ML22061A206

From kanterella
Jump to navigation Jump to search

Issuance of Amendments TSTF-471, Rev. 1, TSTF-571-T, and Administrative Changes to Technical Specification Section 5.0
ML22061A206
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 04/01/2022
From: Robert Kuntz
Plant Licensing Branch III
To: Domingos C
Northern States Power Company, Minnesota
Kuntz R
References
EPID L-2021-LLA-0069
Download: ML22061A206 (70)
v • d • e


Text

April 1, 2022 Mr. Christopher P. Domingos Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -

ISSUANCE OF AMENDMENTS RE: TSTF-471, REVISION 1 ELIMINATE USE OF TERM CORE ALTERATIONS IN ACTIONS AND NOTES, TSTF-571-T, REVISE ACTIONS FOR INOPERABLE SOURCE RANGE NEUTRON FLUX MONITOR, AND ADMINISTRATIVE CHANGES TO TECHNICAL SPECIFICATION SECTION 5.0 (EPID L-2021-LLA-0069)

Dear Mr. Domingos:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 238 to Renewed Facility Operating License No. DPR-42 and Amendment No. 226 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated April 19, 2021 as supplemented by letter dated September 30, 2021.

The amendments modify the TSs to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-471, Revision 1, Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes, TSTF-571-T, Revise Actions for Inoperable Source Range Neutron Flux Monitor, and make an administrative change to reformat numbering of TSs Section 5.0 and remove unused pages.

C. Domingos A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosures:

1. Amendment No. 238 to DPR-42
2. Amendment No. 226 to DPR-60
3. Safety Evaluation cc: Listserv

NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 Renewed License No. DPR-42

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated April 19, 2021, as supplemented by letter dated September 30, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 238, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Scott Scott P. Wall P.Date:

Wall 2022.04.01 13:00:05 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: April 1, 2022

NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 Renewed License No. DPR-60

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated April 19, 2021, as supplemented by letter dated September 30, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Scott P. Wall Scott P. Wall Date: 2022.04.01 12:59:39 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: April 1, 2022

ATTACHMENT TO LICENSE AMENDMENT NOS. 238 AND 226 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicate the areas of change.

Renewed Facility Operating License No. DPR-42 REMOVE INSERT Page 3 Page 3 Renewed Facility Operating License No. DPR-60 REMOVE INSERT Page 3 Page 3 Technical Specifications Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT REMOVE INSERT REMOVE INSERT 1.1-2 1.1-2 5.0-9 5.5-3 5.0-28 5.5-18 3.8.2-2 3.8.2-2 5.0-10 5.5-4 5.0-29 5.5-19 3.8.2-3 3.8.2-3 5.0-11 5.5-5 5.0-30 5.5-20 3.8.5-2 3.8.5-2 5.0-12 5.5-6 5.0-31 5.5-21 3.8.8-1 3.8.8-1 5.0-13 5.5-7 5.0-32 5.6-1 3.8.8-2 3.8.8-2 5.0-14 5.5-8 5.0-33 5.6-2 3.8.10-1 3.8.10-1 5.0-15 5.5-9 5.0-34 5.6-3 3.8.10-2 3.8.10-2 5.0-16 5.5-10 5.0-35 5.6-4 3.9.1-1 3.9.1-1 5.0-17 -------- 5.0-36 5.6-5 3.9.3-1 3.9.3-1 5.0-18 -------- 5.0-37 5.6-6 3.9.3-2 3.9.3-2 5.0-19 -------- 5.0-38 5.6-7 5.0-1 5.1-1 5.0-20 -------- 5.0-39 5.6-8 5.0-2 5.2-1 5.0-21 5.5-11 5.0-40 5.6-9 5.0-3 5.2-2 5.0-22 5.5-12 5.0-41 5.7-1 5.0-4 5.2-3 5.0-23 5.5-13 5.0-42 5.7-2 5.0-5 5.3-1 5.0-24 5.5-14 5.0-43 5.7-3 5.0-6 5.4-1 5.0-25 5.5-15 5.0-44 5.7-4 5.0-7 5.5-1 5.0-26 5.5-16 5.0-45 5.7-5 5.0.8 5.5-2 5.0-27 5.5-17

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 238, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-42 Amendment No. 238

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purposes of volume reduction and decontamination.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 226

Definitions 1.1 1.1 Definitions (continued)

CHANNEL A CHANNEL CHECK shall be the qualitative assessment, by CHECK observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL A COT shall be the injection of a simulated or actual signal into OPERATIONAL the channel as close to the sensor output as practicable to verify TEST (COT) OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE The COLR is the unit specific document that provides cycle specific OPERATING parameter limits for the current reload cycle. These cycle specific LIMITS parameter limits shall be determined for each reload cycle in REPORT accordance with Specification 5.6.5. Plant operation within these (COLR) limits is addressed in individual Specifications.

