ML14290A360

Issuance of Amendment No. 228, Request to Adopt Technical Specification Task Force (TSTF)-535, Revision 0, Revise Shutdown Margin Definition to Address Advanced Fuel Designs

ML14290A360
Person / Time
Site:	Columbia
Issue date:	11/12/2014
From:	Andrea George Plant Licensing Branch IV
To:	Reddemann M Energy Northwest
George A
References
TAC MF3649
Download: ML14290A360 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 12, 2014 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION -ISSUANCE OF AMENDMENT RE:

ADOPTION OF TSTF-535, REVISION 0 (TAG NO. MF3649)

Dear Mr. Reddemann:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 228 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station (CGS). The amendment consists of changes to the Technical Specifications (TS) in response to your application dated March 18, 2014.

The amendment proposed to adopt TS Task Force (TSTF) change traveler TSTF-535, Revision 1, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs" at CGS.

The notice of ,availability of TSTF-535, Revision 0, was announced in the Federal Register on February 26, 2013 (78 FR 131 00),

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Andrea E. George, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 228 to NPF-21

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION VIJASHINGTON, D.C. 20555-9001 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 228 License No. NPF-21

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

A. The application for amendment by Energy Northwest (licensee), dated March 18, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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.

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. NPF-21 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 228 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license. The licensee shall operate the facility ih accordance with the Technical Specifications and the Environmental Protection Plan ..

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-21 and Technical Specifications Date of Issuance: November 12, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 228 RENEWED FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Renewed Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 1.1-5 1.1-5

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 228 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.

(3) Deleted.

(4) Deleted.

(5) Deleted.

(6) Deleted.

(7) Deleted.

(8) Deleted.

(9) Deleted.

(10) Deleted.

(11) Shield Wall Deferral (Section 12.3.2. SSER #4, License Amendment #7)

The liceQsee shall complete construction of the deferred shield walls ana window as identified in Attachment 3, as amended by this license amendment. *

(12) Deleted.

(13) Deleted.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-21 Amendment No. 228

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heattransfer rate to the

_(RTP) reactor coolant of 3486 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM(RPS)RESPONSE when the monitored parameter exceeds its RPS trip setpoint at TIME the channel sensor until de-energization of tlie scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is c: 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

• THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

Columbia Generating Station 1.1-5 Amendment No. +69 2ae 228

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELATED TO AMENDMENT NO. 228 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By application dated March 18, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14086A389), Energy Northwest (the licensee), proposed changes to the Technical Specifications (TSs) for Columbia Generating Station (CGS).

Specifically, the licensee requested to adopt U.S. Nuclear Regulatory Commission (NRC)-

approved TS Task Force (TSTF) Standard Technical Specifications (STS) change traveler TSTF-535, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs," dated August 8, 2011 (ADAMS Accession No. ML112200436).

The proposed amendment modifies the TS definition of "Shutdown Margin" (SDM) to require calculation of the SDM at a reactor moderator temperature of 68 degrees Fahrenheit CF) or a higher temperature that represents the most reactive state throughout the operating cycle. This change is needed to address new boiling-water reactor (BWR) fuel designs which may be more reactive at shutdown temperatures above 68 °F.

The licensee stated that the license amendment request is consistent with NRC-approved TSTF traveler TSTF-535. The availability of this TS improvement was announced in the Federal Register on February 26, 2013 (78 FR 131 00), as part of the consolidated line item improvement process.

2.0 REGULATORY EVALUATION

2.1 Background In water-moderated ~eactors, water is used to slow down, or moderate, high-energy fast neutrons to low-energy thermal neutrons through multiple scattering interactions. The low-energy thermal neutrons are much more likely to cause fission when absorbed by the fuel.

However, not all of the thermal neutrons are absorbed by the fuel; a portion of them are instead absorbed by the water moderator. The amount of moderator and fuel that is present in the core heavily influences the fractions of thermal neutrons that are absorbed in each.

Enclosure 2

Water-moderated reactors are designed such that they tend to operate in what is known as an under-moderated condition. In this condition, the ratio of the moderator-to-fuel in the core is small enough that the overall effectiveness of water as a moderator decreases with increasing temperature; fewer neutrons are absorbed in the moderator due to the decrease in its density, but this is overshadowed by the reduction in the number of neutrons that moderate from high fission energy to the lower energy level needed to cause fission. The result is a decrease in power and temperature: a negative reactivity feedback effect where the reactor becomes self-regulating. However, if the amount of moderator becomes too large with respect to the amount of fuel, the reactor can enter an over-moderated condition. In this condition, the overall effectiveness of water as a moderator increases with increasing temperature; the reduction in the number of neutrons absorbed in the moderator outweighs the loss in neutrons reaching lower energies. This causes an increase in power that leads to a further increase in temperature creating a potentially dangerous positive reactivity feedback cycle.

