ML18109A396

Enclosure 2; Reactor Oversight Process Task Force FAQ Log, April 05, 2018

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Enclosure 2 Reactor Oversight Process Task Force FAQ Log April 5, 2018 Document Date: 19 April 2018

FAQ Log April 2018 FAQ No. PI Topic Status Plant/Co. Point of Contact 18-01 IE Definition of Introduced January 31 Generic Ken Heffner Initial (Certrec)

Transient Discussed March 1 Alex Garmoe Tentatively Approved April 5 (NRC) 18-02 IE Watts Bar 2 Introduced March 1 Watts Bar Kim Hulvey/Beth IE01 and Unit 2 Wetzel (TVA)

IE03 Proposed Response Effectiveness Discussed April 5 Alex Garmoe Date (NRC) 18-03 IE Plant-specific Introduced March 1 Columbia Desirée exemption Wolfgramm from Discussed April 5 (Energy guidance Nothwest)

Alex Garmoe (NRC)

For more information, contact: James Slider, (202) 739-8015, jes@nei.org

FAQ 18-01 Definition of Initial Transient - Tentatively Approved NOTE This FAQ would implement a whitepaper that proposed clarifications of the definition of Initial Transient. The whitepaper was discussed with the NRC staff in public ROP meetings in 2013-2014. The final discussion of the whitepaper occurred at a May 14, 2014 public meeting. The NRC staff member who had the lead on performance indicators at the time was Andrew Waugh, who is listed below as the NRC Contact. The concluding discussion is documented in an NRC meeting summary available in ADAMS at accession number ML14149A293. The proposed text changes presented below reflect NRC comments and suggested edits for agency approval presented in a mark-up of the whitepaper attached to the aforementioned meeting summary. The marked-up whitepaper is available under ADAMS accession number ML14149A278.

Plant: Generic Date of Event: September 11, 2014 Submittal Date: September 11, 2014 Licensee

Contact:

Lenny Sueper Tel/email: 612-330-6917 / Leonard.Sueper@xenuclear.com NRC

Contact:

Andrew Waugh Tel/email: (301) 415-5601 / andrew.waugh@nrc.gov Performance Indicator: IE04 - Unplanned Scrams with Complications Site-Specific FAQ (see Appendix D)? No - this is generic FAQ to become effective: When approved Question Section NEI 99-02 Guidance needing interpretation (include page and line citation):

Page 23 Line 20:

20 Was pressure control unable to be established following the initial transient?

Page 24 Lines 39 - 40:

39 Following initial transient, did stabilization of reactor pressure/level and drywell pressure 40 meet the entry conditions for EOPs?

Event or circumstances requiring guidance interpretation:

Two of the questions in NEI 99-02 used to determine if a BWR reactor trip was an Unplanned Scram with Complications include the undefined term initial transient; Was pressure control unable to be established following initial transient? and Following initial transient did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? The failure to define the term has resulted in confusion, with some licensees interpreting initial transient to be equivalent to scram response.

If licensee and NRC resident/region do not agree on the facts and circumstances, explain:

N/A Potentially relevant FAQs: None Response Section Proposed Resolution of FAQ:

The following is proposed to be added in the Definition of Terms section of this indicator:

Initial Transient is intended to envelope the immediate and expected changes to BWR parameters as a result of a scram (e.g., pressure, level, etc.) because of the collapsing of voids in the core and the routine response of the main feedwater and turbine control systems. For example, at some BWRs the reflected pressure wave resulting from the rapid closure of turbine valves during a turbine trip may Page 1 of 2 20180207

FAQ 18-01 Definition of Initial Transient - Tentatively Approved result in a pressure spike in the reactor vessel that causes one or more safety-relief valves (SRVs) to briefly lift. The intent is to allow a licensee to exclude the momentary operation of SRVs when answering Was pressure control unable to be established? The sustained or repeated operation of SRVs in response to turbine control bypass valve failures or Main Steam Isolation Valve (Group I) isolations are not a part of routine BWR scram responses and are therefore not considered to occur within the initial transient. Similarly, a reactor level decrease to Level 3 following a reactor trip due to the expected collapsing of voids in the core can be excluded when answering the question Following initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? as long as the feedwater control system and at least one feedwater pump were operating as designed. Initial transient is different from scram response. The initial transient is a subset of the overall scram response time.

If appropriate, provide proposed rewording of guidance for inclusion in next revision:

See above.

PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No Proposed NRC Response NRC agrees with the proposed language for initial transient with the addition of the following words:

Initial Transient is intended to envelope the immediate and expected changes to BWR parameters as a result of a scram (e.g., pressure, level, etc.) because of the collapsing of voids in the core and the routine response of the main feedwater and turbine control systems. For example, at some BWRs the reflected pressure wave resulting from the rapid closure of turbine valves during a turbine trip may result in a pressure spike in the reactor vessel that causes one or more safety-relief valves (SRVs) to briefly lift. The intent is to allow a licensee to exclude the momentary operation of SRVs when answering Was pressure control unable to be established? The sustained or repeated operation of SRVs in response to turbine control bypass valve failures or Main Steam Isolation Valve (Group I) isolations are not a part of routine BWR scram responses and are therefore not considered to occur within the initial transient. Similarly, an initial reactor level decrease to Level 3 immediately following a reactor trip due to the expected collapsing of voids in the core can be excluded when answering the question Following initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? as long as the feedwater control system and at least one feedwater pump were operating as designed. Initial transient is different from scram response. The initial transient is a subset of the overall scram response time.

Page 2 of 2 20180207

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response Plant: Watts Bar Nuclear Plant, Unit 2 (WBN 2)

Date of Event: 12/31/2017 Submittal Date: 2/21/2018 Engineer/Licensee

Contact:

Kim Hulvey/Beth Wetzel Tel/email: (423) 365-7720/(423)751-2403 NRC Contact Jared Nadel Watts Bar Tel/email: (423) 365-1776 Performance Indicators:

IE01 WBNU2 Unplanned Scrams per 7000 Critical Hours (automatic and manual scrams during the previous four quarters)

IE03 WBNU2 Unplanned Power Changes per 7000 Critical Hours (over previous four quarters)

Site-Specific FAQ (Appendix D)? - Yes FAQ to become effective when approved.

Question Section:

TVA requests the effective date of Watts Bar Unit 2 Unplanned Scrams per 7000 Critical Hours (IE01) and (IE03) Unplanned Power Changes per 7000 Critical Hours be extended until 3Q18 (through Jun 30, 2018) to allow sufficient data for an accurate assessment value. This request is based upon a October 22, 2015 NRC letter to TVA stating If, as the licensee approaches four quarters after either the IE or MS cornerstones become monitored, new information shows that a PI may still not provide accurate assessment value, the Frequently Asked Questions process will be utilized in accordance with NEI 99-02 to reach a conclusion on how to proceed.

NEI 99-02 Guidance needing interpretation:

NRC Letters to TVA dated November 21, 2016 (ML16326A210) and October 22, 2015 (ML15295A253).

NEI 99-02 Page 10 line 25 The number of unplanned scrams during the previous four quarters, both manual and automatic, while critical per 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

NEI 99-02 Page 14 line 9 The number of unplanned changes in reactor power of greater than 20% of full-power, per 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of critical operation excluding manual and automatic scrams.

NEI 99-02 Page E-1 line 12 There are several reasons for submitting an FAQ:

NEI 99-02 Page E-1 line18

3. To request an exemption from the guidance for plant-specific circumstances, such as design features, procedures, or unique conditions.

Event or circumstances requiring guidance interpretation:

Page 1 of 4 20180301

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response This FAQ concerns the Watts Bar Unit 2 new plant startup and subsequent March 23, 2017 Main Condenser failure that resulted in an estimated loss of 3100 critical hours for repair. The reactor was shut down from March 23, 2017 until July 30, 2017 while extensive repairs were completed to the Main Condenser. The cause of the failure was inadequate vendor design (1970s vintage) of the condenser wall support structure leading to support and wall failure. In addition, an extended 39 day refueling outage was completed in the fourth Quarter of 2017. This resulted in an additional estimated loss of 930 critical hours. Being the first refueling outage following WBN Unit 2 commercial operation, many additional tests were required to meet commitments as dictated by the operating license. This resulted in a longer than baseline outage.

