ML22299A095

From kanterella
Jump to navigation Jump to search
Smirt 26_Post-Conference Fire Seminar_Paper_Weerakkody
ML22299A095
Person / Time
Issue date: 10/26/2022
From: Sunil Weerakkody
NRC/NRR/DRA
To:
References
Download: ML22299A095 (12)
v • d • e


Text

26thInternational Conference on Structural Mechanics in Reactor Technology (SMiRT 26) 17th International Post-Conference Seminar on FIRE SAFETY IN NUCLEAR POWER PLANTS AND INSTALLATIONS

Using An Integrated Risk-Informed Decision-Making Process to Address High Energy Arcing Fault (HEAF) Issues at United States Nuclear Power Plants

Sunil D. Weerakkody

United States Nuclear Regulatory Commission, Washington, DC, 20 555-0001, United States of America

ABSTR ACT

In June 2013, the Organization for Economic Co-operation and De velopment (OECD) Nuclear Energy Agency (NEA) issued a report entitled Analysis of High-Energy Arcing Fault Fire Events (HEAF) [1], describing the international operating expe rience for 48 high energy arc-ing fault (HEAF) events occurred at nuclear power plants (NPPs). At that time, these HEAF events accounted for approximately ten percent of all fire even ts collected in the OECD/NEA FIRE (Fire Events Records Exchange) Database. This effort highlighted concerns about the magnitude of HEAF risk to overall risk at nuclear power plants (NPPs). The United States Nuclear Regulatory Commission (U.S. NRC) has a special interest in fire events, and particu-l a r l y H E A F r i s k s, s i n c e ( a ) r i s k s a s s o c i a t e d w i t h f i r e s c o n s t i tute a large fraction of the total core damage frequency, and (b) approximately 50 % of the U.S. NPPs h ave adopted NFPA 05, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants [2], which relies on Fire PRAs (Probabilistic Risk Assessments). In addition, testing conducted under the OECD/NEA HEAF Project identified that PRA m ethods may not ade-quately address the zones of influence (ZOIs) associated with c ertain HEAF events.

Therefore, NRCs Office of Nuclear Regulatory Research (RES), i n collaboration with the Elec-tric Power Research Institute (EPRI) and the OECD/NEA embarked on an initiative to enhance the state-of-the art technology in using PRA in assessing the impacts due to HEAFs. Specifi-cally, one objective of this initiative was to examine the vali dity of the PRA method documented in NUREG/CR-6850 [3] entitled Fire Probabilistic Risk Assessme nt Methods, using addi-tional operating experience, tests, and analyses performed sinc e the publication of that docu-ment. In parallel with this research effort, in 2021, the NRC i nvestigated the HEAF issue using NRCs Office of Nuclear Reactor Regulation (NRR) Office Instruc tion, LIC-504 entitled Inte-grated Risk-Informed Decisionmaking Process for Emergent Issues [4] to apply best available information and NRC risk assessment tools to determine whether the NRC should take prompt and/or longer-term regulatory actions to ensure that HEAF risks to the public remain at ac-ceptable levels. This paper explains how the NRC used its Integ rated Risk-Informed Decision-Making (RIDM) process via the LIC-504 process to address potent ial safety concerns associ-ated with HEAF.

INTRODUCTION

The current HEAF PRA modelling methodology accepted by the NRC as documented in NUREG/CR-6850 was first published in 2005 and addresses HEAFs a ssociated with electrical switchgears. That method was based primarily on the evaluation of a limited number of HEAF events. Supplement 1 to NUREG/CR-6850 [3] was published in 2010 and addresses HEAFs from bus ducts. Since the publication of these HEAF methods, th e NRC, in collaboration with EPRI, has updated and issued the HEAF PRA methodology [5] for p ublic comment using more recent operating experience, testing, and other enhancements to fire modelling. (NRC expects to issue the final report after dispositioning public comments during the fiscal year 2022). Some of the key advances to the new HEAF PRA methodology include the following:

  • Changes to HEAF frequencies and non-suppression failure probab ilities;
  • Substantial changes to ZOIs for non-isophase bus ducts and for low and medium volt-age switchgears;
  • Crediting qualified electrical raceway fire barrier systems (E R F B S ) i n t h e H E A F Z O I a s a means of preventing damage from HEAF effects;
  • Changes to HEAF frequencies;
  • More realistic HEAF damage potential that considers factors su ch as arc duration.

The above list constitutes significant changes to the PRA asses sment methodologies of HEAF. The NRC used the updated HEAF PRA method above to examine changes to the estimated HEAF risks by comparing the current HEAF PRA methodol ogy described in NUREG/CR-6850 [3] to the updated HEAF PRA methodology. In addit ion to comparing quan-tified risks, consistent with NRC risk-informed decision-making (RIDM) practices, the LIC-504 teams analysis also included a review of the HEAF related info rmation to develop recommen-dations that could assist plant operators to maintain or reduce HEAF related risks at their facilities and to assist the NRCs inspection staff to further risk-inform HEAF related oversight activities.

