Click for next page ( 318


The National Academies | 500 Fifth St. N.W. | Washington, D.C. 20001
Copyright © National Academy of Sciences. All rights reserved.
Terms of Use and Privacy Statement



Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.

Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.

OCR for page 317
APPENDIX H NUCLEAR PLANT EMERGENCY PROCEDURES AND GUIDELINES This appendix describes emergency procedures and guidance used at U.S. nuclear plants and how they are being revised in response to the Fukushima Daiichi accident. H.1 EMERGENCY OPERATING PROCEDURES Emergency Operating Procedures (EOPs) are “plant procedures that direct operators' actions necessary to mitigate the consequences of transients and accidents that have caused plant parameters to exceed reactor protection system set points or engineered safety feature set points, or other established limits” (USNRC, 1982, p. 3). For example, station blackout (i.e., loss of all AC power) situations or loss of ultimate heat sink can be handled within EOPs as long as reactor pressure and water level can be monitored and remain within acceptable ranges. An example of the successful use of EOPs in Japan is the response at the Fukushima Daini plant (see Sidebar 4.2 in Chapter 4). EOPs have always been a part of operational practice in the United States and are based around transient events or accidents that the plant was designed to handle, in some cases with operator actions, i.e., the design basis events—although a larger range of events was originally considered in the EOPs for boiling water reactors than for pressurized water reactors. EOPs have long been part of the USNRC’s safety requirements. These requirements are provided in 10 CFR Part 50 and in the technical specifications for each plant. Numerous technical reports (e.g., USNRC, 1980a,b, 1982, 1983) also help guide the development of EOPs. Training and both written and simulator exams for licensing reactor operators and senior reactor operators include EOPs. The shift supervisor, who is stationed in the control room, and the plant manager have command-and control responsibilities for implementing EOPs. (Both individuals possess senior reactor operator licenses.) H.2 SEVERE ACCIDENT MANAGEMENT GUIDELINES Severe Accident Management Guidelines (SAMG) is intended to address “beyond- design-basis” situations in which the core has or is becoming damaged. The goals of the SAMG are to stabilize a degraded core, maintain containment, and minimize the release of the core’s fission products. SAMG is much less specific than the EOPs because they cover a wide range of possibilities of the reactor damage state after significant fuel damage occurs. The phenomenology of severe accidents in light-water reactors is too complex and highly dependent upon the timing of mitigation actions to be fully predictable in advance. An extensive discussion of the SAMG can be found in Chapter 6 of Sehgal (2012). Prepublication Copy H-1

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines Events involving the loss of core cooling are considered to be beyond the nuclear plant’s design basis and are covered by SAMG. The requirements in 10 CFR 50.631 (Loss of all Alternating Current Power) address the conditions that can lead to the loss of core cooling. Licensees are required to provide an additional source of electrical power or otherwise demonstrate that the plant could cope with the loss of all AC power through other means for removing decay heat from the reactor for a specified period of time. In events involving the loss of all AC power, operators would follow the procedures required under 10 CFR 50.63(c)(ii)-(iii): “(ii) A description of the procedures that will be implemented for station blackout events for the duration determined in paragraph (c)(1)(i) of this section and for recovery therefrom; and” “(iii) A list of modifications to equipment and associated procedures, if any, necessary to meet the requirements of paragraph (a) of this section, for the specified station blackout duration determined in paragraph (c)(1)(i) of this section, and a proposed schedule for implementing the stated modifications.” The procedures developed to address (ii) above would address the maintenance of cooling functions using an alternate AC power source or coping strategies. The procedures would also address the restoration of onsite and offsite AC power sources. A key difference between EOPs and SAMG is that the former are subject to regulatory oversight (see NUREG-08992) whereas SAMG is a voluntary industry program. Another important difference is that SAMG anticipates that the engineering staff in the Technical support center will be available to guide reactor operators in applying the guidance and evaluating trade- offs that inevitably occur in severe accident management, whereas EOPs enable control room staff to engage in immediate symptom-based responses. Transition points between EOPs and SAMGs are defined, but some element of judgment is required to determine whether the transition criteria have been met. Consequently, operator training and education play an important role in making timely decisions. SAMG makes use of both standard and non-standard plant systems. It includes approaches to evaluate plant conditions, select the appropriate guidance, and evaluate the effectiveness of the selected guidance during a severe event. It also includes training plans for staff expected to be involved in any of the following three activities: (1) evaluation of plant damage, (2) making decisions on which strategies to implement, or (3) implementing the selected strategies. NEI 91-04 recommends that plants self-evaluate their strategies through use of periodic mini-drills that ensure that personnel who would be involved in the emergency response are familiar with the implementation of SAMG. However, since SAMG is considered an industry initiative, the USNRC has no specific regulatory control. Instead, USNRC has accepted the industry’s commitment to assess its capabilities and implement appropriate improvements within the constraints of existing personnel and hardware (Taylor 1996). In other words, the range of severe accidents scenarios that could be managed with the training and steps outlined in the 1 Available at http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0063.html. 2 Guidelines for the Preparation of Emergency Operating Procedures (1982). Available at http://pbadupws.nrc.gov/docs/ML1025/ML102560007.pdf Prepublication Copy H-2

