ML13168A025

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Issuance of Amendment No. 173 to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-476, Revision 1, Improved BPWS Control Rod Insertion Process (NEDO-33091)
ML13168A025
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 07/15/2013
From: Beltz T
Plant Licensing Branch III
To: Schimmel M
Northern States Power Co
Beltz T
References
TAC ME9600
Download: ML13168A025 (12)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 15, 2013 Mr. Mark A. Schimmel Site Vice President lVIonticelio Nuclear Generating Plant Northern States Power Company - Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT -ISSUANCE OF AMENDMENT NO. 173 TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE (TSTF)

TRAVELER TSTF-476, REVISION 1, "IMPROVED BPWS CONTROL ROD INSERTION PROCESS (NEDO-33091)" (TAC NO. ME9600)

Dear Mr. Schimmel:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 173 to Renewed Facility Operating License No. DPR-22 forthe Monticello Nuclear Generating Plant (MNGP). The amendment consists of. changes to the technical specifications (TSs) in response to your application dated September 18, 2012.

The amendment approves a revision to the MNGP TS Sections 3.1.6, "Rod Pattern Control,"

and 3.3.2.1, "Control Rod Block Instrumentation," to allow MNGP to reference a new Banked Position Withdrawal Sequence (BPWS) shutdown sequence in the TS Bases. Additionally, a footnote is revised in Table 3.3.2.1-1, "Control Rod Block Instrumentation," to allow operators to bypass the rod worth minimizer if conditions for the optional BPWS shutdown process are satisfied. The changes are consistent with NRC-approved TSTF-476, Revision 1, "Improved BPWS Control Rod Insertion Process (NEDO-33091 )."

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 173 to DPR-22
2. Safety Evaluation cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 173 License No. DPR-22

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

A. The application for amendment by Northern States Power Company (NSPM, the licensee), dated September 18, 2012, 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-22 is hereby amended to read as follows:

Technical SpeCifications The Technical Specifications contained in Appendix A. as revised through Amendment No. 173, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

- 2

3. This 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 REGULATORY COMMISSION

l Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Re.actor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License DPR-22 and Technical Specifications Date of Issuance:uLily 15, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 173 RENEWED FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following page of Renewed*Facility Operating License DPR-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. .

REMOVE INSERT 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 a marginal line \

indicating the area of change.

REMOVE INSERT 3.3.2.1-5 3.3.2.1-5

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2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 1974 (those portions dealing with handling of reactor fuel) and August 17, 1977 (those portions dealing with fuel assembly storage capacity);
3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for

, reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, .

possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

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

C. This renewed operating license shall be deemed to <;:ontain and is subject to the conditions specified in the Commission's regulations in, 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the'rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: . .

1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1775 megawatts (thermal).
2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 173, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.
3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 4--tRru 173

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FI,JNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor
a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 As specified in SR 3.3.2.1.4(h)(i) COLR SR 3.3.2.1.5
b. Intermediate Power Range (b) 2 SR 3.3.2.1.1 As specified in Upscale SR 3.3.2.1.4(h)(i) COLR SR 3.3.2.1.5
c. High Power Range - Upscale (c), (d) 2 SR 3.3.2.1.1 As specified in SR 3.3.2.1.4(h)(i) COLR SR 3.3.2.1.5
d. Inop (d), (e) 2 SR 3.3.2.1.1 NA
2. Rod Worth Minimizer 1(1),2(1} SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8
3. Reactor Mode Switch - Shutdown (g) 2 SR 3.3.2.1.7 NA Position (a) THERMAL POWER ~ 30% and < 65% RTP and MCPR is below the limit specified in COLR.

(b) THERMAL POWER ~ 65% and < 85% RTP and MCPR is below the limit specified in COLR.

(c) THERMAL POWER ~ 85% and < 90% RTP and MCPR is below the limit specified in COLR.

(d) THERMAL POWER ~ 90% RTP and MCPR is below the limit specified in COLR.

(e) THERMAL POWER ~ 30% and < 90% RTP and MCPR is below the limit specified in COLR.

(f) With THERMAL POWER ~ 10% RTP, except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed.

(g) Reactor mode switch in the shutdown position.

(h) If the as-found channel setpoint is not the Nominal Trip Setpoint (NTSP) but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(i) The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP shall be specified in the COLR.

The methodology used to determine the NTSP is specified in the Technical Requirements Manual.