The main condenser repairs coupled with the extended refueling outage has resulted in a low number of critical hours (approximately 4588) for the period defined in the Oct 22, 2015 letter. For related background, WBN Unit 2 experienced two scrams and one unplanned power change for the previous 4 quarters. Details are as follows:

• A 1Q17 scram was caused when workers inadvertently depressed a local trip pushbutton on a Hotwell Pump. The pump trip resulted in a secondary plant transient and subsequent reactor scram. The event was attributed to human performance in that workers failed to practice situational awareness around scram sensitive equipment. Corrective actions included coaching Operations personnel on the need to control work activities near operating equipment and installation of bump guard covers on local pushbuttons for a number of Unit 2 secondary pumps.

• A 4Q17 scram was caused by an intermittent circuit card connection in the 2AC Rod Control Power Cabinet. The equipment malfunction resulted in 4 dropped control rods and a subsequent manual reactor scram by control room operators. Corrective actions included a 100% inspection of circuit card connections in the Rod Control Power Cabinets and replacement of suspect cards. No common cause was assessed to exist between the two scrams.

• A 3Q17 unplanned power change was caused by a Main Turbine steam leak.

If licensee and NRC resident/region do not agree on the facts and circumstances explain:

The NRC Watts Bar Site Resident Inspector was informed of this FAQ.

Potentially relevant FAQs:

FAQ 13-01 Turkey Point Unplanned Scrams per 7000 Hours Critical FAQ 17-04 Watts Bar Unit 2 MSPI Effectiveness Date Response Section:

Proposed Resolution of FAQ:

Due to the uniqueness of new construction and starting-up a new unit, TVA requests a two quarter extension to the effective date for WBN Unit 2 IE01 and IE03 indicators (July 1, 2018) due to the loss of a significant number of critical hours. The IE01 indicator objective is to limit the frequency of those events that upset plant stability and challenge critical safety functions during power operations. The IE03 indicator monitors the number of unplanned power changes that could challenge safety functions. NEI 99-02 states that the indicators are based on 7000 critical hours which provides allowance for a routine outage. As of December 31, 2017, the total number of reported critical hours Page 2 of 4 20180301

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response for 2017 was 4588. Extending the effective date to July 1, 2018 will allow four quarters of operation after the extended main condenser repair shutdown to provide a representative assessment result.

Additionally and unique to WBN Unit 2 as a newly licensed plant and in an NRC letter dated October 22, 2015 titled Watts Bar Nuclear Plant, Unit 2 - Reactor Oversight Process Implementation and Partial Cornerstone Transition - Docket No. 50-0391, the NRC provided a ROP transition plan. The plan stated IE01, IE03 and some MS performance indicators will not become valid (monitored only) until at least four (4) quarters after the cornerstone has been transitioned to the ROP. WBN Unit 2 transitioned to full ROP oversight on November 21, 2016. The 2015 letter also stated If, as the licensee approaches four quarters after either the IE or MS cornerstones become monitored, new information shows that a PI may still not provide accurate assessment value, the Frequently Asked Questions process will be utilized in accordance with NEI 99-02 to reach a conclusion on how to proceed.

Similar to this FAQ request, FAQ 17-04, Watts Bar Unit 2 MSPI Effectiveness Date, was recently approved by the NRC to grant an extension for MS01 (Emergency AC Power System), MS07 (High Pressure Injection System), MS08 (Heat Removal System) and MS10 (Cooling Water Systems). The basis for this extension was the loss of critical hours within the first 12 months of operation due to the main condenser repair outage.

If appropriate, provide proposed rewording of guidance for inclusion in next revision: None PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No Proposed NRC Response Both the IE01 and IE03 performance indicators are baselined to an occurrence rate per 7,000 critical hours and include a built-in lower limit of 2,400 critical hours, under which the indicator output is N/A to preclude misleadingly high values at low critical hours. The ROP transition letter dated October 22, 2015, noted that, in order to establish the necessary baseline of critical hours to prevent falsely inflating the data, the IE01 and IE03 performance indicators would become valid after four full calendar quarters have passed following cornerstone transition to the ROP. Since the IE cornerstone was transitioned in November 2016, the IE01 and IE03 performance indicators became effective when 4Q2017 data was submitted. As of the end of 4Q2017, Watts Bar Unit 2 had accumulated more than 6,200 critical hours since initial startup, almost 2,700 critical hours following the main condenser maintenance outage, and the IE01 and IE03 performance indicator calculations for 4Q2017 included nearly 4,600 critical hours in the prior four quarters. Given that the 2,400 critical hour lower limit to preclude unreasonably high performance indicator values was vastly exceeded by both critical hours accrued since initial startup and the number of critical hours included in the 4Q2017 IE01 and IE03 reporting period, the staff views the IE01 and IE03 performance indicators as providing reasonable assessment values.

As noted by the licensee in this FAQ, the NRC previously approved an extension request of the MSPI performance indicators in FAQ 17-04. The ROP transition letter from October 22, 2015, referred to a sensitivity study on the impact of low critical hours on MSPI. The study concluded that MSPI, which nominally uses the prior 36 months of data in its calculation, would produce relatively normal values after 12 months of data. The ROP transition letter used this information in determining that the MSPI performance indicators would become valid for Watts Bar Unit 2 after 12 months of operation.

Because of the roughly three months of lost critical hours, the licensee requested an extension of the Page 3 of 4 20180301

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response MSPI performance indicators by one quarter. The staff approved that request since the plant had not yet accrued the 12 months of minimum critical hours necessary for the performance indicators to provide a reasonably accurate assessment value.

The NRC reviewed FAQ 13-01, in which a similar request was made to extend the effective date of the IE01 performance indicator due to low critical hours following an extended outage. The final NRC response to that FAQ denied the request, noted that the IE01 performance indicator already had a built-in minimum limit before it became effective to preclude misleadingly high values, and concluded that low critical hours does not represent a unique condition that would warrant an exemption.

Furthermore, the staff does not view the three month maintenance outage as an outage of sufficient length to reset performance indicator effective dates. An extended shutdown is defined in IMC 0608 as a condition in which a nuclear power reactor has been subcritical for at least six months.

Additionally, the MSPI sensitivity study, while silent on the IE cornerstone, recommended that MSPI performance indicators be grayed out after a six month shutdown or greater. The staff also identified a recent three month refueling and maintenance outage (Grand Gulf in fall 2016) in which no performance indicators were grayed out or otherwise made invalid. The NRC has found no basis in this case to deviate from these prior staff positions.

The staff reviewed the two scrams that were included as inputs to the 4Q2017 performance indicator to determine whether they were indicative of issues unique to a new plant. In one instance workers inadvertently depressed a local trip pushbutton on a hotwell pump, with the resulting transient ultimately causing a plant scram. The second instance involved an intermittent circuit card connection that resulted in four dropped rods and a subsequent manual scram. The staff did not find these scrams to be situations unique to a new plant.

In summary, the NRC staff does not support granting Watts Bar Unit 2 a two quarter extension to the Unplanned Scrams per 7,000 Critical Hours performance indicator or the Unplanned Power Changes per 7,000 Critical Hours performance indicator.

Page 4 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request Plant: Columbia Generating Station Date of Event: August 20, 2017 Submittal Date:

Licensee

Contact:

Desirée Wolfgramm Tel/email: 509-377-4792 dmwolfgramm@energy-northwest.com NRC

Contact:

Alex Garmoe Tel/email: 301-415-3814 alexander.garmoe@nrc.gov Performance Indicator: Unplanned Scrams with Complications Site-Specific FAQ (see Appendix D)? Yes FAQ to become effective when approved.