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines SAMG is limited to those situations that do not require additional resources in staffing or equipment. Within the last decade, new requirements going beyond this limited approach have been created to respond to potential terrorist attacks. The events at the Fukushima Daiichi plant have further emphasized the need for a more comprehensive approach to severe accident management. Indeed, industry in in the process of developing and implementing new SAMG and associated physical resources. H.3 EXTENSIVE DAMAGE MITIGATION GUIDELINES Following the terrorist attacks of Sept 11, 2001, there was significant concern in the United States about attacks on nuclear power plants using hijacked airplanes or other means (e.g., NAS 2004a). The USNRC and national laboratories analyzed terrorist attack scenarios on nuclear plants and their spent fuel pools and concluded that additional security and mitigation measures were needed.3 The USNRC issued an Interim Compensatory Measure (ICM) Order in 2002 modifying the operating licenses of all plants. Section B.5.b of that order directed plant licensees to take certain actions: “Section B.5.b of the ICM Order requires licensees to adopt mitigation strategies using readily available resources to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities to cope with the loss of large areas of the facility due to large fires and explosions from any cause, including beyond- design-basis aircraft impacts.” The utilities, working through the Nuclear Energy Institute, developed detailed guidance (NEI 06-12 Rev 24) for B.5.b response procedures, termed extensive damage mitigation guidelines (EDMG), and additional equipment to be located at each site. The guidance assumed conditions far beyond design-basis accidents including loss of all AC and DC power, denial of access to structures including the control room, and loss of plant control and monitoring capability. EDMG play a different role than the emergency operating procedures (EOPs). EDMG are intended to provide operators with a “toolbox” of capabilities that can be used to respond to unpredictable damage from large fires and explosions. EDMG also serve as a bridge between the plant operational command and control and the command and control that is provided by the plant’s emergency response organization. Little was publically known about these B.5.b activities because they were initially protected as Safeguards Information. Additional details about the program became public knowledge after the March 2009 rulemaking that codified the B.5.b requirements contained in 3 See The Evolution of Mitigating Measures For Large Fire and Explosions A Chronological History from September 11, 2001 Through October 7, 2009. Available at http://pbadupws.nrc.gov/docs/ML0929/ML092990417.pdf (summary) and http://pbadupws.nrc.gov/docs/ML0929/ML092990417.pdf (detailed chronology). 4 B.5.b Phase 2 & 3 Submittal Guideline, Revision 2 (2006). Available at http://pbadupws.nrc.gov/docs/ML0700/ML070090060.pdf Prepublication Copy H-3

