation of Long-Term Cooling Considering Particulate and Chemical Debris in the Recirculating Fluid" to consider the effects that debris carried downstream of the containment NOC-.AE-1 5003241.Attachment 3 Page 3 of 22 sump screen and into the reactor vessel has on core cooling, including fuel and vessel blockage.

STPNOC also evaluated the type and expected quantity of chemical products that would be expected to form in the recirculation fluid specifically for STP using the methodology developed in WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-l191".

To support the sump performance evaluation, STPNOC performed containment walkdowns using the guidance of NEI 02-01.The UFSAR changes attached to this application include a description of ECCS sump performance evaluations performed to address GSI-191 and respond to GL 2004-02 prior to this risk-informed licensing application.

Those evaluations account for previously implemented hardware modifications and plant procedures and processes to provide high confidence that the sump design supports long-term core cooling following a design basis loss of coolant accident.

Those design modifications and procedure changes were implemented in accordance with 10OCFR50.59.

STP is not requesting approval for those prior changes as part of this license amendment request. The risk-informed evaluation used in the application appropriately accounts for those earlier design modification and procedure changes.Installation of New Sump Strainer Assemblies To address debris related concerns associated with GS1-191 and in response to the debris issues identified in GL 2004-02, new ECCS containment sump strainer assemblies were installed in STP Unit 1 in October 2006 and in STP Unit 2 in April 2007. The surface area of the strainers was increased from approximately 155 square feet per sump to approximately 1818 square feet per sump. The screen-hole size of the strainers was reduced from 0.25 inches diameter to 0.095 inches diameter.

Small particles in water entering the suction pipe from the sump cannot clog the containment spray nozzles (3/8-inch orifice diameter).

Installation of the new strainers did not affect the independence and redundancy of the sumps.The sump strainer design implemented by these modifications meets the current design basis requirements with respect to net positive suction head and ECCS performance.

The sumps are designed according to RG 1.82 proposed Revision 1, dated May 1983, which recommends a calculation of sump screen head loss due to debris blockage.

Utilizing the current licensing basis methodology, the pump NPSH is sufficient to accommodate this head loss. The STP sumps meet the function to preclude passage of debris particles large enough to damage downstream components in the ECOS and CSS. The sump strainer design has been evaluated to meet the current design basis assumptions for analyzing the effects of post-accident debris blockage and for compliance with 10CFR50.46 for long term cooling, General Design Criterion 35 for emergency core cooling, GDC 36 for inspection of ECCS, GDC 38 for containment heat removal, GDC 39 for inspection of containment heat removal system, and GDC 41 for containment atmosphere cleanup.The new strainer design is a safety improvement that contributed to meeting the RG 1.174 criteria for Region Ill, "Very Small Changes," for the results from the risk-informed methodology.

Following installation of the new sump strainers, protective gratings were installed in front of the strainers to preclude inadvertent damage to these components.

The framing structure for the protective gratings consists of vertical grating panels attached to metal columns that are NOC-AE-15003241 Attachment 3 Page 4 of 22 welded to base plates that are anchored into the concrete floor, and is qualified for Seismic Il/I loading to ensure the maximum stresses are below the allowable limits. The structure is made of carbon steel and has a qualified coating applied.Replacement of Fiber Insulation Per the original plant design, the reactor vessel nozzles were insulated with a non-crush insulation material composed of calcium silicate (brand name Marinite).

Marinite insulation was identified as a significant contributor to the debris loading associated with one of the worst case LOCA scenarios for strainer head loss based on previous evaluations.

As a result, all of the Marinite insulation was replaced with NUKON fiberglass insulation during the Fall 2009 refueling outage for STP Unit 1 and the Spring 2010 refueling outage for Unit 2. The presence of Marinite in the plant-specific testing conducted in July 2008 is a conservative element of the testing.Plant-Specific Testingq STP conducted successful plant-specific testing in July 2008 using generally prototypical debris, conservative chemical effects, prototypical simulation of strainer approach flow conditions, and a STP strainer module. This plant-specific test is described in more detail in Attachment 1 and forms the basis for the deterministic scope of the proposed methodology change.Use of a Risk-Informed Approach to Address GS1-191 The risk associated with GS1-1 91 issues has been quantified as described in Attachment 1 and is "very small" as defined by Region Ill in RG 1.174. The proposed UFSAR Appendix 6A describes the risk-informed approach used to confirm that the ECCS and CSS will operate with a high probability following a LOCA when considering the impacts and effects described by GS1-1 91. Therefore no further physical modifications to STP Units 1 and 2 are proposed as part of this license amendment request to implement the risk-informed approach.Attachment 4 provides the Licensee Commitment to implement the proposed amendment following approval and to revise affected sections of the UFSAR identified in this Attachment.

Upon approval of the proposed amendment, applicable UFSAR safety system and design bases descriptions that take credit for the evaluation described in Appendix 6A will be revised.In addition, conforming changes to the Technical Specification Bases are provided in Attachment 3 to this Attachment for information only, to be implemented following NRC approval of the LAR.The performance evaluations for accidents requiring ECOS operation described in Chapters 6 and 15, based on the South Texas Project Units I and 2 Appendix K Large-Break Loss-of-Coolant Accident analysis, demonstrate that for breaks up to and including the double-ended guillotine break of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in 10CFR50.46, and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

NOC-AE-1 5003241 Attachment 3 Page 5 of 22 System redundancy, independence, and diversity features are not changed for those safety systems credited in the accident analyses.

No new programmatic compensatory activities or reliance on manual operator actions are required to implement this change.2.2 Technical Evaluation

-Methodology Change 2.2.1 Current System Descriptions The methodology change affects the analysis of systems and functions that are susceptible to the effects of LOCA debris. The affected systems are those that are supported by' the emergency strainers and sumps during the recirculation phase of LOCA mitigation, which are ECCS (LHSI and HHSI) and 0SS. The associated functions and associated regulations are:* ECCS: 10CFR5O.46(d), GDC 35* Containment Heat Removal: GDC 38* Containment Atmosphere Cleanup: GDC 41 Emeraqency Core Coolinq System The Emergency Core Cooling System is designed to cool the reactor core and provide shutdown capability subsequent to the following accident conditions:

1. Pipe breaks in the Reactor Coolant System which cause a discharge larger than that which can be made up by the normal makeup system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS.2. Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident.3. Pipe breaks in the steam system, up to and including the instantaneous circumferential rupture of the largest pipe in the steam system.4. A steam generator tube rupture.The primary function of the ECCS is to remove the stored and fission product decay heat from the reactor core and to provide shutdown capability during accident conditions.

The ECCS provides shutdown capability for the accidents above by means of boron injection.

It is designed to tolerate a single active failure in the short term or a single active or passive failure in the long term. The system meets its minimum required performance level with onsite or offsite electrical power.The ECCS consists of the high head safety injection and low head safety injection pumps, Safety Injection System accumulators, residual heat removal heat exchangers, the refueling water storage tank along with the associated piping, valves, instrumentation, and other related equipment.

-NOC-AE-1 5003241 Attachment 3 Page 6 of 22 The design bases for selecting the functional requirements of the ECCS, such as peak fuel cladding temperature, etc., are derived from Appendix K limits as delineated in 10CFR50.46.

