L-PI-18-064, License Amendment Request (LAR) to Revise Current Licensing Basis, Supplement 1

License Amendment Request (LAR) to Revise Current Licensing Basis, Supplement 1

ML18338A431
Person / Time
Site:	Prairie Island
Issue date:	12/04/2018
From:	Sharp S Northern States Power Company, Minnesota, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-18-064
Download: ML18338A431 (17)

Text

1717 Wakonade Drive Welch, MN 55089 fl Xcel Energy*

RESPONSIBLE BY NATURE 800.895.4999 xcelenergy.com December 4, 2018 L-PI-18-064 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 License Amendment Request (LAR) to Revise Current Licensing Basis, Supplement 1

References:

1. Letter from NSPM to NRC, License Amendment Request (LAR) to Revise Current Licensing Basis, dated October 2, 2018. (ADAMS Accession ML18275A370)

2. Letter from NRC to NSPM, Prairie Island Nuclear Generating Plant, Units 1 and 2 -

Supplemental Information Needed for Acceptance of Requested Licensing Action re:

Amendment to Reclassify Certain Fuel Handling Equipment. (ADAMS Accession ML18270A015)

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), submitted a request for approval for changes to the Prairie Island Nuclear Generating Plant (PINGP) licensing basis regarding the safety classification of certain fuel handling equipment. As a result of acceptance review of Reference 1, the NRC concluded that additional information is necessary to enable the staff to make an independent assessment regarding acceptability of the proposed amendment. In a teleconference on November 15, 2018, and in Reference 2 the NRC clarified the additional information being requested.

NSPM is providing this supplement to the LAR to revise and clarify the LAR in response to the documented insufficiencies. Specific changes to the LAR include:

Removal of the Auxiliary Building Crane from the scope of the LAR. Also, providing additional information on the seismic design requirements of the manipulator and spent fuel pool cranes and clarification that these design requirements will not be changed as a result of the proposed amendment, so there would remain no need to postulate a crane collapse or evaluate dose consequences of such a collapse.

Clarify that the proposed use of the definition of substantial amount of radioactivity of 10% of 10 CFR 100 (consistent with industry standards) in lieu of the current site-

Document Control Desk Page 2 specific definition of 1% of 10 CFR 100 will apply only to equipment used to handle single fuel assemblies. Further clarify that the quality assurance requirements of NSPM commitments to the application of NUREG-0554 for single failure proof cranes will not be changed as a result of the proposed amendment.

• Clarify that the proposed change is to allow the application of a definition of substantial amount of radioactivity consistent with ANSI/ANS 58.14 as it applies to safety classification (e.g., Safety Related, Augmented Quality, or Non-Safety Related) of the identified fuel handling equipment as opposed to PINGP historic designation of QA Type (e.g., QA Type I, QA Type II, or QA Type Ill). As such, one of the proposed changes to the PINGP Updated Safety Analysis Report (USAR) in Reference 1 was determined to be unnecessary as the statement in the USAR is historical in nature. provides a revised description and assessment of the proposed changes. provides the marked up USAR page that reflects the proposed changes as revised.

Approval of the proposed amendment is requested within 12 months of the acceptance of this request. Once approved, the amendment shall be implemented within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Minnesota Official.

If there are any questions or if additional information is needed, please contact Mr. Jeff Kivi at (612) 330-5788.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury, that the foregoing is true and correct.

Executed on D&-t-GrtB6e.. H. , .z.o 18 .

,PJ~

Scott Sharp Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota

Document Control Desk Page 3 ENCLOSURES:

1. Evaluation of Proposed Change

2. Markup of Updated Safety Analysis Report Page cc: Administrator, Region III, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State of Minnesota

ENCLOSURE 1 PRAIRE ISLAND NUCLEAR GENERATING PLANT Evaluation of Proposed Change License Amendment Request (LAR) to Revise Current Licensing Basis

1.