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines the order into regulations (10 CFR 50.54(hh)(2)5) and the post-Fukushima acknowledgment (USNRC Bulletin 2011-016) of the potential importance of B.5.b capabilities for responding to beyond-design-basis events. Because the B.5.b order was determined to be Safeguards Information, the nuclear utilities in Japan were unaware of some of its content although the Nuclear Safety Commission of Japan apparently was notified of its requirements. Even after the B.5.b. requirements became public knowledge, however, Japanese authorities did not recognize the change of policy and therefore did not initiate any consultations on the requirements with Japanese nuclear utilities. Many of the B.5.b capabilities and accident mitigation measures were needed or used at the Fukushima Daiichi and Daini plants following the March 11, 201,1 earthquake and tsunami. The pre-positioned equipment resources for B.5.b include portable generators, fire truck or other portable water pumps, batteries, cables, tools, fuel and firefighting equipment, all of which were part of the these plant’s responses. The mitigation strategies that the EDMG are intended to cover are listed in Table H.1. At least four of these boiling water strategies were utilized at the Fukushima Daiichi plant, supporting the claim by the USNRC (2013d, p.21) that: “…the mitigating strategies implemented at U.S. nuclear plants following the terrorist attacks of September 11, 2001, to cope with large fires and explosions may have helped in responding to an extended loss of electrical power and core cooling capability that occurred at Fukushima if the equipment was stored in an area of the plant that was not inundated by the tsunami.” TABLE H.1 EDMG Mitigation Strategies BWR mitigation strategies Manual Operation of RCIC or Isolation Condenser DC Power Supplies to Allow Depressurization of RPV & Injection with portable pump Utilize Feedwater and Condensate Makeup to Hotwell Makeup to CST Maximize CRD Procedure to Isolate RWCU Manually Open Containment Vent Lines Inject water into Drywell Portable Sprays 5 10 CFR 50.54. Conditions of Licenses. Available at http://www.nrc.gov/reading-rm/doc- collections/cfr/part050/part050-0054.html. 6 Mitigating Strategies (2011). Available at http://pbadupws.nrc.gov/docs/ML1112/ML111250360.pdf Prepublication Copy H-4

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines PWR mitigation strategies Makeup to RWST Manually Depressurize SGs to Reduce Inventory Loss Manual operation of Turbine- (or Diesel-) Driven AFW Pump Manually Depressurize SGs and Use Portable Pump Makeup to CST Containment Flooding with Portable Pump Portable Sprays Note: AFW = auxiliary feed water BWR = boiling water reactor; CRD = control rod drive; CST = condensate storage tank; PWR = pressurized water reactor; RCIC = reactor core isolation cooling system; RPV = reactor pressure vessel; RWCU = reactor water cleanup; RWST = reactor water storage tank; SG = steam generator SOURCE: NEI (2012) H.4 POST-FUKUSHIMA CHANGES By definition, severe accidents are considered to result in plant conditions that are beyond design basis and outside of the traditional regulatory scope. Nevertheless, the USNRC does have the ability to inspect individual plants to verify that licensees have implemented SAMG. The USNRC used this authority following the Fukushima Daiichi accident to collect information on the implementation, training, and maintenance of SAMG. The USNRC Near-Term Task Force (USNRC NTTF, 2011, p. 64) noted that, while some plants have maintained this important safety program, others have treated the volunteer initiative in a “…significantly less rigorous and formal manner, so much so that the SAMG inspection would have resulted in multiple violations had it been associated with a required program.” The USNRC is currently proposing new rules which would place SAMG under its oversight authority (USNRC, 2012). The industry has also taken a series of actions following the Fukushima Daiichi accident (see Appendix F). The 2012 EPRI/NEI/INPO report “The Way Forward’ (NEI/EPRI/INPO, 2012) outlines a set of goals and actions that the industry has committed to undertake to improve nuclear safety and apply lessons learned from the Fukushima Daiichi accident. These efforts are voluntary, remaining subject to inspection but outside of regulatory requirements. The industry is currently actively engaged with the USNRC in discussing how the industry response will fit in with the proposed changes in the regulatory framework mentioned above. Prepublication Copy H-5