Question Section Nuclear Energy Institute (NEI) 99-02 Guidance needing interpretation (include page and line citation):

- NEI 99-02, Revision 7, Page 24, lines 45-46, and

- NEI 99-02, Revision 7, Page 25, lines 1-3 Event or circumstances requiring guidance interpretation:

This FAQ this is being submitted to request a plant-specific exemption from the guidance related to Unplanned Scrams with Complications (USwC) for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the reactor pressure vessel (RPV) post scram resulting in a second +13 inch scram common to Boiling Water Reactor (BWR) designs.

On August 20, 2017, Columbia Generating Station (Columbia) operators manually scrammed the reactor in response to condenser vacuum degradation following an air removal valve closure. The scram was performed per procedure to prevent an automatic turbine trip (resulting in an automatic reactor scram), main steam isolation valve (MSIV) closure, and reactor feed turbine trip - all of which occur at various low condenser vacuum setpoints. Condenser vacuum was recovered before any of these actions could occur.

For BWRs, RPV water level responds to changes in RPV pressure following a reactor scram.

Specifically, RPV water level reaches the Emergency Operating Procedure (EOP) entry criteria of Level 3, +13 inches, due to collapsing of voids when the turbine throttle and governor valves go closed to isolate steam flow to the turbine, reference NEI 99-02 FAQ 18-01, Definition of Initial Transient. During this pressure transient, there is no loss of water inventory in the vessel. In response to this, operators will enter the EOPs (Plant Procedures Manual (PPM) 5.1.1 for Columbia) and initial RPV level control will be with the feedwater level control system in automatic and RPV pressure control will be with the turbine bypass valve control system in automatic. Operators will then transition level control from reactor feed turbine automatic control to throttling with the start-up level control valves in automatic control. This transition is directed per procedures and allows for more precise level control in low feed flow conditions. During this transition, operators take manual control of the reactor feed turbine speed and the level control valve position per procedure. An initial level band is established using available injection systems in order of preference. For Columbia this band is from

+13 inches to +54 inches. The feedwater level control valves control level in automatic. During a normal BWR scram, pressure drops during the initial void collapse and will restore automatically once the turbine bypass valve control system responds.

For the event on August 20, 2017, following the reactor scram, condenser vacuum continued to slowly deteriorate. Continued degradation of condenser vacuum could result in closure of the MSIVs Page 1 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request and a loss of reactor feed pumps. Prior to exiting the normal scram response procedure (PPM 3.3.1) and EOP (PPM 5.1.1) approximately 18 minutes after the initial scram, operators took action to lower RPV pressure to maintain the availability of the condensate system to control RPV level. This was an intentional operator action to reduce reactor pressure to maximize the time that the condenser could be used to reject energy from the RPV. It also rejected energy into the main condenser that would otherwise have been rejected to the suppression pool through the main steam safety/relief valves (SRVs) following MSIV closure. During the pressure reduction, the allowable cool down rates were not exceeded.

At the beginning of the pressure reduction, operators controlled RPV level in automatic midway between the established level band of +13 inches to +54 inches at the normal value of +36 inches.

This value can allow for RPV level swell, which occurs at the beginning of the pressure reduction, similar to effects seen in a steam generator for a Pressurized Water Reactor (PWR). Control of initial RPV level is crucial to prevent the RPV level swell from reaching the Level 8 setpoint, which occurs at

+54.5 inches. At Level 8, the reactor feed pumps trip and the high pressure core spray (HPCS) and reactor core isolation and cooling (RCIC) system injections automatically terminate, if running.

When the desired pressure was attained the turbine bypass valves were throttled closed to terminate the pressure reduction thereby maintaining RPV pressure in the specified band. This resulted in an expected shrinkage of RPV water level due to collapsing of voids, similar to what occurs following a BWR scram. The RPV water level again momentarily dropped below the +13 inch (Level 3) setpoint while the feedwater level control system responded. Reactor water level was restored to normal levels automatically in less than a minute without operator action. The effects of swelling and shrinkage do not represent a loss of inventory in the reactor pressure vessel.

BWR operators need to account for swelling and shrinkage when depressurizing the RPV. However, while reaching the Level 8 setpoint upon water level swell will result in undesirable termination of inventory injection systems, reaching the Level 3 setpoint upon water level shrinkage does not result in any undesirable effects. That is, for BWRs, following a scram and initial RPV level excursion below

+13 inches, there is no operational impact to a subsequent momentary level excursion below +13 inches during a controlled fast reduction in RPV pressure when the scram has not yet been reset since there are no additional actuations or complications. Inventory is not lost during shrinkage, all feedwater capability is still available and condensate is in service. Due to the condensate system in service and the feedwater system availability, reaching the Level 3 setpoint has no operational impact. In comparison, at Level 2, -50 inches, HPCS and RCIC, among other systems, will initiate automatically when accident conditions exist. The actuations at Level 2 create additional operator action to secure systems once started.

Per NEI 99-02 Rev 7 guidance, page 25, this scram was counted as an USwC due to the second reactor water level scram signal during the scram response. Energy Northwest requests an exemption from reporting as an USwC due to the unique circumstances of this event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common to BWR designs. For BWRs post-scram responses in which a rapid RPV pressure reduction results in a subsequent reactor water level scram signal provides no additional operational challenges when the RPV level response was the result of intended operator actions, no accident conditions exist, available systems automatically recover level to above the Level 3 (scram) setpoint, and no additional actuations or complications occur.

NEI 99-02 Rev 7 page 19 states that the purpose for the USwC is to monitor scrams that either require additional operator actions beyond that of the normal scram or involve the unavailability of or inability to recover main feedwater. Common to BWR plant designs, a controlled fast reduction in Page 2 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request RPV pressure performed as part of approved procedures using forethought and operational knowledge which results in a momentary low level below the scram setpoint presents no additional operator action to restore RPV level. Feedwater flow remains available and able to automatically recover RPV level. Therefore, for a BWR, this event does not meet the intent of the complicated scram Performance Indicator (PI). As discussed in FAQ 18-01, Definition of Initial Transient, the expected collapsing of voids did not represent an inventory loss and feedwater from both main feedwater pumps was available during the transient, therefore no abnormal condition pertaining to water inventory existed.

NEI 99-02 Rev 7 page 24 lines 45-46 and page 25 line 1 states the following:

The requirement to remain in the EOPs because of reactor pressure/water level and drywell pressure following the initial transient indicates complications beyond the typical reactor scram.

As described in the event for Columbia and typical of BWR plant response, the initial expected level excursion below +13 inches requires entry into the EOPs as discussed in FAQ 18-01, Definition of Initial Transient. However, no additional actions were taken in the EOPs to restore RPV level for the expected first or second level excursion as no emergency existed, and the feedwater level control system operated as designed; therefore, there was no requirement to remain in the EOPs.

From page 25 lines 2-3, Additionally, reactor water level scram signal(s) during the scram response indicate level could not be stabilized and require this question be answered Yes. Although a BWR experiences a reactor water level scram signal at the +13 inch setpoint during a controlled fast reduction in RPV pressure due to void collapse, this does not indicate that RPV level cannot be stabilized. As experienced by Columbias event on August 20, 2017, and then subsequently demonstrated in Columbias simulator, the subsequent +13 inch RPV water level excursion is an expected evolution that lasts for less than a minute and is automatically restored and stabilized by the feedwater level control system.

If licensee and NRC resident/region do not agree on the facts and circumstances, explain:

This event was counted as an Unplanned Scram with Complications due to the second reactor water level scram signal during the scram response. The licensee asks that the NRC reconsider this event as an uncomplicated scram for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common to BWR designs. The language in NEI 99-02 for this PI is overly restrictive and does not allow for events such as this where there are no operational impacts of momentarily reaching additional reactor water level scram signals where no emergency exists. As described above no emergency existed pertaining to reactor pressure or level, and operators were not required to remain in the EOPs. This level excursion below +13 inches was an expected evolution and did not present additional challenges to the plant operators NEI 99-02 page 19 line 6.

