ML112870012

Issuance of Amendment Revision to APRM Operability Requirements

ML112870012
Person / Time
Site:	Nine Mile Point  (DPR-063)
Issue date:	10/31/2011
From:	Richard Guzman Division of Operating Reactor Licensing
To:	Langdon K Nine Mile Point
Guzman R, NRR/DORL 415-1030
References
TAC ME5010
Download: ML112870012 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 October 31, 2011 Mr. Kenneth Langdon Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 SUB..IECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO.1-ISSUANCE OF AMENDMENT REGARDING REVISIONS TO AVERAGE POWER RANGE MONITOR INSTRUMENTATION SYSTEM OPERABILITY REQUIREMENTS TECHNICAL SPECIFICATION 3.6.2, PROTECTIVE INSTRUMENTATION (TAC NO. ME5010)

Dear Mr. Langdon:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 211 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1), in response to your application dated November 2, 2010, (Agencywide Documents and Management System (ADAMS) Accession No. ML103130037), as supplemented on January 27,2011 (ADAMS Accession No. ML110400115).

This amendment revises the NMP1 Technical Specifications (TSs) Section 3.6.2, "Protective Instrumentation," by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The amendment eliminates the requirements that the APRM "Upscale' and "Inoperative' scram and control rod withdrawal block functions be operable when the reactor mode switch is in the Refuel position. The amendment also clarifies the operability requirements for the APRM Downscale control rod withdrawal block function when the reactor mode switch is in the Startup and Refuel positions.

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

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 211 to DPR-63

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC (NMPNS)

DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 211 Renewed License No. DPR-63

1.

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

A.

The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee) dated November 2,2010, as supplemented on January 27, 2011, 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-63 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 211, is hereby incorporated into this license.

Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications

- 2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Nancy L. Salgado, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: October 31, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 211 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following pages of 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 Pages Insert Pages 200 200 227 227 232 232

-3 (3)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70 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, 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, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

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; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license 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 also subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 211, is hereby incorporated into this license. Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical SpeCifications.

(3)

Deleted Renewed License No. DPR 63 Amendment No. 191 to 210, 211

TABLE 3.6.2a (cont'd)

INSTRUMENTATION THAT INITIATES SCRAM Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No.

Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System Set Point Operable c

~

0 c.

'i

I J3 1::

I

.2 c

.c G)

J!

I en 0::

en 0::

(ii) Inoperative (b) APRM (i) Upscale (ii) Inoperative 2

3(d)(0) 2 3(e)(0) 2 3(e)(0)

(10)

Turbine Stop Valve Closure 2

4(0)

(11)

Generator Load Rejection 2

2(0)

AMENDMENT NO. 142. 143.153, 211 Specification 2.1.2a S 10% valve closure (j) x x

x x

x x

(i)

(i) 200

TABLE 3.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWAL BLOCK Limiting Condition for Operation Minimum No. of Operable Instrument Reactor Mode Switch Minimum No.

Channels per Position in Which of Tripped or Operable Function Must Be Parameter Operable Trip Systems Trip System (i)

Set Point Operable c

~

Q.

0

"'C CD

I

I 1::

I C

CIS

.c 0>

I t/)

0::

t/)

0::

c.

Downscale 2

3{b)

~ 5 percent of full scale for each scale x

x

d.

Upscale 2

3(b)

S 88 percent of full scale for each scale x

x (3)

APRM

a.

Inoperative 2{h) 3{c) x x

b.

Upscale (Biased by Recirculation Flow) 2(h) 3(c)

Specification 2.1.2a(h) x x

c.

Downscale 2{h) 3{c)

~ [5.28/125] divisions of full scale x

AMENDMENT NO. 142, 143, 153,211 227

NOTES FOR TABLES 3.6.29 and 4.6.2g (a)

Deleted (b)

No more than one of the four IRM inputs to each instrument channel shall be bypassed. These signals may be bypassed when the APRMs are onscale.

(c)

No more than one of the four APRM inputs to each instrument channel shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed. No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to only one APRM shall be bypassed in order for the APRM to be considered operable. In the Run mode of operation, bypass of two chambers from one radial core location in anyone APRM shall cause that APRM to be considered inoperative. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM it is replacing. If one APRM in a quadrant is bypassed and meets all requirements for operability with the exception of the requirement of at least one operable chamber at each radial location, it may be returned to service and the other APRM in that quadrant may be removed from service for test and/or calibration only if no control rod is withdrawn during the calibration and/or test.

(d)

Deleted (e)

Deleted (f)

One sensor provides input to each of two instrument channels. Each instrument channel is in a separate trip system.

(g)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days. Not required to be performed during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run.

(h)

The actuation of either or both trip systems will result in a rod block.

