ML12249A317

Issuance of Amendment One-Time Technical Specification Change Regarding Core Spray Operability During Shutdown

ML12249A317
Person / Time
Site:	Duane Arnold
Issue date:	09/27/2012
From:	Feintuch K Plant Licensing Branch III
To:	Richard Anderson NextEra Energy Duane Arnold
Feintuch K
References
TAC ME8572
Download: ML12249A317 (22)

Text

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"'I", °11 NUCLEAR REGULATORY COMMISSION t:! 0 WASHINGTON, D.C. 20555-0001

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"7 ~O Mr. Richard L. Anderson Vice President Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324-9785

SUBJECT:

DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT RE:

ONE-TIME TECHNICAL SPECIFICATION CHANGE REGARDING CORE SPRAY OPERABILITY DURING SHUTDOWN (TAC NO. ME8572)

Dear Mr. Anderson:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 283 to Renewed Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC).

This amendment consists of a change to the Technical Specifications (TSs) and Renewed Facility Operating License in response to your application dated May 1, 2012, as supplemented by letters dated June 27,2012, and July 26,2012.

The amendment revises existing TS 3.3.5.1, on a one-time basis only, by adding a note to TS Table 3.3.5.1-1, Function 1d, Modes 4 and 5. This one-time amendment enables DAEC to re coat the internal surface of the Suppression Chamber during Refueling Outage 23.

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

Sincerely, Karl D. Feintuch, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 283 to License No. DPR-49

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

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. 283 Renewed 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 May 1, 2012, as supplemented by letters dated June 27,2012, and July 26, 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;

8. 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

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. 283, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC, shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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I Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Attachments: Changes to the License and Technical Specifications Date of Issuance: September 27, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 283 RENEWED FACILITY OPERATING LICENSE NO. DPR-49 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 marginal lines indicating the areas of change.

Remove 3 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove 3.3-41 3.3-41

-3 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. 283, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC 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 NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the Duane Arnold Energy Center and as approved in the SER dated June 1, 1978, and Supplement dated February 10, 1981, subject to the following provision:

NextEra Energy Duane Arnold, LLC may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(4) The licensee is authorized to operate the Duane Arnold Energy Center following installation of modified safe-ends on the eight primary recirculation system inlet lines which are described in the licensee letter dated July 31, 1978, and supplemented by letter dated December 8, 1978.

(5) Physical Protection NextEra Energy Duane Arnold, LLC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, Renewed License No. DPR-49 Amendment 283

EGGS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A1 REQUIREMENTS VALUE

1. Core Spray System

a. Reactor Vessel Water B SR 3.3.5.1.1  ::: 38.3 inches Level - Low Low Low SR 3.3.5.1.3 SR 3.3.5.1.8 SR 3.3.5.1.9

b. Drywell Pressure 1,2,3 4(b) B SR 3.3.5.1.3 High SR 3.3.5.1.8 SR 3.3.5.1.9

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.3  ::: 363.3 psig Pressure - Low SR 3.3.5.1.8 and::: 4851 psig (Injection Permissive) SR 3.3.5.1.9 4 B SR 3.3.5.1.3 "'- 363.3 psig SR 3.3.5.1.8 SR 3.3.5.1.9 and :s. 485.1 psig

d. Core Spray Pump 1 per E SR 3.3.5.1.3  :: 256.6 gpm Discharge Flow - Low pump(e) SR 3.3.5.1.8 and (Bypass) SR 3.3.5.1.9 ~ 2382.1 gpm

e. Core Spray Pump Start 1,2,3, 1 per C SR 3.3.5.1.8  ?: 2.6 seconds Time Delay Relay 4(a),5(a) pump SR 3.3.5.1.9 and :s. 6.8 seconds

f. 4.16 kV Emergency Bus 1,2,3, 1 per F SR 3.3.5.1.5 ~ 3500 V Sequential Loading 4(a),5(a) pump SR 3.3.5.1.6 Relay SR 3.3.5.1.9

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel Water 4 B SR 3.3.5.1.1 ': 38.3 inches Level- Low Low Low SR 3.3.5.1.3 SR 3.3.5.1.8 SR 3.3.5.1.9

b. Drywell Pressure 1,2,3 4 B SR 3.3.5.1.3 High SR 3.3.5.1.8 SR 3.3.5.1.9 (continued)

(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, ECCS-Shutdown.

(b) Also required to initiate the associated Diesel Generator (DG).

(e) During Refuel Outage (RFO) 23, the MODE 4 and 5 requirement for Function 1.d is revised to be zero (0) required channels per pump.