SUMMARY

DESCRIPTION 2

2. DETAILED DESCRIPTION 2 2.1 Proposed Change 2 2.2 Background 3 2.3 Reason for the Proposed Change 6 2.4 Description of Equipment 6

3. REGULATORY EVALUATION 8 3.1 Applicable Regulatory Requirements/Criteria 8 3.2 No Significant Hazards Consideration Analysis 11 3.3 Conclusions 12

4. ENVIRONMENTAL EVALUATION 12

5. REFERENCES 12 Page 1 of 12

DESCRIPTION AND ASSESSMENT 1.0

SUMMARY

DESCRIPTION Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), is submitting a request for approval of changes to the Prairie Island Nuclear Generating Plant (PINGP) licensing basis regarding the safety classification of certain fuel handling equipment. The PINGP-specific definition of Quality Assurance (QA) Type I included a definition of a substantial amount of radioactivity being 1% of 10 CFR 100 limits.

The industry standard, ANSI/ANS 58.14, Safety and Pressure Integrity Classification Criteria for Light Water Reactors, establishes 10% of 10 CFR 100 limits as a comparable offsite exposure. The original PINGP specific classification system (QA Type) has been replaced by a system (Safety Classification) utilizing industry standard and regulatory documents as input.

Thus, the current classification system is based on SSC functions and uses standard industry terminology, such as safety related. In implementing the current classification system, the definition of substantial amount of radiation was not changed to 10% of 10 CFR 100 to reflect the ANSI/ANS 58.14 comparable offsite exposure.

The proposed changes would allow the fuel handling equipment identified in this request that are currently classified as safety-related based solely on the plant-specific definition of a substantial amount of radioactivity to be reclassified as augmented quality1 or non-safety related depending upon applicable requirements. These proposed changes are not intended to change the design requirements of the affected fuel handling equipment nor are the changes intended to relax any quality assurance requirements required to meet the requirements of NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants," or seismic category II/I design requirements applicable to PINGP.

2.0 DETAILED DESCRIPTION 2.1 Proposed Change This license amendment request (LAR) proposes changes to the PINGP licensing basis and does not involve changes to the Facility Operating License, Technical Specifications (TS), or TS Bases. Upon approval, the current licensing basis for fuel handling equipment within the Updated Safety Analysis Report (USAR) will be modified by changing the footnote on USAR Page 1.5-3. Draft changes to the USAR are provided in Enclosure 2.

This amendment would revise the PINGP USAR regarding specific fuel handling equipment to apply the ANSI/ANS 58.14 definition of substantial amount of radiation in lieu of the PINGP-specific definition. No change to the Design Class or design requirements of components is proposed in this amendment. The change is requested specific to fuel handling systems/tools that are required for handling individual irradiated fuel assemblies. No change is requested for existing QA requirements imposed by NSPM commitments to the application of NUREG-0554.

Specific fuel handling components are:2 1

Augmented Quality within the NSPM process is equivalent to ANSI/ANS 58.14 Supplemented Grade Items.

2 NSPM no longer proposes to apply this amendment to the Auxiliary Building Crane.

Page 2 of 12

Manipulator Cranes, including the Load Cells Spent Fuel Pool Bridge Crane Spent Fuel Transfer System and its constituent components (conveyor, upenders, and related equipment), exclusive of the transfer tube and the blind flange Rod Cluster Control Changing Fixtures Spent Fuel Assembly Handling Tools The change in definition of substantial amount of radioactivity applied to safety classification for the identified equipment is justified because its failure could not cause or increase the severity of a loss of coolant accident and is not vital to the safe shutdown and isolation of the reactor. Absent any other QA requirements, the change in definition would allow the identified equipment to be reclassified because results of the fuel handling accident analysis indicate failure of any of the identified equipment could result in the uncontrolled release of radioactivity between 1% and 10% of limits described in 10 CFR 100. The change does not affect QA requirements imposed for structural seismic II/I requirements that apply to PINGP. There is no intended change to the QA requirements associated with handling of heavy loads, such as those imposed by the NSPM commitments to the application of NUREG-0554 at the PINGP.

The proposed change will not result in any change in quality requirements for equipment used to handle heavy loads, so there would be no potential adverse effects on the safety-related functions of nearby equipment during critical load handling activities.