The subsystem functional parameters are integrated so that the Appendix K requirements are met over the range of anticipated accidents and single failure assumptions.

Reliability of the ECCS has been considered in Selection of the functional requirements, selections of the particular components, and location of components and connected piping.Redundant components are provided where the loss of one component would impair reliability.

Valves are provided in series where isolation is desired. Redundant sources of the ECCS actuation signal are available so that the proper and timely operation of the ECCS is not inhibited.

Sufficient instrumentation is available so that a failure of an instrument does not impair readiness of the system. The active components of the ECCS are powered from separate buses which are energized from offsite power supplies.In addition, the standby diesel generators assure that adequate redundant sources of auxiliary onsite power are available to meet all ECCS power requirements.

Each diesel is capable of driving all pumps, valves, and necessary instruments associated with one train of the ECCS.The elevated temperature of the sump solution during recirculation is well within the design temperature of all ECCS components.

In addition, consideration has been given to the potential for corrosion of various types of metals exposed to the fluid conditions prevalent immediately after the accident and during long-term recirculation operations.

I0CFR50.46(b) provides the following criteria to judge the adequacy of the ECCS.1. Peak clad temperature calculated shall not exceed 2,2000°F.2. The calculated total oxidation of the clad shall nowhere exceed 0.17 times the total clad thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the clad with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the clad cylinders surrounding the fuel, excluding the clad around the plenum volume, were to react.4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.5. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by long-lived radioactivity remaining in the core.10CFR50.46(d) provides that the criteria set forth in paragraph (b), with cooling performance calculated in accordance with an acceptable evaluation model, are in implementation of the general requirements with respect to ECCS cooling performance design set forth in this part, including in particular Criterion 35 of appendix A. STPNOC is not proposing to change the basis described in 10CFR50.46(b), but is proposing to revise the methodology for showing NOC-AE-1 5003241 Attachment 3 Page 7 of 22 those requirements are met. The methodology is governed by 10CFR50.46(d) by its incorporation of GDC 35. The LAR supports the exemptions to 10CFR50.46(d) and GDC 35 described in Attachments 2-1, 2-2, and 2-3.Containment Heat Removal System The CHRS is designed to meet the requirements of GDC 38. The CHRS consists of the Reactor Containment Fan Cooler Subsystem, which is a part of the Reactor Containment HVAC System, the RHR heat exchangers, and the Containment Spray System. The ECCS assists the CHRS by transferring heat from the reactor core to the Containment sump. The Residual Heat Removal heat exchangers, in conjunction with the ECCS low-head Safety Injection pumps, are used to transfer heat from the Containment sumps to the Component Cooling Water System. The RCFCs are also cooled by the CCWS following an SI signal. The CCWS rejects this heat to the ultimate heat sink via the Essential Cooling Water System.The CHRS meets the following design bases: 1. The CHRS is capable of removing sufficient energy to limit the peak Containment pressure and to limit the Containment pressure to a low value at the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a DBA.2. In order to ensure the satisfactory operation of the systems after a DBA, each active component is testable during reactor power operation.

3. The system is divided into three trains, with each train receiving power from a separate Emergency Safety Features power supply.As discussed in Attachment 2-1 and 2-4 GDC 38 requires use of deterministic methodology.

The LAR complements the proposed exemption by proposing an amendment to the license to allow a risk-informed methodology.

Containment Atmosphere Cleanup System Containment Atmosphere Cleanup System is designed to meet the requirements of GDC 41.The CSS is provided to reduce the concentration and quantity of fission products in the Containment atmosphere following a LOCA. Per 10CFR50.44, hydrogen recombiners are no longer required for design basis accidents.

The equilibrium sump pH is maintained by trisodium phosphate contained in baskets on the containment floor. The initial CSS water and spilled RCS water dissolves the TSP into the containment sump allowing recirculation of the alkaline fluid. Each unit is equipped with three 50-percent spray trains taking suction from the Containment sump. Each Containment spray train is supplied power from a separate bus. Each bus is connected to both the Offsite and the Standby Power Supply Systems. This assures that for Onsite or for Offsite Electrical Power System failure, their safety function can be accomplished, assuming a single failure.The 055 is:

_NOC-AE-1 5003241 Attachment 3 Page 8 of 22 1. Designed such that it will tolerate a single active failure.2. Designed to accommodate the operating basis earthquake within stress limits of applicable codes and to withstand the safe shutdown earthquake without loss of function.3. Designed to assist in reducing offsite exposures resulting from a design basis accident (DBA) to less than the limits of 10CFR50.67 by rapidly reducing the airborne elemental iodine and particulate concentrations in the Containment following a DBA.As discussed in Attachment 2-1 and 2-5 GDC 41 requires use of deterministic methodology.

The LAR complements the proposed exemption by proposing an amendment to the license to allow a risk-informed methodology.

Containment Emeraqency Sump The Containment emergency sump meets the following design bases: 1. Sufficient capacity and redundancy to satisfy the single-failure criteria.

To achieve this, each CSS/ECCS train draws water from a separate Containment emergency sump.2. Capable of satisfying the flow and net positive suction head requirements of the ECCS and the CSS under the most adverse combination of credible occurrences.

This includes minimizing the possibility of vortexing in the sump.3, Minimizes entry of high-density particles (specific gravity of 1.05 or more) or floating debris into the sump and recirculating lines.4, Sumps are designed in accordance with RG 1.82, proposed revision 1, May 1983.Three independent sumps serve as reservoirs to the ECCS and CSS pumps during the recirculation phase post-DBA.

Each sump is stainless steel lined, contains a Vortex Suppressor, and is provided with four stainless steel strainer assemblies.

The strainer assemblies for each sump consist of two 5-module assemblies, one 4-module assembly, and one 6-module assembly with each module made up of eleven strainer discs. The strainer screen consists of perforated plate with nominal 0.095 inch diameter openings.

Flow leaving the strainer enters a four inlet plenum box (one inlet for each strainer assembly).

The plenum box collects the flow from the strainer assemblies and directs the flow vertically downward directly into the sump pit. An access cover is provided on the plenum box for internal inspections of the sump structures, vortex suppressor, and the strainer assemblies.

The sumps are located at the -11 feet-3 inch level of Reactor Containment Building (RCB).The sumps are physically separated from each other. The floor around the emergency sumps slopes away from them and toward normal sumps located in the area. The drains from the upper levels of the RCB do not terminate in the immediate area of the sumps.The sump structures are designed to withstand the SSE without loss of structural integrity.

NOC-AE-15003241 Attachment 3 Page 9 of 22 Water entering the suction pipe from the sump may contain a small amount of particulate and fiber debris (less than 0.095-inch in diameter).

This debris cannot clog the spray nozzles (3/8-in. orifice diameter) which are the limiting restrictions in the 0S5 system served by the sump.At the beginning of the recirculation phase, the minimum water level above the RCB floor is adequate to provide the required NPSH for the ECCS and CSS pumps.The sumps are designed to NRC RG 1.82, proposed revision 1, May 1983. The sump structures are designed to limit approach flow velocities to less than 0.009 feet/second permitting high-density particles to settle out on the floor and minimize the possibility of clogging the strainers.