As practical examples in support of the proposed changes to the definition of SDM, TSTF-535 discussed SDM with regard to GE14 and GNF2 fuels. TSTF-535 indicated that for historical fuel products through GE14, the maximum reactivity condition for SDM always occurred at a moderator temperature of 68 oF because these fuel products were designed so that the core is always under-moderated when all control rods are inserted, except for the single most reactive rod. In cores with GNF2 fuel, TSTF-535 stated that it is expected that the maximum reactivity condition at beginning of cycle will remain at 68 °F, but that later in cycle, the most limiting SDM may occur at a temperature greater than this, indicating that with this fuel design the core could potentially achieve an over-moderated condition.

In its letter dated March 18, 2014, the licensee stated, in part, that Energy Northwest has concluded that the justifications presented in the TSTF-535 proposal and the model safety evaluation prepared by the NRC staff are applicable to Columbia Generating Station (Columbia) and justify this amendment for the incorporation of the changes to the Columbia TS.

2.2 Technical Specification Changes The licensee's adoption of TSTF-535 for CGS proposes to revise the TS definiti9n of SDM to require calculation of SDM at the reactor moderator temperature corresponding to the most reactive state throughout the operating cycle (68 oF or higher).

The current definition of SDM in* Section 1.1, "Definitions," of the CGS TS is:

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F; and

c. All control rods are fully inserted except for the single control.rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

The licensee proposes the following changes (shown in bold) to the definition of SDM in accordance with TSTF-535:

• SDM shall be tht3 amount of reactivity by which the reactor is evbcritioal or would be subcritical throughout the operating cycle assuming that:

a. The remoter ie xenon free;

b. The moderator temperature ia i!! 68°F, corresponding to the most reactive state; and

e. All control rods are fully inserted except for the single control rod ef highest reaetivity worth, whieh is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

2.3 Regulatory Review Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power pl_ant operating licenses to include TS as part of the license application. The Commission's regulatory requirements related to the content of the TS are contained in Section 50.36(c) of Title 10 of the Code of Federal Regulations (1 0 CFR). That regulation requires that the TS include, among other things, items in the following categories: (1) safety limits, limiting safety systems settings,

• and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.

The regulations in 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 26, "Reactivity control system redundancy and capability," and GDC 27, "Combined reactivity control systems capability," respectively, require that reactivity within the core be controllable to ensure subcriticality is achievable and maintainable under cold conditions, with appropriate margin for stuck rods; and that reactivity within the. core be controllable to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

Among other things, 1.0 CFR. 50.36(c)(2)(1i)(l'3) requires the establishtrH~nt of a.limlting condition for operation (LCO) for a "process variable, design feature, or operating restriction that Is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The TS definition of SDM and the LCOs placed on SDM serve, in part, to satisfy GDCs 26 and 27 by ensuring there is always sufficient negative reactivity worth available to offset the positive reactivity worth of changes in moderator and fuel temperature, the decay of fission product poisons, the failure of a control rod to insert, and reactivity insertion accidents. Given this margin, the core can be held subcritical for conditions of normal operation, including anticipated operational occurrences.

Pe.r the licensee's Final Safety Analysis Report (FSAR) Amendment 62, the licensee de~ermined that requirements of the subject GDCs are met based upon the following plant specifics:

The Licensee's Evaluation Against Criterion 26 Two independent reactivity control systems using different design principles are

• provided. The normal method of reactivity control employs control rod assemblies. Positive insertion of these control rods is provided by means of the CRD [control rod drive] hydraulic system. The control rods are capab.le of reliably controlling reactivity changes during normal operation (e.g., power changes, power shaping, xenon burnout, normal startup and shutdown) via operator-controlled insertions and withdrawals. The control rods are also capable of maintaining the core within acceptable fuel design limits during anticipated operational occurrences via the automatic scram function. The unlikely occurrence of a limited number of stuck rods during a scram will not adversely affect the capability to maintain the core within fuel design limits.

The circuitry for manual insertion or withdrawal of control rods is completely independent of the circuitry for reactor scram. This separation of the scram and normal rod control functions prevents failures in the reactor manual control circuitry from affecting the scram circuitry. Two sources of scram energy (accumulator pressure and reactor vessel pressure) provide needed scram performance over the entire range of reactor pressure, i.e., from operating conditions to cold shutdown. The design of the control rod system includes appropriate margin for malfunctions such as stuck rods. Control rod withdrawal sequences and patterns are selected prior to operation to achieve optimum core performance, and simultaneously, low individual rod worths. The operating procedures to accomplish such patterns are supplemented by the rod worth minimizer, which prevents rod withdrawals yielding a rod worth greater than permitted by the preselected rod withdrawal pattern. Because of the carefully planned and regulated rod withdrawal sequence, prompt shutdown of the reactor can be achieved with the insertion of a small number of the many independent control rods. In the event that a reactor scram is necessary, the unlikely occurrence* of a limited number of stuck rods will not hinder the capability of the control rod system to render the core subcritical The second independent reactivity control system is provided by the reactor coolant recirculation system. By varying reactor flow, it is possible to affect the type of reactivity changes necessary for planned, normal power changes (including xenon burnout). In the event that reactor flow is suddenly increased to its maximum value (pump runout), the core will not exceed fuel design limits because the power flow map defines the allowable initial operating states such that the purrip runout will not violate these limits:

The control rod system is capable of holding the reactor core subcritical under cold conditions, even when the control rod of highest worth is assumed to be

stuck in the fully withdrawn position. This shutdown capability of the control rod system is made possible by designing the fuel with burnable poison (Gd 2 0 3) to control the high reactivity of fresh-fuel. In addition, the standby liquid control system is available to add soluble boron to the core and render it subcritical, as discussed in Section 3.1.2.3.8.

The redundancy and capabilities of the reactivity control systems for the BWR satisfy the requirements of Criterion 26.

Evaluation Against Criterion 27 There is no credible event applicable to the BWR which requires combined capability of the control rod system and poison additions by the emergency core cooling network. The BWR design is capable of maintaining the reactor core

• subcritical, including allowance for a stuck rod, without addition of any poison to the reactor coolant. The primary reactivity contr.ol system for the BWR during postulated accident conditions is the control rod system. Abnormalities are sensed and, if protection system limits are reached, corrective action is initiated through automatic insertion of control rods. High integrity of the protection system is achieved through the combination of logic arrangement, actuator redundancy, power supply redundancy, and physical separation. High reliability of reactor scram is further achieved by separation of scram and manual control circuitry, in-dividual control units for each control rod, and fail-safe design features built into the rod drive system. Response by the RPS [reactor protection system]

is prompt and the total scram time is short.

In the event that more than one control rod fails to insert, and the core cannot be maintained in a subcritical condition by control rods alone .as the reactor is cooled down subsequent to initial shutdown, the standby liquid control (SLC) system will be actuated to insert soluble boron into the reactor core. The SLC system has sufficient capacity to ensure that the reactor can always be maintained subcritical; and hence, only decay heat will be generated by the core which can be removed by the RHR [residual heat removal] system, thereby ensuring that the core will always be coolable.

The design of the reactivity control system ensures reliable control of reactivity under postulated accident conditions with appropriate margin for stuck rods. The capability to cool is maintained under all postulated accident conditions; thus, Criterion 27 is satisfied.

3.0 TECHNICAL EVALUATION

3.1 Current Definition of Shutdown Margin In BWR plants, the control rods are used to hold the reactor core subcritical under cold conditions. The control rod negative reactivity worth must be sufficient to ensure the core is subcritical by a margin known as the SDM. It is the additional amount of negative reactivity worth needed to maintain the core subcritical by offsetting the positive reactivity worth that can

occur during the operating cycle due to changes in moderator and fuel temperature, the decay of fission product poisons, the failure of a control rod to insert, and reactivity insertion accidents.

Specifically, Section 1.1, "Definitions," of the STS defines SDM as the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that the reactor is ( 1) xenon free, (2) the moderator is 68 °F, and (3) all control rods are fully inserted except for the rod of highest worth, which is assumed to be fully withdrawn.

The three criteria provided in the definition help exemplify what has traditionally been the most reactive design condition for a reactor core. Xenon is a neutron poison produced by fission product decay and its presence in the core adds negative reactivity worth. Assuming the core is xenon free removes a positive reactivity offset and is representative of fresh fuel at the beginning of cycle. The minimum temperature the reactor moderator is anticipated to experience is 68 °F, making it the point at which the moderator will be at its densest and therefore capable of providing the highest positive reactivity worth. By assuming the highest worth rod is fully withdrawn, the core can be designed with adequate SDM to ensure it remains safely shut down even in the event of a stuck control rod, as required by GDCs 26 and 27.

Determination of the SDM under the aforementioned conditions yields a conservative result that, along with the requirements set forth in Section 3.1.1 of the TS, helps ensure:

a. the reactor can be made subcritical from all operating conditions and transients and design basis events,

b. the reactivity transients associated-with postulated accident conditions are controllable within acceptable limits, and

.*c. the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in* the shutdown condition ..