This paper includes the following information:

The first section describes the motivation for and development of the NRCs LIC-504 process.

The second section of this paper summarizes the approach, resul ts, risk-informed insights, and observations obtained by comparing estimated risks for two reference U.S. NPPs) via the LIC-504 process. In the third section the NRC LIC-504 teams ap proach, results, risk-informed insights, and observation obtained by reviewing other HEAF rela ted operating experience are summarized. A fourth section provides the regulatory processes that the LIC-504 team used to generate its risk-informed recommendations. The fifth sectio n of the paper provides the recommendations developed by the LIC-504 team, and finally Conc lusions are provided.

INCEPTION OF THE LIC-504 PROCESS

The LIC-504 process grew out of a lesson learned initiative fro m a risk significant event that occurred at the Davis Besse NPP. Specifically, during an inspec tion of the control rod drive mechanism (CRDM) nozzles in February 2002 at the Davis Besse NP P, the licensee discov-ered significant degradation of the reactor pressure boundary [ 6]. Subsequent investigation revealed that a circumferential crack in one of the CRDM nozzle s had led to leakage and boric acid corrosion that formed a cavity around the nozzle in the lo w-alloy steel portion of the re-actor pressure vessel (RPV) head. This left only the stainless steel-clad material to maintain the reactor coolant pressure boundary over an area of approxima tely 16.5 square-inches.

The risk significance of this event was analysed under NRCs Ac cident Sequence Precursor (ASP) study program. The ASP program is described in more detai l in below. The NRC staff estimated that the degraded condition that existed imposed an a dditional core damage prob-ability (CDP) of 6 x 10-3 during a one-year period. Since this value exceeded the ASP pr o-gram significant precursor threshold (i.e., greater than or e qual to 1 x 10-3 CDP), this event

2 was reportable in the NRCs annual Abnormal Occurrence Report [ 7]. Subsequently, the U.S.

General Accounting Office (GAO) (now known as the Government Ac countability Office), in 2004, documented its findings pertaining to the Davis Bessie ev ent in its report GAO-04-415, entitled Nuclear Regulation - NRC Needs to More Aggressively a nd Comprehensively Re-solve Issues Related to the Davis-Besse Nuclear Power Plants S hutdown [8]. In the areas of risk evaluation, communication, and the decision-making process for determining if plant shut-down is warranted, the GAO made two recommendations:

1. Develop specific guidance and a well-defined process for dec iding when to shut down a nuclear power plant. The guidance should clearly set out the pr ocess to be used, the safety related factors to be considered, the weight that should be assigned to each factor, and the standards for judging the quality of the evidence consi dered.
2. Improve the NRCs use of PRA estimates in decision-making by ensuring that the risk estimates, uncertainties, and assumptions made in developing th e estimates are fully de-fined, documented, and communicated to NRC decisionmakers and p rovide guidance to decisionmakers on how to consider the relative importance, vali dity, and reliability of quan-titative risk estimates in conjunction with other qualitative s afety related factors.

In response to these recommendations, the NRC developed office instructions entitled LIC-504, Integrated Risk-Informed Decision Making for Emergent Issu es [4] and LIC-106: Issu-ance of Safety Orders [9].

APPROACH, RESULTS, AND RISK-INFORMED INSIGHTS FROM THE ANALYSES OF TWO REFERENCE PLANTS

This section summarizes the approach, results, and risk insight s obtained from the quantita-tive risk analyses performed by the LIC-504 team with the assis tance from the PRA practition-ers at two reference plants. Details of these analyses are provided in Enclosure 1 to the NRCs LIC-504 Team Memorandum [10].

The staff secured the support of two licensees and selected two reference plants, including a Boiling Water Reactor 4 with a Mark I containment and a three-l oop Pressurized Water Reactor (PWR) with a large dry containment. The use of these reference plants enabled the NRC staff to compare the estimated risks between the current NUREG/CR-685 0 HEAF PRA methodol-ogy [3] and the new HEAF PRA methodology.

The staff noted that there was some variation in how each refer ence plant addressed HEAF modelling. For example, one reference plant credited post-Fukus hima Daiichi FLEX strategies (diverse and flexible coping strategies) added to their facilit y in response to a NRC post-Fuku-shima order with regards to beyond-design-basis (BDB) external events EA-12-049 [12] in their PRA. The other reference plant did not. The reference pla nts also used different PRA methods and levels of refinement to develop the HEAF risk. The staff attributes this latter difference primarily to the different reference plant philosoph ies for evaluating fire risk; one reference plant exercised its model extensively to refine its f ire risk, while the other plant con-cluded that a simpler level of detail was adequate to meet thei r intended objectives.

