Safety Evaluation of Reactor Plants (PWR, BWR) Masayuki Nakatsuji Engineering & Safety Group Plant Management Department The Japan Atomic Power Company August 26th, 2015 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited Contents 1. Overview of Nuclear Power Safety Regulation 2. Concept of Safety Evaluation during Safety Reviews 3. Evaluation of Exposure Dose Equivalents during Normal Operation 4. Safety Design Evaluation The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1. Overview of Nuclear Power Safety Regulations 1 (1) Unification of the Administration Related to Nuclear Safety Regulatory Affairs Divorcing Nuclear Regulation from Nuclear Promotion In the old organizational structure, the Nuclear and Industrial Safety Agency, an organization in charge of nuclear safety regulation, was placed under the Ministry of Economy, Trade and Industry, which was supposed to promote nuclear energy use. In order to settle the problem where nuclear regulation affairs and nuclear promotional affairs were under the administration of the same organization, the nuclear safety regulation section was separated from the Ministry of Economy, Trade and Industry, and the Nuclear Regulation Authority was newly established as an external Bureau of the Ministry of the Environment. [Old Regulatory System] Cabinet office Atomic Energy Commission Overall coordination of measures for the physical protection of nuclear material METI MEXT (Ministry of Economy, Trade and Industry) (Ministry of Education, Culture, Sports, Science and Technology) Agency for Natural Resources and Energy Nuclear Safety Commission Nuclear and Industrial Safety Agency Double checking of nuclear reactor safety reviews Safety regulations on nuclear power reactors • Safety regulations on test and research nuclear reactors • Safeguards* • Utilization of SPEEDI (System for Prediction of Environmental Emergency Dose Information) for radiation monitoring • Regulations on use of radioisotopes* [New Regulatory System] Nuclear Regulation Authority Chairman and four commissioners (appointed with the approval of the Diet) METI Nuclear Regulation Agency (secretariat) *Transferred on April 1, 2013 Double checking to enhance regulations Electric utilities, etc. Regulation Research institutes, universities, etc. The Japan Atomic Power Company MOE (Ministry of the Environment) (Ministry of Economy, Trade and Industry) Agency for Natural Resources and Energy Regulation Electric utilities, research institutes, universities, etc. Source: NRA brochure Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1. 原子力安全規制の概要 1 （１） 原子力規制の一元化 「規制｣と「利用」の分離 これまでは、原子力「利用」の推進を担う経済産業省の下に、原子力の安全「規制」を担う原子力安全・保安院が設置 されていた。そうした「利用の推進」と「安全規制」を同じ組織の下で行うことによる問題を解消するため、経済産業省 から、安全規制部門を分離し、環境省の外局組織として原子力規制委員会を新設した。 【 これまでの規制体制 】 内閣府 原子力委員会 核物質等を守るため の対策の総合調整 原子力安全 委員会 原子炉の安全 審査のダブル チェック等 経済 産業省 資源 エネルギー庁 原子力 安全・保安院 発電用原子 炉の安全規 制等 ダブルチェックで規制 電力会社等 文部 科学省 • 試験研究炉等の安 全規制 • 保障措置* • 放射線のモニタリン グ・SPEEDI（緊急時 迅速放射能影響予測 ネットワークシステム） の運用 • 放射性同位元素の 使用等の規制* 【 新しい規制体制 】 原子力 規制委員会 環境省 委員長＋委員４名 （国会同意人事） 原子力規制庁 （事務局） ＊H25.4.1より移管 経済 産業省 資源 エネルギー庁 規制 研究機関 大学等 The Japan Atomic Power Company 電力会社・研究機関・ 大学等 出典：NRAパンフレット Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1. Overview of Nuclear Power Safety Regulations 2 (2) Nuclear Power Safety Regulation Flow The Japan Atomic Power Company Decommissioning Periodic inspection Examination on physical protection of nuclear material Inspection on safety management Source: Convention on Nuclear Safety National Report for the Sixth Review Meeting Operation Phase Certificate issuance Completion of construction Inspection before commercial operation Approval of compliance with regulations on physical protection of nuclear material Approval of compliance with safety regulations Permission of reactor installation Design Phase Power Reactor Licensee Nuclear Regulation Authority Construction Phase Approval of the construction plan According to the Nuclear Regulation Law, permission from the Nuclear Regulation Authority is necessary