ML18241A383

Issuance of Amendment 307 to Revise Technical Specifications to Adopt TSTF-551, Revision 3

ML18241A383
Person / Time
Site:	Duane Arnold
Issue date:	10/31/2018
From:	Chawla M L Plant Licensing Branch III
To:	Dean Curtland Nextera Energy
Chawla M L NRR/DORL/LPL3-1 301-415-8371
References
EPID L-2017-LLA-0385
Download: ML18241A383 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Dean Curtland NextEra Energy Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324-9785 October 31, 2018

SUBJECT:

DUANE ARNOLD ENERGY CENTER -ISSUANCE OF AMENDMENT NO. 307 TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-551, REVISION 3 (EPID L-2017-LLA-0385)

Dear Mr. Ci.Jrtland:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 307 to Renewed Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC). The amendment consists of changes to the technical specifications (TS) for DAEC to adopt Technical Specifications Task Force (TSTF) traveler TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements," in response to your application dated November 10, 2017. A copy of the related Safety Evaluation is also enclosed.

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-331

Enclosures:

1. Amendment No. 307 to License No. DPR-49 2. Safety Evaluation cc: ListServ Sincerely, Mahesh L. Chawla, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY DUANE ARNOLD, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 307 License No. DPR-49 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by NextEra Energy Duane Arnold, LLC dated November 10, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

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-49 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC shall operate the facility in accordance with the Technical Specifications.

Enclosure 1 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

Attachment:

Changes to Renewed Facility Operating License No. DPR-49 and Technical Specifications Date of Issuance:

October 31 , 2 O 1 8 FOR THE NUCLEAR REGULATORY COMMISSION Q~cz ~/ ---Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 307 RENEWED FACILITY OPERATING LICENSE NO. DPR-49 DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331 Replace the following page of Renewed Facility Operating License No. DPR-49 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. REMOVE 3 INSERT 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. REMOVE 3.6-36 INSERT 3.6-36 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 I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, 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 NextEra Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC I shall operate the facility in accordance with the Technical Specifications. (a) For Surveillance Requirements (SRs) whose acceptance criteria are modified, either directly or indirectly, by the increase in authorized maximum power level in 2.C.(1) above, in accordance with Amendment No. 243 to Facility Operating License DPR-49, those SRs are not required to be performed until their next scheduled performance, which is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment No. 243. (b) Deleted. (3) Fire Protection Program NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated August 5, 2011 (a.nd supplements dated October 14, 2011, April 23, 2012, May 23, 2012, July 9, 2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29, 2013, two supplements dated July 2, 2013, and supplements dated August 5, 2013 and August 28, 2013) and as approved in the safety evaluation report dated September 10, 2013. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. DPR-49 Amendment No. 307 Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SR 3.6.4.1.1 SR 3.6.4.1.2 SR 3.6.4.1.3 DAEC SURVEILLANCE Verify all secondary containment equipment hatches are closed. -----------------------N()TE----------------------------

Doors in high radiation areas may be verified by administrative means. Verify that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed, except when the access opening is being used for entry and exit. Verify each SBGT subsystem can maintain 0.25 inch of vacuum water gauge in the secondary containment at a flow rate 4000 cfm. 3.6-36 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment 307 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 307 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

By application dated November 10, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 17318A240), NextEra Energy Duane Arnold, LLC (NextEra or the licensee) requested changes to the technical specifications (TS} for the Duane Arnold Energy Center (DAEC). Specifically, the licensee requested changes to the TSs to adopt Technical Specifications Task Force (TSTF) traveler TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements," dated October 3, 2016 (ADAMS Accession No. ML 16277A226).

The U.S. Nuclear Regulatory Commission (NRC or the Commission) approved the traveler on September 21, 2017 (ADAMS Package Accession No. ML 17236A365).

The proposed change would revise the TS to permit secondary containment access openings to be open to permit entry and exit.

2.0 REGULATORY EVALUATION

2.1 System Description The secondary containment is a structure that encloses the primary containment including components that may contain primary system fluid. The safety function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a design-basis accident (OBA) to ensure that the control room operator and offsite doses are within the regulatory limits. There is no redundant train or system that can perform the secondary containment function should the secondary containment be inoperable.

The secondary containment boundary is the combination of walls, floor, roof, ducting, doors, hatches, penetrations, and equipment that physically form the secondary containment.