DOSE DOSE EQUIVALENT I-131 shall be that concentration of I-131 EQUIVALENT (microcuries/gram) that alone would produce the same dose I-131 when inhaled as the combined activities of isotopes I-131, I-132, I-133, I-134 and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake And Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 1.1-2 Unit 2 - Amendment No. 226

AC Sources - Shutdown 3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. Required path inoperable. ------------------NOTE---------------

Enter applicable Conditions and Required Actions of LCO 3.8.10, with one required train de-energized as a result of Condition A.

A.1 Suspend movement of Immediately irradiated fuel assemblies.

AND A.2 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND A.3 Initiate action to restore Immediately required path to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.2-2 Unit 2 - Amendment No. 226

AC Sources - Shutdown 3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Suspend movement of Immediately inoperable. irradiated fuel assemblies.

AND B.2 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND B.3 Initiate action to restore Immediately required DG to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.2-3 Unit 2 - Amendment No. 226

DC Sources-Shutdown 3.8.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One required DC B.1 Suspend movement of Immediately electrical power irradiated fuel assemblies.

subsystem inoperable for reasons other than AND Condition A.

B.2 Suspend operations Immediately OR involving positive reactivity additions that Required Action and could result in loss of associated Completion required SDM or boron Time of Condition A not concentration.

met.

AND B.3 Initiate action to restore Immediately required DC electrical power subsystems to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.5-2 Unit 2 - Amendment No. 226

Inverters-Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Inverters-Shutdown LCO 3.8.8 One Reactor Protection Instrument AC inverter shall be OPERABLE.

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

NOTE-------------------------------------------------

LCO 3.0.3 not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required inverter A.1 Suspend movement of Immediately inoperable. irradiated fuel assemblies.

AND A.2 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND A.3 Initiate action to restore Immediately required inverter to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.8-1 Unit 2 - Amendment No. 226

Inverters-Shutdown 3.8.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage and alignment to In accordance with required Reactor Protection Instrument AC panel. the Surveillance Frequency Control Program Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.8-2 Unit 2 - Amendment No. 226

Distribution Systems-Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems-Shutdown LCO 3.8.10 The necessary portion of safeguards AC, DC, and Reactor Protection Instrument AC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

NOTE--------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately safeguards AC, DC, or supported required Reactor Protection feature(s) inoperable.

Instrument AC electrical power distribution OR subsystems inoperable.

A.2.1 Suspend movement of Immediately irradiated fuel assemblies.

AND Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.10-1 Unit 2 - Amendment No. 226

Distribution Systems-Shutdown 3.8.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND A.2.3 Initiate actions to restore Immediately required safeguards AC, DC, and Reactor Protection Instrument AC electrical power distribution subsystems to OPERABLE status.

AND A.2.4 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.8.10-2 Unit 2 - Amendment No. 226

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System and the refueling cavity shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 6.

NOTE--------------------------------------

Only applicable to the refueling cavity when connected to the RCS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A.1 Suspend positive reactivity Immediately not within limits. additions.

AND A.2 Initiate action to restore Immediately boron concentration to within limits.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.9.1-1 Unit 2 - Amendment No. 226

Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Nuclear Instrumentation LCO 3.9.3 Two core subcritical neutron flux monitors shall be OPERABLE.

AND One core subcritical neutron flux monitor audible count rate circuit shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required core A.1 Suspend positive reactivity Immediately subcritical neutron flux additions.

monitor inoperable.

AND

NOTE ----------------

Fuel assemblies, sources, and reactivity control components may be moved if necessary to restore an inoperable core subcritical neutron flux monitor or to complete movement of a component to a safe condition.

A.2 Suspend movement of Immediately fuel, sources, and reactivity control components within the reactor vessel.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.9.3-1 Unit 2 - Amendment No. 226

Nuclear Instrumentation 3.9.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Two required core B.1 Initiate action to restore Immediately subcritical neutron flux one core subcritical monitors inoperable. neutron flux monitor to OPERABLE status.

AND B.2 Perform SR 3.9.1.1. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Required core subcritical C.1 Initiate action to isolate Immediately neutron flux monitor unborated water sources.

audible count rate circuit inoperable.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 3.9.3-2 Unit 2 - Amendment No. 226

Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The shift supervisor shall be responsible for the control room command function. During any absence of the shift supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active senior reactor operator (SRO) license shall be designated to assume the control room command function. During any absence of the shift supervisor from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or reactor operator (RO) license shall be designated to assume the control room command function.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.1-1 Unit 2 - Amendment No. 226

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Safety Analysis Report (USAR) or Quality Assurance Topical Report;
b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;
c. A corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and
d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.2-1 Unit 2 - Amendment No. 226

Organization 5.2 5.2 Organization (continued) 5.2.2 Plant Staff The plant staff organization shall include the following:

a. An operator to perform non-licensed duties shall be assigned to each reactor containing fuel and one additional operator to perform non-licensed duties shall be assigned when either or both reactors are operating in MODES 1, 2, 3, or 4.
b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.f for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
c. An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
d. Not Used.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.2-2 Unit 2 - Amendment No. 226

Organization 5.2 5.2 Organization 5.2.2 Plant Staff (continued)

e. The operations manager or assistant operations manager shall hold an SRO license. In addition, the duty shift manager shall hold an SRO license.
f. In MODES 1, 2, 3, and 4, the shift technical advisor shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.2-3 Unit 2 - Amendment No. 226

Plant Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Plant Staff Qualifications 5.3.1 Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the following:

The radiation protection manager shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1, September 1975.