DAEG 3,3-41 Amendment 283

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 283 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 NEXTERA ENERGY DUANE ARNOLD, LLC. FOR DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

By letter dated May 1, 2012 (Reference 1), NextEra Energy Duane Arnold, LLC (the licensee) has requested an amendment to Renewed Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC). The licensee provided supplemental information to its request by letters dated June 27,2012, and July 26,2012 (References 2 and 3, respectively), in response to questions from the U.S. Nuclear Regulatory Commission (NRC) staff. The proposed change would revise the DAEC Technical Specifications (TSs), on a one-time basis only, by adding a note to TS Table 3.3.5.1-1, Function 1.d, Modes 4 and 5.

The supplements dated June 27, 2012, and July 26,2012, 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 or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 10, 2012 (77 FR 40654).

Background

The licensee has a license renewal commitment for DAEC to re-coat the internal surface (outer shell walls) of the Suppression Chamber (also identified as the Torus) prior to entering the period of extended operation in February 2014. The licensee stated that the upcoming Refueling Outage (RFO) 23 is the only planned outage available to fulfill that commitment.

During the Torus re-coat project, all of the water will be removed from the Torus such that there will be no Suppression Pool (SP) available as an Emergency Core Cooling System (ECCS) suction source for a Significant portion of the outage. The Residual Heat Removal (RHR) system suction piping cannot be cross-tied to the Condensate Storage Tanks (CSTs); therefore, no RHR pumps will be available to satisfy Limiting Condition for Operation (LCO) 3.5.2, "ECCS

- Shutdown," as the SP will be drained below the water level specified in TS Surveillance Requirement (SR) 3.5.2.1 (i.e.,,:::: 7.0 ft). In order to satisfy LCO 3.5.2 during the stated Applicability (MODE 4 and MODE 5, except with the spent fuel storage pool gates removed and water level.:::: 21 feet, 1 inch over the top of the reactor pressure vessel flange), the licensee will need to rely exclusively on the two Core Spray (CS) subsystems for meeting this LCO. During Enclosure

- 2 the Torus re-coat project, however, it will be necessary to isolate the CS minimum flow path return line to the Torus to preclude introducing water from the CSTs during the re-coating process. The Torus must be completely drained to strip off the existing coating, to re-apply the new coating, and to allow the time necessary for the curing process.

The CS minimum flow valves are "normally open" and only close when CS flow is > 600 gallons per minute (gpm) (nominal) is sensed. When the CS pump is not running, the minimum flow instruments send an open signal to the valves. The current TS requires this minimum flow path to be available in order to declare the CS subsystems Operable - LCO 3.3.5.1 (ECCS Instrumentation) in Table 3.3.5.1-1, Function 1.d, requires the CS minimum flow logic to be Operable in Modes 4 and 5 whenever CS is required to be Operable per LCO 3.5.2 and SR 3.5.2.4 requires the minimum flow valve to be Operable, as well. Thus, during the Torus re coat process, Le., with the SP drained, and the two CS subsystems aligned to the CSTs for suction, there will be no Operable ECCS subsystems. It should be noted, however, that while the two CS sUbsystems will be declared inoperable based on current TS language, the two CS subsystems will be fully functional for injection during this time period. Given the specific hardship that this imposes on the DAEC, the licensee is requesting a one-time only TS change to no longer require the CS minimum flow logic (3.3.5.1, Function 1.d) and the companion minimum flow valves (SR 3.5.2.4) to be Operable in Modes 4 and 5 in order to declare the CS subsystems Operable for meeting LCO 3.5.2 during RFO 23.

Proposed Change The proposed amendment would revise the DAEC TS, on a one-time basis only, by adding a note to TS Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation," Function 1.d, Modes 4 and 5. Specifically, the note would state the following:

(e) During Refuel Outage (RFO) 23, the MODE 4 and 5 requirement for Function 1.d is revised to be zero (0) required channels per pump.

2.0 REGULATORY EVALUATION

2.1 Reactor Systems Safety Aspects of the Regulatory Evaluation Title 10 of the Code of Federal Regulations (10 CFR) 50.36, 'Technical specifications," requires that the facility's TS will include a section addressing LCO. In accordance with 10 CFR 50.36(c)(2)(ii), the LCO of a nuclear reactor must be established for each item meeting one or more of the specified criteria. One of these criteria is Criterion 3 which requires an LCO 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. Further, paragraph (c)(3) states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

The proposed one-time change during RFO 23 to add a note to TS Table 3.3.5.1-1, Function 1.d, Modes 4 and 5, does not invalidate the requirements delineated in LCO 3.5.2 and SR 3.5.2.2 during Operations with Potential for Draining the Reactor Vessel (OPDRV) and, therefore, is consistent with 10 CFR 50.36(c)(3).

-3 2.2 Human Performance Aspects of the Regulatory Evaluation The regulatory requirements and guidance which the NRC staff considered in its review of the License Amendment Request (LAR) are as follows:

2.2.1 Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria (GDC)," Criterion 19 - Control room. "A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents .... Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures."