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines H.4.1 Diverse and Flexible Coping Strategies (FLEX) An important component of the industry’s response is the FLEX program, a set of prepositioned capabilities designed to extend the coping period in the event of an extended AC power loss and other adverse situations such as occurred at the Fukushima Daiichi plant. These capabilities are intended to be used in conjunction with revised SAMG. The USNRC reviewed FLEX and ordered7 that each U.S. nuclear plant develop a site-specific plan to mitigate severe accidents of the type experienced at Fukushima Daiichi using FLEX-type capabilities. The order requires a phased approach with the following elements (the following text is taken directly from Attachment 2 of the Order, p.4):  The initial phase requires the use of installed equipment and resources to maintain or restore core cooling, containment and SFP [spent fuel pool] cooling capabilities.  The transition phase requires providing sufficient, portable, onsite equipment and consumables to maintain or restore these functions until they can be accomplished with resources brought from off site.  The final phase requires obtaining sufficient offsite resources to sustain those functions indefinitely. The FLEX implementation guide8 contains these elements and was endorsed in USNRC Interim Staff Guidance9 as being an acceptable means of complying with the Mitigation strategies order. The only caveat was that for the initial phase of the response, a determination of appropriate response time had to be made and used in the selection of storage location and readiness of equipment. The USNRC will review each plant’s FLEX installation and guidance as they are being completed (which will be no later than the end of 2016) and will issue a Safety Evaluation Report. H.4.2 Revision of SAMG The Electric Power Research Institute commissioned a revision to the Severe Accident Management Guidance Technical Basis Report (EPRI, 2012c). The revised report was published in October 2012.10. This report is the first update of the original 1991 version, adding additional Candidate High Level Actions in Volume 1 and providing supporting technical information in Volume 2. New material addresses using sea water injection for reactor core cooling, common cause failures due to external events, cooling spent fuel pools, setting priorities in multi-unit 7 EA-12-049, Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (March 2012). Available at http://pbadupws.nrc.gov/docs/ML1205/ML12054A735.pdf. 8 NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. B1 (May 2012) Available at http://pbadupws.nrc.gov/docs/ML1214/ML12143A232.pdf. 9 Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events. JLD-ISG-2012-01. Available at http://pbadupws.nrc.gov/docs/ML1222/ML12229A174.pdf 10 Available at http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=000000000001025295 Prepublication Copy H-6

OCR for page 317
Appendix H: Nuclear Plant Emergency Procedures and Guidelines events, containment isolation failure and hydrogen combustion within plant buildings. The intent, as with the original report is to guide owners groups in developing new SAMG. Efforts are currently underway to develop revised versions of the SAMG for both generic and plant-specific guidance. The Boiling Water Reactor Owners Group, Emergency Procedures Group has been meeting quarterly since the Fukushima Daiichi accident and completed Revision 3 of the generic guidelines in 2013. These are integrated guidance for emergency procedures and severe accident, referred to as Emergency Procedure Guidelince/Severe Accident Guidelince. According to the Nuclear Energy Institute (18 Nov 2013), this revision utilizes both FLEX and EDMG capabilities and guidance to provide core and spent fuel pool cooling and maintain containment functions. Individual plants are developing EOPs and SAMGs based on this generic guidance but tailored to their specific situations. The generic guidance is in the process of being implemented for each plant, and industry workshops are being held in the United States, Europe, Mexico, Japan and Taiwan to assist with this process. The USNRC has formally requested that the guidelines be submitted so that they can be reviewed by staff in 2014 to support ongoing rulemaking activities. H.4.3 Response in Japan TEPCO (2012b, p. 471) has proposed a program of countermeasures similar to FLEX. The strategies are to: “…consider capabilities for accident control assuming situations where almost all station facilities used to control the accident lose their functions. This is in addition to the basic approach of assuming a certain scale of an external event, including tsunamis which caused the Fukushima accident, and taking complete countermeasures against it to prevent accidents from occurring.” Examples of the type of equipment and guidance documents are provided in the Kawano (2012). The descriptions of the equipment and capabilities are plant-specific and designed to address the situations encountered at the Fukushima Daichi plant following the March 11, 2011, earthquake and tsunami. Prepublication Copy H-7

OCR for page 317