In addition, the operations manager shall be qualified as required by TS 5.2.2.e.

The licensed operators shall comply only with the requirements of 10 CFR 55.

5.3.2 For the purpose of 10 CFR 55.4, a licensed SRO and a licensed RO are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m).

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.3-1 Unit 2 - Amendment No. 226

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;
c. Quality control for effluent and environmental monitoring;
d. Not used; and
e. All programs specified in Specification 5.5.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.4-1 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Monitoring, and Radioactive Effluent Reports required by Specification 5.6.2 and Specification 5.6.3.

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and
2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;
b. Shall become effective after the approval by a member of plant management designated by the plant manager; and Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-1 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) (continued)

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed. The date (i.e., month and year) the change was implemented shall be indicated.

5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practical. The systems include portions of the Residual Heat Removal and Safety Injection Systems. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and
b. Integrated leak test requirements for each system at refueling cycle intervals or less.

5.5.3 Post Accident Sampling This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.

The program shall include the following:

a. Training of personnel;
b. Procedures for sampling and analysis; and
c. Provisions for maintenance of sampling and analysis equipment.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-2 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable.

This program shall allocate releases equally to each unit. The liquid radwaste treatment system, waste gas treatment system, containment purge release vent, and spent fuel pool vent are shared by both units. Experience has also shown that contributions from both units are released from each auxiliary building vent. Therefore, all releases will be allocated equally in determining conformance to the design objectives of 10 CFR 50, Appendix I.

The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402;
c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-3 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

Determination of projected dose contributions for radioactive effluents in accordance with the methodology in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days from the liquid effluent releases would exceed 0.12 mrem to the total body or 0.4 mrem to any organ; or from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose, 0.8 mrad for beta air dose, or 0.6 mrem organ dose;
g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:
1. for noble gases: a dose rate < 500 mrem/yr to the whole body and a dose rate < 3000 mrem/yr to the skin, and
2. for iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ;
h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-4 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the USAR, Section 4.1.4, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT or PT) of exposed surfaces of the removed flywheels may be conducted at 20 intervals.

5.5.7 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-5 Unit 2 - Amendment No. 226 Corrected by letter dated June 21, 2016

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Inservice Testing Program (continued)

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Semiquarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified 2 years or less in the Inservice Testing Program for performing inservice testing activities.
c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and
d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-6 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-7 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued)

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.
3. The operational LEAKAGE performance criterion is specified in LCO 3.4.14, RCS Operational LEAKAGE.
c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2 and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-8 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-9 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
e. Provisions for monitoring operational primary to secondary LEAKAGE.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-10 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of the Control Room Special Ventilation System (CRSVS), Auxiliary Building Special Ventilation System (ABSVS), and Shield Building Ventilation System (SBVS) at least once each 24 months.

Demonstrate for the ABSVS, SBVS, and CRSVS systems that:

a. An inplace DOP test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% (for DOP, particles having a mean diameter of 0.7 microns);
b. A halogenated hydrocarbon test of the inplace charcoal adsorber shows a penetration and system bypass < 0.05% (SBVS not applicable);
c. A laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than: 1) 10% penetration for ABSVS, and 2) 2.5% penetration for the CRSVS when tested in accordance with ASTM D3803-1989 at a temperature of 30C and 95% relative humidity (RH);
d. The pressure drop across the combined HEPA filters and the charcoal adsorbers (SBVS not applicable to charcoal adsorbers) is less than 6 inches of water at the system flowrate + 10%; and
e. A laboratory test of a sample of the charcoal adsorber shall have filter test face velocities greater than or equal to the following values for each system: 1) 54 fpm for the CRSVS, and 2) 72 fpm for the ABSVS.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-11 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of oxygen in the waste gas holdup system and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria;
b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 78,800 Curies of noble gas (considered as dose equivalent Xe-133); and
c. A surveillance program to ensure that the quantity of radioactivity contained in each of the following tanks shall be limited to 10 Curies, excluding tritium and dissolved or entrained noble gases:

Condensate storage tanks Outside temporary tanks The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance Frequencies.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-12 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. Acceptability of new fuel oil shall be determined prior to addition to the safeguards storage tanks.

Testing of diesel fuel oil stored in the safeguards storage tanks shall be performed at least every 31 days.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.

