L-MT-17-063, Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times

Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times

ML17263B194
Person / Time
Site:	Monticello
Issue date:	09/20/2017
From:	Church C Northern States Power Company, Minnesota, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9467, L-MT-17-063
Download: ML17263B194 (18)

Text

2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com September 20, 2017 L-MT-17-063 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times (CAC No. MF9467)

References:

1) Letter from NSPM to the NRC, Monticello Nuclear Generating Plant License Amendment Request: Revision to Emergency Plan Staff Augmentation Response Times, dated March 24, 2017. (ADAMS Accession Number ML17083A083)

2) Email from the NRC to NSPM, Request for Additional Information for Monticello Nuclear Generating Plant License Amendment Request dated March 24, 2017 RE: Emergency Response Organization, dated August 7, 2017. (ADAMS Accession Number ML17219A075)

Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License. The proposed change would revise the MNGP Emergency Plan (E-Plan) to extend staff augmentation times for Emergency Response Organization (ERO) response functions. In Reference 2, the NRC provided NSPM with Requests for Additional Information (RAIs) pertaining to information needed to support the staffs review of NSPMs application in Reference 1. The enclosure to this letter provides NSPMs response to the NRC RAIs.

The supplemental information provided herein does not change the conclusions of the No Significant Hazards Consideration and the Environmental Consideration evaluations provided in Reference 1.

If there are any questions or if additional information is needed, please contact Mr. Peter Gohdes at (612) 330-6503 or Peter.Gohdes@xenuclear.com.

Document Control Desk Page 2 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury, that the foregoing is true and correct.

Executed on September 20, 2017.

~J{~

Christopher R. Church Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC State of Minnesota

L-MT-17-063 NSPM Enclosure Response to Request for Additional Information:

License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times On March 24, 2017, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), submitted an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (Reference 1). The proposed change would revise the MNGP Emergency Plan (E-Plan) to extend staff augmentation times for Emergency Response Organization (ERO) response functions. By email dated August 7, 2017, the NRC requested the following additional information (Reference 2). The responses to this request for additional information (RAI) are provided below.

MNGP RAI-1 LAR Enclosure 1, Section 3.2.5, Plant System Engineering, Repair and Corrective Actions Function,"Technical Support Major Task c, [sic] states, in part (on page 26 of 42):

A review of procedural actions for this position demonstrated that failed fuel determinations as well as establishing recovery/reentry priorities would not be required during the first 60 minutes of the event. Initial reactor core stabilization activities are performed by the Operations crew under the direction of an SRO

[Senior Reactor Operator].

Provide further justification for the extension in augmentation timing for the Core Thermal Engineer position. Specifically, clarify whether the SRO has the necessary expertise with core/thermal hydraulics, and possesses the ability to offer adequate functional oversight to assess core conditions. This justification should include a description of any procedure and information technology advances since the implementation of NUREG-0654, such as improvements that allow using a symptom-based emergency operating procedure network, and computerized or automated systems for the acquisition and display of parameters used to evaluate core conditions.

NSPM Response Section 5.3.1.5 of the E-Plan states that technical support for the shift Operations staff is initially provided by the duty Shift Manager [SRO] or Shift Technical Advisor [STA] (when staffed separately on-shift). Enhancements in the STAs ability to perform required actions on-shift have resulted from improvements in training, adoption of symptom-based Emergency Operating Procedures (EOPs), and significant improvements in the ability to use computer parameters to efficiently monitor core conditions.

STA Training The required educational qualifications for individuals filling this function include one of the following:

Page 1 of 13

L-MT-17-063 NSPM Enclosure Bachelor degree in Engineering from an accredited institution.

Bachelor degree in Engineering Technology from an accredited institution, including course work in the physical, mathematical, or engineering sciences.

Bachelor degree in Physical Science from an accredited institution, including course work in the physical, mathematical, or engineering sciences.

Professional Engineer License by completion of the PE examination.

An STA candidate must meet one of the following entry-level training requirements:

Successfully completed the MNGP Initial License Training SRO Program.

Hold a current MNGP NRC SRO License and be current in the Operations Training requalification program.

Additionally, the STA candidate must have at least one year of experience in the operation or engineering support of a nuclear power plant with at least six months of experience at MNGP.