The LIC-504 team leveraged insights derived from the reference plants HEAF PRA analyses to support the quantitative analysis. For example, the dominant sequences of the fire PRA HEAF scenarios were a key input that the staff used to select t he areas for the plant walkdowns. The plant walkdowns enabled the LIC-504 team to dete rmine how HEAF scenar-ios and associated frequencies could be modified to capture the changes related to the up-dated HEAF Fire PRA methodology. The walkdowns were instrumenta l in identifying which additional targets would be impacted and which could be elimina ted. The larger ZOIs for cer-tain configurations associated with the updated HEAF PRA method ology expanded the

3 number of targets in some areas. Whereas some ERFBS protected s cenario targets in the HEAF ZOIs were eliminated, as they were previously assumed to f ail according to the current NUREG/CR-6850 guidance.

Table 1 below summarizes the results for the two reference plan ts base HEAF related risks using the current NUREG/CR-6850 guidance [3] versus the updated HEAF methodology.

Table 1 Comparison of core damage frequency (CDF) and large early relea se fre-quency (LERF) using current versus new HEAF PRA guidance (all n umbers are events/year)

Description CDF CDF CDF LERF LERF LERF (Updated (Current (Updated (Current Method) NUREG/CR-method) NUREG/CR-6850 Method) 6850 Method)

Reference Plant No. 1

SWGR related 1.7 E-06 1.3E-05 -1.1 E-05 3.9E-08 3.5 E-07 -3.2 E-07 Bus Duct related 5.0 E-07 4.6 E-07 4.5 E-08 3.6 E-08 1.5 E-08 2.0 E -08 Total HEAF risk 2.2 E-06 1.4 E-05 -1.1 E-05 7.5 E-0 8 3.7 E-08 -3.0 E 07

Reference Plant No. 2

SWGR related 8.7 E-07 3.7 E-07 5.0 E-07 2.2 E-08 1.2 E-08 9.2 E-09 Bus duct related 3.3 E-05 1.4 E-07 3.3 E-05 3.7 E-06 7.4 E-09 3.7 E-06 Total HEAF risk 3.4 E-05 5.1 E-07 3.4 E-05 3.7 E-06 1.9 E-08 3.7 E-06

For Reference Plant No. 1, the reduction in HEAF risk associate d with the new method is largely driven by the reduction in switchgear related HEAF risk s. The ability of the new method to credit protection from the ERFBS and the relatively small ar c duration time for the reference plant (i.e., short electrical fault clearing time), which reduc es the energy released from the HEAFs and, consequently reduces the ZOIs. For Reference Plant N o. 2, the increase in HEAF risk is dominated by the estimated risk increases associated wi th the bus ducts due to in-creased ZOIs, and the potential for damaging additional targets.

To further refine the staffs perspective on the risk significa nce, several sensitivity studies were also performed. Details on these sensitivity studies are provid ed in [10], which led to risk-informed insights.

The risk-informed insights given below are based on the informa tion obtained from the two reference plants. It is important to emphasize that since the H EAF related risks are highly plant specific, they may not be applicable to other plants. Also, as conveyed below, staff noted some increases as well as some decreases in risk when the new HEAF m ethodology was applied.

However, it is important to note that based on the results of t he overall staffs assessments, the staff concluded that there was no significant increase in t otal HEAF risk, warranting the need for any additional regulatory requirements.

  • Application of the new methodology for bus duct HEAFs provided a significant increase in estimated risk in many, but not all, cases. The instances that showed significant increases in risk were attributed to larger ZOIs resulting from the new H EAF PRA methodology. The major difference between the new HEAF PRA methodology and the e xisting NUREG/CR-6850 Supplement 1 methodology is the assignment of larger ZOIs for long fault duration times. Thus, the staff concludes that those plants with relativ ely long fault clearing times, and consequently larger ZOIs for bus ducts, could experience a significant increase in risk due to HEAFs.

4

  • Application of the updated HEAF methodology for switchgear HEA Fs showed an increase in estimated risk for certain configurations. The change in ris k from Reference Plant No. 2 is larger than that for Reference Plant No. 1. However, the sta ff performed a more simpli-fied analysis for Reference Plant No. 2 relative to Reference P lant No. 1. Furthermore, the vertical ZOI above the switchgear for the new HEAF PRA methodol ogy is always smaller than the value from NUREG/CR-6850 [3]. Additionally, the new me thodology predicts fire damage from HEAF in a region near the cabinet (just above and i n front of) not covered by NUREG/CR-6850. For plant configurations with additional targ ets in this region, the switchgear could see a significant increase in risk with the ne w PRA HEAF methodology.