5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews;
b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
1. a change in the TS incorporated in the license, or
2. a change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59;
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR; and Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-13 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)

d. Proposed changes that meet the criteria of Specification 5.5.12 b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

5.5.13 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
d. Other appropriate limitations and remedial or compensatory actions.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-14 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-15 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exception:
1. Unit 1 and Unit 2 (steam generator (SG) replacement commencing Fall 2013) are excepted from post-modification integrated leakage rate testing requirements associated with SG replacement.
b. The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, Pa, of 46 psig.
c. The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.15% of primary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.06% of primary containment air weight per day at pressure Pa. For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.006% of primary containment air weight per day at pressure Pa.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-16 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program (continued)

d. Leakage Rate acceptance criteria are:
1. Primary containment leakage rate acceptance criterion is < 1.0 La.

Prior to unit startup, following testing in accordance with the program, the combined leakage rate acceptance criteria are < 0.60 La for all components subject to Type B and Type C tests and

< 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is < 0.05 La when tested at > 46 psig.

b) For each door intergasket test, leakage rate is < 0.01 La when pressurized to > 10 psig.

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.15 Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance of the 125V plant safeguards batteries and service building batteries, which may be used instead of the safeguards batteries during shutdown conditions in accordance with manufacturers recommendations, as follows:

a. Actions to restore battery cells with float voltage < 2.13 V will be in accordance with manufacturers recommendations, and
b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-17 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Special Ventilation System (CRSVS),

CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.
b. Requirements for maintaining the CRE boundary in its design conditions including configuration control and preventive maintenance.
c. Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
d. Licensee controlled programs that will be used to verify the integrity of the CRE boundary. Conditions that generate relevant information from those programs will be entered into the corrective action process and shall be trended and used as part of the periodic assessments of the CRE boundary.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-18 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air in-leakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered in-leakage measured by the testing described in paragraph c.

The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analysis of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions of the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability and determining CRE unfiltered in-leakage as required by paragraph c.

5.5.17 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.
c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-19 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.18 Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."

The program shall include the following:

a. The RICT may not exceed 30 days;
b. A RICT may only be utilized in MODES 1 and 2;
c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.
1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.
3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-20 Unit 2 - Amendment No. 226

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Risk Informed Completion Time Program (continued)

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.5-21 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Not Used.

5.6.2 Annual Radiological Environmental Monitoring Report

NOTE----------------------------------------

A single submittal may be made for the plant. The submittal should combine sections common to both units.

The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Monitoring Report shall include summarized and tabulated results, in the format of Regulatory Guide 4.8, December 1975, of all radiological environmental samples taken during the report period. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

The report shall also include the following: a summary description of the radiological environmental monitoring program; a map of sampling locations keyed to a table giving distances and directions from the reactor site; and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-1 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Report

NOTE-----------------------------------------

A single submittal may be made for the plant. The submittal shall combine sections common to both units.

The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Not Used.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

TS 2.1.1, Reactor Core SLs; LCO 3.1.1, SHUTDOWN MARGIN (SDM);

LCO 3.1.3, Isothermal Temperature Coefficient (ITC);

LCO 3.1.5, Shutdown Bank Insertion Limits; LCO 3.1.6, Control Bank Insertion Limits; LCO 3.1.8, PHYSICS TESTS Exceptions - MODE 2; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-2 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

LCO 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor F LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD);

LCO 3.3.1, Reactor Trip System (RTS) Instrumentation Overtemperature T and Overpower T Parameter Values for Table 3.3.1-1; LCO 3.4.1, RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits; and LCO 3.9.1, Boron Concentration.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. NSPNAD-8101-A, Qualification of Reactor Physics Methods for Application to PI Units (latest approved version);
2. NSPNAD-8102-PA, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units (latest approved version);
3. NSPNAD-97002-PA, Northern States Power Companys Steam Line Break Methodology, (latest approved version);
4. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology;
5. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code;
6. Deleted;
7. WCAP-10924-P-A, Westinghouse Large Break LOCA Best Estimate Methodology; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-3 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II;
9. WCAP-13677-P-A, 10 CFR 50.46 Evaluation Model Report:

W-COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOTM Cladding Options;

10. NSPNAD-93003-A, Transient Power Distribution Methodology, (latest approved version);
11. NAD-PI-003, Prairie Island Nuclear Power Plant Required Shutdown Margin During Physics Tests;
12. NAD-PI-004, Prairie Island Nuclear Power Plant F QW (Z) Penalty With Increasing FCQ (Z) / K(Z) Trend;
13. WCAP-10216-P-A, Revision 1A, Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification;
14. WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions;
15. WCAP-11397-P-A, Revised Thermal Design Procedure;
16. WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report;
17. WCAP-7588 Rev. 1-A, An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-4 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

18. WCAP-7908-A, FACTRAN - A FORTRAN IV Code for Thermal Transients in a UO2 Fuel Rod;
19. WCAP-7907-P-A, LOFTRAN Code Description;
20. WCAP-7979-P-A, TWINKLE - A Multidimensional Neutron Kinetics Computer Code;
21. WCAP-10965-P-A, ANC: A Westinghouse Advanced Nodal Computer Code;
22. WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event;
23. WCAP-11596-P-A, Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores;
24. WCAP-12910 Rev. 1-A, Pressurizer Safety Valve Set Pressure Shift;
25. WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis;
26. WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses;
27. WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM);
28. Caldon Engineering Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM System; Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-5 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