(i)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the Trip System in the tripped condition, provided at least one other operable channel in the same Trip System is monitoring that Parameter.

G)

Neutron detectors are excluded.

AMENDMENT NO. 142.186, 211 232

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 211 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT NUCLEAR STATION, UNIT NO.1 DOCKET NO. 50-220

1.0 INTRODUCTION

By letter dated November 2,2010 (Agencywide Documents Access Management System (ADAMS) Accession No. ML103130037), as supplemented on January 27, 2011 (ADAMS Accession No. ML110400115), Nine Mile Point Nuclear Station, LLC (NMPNS or the licensee) submitted a request for changes to the Nine Mile Point, Unit No.1 (NMP1) Technical Specifications (TSs).

The supplement dated January 27, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination noticed in the Federal Register on March 22, 2011 (76 FR 16007).

The proposed change would modify NMP1 TS Section 3.6.2, "Protective Instrumentation," by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The proposed amendment would eliminate the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable when the reactor mode switch is in the Refuel position. The proposed amendment would also clarify the operability requirements for the APRM Downscale control rod withdrawal block function when the reactor mode switch is in the Startup and Refuel positions.

There is a number of neutron monitoring instrumentation which provides adequate monitoring and protection for reactivity events occurring during the Refuel operating condition; however, these APRM reactor scram and control rod withdrawal block functions, in the refuel operating condition, do not provide any meaningful core protection. In addition, TS Table 3.6.2g would be revised to delete the reference to Note (d) regarding operability of the APRM "Downscale" control rod withdrawal block function with the reactor mode switch in the Startup and Refuel positions. The APRM "Downscale" control rod withdrawal block function is automatically bypassed when the reactor mode switch is in the Startup or Refuel positions; thus, by design, this function is not required when operating with the reactor mode switch in these positions.

- 2 The proposed TS changes will simplify application of the TSs during refueling and startup operations and will facilitate performance of maintenance on Local Power Range Monitoring (LPRM) and APRM equipment.

2.0 REGULATORY EVALUATION

In Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, the Commission established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs.

As stated in 10 CFR 50.36(c)(2)(i), the "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. 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 technical specification... " Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires an LCO to be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3.0 TECHNICAL EVALUATION

3.1 Background

The Neutron Monitoring System monitors the neutron flux level in the reactor using in-core instrumentation in three separate, overlapping ranges: (1) source range monitoring (SRM), (2) intermediate range monitoring (IRM), and (3) power range monitoring. The power range monitoring is accomplished by the LPRM and APRM systems.

The SRM system monitors the reactor neutron flux level during refueling, plant startup, and low power operations over the neutron flux level range from the source level (approximately 3 x 104 neutron volts (nv>> up to 3 X 1012 nv (approximately 8 percent of rated thermal power). The SRM system utilizes 4 instrumentation channels to monitor reactor neutron flux level.

The IRM system monitors reactor neutron flux level during plant startup and the initial phase of power operation (to approximately 40 percent of rated thermal power), over the neutron flux level range from approximately 1.2 x 108 nv to 3.6 x 1012 nv. The IRM system initiates appropriate scram and control rod withdrawal block signals to prevent damage to the fuel from certain abnormal operational transients or operator errors while operating within the intermediate range of power. There are 8 IRM instrument channels. The 8 channels are divided into 2 trip systems, with 4 channels inputting to each trip system.

The APRM system receives the amplified output from 64 of the LPRM detectors, which provide an indication of the power distribution and local power changes. The APRM system automatically initiates reactor scram signals and control rod withdrawal block signals when

- 3 power limits are exceeded or malfunctions occur, thereby protecting the reactor core from high power levels that could cause fuel damage. There are 8 APRM channels, with 2 APRM channels monitoring each core quadrant. Each APRM channel averages 8 LPRM output signals to provide a continuous indication of average reactor power from a few percent to greater than rated thermal power.

3.2 Description of Proposed TS Change Table 3.6.2a. "Instrumentation that Initiates Scram" For Parameter (9)(b), "APRM," TS Table 3.6.2a currently requires that the "Upscale" and "Inoperative" scram functions be operable with the reactor mode switch in the Refuel position.

The proposed amendment would eliminate the requirement that these two APRM scram functions be operable with the reactor mode switch in the Refuel position.

Table 3.6.2g, "Instrumentation that Initiates Control Rod Withdrawal Block" For Parameter (3), "APRM," TS Table 3.6.2g currently requires that the "Inoperative" and "Upscale (Biased by Recirculation Flow)" control rod withdrawal block functions be operable with the reactor mode switch in the Refuel position. The proposed amendment would eliminate the requirement that these two APRM control rod withdrawal block functions be operable with the reactor mode switch in the Refuel position.