Additionally, in a few cases the ZOI other than the vertical ZO I increased in the new meth-odology. Finally, longer fault clearing times lead to multiple, simultaneous switchgear HEAF fires, which may expose additional cables to fire damage.

  • The updated HEAF PRA methodology credits ERFBS for preventing damage to cables within the new ZOI of the bus ducts and switchgear, unlike NURE G/CR-6850 and its Sup-plement 1. The staff noted that the risk decrease for the switc hgear in Reference Plant No. 1 was primarily attributed to the credit given for ERFBS in the new methodology. De-pending on the plant-specific configurations, fault clearing ti mes, and risk profiles, appli-cation of the new methodology, including credit for preventing damage by ERF BS, m ay r e-sult in an estimated risk reduction.
  • The changes in risk from the application of the updated HEAF m ethod in Reference Plant No. 2, including the sensitivities, were generally larger than those for Reference Plant No.
1. Reference Plant No. 2 had rooms with larger amounts of cabli ng that were more sensi-tive to effects from HEAFs. Because HEAF risk, as in general fo r fire risk, is configuration dependent, this resulted in larger risk impacts for Reference P lant No. 2. As demonstrated by the sensitivities from Reference Plant No. 2 for the switchg ear, protecting important cabling from fire damage is important to mitigate fire risk.
  • A review of HEAF scenarios from the two reference plants provi ded additional risk-in-formed insights that could assist licensees in reducing their H EAF risks. Specifically, the team noted that a significant fraction of HEAF related risks we re associated with only a handful of HEAF scenarios while reviewing HEAF scenarios includ ed in their PRAs both plants. Since significant fractions of the HEAF related risk is distributed among a very small number of HEAF scenarios, it may be possible to use these scenarios to identify the subset of components that dominate the HEAF risks and focus mai ntenance or other re-lated resources on that subset.

APPROACH, RESULTS, AND INSIGHTS FROM OTHER SOURCES OF OPERATING EXPERIENCE

The NRC staff reviewed information from several other operating experience sources to obtain qualitative observations related to HEAF events. Each of the ev ents reviewed provided one or more observations relating to measures that a licensee may adop t to minimize the likelihood of HEAFs or to mitigate the consequences if a HEAF were to occu r. Since the staff reviewed many events, there was the potential to generate and list a lar ge number of observations.

However, a lengthy list of observations might be too unwieldy a nd inhibit the readers ability to bring focus on a handful of risk-informed insights. Therefore, the staff focused on the more risk significant issues.

5 Risk-Informed Insights and Observations from the ASP Event and the Maanshan NPP Station Blackout Event

The NRCs ASP program evaluates potentially risk significant events and degraded conditions that occur at NPPs. To assess the risk significance of events, the ASP uses conditional core damage probability (CCDP). To assess the risk significance of d egraded conditions that exist for a specific exposure time, the ASP program uses the change i n core damage probability (CDP). Events or degraded conditions for which CCDP or CDP ex ceed a set threshold are identified as precursors and saved in the ASP database. Irrespe ctive of the metric used, events documented in the ASP Program provide a basis to identify the subset of risk significant HEAF events, and consequently, to generate risk-informed insigh ts. Therefore, HEAF events or degraded conditions associated with HEAFs in the ASP databas e can be characterized as the subset of HEAF events that had the highest impact on safety. of the LIC-504 memorandum [10] provides details of nine HEAF events in the ASP database as well as the 2001 Maanshan NPP HEAF event that were risk significant enough to be characterized as accident sequence precursors [11]. The staff added the 2001 Maanshan NPP event to the mix of the ASP database events becaus e (1) the Maanshan NPP design (a power plant with two Westinghouse three loop PWRs is similar to a number of U.S.

plant designs, (2) the event constitutes the most risk signific ant HEAF event (highest estimated CCDP) and as such has the potential to be a rich source of risk -insights, and (3) an ASP-like analysis had been performed on the Maanshan NPP event.

Details on the HEAF event that occurred at Maanshan, Unit 1 in 2001 are provided in [11]

which describes several significant HEAF events that occurred b etween 1986 and 2001. In summary, a fire started as the result of a fault in the safety related 4 kV switchgear supply circuit breaker. The initial fault caused explosions, arcing, s moke, and ionized gases, which propagated to adjacent safety related 4 kV switchgear and damag ed six switchgear compart-ments. The damage resulted in the complete loss of the faulted safety bus and its emergency diesel generator (EDG) and a loss of offsite power (LOOP) to th e undamaged safety bus be-cause of faulting of its offsite electrical feeder circuit. An independent failure of the redundant EDG resulted in a loss of all alternating current (AC) power. S moke hindered access to equip-ment, delaying the investigation and repair of the failures. Th e station blackout (SBO) was terminated after about two hours when an alternate AC EDG was s tarted and connected to the undamaged safety bus. This event prompted the following ris k-informed insight:

  • HEAFs that can lead to SBOs are likely to initiate at buses or switchgear that are essential to supply AC power from both offsite power and emergency diesel s (or another emergency supply). Resources focused to minimize the likelihood of HEAF o ccurrence at those switchgear and buses (e.g., improved preventive and predictive electrical maintenance) can reduce HEAF related risks. Measures taken to minimize the p ossibility of a HEAF at one emergency bus, causing failure of the redundant electrical train due to consequential failures (e.g., due to smoke, or design deficiencies), will als o minimize the SBO related HEAF risks.

The plant impacts associated with the ten events identified in the Enclosure 2 of the HEAF LIC-504 memorandum [10] which documents nine ASP events and the Maanshan event included full or partial LOOP events, and the loss of a single 4 kV emer gency bus. These events, in conjunction with other consequential failures have the potentia l to lead to SBO events such as that at Maanshan. Therefore, plant features that could mitig ate SBOs can be used to further mitigate SBO related HEAF risks. In light of that, the LIC-504 team offered the following risk-informed insight:

  • In general, HEAFs leading to SBOs constitute the highest HEAF related risks. Plant de-sign and operational changes that have been adopted to enhance the mitigation of BDB accidents order [12] are likely to reduce HEAF related risks.

6 In addition to the risk-informed insights, based on review of t he ASP events, the LIC-504 team offered the following additional observations:

  • Of the nine events screened into the ASP database, eight event s occurred in high-or me-dium-voltage equipment. The other event occurred at a 480 V loa d center.
  • The staff investigated whether there were predominant root cau ses of the HEAFs that ap-peared in the ASP database. The root causes varied: four of the events occurred because of inadequate maintenance {two due to presence of foreign mater ial (carbon fiber, alumi-num debris), two events occurred due to other unspecified inade quate maintenance prac-tices}; and other causes included deficient design controls, wa ter intrusion, random fail-ures, and faulty protective relay coordination.
  • Low voltage (480 V or less) components cannot be screened out as negligibly risk signif-icant. Particularly, HEAFs at low voltage load centers can lead to moderately risk signifi-cant events unless the systems are designed to prevent long dur ation arcing.
  • Ingestion of dust or any other material to bus ducts creates t he potential for multiple con-current HEAFs.

To assess the risk of HEAF events in a more generic manner, the staff used a subset of the nine ASP events, and outputs of the NRCs suite of Standard Plant Analysis Risk (SPAR) models to develop a HEAF related average core damage frequency (CDF) for U.S. NPPs. The estimate is based on the frequency of risk significant ASP even ts multiplied by a suitably bounding CCDP. That estimate, however, is simply an approximati on, and is not representa-tive of HEAF related risks at any U.S. NPP since HEAF risks are highly plant specific. Further, as illustrated by the HEAF operating experience, the plant and operator response to the HEAF event can lead to other failures and conditions that are unrela ted to the initial HEAF and are difficult to capture in a risk assessment. However, this approx imation approach provides some insights regarding the relative magnitude of HEAF related risks in a general sense.

Of the nine ASP events, six occurred between 2010 and 2021. One occurred between 2000 and 2009 and two occurred before 2000. There could be a variety of possible explanations for this, including under-reporting of HEAF events before 2010 or c hanges in the ASP risk as-sessment process over time. Although the staff did not investig ate the reason for this trend, the staff is confident that risk significant HEAF events occurr ing since 2010 have been appro-priately captured in the ASP database. Therefore, to prevent in appropriate biasing of the risk significant HEAF event frequency, the staff assumed operating e xperience of the last twelve years is most representative of the current risk. That assumpti on yields 6 events over approx-imately 1200 reactor years (or ~ 5 x 10 -3 events/year).

The staff noted that the ASP HEAF events led to a variety of in itiating events, including tran-sients (reactor or turbine generator trips), LOOPs, or loss of a vital emergency AC power bus.

Based on a review of SPAR model results, the most limiting CCDP for these initiating events is associated with a loss of a vital AC bus with a CCDP value o f ~ 1 x 10-3 (representing a 95 % upper bound value for all SPAR model results). The SPAR mo del CCDP results for transients and LOOPs were all below a CCDP value of 1 x 10-3. Based on these estimates, the staff concluded that a reasonably bounding average HEAF NPP CDF value, based on ASP events, is approximately 5 x 10 -6 per reactor year. This value is generally considered to be a small risk impact, compared to the NRCs safety goals [13], but constitutes a non-negligible fraction of the risk. Furthermore, on a plant-specific basis, H EAFs may contribute to a sub-stantial fraction of the fire risk. As mentioned earlier, the H EAF related risk is highly plant specific. For instance, for Reference Plant No. 1, the HEAF rel ated CDF was about 2 x 10-6 per reactor year.