29. Caldon Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM or LEFM CheckPlus System;
30. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report;
31. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM;
32. Commencing Unit 1 Cycle 30 and Unit 2 Cycle 30, this reference shall be used in lieu of reference 23: WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004;
33. Commencing Unit 1 Cycle 30 and Unit 2 Cycle 30, this reference shall be used in lieu of reference 23: WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007;
34. WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Technical Specifications, February 2019.
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-6 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.7, RCS Loops - MODE 5, Loops Filled; LCO 3.4.10, Pressurizer Safety Valves; LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature; LCO 3.4.13, Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature; and LCO 3.5.3, ECCS - Shutdown.

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (includes any exemption granted by NRC to ASME Code Case N-514).

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-7 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,
b. Degradation mechanisms found,
c. Nondestructive examination techniques utilized for each degradation mechanism,
d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
e. Number of tubes plugged during the inspection outage for each degradation mechanism,
f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and
g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-8 Unit 2 - Amendment No. 226

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.8 EM Report When a report is required by Condition C or I of LCO 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.6-9 Unit 2 - Amendment No. 226

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied in place of the controls required by paragraph 10 CFR 20.1601(a) and (b) of 10 CFR 20:

5.7.1 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
b. Access to, and activities in each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously displays radiation dose rates in the area; or
2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint; or Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.7-1 Unit 2 - Amendment No. 226

High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates (continued)

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area; or
4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.7-2 Unit 2 - Amendment No. 226

High Radiation Area 5.7 5.7 High Radiation Area (continued) 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or their designee.
2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint; or Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.7-3 Unit 2 - Amendment No. 226

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source (continued)

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area; or
3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.
4. In those cases where options (2) and (3) above are impractical or determined to be inconsistent with the As Low As is Reasonably Achievable principle, a radiation monitoring device shall be used that continuously displays radiation dose rates in the area.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.7-4 Unit 2 - Amendment No. 226

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source (continued)

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
f. Such individual areas that are located within a larger area where no enclosure exists for the purpose of locking and where no enclosure can be reasonably constructed around the individual area, that individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a flashing light shall be activated at the area as a warning device.

Prairie Island Unit 1 - Amendment No. 238 Units 1 and 2 5.7-5 Unit 2 - Amendment No. 226

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 238 TO RENEWED FACILITY 1OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 226 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated April 19, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21109A385), as supplemented by letter dated September 30, 2021 (ADAMS Accession No. ML21273A191), Northern States Power Company, doing business as Xcel Energy (the licensee) submitted a license amendment request (LAR) for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Prairie Island). The LAR proposes to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-471, Revision 1, Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes, and makes an administrative change to reformat the page numbering of Technical Specifications (TSs) Section 5.0 and remove unused pages. The LAR also proposes to adopt TSTF-571-T, Revise Actions for Inoperable Source Range Neutron Flux Monitor in response to a U. S. Nuclear Regulatory Commission (NRC or Commission) letter dated October 4, 2018, to the TSTF (ADAMS Accession No. ML17346A587).

TSTF-471 removes the term CORE ALTERATIONS from the TSs. The following definition is included in the Prairie Island TSs, Section 1.1: CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. Several TSs require suspension of CORE ALTERATIONS, which applies an operational burden with no corresponding safety benefit over potential alternative actions.

The supplemental letter dated September 30, 2021, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 15, 2021 (86 FR 31743).

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 Proposed Changes The LAR proposed administrative changes to TS Section 5.0. The proposed changes to TS Section 5.0 reformat the page numbering and remove unused pages to ease future changes.

These changes do not change any text in TS Section 5.0.

The LAR proposed to remove the term CORE ALTERATION from the following TSs consistent with TSTF-471:

3.8.2, AC[alternating current] Sources - Shutdown 3.8.5, DC [direct current] Sources - Shutdown 3.8.8, Inverters - Shutdown" 3.8.10, Distribution Systems - Shutdown" 3.9.1, Boron Concentration 3.9.3, Nuclear Instrumentation The LAR proposed to revise TS 3.9.3, Required Action A.2, with a note to add allowance to move fuel assemblies, sources, and reactivity control components, if necessary, to restore operability or to complete movement of a component to a safe condition consistent with TSTF-571-T.

2.1 Regulatory Requirements Section 182a of the Atomic Energy Act (the Act) requires applicants for nuclear power plant operating licenses to include TS as part of the license. The TS ensures the operational capability of structures, systems and components that are required to protect the health and safety of the public. The NRCs regulatory requirements related to the content of the TS are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36.

Regulation 10 CFR 50.36(b) states, in part, that: The technical specifications will be derived from the analyses and evaluation included in the safety evaluation report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information].

The Commission may include such additional TSs as the Commission finds appropriate.

