L-PI-18-014, License Amendment Request (LAR) to Revise Current Licensing Basis

License Amendment Request (LAR) to Revise Current Licensing Basis

ML18275A370
Person / Time
Site:	Prairie Island
Issue date:	10/02/2018
From:	Sharp S Northern States Power Company, Minnesota, Xcel Energy Inc
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-18-014
Download: ML18275A370 (16)

Text

1717 Wakonade Drive Welch, MN 55089

(}, Xcel Energy*

RESPONSIBLE BY NATURE

• 800.895.4999 xcelenergy.com October 2, 2018 L-PI-18-014 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 Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), is submitting 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.

The changes would allow certain fuel handling equipment to be reclassified from QA Type I (safety-related) to QA Type III (non-safety related). provides a description and assessment of the proposed changes. Enclosure 2 provides the marked up USAR pages that reflect the proposed changes.

Approval of the proposed amendment is requested within one year of the date 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.

Document Control Desk Page 2 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 Ot:.-r~~ t., zo,e.

d:::!a?

Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota ENCLOSURES:

1. Evaluation of Proposed Change

2. Markup of Updated Safety Analysis Report Pages cc: Administrator, Region Ill, 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. DETAILED DESCRIPTION 2.1 Proposed Change 2.2 Background 2.3 Reason for the Proposed Change 2.4 Description of Equipment

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

4. ENVIRONMENTAL EVALUATION

5. REFERENCES Page 1 of 11

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 for clarification and changes to the Prairie Island Nuclear Generating Plant (PINGP) licensing basis regarding the quality assurance type and design classification of fuel handling equipment. The changes would allow certain fuel handling equipment to be reclassified from QA Type I (safety-related) to QA Type III (non-safety related).

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 and by making a corresponding change to USAR Page 12.2-4 that describes the correlation between Design Class and QA Type.

Draft changes to the USAR are provided in Enclosure 2.

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:

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 Page 2 of 11

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, Safety and Pressure Integrity Classification Criteria for Light Water Reactors (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 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.

Page 3 of 11

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.

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-Page 4 of 11

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 QA Type I, safety related.

This amendment would revise the PINGP USAR regarding specific fuel handling equipment to relax the PINGP-specific classification scheme to allow them to be classified as QA Type III, non-safety related. No change to allow relaxation of the Design Class of components is proposed in this amendment. The affected equipment includes:

Manipulator Cranes, including the Load Cells Spent Fuel Pool Bridge Crane Auxiliary Building 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 QA Type for the affected 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. Although, by analysis, the failure of any of the affected equipment could result in the uncontrolled release of radioactivity in excess of 1% of limits described in 10 CFR 100 (current PINGP license basis definition of QA Type I), the fuel handling accident analysis shows the dose consequences are below the 10% of 10 CFR 100.11 criterion of ANSI/ANS-58.14-1993.

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 QA Type I, 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 Auxiliary Building Crane The Auxiliary Building Crane is used for handling spent fuel casks and has been upgraded to be in compliance with Section 5.1.6 and Appendix C of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants. This upgrade to the auxiliary building crane has been made to provide a handling system for handling heavy loads in the spent fuel pool area that satisfies the single-failure-proof guidelines of Section 5.1.6 of NUREG-0612, and thus eliminates the need to analyze the effects of drops of heavy loads per the evaluation criteria of Section 5.1 of NUREG-0612. The original crane was designed to handle a load of 125 Tons. The main hoist capacity has not changed with the upgrade, but the auxiliary hoist capacity has been changed Page 5 of 11

from 25 Tons to 15 Tons. The single-failure-proof features have not been incorporated into the design of the auxiliary hoist. All crane structural members have been designed to withstand impact loads per applicable specifications. A seismic evaluation has been performed for the loaded condition.

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 West Hoist of the Spent Fuel Pool Bridge Crane has been upgraded to single failure proof.

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.

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 Page 6 of 11

lowers the RCC element and releases it. The manipulator crane loads and removes the fuel assemblies into and out of the carriage.

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 Page 7 of 11

codes, standards, quality assurance programs, test procedures, and inspection acceptance levels used is required.

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 portions of the fuel handling system proposed to be designated as QA Type III (non-safety related) will be classified in accordance with NRC accepted codes and standards. The other portions of fuel handling system, such as the fuel transfer tube, will remain classified as QA Type I (safety related).

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.

Page 8 of 11

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-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 as non-safety related.

Page 9 of 11

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 proposed amendment does not result in a significant increase in the probability of an accident because the reclassification of the specified fuel handling equipment from safety-related to non-safety related will not affect the design, testing, operation or maintenance of the affected equipment.

The change in equipment classification 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.

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 quality assurance type. 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.

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.

Page 10 of 11

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 Pages 2 pages follow

PRAIRIE ISLAND UPDATED SAFETY AN:AL. YSIS REPOR'T USAR Sl!diion, 1 Re-wsion:U Pallil! 1laS.:3 Qualify Assura:111ce Types QA.Type - Titmse items for which title Quatity Assurance Program must assufie the

~ig'.l'lest feasible degree of QUlality standards co11Sistent with 1il'le importance of tl'le safety function to be performed. This category indudes those items of the plant wh*dti are essentEal to tl:le prevention of acctdents *which could affect the pub[ic l 'lealtl'l and safety by the Irelease of quaTI1tities[1l of radtoactMty: or aie iequtred in the mitigation of the 1

ronsequooces of such acctdents.