2.2.2 10 CFR 50.120, "Training and qualification of nuclear power plant personnel" 2.2.3 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition":

Chapter 13 addresses "Conduct of Operations", specific sub-chapters considered in this review were Chapters 13.2.1, "Reactor Operator Requalification Program; Reactor Operator Training", and 13.5.2.1, "Operating and Emergency Operating Procedures" .

Chapter 18 provides review guidance for "Human Factors Engineering" 2.2.4 NUREG-1764, "Guidance for the Review of Changes to Human Actions" 2.2.5 GL 82-33, "Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability" 2.2.6 NUREG-0700, "Human-System Interface Design Review Guidelines," Revision 2 2.2.7 NUREG-0711, "Human Factors Engineering Program Review Model," Revision 2 2.2.8 IN 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times" In accordance with the generic risk categories established in Appendix A to NUREG-1764, the task sequence of, "initiating Core Spray", reviewed herein, is considered "risk-important" due to the fact that it is required to prevent core uncovering when the decay heat removal function is lost. Because of its risk importance, the NRC staff performed a "Level One" review (I.e., the most stringent of the graded reviews possible under the guidance of NUREG-1764).

Note: The NRC assessment of risk applied to this Safety Evaluation is only for purposes of scoping the NRC review and may conflict with the licensee's assessment of risk importance or that of other NRC Branches applied for other purposes. It should not be considered as an

-4 accurate assessment of risk when compared to other methods, especially those using plant specific data and NRC-accepted methods of Probabilistic Risk Analysis and Human Reliability Analysis.

The NRC staff's technical evaluation of the proposed TS changes is provided below.

3.0 TECHNICAL EVALUATION

3.1 Reactor Systems Safety Aspects of the Technical Evaluation In the "Technical Evaluation" section of its May 1, 2012 submittal, the licensee stated that during MODEs 4 and 5, the reactor is fully depressurized and the low pressure permissive (::; 450 pounds per square inch gauge nominal) for the CS injection valves is always satisfied, i.e., there is virtually no time between the CS initiation signal (low-low-low reactor vessel water level::; 64" nominal) and injection valve opening. The CS pump starts are delayed for approximately five seconds (LCO 3.3.5.1, Function 1.e). The CS injection motor-operated valves are expected to be substantially open prior to the CS pump start to allow full flow to the reactor, and the likelihood of "dead-heading" the CS pump is negligible. Consequently, there is no functional need for the minimum flow logic and valve{s) to be operable in MODEs 4 and 5 as this pump start sequence poses no risk to pump overheating. In addition, with the minimum flow path secured (manual valves closed), it is expected that there will be no diversion of CS flow away from the reactor.

The licensee further stated that plant procedures will be enhanced to ensure a CS pump flow path is established and flow verified to be 2: 600 gpm immediately following CS initiation and maintained as such during pump operation. The manual startup/initiation procedure will be temporarily revised for RFO 23. Direct indication for CS pump flow is available and alternate indications for confirming CS pump flow 2: 600 gpm, such as minimum flow valve position which will be available as long as the associated instrumentation remains in service, will be used. If a flow path and required flow rate cannot be confirmed, the pump shall promptly be secured.

Short duration operation of a CS pump at pump shutoff head (i.e. dead headed), sufficient for the operators to diagnose the condition and secure the pump, is not expected to be detrimental to the pump. However, it is expected that extended operation beyond a few minutes can cause pump damage.

As a consequence, potential human errors that could result from this activity are:

1. CS pump damage from prolonged low flow or dead-head operation.

2. Diversion of CS injection flow through the minimum flow line as a result of failing to ensure the line is isolated.

In order to minimize potential human error during RFO 23, the licensee committed to strict administrative and procedural controls, operator training, and use of human performance tools that are essential in preventing these types of consequential human errors. The licensee further committed to "guard" both CS subsystems, such that no work or testing will be permitted on either of the CS subsystems during RFO 23 when both CS subsystems are required to be OPERABLE to comply with LCO 3.5.2.

- 5 In SR 3.5.2.2.b, a NOTE states that only one required CS subsystem (out of two required low pressure subsystem) may take credit for aligning to the CST (required water level in one CST is

.:::. 11 ft or.:::. 7 ft in both CSTs) during OPDRVs. In response to a question from the NRC staff (Reference 2), the licensee stated that DAEC will continue to comply with the existing TS requirements of LCO 3.5.2 and consequently, the licensee did not request relief specifically from this SR requirement in their application. In their response, the licensee further confirmed that the proposed TS change only affects the minimum flow path (logic and valves) for the CS system, and that there are no requested changes in how the CSTs can be credited as a suction source for complying with LCO 3.5.2.