Regulation 10 CFR 50.36(c) specifies the categories of items required to be in TSs. One such category is limiting conditions for operation (LCOs). LCOs are defined in 10 CFR 50.36(c)(2)(i) as the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee must shut down the reactor or follow any remedial or Required Action (e.g., testing, maintenance, or repair activity) permitted by the TSs until the condition can be met. The remedial actions (i.e., ACTIONS) associated with an LCO contain conditions that typically describe the ways in which the requirements of the LCO may fail to be met.

Appendix A to 10 CFR 50 describes general design criteria (GDC) that establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.

NUREG-1431, Revision 5, Standard Technical Specifications [STS]- Westinghouse Plants, (ADAMS Accession No. ML21259A155) contains a template for Westinghouse plants that the NRC staff has found to include the necessary information for compliance with 10 CFR 50.36.

The STS is revised through approval of various TSTF travelers, which licensees may adopt by showing that they meet the technical basis described in the model safety evaluations corresponding to each traveler. The applicable information can be found for the TSTF travelers that Prairie Island proposed to implement in the following documents:

TSTF-471-A, Revision 1 (ADAMS Accession No. ML19101A215), was approved for use by the NRC staff in a letter dated December 7, 2006 (ADAMS Accession No. ML062860320).

TSTF-471 corrected an oversight related to implementation of TSTF-51 which was approved by NRC letter dated November 1, 1999 (ADAMS Accession No. ML993190284). TSTF-51 was intended to delete any reference to the term CORE ALTERATIONS in the TS. Following TSTF-51 implementation, the industry noted other places in the TS where the term continued to exist. TSTF-471 (a) eliminated the term CORE ALTERATIONS from these remaining references, and (b) removed the requirement to Suspend CORE ALTERATIONS from several specific STS sections on electrical and refueling operations related to the plant refueling mode of operations.

TSTF-571-T, Revision 0, insofar as it relates to the revised actions for inoperable source range neutron flux monitor (pages 6 to 30 of the document in ADAMS Accession No. ML18221A561).

TSTF-571-T revises the REQUIRED ACTIONS for an inoperable source range neutron flux monitor to prohibit the movement of fuel assemblies, sources, and reactivity control components when a core subcritical neutron flux monitor is inoperable. A provision is included to allow such movement if it is needed to repair the core subcritical neutron flux monitor.

3.0 TECHNICAL EVALUATION

3.1 Accidents and Transients When the reactor vessel head is unbolted and removed, core alterations take place during operating Mode 6 (refueling operation). There are only two accidents considered during Mode 6. These are: (1) a fuel-handling accident (FHA), and (2) a boron dilution accident. An FHA is initiated by the dropping of an irradiated fuel assembly, either in the containment or in the spent fuel pool. There are no mitigation actions, except for taking credit for ventilation systems to reduce the dose consequences. Thus, the suspension of core alterations, except for suspension of movement of irradiated fuel, will not prevent or impair the mitigation of an FHA.

The analysis for an FHA assumes that a fuel assembly is dropped during fuel handling in the containment or the spent fuel pool. Interlocks and procedural and administrative controls make such an event highly unlikely. However, if an assembly were damaged to the extent that one or more fuel rods were broken, the accumulated fission product gases and iodines in the fuel element gap would be released to the surrounding water. Release of the solid fission products in the fuel would be negligible because of the low fuel temperature during refueling, which greatly limits their diffusion.

A boron dilution accident is initiated by a dilution source that results in the boron concentration dropping below the value required to maintain the shutdown margin. TS 3.9.1, Boron Concentration, applies in Mode 6, and the refueling boron concentration limit is specified in the core operating limits report (COLR). This accident is mitigated by stopping the dilution. The

suspension of core alterations has no effect on the mitigation of a boron dilution accident. Also, the control rods or fuel do not affect the initial conditions of a boron dilution accident.

3.2 TS Changes 3.2.1 TS 1.1 Definitions The licensee proposed to remove the definition for CORE ALTERATIONS from TS Section 1.1.

Since a TS Definition has no actions or surveillance requirements, the removal of this definition to coincide with the deletion from other TS sections is acceptable.

3.2.2 TSs 3.8.2, 3.8.5, 3.8.8, and 3.8.10 If a required train of electrical power (AC, DC, Inverters, or Distribution Systems) is inoperable, the Required Actions require:

suspension of CORE ALTERATIONS, suspension of movement of irradiated fuel assemblies, and suspension of operations involving positive reactivity additions that could result in loss of required shutdown margin (SDM) or boron concentration.

Because the Required Actions require the suspension of movement of irradiated fuel assemblies, the initiating conditions for an FHA are prohibited. Because the Required Actions require the suspension of positive reactivity additions that could result in a loss of SDM, the initial conditions for a boron dilution accident are prevented. Therefore, the action to suspend CORE ALTERATIONS in these TSs provides no safety benefit and is not needed.

3.2.3 TS 3.9.1 - Boron Concentration Current TS LCO 3.9.1 requires that the boron concentration be within the limits specified in the COLR during operation in Mode 6. When the boron concentration is not within the limits, Required Actions A.1, A.2, and A.3, require the operator to immediately suspend CORE ALTERATIONS, suspend positive reactivity addition, and initiate action to restore boron concentration to within the limits, respectively.