Monticello 3.3.2.1-5 Amendment No. 146, 159, 173

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 173 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC) dated September 18,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML122630063), Northern States Power Company - Minnesota (NSPM), doing business as Xcel Energy, Inc (the licensee), submitted a license amendment request regarding changes to Monticello Nuclear Generating Plant (MNGP) technical specifications (TS) Sections 3.1.6, "Rod Pattern Control," and 3.3.2.1, "Control Rod Block Instrumentation." According to the licensee's letter, the proposed changes would revise the Bases section of TS 3.1.6, "Rod Pattern Control,"

and 3.3.2.1, "Control Rod Block Instrumentation," in addition to adding a footnote to TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," to allow reference to an improved, optional Banked Position Withdrawal Sequence (BPWS) for use during reactor shutdown.

The licensee states that the proposed changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-476, Revision 1, "Improved BPWS Control Rod Insertion Process (NEDO-33091 )." The availability of this TS improvement was announced in the Federal Register on May 23, 2007 (72 FR 29004) as part of the consolidated line item improvement process.

2.0 REGULATORY EVALUATION

The control rod drop accident (CRDA) is the design basis accident for the subject TS changes.

In order to minimize the impact of a CRDA, the BPWS process was developed to minimize control rod reactivity worth for boiling-water reactor (BWR) plants. The proposed improved BPWS further simplifies the control rod insertion process and, in order to evaluate it, the NRC staff followed the guidelines of Standard Review Plan Section 15.4.9, and referred to General Design Criterion (GDC) 28 of Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 as its regulatory requirement. GDC 28 states that the reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity Enclosure

- 2 increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.

In its application, the licensee provided the following plant-specific discussion for meeting the NRC staff's TSTF-476 regulatory evaluation stated above:

The model Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDG).

MNGP was designed largely before the publishing of the 70 General Design Criteria (GDG) for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission (AEC) for public comment in July 1967, and construCted prior, to the 1971 publication of the 10 CFR 50, Appendix A, GDC.

As such, MNGP was not licensed ~o the Appendix A, GDC.

The MNGP USAR [Updated Safety Analysis Report], Section 1.2, lists the Principal Design Criteria (PDC) for the design, construction and operation of the plant. MNGP USAR Appendix E provides a plant comparative evaluation to the 70 proposed AEC design criteria. It was concluded that the plant conforms to the intent of the GDC. The applicable GDC and PDC are discussed below.

  • PDC 1.2.2 - Reactor Core
d. Power excursions which could result from any credible reactivity addition accident do not cause damage, either by motion or rupture, to the reactor' vessel or impair operation of required safeguards.

The applicable 70 Draft AEC General Design Criteria (AEC-GDG) are:

  • Criterion 31 - Reactivity Control Systems Malfunction (Category B)

The reactivity control systems shall be capable of sustaining any single malfunction, such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.

  • Criterion 32 - Maximum Reactivitv Worth of Control Rods (Category A)

Limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.

The licensee further stated that "as discussed in the model Safety Evaluation for TSTF-476, Rev. 1 (72 FR 29004), 10 CFR 50, Appendix A, Criterion 28 states that the reactivity control

-3 systems shall be designed with appropriate .limits on the poten~ial amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. The NRC staff concluded that proper use of the improved BPWS will prevent" a Control Rod Drop Accident (CRDA) from occurring while power is below the Low Power Setpoint (lPSP). The proposed TS changes do not alter the means of compliance with GDC28."

In its application, the licensee confirmed compliance with the NRC staff's regulatory evaluation by stating that "NSPM has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. The proposed TS changes will continue to assure that the design requirements'and acceptance criteria for MNGP are met."

3.0 TECHNICAL EVALUATION

3.1 Licensee's Proposed Changes Consistent with NRC-approved TSTF-476, Revision 1, the licensee proposed the following plant specific changes to its TSs and Bases:

  • Revise TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," to revise a footnote that allows operators to bypass the rod worth minimizer if conditions for the optional BPWS shutdown process are satisfied.

3.2 Staff Evaluation The TSTF-476, Revision 1, states that the improved BPWS provides the following benefits:

(1) allows the plant to reach the all-rods-in condition prior to significant reactor cool down, which reduces the potential for re-criticality as the reactor cools down; (2) reduces the potential for an operator reactivity control error by reducing the total number of control rod manipula,ions; (3) minimizes the need for manual scrams during plant shutdowns, resulting in less wear on control rod drive (CRD) system components and CRD mechanisms; and, (4) eliminates unnecessary control rod manipulations at low power, resulting in less wear on reactor manual control and CRD system components.

As part of the implementation process for the license amendment that revises MNGP's TS Table. 3.3.2.1-1, "Control Rod Block Instrumentation," footnote (f), "except during the reactor shutdown process if the coupling of each withdrawn control rod has been confirmed," the staff's evaluation for the adoption of TSTF-476 requires that appropriate detailed operational procedures will be established by a licensee to implement the controls included in Section 4

-4 "Safety and Technical Evaluations" and Section 5 "Plant Implementation" of NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process."