Routinely used secondary containment access openings contain at least one inner and one outer door in an airlock configuration.

In some cases, secondary containment access openings are shared such that there are multiple inner or outer doors. All secondary containment access Enclosure 2 doors are normally kept closed, except when the access opening is being used for entry and exit of personnel, equipment, or material.

Secondary containment operability is based on its ability to contain, dilute, and hold up fission products that may leak from primary containment following a OBA. To prevent ground level exfiltration of radioactive material while allowing the secondary containment to be designed as a mostly conventional structure, the containment requires support systems to maintain the pressure at less than atmospheric pressure.

During normal operation, nonsafety-related systems are used to maintain the secondary containment at a slight negative pressure to ensure that any leakage is into the building and that any containment atmosphere exiting is via a pathway monitored for radioactive material.

However, during normal operation, it is possible for the secondary containment vacuum to be momentarily less than the required vacuum for a number of reasons such as during wind gusts or swapping of the normal ventilation subsystems.

During emergency conditions, the standby gas treatment (SGT) system is designed to be capable of drawing down the secondary containment to a required vacuum within a prescribed time and continue to maintain the negative pressure as assumed in the accident analysis.

For DAEC, the SGT must be able to establish the required vacuum within 5 minutes. The leak tightness of the secondary containment together with the SGT system ensure that radioactive material is either contained in the secondary containment or filtered through the SGT system filter trains before being discharged to the outside environment via the elevated release point. 2.2 Proposed Technical Specification Change DAEC surveillance requirement (SR) 3.6.4.1.2 requires verification that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed. The proposed change would modify this SR by adding the following phrase to the end of the SR statement, " ... except when the access opening is being used for entry and exit." The proposed change would allow for the temporary opening of the inner and outer doors of secondary containment for the purpose of entry and exit (i.e., normal opening and prompt closure of a door for transit).

2.3 Regulatory Requirements and Guidance The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TS in accordance with the requirements of 10 CFR 50.36. The applicant must include in the , application a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls." However, per 10 CFR 50.36(a)(1), these TS bases "shall not become part of the technical specifications." Additionally, 10 CFR 50.36(b) states: Each license authorizing operation of a ... utilization facility ... will include technical specifications.

The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"].

The Commission may include such additional technical specifications as the Commission finds appropriate. The categories of items required to be included in the TSs are provided in 10 CFR 50.36(c).

As required by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCO will be met. The NRC staff's guidance for review of TSs is in Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP), dated March 2010 (ADAMS Accession No. ML 100351425).

SRP, Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, dated July 2000 (ADAMS Accession No. ML003734190), provides guidance to the NRC staff for the review of alternate source term (AST) amendment requests.

It states that the NRC reviewer should evaluate the proposed change against the guidance in NRC Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Revision 0, dated July 2000 (ADAMS Accession No. ML003716792).

RG 1.183 provides an acceptable methodology for analyzing the radiological consequences of several design basis accidents to show compliance with 10 CFR 50.67, "Accident source term." RG 1.183 provides guidance to licensees on acceptable application of AST (also known as the accident source term) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST. The regulation at 10 CFR 50.67(b )(2) states: The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that: (i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [Sievert]

(25 rem [roentgen equivalent man]}2 total effective dose equivalent (TEDE). (ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive 2 The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions.

Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation. a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE). (iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

3.0 TECHNICAL EVALUATION The NRC staff evaluated the licensee's application in accordance with the regulations and guidance discussed in Section 2.3 of this safety evaluation (SE) and the NRG-approved traveler TSTF-551, Revision 3. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate.

In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 1 O CFR 50.54. The application also included proposed changes to the TS bases. 3.1 Proposed Change to SR 3.6.4.1.2 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions.

Normal entry and exit conditions do not include planned activities that could result in the simultaneous opening of redundant secondary containment access openings, such as maintenance of a secondary containment personnel access door or movement of large equipment through the openings that would take longer than the normal transit time. The NRC staff reviewed the proposed change to SR 3.6.4.1.2.

The NRC staff determined that the SR continues to provide appropriate confirmation that secondary containment boundary doors are properly positioned and capable of performing their function in preserving the secondary containment boundary.

The NRC staff determined that the SR continues to appropriately verify the operability of the secondary containment and provides assurance that the necessary quality of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).