CORE ALTERATIONS are defined in the Prairie Island TSs as the movement of any fuel, sources, or reactivity, control components within the reactor vessel with the vessel head removed and fuel in the vessel. CORE ALTERATIONS could involve the addition of fuel assemblies to the reactor vessel and the withdrawal of control rods, resulting in positive reactivity additions. With the proposed deletion of the action to immediately suspend CORE ALTERATIONS, the proposed Required Action A.1 (current Required Action A.2) still requires immediate suspension of positive reactivity additions, which would prohibit diluting the boron concentration of the coolant in the reactor coolant system, adding fuel assemblies to the reactor vessel, or removing reactivity control components, and proposed Required Action A.2 (current Required Action A.3) continues to require the operator to initiate action to restore boron concentration to within the limits in the COLR. Therefore, the proposed Required Actions A.1 and A.2 would continue to assure that the requirements of boron concentration limits are met and is consistent with STS 3.9.1 in NUREG-1431, Revision 5. Because the proposed Required Actions would continue to provide assurance that the boron concentration is less limiting than the initial condition of the boron dilution analysis, the NRC staff finds that the proposed TS 3.9.1

meets the requirements of 10 CFR 50.36(c)(2) in that following the Required Actions will allow for safe operation of the plant and is therefore acceptable.

3.2.4 TS 3.9.3 - Nuclear Instrumentation Action A.1 Current TS 3.9.3, Action A.1, requires that CORE ALTERATIONS be suspended if the required nuclear instrumentation is determined to be inoperable. Consistent with TSTF-471, the LAR proposed to delete CORE ALTERATIONS from TS 3.9.3, Required Action A.1. The revised Action would be to Suspend Positive Reactivity Additions when required nuclear instrumentation is inoperable.

Two core subcritical neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed core subcritical neutron flux monitors are part of the nuclear instrumentation system. These detectors are located external to the reactor vessel and detect neutrons leaking from the core.

Since these instruments are the only direct means of monitoring core reactivity conditions, if the core subcritical neutron flux monitors are inoperable, positive reactivity additions must be suspended immediately to preclude an accidental criticality. These requirements provide reasonable assurance that an accidental criticality will be avoided and is acceptable to the NRC staff.

The amendment request proposed to replace the Required Action to Suspend CORE ALTERATIONS with the Action to Suspend Positive Reactivity Additions, when required core subcritical neutron flux monitor instrumentation is inoperable. The requirement provides reasonable assurance that an accidental criticality will be avoided.

Action A.2 TSTF-286 Operations Involving Positive Reactivity Additions, (ADAMS Accession No. ML003730788) was adopted into the Prairie Island TSs with the conversion to improved STS in a license amendment issued on July 26, 2002 (ADAMS Accession No. ML022070661).

In a letter dated November 7, 2013, to the TSTF, the NRC staff raised a concern with the Required Actions in the nuclear instrumentation specification when one core subcritical neutron flux monitor is inoperable (ADAMS Accession No. ML13246A358). Resolution of this concern is discussed in further detail in the Section titled TSTF-571-T below. The concern (stated in Section 4 of TSTF-571-T) is that during the movement of fuel assemblies, sources, and reactivity, control components with one core subcritical neutron flux monitor inoperable, there is the potential for the operable core subcritical neutron flux monitor to become effectively decoupled from the core reactivity condition (hereafter referred to as decouple-effect). For example, if one core subcritical neutron flux monitor is inoperable and certain strategically located fuel assemblies are removed, then the operable core subcritical neutron flux monitor may no longer be capable of monitoring the reactivity condition of fuel assemblies that are located in the far half of the core. Therefore, the changes made in TSTF-286 may create a situation in which an increase in reactivity in part of the reactor core might not be detected.

In addressing the NRCs concern regarding the decouple-effect, the licensee added the requirement to suspend movement of fuel, sources, and reactivity control components within the reactor vessel, modified by a note stating that Fuel assemblies, sources, and reactivity

control components may be moved, if necessary, to restore an inoperable core subcritical neutron flux monitor or to complete movement of a component to a safe condition. The NRC staff finds that the addition of this requirement for one or more required core subcritical neutron flux monitors inoperable would avoid the reactivity change and adequately addresses the NRC staffs concern regarding the decouple-effect in TSTF-571-T.

Action C.2 The audible count rate from the core subcritical neutron flux monitors provides prompt and definite indication to the operators of a boron dilution event. If the required core subcritical neutron flux monitor audible count rate circuit is inoperable, TS 3.9.3, Required Action C.2, current requirement is to immediately suspend CORE ALTERATIONS. The amendment request proposed to remove this Required Action from the TSs. This Required Action is not part of the STS so it is not addressed by TSTF-471 or TSTF-571-T.