The sump structures are designed to withstand the maximum expected differential pressure imposed by the accumulation of debris.Most potential sources of debris are remote from the emergency sumps and are separated by shield walls or other partitions.

Expected debris constituents are pieces of insulation and paint.The possibility of paint chips peeling off has been reduced by requiring proper surface preparation and by painting large surface components (such as: the Containment liner, RCS supports, floors, and structural steel) with coatings which have been qualified under DBA conditions.

The stainless steel reflective insulation is used exclusively on the Reactor Vessel including the head. The blanket fiberglass type is used on the hot piping, valves, and other equipment including the Steam Generators, the Pressurizer, and the Reactor Coolant Pumps. Cellular glass insulation is used on cold piping for antisweat purposes.

Microtherm is also used for piping in the wall penetrations.

Containment emergency sumps are inspected periodically as delineated in the Technical Specifications.

2.3 Background

GS1-191 concerns the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized-water reactors and result in loss of net positive suction head for the ECCS and CSS pumps, impeding the flow of water from the sump. GL 2004-02 requested licensees to address GS1-191 issues, focused on demonstrating compliance with the ECCS acceptance criteria in I0CFR50.46.

GL 2004-02 requested licensees to perform new, more realistic analyses using an NRC-approved methodology and to confirm the functionality of the ECCS and CSS during design basis accidents that require containment sump recirculation.

K The current design for the containment sumps and strainers was assessed in response to NRC Generic Letter 2004-02. STPNOC provided specific information regarding the deterministic methodology for demonstrating compliance by applying industry and NRC guidance.

However, since this methodology has not been demonstrated to fully address GSI-191 without the need for additional changes to the plant design, such as extensive modifications to insulation in the containment, a risk-informed approach is applied to evaluate effects of LOCA debris using the guidance in RG 1.174.

NOC-AE-1 5003241 Attachment 3 Page 10 of 22 Larger containment sump strainers have been installed that greatly reduce the potential for loss of net positive suction head. Problematic insulation was removed; reactor coolant system pipe welds where a break could generate large amounts of insulation debris and which were subject to PWSCC were mitigated.

The STP piloted risk-informed approach maintains the defense-in-depth measures as described in Attachment 1-4. These measures include those identified in response to NRC Bulletin 2003-01 and GL 2004-02 to address the potential for sump strainer clogging and other concerns associated with GSI-191.The Commission issued Staff Requirements Memorandum (SRM)-SECY-10-01 13, "Closure Options for Generic Safety Issue (GSI) -191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance," directing the staff to consider alternative options for resolving GS1-191 that are innovative and creative, as well as risk-informed and safety conscious.

Subsequently, STPNOC, through interactions with the staff, developed a risk-informed approach to address GS1-191 based on the guidance in RG 1.174. STPNOC submitted to the NRC the preliminary results showing that the risks, Core Damage Frequency and Large Early Release Frequency, associated with GS1-191 concerns are less than the threshold for Region Ill, "Very Small Changes," of RG 1.174, and notified the NRC of the intent to seek exemption from certain requirements of 10CFR50.46 (ML1 1354A386).

SECY-12-0093, "Closure Options for Generic Safety Issue -191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance," described the staff plans to use STPNOC as a pilot for other licensees choosing to use this approach.

The exemption request from certain requirements of I0CFR50.46 including impacted General Design Criteria of Appendix A of 10CFR50 is provided in Attachment 2 to this letter.2.4 Engineering Analysis Evaluation Overview The design and licensing basis descriptions of accidents requiring ECCS and CSS operation, including analysis methods, assumptions, and results provided in UFSAR Chapters 6 and 15 remain unchanged.

This is based on the functionality of the ECCS and CSS during design basis accidents being confirmed by demonstrating that safety margin and defense-in-depth are maintained with high probability (Region Ill RG 1.174).The performance evaluations for accidents requiring ECCS operation described in Chapters 6 and 15, based on the South Texas Project Units I and 2 Appendix K Large-Break Loss-of-Coolant Accident analysis, demonstrate that for breaks up to and including the double-ended guillotine break of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in 10CFR5O.46, and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

The methodology for calculating the risk associated with GS1-191 concerns evaluates a full spectrum of breaks up to and including double-ended guillotine breaks, for all RCS pipe sizes.The results show the risk associated with GS1-191 concerns for STP Units I and 2 is "very small" as defined by Region Ill in RG 1.174. Based on evaluation of defense-in-depth, safety margin, and a high probability of successful performance, the functionality of the ECCS and CSS is confirmed.

The detailed technical evaluation is presented in Attachment

1.

NOC-AE-1 5003241 Attachment 3 Page 11 of 22 The LAR is requested for the scope of breaks that can generate fiber debris on the ECCS sump strainer that exceeds the amount of fiber debris bounded by the plant-specific testing.In Attachment 1-2 and 1-3, STPNOO determined that only large breaks were in this scope and listed 45 examples.

STPNOC is requesting a LAR for this scope of breaks to allow evaluation of the debris effects using a risk-informed methodology because there is no practical deterministic methodology currently available.

Section 5 of Attachment 1-3, RoverD includes a thermal-hydraulic analysis that confirms that there is adequate core cooling for the entire spectrum of breaks. The thermal-hydraulics analysis is part of a risk-informed evaluation for downstream effects (see Section 2 of Attachment 1-3) and applies an 800°F success criteria for POT. The detailed technical description of the risk-informed screening process is presented in Section 5 of Attachment 1-3.The risk-informed thermal-hydraulics screening analysis does not replace the ECOS evaluation methodology in the STP UFSAR Oh. 15.6, which applies only through the LOCA reflood phase (approximately 5 minutes) and is not used for the assessment of long-term' cooling (after about 1/ hour and beyond) required by the risk-informed assessment of debris effects.2.5 Technical Specification Changes STPNOC is proposing to change the TS for ECCS and CSS to specifically address the potential effects of debris. The ECCS and 0SS are the only systems potentially affected by debris effects because they are the only systems that are supported by the containment emergency sumps and strainers.

2.5.1 TS 3/4.5.2, "ECCS Subsystems

-Tavg Greater Than or Equal to 350°FSystem Description

-see Section 2.2.1, above The change to the EGOS TS is proposed only for TS 3.5.2, which applies in MODE 1, 2, and 3.The risk-informed debris evaluation is performed with the plant assumed to be at-power when the initiating event LOCA occurs. Consequently, the debris generation and transport is initiated from normal operating temperature and pressure conditions.