For Reference Plant No. 2, the HEAF related CDF was about 3 x 1 0-5 per reactor year. How-ever, the Fire PRA associated with Reference Plant No. 2 includ ed several more challenging fire scenarios and used a more simplified and bounding modelling approach compared to Plant No. 1.

7 Summary of Risk-Informed Insights from EPRI 3002015459

In March 2019, EPRI published a report entitled Critical Maint enance Insights on Preventing HEAFs [14]. The Executive Summary of that report noted that HE AFs can occur, and when combined with latent protective device or switchgear issues, co uld escalate, and cause signif-icant equipment damage and impact to the licensees capability to generate electrical power at the NPP. The Executive Summary also noted that (1) an analys is of industry data demon-strated that an effective preventive maintenance program is imp ortant in minimizing the likeli-hood and severity of HEAF events, (2) 64%of HEAF events were co nsidered preventable, and (3) the most prevalent cause of failure due to HEAFs was inadeq uate maintenance.

The report examined four types of electrical equipment: circuit breakers/switchgear, bus ducts, protective relays, and cables. In addition to discussing the ge neral importance of maintenance, the report provided insights on circuit breakers/switchgear. Th e staff characterizes two key findings of the EPRI report as risk-informed insights becaus e these insights are focused on a subset of components that are likely to be of relatively high -risk significance. These two risk-informed insights from the EPRI report are provided below:

  • With respect to circuit breakers, the report noted that mainte nance of the Unit Auxiliary Transformer (UAT) breaker is particularly important because its failure can lead to an ex-tended duration generator-fed fault at the first switchgear bus. Operating experience has shown this breaker to fail during automatic bus transfers. The report acknowledged the challenges that licensees confront in performing preventive mai ntenance due to con-straints associated with outage schedules and offered risk-info rmed guidance so that li-censees may focus their maintenance on the risk critical subset of maintenance activities.
  • With respect to switchgear, the report noted that for critical switchgear, such as feeder circuit breakers that carry higher currents and switchgear that is part of a bus transfer scheme, proper maintenance of connections on both the bus duct side and the circuit breaker side is especially important.

Observations from the OECD/NEA HEAF Fire Events Report

The staff reviewed the OECD/NEA report on HEAF fire events from 2013 [1] detailing 48 HEAF events, eleven of which occurred in the U.S. The definition of HEAF events used by the NRC is narrower than that used in the OECD/NEA report. For example, the OECD/NEA report in-cludes several HEAF events that took place within large transfo rmers installed outdoors, which are not included in the NRC HEAF definition. The large number o f events included in the OECD/NEA report generated several potential observations. Based on the review of the events from this report, the LIC-504 team identified the following observations relating to HEAF event prevention and mitigation:

Equipment Side

  • Proper maintenance practices: several HEAF events were attribu ted to poor, or lack of maintenance.
  • Aging management for electrical components: some HEAF events w ere caused by age related degradation of protective components, for example of bu s insulation.
  • Post-maintenance testing and inspection to ensure as-left cond itions: the root cause of some HEAF events was identified as components not being left in the correct condition post-maintenance.

8 Operations Side

  • Housekeeping to prevent dust and other foreign matter accumula tion: the root cause of many events was identified as the build-up and presence of dust, debris, and other foreign material inside bus ducts or breaker enclosures.
  • Identification and correction of existing design issues: the s everity of many of the reported events was exacerbated by long-standing design errors or proble ms.
  • Understanding of the electrical system and event conditions to prevent incorrect operator actions: the severity of some of the reported events was increa sed by operators taking incorrect actions or not understanding what the correct actions were.

APPROACH USED TO LIC-504 TEAM RECOMMENDATIONS

of the HEAF LIC-504 memorandum [10] provides recomm endations for NRCs senior managements consideration. Some of the guidance that th e LIC-504 team used to generate their recommendations are included in the LIC-504 Offi ce Instruction itself. For in-stance, Section 4.2.1 of the LIC-504 Office Instruction describ ed when the NRC should con-sider issuing prompt regulatory requirements such as Orders to shutdown units based on risk insights. LIC-504 also provides guidance on the nature of gener ic communications to licen-sees that NRC should consider based on risk significance. The L IC-504 team considers other NRC guidance documents also to generate its recommendations suc h as:

  • NRCs Management Directive (MD) 8.18 entitled Generic Communi cations Program [15]

to further inform on whether NRC should consider issuing a Bull etin, a Generic Letter, a Regulatory Issue Summary, or an Information Notice to address t he emerging issue;

  • MD 6.3 entitled Rulemaking Process [16] and the associated g uidance document NUREG/BR-0058 entitled, Regulatory Analysis Guidelines of the U.S. Nuclear Regula-tory Commission [17] to determine whether the issue warrants t he LIC-504 team to rec-ommend new rulemaking (or modifying an existing rule).