Additionally, the licensee confirmed (Reference 2) that they will follow the actions contained in NRC Enforcement Guidance Memorandum (EGM) 11-003 regarding OPDRVs when Secondary Containment is not Operable in MODE 5 (Reference 4). Accordingly, the licensee stated that all OPDRV activities at DAEC during RFO 23 will be scheduled to take place when either:

1. a. The Reactor Pressure Vessel (RPV) cavity is fully flooded up (I.e., when LCO 3.5.2 is no longer required to be met and the SR 3.5.2.2 Note is no longer applicable),

AND

b. Secondary Containment is Operable.

OR

2. If the cavity is not flooded up (I.e., the LCO 3.5.2 Applicability is met and the Note to SR 3.5.2.2 applies):

a. The minimum water level in the Suppression Pool is adequate, per SR 3.5.2.2, to support Residual Heat Removal (RHR) pump(s} for meeting LCO 3.5.2 (I.e., the Torus is not completely drained),

AND

b. Secondary Containment is Operable.

In summary, the licensee confirmed that no credit will be taken during RFO 23 for both CS pumps, with their suction piping aligned to the CST, for complying with LCO 3.5.2 during OPDRVs (Reference 2).

Further in their response (Reference 2), the licensee clarified that the draining of the SP in MODEs 4 and 5 does not necessarily result in a loss of decay heat removal to the ultimate heat sink, as the RHR System will be Operable for Shutdown Cooling mode in accordance with TS LCOs, 3.4.8, 3.9.7, and 3.9.8. In the event an unexpected drain down event does occur when the Torus is drained, there is a large inventory of makeup water available for mitigation.

Because DAEC is a "zero release" plant for liquid radioactive effluents, none of the primary reactor system water inventory is discarded; particularly, the approximately 400,000 gallons of water removed from the Torus during the re-coat project. At various times during the RFO, a total of approximately 1.1 Million gallons of water is stored in a combination of locations, such as

-6 the CSTs, RPV cavity, main condenser, Torus, and Radioactive Waste building. Therefore, the licensee states that DAEC is not solely reliant on the fixed amount of inventory in the CSTs for make-up capability during RFO 23.

The NRC staff has reviewed the licensee's submittal, supplemental information provided in response to the NRC staff's questions (Reference 2), and related documentation. The NRC staff understands that the proposed TS change for DAEC applies only for RFO 23, and that it does not invalidate or contradict LCO 3.5.2, SR 3.5.2.2, related TS requirements, and the requirements contained in NRC's EGM 11-003 regarding OPDRV. The NRC staff also understands that the proposed TS change for RFO 23 has potential for allowing human errors to occur that can cause CS pump damage from prolonged low flow or dead-head operation, or diversion of CS injection flow through the minimum flow line. In order to prevent and mitigate such human errors during RFO 23, the licensee committed to strict administrative and procedural controls, operator training, and use of human performance tools. The licensee also committed to guard both CS subsystems, such that no work or testing will be permitted on either of the CS subsystems during RFO 23 when both CS subsystems are needed to be OPERABLE to meet the requirements of LCO 3.5.2.

The NRC staff finds that the proposed license amendment request is acceptable based on the following considerations: 1) The proposed TS change applies only for RFO 23 and the change does not invalidate or contradict other TS requirements applicable for OPDRV, 2) the TS change applies only for a limited period of time (i.e., RFO 23 only) and that the licensee provides reasonable assurance that specific and timely measures will be taken during this period to prevent and minimize potential human errors, 3) draining of SP in MODEs 4 and 5 does not result in a loss of decay heat removal to the ultimate heat sink, as the RHR system will be Operable for Shutdown Cooling mode, and that 4) in the event an unexpected drain down event should occur when the Torus is drained, there is adequate inventory of makeup water available at DAEC in addition to the fixed amount of inventory in the CSTs. The NRC staff concludes that the proposed amendment of the TS continue to meet the requirements delineated in LCO 3.5.2, SR 3.5.2.2, and related regulatory requirements during OPDRVs in RFO 23.

Therefore, the proposed TS amendment is acceptable from a technical perspective.

3.2 Human Performance Safety Aspects of the Technical Evaluation 3.2.1 Description of Operator Action(s) Added/Changed/Deleted Due to the fact that Core Spray will be the only safety system available for accident mitigation while the internal Suppression Chamber surface is being re-coated during RFO 23, the following actions will be required. Additionally, a note is to be added to TSs to clarify that these actions are allowed/required only during RFO 23.

Pre-staged actions:

1. a. Isolate the CS minimum flow path to the Suppression Chamber;

1. b. Ensure that one CS pump flow path to the reactor is established;

-7 Post-initiation actions:

1. c. Verify flow to be > 600 gpm immediately following CS initiation;

1. d. If CST level falls too low, align alternate sources of coolant.