Potentially relevant FAQs:

FAQ 18-01, Definition of Initial Transient Response Section Proposed Resolution of FAQ:

This FAQ is proposed as a plant-specific exemption for this event as an uncomplicated scram for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common Page 3 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request to BWR designs. This event was the result of intended operator actions, no accident conditions existed, available systems automatically recovered reactor water level above the scram setpoint, and no additional actuations or complications occurred.

If appropriate, provide proposed rewording of guidance for inclusion in next revision:

NA PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No

Attachment:

August Scram Timeline Page 4 of 4 20180301

NEI 99-02 FAQ 18-01 Definition of Initial Transient - Tentatively Approved NOTE This FAQ would implement a whitepaper that proposed clarifications of the definition of Initial Transient. The whitepaper was discussed with the NRC staff in public ROP meetings in 2013-2014. The final discussion of the whitepaper occurred at a May 14, 2014 public meeting.

The NRC staff member who had the lead on performance indicators at the time was Andrew Waugh, who is listed below as the NRC Contact. The concluding discussion is documented in an NRC meeting summary available in ADAMS at accession number ML14149A293.

The proposed text changes presented below reflect NRC comments and suggested edits for agency approval presented in a mark-up of the whitepaper attached to the aforementioned meeting summary. The marked-up whitepaper is available under ADAMS accession number ML14149A278.

Plant: Generic Date of Event: September 11, 2014 Submittal Date: September 11, 2014 Licensee

Contact:

Lenny Sueper Tel/email: 612-330-6917 / Leonard.Sueper@xenuclear.com NRC

Contact:

Andrew Waugh Tel/email: (301) 415-5601 / andrew.waugh@nrc.gov Performance Indicator: IE04 - Unplanned Scrams with Complications Site-Specific FAQ (see Appendix D)? No - this is generic FAQ to become effective: When approved Question Section NEI 99-02 Guidance needing interpretation (include page and line citation):

Page 23 Line 20:

20 Was pressure control unable to be established following the initial transient?

Page 24 Lines 39 - 40:

39 Following initial transient, did stabilization of reactor pressure/level and drywell pressure 40 meet the entry conditions for EOPs?

Event or circumstances requiring guidance interpretation:

Two of the questions in NEI 99-02 used to determine if a BWR reactor trip was an Unplanned Scram with Complications include the undefined term initial transient; Was pressure control unable to be established following initial transient? and Following initial transient did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? The failure to define the term has resulted in confusion, with some licensees interpreting initial transient to be equivalent to scram response.

If licensee and NRC resident/region do not agree on the facts and circumstances, explain:

N/A Potentially relevant FAQs: None Response Section Proposed Resolution of FAQ:

The following is proposed to be added in the Definition of Terms section of this indicator:

Initial Transient is intended to envelope the immediate and expected changes to BWR parameters as a result of a scram (e.g., pressure, level, etc.) because of the collapsing of voids in the core and the routine response of the main feedwater and turbine control systems. For example, at some BWRs the reflected pressure wave Page 1 of 2 20180207

NEI 99-02 FAQ 18-01 Definition of Initial Transient - Tentatively Approved resulting from the rapid closure of turbine valves during a turbine trip may result in a pressure spike in the reactor vessel that causes one or more safety-relief valves (SRVs) to briefly lift. The intent is to allow a licensee to exclude the momentary operation of SRVs when answering Was pressure control unable to be established? The sustained or repeated operation of SRVs in response to turbine control bypass valve failures or Main Steam Isolation Valve (Group I) isolations are not a part of routine BWR scram responses and are therefore not considered to occur within the initial transient. Similarly, a reactor level decrease to Level 3 following a reactor trip due to the expected collapsing of voids in the core can be excluded when answering the question Following initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? as long as the feedwater control system and at least one feedwater pump were operating as designed. Initial transient is different from scram response. The initial transient is a subset of the overall scram response time.

If appropriate, provide proposed rewording of guidance for inclusion in next revision:

See above.

PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No Proposed NRC Response:

NRC agrees with the proposed language for initial transient with the addition of the following words:

Initial Transient is intended to envelope the immediate and expected changes to BWR parameters as a result of a scram (e.g., pressure, level, etc.) because of the collapsing of voids in the core and the routine response of the main feedwater and turbine control systems. For example, at some BWRs the reflected pressure wave resulting from the rapid closure of turbine valves during a turbine trip may result in a pressure spike in the reactor vessel that causes one or more safety-relief valves (SRVs) to briefly lift. The intent is to allow a licensee to exclude the momentary operation of SRVs when answering Was pressure control unable to be established? The sustained or repeated operation of SRVs in response to turbine control bypass valve failures or Main Steam Isolation Valve (Group I) isolations are not a part of routine BWR scram responses and are therefore not considered to occur within the initial transient. Similarly, an initial reactor level decrease to Level 3 immediately following a reactor trip due to the expected collapsing of voids in the core can be excluded when answering the question Following initial transient, did stabilization of reactor pressure/level and drywell pressure meet the entry conditions for EOPs? as long as the feedwater control system and at least one feedwater pump were operating as designed. Initial transient is different from scram response.

The initial transient is a subset of the overall scram response time.

Page 2 of 2 20180207

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response Plant: Watts Bar Nuclear Plant, Unit 2 (WBN 2)

Date of Event: 12/31/2017 Submittal Date: 2/21/2018 Engineer/Licensee

Contact:

Kim Hulvey/Beth Wetzel Tel/email: (423) 365-7720/(423)751-2403 NRC Contact Jared Nadel Watts Bar Tel/email: (423) 365-1776 Performance Indicators:

IE01 WBNU2 Unplanned Scrams per 7000 Critical Hours (automatic and manual scrams during the previous four quarters)

IE03 WBNU2 Unplanned Power Changes per 7000 Critical Hours (over previous four quarters)

Site-Specific FAQ (Appendix D)? - Yes FAQ to become effective when approved.

Question Section:

TVA requests the effective date of Watts Bar Unit 2 Unplanned Scrams per 7000 Critical Hours (IE01) and (IE03) Unplanned Power Changes per 7000 Critical Hours be extended until 3Q18 (through Jun 30, 2018) to allow sufficient data for an accurate assessment value. This request is based upon a October 22, 2015 NRC letter to TVA stating If, as the licensee approaches four quarters after either the IE or MS cornerstones become monitored, new information shows that a PI may still not provide accurate assessment value, the Frequently Asked Questions process will be utilized in accordance with NEI 99-02 to reach a conclusion on how to proceed.

NEI 99-02 Guidance needing interpretation:

NRC Letters to TVA dated November 21, 2016 (ML16326A210) and October 22, 2015 (ML15295A253).

NEI 99-02 Page 10 line 25 The number of unplanned scrams during the previous four quarters, both manual and automatic, while critical per 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

NEI 99-02 Page 14 line 9 The number of unplanned changes in reactor power of greater than 20% of full-power, per 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of critical operation excluding manual and automatic scrams.

NEI 99-02 Page E-1 line 12 There are several reasons for submitting an FAQ:

NEI 99-02 Page E-1 line18

3. To request an exemption from the guidance for plant-specific circumstances, such as design features, procedures, or unique conditions.

Event or circumstances requiring guidance interpretation:

This FAQ concerns the Watts Bar Unit 2 new plant startup and subsequent March 23, 2017 Main Condenser failure that resulted in an estimated loss of 3100 critical hours for repair. The Page 1 of 4 20180301

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response reactor was shut down from March 23, 2017 until July 30, 2017 while extensive repairs were completed to the Main Condenser. The cause of the failure was inadequate vendor design (1970s vintage) of the condenser wall support structure leading to support and wall failure. In addition, an extended 39 day refueling outage was completed in the fourth Quarter of 2017.

This resulted in an additional estimated loss of 930 critical hours. Being the first refueling outage following WBN Unit 2 commercial operation, many additional tests were required to meet commitments as dictated by the operating license. This resulted in a longer than baseline outage.

The main condenser repairs coupled with the extended refueling outage has resulted in a low number of critical hours (approximately 4588) for the period defined in the Oct 22, 2015 letter.