STA candidates receive detailed initial training in the following areas:

Introduction to Critical Safety Functions Reactivity Control Radioactivity Containment Core Heat Removal Ultimate Heat Sink Critical Safety Function Case Studies Tracking and Trending of Critical Plant Parameters Subsequently qualified STA personnel receive continuing training including:

Reactor Physics Core Characteristics over Core Life Reactivity Control during Normal and Emergency Operating Conditions STA Functions Personnel performing the STA function on-shift are responsible for maintaining a broad perspective of operating activities and operating conditions during accident conditions and for providing assessment and advice to the Shift Manager responsible for plant operation.

Procedure and Technical Improvements Since the approval of Revision 2 of the E-Plan on January 20, 1983 (Reference 3), the MNGP EOPs have been improved based upon internal operating experience and industry initiatives.

EOPs now use a symptom-based approach that demands less assessment and interpretation of plant conditions by the operating crews. The EOPs interface well with new technology, such Page 2 of 13

L-MT-17-063 NSPM Enclosure as the plant process computer system (PPCS) and the Safety Parameter Display System (SPDS). In addition, since Revision 2 of the E-Plan, the EOPs have been put into flowcharts, revised for human factoring, and modified for improved layout resulting in more consistent implementation.

Upgrades in plant computer capability have also contributed to the reduction in burden for the performance of STA related activities. Power Range Monitor (PRM) equipment has been replaced by a digital Power Range Neutron Monitor (PRNM) System. PRNM provides a functional replacement of the Local Power Range Monitor (LPRM), Average Range Monitor (APRM), and Rod Block Monitor (RBM) functions and upgrades the interfaces to the Rod Worth Minimizer (RWM) and Automated Traversing In-Core Probe (ATIP). This new capability is available via displays in the Control Room located above the control panels, at the EOP/Control Room Supervisor (CRS) desk, the Reactor Operator (RO) desk and at the Shift Manager desk. The SPDS consists of three primary displays that are designed to support the information needs of the Emergency Procedure Guidelines (EPGs). These displays, RPV Control Display, Containment Control Display, and Critical Plant Variables Display, are elaborated in special function displays. The special function displays provide: 1) two-dimensional plots of the limiting conditions defined in the Emergency Operating Procedures (EOPs), e.g., Drywell Design Pressure Curve, 2) trend plots of all control parameters, showing data from the most recent 30 minutes, 3) the validation status of SPDS input data, and 4) radiation monitoring displays. These display capabilities which were not available at the time of the approval of Revision 2 of the E-Plan enhance the STAs ability to monitor core thermal conditions and provide advice to the operating crew.

Summary Plant data accessibility, combined with additional trending and core monitoring capability, and the use of symptom-based procedures and flow charts have significantly reduced on-shift burden for performance of STA functions. This allows the individual performing these functions to maintain this responsibility for the first 60 minutes without adversely impacting response to an event.

MNGP RAI-2 LAR Enclosure 1, Section 3.2.5, Repair and Corrective Actions Major Task c, states, in part (on page 26 of 42):

Historically, the repair functions associated with an event have been completed by Plant Equipment Operation (PEO) personnel on-shift who are qualified to respond to plant events and perform actions to stabilize the plant.

Describe the training and qualifications provided to the PEOs that would allow them to perform repair and corrective actions to justify the extension in augmentation times for the maintenance technicians. This justification should include a review that there will not be any conflict between the added collateral duties and other assigned emergency response functions.

Page 3 of 13

L-MT-17-063 NSPM Enclosure NSPM Response NSPM proposes to revise Table 1 (Attachment 1 of this Enclosure) of the MNGP E-Plan to remove the reference to performance of Mechanical and Electrical Maintenance expertise as a duty performed by shift personnel assigned other functions. The basis for the proposed change is the robust nature of the Emergency Core Cooling Systems (ECCS), which precludes the need for maintenance activities during the initial stages of the event, and the training provided to Licensed and Non-Licensed Operations personnel to respond to plant conditions through the use of EOPs and Abnormal Operating Procedures (AOPs).

MNGP ECCS In accordance with the MNGP Updated Safety Analysis Report (USAR), MNGP was designed and licensed in conformance with the proposed Atomic Energy Commission 70 General Design Criteria for Nuclear Power Plant Construction, Group VII Engineered Safety Features.

Group VII features that apply to the MNGP ECCS include:

Criterion 41, Engineered Safety Features Components Capability (Category A), states:

Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.