These actions are required for the following reasons:

Action 1.a. - Performed to ensure that personnel in the Suppression Chamber are not harmed by CS flow being directed to the Suppression Chamber where they will be working, and to prevent flow from being diverted from the reactor.

Action 1.b. - Performed to ensure that a flow path is pre-staged for immediate injection upon a CS initiation signal.

Action 1.c. - Needed to confirm that CS has initiated and is providing sufficient flow.

Action 1.d. - Needed to monitor the inventory of coolant in the CST, and to re-configure to an alternate source before the inventory in the CST falls to the point where adequate flow is jeopardized and air entrainment is possible.

The licensee has stated that no other actions are required to support the LAR. The NRC staff agrees that the descriptions of the actions added, changed, and deleted are adequate.

3.2.2 Operating Experience Review The licensee stated in its LAR that its review of operating experience revealed that a Licensee Event Report (LER) 266/01-005, dated January 28,2002, associated with the Point Beach Nuclear Plant (Agencywide Documents Access and Management System (ADAMS) Accession No. ML020560352) was applicable to the proposed LAR. The LER provided insight regarding the danger of actuating a pump with the flow path isolated (dead-heading). The LER highlighted the possibility of destroying the CS injection pumps (the only available accident mitigation system) by inadvertently dead-heading them. As a result, actions were added to the relevant procedures to minimize such errors.

The NRC staff agrees with the licensee's assessment and finds the licensee's operating experience review to be acceptable.

3.2.3 Functional Requirements Analysis and Function Allocation Because the operator actions are not new actions, a re-analysis of the functional requirements analysis is not necessary. Additionally, in this case, a review of allocation of function is not necessary because the actions are backup actions to the automatic CS initiation and are by their nature manual actions and could not be allocated otherwise. So, there was no need for either a new or revised functional requirements analysis or a reallocation of function.

The NRC staff finds the licensee's approach acceptable based on the fact that there are no changes to any of the actions associated with the proposed LAR.

-8 3.2.4 Task Analysis Because the operator actions are not new actions, the task analysis that provides the operator requirements need not be revised. Based on this review, and reviews of station procedures and standards, no issues were identified that could add to the workload of operators in a manner that would prevent them from timely initiation of Core Spray if automatic initiation failed.

The NRC staff concludes that revision of the licensee's task analysis is not necessary, because the actions associated with this proposed LAR are not new and have not changed.

3.2.5 Staffing Licensee staffing and qualifications are not affected by the proposed LAR. No new or additional staff is required, nor are there any new or additional qualifications necessary to perform the action sequence within the time constraints established.

The NRC staff agrees that no additional staffing or qualifications, or changes thereto, are required, and finds this human performance aspect of the LAR to be acceptable.

3.2.6 Probabilistic Risk and Human Reliability Analyses The Licensee submitted an application based on deterministic analysis.

Therefore, no probabilistic risk and human reliability review was performed.

3.2.7 Human-System Interface Design The licensee stated in its July 26, 2012 submittal that all required controls, displays, and alarms are available in the control room and will not be changed. Human-System Interface (HSI) design of the control room, remote shutdown area, and the simulator, including the design of the Safety Parameter Display System will not be affected by the proposed LAR. The same controls, displays, and alarms that have been successfully used in the past will continue to be used under the proposed LAR.

Based on the fact that no changes are needed to the HSI design, the NRC staff finds this aspect of the LAR to be acceptable.

3.2.8 Procedure Design The licensee stated in its July 26, 2012 submittal that one procedure is affected by the proposed LAR. The changes affect Operating Instruction 151, Core Spray System. The procedure will be revised for Refueling Outage 23 (RFO 23) to note the change in CS operability status and to detail strategies and precautions for operating the system with no minimum flow line available.

Additionally, the licensee is guarding the affected equipment using administrative controls outlined in operating procedures, OP-AA-1 02-1 003, "Guarded Equipment" and OP-AA-102 1003, "Protected Trains and Guarded Equipment".

The NRC staff finds the licensee's position is acceptable based on the fact that personnel actions required to support the proposed LAR have been incorporated into the draft procedure

- 9 which was attached to the licensee's July 26,2012, submittal, and that the steps in the draft procedure are clear and easy to understand.

3.2.9 Training Program Design Because the proposed LAR is a one time change, operator training programs will not be permanently modified. However, training will be provided in preparation for RFO 23 and will include a review of applicable revisions to the CS operating procedure. The Shutdown Safety Plan for RFO 23, which delineates available alternate injection sources and emphasizes the importance of minimum flow protection during times when CS is a primary makeup source, will be discussed during training.