Those conditions maximize the break zones of influence and initial decay heat. STPNOC considered application of a debris-specific TS requirement to TS 3.5.3, which applies in MODE 4 and determined that application of the CLB for debris is adequate for the following reasons:* STPNOC believes the deterministic testing is adequate for any MODE 4 LOCA because MODE 4 plant conditions do not support generation or transport of comparable quantities of debris.o The pressure and temperature (RCS pressure < COMS setpoint and temperature

< 350°F) do not produce ZOIs comparable to MODE 3 and higher o Cooling requirements are reduced because decay heat is much lower in MODE 4* There is no STP risk model established for MODE 4 on which to base a risk-informed completion time; however, the pipe break and associated debris risk is qualitatively NOC-AE-1 5003241 Attachment 3 Page 12 of 22 assessed to be lower than for MODE 3 and above o Pipe break frequencies would be significantly lower based on lower operating stresses and much shorter lengths of time in MODE 4 compared to MODE 3 and higher.Debris Related Required Action STPNOC proposes the required action below.c. With less than the required flow paths OPERABLE solely due to potential effects of LOCA generated and transported debris that exceeds analyzed amounts, perform the following:

1. Immediately initiate action to implement compensatory actions, AND 2. Within 90 days restore the affected subsystems to OPERABLE status,* OR Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> The proposed required action applies to the LCO 3.5.2.d requirement to have an operable flow path through the containment sump. It applies only for the potential effects of LOCA generated and transported debris that exceeds the amount that has been analyzed.

The operability with respect to debris is based on a quantity of debris; therefore, emergent nonconforming or degraded conditions can be evaluated in a deterministic process based on analyzed debris conditions.

No quantitative risk assessment is necessary, so the evaluation process is consistent with the guidance in Part 9900 of the NRC Inspection Manual that does not allow the use of risk in an operability determination.

The required action to implement compensatory action is based on the very low contribution by LOCA generated debris to the risk of core damage, and is a reasonable response to minimize the potential increase in risk from the debris source. Typical compensatory action would include actions such as:* Remove the debris or source of debris or take action that would prevent transport of the debris to the emergency sump* Defer maintenance that would affect availability of the affected systems and strainers* Increase frequency of RCS leak detection monitoring

• Brief operators on LOCA debris management actions The requirement to be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> is consistent with the other required actions in the ECCS TS.

NOC-AE-1 5003241 Attachment 3 Page 13 of 22 Existing action c will be administratively redesignated as action d, and the ECCS TS pages will be renumbered to accommodate the new action statement.

Debris Related Required Completion Time The action applies only for the potential effects of debris. Under the current TS, STPNOC would likely apply the Risk Managed Technical Specification (RMTS) PRA Functional allowance in NEL 06-09 Section 2.3.1 since the ECCS would still be functional for small and medium break LOCAs. Based on the risk evaluation in RoverD, STPNOC expects the 30-day "backstop" RMTS completion time would apply. Under CTS, STPNOC would monitor and manage risk based on plant configuration to until operability was restored or the 1 E-05 ICDP limit was reached.The proposed required completion time of 90 days is based on there being very small incremental increase in risk by increasing the completion time from 30 days to 90 days.Unlike RMTS, it is a set time and not subject to the risk management requirements of RMTS. Also, because the required completion time is longer than the 30-day RMTS backstop time, RMTS does not apply for the proposed action. The advantages of the 90-day completion time are: 1. It provides clarity for the operators with regard to application of the TS for degraded or nonconforming conditions associated with the potential effects of LOCA debris.2. It eliminates the need to apply the "PRA Functional" feature of RMTS, which simplifies application of the TS.3. It improves the usefulness of the STP pilot application as a precedent for other licensees.

4. A 90-day required completion time has precedent for Control Room Envelope (TSTF-448, Federal Register Notice, Volume 72, Pages 2022-2033, Technical Specification Improvement To Modify Requirements Regarding Control Room Envelope Habitability Using the Consolidated Line Item Improvement Process, dated January 17, 2007)2.5.2 TS 3/4.6.2.1, "Depressurization and Cooling Systems -Containment Spray System" System Description

-see Section 2.2.1, above CSS applies in MODE 1, 2, 3, and 4. New Action c is added to address effects of debris in MODE 1, 2, and 3. As for the ECCS TS, the deterministic testing is considered to bound the LOCA debris for any LOCA in MODE 4. STPNOC is incorporating wording in the CSS TS Bases to clarify the difference between the debris related requirements for MODE 1, 2, and 3 and MODE 4.

NOC-AE-1 5003241 Attachment 3 Page 14 of 22 Debris Related Required Action The proposed required action for CSS is essentially the same as that for ECCS: c. With one or more Containment Spray Systems inoperable in MODE 1, 2, or 3 solely due to potential effects of LOCA generated and transported debris that exceeds analyzed amounts, perform the following:

1. Immediately initiate action to implement compensatory actions, AND 2. Within 90 days restore the affected system(s) to OPERABLE status, OR Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Debris Related Required Completion Time As with ECCS, STPNOC is proposing a 90-day required completion time. Also, similar to ECCS, RMTS applies to CSS, so the same elements of the ECCS discussion above apply for CSS.The CSS TS pages will be renumbered to accommodate the new action statement.

2.5.3 Technical

Specification Change Risk Assessment The primary purpose of the proposed changes to the TS is to provide the operator with guidance to deal with debris-specific conditions and to provide a required completion time that is commensurate with the very small risk from the effects of debris. The RoverD description in Attachment 1 provides the overall quantification of ACDF and ALERF for the methodology change. These metrics represent the change in risk associated with assuming failure for the scope of LOCAs that generate amounts of debris that is not bounded by plant-specific testing.The risk associated with the proposed change to the TS is the risk associated with allowing a longer completion time to restore nonconforming or degraded conditions caused by potential effects of debris.

NOC-AE-1 5003241 Attachment 3 Page 15 of 22 STPNOC has evaluated the proposed change in accordance with the guidance of RG 1.177 with respect to Tier 1, Tier 2 and Tier 3. RG 1.177 references RG 1.174 numeric guidelines regarding changes in risk.Tier 1, PRA Capability and Insights STPNOC evaluated the ODE, LERF, ICCDP, and ICLERP for the proposed change. The calculated increase in ODE is in the range of 10.i per reactor year to 10-6 per reactor year, and the total ODE is less than 104 per reactor year (Region Ill in RG 1.174). The best-estimate ACDF is 7.0E-07, which is less than the 1 .0E-06 limit, whereas the corresponding best-estimate ALERE of 1 .8E-09 is well within the 1 .0E-07 limit and the total LERF is less than 10-s per reactor year (Region III in RG 1.174). The ACDF is conservatively based on the exceedance frequency from NUREG 1829 as opposed to the STP PRA Version 7.2 LLOCA exceedance frequency of 1 .34E-6 per year. The LLOCA frequencies assumed are for breaks of 6 inches or greater. Discussion of the STP PRA capability is provided in Enclosure 4-2 of Reference 1 to the cover letter. However, the STP PRA is only used in this analysis to develop the conditional probability of LERF, given strainer failure for LLOCA ODE.* ACDF: 7.0E-07/yr

• ALERE: 1 .8E-09/yr* ICCDP: 7.7E-07* ICLERP: 1.9E-09 Tier 2, Avoidance of Risk-Significant Plant Configurations STPNOC avoids and manages potentially risk-significant configurations by the application of the TS and through its risk monitoring and risk management processes for TS and the Maintenance Rule. Degraded or nonconforming ECCS and CSS configurations that cause the systems to be inoperable are managed with the non-debris related actions in the TS, which limit the time the plant can be operated with the system in a degraded condition.