3.2 Proposed Definition of Shutdown Margin*

• The specified moderator temperature of 68°F facilitates the maximum reactivity condition only if the core exists in an urider-moderated condition. In addition fo:burnable poisons; many modern fuel designs also incorporate partial _length rods for increased neutron economy which are employed in *order to extend the operating cycle. Both of these affect the ratio of moderator to fuel. The strong local absorption* effects of the bu*rnable poisons in fresh fuel' make the core under--moderated. As burnable poisons are depleted during the fuelcycle, the core becomes less under-moderated, potentially leading to a slightly over-moderated condition wherein the core will be more reactive at a moderator temperature higher than the 68 oF specified in the SDM definition. Thus, the maximum core reactivity conditior:i and the most limiting SDM may occur later inthe fuel cycle at a temperature greater than 68 °F. Consequently, calculation of the SDM at the currently defined moderator temperature of 68 oF may not accurately determine the available margin. . * *

• TSTF-535, therefore, proposed a change to the definition of SDM to enable calculation of the SDM at a reactor moderator temperature*of 68*"F or a higher temperature corresponding to the most reactive state throughout the operating cycle. SDM would be calculated using the appropriate limiting conditions for .all fuel types at any time in core life.

In $Upport of the proposed enange, 'fSTF~535 eited the ret:~uirements for SE>M as specified in Topical Report NED0~2401.1~A, Revision 18, "General Electric Standard Application for Reactor Fuel (GESTAR II)," dated April 2011 (ADAMS Accession No. ML111120046). Section 3.2.4.1, "Shutdown Reactivity," of GESTAR II states, in part, that:

The core must be capable ~f being made suberitieal, with margin, in the most reC!etive condition throughout the operating cycle with the most reactive eontrol rod fully withdrawn and all other rods fully inserted.

The Traveler also cited NUREG*OBOO, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (SRP) Section 4.3, "Nuclear Design" (ADAMS Accession No. ML070740003), whic~ states the following specific area of review for reactivity control system~ and SDM:

The adequacy of the control systems to assure that the reactor can be returned to and maintained in the cold shutdown condition at any time during operation.

The applicant shall discuss shutdown margins (SDM). Shutdown margins need to be demonstratec:l by the applicant throughout the fuel cycle.

Although the licensing basis for SDM in GESTAR II are only applicable for cores licensed with Global Nuclear Fuels methods, they are consistent with the review procedures set forth in the SRP, which are provided to help ensure compliance with GDCs 26 and 27. TSTF-535 stated that while the SRP doe~ not prescribe the temperature at which the minimum SDM should be determined, the requirement of shutting down the reactor and maintaining it in a shutdown condition "at any time during operation" suggests that considering a range of thermal and exposure conditions would be appropriate in the determination of the minimum SDM. Because newer fuel designs employ elements s*uch as partial length rods and burnable absorbers, which may cause the maximum core reactivity conditions and the most limiting SDM to occur later in the fuel cycle at a temperature greater than 68 °F, the NRC staff agrees with the topical report assessment in this regard. Additionally, the NRC staff concludes that allowing calculation of the SDM at the most limiting core reactivity condition is prudent with respect to ensuring compliance with GDCs 26 and 27 or their plant-specific equivalent, and concludes that the proposed changes to the CGS TSs are acceptable.

The impetus for TSiF-5:35 was to provide for a more broadly applicable SDM definition in recognition of modern fuel designs, for which the core may not be in its most reactive condition at 68 °F. The proposed language will require the licensee to consider all temperatures equal to or exceeding 68 °F, and all times in the operating cycle. This change places an additional responsibility on the licensee to identify the most limiting time-in-cycle and temperature, a change that is more conservative than the current definition and will ensure the licensee maintains adequate SDM as required by their current licensing basis. Therefore, the change is acceptable for CGS. The NRC staff also concludes that the revised definition is consistent with the 10 CFR 50.36 r~quir~ments pertaining to LCOs, because it ensures that the LCOs for SDM consider a broadly conservative range of potential initial conditions in the anticipated operational occurrence analyses.

3.3 Summary The NRC staff has reviewed the licensee's implementation of TSTF-535 proposed revisions to the definition of SDM. Based on the considerations discussed above, the NRC staff concludes that the proposed revisions are acceptable and will provide a conservative and improved approach to the calculation of SDM that ensures use of the appropriate limiting conditions for all fuel types at any time in the life of the core. Additionally, the NRC staff concludes the proposed changes to the definition of SDM will require the licensee to calculate SDM in consideration of the most limiting conditions in the core. Therefore, the revised SDM definition is acceptable for use with any current fuel design.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the 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 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 July 22, 2014 (79 FR 42544). 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 r)lanner, (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 Contributor: R. Grover Date: November 12, 2014

• .. ML14290A360 *via memo **via email OFFICE NRR/DORLILPL4-1 /PM NRR/DORLILPL4-1 /LA NRR/DSS/STSB/BC NAME A George JBurkhardt** REIIiott*

DATE 10/23/14 10/23/14 08/28/14 OFFICE. OGC INLO w/cori?ments NRR/DORL/LPL4-1 /BC NRR/DORLILPL4-1 /PM NAME Jlindell MMarkley A George DATE 10/30/14 . 11/06/14 11/12/14