The licensee stated that the practice of using CS as a makeup source without minimum flow protection is not new. The amendment request only changes the operability state of CS, and not how the system is operated. Based on the fact that the action sequence will continue to be included in the training program and that the training will be implemented prior to amending the TS, the NRC staff finds that the training to be provided is acceptable.

3.2.10 Human Factors Verification and Validation Per NUREG-0711, the scope of human factors engineering design verification may be restricted to the modified HSls and their interactions with the rest of the HSls. NUREG-0711 also states that, "Integrated system validation may not be needed when a modification results in minor changes to personnel tasks such that they may reasonably be expected to have little or no overall effect on workload and the likelihood of error." In this case, there were no plant modifications required to support the proposed change to TS and the operating procedure was only changed slightly to add previously practiced actions. Therefore, the NRC staff concludes that no additional design verification or validation is necessary.

3.2.11 Human Performance Monitoring Strategy Because this is a "one-time only" change, long-term monitoring is not required. Based on the administrative protections afforded by the licensee's equipment guarding program, and by the administrative barriers of "flagging" and tagging of affected equipment, the NRC staff finds the intermediate (RFO only) monitoring strategy acceptable.

3.2.12 Human Performance Aspect Conclusion Based on the statements provided by the licensee (Le., that appropriate changes to procedures and training that are required to support the proposed LAR will be implemented prior to implementation of the proposed LAR), and; that appropriate administrative controls will be applied to procedures, training, and human interface design to prevent inadvertent changes and errors, the NRC staff concludes that the proposed LAR is acceptable from the human performance point of view.

4.0 COMMITMENT In its cover letter dated May 1, 2012, as restated in Attachment 5 of the same letter, the licensee provided the following commitment:

- 10 NextEra Energy Duane Arnold will guard both CS subsystems and will not perform any work or testing on either of the CS subsystems during RFO 23 when both CS subsystems are needed to be Operable to meet the requirements of LCO 3.5.2.

In the same letter (Attachment 1, Section 3.3 Technical Evaluation, page 8 of 12), the licensee stated the following:

Strict administrative and procedural controls, operator training, and use of human performance tools will be essential to preventing these types of consequential human errors. Furthermore, both CS subsystems will be guarded and no work or testing will be permitted on either of the CS subsystems during RFO 23 when both CS subsystems are needed to be Operable to meet the requirements of LCO 3.5.2.

The licensee provided further details to support the above statement on pages 7 and 8 of 8, in its July 26, 2012, letter.

The NRC staff reviewed the information and finds that the licensee's planned actions appear reasonable and adequate, and to be acceptable. The NRC staff arrived at this finding based on the supporting information and details provided in the licensee's application and supplements, rather than solely on the licensee's commitment statement.

5.0 STATE CONSULTATION

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

6.0 PUBLIC COMMENT 6.1 Background to the Public Comment On July 10, 2012, the NRC staff published a "Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing," in the Federal Register associated with the proposed amendment request (77 FR 40654). In accordance with the requirements in 10 CFR 50.91, the notice provided a 30-day period for public comment on the proposed no significant hazards consideration (NSHC) determination. On August 10, 2012, a public comment was received by email message. The comment did not specifically address the NSHC determination, but focused on the merits of the licensee's application.

The public comment is reproduced in its entirety below. The author of the comment structured his remarks into numbered items. Although not within scope of the NSHC determination, the NRC staff addressed the items in the comment as grouped by its author. The NRC response follows the corresponding quoted excerpt (that is, the three questions posed by the author of the public comment).

Bracketed information is authored by the NRC staff to facilitate reading of the comment and corresponding responses. The term "RAI" refers to a Request for Additional Information

- 11 message from the NRC staff requesting more details or clarification to facilitate its decisions.

The term "RAI response" refers to the licensee's corresponding response. The term "BWR/4" refers to the version of the Boiling Water Reactor constructed at the Duane Arnold Energy Center.

6.2 Public Comment in its entirety, including bracketed [ ] remarks by the NRC staff:

From: Dave Lochbaum [mailto: DLochbaum@ucsusa.org]

Sent Friday, August 10, 20129:11 AM To: Feintuch, Karl Cc: Leeds, Eric; Witte, Ulrich; Casto, Chuck

Subject:

Proposed refueling plan at Duane Arnold Hello Mr. Feintuch:

Ulrich Witte, a colleague and friend, brought to my attention the correspondence exchange between the Duane Arnold licensee and the NRC (e.g., ADAMS Accession Nos. ML121000327, ML12179A299, ML12209A420, ML12122A212, and ML12199A130).

[These document references, in the order in which they are listed, to an RAI response dated April 5,2012, for an unrelated amendment request to change the licensee's emergency action levels (EALs); two RAI responses for this requested licensing action dated June 27,2012, and July 26,2012; the application letter dated May 1, 2012; and the RAI dated July 16, 2012, to which the licensee responded on July 26, 2012].