For related background, WBN Unit 2 experienced two scrams and one unplanned power change for the previous 4 quarters. Details are as follows:

• A 1Q17 scram was caused when workers inadvertently depressed a local trip pushbutton on a Hotwell Pump. The pump trip resulted in a secondary plant transient and subsequent reactor scram. The event was attributed to human performance in that workers failed to practice situational awareness around scram sensitive equipment.

Corrective actions included coaching Operations personnel on the need to control work activities near operating equipment and installation of bump guard covers on local pushbuttons for a number of Unit 2 secondary pumps.

• A 4Q17 scram was caused by an intermittent circuit card connection in the 2AC Rod Control Power Cabinet. The equipment malfunction resulted in 4 dropped control rods and a subsequent manual reactor scram by control room operators. Corrective actions included a 100% inspection of circuit card connections in the Rod Control Power Cabinets and replacement of suspect cards. No common cause was assessed to exist between the two scrams.

• A 3Q17 unplanned power change was caused by a Main Turbine steam leak.

If licensee and NRC resident/region do not agree on the facts and circumstances explain:

The NRC Watts Bar Site Resident Inspector was informed of this FAQ.

Potentially relevant FAQs:

FAQ 13-01 Turkey Point Unplanned Scrams per 7000 Hours Critical FAQ 17-04 Watts Bar Unit 2 MSPI Effectiveness Date Response Section:

Proposed Resolution of FAQ:

Due to the uniqueness of new construction and starting-up a new unit, TVA requests a two quarter extension to the effective date for WBN Unit 2 IE01 and IE03 indicators (July 1, 2018) due to the loss of a significant number of critical hours. The IE01 indicator objective is to limit the frequency of those events that upset plant stability and challenge critical safety functions during power operations. The IE03 indicator monitors the number of unplanned power changes that could challenge safety functions. NEI 99-02 states that the indicators are based on 7000 critical hours which provides allowance for a routine outage. As of December 31, 2017, the total number of reported critical hours for 2017 was 4588. Extending the effective date to July 1, 2018 will allow four quarters of operation after the extended main condenser repair shutdown to provide a representative assessment result.

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FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response Additionally and unique to WBN Unit 2 as a newly licensed plant and in an NRC letter dated October 22, 2015 titled Watts Bar Nuclear Plant, Unit 2 - Reactor Oversight Process Implementation and Partial Cornerstone Transition - Docket No. 50-0391, the NRC provided a ROP transition plan. The plan stated IE01, IE03 and some MS performance indicators will not become valid (monitored only) until at least four (4) quarters after the cornerstone has been transitioned to the ROP. WBN Unit 2 transitioned to full ROP oversight on November 21, 2016.

The 2015 letter also stated If, as the licensee approaches four quarters after either the IE or MS cornerstones become monitored, new information shows that a PI may still not provide accurate assessment value, the Frequently Asked Questions process will be utilized in accordance with NEI 99-02 to reach a conclusion on how to proceed.

Similar to this FAQ request, FAQ 17-04, Watts Bar Unit 2 MSPI Effectiveness Date, was recently approved by the NRC to grant an extension for MS01 (Emergency AC Power System),

MS07 (High Pressure Injection System), MS08 (Heat Removal System) and MS10 (Cooling Water Systems). The basis for this extension was the loss of critical hours within the first 12 months of operation due to the main condenser repair outage.

If appropriate, provide proposed rewording of guidance for inclusion in next revision:

None PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No Proposed NRC Response:

Both the IE01 and IE03 performance indicators are baselined to an occurrence rate per 7,000 critical hours and include a built-in lower limit of 2,400 critical hours, under which the indicator output is N/A to preclude misleadingly high values at low critical hours. The ROP transition letter dated October 22, 2015, noted that, in order to establish the necessary baseline of critical hours to prevent falsely inflating the data, the IE01 and IE03 performance indicators would become valid after four full calendar quarters have passed following cornerstone transition to the ROP. Since the IE cornerstone was transitioned in November 2016, the IE01 and IE03 performance indicators became effective when 4Q2017 data was submitted. As of the end of 4Q2017, Watts Bar Unit 2 had accumulated more than 6,200 critical hours since initial startup, almost 2,700 critical hours following the main condenser maintenance outage, and the IE01 and IE03 performance indicator calculations for 4Q2017 included nearly 4,600 critical hours in the prior four quarters. Given that the 2,400 critical hour lower limit to preclude unreasonably high performance indicator values was vastly exceeded by both critical hours accrued since initial startup and the number of critical hours included in the 4Q2017 IE01 and IE03 reporting period, the staff views the IE01 and IE03 performance indicators as providing reasonable assessment values.

As noted by the licensee in this FAQ, the NRC previously approved an extension request of the MSPI performance indicators in FAQ 17-04. The ROP transition letter from October 22, 2015, referred to a sensitivity study on the impact of low critical hours on MSPI. The study concluded that MSPI, which nominally uses the prior 36 months of data in its calculation, would produce relatively normal values after 12 months of data. The ROP transition letter used this information in determining that the MSPI performance indicators would become valid for Watts Bar Unit 2 after 12 months of operation. Because of the roughly three months of lost critical hours, the Page 3 of 4 20180301

FAQ 18-02 Watts Bar Critical Hours - Proposed NRC Response licensee requested an extension of the MSPI performance indicators by one quarter. The staff approved that request since the plant had not yet accrued the 12 months of minimum critical hours necessary for the performance indicators to provide a reasonably accurate assessment value.

The NRC reviewed FAQ 13-01, in which a similar request was made to extend the effective date of the IE01 performance indicator due to low critical hours following an extended outage. The final NRC response to that FAQ denied the request, noted that the IE01 performance indicator already had a built-in minimum limit before it became effective to preclude misleadingly high values, and concluded that low critical hours does not represent a unique condition that would warrant an exemption. Furthermore, the staff does not view the three month maintenance outage as an outage of sufficient length to reset performance indicator effective dates. An extended shutdown is defined in IMC 0608 as a condition in which a nuclear power reactor has been subcritical for at least six months. Additionally, the MSPI sensitivity study, while silent on the IE cornerstone, recommended that MSPI performance indicators be grayed out after a six month shutdown or greater. The staff also identified a recent three month refueling and maintenance outage (Grand Gulf in fall 2016) in which no performance indicators were grayed out or otherwise made invalid. The NRC has found no basis in this case to deviate from these prior staff positions.

The staff reviewed the two scrams that were included as inputs to the 4Q2017 performance indicator to determine whether they were indicative of issues unique to a new plant. In one instance workers inadvertently depressed a local trip pushbutton on a hotwell pump, with the resulting transient ultimately causing a plant scram. The second instance involved an intermittent circuit card connection that resulted in four dropped rods and a subsequent manual scram. The staff did not find these scrams to be situations unique to a new plant.

In summary, the NRC staff does not support granting Watts Bar Unit 2 a two quarter extension to the Unplanned Scrams per 7,000 Critical Hours performance indicator or the Unplanned Power Changes per 7,000 Critical Hours performance indicator.

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FAQ 18-03 Columbia Scram Exemption Request Plant: Columbia Generating Station Date of Event: August 20, 2017 Submittal Date:

Licensee

Contact:

Desirée Wolfgramm Tel/email: 509-377-4792 dmwolfgramm@energy-northwest.com NRC

Contact:

Alex Garmoe Tel/email: 301-415-3814 alexander.garmoe@nrc.gov Performance Indicator: Unplanned Scrams with Complications Site-Specific FAQ (see Appendix D)? Yes FAQ to become effective when approved.

Question Section Nuclear Energy Institute (NEI) 99-02 Guidance needing interpretation (include page and line citation):

- NEI 99-02, Revision 7, Page 24, lines 45-46, and

- NEI 99-02, Revision 7, Page 25, lines 1-3 Event or circumstances requiring guidance interpretation:

This FAQ this is being submitted to request a plant-specific exemption from the guidance related to Unplanned Scrams with Complications (USwC) for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the reactor pressure vessel (RPV) post scram resulting in a second +13 inch scram common to Boiling Water Reactor (BWR) designs.