The risk associated with overall plant configurations is continuously monitored with STPNOC's configuration risk management tool in accordance with Maintenance Rule requirements.

These programs include requirements for appropriate compensatory actions and are governed by station procedures.

Tier 3, Risk-Informed Configuration Management The ORMP implemented for the Maintenance Rule and for STP's Risk Managed TS quantify configuration risk based on actual equipment availability and provide guidance for minimizing time in higher risk configurations and provide guidance for use of compensatory action.3.0 Evaluation of Defense-in Depth (DID) and Safety Margin NOC-AE-1 5003241 Attachment 3 Page 16 of 22 Detailed evaluations of DID and Safety Margin are presented in Attachment 1-4 and can be applied to both the methodology change and the TS change.4.0 Implementation and Monitoring Program Design modifications addressing GSI-191 concerns, including installation of new sump strainers and replacement of problematic insulation, have been previously implemented using the STP design change process.STPNOC has implemented procedures and programs for monitoring, controlling and assessing changes to the plant that have a potential impact on plant performance related to GS1-191 concerns.

These provide the capability to monitor the performance of the sump strainers and the ability to assess impacts to the inputs and assumptions used in the PRA and the associated engineering analysis that support the proposed change. Programmatic requirements ensure that the potential for debris loading on the sump does not materially increase.

These include:*Programs and procedures have been implemented to evaluate and control potential sources of debris in containment:

o Technical Specification Surveillance Requirements implemented by STP procedures require visual inspections of all accessible areas of the containment to check for loose debris, and each containment sump to check for debris, as described in Section 4.1.3.oThe STP Design Change Package procedure includes provisions for managing potential debris sources such as insulation, qualified coatings, addition of aluminum or zinc, and potential effects of post-LOCA debris on recirculation flow paths and downstream components.

The procedure has been augmented to explicitly require changes that involve any work or activity inside the containment be evaluated for the potential to affect the following:

• Reactor coolant pressure boundary integrity* Accident or post-accident equipment inside containment

• Quantity of metal inside containment

• Quantity or type of coatings inside containment

• Thermal insulation changed or added* Post-LOCA recirculation flow paths to the emergency sumps* Post-LOCA recirculation debris impact on internals of fluid components

• Addition or deletion of cable A 10CFR50.59 screening or evaluation is required to be completed for all design changes. This process ensures that new insulation material that may differ from the initial design is evaluated for GS1-1 91 concerns.o Programs to ensure that Service Level 1 protective coatings used inside containment are procured, applied, and maintained in compliance with applicable regulatory

NOC-AE-1 5003241 Attachment 3 Page 17 of 22 requirements.

Procedures have been implemented to govern the use of signs and labels inside containment.

• As part of the STP Condition Reporting Process, condition reports are written due to adverse conditions identified during the containment inspections or containment emergency sumps and strainers surveillances.

• The STP Maintenance Rule program includes performance monitoring of functions associated with ECCS and CSS. The inclusion of the ECCS and CSS into the Maintenance Rule program and the assessment of acceptable system performance provide continued assurance of the availability for performance of the required functions.

• Quality Assurance (QA): The STP QA program is implemented and controlled in accordance with the Operations Quality Assurance Plan and is applicable to 550s to an extent consistent with their importance to safety, and complies with the requirements of 10CFR50, Appendix B and other program commitments as appropriate.

The QA Program is implemented with documented instructions, procedures, and drawings which include appropriate quantitative and qualitative acceptance criteria for determining that prescribed activities have been satisfactorily accomplished.

Procedures control the sequence of required inspections, tests, and other operations when important to quality.To change these controls, the individual procedure must be changed and a similar level of review and approval given to the original procedure is required.

Such instructions, procedures, and drawings are reviewed and approved for compliance with requirements appropriate to their safety significance.

QA program controls are applied to safety-related SSCs to provide a high degree of confidence that they perform safely and activities are performed as expected.

The rigorous controls imposed by the QA program provide adequate quality control elements to ensure system component reliability for the required functions.

The proposed change does not involve any changes to ASME Section Xl inspection programs or mitigation strategies that have been shown effective in early detection and mitigation of weld and material degradation in Class I piping applications.

5.0 Technical

Evaluation Conclusion The technical evaluation shows that the functionality of the ECCS and CSS during design basis accidents is confirmed by demonstrating that safety margin and defense-in-depth are maintained with high probability (Region Ill RG 1.174).

NOC-AE-1 5003241 Attachment 3 Page 18 of 22 6.0 Regulatory Evaluation

6.1 Regulatory

Requirements The following regulations apply to the proposed amendment.

Approval of the proposed amendment is contingent upon approval of the requests for exemptions from these regulations as provided and justified in Attachments 2-1 through 2-5.* 10CFR50.46(d)

• GDC 35, "Emergency core cooling"* GDC 38, "Containment heat removal,"* GDC 41, "Containment atmosphere cleanup," 6.2 Regulatory Guidance NRC Regqulatory Guide 1.174, 'An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides the NRC staff's recommendations for using risk information in support of licensee-initiated Licensing Basis changes to a nuclear power plant that require NRC review and approval.

This regulatory guide describes an acceptable approach for assessing the nature and impact of proposed Licensing Basis changes by considering engineering issues and applying risk insights.In implementing risk-informed decision making, Licensing Basis changes are expected to meet a set of key principles.

These principles include the following:

1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under IOCFR5O. 12, "Specific Exemptions").

The exemptions requested in Attachment 2 to this letter implement this requirement.

2. The proposed change is consistent with a defense-in-depth philosophy.

Defense-in-depth is presented in detail in Attachment 1-4. The proposed change is consistent with the defense-in-depth philosophy in that the following aspects of the facility design and operation are unaffected:

• Functional requirements and the design configuration of systems* Existing plant barriers to the release of fission products* Design provisions for redundancy, diversity, and independence

• Plant's response to transients or other initiating events* Preventive and mitigative capabilities of plant design features The STP risk-informed approach analyzes a full spectrum of LOCAs, including double-ended guillotine breaks for all piping sizes up to and including the largest pipe in the reactor coolant system. By requiring that mitigative capability be maintained in a realistic NOC-AE-1 5003241 Attachment 3 Page 19 of 22 and risk-informed evaluation of GS1-191 for a full spectrum of LOCAs, the approach ensures that defense-in-depth is maintained.

3. The proposed change maintains sufficient safety margins.As described in Attachment 1-4, sufficient safety margins associated with the design will be maintained by the proposed change.4. When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

The proposed change is defined as the risk associated with effects of LOCA debris. Using engineering analysis and the PRA, this risk has been calculated and shown to be "very small" as defined by Region Ill in RG 1.174 and is therefore consistent with the Commission's Safety Goal Policy Statement 5. The impact of the proposed change should be monitored using performance measurement strategies.

Performance monitoring is discussed in Section 4.0 above.NRC Regqulatory Guide 1.177. "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications", describes an ~acceptable approach for evaluating risk-informed changes to TS. STPNOC applied the guidance of RG 1.177 to the evaluation of the proposed TS changes for ECCS and CSS.NRC Regqulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

The STPNOC PRA model used for the risk-informed approach for addressing GS1-191 concerns is in compliance with Revision 1 of RG 1.200 (ML13323A183, Enclosure 4-2; ML14178A481, Attachment 1, pg. 2).6.3 Precedents STPNOC identified no precedents.