Currently, the Prairie Island TS bases state that the LCO ensures that audible indication is available to alert the operator in containment in the event of a dilution accident or improperly loaded fuel assembly. In the event of a boron dilution accident, the audible count rate monitors provide audio indication that allows the operators to quickly respond and mitigate the accident.

If the audible monitor is inoperable, then TS 3.9.3, Required Action C.1, would require the licensee to initiate action to isolate sources of unborated water. The LAR stated that isolating unborated sources of water precludes a boron dilution accident. The NRC staff notes that proposed removal of TS 3.9.3, Required Action C.2, would allow for positive reactivity additions when one required audible count rate circuit is inoperable. However, the overall core reactivity will be conservatively less than that assumed in the cycle-specific boron dilution analysis, since this limit was developed assuming that control rods and fuel assemblies are in the most adverse conditions. Therefore, the NRC staff finds that the proposed changes would not permit entry into conditions less favorable than the initial conditions of the boron dilution analysis.

The LAR also proposes to revise the TS 3.9.3 bases to remove references to an improperly loaded fuel assembly. The letter dated September 30, 2021, stated that a review of the TS bases and Updated Safety Analysis Report (USAR) determined that improperly loading a fuel assembly in the reactor vessel is only addressed in TS 3.9.3 bases for the LCO and without reference to any safety analyses. The letter dated September 30, 2021, also states that none of the current Required Actions associated with TS 3.9.3 provide a tie to a misloaded fuel assembly in the reactor vessel.

The NRC staff reviewed Revision 36 of Prairie Island USAR and did not find an analysis considering improper loading of a fuel assembly in the reactor vessel. The NRC staff notes that a fuel misloading accident was generically dispositioned in NSPNAD-8102-PA, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Applications to PI Units, (ADAMS Accession No. ML003676173) which is reference 2 under item b of TS 5.6.5, Core Operating Limits Report. The evaluation concluded that the incore moveable detector system would detect all core loading errors that violate Nuclear Heat Flux Hot Channel Factor (FQ) and Nuclear Enthalpy Rise Hot Channel Factor (FH) limits. The incore moveable detector system does not include the audible count rate circuit or the subcritical neutron flux monitors. Based on this review, the NRC staff agrees that the proposed removal of TS 3.9.3, Required Action C.2, does not relate to detection of an improperly loaded fuel assembly.

Because the proposed change is consistent with the STS and would not permit entry into conditions less favorable than the initial conditions of the boron dilution analysis, the NRC staff

finds that the proposed action provides reasonable assurance that an accidental criticality will be avoided.

TS 3.9.3 Conclusion The proposed changes to Required Actions A.1 and A.2 and C.3 would continue to assure that the requirements of boron concentration limits are met and is consistent with STS 3.9.3 in NUREG-1431, Revision 5. Because of this, the NRC staff finds that the proposed TS 3.9.3 meets the requirements of 10 CFR 50.36(c)(2) in that following the Required Actions will allow for safe operation of the plant and is therefore acceptable.

3.3 TSTF-571-T On November 7, 2013, NRC staff provided a letter to the TSTF expressing concerns with travelers TSTF-51, TSTF-286, and TSTF-471. On October 4, 2018, NRC staff provided a letter to the TSTF detailing the resolution of these concerns and corresponding information that should be provided in future applications to adopt these TSTF travelers. TSTF-571-T addresses these concerns with TSTF-51, TSTF-286, and TSTF-471.

The letters discuss three concerns. The first reiterates a reviewers note in TSTF-51. This note was present in the version of TSTF-51 adopted by the licensee and the licensee addressed this note when they adopted TSTF-51. Because this concern was addressed in a previous licensing action, it was not reviewed further in the current review.

The October 4, 2018, letter also requested specific discussion to be included in the application when the term CORE ALTERATIONS is removed from the applicability of various TSs. The licensee did not include this discussion in the current application, since the LAR does not propose to remove the term CORE ALTERATIONS from the applicability of any TSs. NRC staff notes that the licensees previously approved partial adoption of TSTF-51 did not remove the term CORE ALTERATIONS from the applicability of any TSs either. Because CORE ALTERATIONS is not being removed from applicability of any TSs, this NRC staff concern is adequately addressed by the current application and is acceptable.

The third concern discussed in the October 4, 2018, letter is discussed above for TS 3.9, Action A.2, of this safety evaluation.

Based on the above, the LAR has addressed the concerns in TSTF-51, TSTF-286, and TSTF-471 and is therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Minnesota State official was notified of the proposed issuance of the amendments on February 25, 2022. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no

significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on June 15, 2021 (86 FR 31743). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: A. Rau, NRR Date of Issuance: April 1, 2022

ML22061A206 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME RKuntz SRohrer SKrepel VCusumano DATE 3/2/2022 3/3/2022 1/28/2022 3/4/2022 OFFICE OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME JAzeizat NSalgado (SWall for RKuntz w/comments)

DATE 3/25/2022 4/1/2022 4/1/2022

Retrieved from "https://kanterella.com/w/index.php?title=ML22061A206&oldid=3439785"