QA.Type Ill - Those items for whic'l'l the Qualrly Assura:nce ~rograrn must ,engerider a l'lig'.l'l ronfidenoe that tl'le item wm perform satisfacto:riily. This category in dudes those items whose fa'.il re*would not dlrectly affect the health and safety of the pubr c, but the fa'.ilure of W:hich could cause severe economic loss or cauiSle th:e*plant to experience an extended outage.

QA.Type 1111 - This category indludes all other items 11101 i nc udecl i n Types .and Ill.

CRITEIRll:ONI 2 - PEIIRFORMANCIE STANDAIRDS Those systems and components of reactor faoilifies whidh are essential to the prevention of acciclents wh ich could affect the pubtic lhealth and safety orto mitigation of their consequences shall be*destgned, fab:liicated, and erected to peliformanoe standards,tl:lal will ,enab:le the faci[ity to *withs'land, Wlllhout oss of the capability to protect the pub:1ic, the add'itional forces that m'igllt be lmposecl by natural pheriomena such as earthqualffis, tornadoes, flooding conditions, winds, ioe, and ofher local site effects. The design lbases so ,established shrall fieflect (a) app.ropnale ,oonsideration of tl:le most severe of th:ese natural plleno:mena that have lbeen irecolided for the site and tl:le surrounding area andl (b) ,an appropriate margin fo:rwithslarnfing foroes greater than tl:lose recorded to reflect unoertainties abol.lt the historical data and their suitability as a basis for design.

ANSWER The*systems arnd components ,designated! C~ass 111 ln Sectio111 12, in conjunction w;ith administrative controls and analysis, as app.1icable, are designed to wruistand, witholJlt loss of capabi[ity to protect the pu'.blic, the most sevefie environme111tal pheriomena ever 1

experiienood at the site with appmpriiate ma1gins indluded in the design fo r oooerlainlies in historical ,di.ala.. Pote11t;a1 environmental hazards are d'iscussecl and ,aTilalyzed in 1

Sections 2 ,aTild 4 of ~tie report a:rnd ~tie influence of these l 'lazards on various aspects of tl:le 1Planrt design is discussed iru the sectrons covering the specific systems andl rornpo.nrents concerned.. An outline of ftle design ptiilosophy for Crass I systems andl rornpo111ents and a listi11g of the .app~callle report sections desoriibing the systems and rompo.nrents covered by this aiterion afie induded in Section 1.2.

11) A substantial amount of radioaclivil:y is defined as that a mount of radioactive m aterial which -,ukl produce. radiation levels at the site boundary in excess ct 11'!1, al 10 CFRHI[)

In accordance with Amendments ###/### the Manipulator Cranes including the load cells, the Spent Fuel Pool Bridge Crane, the Auxiliary Building 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 are classified as QA Type III.

PRAIRIE ISLAND UPDATED SAFETY AN:AL. YSIS REPOR'T USAR Section *112 Revision 36P P.age- 12:2-4 Ground Floor

1. lJnit 1 Emergency dtesel generators 2 . !Batteries

3. Aiir compressors

4. At1X1"1ia:ry feedwater pumps.

5. Codling water pipes The*main feedwater pumps which are Glass 1111 are located in Olass IW portions of the Tmrbine IB'lli ding.

Mezzanine Floor

1. lJnit 1 4160V a11d 480V safeguards switchgear The above Class I designation applies to 1he walls, floors, cetlings, sbuc1mal support and foundations of s'ln.Jciures that [isolate, support or are assoaia1ed wilh th,e protection of Olass Ill ,equipment The quality 0lassltica.lion for a[I IDeslgn Class I and Design C!a.ss I"' components [is1ed ITT l":able 122-1 are Type 1 and Type 3, respectively.

12.2.1.2 Design Codes 1

The*destgn and consbuclion of lhilS plant lhas been ITT acrorda:111ce wiith 1he folfOwing codes .as .a.pp[icallle:

a. American Conarete Institute Codes; ACI 31-~ , ACI 301-66, ACI 349-tl5 and ofher sections of 1he ACI Codes as ap;plicabte.

b. American Institute of Steel Consbuclion *Specilica~ion for the Design, Fabrication and Eirecton o:f Strucillral Steel Buildings,* 1953 6dition.

~Modlticalions to the plant since origITTal construction have used more ieOOJ11t ediitions.),

c. American WeldITTg Soctety Code D 1.U '"standards for Arc andl Gas Welding1ITT Buitding Construction."'

,d_ lntemational Conferrance of Eluitding Officials *unlform Building Code,n 961 Edition.

The Auxiliary Building Crane, Manipulator Cranes and the Spent Fuel Pool Crane are QA Type 3 in accordance with Amendments ###/###.