Criterion 44, Emergency Core Cooling Systems Capability (Category A), states:

At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features or components unless it can be demonstrated that (a) the capability of the shared feature or component to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a loss-of-coolant accident and is not lost during the entire period this function is required following an accident.

These design criteria are met at MNGP through provision of a redundant and diverse ECCS which assures the required emergency core cooling is provided even if there should be a Page 4 of 13

L-MT-17-063 NSPM Enclosure single failure of any component in the system. Section 6.2 of the USAR provides a description of the ECCS which consists of the following:

High Pressure Coolant Injection System (HPCI)

Automatic Depressurization System (ADS)

Core Spray System (CSS)

Residual Heat Removal System (RHR)

HPCI The HPCI System is designed to pump water into the reactor vessel under loss-of-coolant conditions which do not result in rapid depressurization of the pressure vessel. The loss-of-coolant might be due to loss of reactor feedwater or to small line breaks which do not cause depressurization of the reactor vessel. The HPCI System provides adequate reactor core cooling for small breaks below the capability of the unassisted Core Spray or Low Pressure Coolant Inject (LPCI) and to depressurize the reactor primary system to aid the LPCI and Core Spray. Operation of the HPCI turbine continues as long as reactor pressure is above 150 psig at which point Core cooling is accomplished by either the Core Spray or LPCI Systems. The HPCI turbine is driven with steam from the reactor vessel and only requires power from the plant safety related battery system for operation.

ADS The ADS is designed to depressurize the reactor to permit either LPCI or Core Spray to cool the reactor core during a small break Loss-Of-Coolant Accident (LOCA). The ADS accomplishes reactor vessel depressurization by blowdown through automatic opening safety/relief valves which vent steam to the suppression pool. The ADS consists of three (3) main steam safety relief valves (SRVs) with an automatic circuitry. To assure that the ADS valves are able to operate, provisions are made to power each valve, independently, from alternate power sources. The electrical circuitry is arranged such that single component failures cannot negate the manual operation of these valves.

CSS Two independent core spray loops are provided for use under loss-of-coolant conditions associated with large pipe breaks and reactor vessel depressurization. Each of the two core spray cooling loops consists of a pump, valves, and associated piping. Physical separation of the pumps is achieved by locating pumps in different corners of the reactor building. The power source for each CSS is located on separate emergency buses. Power for these emergency buses can be supplied from the diesel generators. The control system is arranged to provide two independent and separately isolated control and power circuits for operation of the two independent core spray subsystems.

Page 5 of 13

L-MT-17-063 NSPM Enclosure RHR/LPCI The RHR system consists of two loops, each comprised of a heat exchanger, two RHR pumps in parallel and associated piping. These two loops are physically separated in different corners of the Reactor Building to minimize the possibility of a single physical event causing the loss of the entire system. Power to the RHR pumps is supplied from safety related essential buses.

These buses are normally powered from adjacent non-safety buses fed from the auxiliary off-site source. However, if this non-safety bus source becomes unavailable, power is available from an emergency off-site source or the dedicated standby Emergency Diesel Generator (EDG) associated with that essential bus. LPCI is an RHR subsystem which is designed to re-flood the reactor vessel to at least two-thirds core height and to maintain this level. After the core has been flooded to this height, the capacity of one RHR pump is more than sufficient to maintain the level.

The diverse and redundant design of the MNGP ECCS ensures adequate core cooling is provided even in the event of a LOCA with a single failure of an ECCS component or support system coincident with the unavailability of offsite power. The ECCS has been analyzed in accordance with the NRC requirements and the acceptance criteria of 10 CFR 50.46. The LOCA analysis provides assurance that the most limiting break size, break location, and single failure combination have been considered for the plant. Details of the analysis are provided in Section 14.7.2 of the MNGP USAR.

Licensed and Non-Licensed Operator Event Response Actions Licensed and Non-Licensed Operators perform actions in response to plant events as directed by AOPs and EOPs. These actions may be used to restore system capability as initial corrective actions typically attributed to the Technical Support and Repair/Corrective Actions Major Task.