For Parameter (3), "APRM," TS Table 3.6.2g currently refers to Note (d) regarding operability of the "Downscale" control rod withdrawal block function with the reactor mode switch in the Startup and Refuel positions. Note (d) states that this function may be bypassed in the startup and refuel positions of the reactor mode switch when the IRMs are onscale. The proposed amendment would delete these references to Note (d) and would delete Note (d) from the table.

3.3 Evaluation of Proposed TS Changes In the 'Technical Evaluation" section of its November 2,2010, submittal, the licensee stated that the APRM system monitors the reactor neutron flux level in the power operating range from a few percent to greater than rated thermal power. The system generates a scram signal at or below 122 percent of the rated value during bulk neutron flux level transients. The system also is capable of generating a control rod withdrawal block signal to mitigate postulated single control rod withdrawal error events. Both the scram and rod block setpoints are varied as a function of reactor recirculation flow. These APRM automatic protective functions prevent damage to the fuel for postulated reactivity insertion events occurring during power operating conditions (including both the Startup and Run positions of the reactor mode switch), such as the Control Rod Withdrawal Error event and the Control Rod Drop Accident.

The proposed amendment does not have any effect on the updated final safety analysis report (UFSAR) analyses for these postulated at-power reactivity insertion events since the TSs will continue to require that the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions remain operable when the reactor mode switch is in the Startup and Run pOSitions.

-4 In the refueling operating condition (defined in TS Definition 1.1.c), the reactor mode switch is in the Refuel position, the reactor coolant system temperature is less than 212 of, and all control rods are inserted in cells containing fuel. Since reactor neutron flux levels during refueling are below the APRM indicating range, the APRM system does not provide any meaningful core monitoring or protection in the Refueling operating condition. The SRM system and the IRM system provide adequate neutron flux monitoring during refueling and automatically initiate protective actions (scram or control rod withdrawal block) when required during refueling.

Operability of the SRM and IRM systems is required when the reactor mode switch is in the Refuel position, in accordance with the TS and UFSAR requirements. The possibility of inadvertent criticality due to a control rod withdrawal error during refueling is minimized by design features, procedural controls, and existing TS requirements, as delineated in TS Table 3.6.2g, TS Sections 3.1.1.a(1), 4.1.1.a(1), and 3.5.2.

The licensee further stated that TS Table 3.6.2g currently refers to Note (d) regarding operability of the APRM "Downscale" control rod withdrawal block function with the reactor mode switch in the Startup and Refuel positions. Note (d) states that this function may be bypassed in the startup and refuel positions of the reactor mode switch when the IRMs are onscale. This statement does not accurately reflect the actual plant design, as it incorrectly implies an optional or selective bypass feature. As described in the NMP1 UFSAR Section VIII-C.1.1.4, the actual design is such that this bypass occurs automatically; thus, by design, the APRM "Downscale" control rod withdrawal block function is not required to be operable with the reactor mode switch in the Startup and Refuel positions. Hence, Note (d) and the references to this Note have been proposed to be deleted from TS Table 3.6.2g. This change is considered administrative in nature since the actual operability requirements have not changed. The APRM "Downscale" control rod withdrawal block function is not credited in the plant safety analyses.

By letter dated January 5,2011 (ADAMS Accession No. ML103610334), the NRC staff requested additional information from the licensee following its acceptance review. The licensee provided supplemental information by letter dated January 27,2011 (ADAMS.

Accession No. ML110400115), in response to the NRC staff's comments. Specifically, the NRC staff requested the licensee to provide justification or analysis to explain if the APRM can be non-operable (as requested in the LAR) while in other modes (e.g., "Shutdown - Cold" or "Shutdown - Hot") and test conditions, as discussed in Section 1.1 of the NMP1 TSs. In addition, if there were any precedents that the NRC staff should consider that allow the APRMs to be non-operational while in "Shutdown Condition - Hot" and I or while reactor coolant system pressure testing, control rod scram time testing, and scram recovery operations are being performed, as appropriate. The licensee stated that there are two reactor operating conditions other than Refueling during which the reactor mode switch may be placed in the Refuel position.

These are: (1) Shutdown Condition-Cold and (2) Shutdown Condition-Hot. With the reactor mode switch in the Shutdown position, electrical interlocks prevent control rod withdrawal.

The purpose of placing the reactor mode switch in the Refuel position while in the Shutdown Cold condition is to permit withdrawal of a single control rod for testing (e.g., control rod scram time testing) or maintenance. The purpose of plaCing the reactor mode switch in the Refuel position while in the Shutdown-Hot condition is to permit withdrawal of a single control rod during reactor coolant system pressure testing, control rod scram time testing, and scram recovery operations with the reactor coolant temperature greater than 212 OF. Placing the

- 5 reactor mode switch in the Refuel position following a scram facilitates scram recovery by permitting individual control rod motion, thereby allowing the operators to fully insert any control rods that have not settled to the full-in position after the scram. The special operating provisions relating to the Shutdown-Hot condition were approved by the NRC by letter dated June 9, 1988, issuing License Amendment No. 99 (ADAMS Accession No. ML011020016).