In addition to the above, the LIC-504 team leveraged the Teach ing Element of the NRCs Be RiskSMART framework [18]. In doing so, the LIC-504 team co nsidered various types of communication venues that the NRC will leverage to convey the a ctions a licensee may con-sider to mitigative risks associated with the HEAFs.

LIC 504 TEAM RECOMMENDATIONS TO NRC MANAGEMENT

The NRCs LIC-504 team considered and investigated a full range of potential options to rec-ommend under the NRCs licensing, rulemaking, and oversight res ponsibilities. As conveyed above, the team noted some increases as well as some decreases in risk when the new HEAF methodology was applied; however, team concluded that there is no significant increase in risk warranting additional regulatory requirements. In addition, the team evaluated various communication options to share its insights with licensees so t hey can implement effective steps to further reduce and/or mitigate HEAF risks. The final m anagement-endorsed recom-mendations are provided below:

  • Issue an Information Notice (IN) to share information on (1) t he operating experience and risk insights from the LIC-504 assessment, (2) regulatory frame work/license conditions, and (3) the availability of the new HEAF risk assessment method ology for licensee con-sideration.

9

  • Incorporate risk insights obtained from the LIC-504 assessment to inform NRRs ongoing PRA configuration control initiative.
  • Consider incorporating risk insights obtained from the LIC-504 assessment to inform NRRs Reactor Oversight Process.
  • Communicate risk insights gleaned from the HEAF related risks / LIC-504 process with regional inspectors and senior reactor analysts.
  • Share risk insights gained from the HEAF LIC-504 analysis with external stakeholders via public meetings (e.g., workshops), participation at owners grou p meetings, and commu-nications at national and international forums.

CONCLUSIONS

The NRC has successfully developed a process to address safety issues that emerge as a result of world-wide nuclear power plant operating experiences in an efficient and effective Commented [SW1]: Marina: I reworded this section based manner. NRR developed an Office Instruction entitled, LIC-504, Integrated Risk-Informed De-on management feedback.

cisionmaking for Emergent Issues that describes this process, which enables NRC staff to use best available information to assess risk (quantitative or qualitative), defence-in-depth, and safety margins. This process allows for the NRC to disposit ion issues in a timely manner, consistent with risk-informed decision-making principles.

ACKNOWLEDGEMENTS

The significant contributions made by the following staff membe rs of the NRCs HEAF LIC-504 team were instrumental in the authors ability to prepare t his paper: Kevin Coyne, J. S.

Hyslop, Nicholas Melly, Charles Moulton, Reinaldo Rodriguez, Ch ing Ng, Siva Lingam, and Gabriel Taylor.

REFERENCES

[1] Organisation for Economic Co-operation and Development (OEC D) Nuclear Energy Agency (NEA), Committee on the Safety of Nuclear Installations (CSNI): OECD FIRE Project - Topical Report No. 1, Analysis of High Energy Arcing Fault (HEAF) Fire Events, NEA/CSNI/R(2013)6, Paris, France, June 2013, http://www.oecd-nea.org/documents/2013/sin/csni-r2013-6.pdf.

[2] United States Nuclear Regulatory Commission (U.S. NRC): US Code of Federal Regu-lations, Part 10 CFR 50.48(c) National Fire Protection Associa tion Standard NFPA805, Federal Register 65 FR 38190, June 20, 2000; 69 FR 33 550, June 16, 2004; 72 FR 49495, August 28, 2007.

[3] United States Nuclear Regulatory Commission (U.S. NRC) Offi ce of Nuclear Regulatory Research (RES) and Electric Power Research Institute (EPRI): Fi re PRA Methodology for Nuclear Power Facilities, Final Report, Volume 2: Detailed Methodology, Appendix M, EPRI/NRC-RES, NUREG/CR-6850 (EPRI 10191989), Washington, DC, and Palo Alto, CA, USA, 2010, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/ cr6850/index.html.

[4] United States Nuclear Regulatory Commission (U.S. NRC) Offi ce of Nuclear Regulatory Regulation (NRR): Integrated Risk-Informed Decisionmaking for E mergent Issues, U.S.

10 NRCs NRR Office Instruction LIC-504, Revision 5, Washington, D C, USA, March 9, 2020, https://www.nrc.gov/docs/ML19253D401.pdf.