Licensed Operator event response actions directed by AOPs and EOPs include but are not limited to:

Response to Reactor SCRAM Response to an ATWS Response to Plant Fires Response to Security Events Response to LOCA Response to Steam Leaks/Breaks Response to a Loss of Off-Site Power Response to a Station Blackout Response to Loss of Vital DC Power Control of RPV Level and Pressure Control of Primary and Secondary Containment Shutdown of the Plant from Outside the Control Room Page 6 of 13

L-MT-17-063 NSPM Enclosure Non-Licensed Operators (NLOs) event response actions directed by AOPs and EOPs include, but are not limited to:

Starting/Stopping/Operating Pumps Manipulating System Valves Manipulating Electrical Breakers Starting/Stopping/Operating EDGs Licensed Operator Training Initial License Training (ILT) is designed to train license candidates to meet the requirements of 10 CFR 55 and NUREG-1021 for NRC license examinations and to perform the duties of a reactor operator or senior reactor operator. This includes training on AOP and EOP response activities. The knowledge portion of AOP/EOP training for ILT candidates is presented in the classroom or simulator, as appropriate, as well as a discussion of related bases documents.

The candidates knowledge and understanding of the material is evaluated using written exams, scripted questions at end of job performance measures (JPMs), plant or simulator walkthroughs with fundamental questions incorporated, simulator training with fundamental questions incorporated, observation results, etc.

Licensed Operator Requalification Training (LORT) is based upon a systems approach to training and conducted in accordance with 10 CFR 55.59 and National Academy for Nuclear Training document, ACAD-07-001, Guidelines for Continuing Training of Licensed Personnel, which includes AOP and EOP response activities. A Biennial Training Plan (BTP) is developed every two years to support the subsequent biennial LORT cycle. The BTP provides a schedule to ensure regulatory and ACAD training commitments are covered at the prescribed frequency as well as additional material identified by incumbents and various training and/or curriculum oversight committees. Insights from probabilistic risk assessment data is also considered when assigning the relative importance training needs for LORT. Biennial and annual examinations are conducted using written exams, dynamic simulator exams and JPMs in accordance with 10 CFR 55.59 and ACAD-07-001.

Non-Licensed Operator Training Initial Non-Licensed Operator Training (NLOT) is designed to provide the necessary knowledge and skill for candidates to perform the assigned tasks of the NLO. The initial NLOT includes training in general engineering concepts, fundamental power plant operations, specific plant systems, normal, abnormal and emergency plant operations and administrative procedures. The training is broken into three training periods for Fundamentals, Turbine Side and Reactor Side. Periodic written exams are conducted during training periods with a comprehensive exam at the end of each training period. Upon completion of the Turbine Side and Reactor Side training periods, the NLOT candidates complete a period of on the job training (OJT) to include completion of task performance evaluations.

The goal of Non-Licensed Operator Requalification Training (NLORT) is to maintain and enhance the required knowledge, skills, and abilities of personnel in order to perform assigned Page 7 of 13

L-MT-17-063 NSPM Enclosure roles and responsibilities. NLORT is conducted as a continuous two-year training program divided into seven week training cycles. Every two years a two-year training plan will be developed to support the next two years of NLO training. The two-year training plan will be specifically constructed to provide a schedule for ensuring that required tasks and topics selected for continuing training will be accomplished. At a minimum, there will be an evaluation every other training cycle. Evaluations will alternate between a JPMs Walkthrough Exam and a written exam.

Summary Crediting the robust ECCS capability through redundant and diverse system design and protection against loss of ECCS capability due to a single component failure provides the basis for the position that no ECCS repair and corrective actions are necessary for on shift personnel prior to augmentation of maintenance personnel. However, in the unlikely event of an ECCS failure, on-shift licensed and non-licensed personnel will respond to restore ECCS functions as directed by AOPs and EOPs. As discussed above, licensed and non-licensed personnel are trained to perform these actions as part of their assigned duties as such they would not be considered a collateral duty. The robust ECCS capability also provides the basis for the proposed changes to the augmentation times for maintenance personnel as ECCS repair and corrective actions performed by augmented personnel would address restoration of redundant equipment as opposed to restoration of minimum required equipment to provide core cooling capabilities. Therefore, the proposed change will not result in a reduction in response capability for performance of technical support, repair activities or corrective actions.

MNGP RAI-3 LAR Enclosure 1, Section 3.2.6, Protective Actions (In-Plant) Function, Access Control/Dosimetry, states, in part (on page 27 of 42):

Radiation work permits (RWPs) establish the necessary preset warnings/alarms associated with use of the ED [electronic dosimeter]. Specific emergency RWPs have been developed for use during a declared emergency, which automatically provide the ED with emergency dose and dose rate alarms.