Additionally, the NRC staff evaluated the impact of modifying the licensee's TS to allow secondary containment access openings to be open for entry and exit on the licensee's basis radiological consequence dose analyses to ensure that the modification will not result in an increase in the radiation dose consequences and that the resulting calculated radiation doses will remain within the design criteria specified in the current radiological consequence analyses.

The NRC staff review of these DBAs determined that there is one DBA that takes credit for the secondary containment and is possibly impacted by the brief, inadvertent, simultaneous opening of both an inner and outer access door during normal entry and exit conditions, the loss-of-coolant accident (LOCA). 3.1.1 LOCA Following a LOCA, the secondary containment structure is maintained at a negative pressure ensuring that leakage from primary containment to secondary containment can be collected and filtered prior to release to the environment.

The SGT system performs the function of maintaining a negative pressure within the secondary containment, as well as collecting and filtering the leakage from primary containment.

The licensee credits the SGT system for mitigation of the radiological releases from the secondary containment.

In the LOCA analysis, the secondary containment draw down analysis assumes that the SGT system can draw down the secondary containment within 5 minutes. SR 3.6.4.1.3 verifies each SGT subsystem can maintain greater than or equal to 0.25 inches of vacuum water gauge in the secondary containment at a flow rate of less than or equal to 4000 cubic feet per minute. Conservatively, the OBA LOCA radiological consequence analysis in the OAEC updated final safety analysis report (UFSAR), Section 15.2, assumes that following the start of a OBA LOCA, the secondary containment pressure of 0.25 inches of vacuum water gauge is achieved at approximately 5 minutes. The license assumes that releases into the secondary containment prior to the 5-minute draw down time leak directly to the environment as a ground level release with no filtration.

After the assumed 5-minute draw down, these releases are filtered by the SGT system and released via the SGT system exhaust vent. The NRC staff reviewed the information contained in the license amendment request (LAR) and the OAEC UFSAR to ascertain the conservatism in the 5-minute draw down time assumption.

The LAR indicated that in the unlikely event that an accident would occur when both personnel access doors are open for entry and exit, the brief time required to close one of the doors is small compared to the draw down time margin provided by the 5 minute duration assumed in the accident analysis for reducing the post-accident secondary containment pressure to 0.25 inches of vacuum water gage. Based on reviews of similar LARs, the NRC staff concludes that the time both doors may be open simultaneously will be limited to the time it takes to traverse through a door, typically less than 10 seconds. Section 6.5.3.3, "Standby Gas Treatment System," of the OAEC UFSAR states that upon receipt of SGT system initiation signals, the reactor building ventilation supply and exhaust fans are tripped, the normal reactor building ventilation is isolated, and both trains of the SGT systems are started. In addition, the dampers required for reactor building isolation and SGT initiation are designed to go to their required position within 10 seconds, and the SGT fans will be up to speed in less than 10 seconds. Based on this information, the NRC staff concludes that the licensee's OBA LOCA analysis has sufficient conservatism by assuming a draw down time of 5 minutes from the start of the OBA LOCA. Margin exists to ensure that the secondary containment can be reestablished during a brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable assurance that a failure of a safety system needed to control the release of radioactive material to the environment will not result. The brief, inadvertent, simultaneous opening of the secondary containment access doors does not impact the OBAs and will not result in an increase in any onsite or offsite dose. Based on the above discussion, the NRC staff finds that the licensee's proposed change to the TS does not impact the licensee's design basis LOCA radiological consequence analysis and will not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that this change is acceptable with respect to the radiological consequences of the OBAs. The licensee was approved for AST methodology and the radiological dose consequences analyses for DBAs via license Amendment Nos. 237 and 240 for the DAEC. The NRC staff reviewed the impact of the proposed change to the DAEC TS on all DBAs currently analyzed in the DAEC UFSAR that could have the potential for significant dose consequences.

Chapter 15 of the DAEC UFSAR describes the DBAs and their radiological consequence analysis results. See Section 3.2 for the discussion of DBAs other than LOCA (which was discussed above). The NRC staff review was limited to the licensee's request to provide an allowance for the brief, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions.