The NRC staff reviewed the licensee's submittal, the supplemental information, and related documentation (e.g., TS, UFSAR, Improved Standard Technical Specifications (ISTS>>, and conclude that the APRM does not provide any meaningful core monitoring or protection during the Refueling operating condition, as well as, during the special operating conditions discussed above. The SRM and IRM systems (required to be operable when the reactor mode switch is in the Refuel position) provide adequate neutron flux monitoring and automatically initiate protective actions when required. Furthermore, existing design features, procedural controls, and TS requirements minimize the possibility of inadvertent criticality due to a control rod withdrawal error during refueling; and the SRM and IRM systems provide adequate monitoring and core protection if such an event were to occur.

The NMP1 TS allowances to be in the Shutdown-Cold and Shutdown-Hot operating conditions, with the reactor mode switch in the Refuel position, are consistent with the provisions of the BOiling-Water Reactor (BWR) ISTS, NUREG-1433. Specifically, the NUREG-1433 Special Operations Specifications 3.10.3, "Single Control Rod Withdrawal - Hot Shutdown," and 3.10.4, "Single Control Rod Withdrawal-Cold Shutdown," allow placing the reactor mode switch in the Refuel position while in hot shutdown or cold shutdown to permit single rod withdrawal for maintenance and/or testing, by imposing certain restrictions. Those restrictions do not require operability of any APRM functions. The basis for acceptability of these provisions, stated in the NUREG-1433 TS Bases, is that the interlocks that prevent the withdrawal of more than one control rod, together with adequate shutdown margin, will preclude unacceptable reactivity excursions.

The proposed TS changes to the APRM system operability requirements are consistent with 10 CFR 50.36 because the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions are not part of the primary success path to mitigate a design-basis accident or transient during the plant operating conditions allowed by the TSs with the reactor mode switch in the Refuel position. In addition. the APRM "Downscale" control rod withdrawal block function is not part of the primary success path to mitigate a design-basis accident or transient when the reactor mode switch is in the Startup and Refuel positions. The NRC staff, therefore, concludes that the proposed changes are consistent with the regulatory requirements.

In addition, the proposed TS changes are consistent with the ISTS, NUREG-1433, which has been approved by the NRC.

3.4 Conclusion The NRC staff finds the license amendment request acceptable because: (1) the APRM does not provide any meaningful core monitoring or protection during the Refueling operating condition and the special operating conditions discussed above, (2) the SRM and IRM systems provide adequate neutron flux monitoring and automatically initiate protective actions when required, and (3) existing design features, procedural controls, and TS requirements minimize the possibility of inadvertent criticality due to a control rod withdrawal error during refueling; and

-6 (4) the SRM and IRM systems provide adequate monitoring and core protection if such an event were to occur. Therefore, the NRC staff concludes that the proposed amendment for the NMP1 TSs is acceptable.

In addition, the proposed license amendment request was evaluated by the NRC staff to determine whether applicable regulations and requirements continue to be met. It was determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TSs. Applicable regulatory requirements will continue to be met, adequate defense-in-depth will be maintained, and sufficient safety margins will be maintained.

The NRC staff, therefore, finds this license amendment request acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York 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 installation or use of a facility component 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 (76 FR 16007, March 22, 2011). The amendment also relates to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(10). 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) 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: M. Razzaque Date: October 31, 2011

October 31, 2011 Mr. Kenneth Langdon Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 SUB~IECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO.1 - ISSUANCE OF AMENDMENT REGARDING REVISIONS TO AVERAGE POWER RANGE MONITOR INSTRUMENTATION SYSTEM OPERABILITY REQUIREMENTS TECHNICAL SPECIFICATION 3.6.2, PROTECTIVE INSTRUMENTATION (TAC NO. ME5010)

Dear Mr. Langdon:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 211 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No.1 (NMP1), in response to your application dated November 2,2010, (Agencywide Documents and Management System (ADAMS) Accession No. ML103130037), as supplemented on January 27,2011 (ADAMS Accession No. ML110400115).

This amendment revises the NMP1 Technical Specifications (TSs) Section 3.62, "Protective Instrumentation," by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The amendment eliminates the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable when the reactor mode switch is in the Refuel position. The amendment also clarifies the operability requirements for the APRM Downscale control rod withdrawal block function when the reactor mode switch is in the Startup and Refuel positions.

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

Sincerely, IRAI Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 211 to DPR-63

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

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