Provide further description about the tools and processes used for the task of access control, including a description of portal/contamination monitors, self-alarming dosimeters, and automated access control system for the radiologically controlled area (RCA) that maintain active radiation work permits, which are readily available if an emergency is declared (e.g., the system verifies qualifications, dose margins, and access requirements).

NSPM Response Capabilities implemented since Revision 2 of the E-Plan have resulted in improved efficiencies with respect to access control and dosimetry. Enhancements have been made to access control equipment and processes. These improvements have streamlined access control by automating the issuance of dosimetry and by providing Emergency RWPs that program the Page 8 of 13

L-MT-17-063 NSPM Enclosure dosimeters with dose and dose rate thresholds appropriate to post-accident conditions. Finally, turnstiles are provided that verify proper dosimetry prior to allowing RCA access further reducing burden on the Radiation Protection (RP) Technician.

The site computer system provides process and plant radiological data needed to determine protective actions needed for in-plant work.1 There are two different RWP processes available that reduce the burden on RP Technicians during the initial phase of an accident by bypassing or significantly reducing RP planning needed by the normal process. In addition to the RWP processes, the site also maintains a stand-alone set of emergency Self Reading Dosimeters (SRDs) with higher limits bypassing the RWP process and allowing immediate entry if the Shift Manager determines the need.

Once the Operations Support Center (OSC) is activated, the RP Technician in the OSC accesses remote radiological monitoring and can determine RWP and briefing needs. The OSC technician can rapidly determine if normal RWPs or the Emergency RWPs would be used. For special circumstances, the site provides procedural guidance for expedited development of a task specific RWPs. Equipment descriptions supporting the process are described below.

Dosimetry and RWP Equipment Description Electronic SRDs are used at MNGP that track dose and dose rate data and activates audible and visual alarms at independently programmed thresholds. Different audible signals designate dose and dose rate alarms and other faults such as defective or low battery. Alarm messages are displayed either by symbols flashing at the same time as the displayed measurement or by alphanumeric messages alternating with the measurement.

The SRDs are used in conjunction with the access control software to record and track personnel radiation exposure. The dosimeter communicates with the access control software via a reader that downloads RWP thresholds and passes dose and dose rate information to the dosimetry database. Once the SRD has been activated by the access control software, the dosimeter works as a key to unlock turnstiles to gain access to the RCA.

Two RWP processes available to personnel during the initial phase of an accident are as follows:

Urgent Work RWP Emergency RWP Both of these processes bypass the normal RP planning and work request processes in favor of rapid response to emergency needs. Logging into Emergency RWPs uses the same process as routine RWPs. Emergency RWPs require telemetry and set dose and dose rate alarm threshold limits into the SRDs commensurate with accident conditions with elevated 1

MNGP RAI-4 response provides a detailed description of the plant automated radiological monitoring capability.

Page 9 of 13

L-MT-17-063 NSPM Enclosure dose rates in the plant. Emergency RWPs permit access to High Contamination Areas.

Multiple RWP tasks are used at MNGP to allow access to High Radiation Areas and Locked High Radiation Areas.

Automated Access Control System An automated access control system is provided near the main access point to the RCA. The automated access control software verifies Radiation Worker and Plant Access Training are current, determines whether an HRA briefing is required and displays annual dose and dose margin to the administrative limit of 2000 mrem. The system also displays the dose and dose rate alarm thresholds that will be set in the SRD.

Remote Monitoring or Worker Dosimetry The Emergency RWP requires entrants to use telemetry on their SRDs. Transmitters relay dose and dose rate information to receivers connected to the plant local area computer network (LAN). PCs with the software installed are used to monitor, track and trend SRD telemetry data. Remote monitoring software, accessible from designated computers allows real-time monitoring of workers wearing an SRD equipped with a transmitter. PCs with this software are located throughout the RP office, main access control, technicians workstations, and the OSC. Workers wearing a transmitting SRD appear to the remote monitoring software user after they log into the access control software and are removed after log out.

Administrative set points which are lower than RWP alarm set points on the electronic dosimeter may be set by RP personnel. The administrative alarms can assist RP job coverage by providing early notice of accumulating dose and dose rates. Use of the remote monitoring software allows the RP Technician to monitor multiple workers simultaneously reducing the time and effort required for job coverage while maintaining adequate protection for rad workers.