On August 20, 2017, Columbia Generating Station (Columbia) operators manually scrammed the reactor in response to condenser vacuum degradation following an air removal valve closure. The scram was performed per procedure to prevent an automatic turbine trip (resulting in an automatic reactor scram), main steam isolation valve (MSIV) closure, and reactor feed turbine trip - all of which occur at various low condenser vacuum setpoints. Condenser vacuum was recovered before any of these actions could occur.

For BWRs, RPV water level responds to changes in RPV pressure following a reactor scram.

Specifically, RPV water level reaches the Emergency Operating Procedure (EOP) entry criteria of Level 3, +13 inches, due to collapsing of voids when the turbine throttle and governor valves go closed to isolate steam flow to the turbine, reference NEI 99-02 FAQ 18-01, Definition of Initial Transient. During this pressure transient, there is no loss of water inventory in the vessel. In response to this, operators will enter the EOPs (Plant Procedures Manual (PPM) 5.1.1 for Columbia) and initial RPV level control will be with the feedwater level control system in automatic and RPV pressure control will be with the turbine bypass valve control system in automatic. Operators will then transition level control from reactor feed turbine automatic control to throttling with the start-up level control valves in automatic control. This transition is directed per procedures and allows for more precise level control in low feed flow conditions. During this transition, operators take manual Page 1 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request control of the reactor feed turbine speed and the level control valve position per procedure.

An initial level band is established using available injection systems in order of preference.

For Columbia this band is from +13 inches to +54 inches. The feedwater level control valves control level in automatic. During a normal BWR scram, pressure drops during the initial void collapse and will restore automatically once the turbine bypass valve control system responds.

For the event on August 20, 2017, following the reactor scram, condenser vacuum continued to slowly deteriorate. Continued degradation of condenser vacuum could result in closure of the MSIVs and a loss of reactor feed pumps. Prior to exiting the normal scram response procedure (PPM 3.3.1) and EOP (PPM 5.1.1) approximately 18 minutes after the initial scram, operators took action to lower RPV pressure to maintain the availability of the condensate system to control RPV level. This was an intentional operator action to reduce reactor pressure to maximize the time that the condenser could be used to reject energy from the RPV. It also rejected energy into the main condenser that would otherwise have been rejected to the suppression pool through the main steam safety/relief valves (SRVs) following MSIV closure. During the pressure reduction, the allowable cool down rates were not exceeded.

At the beginning of the pressure reduction, operators controlled RPV level in automatic midway between the established level band of +13 inches to +54 inches at the normal value of +36 inches. This value can allow for RPV level swell, which occurs at the beginning of the pressure reduction, similar to effects seen in a steam generator for a Pressurized Water Reactor (PWR). Control of initial RPV level is crucial to prevent the RPV level swell from reaching the Level 8 setpoint, which occurs at +54.5 inches. At Level 8, the reactor feed pumps trip and the high pressure core spray (HPCS) and reactor core isolation and cooling (RCIC) system injections automatically terminate, if running.

When the desired pressure was attained the turbine bypass valves were throttled closed to terminate the pressure reduction thereby maintaining RPV pressure in the specified band.

This resulted in an expected shrinkage of RPV water level due to collapsing of voids, similar to what occurs following a BWR scram. The RPV water level again momentarily dropped below the +13 inch (Level 3) setpoint while the feedwater level control system responded.

Reactor water level was restored to normal levels automatically in less than a minute without operator action. The effects of swelling and shrinkage do not represent a loss of inventory in the reactor pressure vessel.

BWR operators need to account for swelling and shrinkage when depressurizing the RPV.

However, while reaching the Level 8 setpoint upon water level swell will result in undesirable termination of inventory injection systems, reaching the Level 3 setpoint upon water level shrinkage does not result in any undesirable effects. That is, for BWRs, following a scram and initial RPV level excursion below +13 inches, there is no operational impact to a subsequent momentary level excursion below +13 inches during a controlled fast reduction in RPV pressure when the scram has not yet been reset since there are no additional actuations or complications. Inventory is not lost during shrinkage, all feedwater capability is still available and condensate is in service. Due to the condensate system in Page 2 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request service and the feedwater system availability, reaching the Level 3 setpoint has no operational impact. In comparison, at Level 2, -50 inches, HPCS and RCIC, among other systems, will initiate automatically when accident conditions exist. The actuations at Level 2 create additional operator action to secure systems once started.

Per NEI 99-02 Rev 7 guidance, page 25, this scram was counted as an USwC due to the second reactor water level scram signal during the scram response. Energy Northwest requests an exemption from reporting as an USwC due to the unique circumstances of this event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common to BWR designs. For BWRs post-scram responses in which a rapid RPV pressure reduction results in a subsequent reactor water level scram signal provides no additional operational challenges when the RPV level response was the result of intended operator actions, no accident conditions exist, available systems automatically recover level to above the Level 3 (scram) setpoint, and no additional actuations or complications occur.

NEI 99-02 Rev 7 page 19 states that the purpose for the USwC is to monitor scrams that either require additional operator actions beyond that of the normal scram or involve the unavailability of or inability to recover main feedwater. Common to BWR plant designs, a controlled fast reduction in RPV pressure performed as part of approved procedures using forethought and operational knowledge which results in a momentary low level below the scram setpoint presents no additional operator action to restore RPV level. Feedwater flow remains available and able to automatically recover RPV level. Therefore, for a BWR, this event does not meet the intent of the complicated scram Performance Indicator (PI). As discussed in FAQ 18-01, Definition of Initial Transient, the expected collapsing of voids did not represent an inventory loss and feedwater from both main feedwater pumps was available during the transient, therefore no abnormal condition pertaining to water inventory existed.

NEI 99-02 Rev 7 page 24 lines 45-46 and page 25 line 1 states the following: Commented [GA1]: I dont have these lines in NIE 99-02 Rev 7 The requirement to remain in the EOPs because of reactor pressure/water level and drywell pressure following the initial transient indicates complications beyond the typical reactor scram.

As described in the event for Columbia and typical of BWR plant response, the initial expected level excursion below +13 inches requires entry into the EOPs as discussed in FAQ 18-01, Definition of Initial Transient. However, no additional actions were taken in the EOPs to restore RPV level for the expected first or second level excursion as no emergency existed, and the feedwater level control system operated as designed; therefore, there was no requirement to remain in the EOPs.

From page 25 lines 2-3, Additionally, reactor water level scram signal(s) during the scram response indicate level could not be stabilized and require this question be answered Yes. Although a BWR experiences a reactor water level scram signal at the +13 inch setpoint during a controlled fast reduction in RPV pressure due to void collapse, this does Page 3 of 4 20180301

FAQ 18-03 Columbia Scram Exemption Request not indicate that RPV level cannot be stabilized. As experienced by Columbias event on August 20, 2017, and then subsequently demonstrated in Columbias simulator, the subsequent +13 inch RPV water level excursion is an expected evolution that lasts for less than a minute and is automatically restored and stabilized by the feedwater level control system.

If licensee and NRC resident/region do not agree on the facts and circumstances, explain:

This event was counted as an Unplanned Scram with Complications due to the second reactor water level scram signal during the scram response. The licensee asks that the NRC reconsider this event as an uncomplicated scram for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common to BWR designs. The language in NEI 99-02 for this PI is overly restrictive and does not allow for events such as this where there are no operational impacts of momentarily reaching additional reactor water level scram signals where no emergency exists. As described above no emergency existed pertaining to reactor pressure or level, and operators were not required to remain in the EOPs. This level excursion below +13 inches was an expected evolution and did not present additional challenges to the plant operators NEI 99-02 page 19 line 6.

Potentially relevant FAQs:

FAQ 18-01, Definition of Initial Transient Response Section Proposed Resolution of FAQ:

This FAQ is proposed as a plant-specific exemption for this event as an uncomplicated scram for Columbia due to the unique circumstances of the event which led to operators intentionally reducing pressure in the RPV post scram per station procedures resulting in a second +13 inch scram common to BWR designs. This event was the result of intended operator actions, no accident conditions existed, available systems automatically recovered reactor water level above the scram setpoint, and no additional actuations or complications occurred.