The NRC plans to use STP Units 1 and 2 as a pilot for other licensees choosing to a risk-informed approach for closure of GL 2004-02.6.4 No Significant Hazards Consideration Determination STPNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10CFR50.92,"Issuance of amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

NOC-AE-15003241 Attachment 3 Page 20 of 22 Response:

No.The proposed changes are a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the STP licensing basis, changes to the ECCS and CSS TS to extend the required completion time for potential LOCA debris related effects and associated administrative TS changes. The methodology change concludes that the ECCS and CSS will have sufficient defense-in-depth and safety margin and will operate with high probability following a LOCA when considering the impacts and effects of debris accumulation on containment emergency sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss of coolant accidents.

The methodology change also supports the changes to the TS.There is no significant increase in the probability of an accident previously evaluated.

The proposed changes address mitigation of loss of coolant accidents and have no effect on the probability of the occurrence of a loss of coolant accident.

The proposed methodology and TS changes do not implement any physical changes to the facility or any SSCs, and do not implement any changes in plant operation that could lead to a different kind of accident.The proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

The methodology change confirms that required SSCs supported by the containment sumps will perform their safety functions with a high probability, as required, and does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits. The safety analysis acceptance criteria in the UFSAR continue to be met for the proposed methodology change. The evaluation of the changes determined that containment integrity will be maintained.

The dose consequences were considered in the assessment and quantitative evaluation of the effects on dose using input from the risk-informed approach shows the increase in dose consequences is small.Therefore, the proposed change does not involve a significant increase in the probability or consequences of any the accident previously evaluated in the UFSAR.2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.The proposed changes are a methodology change for assessment of debris effects from LOCAs that are already evaluated in the STP UFSAR, an extension of TS required completion time for potential LOCA debris related effects on ECCS and CSS, and associated administrative changes to the TS. No new or different kind accident is being evaluated.

None of the changes install or remove any plant equipment, or alter the design, physical configuration, or mode of operation of any plant structure, system or component.

The proposed changes do not introduce any new failure mechanisms or malfunctions that can initiate an accident.

The proposed changes do not introduce failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident.

NOC-AE-1 5003241 Attachment 3 Page 21 of 22 Therefore, the proposed changes do not create the possibility for a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?Response:

No.The proposed changes are a methodology change for assessment of debris effects from LOCAs that are already evaluated in the STP UFSAR, an extension of TS required completion time for potential LOCA debris related effects on ECCS and CSS, and associated administrative changes to the TS. The effects from a full spectrum of LOCAs, including double-ended guillotine breaks for all piping sizes up to and including the largest pipe in the reactor coolant system, are analyzed.

Appropriate redundancy and consideration of loss of offsite power and worst case single failure are retained, such that defense-in-depth is maintained.

Application of the risk-informed methodology showed that the increase in risk from the contribution of debris effects is very small as defined by RG 1.174 and that there is adequate defense in depth and safety margin. Consequently, STP determined that the risk-informed method demonstrates the containment sumps will continue to support the ability of safety related components to perform their design functions when the effects of debris are considered.

The proposed change does not alter the manner in which safety limits are determined or acceptance criteria associated with a safety limit. The proposed change does not implement any changes to plant operation, and does not significantly affect SSCs that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

The proposed change does not significantly affect the existing safety margins in the barriers for the release of radioactivity.

There are no changes to any of the safety analyses in the UFSAR.Defense in depth and safety margin was extensively evaluated for the methodology change and the associated TS changes. The evaluation determined that there is substantial defense in depth and safety margin that provide a high level of confidence that the calculated risk for the methodology and TS changes is conservative and that the actual risk is likely much lower.Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the above, STPNOC concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10CFR50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

6.5 Conclusion

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations contingent upon approval of the exemption requested in Attachment 2 to this letter, and (3)

NOC-AE-1 5003241 Attachment 3 Page 22 of 22 the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.7.0 Environmental Consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9).

Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

NOC-AE-1 5003241 Attachment 3-1 Technical Specification Page Markups NOC-AE-1 5003241 Attachment 3-1 Page 1 of 4 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS

-TAvG.GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION 3.5.2 Three independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of: a. One OPERABLE High Head Safety Injection pump, b. One OPERABLE Low Head Safety Injection pump, c. One OPERABLE RHR heat exchanger, and d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation through a High Head Safety Injection pump and into the Reactor Coolant System and through a Low Head Safety Injection pump and its respective RHR heat exchanger into the Reactor Coolant System.APPLICABILITY:

MODES 1, 2, and 3.*ACTION: a. With less than the above subsystems OPERABLE, but with at least two High Head Safety Injection pumps in an OPERABLE status, two Low Head Safety Injection pumps and associated RHR heat exchangers in an OPERABLE status, and sufficient flow paths to accommodate these OPERABLE Safety Injection pumps and RHR heat exchangers, **within 7 days restore the inoperable subsystem(s) to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.b. With less than two of the required subsystems OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least two subsystems to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.SOUTH TEXAS -UNITS 1 & 2 3/4 5-3 Unit 1 -Amendment No. 45-,1--7-0 179 Unit 2-Amendment No. !39, !45& 166 NOC-AE-1 5003241 Attachment 3-1 Page 2 of 4 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS

-TArt. GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION c. With less than the required flow paths OPERABLE solely due to potential effects of LOCA generated and transported debris that exceeds analyzed amounts. perform the following:

1. Immediately initiate action to implement compensatory actions, AND 2. Within 90 days restore the affected flowpath(s) to OPERABLE status, OR Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the followingq 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.d. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be submitted within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0. 70.* Entry into MODE 3 is permitted for the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.1.2 provided that the Safety Injection pumps are restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375°F, whichever comes first.**Verify required pumps, heat exchangers and flow paths OPERABLE every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.SOUTH TEXAS -UNITS 1 & 2 CONTAINMENT SYSTEMS 3/4 5-3a Unit 1 -Amendment No. 4151,17O 179 Unit 2 -Amendment No. 4~9-1-58=

166 NOC-AE-1 5003241 Attachment 3-1 Page 3 of 4 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONITAINMECNDTISPRA SYSTE LIMITING CONDITION FOR OPERATION 3.6.2.1 Three independent Containment Spray Systems shall be OPERABLE with each Spray system capable of taking suction from the RWST and transferring suction to the containment sump.APPLICABILITY:

MODES 1, 2, 3, and 4 ACTION: a. With one Containment Spray System inoperable, within 7 days restore the inoperable Spray System to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.b. With more than one Containment Spray System inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least two Spray Systems to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.c. With one or more Containment Spray Systems inoperable in MODE 1. 2. or 3 solely due to note~nti~l nf I OCA dlhriQ thaf ,IQl 2fnlv2II7Ol amounts, perform the following:

1. immediately initiate action to implement compensatory actions, AND 2. within 90 days restore the affected system(s) to OPERABLE status, Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SOUTH TEXAS -UNITS 1 & 2 3/4 6-14 Unit 1 -Amendment No. 56, 4 188 Unit 2 -Amendment No. 444 466 175 NOC-AE-1 5003241 Attachment 3-1 Page 4 of 4 CONTAINMENT SPRAY SYSTEM SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE: a. At a frequency in accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;b. By verifying on a STAGGERED TEST BASIS, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 283 psid when tested pursuant to Specification 4.0.5;c. At a frequency in accordance with the Surveillance Frequency Control Program during shutdown, by: 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure High 3 test signal, and 2) Verifying that each spray pump starts automatically on a Containment Pressure High 3 test signal coincident with a sequencer start signal.d. By verifying each spray nozzle is unobstructed following maintenance activities that could result in spray nozzle blockage.SOUTH TEXAS -UNITS 1 & 2 3/4 6-14a Unit 1 -Amendment No. 4-56, 7- 188 Unit 2 -Amendment No. 4447, 466 175 NOC-AE-1 5003241 Attachment 3-2"Clean" Technical Specification Pages EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS

-TAve GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION 3.5.2 Three independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of: a. One OPERABLE High Head Safety Injection pump, b. One OPERABLE Low Head Safety Injection pump, c. One OPERABLE RHR heat exchanger, and d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation through a High Head Safety Injection pump and into the Reactor Coolant System and through a Low Head Safety Injection pump and its respective RHR heat exchanger into the Reactor Coolant System.APPLICABILITY:

MODES 1, 2, and 3.*ACTION: a. With less than the above subsystems OPERABLE, but with at least two High Head Safety Injection pumps in an OPERABLE status, two Low Head Safety Injection pumps and associated RHR heat exchangers in an OPERABLE status, and sufficient flow paths to accommodate these OPERABLE Safety Injection pumps and RHR heat exchangers, **within 7 days restore the inoperable subsystem(s) to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.b. With less than two of the required subsystems OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least two subsystems to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.SOUTH TEXAS -UNITS 1 & 2 3/4 5-3 Unit 1 -Amendment No. 45-1, !170 179 Unit 2 -Amendment No. !39, !58-46..

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS

-TAVG GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION c. With less than the required flow paths OPERABLE solely due to potential effects of LOCA generated and transported debris that exceeds analyzed amounts, perform the following:

1. Immediately initiate action to implement compensatory actions, AND 2. Within 90 days restore the affected flowpath(s) to OPERABLE status, OR Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.d. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be submitted within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0. 70.* Entry into MODE 3 is permitted for the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.1.2 provided that the Safety Injection pumps are restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375 0 F, whichever comes first.** Verify required pumps, heat exchangers and flow paths OPERABLE every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.SOUTH TEXAS -UNITS 1 & 2 3/4 5-3a Unit 1 -Amendment No. 4&5!,4!70--479 Unit 2 -Amendment No. 3-614 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONITAINMECNDTISPRA SYSTE LIMITING CONDITION FOR OPERATION 3.6.2.1 Three independent Containment Spray Systems shall be OPERABLE with each Spray system capable of taking suction from the RWST and transferring suction to the containment sump.APPLICABILITY:

MODES 1, 2, 3, and 4 ACTION: a. With one Containment Spray System inoperable, within 7 days restore the inoperable Spray System to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.b. With more than one Containment Spray System inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least two Spray Systems to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.c. With one or more Containment Spray Systems inoperable in MODE 1, 2, or 3 solely due to potential effects of LOCA generated and transported debris that exceeds analyzed amounts, perform the following:

1. immediately initiate action to implement compensatory actions, AND 2. within 90 days restore the affected system(s) to OPERABLE status, OR Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SOUTH TEXAS -UNITS 1 & 2 3/4 6-14 Unit 1 -Amendment No. 156,1!70, 188 Unit 2 -Amendment No. 4A-17 SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment spray System shall be demonstrated OPERABLE: a. At a frequency in accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;b. By verifying on a STAGGERED TEST BASIS, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 283 psid when tested pursuant to Specification 4.0.5;c. At a frequency in accordance with the Surveillance Frequency Control Program during shutdown, by: 1) Verifying that each automatic valve in the flo~iv path actuates to its correct position on a Containment Pressure High 3 test signal, and 2) Verifying that each spray pump starts automatically on a Containment Pressure High 3 test signal coincident with a sequencer start signal.d. By verifying each spray nozzle is unobstructed following maintenance activities that could result in spray nozzle blockage.SOUTH TEXAS -UNITS 1 & 2 3/4 6-14a Unit 1 -Amendment No. 156, 170, 188 Unit 2 -Amendment No. !'!,!.,1-6,175 NOC-AE-1 5003241 Attachment 3-3 Technical Specifications Bases Page Markups (Information Only)SOUTH TEXAS -UNITS 1 & 2 3/4 6-14 Unit 1 -Amendment No. 156, 179, 188 Unit 2-Amendment No. 144,- !66,!75 NOC-AE-1 5003241 Attachment 3-3 Page 1 of 5 Technical Specifications Bases Page Markups (Information Only)Add the followingq to the Bases Section for 3/4.5.2 EGOS Subsystems:

The OPERABILITY of the ECCS Subsystems is assured by the capability of the containment emergency sump strainers to limit entry of debris into the sump and recirculating lines. This capability ensures that the flow and net positive suction head requirements of ECCS are satisfied.

Assurance that containment debris will not block the sump strainers and render the ECCS Subsystem inoperable on emergency recirculation during design basis accidents is provided by inspection and engineering evaluation.

UFSAR Appendix 6A provides a risk-informed approach that addresses the potential of debris blockage concluding that long-term core cooling following a design basis loss of coolant accident is assured with high probability.

UFSAR Appendix 6A also provides guidance for assessing the potential impact on Operability due to unexpected material such as loose debris discovered in containment that may contribute to debris loading on the strainers.

Technical Basis: The Licensing Basis with regard to effects of debris is that there is a high probability that ECOS and CSS can perform their design basis functions based on successful plant-specific prototypical testing using deterministic NRC-approved assumptions, and that the risk from breaks that could generate debris that is not bounded by the testing is very small and acceptable in accordance with the criteria of RG 1.174.STP evaluated the risk associated with the effects on long-term cooling due to debris accumulation on Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects of debris that penetrates the strainers.

A full spectrum of postulated LOCAs is analyzed, including double-ended guillotine breaks (DEGBs) for all pipe sizes up to the largest pipe in the reactor coolant system. The changes to CDF and LERF associated with the effects of debris are quantified by applying the LOCA frequencies published in NUREG-1829, and then compared to RG 1.174 acceptance guidelines.

The STP analysis shows that the contribution to risk from the breaks that are not deterministically mitigated is within RG 1.174 Region Ill.Strainer Operability:

The affected ECCS and CSS are OPERABLE with respect to the effects of debris when the expected effects of the debris on the emergency sump strainers are consistent with the analysis.This operability requirement is based on quantity and characteristics and location of the debris in the RCB being consistent with the debris analysis assumptions.