Contamination Monitors Portal style contamination monitors are located at the exit points of the MNGP RCA. The monitors are equipped with gas flow detectors for alpha and beta detection and large plastic scintillators for gamma detection. Indicator lights at the entry show when the monitor is ready to use. The monitor counts the front, back, both feet and both arms/hands in a complete cycle.

Visible and audible alarms are given if contamination is detected. A CONTAMINATED result is shown on a large color LCD display with voice reinforcement and an LED lights up beside each contaminated detector. The display shows the type (alpha, beta or gamma), the quantity and the location of the contamination based on which detector(s) is alarming. The system records data and date / time stamped logs showing the number of times the unit was used, parameters used, calibration settings and fault messages.

An additional display screen that duplicates the screen messages given to the occupant is provided for each of the monitors at access control such that the access control technician can Page 10 of 13

L-MT-17-063 NSPM Enclosure see contamination levels, types and locations when an occupant alarms one of the monitors.

This feature serves to further reduce burden on the access control technician.

Summary The tools and processes described above significantly reduce the burden on duty RP personnel in the performance of access control and dosimetry activities. They provide RP personnel expanded capability to expedite required in-plant response supporting the proposal to delay augmentation of the response of RP Technicians.

MNGP RAI-4 LAR Enclosure 1, Section 3.2.6, Protective Actions (In-Plant) Function Summary, states, in part (on page 28 of 42):

The proposed changes maintain the existing on-shift HP [health physics]

technicians for the HP Coverage task. The proposed changes extend the 30-minute and 60-minute response times for the personnel to 60 minutes and 90 minutes, respectively. NSPM has implemented improvements in technology in the areas of dosimetry and access control at the MNGP which reduced the need for HP Technician actions in each of these areas during the early stages of event response.

Provide further justification for the extensions in time for the HP coverage task, including a description of the availability of installed area, process, airborne and effluent radiation monitors, automated systems and information technology solutions, and enhanced work processes that would be available under accident conditions.

NSPM Response Enhancements have been made in the ability to remotely monitor overall radiological plant conditions supporting in plant emergency response. The remainder of this response will focus on the improvements in radiation monitoring systems limiting the dependency on direct technician monitoring in the plants.2 The radiation monitoring system consists of a wide variety of monitoring covering key areas of the plant. These monitors are integrated into the SPDS and readings are available in the SPDS displays available in the Work Execution Center Office, Duty Shift Manager Office, Technical Support Center (TSC), Emergency Operations Facility (EOF), and throughout the Control Room. The system provides the Emergency Director, RP and chemistry personnel with the ability to rapidly assess overall plant conditions throughout the site. Access to plant radiological data greatly enhances the ability of a single RP Technician to brief multiple response teams, expediting entry into the plant while preserving the ability of the technician to perform other tasks concurrently.

2 See the response to MNGP RAI-3 for details on improvements in overall radiation work processes and monitoring.

Page 11 of 13

L-MT-17-063 NSPM Enclosure Monitoring Equipment Description The Process Radiation Monitoring System is designed to provide continuous monitoring and readout of the radioactivity of all process lines and vents that can release radioactivity directly to the environment. In addition, this system also continuously measures, indicates and records the radioactivity concentration levels of in-plant process streams and vents.

The Process Radiation Monitoring System consists of several subsystems which provide continuous monitoring of the radiation levels of liquid and gaseous processes throughout the plant which can release activity directly to the environment. These subsystems assist in controlling the release of radioactive by-products within the legally prescribed limits as set forth in the Technical Specifications and Offsite Dose Calculation Manual. They also help provide for personnel safety by warning of abnormal radiation release levels and, in some cases, automatically terminating these releases.

The subsystems of the Process Radiation Monitoring System provide plant personnel with visual real time indication of process radiation levels, as well as a permanent record by means of recorders.

Area Radiation Monitors The Area Radiation Monitoring System provides the Control Room operating personnel with a continuous indication of gamma radiation levels at various selected locations throughout the plant buildings. The system contributes plant dose rate information so that correct decisions can be made with respect to deployment of personnel in the event of a radiation incident as well as providing local alarms at key points where substantial radiation level changes may occur.