[5] United States Nuclear Regulatory Commission (U.S. NRC): Hig h Energy Arcing Fault Frequency and Consequence Modelling, Draft Report for Comment, NUREG-2262, Washington, DC, USA, July 2022, https://www.nrc.gov/docs/ML2215/ML22158A071.pdf.

[6] United States Nuclear Regulatory Commission (U.S. NRC): Rea ctor Coolant System Pressure Boundary Leakage Due to Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism Nozzles and Reactor Pressure Vessel Head De gradation, LER 50-346/02-002-00, Washington; DC, USA; April 29, 2002, https://www.nrc.gov/docs/ML2011/ML20112F488.pdf.

[7] Federal Register (FR): Report to Congress on Abnormal Occur rences, 68 FR 19233, Fiscal Year 2002, Dissemination of Information, April 18, 2003, https://casetext.com/federal-register/report-to-congress-on-abn ormal-occurrences-fis-cal-year-2002-dissemination-of-information.[8]U.S. General Acco unting Office (GAO):

Report to Congressional Requesters, Nuclear Regulation - NRC N eeds to More Ag-gressively and Comprehensively Resolve Issues Related to the Da vis-Besse Nuclear Power Plants Shutdown, GAO-04-415, May 2004, http://www.gao.gov/cgi-bin/getrpt?GAO-04-415.

[9] United States Nuclear Regulatory Commission (U.S. NRC) Offi ce of Nuclear Regulatory Regulation (NRR): Issuance of Safety Orders Issues, U.S. NRCs NRR Office Instruction LIC-106, Revision 1 (NRC ADAMS Accession No. ML19283B565), Wash ington, DC, USA, April 23, 2020.

[10] United States Nuclear Regulatory Commission (U.S. NRC): NR C High Energy Arcing Faults LIC-504 Team Recommendat ions, (Memorandum: ADAMS Accession No.

ML22201A000, Enclosure 1: ADAMS Accession No. ML22201A001, Encl osure 2:

ADAMS Accession No ML22201A002, Enclosure 3: ADAMS Accession No.

ML22201A003), Washington, DC, USA, July 27, 2022, https://www.nrc.gov/docs/ML2220/ML22200A272.html.

[11] Raughley, W: S., and G. F. Lanik: Operating Experience Ass essment, Energetic Faults in 4.16 kV to 13.8 kV Switchgear and Bus Ducts That Caused Fire s in Nuclear Power Plants, 1986-2001, United States Nuclear Regulatory Commission (U.S. NRC), Wash-ington, DC, USA, February 2002, https://www.nrc.gov/docs/ML0212 /ML021290358.pdf.

[12] United States Nuclear Regulatory Commission (U.S. NRC): Is suance of Order to Modify Licenses with Regard to Modify Licenses with Regards to Beyond-Design-Basis Exter-nal Events, EA-12-049, Washington, DC, USA, March 12, 2012, https://www.nrc.gov/docs/ML1205/ML12054A735.pdf.

[13] Federal Register (FR): Safety Goals for the Operation of N uclear Power Plants; Policy Statement, 51 FR 30028, Republication, August 21, 1986, https://www.nrc.gov/reading-rm/doc-collections/commission/polic y/51fr30028.pdf.

[14] Electric Power Research Institute (EPRI): Critical Mainten ance Insights on Preventing High-Energy Arcing Faults, EPRI 3002015459, Palo Alto, CA, USA, March 2019.

[15] United States Nuclear Regulatory Commission (U.S. NRC): Ge neric Communication Program, U.S. Nuclear Regulatory Commission Management Directiv e MD 8.18, Vol-ume 8: Licensee Oversight Programs, Washington, DC, USA, Decemb er 9, 2015, https://www.nrc.gov/docs/ML1532/ML15327A372.pdf.

[16] United States Nuclear Regulatory Commission (U.S. NRC): Th e Rule Making Process, U.S. Nuclear Regulatory Commission Management Directive MD 6.3, Volume 6: Internal

11 Management, Washington, DC, USA, July 3, 2019, https://www.nrc.gov/docs/ML1921/ML19211D136.pdf.

[17] United States Nuclear Regulatory Commission (U.S. NRC): Re gulatory Analysis Guide-lines of the U.S. Nuclear Regulatory Commission, Draft Report f or Comment, NUREG/BR-0058, Revision 5 (NRC ADAMS Accession No. ML171001480),

Washington, DC. USA, April 2017, https://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures /br0058/index.html.

[18] United States Nuclear Regulatory Commission (U.S. NRC): Be riskSMART: Guidance for Integrating Risk Insights into NRC Decisions, NUREG/KM-0016, Washington, DC, USA, March 2021, https://www.nrc.gov/reading-rm/doc-collections/nuregs/knowledge /km0016/index.html.

12

Retrieved from "https://kanterella.com/w/index.php?title=ML22299A095&oldid=3718558"