2.2 Background The Atomic Energy Commission (AEC) issued a construction permit for PINGP Units 1 and 2 on June 25, 1968 (ADAMS Accession No. ML022170205). PINGP was designed and constructed to comply with NSPMs understanding of the intent of the Atomic Energy Commissions Proposed General Design Criteria For Nuclear Power Plant Construction Permits dated July 11, 1967, hereafter referred to as the AEC GDC (32 FR 10213). Because the construction of the plant was significantly completed prior to the issuance of the 10 CFR 50, Appendix A General Design Criteria (GDCs), the plant was not reanalyzed and the Final Safety Analysis Report (FSAR) was not revised to reflect these later criteria. However, the AEC Safety Evaluation dated September 28, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and, ... are satisfied that the plant design generally conforms to the intent of these criteria.

PINGP-specific Definitions of Design Class and Quality Assurance Type (QA Type)

PINGP was designed prior to the issuance of current regulatory guidance and industry standards for classifying the quality level of systems, structures, and components (SSCs).

Instead a scheme of Quality Assurance Types was developed during plant construction.

The PINGP definition of QA Type, as described in USAR Section 1.5, discussion of AEC GDC Criterion 1, is given in the following paragraphs:

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The original, PINGP specific, classification system has been replaced by a system utilizing industry standard and regulatory documents as input. Thus, the current classification system is based on SSC functions and uses standard industry terminology, such as safety related.

In general, QA Type I is associated with Safety Related, QA Type II is associated with Augmented Quality (a subset of Non-Safety Related), and QA Type III is associated with standard quality Non-Safety Related. Safety Related, Augmented Quality, and Non-Safety Related are defined in applicable fleet procedures.

Quality Assurance Types QA Type I - Those items for which the Quality Assurance Program must assure the highest feasible degree of quality standards consistent with the importance of the safety function to be performed. This category includes those items of the plant which are essential to the prevention of accidents which could affect the public health and safety by the release of quantities(1) of radioactivity or are required in the mitigation of the consequences of such accidents.

QA Type II - Those items for which the Quality Assurance Program must engender a high confidence that the item will perform satisfactorily. This category includes those items whose failure would not directly affect the health and safety of the public, but the failure of which could cause severe economic loss or cause the plant to experience an extended outage.

QA Type III - This category includes all other items not included in Types I and II.

(1)

A substantial amount of radioactivity is defined as that amount of radioactive material which would produce radiation levels at the site boundary in excess of 1% of 10 CFR100.

The site-specific definition of substantial amount of radioactivity as 1% of the 10CFR100 limits to determine the classification of QA Type I SSCs (safety-related) was developed during original construction and is more conservative than the 10% of 10 CFR 100.11 limits eventually established in ANSI/ANS 58.14 (Reference 1).

The PINGP definition of the Nuclear Safety Design Classifications, as described in USAR Section 12.2.1.1, Classification of Structures and Components, is given in the following paragraphs:

All structures (including the Reactor Building), systems (including instruments and controls), and components were classified as Design Class I, Il or III according to their function and importance in relation to the safe operation of the reactor, with emphasis on the degree of integrity required to protect the public. These are listed in Table 12.2-1.

a. Design Class I Page 4 of 12

Those structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of substantial1amounts of radioactivity, and those structures and components vital to safe shutdown and isolation of the reactor.

b. Design Class I*

Some items in Table 12.2-1 are designated as Design Class I* indicating that these items have been originally designed or have been subsequently analyzed or tested to Design Class I, Design Basis Earthquake loading (dynamic) only, and that these items are treated as Design Class III items in all other respects.

1 A substantial amount of radioactivity is defined as that amount of radioactive material which would produce radiation levels at the site boundary in excess of 1.0% of 10CFR100 limits.

Both the definitions of Design Class I and QA Type I state that a substantial amount of radioactivity means greater than 1% of 10 CFR 100 limits at the site boundary.

With respect to the correlation between Design Class and QA Type, the PINGP USAR Section 12 notes:

The quality classification for all Design Class I and Design Class I* components listed in Table 12.2-1 are Type 1 and Type 3, respectively.

PINGP USAR Table 12.2-1, Classification of Structures, Systems and Components, gives the Design Class of the various plant structures, systems, and components, including a number of cranes used to handle irradiated fuel.