There are Area Radiation Monitors located throughout the plant, Recombiner Building, and Off-Gas Storage Building. Each consists of a detector coupled to an indicator located in the Control Room. Indications are available for:

Containment High Range Monitors Post-Accident Sample System (PASS)

Off-Gas Storage Building Drywell Monitors The area monitors are X-ray and gamma sensing devices. The Area Radiation Monitors connect to 12 Control Room annunciators. In addition to alarms in the Control Room, when the trip set points are reached, 16 of the area monitors have local alarms provided to warn people of high radiation levels in those areas where there is a potential for substantial changes.

Page 12 of 13

L-MT-17-063 NSPM Enclosure Summary Integration of the radiation monitoring systems with the plant computer system provides expanded availability of normal and accident condition radiological information to critical locations staffed by on-shift minimum staff and Emergency Response Facilities. The improved availability has reduced the need for direct in plant RP Technician support.

References

1. Letter from NSPM to the NRC, Monticello Nuclear Generating Plant License Amendment Request: Revision to Emergency Plan Staff Augmentation Response Times, dated March 24, 2017. (ADAMS Accession Number ML17083A083)

2. Email from the NRC to NSPM, Request for Additional Information for Monticello Nuclear Generating Plant License Amendment Request dated March 24, 2017 RE: Emergency Response Organization, dated August 7, 2017. (ADAMS Accession Number ML17219A075)

3. Letter from the NRC to NSPM, Emergency Preparedness Safety Evaluation Report Related to the Operation of Monticello Nuclear Generating Plant, dated January 20, 1983.

(Legacy ADAMS Accession Number 8302040207)

Page 13 of 13

ENCLOSURE, ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Regarding License Amendment Request to Revise Emergency Plan Staff Augmentation Response Times MARKED-UP COPY PAGES OF EMERGENCY PLAN (2 Pages Follow)

Monticello E-PLAN EMERGENCY PLAN 33 of 124 Table 1 MINIMUM SHIFT STAFFING AND CAPABILITY FOR ADDITIONS FOR NUCLEAR POWER PLANT EMERGENCIES Capability for Additions Position Title or On 360 690 Major Functional Area Major Tasks Expertise Shift min min Plant Operations and Shift Manager 1 --- ---

Assessment of Control Room Supv 1 --- ---

Operational Aspects SRO 1 --- ---

Nuclear Lead 1 --- ---

PE&RO (RO)

Nuclear PE&RO 2 --- ---

Nuclear Asst. PEO 3 --- ---

Emergency Direction Emergency Director 1* --- ---

and Control (Shift Manager until relieved)

Notification/ Notify licensee, Shift Emergency 1 --- ---

Communication Local, State, and Communicator Federal personnel &

agencies Maintain Emergency --- 12 2 Communications Communicators Radiological Accident Emergency Emergency --- --- 1 Assessment and Operations Facility & Manager (EOF)

Support of Operational TSC Leads ED (TSC)OF Coord --- 1 ---

Accident Assessment Coordinate EOF prior to arrival of Emergency Manager Off-Site Dose Radiological --- 1 ---

Assessment Emergency Coord RPSS --- --- 1 Off-Site Surveys Radiation Protection --- 1 1

/Support --- 1 1 On-Site (out-of-plant)/ 2 ---1 1 In-Plant Surveys Chemistry/Radio- Chemistry 1 1 ---1 Chemistry

Monticello E-PLAN EMERGENCY PLAN 34 of 124 Table 1 MINIMUM SHIFT STAFFING AND CAPABILITY FOR ADDITIONS FOR NUCLEAR POWER PLANT EMERGENCIES (CONTD)

Capability for Additions Position Title or On 360 690 Major Functional Area Major Tasks Expertise Shift min min Plant System Technical Support Technical Advisors 1* --- ---

Engineering, Repair and Core/Thermal Hyd. --- 1 ---

Corrective actions Electrical --- 1--- ---1 Mechanical --- 1--- ---1 Repair & Corrective Mech Maint ---1* ---1 ---1 Actions Elec Maint ---1* 1 ---1 I&C --- ---1 ---1 Protective Actions Radiation Protection Radiation Protection 2* 1 1 (In-Plant) a. Access Control

b. HP Coverage for response actions

c. Personnel monitoring

d. Dosimetry Fire Fighting Fire Brigade per Local 4 AWI-08.01.01 Support Rescue Operations and 2* Local First Aid Support Site Access Control and Security, Security Force All per Personnel Accountability Communications, Security Plan Personnel Accountability Total 13 914 129

• May be provided by shift personnel assigned other functions.