The licensee wants to apply a protective coating to the inside surface of the torus at Duane Arnold during the upcoming refueling outage 23 (RFO 23). Doing so requires draining water from the torus and preventing water entering the torus during the work activities.

The licensee proposes to rely on one train of the core spray system to satisfy reactor core cooling regulatory requirements during this work. But core spray normally takes suction from the torus and its minimum 'flow recirculation line returns water to the torus. The licensee proposes to realign this core spray subsystem to take suction from the condensate storage tank and to disable the minimum flow recirculation function.

In reviewing the available materials regarding this proposal, I was unable to find answers to the following questions:

1) Why doesn't the licensee offload the entire reactor core to the spent fuel pool prior to and throughout the torus activities? Doing so would eliminate the need for the core spray subsystem. Has this licensee placed schedule (i.e., shorter refueling outages) ahead of safety?

2) Has the seismic risk from core spray using suction from the condensate storage. tank vice the torus been evaluated? I have never worked at Duane

- 12 Arnold, but I've worked at several boiling water reactors similar in design to it.

The condensate storage tank - the normal suction for the high pressure coolant injection and reactor core isolation cooling systems and alternate suction for the core spray and residual heat removal systems - is typically not credited in safety studies because of the challenge in seismically qualifying the condensate storage tank and its associated piping. Here, the licensee proposes to swap the torus - a more seismically robust source - to the condensate storage tank - a more seismically fragile source - without apparently justifying that margin reduction.

3) What about reactor vessel pressurization events during the torus work? The licensee provided incomplete justification for its proposal to eliminate the minimum flow recirculation function for the core spray pump during the proposed torus outage. The licensee's justification argued that reactor vessel pressure is always below the core spray injection valve opening pressure setpoint during

[during] Modes 4 and 5, thus providing complete assurance that this valve will open in time to prevent damage to the core spray pump from operation at no or low flow conditions. But that logic is simply not reflected in the BWR/4 Standard Technical Specifications (STS) and their bases issued by the NRC staff in April 2012 (ML12104A192 and ML12104A193). Minimum flow recirculation is required for the core spray pumps in Modes 4 and 5 in the STS. The flaw in the licensee's logic is that a loss of shutdown cooling could cause the reactor vessel pressure to increase above the opening pressure setpoint for the core spray injection valve. Last year's tragedy at Fukushima vividly illustrated the consequences from reactor vessel pressure being higher than the discharge pressure of makeup sources. The proposal by the licensee sets the stage for replicating elements of that disaster: (1) shutdown cooling is lost, (2) the ensuing heatup of the reactor vessel water increases the reactor vessel pressure above the injection valve's opening setpoint, and (3) the solitary core spray pump dead heads until it is irrevocably damaged.

UCS urges the NRC staff not to approve this amendment request until all applicable questions have been properly answered.

Thanks, David Lochbaum Director, Nuclear Safety Project Union of Concerned Scientists [UCS]

PO Box 15316 Chattanooga, TN 37415 (423) 468-9272 office (423) 488-8318 cell dlochbaum@ucsusa.org 6.3 NRC staff response to the Three Questions in the Public Comment The NRC staff acknowledges without comment that the author of the Public Comment paraphrased the licensee's proposed action in the preamble to his three questions.

- 13 NRC Staff Response to Question 1:

The only TS change the licensee has requested for RFO 23 is to add a footnote to Table 3.3.5.1-1 which allows them to remove the minimum flow logic that controls the CS minimum flow valve's function during Modes 4 and 5, and to operate the valve manually instead.

Consequently, the operators will operate this valve, as needed, during the period of torus re coating activity. This valve will remain available, although technically inoperable according to the current TS language. The proposed change did not request for waiver to any other TS requirements that apply to Modes 4 and 5 (note that the RHR system will remain operable for shutdown cooling mode in accordance with TS LCOs, 3.4.8, 3.9.7, and 3.9.8) nor did the licensee request any hardware changes during this period. As a result, the NRC staff realized that adequacy of human performance to operate the subject valve and related activities in timely manner, if needed, during this period will be the key factors affecting plant safety. Therefore, the NRC staff evaluated the role of human performance in order to determine whether human activities can be relied upon for accident mitigation during this limited period of time.

As discussed in the SE Section 3.1, the NRC staff believes that the licensee's enhanced plant procedures combined with strict administrative and procedural controls, operator training, and use of human performance tools during RFO 23 provides a reasonable assurance that specific and timely actions will be taken during this period to prevent and minimize potential human errors. Contrary to the suggestion made by Mr. Lochbaum that offloading the entire reactor core to the spent fuel pool during RFO 23 is a safer option, the NRC staff believes that the proposed actions are sufficiently safe, if not safer. Repetitious human actions involved in offloading an entire core have their own risk due to potential human errors; as well as, potential equipment failures, such as crane failures, during the process. Offloading the entire core also increases the analyses needed to safely re-arrange the spent fuel pool before and after the torus re-coating.