Related PINGP Licensing History In 1998, PINGP USAR Revision 16 changed section 12.2.1.1 to the current statement that correlated Design Class I to QA Type 1 and Design Class I* to QA Type 3. Prior to USAR Revision 16, PINGP USAR section 12.2.1.1 correlated both Design Class I and Design Class I* to QA Type I. The change was made under PINGP License Amendments 140/131, Cooling Water System Emergency Intake Design Bases (Reference 2).

In 2004, NSPM undertook a voluntary initiative to change the PINGP licensing basis to adopt ANSI/ANS-58.14-1993, for the classification of SSCs. The exception to this adoption of ANSI/ANS-58.14-1993 is that PINGP retained the existing definition of a substantial amount of radiation for classifying QA Type I SSCs rather than adopting the criteria that is referenced in ANSI/ANS-58.14-1993. ANSI/ANS-58.14-1993 ties classification of safety-related to functions relied upon in the safety analyses of design basis events (DBEs) to prevent or mitigate DBEs whose consequences could result in potential off-site exposures comparable to the guideline exposures of 10 CFR 100.11, where comparable to is defined as greater than or equal to 10%

of the limit.

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During the same time frame, PINGP was granted a license amendment in 2004 for a selective scope implementation of Alternate Source Term (AST) application for a fuel handling accident.

(Reference 3) PINGP was issued a license amendment for a full scope AST application in 2013 (Reference 4). For full implementation of the AST design basis accident analysis methodology, the dose acceptance criteria specified in 10 CFR 50.67 provides an alternative to the previous whole body and thyroid dose guidelines in stated in 10 CFR 100.11. The post-fuel handling accident TEDE dose was determined to be 2.28 rem at the Exclusion Area Boundary. This is less than 10% of the 10 CFR 50.67 limit of 25 rem. However, because the radiological consequences of a fuel handling accident are greater than 1% of the dose limits of 10 CFR 50.67 and 10 CFR 100, certain fuel handling equipment remains classified as safety related.

2.3 Reason for the Proposed Change The existing definition of a substantial amount of radiation in the PINGP USAR results in certain fuel handling equipment being conservatively classified as safety related, whereas accepted industry standards such as ANSI/ANS-58.14 would not require the same fuel handling equipment to be safety related. This results in additional expense and hardship in obtaining qualified replacement parts with no corresponding benefit to public health and safety.

2.4 Description of Equipment Spent Fuel Pool Bridge Crane The spent fuel pool bridge crane is a wheel-mounted walkway, spanning the spent fuel pool which carries electric monorail hoists on an overhead structure. The fuel assemblies are moved within the spent fuel pool by means of a long handled tool suspended from the hoist.

The hoist travel and tool length are designed to limit the maximum lift of a fuel assembly to a safe shielding depth. The crane was replaced and the West Hoist was upgraded to single failure proof in accordance with NUREG-0554.

Under the design change that replaced the crane, it and the west hoist were designed to meet the requirements of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and NUREG-0554, "Single-Failure-Proof cranes for Nuclear Power Plants." There are two heavy load items carried by the spent fuel pool bridge crane, the pool divider gates and the pool covers.

The design includes "seismic restraints" that will prevent X-Y motion of the crane when it is parked. These restraints are pins that will lock the crane end trucks to the west end stops mounted in the floor. An analysis was performed on the crane to calculate displacements, stresses, and forces in the structural members of the crane under static and seismic conditions. Any change in safety classification as a result of the proposed amendment would not change these design requirements.

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Manipulator Crane and Load Cell The manipulator crane is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling water. The bridge spans the refueling cavity and runs on rails set into the floor along the edge of the refueling cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. A long tube with a pneumatic gripper on the end is lowered out of the mast to grip the fuel assembly. The gripper tube is long enough so the upper end is still contained in the mast when the gripper end contacts the fuel.

A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel is transported while inside the mast tube to its new position. The manipulator can lift only one fuel assembly at a time.

The manipulator crane is designed to withstand an earthquake. It is not a requirement of the manipulator crane that it be able to operate during an earthquake nor that it be able to operate after an earthquake. The following requirements were applied to the design:

a. The crane must not drop a fuel assembly if one is in the gripper during an earthquake.

b. The bridge and trolley must not be derailed.

c. No component part of the machine shall break off and fall into the refueling canal.