Normal entry and exit conditions do not include planned activities that could result in the simultaneous opening of redundant secondary containment access openings, such as maintenance of a secondary containment personnel access door or movement of large equipment through the openings that would take longer than the normal transit time. 3.1.2 Conclusion As described above, the NRC staff reviewed the technical basis provided by the licensee to assess the radiological impacts of the proposed change to SR 3.6.4.1.2.

The NRC staff finds that the proposed change is consistent with the regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds that, with the proposed change, the TSs will continue to comply with these criteria and the licensee's estimates of the dose consequences of a design basis LOCA will continue to comply with the requirements of the current radiological consequence analyses.

Therefore, the proposed change is acceptable with regard to the radiological consequences of the postulated DBAs. 3.2 Variations from the Approved Traveler The licensee is proposing the following variations from the TS changes described in the NRG-approved TSTF-551, Revision 3, or the applicable parts of TSTF-551 or the NRC staff's SE for TSTF-551.

These variations do not affect the applicability of TSTF-551 or the NRC staffs SE for TSTF-551 to the proposed LAR. The DAEC TS does not contain an SR equivalent to SR 3.6.4.1.1 in TSTF-551.

Therefore, the addition of the SR 3.6.4.1.1 Note, as provided by TSTF-551, is not applicable.

The DAEC TS does not contain an SR equivalent to SR 3.6.4.1.4 in TSTF-551.

Therefore, the editorial change to SR 3.6.4.1.4, as provided by TSTF-551, is not applicable.

The DAEC TS utilizes different SR numbering and wording than the Standard Technical Specifications (STS) on which TSTF-551 was based. Specifically, DAEC SR 3.6.4.1.2 verifies that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed whereas STS SR 3.6.4.1.3 verifies one secondary containment access door in each access opening is closed. These differences are editorial and do not affect the applicability of TSTF-551 to the DAEC TS. TSTF-551 discusses the applicable regulatory requirements and guidance, including the 10 CFR Part 50, Appendix A, General Design Criteria (GDC). DAEC was not licensed to the 1 O CFR Part 50, Appendix A, GDC. The DAEC equivalents of the referenced GDC are Atomic Energy Commission GDC for nuclear power plants, Appendix A, of 10 CFR Part 50, effective May 21, 1971, and subsequently amended July 7, 1971, as discussed in DAEC UFSAR, Section 3.1. This difference does not alter the conclusion that the proposed change is applicable to DAEC. The final model SE for TSTF-551 discusses that the NRC staff review determined that there are two DBAs that take credit for the secondary containment and are possibly impacted by the brief, inadvertent, and simultaneous opening of both an inner and outer access door during normal entry and exit conditions (i.e., LOCA and the fuel handling accident (FHA) in secondary containment).

The DAEC FHA does not credit the secondary containment or the SGT system for mitigation of FHAs. Because the DAEC FHA radiological consequence analysis does not credit the secondary containment or the SGT system, the FHA in secondary containment analysis is not impacted by the brief, inadvertent, and simultaneous opening of both an inner and outer access door during normal entry and exit conditions.

This difference does not alter the conclusion that the proposed change is applicable to DAEC. 3.3 Summary The NRC staff reviewed the proposed change and determined that it meets the standards in 10 CFR 50.36(b).

The proposed SR assures that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met and satisfy 10 CFR 50.36(c)(3).

Additionally, the change to the TS was reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP, Chapter 16. . Additionally, the NRC staff evaluated the impact of the proposed change on the design basis radiological consequence analyses against the regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds that with the proposed change, the TSs will continue to comply with the requirements of the current radiological consequence analyses.

Therefore, the proposed change is acceptable with regard to the radiological consequences of the postulated DBAs.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment on September 4, 2018. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on February 27, 2018 (83 FR 8517). Accordingly, the amendment meets 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 amendment.

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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors:

Kristy Bucholtz, NRR/DRA/ARCB Nageswara Karipineni, NRR/DSS/SBPB Jerome Bettle, NRR/DSS/SCPB Margaret Chernoff, NRR/DSS/STSB Andrea Russell, NRR/DSS/STSB Date of issuance:

October 31, 2018

SUBJECT:

DUANE ARNOLD ENERGY CENTER -ISSUANCE OF AMENDMENT NO. 307 TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-551, REVISION 3 (EPID L-2017-LLA-0385)

DATED OCTOBER 31, 2018 DISTRIBUTION:

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