If appropriate, provide proposed rewording of guidance for inclusion in next revision:

NA PRA update required to implement this FAQ? No MSPI Basis Document update required to implement this FAQ? No

Attachment:

August Scram Timeline Page 4 of 4 20180301

Attachment to FAQ 18-03 AR-V-1 Closure August 20th, 2017 15:47:15 Sequence of events (times are in seconds):

Time after Actual Clock Description AR-V-1 Time Closure (hh:mm:ss)

(hh:mm:ss) 00:00:00 15:47:15 Air Removal Valve (AR-V-1) closed as observed by vacuum degradation. Vacuum degrades at about 0.26 in Hg per minute.

00:06:43 15:53:58 Operators begin reducing flow from 96Mlb/hr to 70Mlb/hr.This is expected operator response due to degrading condenser vacuum.

00:09:59 15:57:14 Operators complete reducing flow to 70Mlb/hr 00:13:39 16:00:54 Operators begin reducing flow from 70Mlb/hr to 67Mlb/hr 00:14:13 16:01:29 Operators complete reducing flow to 67Mlb/hr 00:18:25 16:05:39 Reactor SCRAM. Reactor Recirculation (RRC) pumps trip on End of Cycle Recirculation Pump Trip (EOC-RPT). Turbine Trips. This is an expected plant response for a scram due to elevated backpressure.

00:18:25 16:05:43 + 13 inch low Reactor Pressure Vessel (RPV) level. Normal plant response as expected for a scram at power.

00:18:47 16:06:02 + 13 inch low RPV level clear. Expected plant response.

Operators begin opening Reactor Feedwater Flow Control Valve (RFW-00:23:57 16:11:13 FCV-10A) manually. Expected operator response per procedure.

Operators manually close RFW-FCV-10A and RFW-FCV-10B. Valves are closed at 16:15:24 (1749). Operators place the startup level control 00:27:15 16:14:30 system in AUTO. Observed based on startup controller demand vs RPV level and obvious tracking of an automatic system.

00:28:14 16:15:29 Operators reduce pressure at 50psig per minute. Final pressure is 550psig. Intentional pressure reduction per operator procedure.

Attachment to FAQ 18-03 00:31:34 16:18:49 Operators start RRC-P-1A. The feedwater control system inserts a large amount of feedwater to compensate for lowering RPV level.

Operators start RRC-P-1A.

00:35:46 16:22:01 Feedwater Pumps trip on Level 8. This is based on plant data from RFW-DPT-4B (B028) and RFW.

00:37:07 16:23:22 Turbine Bypass valves rapidly close with RPV pressure at 550psig. This is expected plant response.

+ 13 inch low RPV level. This is expected plant response after the 00:37:50 16:24:05 closure of the turbine bypass valves after securing the pressure reduction.

00:38:41 16:24:56 + 13 inch low RPV level clear. Expected plant response for systems in automatic.

Turbine Bypass valves go Attachment to FAQ 18-03 closed at the end of the pressure reduction causing large shrink.

Reactor Level (in) 60 RRC-P-1A is started Pressure 50 reduction starts at 50psig/min.

40 30 PLANT DATA 20 10 Operators place the startup level control system in AUTO.

0 1100 1300 1500 1700 1900 2100 2300 2500 2700 Time after AR-V-1 Closure (seconds)

Attachment to FAQ 18-03 Reactor Pressure (psig) 1200 1000 800 600 PLANT DATA 400 200 0

0 500 1000 1500 2000 2500 3000 Time after AR-V-1 Closure (seconds)

Attachment to FAQ 18-03 Reactor Level (in) 60 50 40 30 PLANT DATA SIMULATOR 20 10 0

0 500 1000 1500 2000 2500 3000 Time after AR-V-1 Closure (seconds)

Whitepaper to Change Text of NEI 99-02 Reporting ANS Data Following a Transition to IPAWS Introduction The U.S. Federal Emergency Management Agency (FEMA) has issued policy guidance indicating that the Integrated Public Alert and Warning System (IPAWS) may be used by a State, Tribal, and Local government as a primary or backup means of public alerting and notification; refer to FEMA memorandum, IPAWS Implementation Guidance, dated September 13, 2017. A description of IPAWS may be found here and IPAWS testing is discussed here. Some sites, in conjunction with their offsite response organization (ORO) partners, intend to replace their current siren-based prompt public Alert and Notification System (ANS) with IPAWS (i.e., IPAWS would be the primary means of alert and notification). This whitepaper proposes changes to NEI 99-02 that would clarify the applicability of the ANS Reliability performance indicator (EP03) to a site that has implemented IPAWS as a FEMA-approved prompt public ANS.

NEI 99-02 Section Affected The guidance to be changed is found in the Clarifying Notes section beginning on page 58 at line 2 (in the so-called Clean Copy of NEI 99-02, Rev. 7), or page 61 at line 2 (in the line-in/line-out mark-up version of Rev. 7).

Background

Title 10, Code of Federal Regulations, Part 50, Appendix E,Section IV.D.3, states,

. . . The design objective of the prompt public alert and notification system shall be to have the capability to essentially complete the initial alerting and initiate notification of the public within the plume exposure pathway EPZ within about 15 minutes.

The alerting and notification capability shall additionally include administrative and physical means for a backup method of public alerting and notification capable of being used in the event the primary method of alerting and notification is unavailable during an emergency to alert or notify all or portions of the plume exposure pathway EPZ population. [Emphasis added]

Included is a requirement that, The backup method shall have the capability to alert and notify the public within the plume exposure pathway EPZ, but does not need to meet the 15-minute design objective for the primary prompt public alert and notification system. [Emphasis added]

In 76 Fed. Reg. 72,560, Enhancements to Emergency Preparedness Regulations, dated November 23, 2011, the NRC states:

The intent of the final rule is not to have a duplicate primary ANS, but to have a means of backup alerting and notification in place so the public can be alerted in sufficient time to allow offsite officials to consider a range of protective actions for the public to take in the event of a severe accident with potential offsite radiological consequences.

With respect to the ANS performance indicator, NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, states (top of page 58),

This indicator monitors the reliability of the offsite Alert and Notification System (ANS), a critical link for alerting and notifying the public of the need to take protective actions. It provides the percentage of the sirens that are capable of performing their safety function based on regularly scheduled tests.

Page 1 of 2 20180320

Whitepaper to Change Text of NEI 99-02 Reporting ANS Data Following a Transition to IPAWS The associated Clarifying Notes include (top of page 59):

For those sites that do not have sirens, the performance of the licensees alert and notification system will be evaluated through the NRC baseline inspection program. A site that does not have sirens does not report data for this indicator.

It should also be noted that monitoring of ANS Reliability data formally began with implementation of the Reactor Oversight Process in April, 2000. At that time, there was no regulatory requirement for a backup method of public alerting and notification.

The above discussion makes clear that the ANS Reliability performance indicator applies to sirens comprising a primary prompt public ANS with a design objective to essentially complete the initial alerting of the public within the plume exposure pathway emergency planning zone (EPZ) within about 15 minutes. Upon implementation of a FEMA-approved primary ANS that does not use sirens for prompt public alerting (e.g., one based on IPAWS), the ANS Reliability performance indicator is no longer applicable to the site because sirens, the technology that underlies the indicator, are no longer part of the primary prompt public ANS. The licensee can stop reporting siren test data beginning with the quarter during which IPAWS is implemented as the FEMA-approved primary prompt public ANS.

Thereafter, the performance of the licensees ANS will be evaluated through the NRC baseline inspection program (e.g., NRC Inspection Procedure 71114.02, Alert and Notification System Evaluation) and/or other methods determined by the NRC in conjunction with FEMA.

Proposed Changes to NEI 99-02 Starting on page 59 (Clean version of NEI 99-02), beginning at line 2; or page 61 (line-in/line-out version) also at line 2, replace the existing text with following text.

For those sites that do not have sirens, the performance of the licensees alert and notification system will be evaluated through the NRC baseline inspection program. A site that does not have sirens does not report data for this indicator.