for the installation of a nuclear reactor. The nuclear power reactor licensee is required to get reactor installation approval, approval of the reactor construction plan, and inspection of the nuclear fuel assemblies at the design and construction stages. At the operation stage, the nuclear power reactor licensee is required to conduct periodic inspections and get examinations for compliance with the regulations on physical protection of nuclear material. Examinations for compliance with safety regulations are conducted by Inspectors for the Safety Management of Nuclear Installations. The nuclear power reactor licensee is required to evaluate safety management activities for the reactor facilities and to verify that new findings are incorporated into the safety management activities. In order to extend the period of operation, the results of the component deterioration evaluation and the maintenance management policy must be approved by the Nuclear Regulation Authority every ten years. Detailed design Basic design Periodic evaluation Evaluation of aging degradation management technology Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1.原子力安全規制の概要 2 (2) Nuclear Power Safety Regulation Flow 出典:原子力の安全に関する条約 日本国第6回国別報告 The Japan Atomic Power Company Decommissioning 定期検査 核物質防護検査 保安検査 Operation Phase 合格証交付 竣工 使用前検査等 核物質防護規程の認可 保安規定の認可 Construction Phase 工事計画の認可 原 子 炉 設 置 者 Design Phase 原 子 力 規 制 委 員 会 詳細設計 原子炉設置許可 原子炉を設置するには、原子炉等規制法に基づき、原子力規制委員会の許可を受けなければならない。 原子炉設置者は、設計・建設段階では、原子炉設置許可、工事計画認可、燃料体検査等が必要である。 運転段階では、原子炉設置者は、定期検査、核物質防護規程の遵守状況検査が行われる。また、保安規定の遵守状況検査 には、原子力保安検査官が実施している。 原子炉設置者は、原子炉施設の保安活動の評価、保安活動への最新知見の反映評価を行う。 運転期間延長する場合、10年ごとに機器の劣化評価及び保守管理方針に対する原子力規制委員会による確認が必要である。 基本設計 定期的な評価 高経年化技術評価 Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1. Overview of Nuclear Power Safety Regulations 3 (3) Overall Image of Safety Reviews To keep adequate distance between a given reactor and the public. To reduce radioactive releases into the environment during normal operation. Measures to prevent the occurrence of abnormal events. Safety Security measures To confine radioactive materials based on defense-indepth concept Measures to mitigate the development of abnormal events into the accidents. To confirm suitability of measures. (safety assessment) Measures to prevent abnormal release of radioactive materials to the environment. Measures to prevent Severe Accidents The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 1. 原子力安全規制の概要 3 (3) 安全審査の全体像 原子炉と公衆との適切な距離の 確保 通常運転時における公衆の被ば く低減対策 異常事象の発生を未 然に防止するための 対策 安全確保 対 策 多重防護の考え方による放射性 物質の閉じ込め 異常事象の拡大防止 および事故への発展 を防止するための対 策 対策の適正 の確認 （安全評価) 外部への放射性物質 の異常な放出を防止 するための対策 重大事故等を防止す るための対策 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 2. Concept of Safety Evaluation during Safety Reviews 4 (1) Concept of Safety Evaluation during Safety Reviews Evaluation Category Safety Design Evaluation Evaluation of dose equivalents during normal operation Unexpecte d transients during operation Accidents Purpose of Evaluation Confirmation that the radiation dose to which the local populace is exposed during normal operation is maintained at a sufficiently low level Confirmation the functions required to assure safety are realized by elements such as the structures and devices in nuclear power plants not only during normal operation, but also in abnormal statuses Phenomena to be evaluated Judgment Criteria Amount of radioactive substances released into the environment from exhaust stacks and waste water ports Confirmation that the dose equivalent to which the local populace in the vicinity of the power plant is exposed in below the target value (referred to as dose target value, indicating an annual effective dose equivalent of 0.05 mSv) Equipment failures and malfunctions or operational errors by operators and phenomena that may lead to an abnormal status caused by disturbances of external origin foreseen as occurring at a similar frequency during the life of the nuclear power plant. Confirmation that there is no damage to the