Any change in safety classification as a result of the proposed amendment would not change these design requirements.

Fuel Transfer System The fuel transfer system is an underwater conveyor car that runs on tracks extending from the refueling canal through the transfer tube and into the fuel transfer canal. The conveyor car receives a fuel assembly in the vertical position from the manipulator crane. The fuel assembly is lowered to a horizontal position for passage through the tube and then is raised to a vertical position in the fuel transfer canal.

During plant operation, the conveyor car is normally stored in the fuel transfer canal. A blind flange is bolted on the refueling canal end of transfer tube to seal the reactor containment. The terminus of the tube outside the containment is closed by a gate valve.

Rod Cluster Control Changing Fixture A fixture is mounted on the refueling cavity wall for removing rod cluster control (RCC) assemblies from spent fuel assemblies and inserting them into new fuel assemblies. The fixture consists of two main components; a guide tube mounted to the wall for containing and guiding the RCC assemblies, and a wheel-mounted carriage for holding the fuel assemblies and positioning fuel assemblies under the guide tube. The guide tube contains a pneumatic gripper on a winch that grips the RCC assembly and lifts it out of the fuel assembly. By repositioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers the RCC element and releases it. The manipulator crane loads and removes the fuel assemblies into and out of the carriage.

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Spent Fuel Assembly Handling Tools The Spent Fuel Handling Tools are used with the Fuel Pool Bridge Crane to move spent fuel assemblies in the spent fuel pools. They are manually operated through a mechanical linkage and use four cam-actuated fingers. The shank of the tools is long enough to prevent raising spent fuel elements to a height in the pool where insufficient radiation shielding is available for personnel.

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements/Criteria 3.1.1 10 CFR 50.2 Definitions The definition of safety-related structures in 10 CFR 50.2 states, in part:

Safety-related structures, systems and components means those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 100.11 of this chapter, as applicable.

3.1.2 10 CFR 50 Appendix A General Design Criteria As stated in Section 2.2 of this enclosure, PINGP was designed and constructed to comply with NSPMs understanding of the intent of the AEC GDC for Nuclear Power Plant Construction Permits, as proposed on July 11, 1967. Therefore, the PINGP Licensing Basis requires conformance to the AEC GDC, as reflected in the PINGP Updated Safety Analysis Report (USAR) Section 1.2. PINGP was not licensed to NUREG-0800, Standard Review Plan.

The following AEC GDCs are applicable to the proposed changes:

Criterion 1 - Quality Standards. Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety functions, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures, and inspection acceptance levels used is required.

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The PINGP systems and components will continue to be classified according to their importance in the prevention and mitigation of accidents which could cause undue risk to the health and safety of the public.

The quality classification standard of ANSI/ANS 58.14 will be applied to the specified fuel handling equipment. The other portions of fuel handling system, such as the fuel transfer tube, will remain classified as safety related consistent with ANSI/ANS 57.1.

Criterion 4 - Sharing of Systems. Reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing.

The only shared components in the fuel handling system are located within the PINGP Auxiliary Building and are associated with the common spent fuel pool. The safety-related fuel transfer tubes that connect each unit to the spent fuel pool are normally flanged closed. During refueling operations in one unit, the fuel transfer canal in the remaining unit will remain isolated to prevent both units from communicating with the spent fuel pool simultaneously. The design class and quality type of the fuel transfer tubes are unaffected by the proposed change.

Therefore a fuel handling accident in one unit will not impact the ability of the remaining unit to initiate an orderly shutdown and cooldown.

Criterion 18 - Monitoring Fuel and Waste Storage Monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures.

No instrumentation associated with spent fuel storage and handling that monitors and alarms the loss of decay heat removal or associated with radiation exposures is being reclassified or will otherwise be affected by this amendment.