Therefore, the NRC staff concludes that the course of action proposed by DAEC during the torus re-coating project is a prudent and acceptable approach.

NRC Staff Response to Question 2:

In the unlikely event that the CST is not available when the Torus is drained, the CST is not the sole source of water providing makeup capability. There remains a large inventory of makeup water available for accident mitigation. DAEC is a "zero release" plant for liquid radioactive effluents, and none of the primary reactor system water inventory is discarded; particularly, the approximately 400,000 gallons of water removed from the Torus during the re-coat project. At various times during the RFP, a total of approximately 1.1 Million gallons of water is stored in a combination of locations such as the CSTs, the RPV cavity, the main condenser, the Torus, and the Radioactive Waste building. Therefore, DAEC is not totally reliant on just the fixed amount of inventory in the CSTs for makeup capability during the Torus activity.

NRC Staff Response to Question 3:

As stated earlier, the proposed LAR does not include any change in the operability requirements of the RHR System. The RHR system will remain Operable for Shutdown Cooling mode during the Torus activity in accordance with DAEC TS LCOs, 3.4.8, 3.9.7, and 3.9.8; therefore, reactor vessel pressurization is not expected during this period.

- 14 The NRC staff believes that the likelihood of "dead-heading" the CS pump is expected to be small because of the CS pump start sequence, and poses little risk to pump overheating. The licensee is implementing enhanced plant procedures to provide assurance that a CS pump flow path is established and flow is verified to be ~ 600 gpm immediately following CS initiation and thereafter maintained as such during pump operation. The manual startup/initiation procedure will be temporarily revised for RFO 23. Direct indication for CS pump flow is available and alternate indications for confirming CS pump flow ~ 600 gpm will be used, such as minimum flow valve position will be available as long as the associated instrumentation remains in service. If a flow path and required flow rate cannot be confirmed, the pump shall be promptly secured. Short duration operation of a CS pump at shutoff head (Le., dead-headed), sufficient time for the operators to diagnose the condition and secure the pump, is not expected to be detrimental to the pump. The NRC staff understands that extended operation beyond a few minutes can possibly damage the pump; therefore, to minimize potential human error during RFO 23, the licensee is implementing strict administrative and procedural controls, operator training, and use of human performance tools that are effective in preventing these types of consequential human errors, and thus provides reasonable assurance that pump "dead heading" is avoided.

7.0 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. 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 10, 2012 (77 FR 40654). 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.

A public comment was received on August 10, 2012, addressing the licensee's proposed action (not associated with the Commission's proposed no significant hazards consideration determination finding). The comment is reproduced in its entirety in Section 6.0 of this Safety Evaluation, in addition to the NRC staff's response to three questions included within the public comment.

8.0 CONCLUSION

S The proposed license amendment request was evaluated by the NRC staff. It was determined that the 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, concludes that this license amendment request is acceptable for RFO 23 only.

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

- 15 amendment will not be inimical to the common defense and security or to the health and safety of the public.

9.0 REFERENCES

1. Letter from P. Wells (DAEC) to USNRC, "Application for One-Time Technical Specification Change Regarding Core Spray Operability during Shutdown," dated May 1, 2012 (ADAMS Accession No. ML12122A212).

2. Letter from R.L Anderson (DAEC) to USNRC, "Response to Request for Additional Information: Application for One-Time Technical Specification Change Regarding Core Spray Operability during Shutdown," dated June 27,2012 (ADAMS Accession No. ML12179A299).

3. Letter from R. L Anderson (DAEC) to USNRC, "Response to Request for Additional Information: Application for One-Time Technical Specification Change Regarding Core Spray Operability during Shutdown," dated July 26, 2012 (ADAMS Accession No. ML12209A420).

4. EGM-11-003, "Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel," dated October 4,2011 (ADAMS Accession No. ML11251A230).

Principal Contributors: M. Razzaque G. Lapinsky K. Feintuch Date of issuance: September 27,2012

ML12249A317 ** SE transmitted by memo dated August 30,2012

.** w/comments OFFICE DORULPL3-1/PM DORULPL3-11 LA DE/SRXB/BC

• DRAlAHPB/BC **

NAME TBeitz BTuily SMiranda UShoop DATE 09110/12 09/10/12 07/10/12 08/30/12 OFFICE DSS/STSB/BC*** OGC NLO wIcomments DORULPL3-1/BC DORULPL3-1/PM NAME RElliott LSubin IFrankl KFeintuch DATE 09/11/12 09/20/12 09/27/12 09/27/12