In order to comply with the above requirement, the licensee states in its application that, "As discussed in the model SE published in the Federal Register on May 23, 2007 (72 FR 29004) for this Technical Specifications improvement, the following plant-specific verification/commitments were performed." The safety evaluation for NEDO-33091-A explained that the potential for a control rod drop accident (CRDA) will be eliminated by changes to operational procedures.

The licensee committed to the following procedural changes prior to implementation:

1. Before reducing power to the low power setpoint (LPSP), operators shall confirm control rod coupling integrity for all rods that are fully withdrawn. Control rods that have not been confirmed coupled and which are in intermediate positions must be fully inserted prior to power reduction to the LPSP. No action is required for fully-inserted control rods.

If a shutdown is required and all rods which are not confirmed coupled cannot be fully inserted prior to the power dropping below the LPSP, then the original/standard BPWS must be adhered to.1

2. After reactor power drops below the LPSP, rods may be inserted from notch position 48 to notch position 00 without stopping at the intermediate positions.

However, it is recommended that operators insert control rods in the same order as specified for the original/standard BPWS as much as reasonably possible.

When in the process of shutting down following improved BPWS with the power below the LPSP, no control rod shall be withdrawn unless the control rod pattern is in compliance with standard BPWS'requirements.

The licensee further states, "In addition to the procedure changes specified above, the staff previously concluded, based on its review of NEDO-33091-A, that no single failure of the boiling water reactor control rod drive (CRD) mecha'nical or hydraulic system can cause a control rod to drop completely out of the reactor core during the shutdown process. Therefore, the proper use of the improved BPWS will prevent a CRDAfrom occurring while power is below the LPSP."

The licensee confirms by stating, "NSPM /las verified, in accordance wit Ii NEDO-33091-A, Revision 2, that no single failure of the boiling water reactor CRD mechanical or hydraulic system can cause a control rod to drop completely out of the reactor core during the shutdown process." ' ,

The NRC staff has reviewed the NSPM's change to its TS Table 3.3.2.1-1, item f. The staff found that the licensee's above commitments to change its operational procedures are consistent with the staff's approved TSTF-476 evaluation. The NRC staff also reviewed the TSTF-476, Revision 1, bases and similarly concluded that the TSTF is applicable to NSPM.

1 The original/standard BPWS can be found in Licensing Topical Report (LTR) NEDO-21231, "Banked Position Withdrawal Sequence," January 1977, and is referred to in NUREG-1433 and NUREG-1434,

-5 Therefore, the TSTF is appropriate for adoption by the licensee. In addition, the NRC staff reviewed the licensee's proposed changes against the corresponding changes made to the Standard Technical Specifications by TSTF-476-A, Revision 1, which the staff found to satisfy applicable regulatory requirements as described above. The NRC staff found that the proposed changes are consistent with those approved by the staff in the subject TSTF. The NRC staff, therefore, has determined that the changes are acceptable.

4.0 STATE CONSULTATION

In accordCince with the Commission's regulations, the Minnesota 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 use of facility components 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 (77 FR 73688, dated December 11, 2012). 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 Contributor: R. P. Grover, NRR Date of issuance: July 15, 2013"

July 15, 2013 Mr. Mark A. Schimmel Site Vice President Monticello Nuclear Generating Plant Northern States Power Company - Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT NO. 173 TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE (TSTF)

TRAVELER TSTF-476, REVISION 1, "IMPROVED BPWS CONTROL ROD INSERTION PROCESS (NEDO-33091)" (TAC NO. ME9600)

Dear Mr. Schimmel:

~he U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 173 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant (MNGP). The amendment consists of changes to the technical specifications (TSs) in response to your application dated September 18, 2012.

The amendment approves a revision to the MNGP TS Sections 3.1.6, "Rod Pattern Control,"

and 3.3.2 . 1, "Control Rod Block Instrumentation," to allow MNGP to reference a new Banked Position Withdrawal Sequence (BPWS) shutdown sequence in the TS Bases. Additionally, a footnote is revised in Table 3.3.2.1-1, "Control Rod Block Instrumentation," to allow operators to bypass the rod worth minimizer if conditionsfor the optional BPWS shutdown process are satisfied. The changes are consistent with NRC-approved TSTF-476, Revision 1, "Improved BPWS Control Rod Insertion Process (NEDO-33091)."

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRAJ Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 173 to DPR-22
2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:

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DATE 06/17/13 06/27/H 02/21/13 07/10/13 07/15/13 07/15/13 OFFICIAL RECORD COPY

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