3.1.3 10 Industry Codes and Standards ANSI/ANS-58.14-1993 Section 4.2.2 Determine Plant-Level Safety-Related Functions. The plant-level functions (e.g., emergency core cooling) relied upon in the safety analyses of DBEs to prevent or mitigate those DBEs whose consequences could result in potential off-site exposures (as described in the DBE analyses documented in the plant LBD) comparable (i.e.,

greater than or equal to 10%) to the guideline exposures of 10 CFR 100.11 shall be determined. These functions shall be classified safety-related.

Section 5.6.1 Supplemented Grade Items. Items that are not classified safety-related through application of the criteria in 5.1 through 5.5, but to which a significant licensing requirement or commitment applies, shall be classified supplemented grade. This includes those items that are committed by statements in the plant licensing basis documentation to be designated and treated as safety-Page 9 of 12

related items. A significant licensing requirement or commitment is one that is based on an NRC regulation or licensing guidance. Items typically classified supplemented grade include:

(9) Items required to handle, store, or cool new and spent fuel that are subject to the requirements of Criterion 61 of Appendix A of 10 CFR 50. For guidance, see ANSI/ANS-57.1-1992, American National Standard Design Requirements for Light Water Reactor Fuel Handling Systems; ANSI/ANS-57.2-1983, American National Standard Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants; and ANSI/ANS-57.3-1983, American National Standard Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants.

As explained below, this amendment does not affect the classification of the fuel transfer tube and blind flanges which ANSI/ANS-57.1-1992 (Reference 5) designates as safety-related. In accordance with ANSI/ANS-57.1-1992, the equipment to be reclassified within the scope of this amendment may be purchased to commercial codes and standards. ANSI/ANS-57.2-1983 and ANSI/ANS- 57.3-1983 are not applicable to this amendment in that they do not inform the classification of the equipment within the scope of this amendment.

ANSI/ANS 57.1-1992 6.2 Safety Classification and Design Standards. The portion of the transfer tube that serves as part of the primary reactor containment shall be designated Safety Class 2 (SC-2) and shall meet the requirements of American National Standard Containment Isolation Provision for Fuel Systems, ANSI/ANS-56.1-1984. Also, it shall be designed and fabricated in accordance with the American National Standard ANSI/ASME Boiler and Pressure Vessel Code-1992,Section III, Rules for Construction of Nuclear Power Plant Components and a quality assurance program meeting the applicable requirements of American National Standard Quality Assurance Program Requirements for Nuclear Power Plants, ANSI/ASME NQA-1-1989. All other components of the fuel handling system shall be designated non-nuclear safety (NNS) and shall be designated and fabricated to commercial codes and standards.

The proposed changes in classification are consistent with ANSI/ANS 57.1-1992. The affected equipment does not include the fuel transfer tubes. Therefore, the equipment within the scope of the amendment will be classified based on the revised definition of a substantial amount of radioactivity that is consistent with ANSI/ANS 58.14 discussion of potential off-site exposures comparable to the guideline exposures of 10 CFR 100.11.

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3.2 No Significant Hazards Consideration Analysis NSPM has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The fuel handling accident is the only previously evaluated accident for the fuel handling equipment being addressed. The proposed amendment does not result in a significant increase in the probability of an accident because the change in definition of substantial amount of radioactivity as applied to determining the safety classification of the specified fuel handling equipment will not alter the results of fuel handling accidents analyzed in Chapter 14 of the PINGP Updated Safety Analysis Report.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed reclassification of specified refueling handling equipment does not alter existing system interactions or introduce new system interactions. The change will not affect how the specified equipment is operated or maintained. Neither will the change affect the QA requirements for equipment that is required to maintain integrity for seismic category II/I requirements, so no new potential accidents need be postulated as a result of the proposed change.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the USAR.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed amendment revises the current licensing basis to apply a criterion for designating equipment as safety-related that is consistent with the definition of Comparable Off-site Exposures in ANSl/ANS-58.14-1993 for the purposes of equipment safety classification. The proposed amendment is consistent with existing regulatory guidance. The proposed amendment does not reduce compliance with AEC GDC 1. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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Based on the above, NSPM concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.3 Conclusions In 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION A review of the proposed amendment 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 significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

5.0 REFERENCES

1. ANSI/ANS 58.14-1993, Safety and Pressure Integrity Classification Criteria for Light-Water Reactors.

2. PINGP License Amendments 140/131 (Unit 1 and Unit 2, respectively), Cooling Water System Emergency Intake Basis, issued November 4, 1998. (ML022260769)

3. PINGP License Amendments 166/156, Approved Alternate Source Term (AST) methodology for Fuel Handling Accident, issued September 10, 2004.