The types of debris considered in the analysis included insulation and latent debris fiber fines, particulate from coatings and latent debris, and chemical precipitates (primarily from aluminum corrosion).

Strainer operability evaluation is fundamentally deterministic and the intent is to not require a risk assessment to make the operability determination.

The criterion recognizes that there is margin NOC-AE-1 5003241 Attachment 3-3 Page 2 of 5 and conservatism in the debris assumptions used for the deterministic testing and in the debris generation and transport analyses that can be applied to account for previously unidentified debris.Guidance for evaluating potential debris is provided in 0PSP03-XC-0002A.

Applicability:

This required action applies only for the potential effects of debris on emergency sump strainer operability or on in-core debris effects. It does not apply for effects other than those caused by debris. Debris effects are conditions caused by transportable debris that could impact the net positive suction head or otherwise degrade pump performance, or cause strainer structural failure by excess accumulation on one of more of the emergency sump strainers.

Gaps or other conditions that are a physical degraded or nonconforming condition of the strainer are to be addressed by the system train-specific, non-debris TS actions a and b.The requirements apply in MODE 1, 2, and 3. In these MODEs, the plant is in normal operating pressure and temperature where generation of design basis quantities of debris can reasonably be postulated.

For lower MODEs of operation, there is less energy in the RCS and reduced capability to generate the zones of influence associated with pipe breaks, and the core is at generally lower levels of decay heat generation.

Consequently, effects of debris are less likely to cause a condition where ECCS or 0SS is inoperable.

Technical Specifications require that all applicable actions must be entered. If concurrent maintenance requirements or a non-debris related degraded or nonconforming condition occurs that would make any system(s) or subsystem(s) inoperable, the non-debris required action for the system(s) or subsystem(s) must be applied. The action from the debris related condition will continue to apply from the time it was initially entered.Required Action: The required action to implement compensatory action is based on the very low contribution by LOCA generated debris to the risk of core damage, and is a reasonable response to minimize the potential increase in risk from the debris source. Typical compensatory action would include actions such as:* Remove the debris or source of debris or take action that would prevent transport of the debris to the emergency sump* Defer maintenance that would affect availability of the affected systems and strainers* Increase frequency of RCS leak detection monitoring

• Brief operators on LOCA debris management actions K NOC-AE-1 5003241 Attachment 3-3 Page 3 of 5 Completion Time: The 90 day completion time is based on the very low contribution to risk from LOCA generated debris. It provides sufficient time to more thoroughly assess the condition and to take corrective action. Operability can be restored by mitigation of the debris such as by removal or making it non-transportable, or by performing an evaluation that demonstrates that the Licensing Basis is maintained.

Add the following to the Bases Section 3/4.6.2.1 Containment Spray System: The OPERABILITY of the ECCS Subsystems is assured by the capability of the containment emergency sump strainers to limit entry of debris into the sump and recirculating lines. This capability ensures that the flow and net positive suction head requirements of ECCS are satisfied.

Assurance that containment debris will not block the sump strainers and render the ECCS Subsystem inoperable on emergency recirculation during design basis accidents is provided by inspection and engineering evaluation.

UFSAR Appendix 6A provides a risk-informed approach that addresses the potential of debris blockage concluding that long-term core cooling following a design basis loss of coolant accident is assured with high probability.

UFSAR Appendix 6A also provides guidance for assessing the potential impact on Operability due to unexpected material such as loose debris discovered in containment that may contribute to debris loading on the strainers.

Technical Basis: The Licensing Basis with regard to effects of debris is that there is a high probability that ECCS and CSS can perform their design basis functions based on successful plant-specific prototypical testing using deterministic NRC-approved assumptions, and that the risk from breaks that could generate debris that is not bounded by the testing is very small and acceptable in accordance with the criteria of RG 1.174.STp evaluated the risk associated with the effects on long-term cooling due to debris accumulation on Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects of debris that penetrates the strainers.

A full spectrum of postulated LOCAs is analyzed, including double-ended guillotine breaks (DEGBs) for all pipe sizes up to the largest pipe in the reactor coolant system. The changes to CDF and LERF associated with the effects of debris are quantified by applying the LOCA frequencies published in NUREG-1829, and then compared to RG 1.174 acceptance guidelines.

The STP analysis shows that the contribution to risk from the breaks that are not deterministically mitigated is within RG 1.174 Region Ill.Strainer Operability:

The affected ECCS and CSS are OPERABLE with respect to the effects of debris when the expected effects of the debris on the emergency sump strainers are consistent with the analysis.This operability requirement is based on quantity and characteristics and location of the debris in the RCB being consistent with the debris analysis assumptions.

The types of debris considered NOC-AE-1 5003241 Attachment 3-3 Page 4 of 5 in the analysis included insulation and latent debris fiber fines, particulate from coatings and latent debris, and chemical precipitates (primarily from aluminum corrosion).

Strainer operability evaluation is fundamentally deterministic and the intent is to not require a risk assessment to make the operability determination.

The criterion recognizes that there is margin and conservatism in the debris assumptions used for the deterministic testing and in the debris generation and transport analyses that can be applied to account for previously unidentified debris.Guidance for evaluating potential debris is provided in 0PSP03-XC-0002A.

Applicability:

This required action applies only for the potential effects of debris on emergency sump strainer operability or on in-core debris effects. It does not apply for effects other than those caused by debris. Debris effects are conditions caused by transportable debris that could impact the net positive suction head or otherwise degrade pump performance, or cause strainer structural failure by excess accumulation on one of more of the emergency sump strainers.

Gaps or other conditions that are a physical degraded or nonconforming condition of the strainer are to be addressed by the system train-specific, non-debris TS actions a and b.The requirements apply in MODE 1, 2, and 3. In these MODEs, the plant is in normal operating pressure and temperature where generation of design basis quantities of debris can reasonably be postulated.

For lower MODEs of operation, there is less energy in the RCS and reduced capability to generate the zones of influence associated with pipe breaks, and the core is at generally lower levels of decay heat generation.

Consequently, effects of debris are less likely to cause a condition where EGOS or CSS is inoperable.

Technical Specifications require that all applicable actions must be entered. If concurrent maintenance requirements or a non-debris related degraded or nonconforming condition occurs that would make any system(s) or subsystem(s) inoperable, the non-debris required action for the system(s) or subsystem(s) must be applied. The action from the debris related condition will continue to apply from the time it was initially entered.Required Action: The required action to implement compensatory action is based on the very low contribution by LOCA generated debris to the risk of core damage, and is a reasonable response to minimize the potential increase in risk from the debris source. Typical compensatory action would include actions such as: , Remove the debris or source of debris or take action that would prevent transport of the debris to the emergency sump* Defer maintenance that would affect availability of the affected systems and strainers* Increase frequency of RCS leak detection monitoring

• Brief operators on LOCA debris management actions NOC-AE-1 5003241 Attachment 3-3 Page 5 of 5 Completion Time: The 90 day completion time is based on the very low contribution to risk from LOCA generated debris, It provides sufficient time to more thoroughly assess the condition and to take corrective action. Operability can be restored by mitigation of the debris such as by removal or making it non-transportable, or by performing an evaluation that demonstrates that the Licensing Basis is maintained.