Sites that do not use sirens in the primary prompt public ANS do not report data for this indicator and may stop reporting data beginning with the quarter that a FEMA-approved primary prompt public ANS without sirens is implemented. The performance of the licensees alert and notification system will be evaluated through the NRC baseline inspection program and/or other methods determined by the NRC in conjunction with FEMA.

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Whitepaper to Change Text of NEI 99-02 DEP Form Accuracy Introduction Industry operating experience has indicated that the information contained on initial notification forms can be bifurcated into entries of greater and lesser risk significance. Form entries that are critical to offsite response organization (ORO) protective action decision-making are in the former category. Other entries not critical to ORO decision-making are in the latter category. The criteria in NEI 99-02, Revision 7 1, for assessing the accuracy of an initial notification form consider all the listed form entries to have equal risk significance. This is based on the presumption that an error in any one of them will result in a missed performance indicator opportunity. Thus, currently, licensees are guided to report as Drill/Exercise Performance (DEP) indicator opportunity failures some incorrect form entries that would not materially impact ORO protective action decision-making (i.e., entries of lesser risk significance).

NEI 99-02 Section Affected The guidance to be changed involves assessing the accuracy of an initial notification form for alerting an ORO of an emergency. The affected text is found in the following pages and lines of NEI 99-02, Revision 7: page 47, lines 38 through 43; page 48, lines 1 through 5 and lines 19 and 24.

Discussion The proposed resolution is to revise the guidance for assessing the accuracy of initial notification forms.

The revised accuracy criteria identify form entries that are always required for ORO protective action decision-making (more risk significant) and those that may not be critical to such decisions (potentially less risk significant). If inaccurate, a less risk significant form entry could be corrected on the spot during a notification or by a subsequent notification/communication with no impact on the effectiveness of ORO decisions. To implement the revised criteria, a licensee would need concurrence from the appropriate ORO (i.e., the governmental authority responsible for protective action decision-making).

The ORO concurrence would indicate that an entry is not critical for protective action decision-making and need not be assessed for DEP indicator accuracy. It should be noted that this change is consistent with the initial notification form content guidance in Section II.E.3 of NUREG-0654/FEMA-REP-1, Revision 1 2 and Draft Revision 2 3.

Related to the above discussion, the term Protective Action Recommendation, or PAR, should be added to the accuracy criteria because it subsumes two of the current initial notification form attributes, Whether offsite protective measures are necessary and Potentially affected population and areas.

In other words, a PAR transmits the protective measures identified as necessary for a given population or area (e.g., evacuate a subarea or shelter a town). It should be noted that the DEP indicator Clarifying Notes section already uses the term PAR in a manner consistent with its addition to the form accuracy criteria.

Proposed Changes to NEI 99-02 Starting on page 47 at line 38, replace existing text with the following new text:

• Initial notification form completed appropriate to the event to include (see clarifying notes):

o Plant 1 Regulatory Assessment Performance Indicator Guideline, NEI 99-02, Revision 7, August 31, 2013. A line-in/line-out version, showing changes from Revision 6 to Revision 7, is available on the NRC web site at https://www.nrc.gov/docs/ML1326/ML13261A116.pdf.

2 Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, NUREG-0654/FEMA-REP-1, Revision 1, November 1980. Available at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0654/r1/

3 NUREG-0654/FEMA-REP-1, Draft Revision 2 is available at https://www.nrc.gov/docs/ML1416/ML14163A605.pdf.

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Whitepaper to Change Text of NEI 99-02 DEP Form Accuracy o Class of emergency o Whether a release is taking place o Protective Action Recommendation o EAL number*

o Description of emergency*

o Wind direction and speed*

o Date and time of declaration of emergency*

o Whether the event is a drill or actual event*

o Unit as applicable*

Starting on page 48 at line 19, replace existing text with the following new text:

• It is understood that initial notification forms are negotiated with offsite authorities. At a minimum, the first four form elements listed above should be assessed for accuracy. Any of the six remaining elements, identified with an asterisk, should also be assessed for accuracy if required by an offsite response organization (ORO) to make accurate and timely protective action decisions. To ensure that valid site-specific accuracy criteria are maintained, a licensee should review each of six asterisked form elements with the applicable ORO(s) and determine which need to be assessed for accuracy. The licensee should document ORO concurrence to not assess the accuracy of a form element and retain the documentation for inspection. If the form includes elements in addition to these, those elements need not be assessed for accuracy when determining the DEP PI. It is, however, expected that any errors, whether involving the six asterisked form elements not assessed for DEP PI accuracy or additional elements, would be critiqued and addressed through the plants corrective action process.

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Whitepaper to Change Text of NEI 99-02 ERO PI Credit for BDB Drills Introduction The NRC is expected to issue new drill/exercise requirements in 10 CFR 50.155, Mitigation of Beyond-Design-Basis Events (refer to SECY-16-0142 1). These requirements are separate from the existing drill/exercise requirements in 10 CFR 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities. In anticipation of the issuance of these requirements, the industry has drafted this proposed change to NEI 99-02.

As a method to enhance and assess the effectiveness of training, and an efficiency measure, a licensee may choose to conduct a periodic beyond-design-basis (BDB) event response drill during a regularly scheduled emergency preparedness (EP) drill (i.e., an emergency response organization (DRO) team would be presented with a BDB event scenario). Because of the nature of BDB event response drills (refer to the guidance in NEI 13-06) 2, it may not be possible to grant ERO Participation credit to certain Key Positions under the existing performance indicator guidance because they may not have a DEP opportunity. The inability to grant credit could require the conduct of additional drills to ensure that all ERO Key Position holders have a credited ERO Participation opportunity within the allowed 8-quarter window. Given the training and efficiency value of periodically using a BDB event scenario in EP drills, and recognizing that emergency plan procedures will be implemented during these drills, there should be an allowance for granting ERO Participation credit to all Key Position holders.

NEI 99-02 Section Affected It is proposed to add new guidance for granting credit for ERO Participation to the Clarifying Notes section; no existing text is being replaced.

Discussion The Clarifying Notes section should be revised to add guidance for granting ERO Participation credit to all Key Position holders participating in a BDB event response drill, including those responsible for the performance of DEP opportunities outside the Control Room. This change would be similar to an earlier one for granting credit for participation in hostile action-based drills required by 10 CFR 50, Appendix E.

This credit provision is discussed NEI 99-02, Revision 7, page 56, lines 28 through 37.

Proposed Changes to NEI 99-02 Insert the following new text on page 56, starting at line 39:

Drills conducted pursuant to the requirements of 10 CFR 50.155 and using a scenario based on a beyond-design-basis (BDB) event may have all DEP indicator opportunities performed solely in the Control Room (e.g., declaration of a General Emergency could occur prior to activation of ERO facilities). In these cases, ERO Participation credit can be granted to Key Position holders performing DEP indicator functions outside the Control Room provided that the following applicable criteria are met:

• Emergency Classification - The individual confirms the accuracy of the emergency classification level in effect around the time their facility is activated.

• Notification - The individual performs at least one update notification to an ORO.

1 SECY-16-0142, Final Rule: Mitigation of Beyond-Design-Basis Events, December 15, 2016, available at the following URL:

https://www.nrc.gov/docs/ML1629/ML16291A186.pdf.

2 NEI 13-06, Revision 1, Enhancements to Emergency Response Capabilities for Beyond Design Basis Events and Severe Accidents, February 2016, available at URL https://www.nrc.gov/docs/ML1622/ML16224A618.pdf.

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Whitepaper to Change Text of NEI 99-02 ERO PI Credit for BDB Drills

• PAR - The individual performs at least one PAR assessment, even if the result does not change the existing PAR (e.g., an assessment to confirm a previously transmitted PAR).

Objective evidence shall be documented to demonstrate that the above requirements were met.

If an individual participates in at least one BDB event response drill and at least one hostile action-based (HAB) drill within a three-year period, and these drills do not present a DEP opportunity, then only one of the drills can be credited.

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