(ML042430504)

4. PINGP License Amendments 206/193, Implement Alternative Source Term (AST)

Additional Conditions, issued January 22, 2013. (ML112521289)

5. ANSI/ANS 57.1-1992, American National Standard Design Requirements for Light Water Reactor Fuel Handling Systems.

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ENCLOSURE 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request (LAR) to Revise Current Licensing Basis Markup of Updated Safety Analysis Report Page 1 page follows

PRAIRIE ISLAND UPDATED SAFETY ANALYSIS REPOR'T UiSAR Section, 1 Re-vision 34 PaQil! 1"5-3 Quality Assurance Types QA Type I - Those items for which title Quality Assu ranoe Program must assui;e the hig'hest feasible degree of quality standlards colliS'istent with title importanoe of title safety fllmctiion to be performed. This category indudes those items of lhe ptant whidh are essential to tl:le preven~ion of accidents wll'ich cou'ld affect the pub[ic lheaUh and safety by the irelease of quantitiesl1J of rad[oactMty or are requtred in the mitigation of the consequences of such acctde11ts.

QA Type Ill - Those items for whidh tl:l.e Qualiity Assur.mce Rrograrn must ,engender.a hig'tl confictenoe that ~he item wm perform satisractoriily. This c.ategmy in dudes those items,whose t.rilure would not drrecily affect title health and safety oUhe pu'blic, buUhe t.rilure of W:hich could cause severe economic loss or cause the*plant to experience an extended outage.

QA Type Ill - This category inc'luctes all other items not i nclllded in Types I and Ill.

CRITERIION 2 - PERFORMANCE STANDARDS Those s,ystems and components of reactor raailities whidh are essential to the prevention of accidents whiich could affect the pubtic health and safety orto mitigation of their consequenoes shall be designed, fabriicated, and erected to performanoe standards,that will ,enalJle title facility to wilhstand, without toss of the capability to protect the public, the additiomil forces ~hat might be imposed y natural phenomena such as earthquakes, tornadoes, floodmg conditioos, W:indS, ioe, and other local site effects. Toe design lbases so ,establislhed shall reflect (a) approprrate ,consideration of tl:le most severe of these natural phenomena that have been 1recollfed for the*site and the surrounding area and (b) an approprtate margm torwithstand'ing tomes greater tha111 tl:lose recorded to reflect unoertainties about the historical data and their suitability as a basis for design.

A!NSWER Th.e systems and components designated Ciass I in Sedion 12, in conjunction W:ilh administrative conb'ols and analysis, as app,1icable, are designed to withstand, wrnhout loss of ,capability to protect ~he pu'bli.c, title most severe ,environmental phenomena ever experiienced at the site wirth approp:riiate margins included in the design for unoertainties in historical data. Potentiial enviiron:mental haizams are discussed and .analyzed m Sections 2 and 14 of the report and the i11fluence of these lhazards on various aspects of tl:le plant design is discussed in the sections covering the specific s,ysterns and components concerned.. An outrine of lhe design philosophy for C'3ss I systems and components and a listi11g of the .a;pp*callle report sections desariibing the systems and components covered by ~his aiterton are induced m Sedion 12.

C1) A substantial amount of radioaclivil:y is defined as that amount of radioaclive m atelial which would produce radiation levels al the site boundary in excess ot '1'!1, of 10 CFRHI[)

In accordance with Amendments ###/### the classification of the Manipulator Cranes including the load cells, the Spent Fuel Pool Bridge Crane, the Spent Fuel Transfer System (exclusive of the transfer tube/blind flange), the Rod Cluster Control Changing Fixtures, and Spent Fuel Assembly Handling Tool classification is based on a definition of substantial amount of radioactivity of: that amount of radioactive material which would produce radiation levels at the site boundary in excess of 10% of 10CFR100.