reactor core and that normal operation can be restored once the phenomena have been eliminated Phenomena that lead to an abnormal status beyond unexpected transient changes occurring during operation and that, although rare, present the risk of release of radioactive substances from the nuclear power plant and that need to be assumed from the standpoint of the safety of the nuclear power plant Confirmation that there is no possibility of a reactor core meltdown or other serious damage, that there is no secondary damage that may cause other abnormal statuses during the process of the phenomena and, additionally, that the design of barriers against the diffusion of radioactive substances is valid Source: Safety Evaluation from the Safety Review Perspective, “Unfailing safety” through “appropriate review” (Agency of Natural Resources and Energy, Ministry of International Trade and Industry) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 2. 安全審査における安全評価の考え方 4 (１) 安全審査における安全評価の考え方 評価の分類 評価の目的 通常運転時の線 量当量評価 通常運転時における周辺公 衆の放射線被ばくが十分低く 保たれることを確認する 運転時の異 常な過渡変 化 安 全 設 計 評 価 原子力発電所の構築物や機 器は通常運転時のみならず， 異常状態においても安全確 保の観点から所定の機能を 果たすことを確認する 事故 評価すべき事象 判断基準 排気筒及び放水口から環境に放出 される放射性物質の量 発電所周辺における公衆の受ける線 量当量が目標値（線量目標値といい， 実効線量当量で年間0.05ミリシーベ ルト）以下であることを確認する 原子炉の運転中において，原子力 発電所の寿命期間中に予想される 機器の故障・誤動作又は運転員の 誤操作，及びこれらと類似の頻度で 発生すると予想される外乱によって 生ずる異常な状態に至る事象 炉心の損傷はなく，かつ，通常運転 に復帰することができる状態で事象 が収束されることを確認する 運転時の異常な過渡変化を超える 異常な状態であって，発生する頻 度はまれであるが，発生した場合 は原子力発電所からの放射性物質 の放出の可能性があり，原子力発 電所の安全性を評価する観点から 想定する必要のある事象 炉心の溶融あるいは著しい損傷のお それがなく，かつ，事象の過程におい て他の異常状態の原因となるような２ 次的損傷が生じなく，さらに放射性物 質の放散に対する障壁の設計が妥 当であることを確認する 出典；安全審査から見た安全評価 「適切な審査」で「確かな安全」を（通商産業省資源エネルギー庁） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 2. Concept of Safety Evaluation during Safety Reviews 5 (1) Concept of Safety Evaluation during Safety Reviews Evaluation Category Site Evaluation Serious accidents Hypothetical accidents Purpose of Evaluation Confirmation that reactor core site conditions are appropriate Phenomena to be evaluated Judgment Criteria Confirmation that setting in a nonpopulated area is appropriate Serious accidents foreseeable in worst-case scenarios as seen from the technical perspective taking into account elements such as phenomena in the vicinity of the site, reactor characteristics and safety and protective facilities Confirmation that the local populace will not be exposed to damage caused by radiation Confirmation that areas with small populace can be ensured and that the power plant is appropriately isolated from areas where the population is concentrated Hypothetical phenomena that go beyond serious accidents the occurrence of which is inconceivable from the technical standpoint (1) Confirmation that the local populace will not be exposed to marked damage caused by radiation (2) Confirmation that the calculated total-body dose value low enough to be acceptable from the collective dose perspective Source: Safety Evaluation from the Safety Review Perspective, “Unfailing safety” through “appropriate review” (Agency of Natural Resources and Energy, Ministry of International Trade and Industry) New Regulatory Standards Evaluation Category Countermeasures against severe accidents Purpose of evaluation Phenomena to be evaluated 【Enforced or newly determined】 Before designing a nuclear reactor, external phenomena, which may possibly occur on site, must be fully examined, and their effect on the reactor must be adequately considered. - It must be verified that the effect of radiation on the populace and environment will be permissibly low if a severe accident occurs and a substantial amount of radioactive substances are released. Events of core damage, containment vessel damage, or the like, which are postulated to occur in case of a severe accident Seismic Events Tsunami Postulated Natural Phenomena Accidental human errors postulated in monitoring areas The Japan Atomic Power Company Judgment Criteria - The reactor must be installed on a foundation which can adequately support it when struck by an earthquake. - The safety functions must not be seriously damaged by a tsunami. - The safety functions must not be damaged by natural phenomena or human errors. - The reactor core must not be seriously damaged, and sufficient core cooling capability must be maintained. - The containment vessel must not be seriously damaged, and an abnormally large amount of radioactive substances must not be released from the site. Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited ２．安全審査における安全評価の考え方 5 (１) 安全審査における安全評価の考え方 評価の分類 重大事故 立 地 評 価 評価の目的 原子炉の立 地条件が適 切であること を確認する 仮想事故 評価すべき事象 判断基準 非居住区域の設 定が適切であるこ とを確認する 敷地周辺の事象，原子炉の特性，安全防 護施設等を考慮し，技術的見地からみて， 最悪の場合には起こるかもしれないと考え られる重大な事故 周辺の公衆に放射線障害を与えないことを 確認する 低人口地帯の確 保及び人口密集 地帯からの離隔が 適切であることを 確認する 重大事故を超えるような技術的見地からは 起こるとは考えられない仮想的な事故 (1)周辺の公衆に著しい放射線災害を与えな いことを確認する (2)全身線量の積算値が集団線量の見地か ら十分受け入れられる程度に小さい値である ことを確認する 出典；安全審査から見た安全評価 「適切な審査」で「確かな安全」を（通商産業省資源エネルギー庁） 新規制基準 評価の分類 評価の目的 評価すべき事象 判断基準 【強化又は新設】 立地地点において発生しうる外 的事象を十分に調査し、原子炉 施設に及ぼす影響を考慮して設 計すること ・地震事象 ・津波事象 ・想定される自然現象 ・周辺監視区域において想定される偶 発的な人為事象 ・地震力に対して十分に施設を支持できる地 盤に設置し、十分耐えれること ・津波に対して、安全機能に大きな影響を及 ぼさないこと ・自然現象及び人為事象に対して安全性を 損なわないこと 重大な事故が発生した場合に放射 性物質の異常な水準の放出に伴う 公衆や環境への放射線影響のリス クが容認可能なほど低いことを確 認する 重大事故が発生した場合に想定される炉 心損傷事象、格納容器破損事象 等 ・炉心の著しい損傷が発生しなく、かつ、炉心 を十分に冷却できること ・格納容器の破損及び放射性物質が異常な 水準で敷地外へ放出されないこと 重大事故対策 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 6 (2) Scheme of new regulation on Severe Accident Huge diffusion Prevention of human suffering Recovery of environment ⑤ Huge venting Containment vessel damage ④ ④ -2 Release-inhibiting and diffusion relaxation Prevention of largescale release containment vessel damage prevention ④ -１ Current design-based supposition Error and failure occur Normal operation Countermeasures for release-inhibiting and diffusion relaxation water cannon truck, etc. ③-2 ③-１ Escalation ② prevention to the accidents ① Mobile RHR etc. Countermeasures for containment vessel damage prevention (SA countermeasureⅡ) Specific Safety facility alternative containment vessel spray, etc. Prevention of severe core damage design base accident＋safety function preservation (single-failure) Emergency Preparedness LRF Severe core damage Loss of safety function of equipment based on design basis (multiple failure) Measures for safety Anomaly and failure occurrence prevention power–supply car, etc. Countermeasures for severe core damage prevention (SA countermeasureⅠ) Specific Safety facility power interchange between units ECCS (Emergency Core Cooling System, etc. Voluntary actions by Reactor owner Purpose of Defense-in-depth Plant condition anomaly detection and halt operation, etc. Interlock, etc. tolerance improvement Design basis Current regulation scope internal event Strength of outer event external event (natural phenomena, aircraft attack, terrorism, etc.) Source：Basic idea on SA countermeasure regulation in the LWR facility issued by NISA (Current progress) on August 27, 2012 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 6 (２) 新たなシビアアクシデント対策規制の枠組みのイメージ 大規模な拡散 深層防護層の目的 ④ 放出抑制・拡散緩和策 放水車等 LRF 大規模な 放出防止 格納容器 ④-１ 破損防止 設計基準に基 づく設備の安 ③-2 全機能喪失 （多重故障) 現行の設計上の想定 防 災 放出抑制・ 拡散緩和 ④-2 大規模な放出 格納容器損傷 著しい 炉心損傷 安全上の措置 人的被害防止 環境回復 ⑤ 設計基準事故 ＋安全機能維 ③-１ 持 （単一故障） 著しい炉心損傷防止 プラント状態 異常・故障の 発生 ② 事故への 拡大防止 通常運転 ① 異常・故障 発生防止 現行の規制範囲 格納容器損傷 防止策（SA対策Ⅱ） 移動式RHR等 特定安全施設 代替格納容器ｽﾌﾟﾚｲ等 著しい炉心損傷 防止策（SA対策Ⅰ） 電源車等 特定安全施設 電源融通等 ECCS （緊急炉心冷却装置）等 原 子 炉 設 置 者 の 自 主 的 取 組 異常検知・停止操作等 インターロック等 内的事象 裕度 向上 設 計 基 準 外的事象の強度 外的事象（自然現象、航空機衝突、テロ等） 出典：原子力安全・保安院「発電用軽水型原子炉施設におけるシビアアクシデント対策規制の基本的考え方について（現時点での検討状況）」（平成24年8月27日） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. Evaluation of Exposure Dose Equivalents during Normal Operation 7 In the reactor facilities, it is crucial to try to keep radiation exposure dose in surrounding public and nuclear plant workers as low as reasonably achievable. This way of thinking is called ALARA, from the first letter of each word. To this end, Nuclear Safety Commission of Japan prescribed “Regulatory Guide for the Annual Dose Target for the Public in Vicinity of Light Water Nuclear Power Reactor Facilities” to manage radioactive material generated by nuclear reactor operation strictly and keep release into surrounding environment as low as reasonably achievable. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. 通常運転時の被ばく線量当量の評価 7 原子炉施設では、平常運転時において、原子炉施設の運転 に伴う周辺公衆及び従事者の被ばく線量が、合理的に達成可 能な限り低くなるように努める（このような考え方をＡｓ Low As Reasonably Achievable（ALARA）の精神という）ことが大事であ るとされている。 そのために、原子力安全委員会により「発電用軽水型原子炉 施設周辺の線量目標値に関する指針」が定められ、原子炉の 運転に伴い発生する放射性物質を厳重に管理し、周辺環境へ の放出を合理的に達成可能な限り抑制するよう設備することと している。 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. Evaluation of Exposure Dose Equivalents during Normal Operation 8 (1) Methods of Radioactive Waste Treatment and Routes of Impact on the Local Populace Gaseous waste [Irradiated undiluted gases (e.g. 16N, 19O, 3H)] Attenuation using filters and hold-up devices, for example, and release from exhaust stacks while monitoring radiation levels (Cause of external and internal exposure to radiation) Liquid waste [Irradiated rust, etc., incorporating water (e.g. 60Co, 58Co, 54Mn)] Attenuation processing using filters, for example, and release into the sea after confirmation that radioactive concentration is equivalent to or less than standard values (Cause of external and internal exposure to radiation) Solid waste [e.g. Contaminated cloths, filter sludge)] Attenuation and volume reduction using incinerators and attenuation tanks, for example, and storage in silos (Cause of external exposure to radiation) Source: Safety Evaluation from the Perspective of Safety Reviews, “Unfailing safety” through “appropriate review” (Agency of Natural Resources and Energy, Ministry of International Trade and Industry) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. 通常運転時の被ばく線量当量の評価 8 (1) 放射性廃棄物の処理方法と周辺公衆への影響経路 気体廃棄物［放射化した非凝縮性ガス（16N，19O，3H 等）］ フィルタ，ホールドアップ装置等により減衰処理し，放射線レベルを監視しながら，排気筒より放出 （外部・内部被ばく要因） 液体廃棄物［放射化した錆等を含む水（60Co,58Co,54Mn 等）］ フィルタ等により減衰処理し，放射能濃度が基準値以下であることを確認後，海洋放出（外部・内部被ばく要因） 固体廃棄物［汚染したウェス，フィルタスラッジ等］ 焼却炉・減衰タンク等により減衰・減容処理し，貯蔵庫に保管（外部被ばく要因） 出典；安全審査から見た安全評価 「適切な審査」で「確かな安全」を（通商産業省資源エネルギー庁） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. Evaluation of Exposure Dose Equivalents during Normal Operation 9 (2) Evaluation of Exposure Dose Equivalents during Normal Operation ALARA(As Low As Reasonably Achievable) Judgment criteria (Dose target value): 0.05 mSv/Year or less Measures to mitigate exposure of the local populace to radiation: Control (Target values) of release of liquid and gaseous waste Evaluation of exposure of the local populace to radiation Volume of radioactive gaseous waste released Volume of radioactive liquid waste released Calculation of annual effective dose equivalent produced by radioactive noble gases Calculation of annual effective dose equivalent produced by radioiodine Total Calculation of the maximum annual effective dose equivalent value to which the local populace is exposed Comparison with dose target value Calculation of annual effective dose equivalent produced by radioactive substances other than iodine (mSv/Year) Annual effective Dose Equivalent Annual effective dose equivalent produced by radioactive noble gases 0.0030 Annual effective dose equivalent produced by radioiodine contained in gaseous and liquid waste 0.0006 Annual effective dose equivalent produced by radioactive substances (not including iodine) contained in liquid waste 0.0045 (mSv/Year) Dose to which the populace in the vicinity of the nuclear power plant is exposed Source: Safety Evaluation from the Perspective of Safety Reviews, “Unfailing safety” through “appropriate review” (Agency of Natural Resources and Energy, Ministry of International Trade and Industry) The Japan Atomic Power Company Maximum annual effective Dose Equivalent Value Dose Target Value 0.0081 0.05 Dose target outside the monitoring zone around a LWR plant (annual)･･･0.05 Measures to mitigate and evaluation of exposure to radiation of personnel engaged in work involving radiation - Work management (Time, distance, shielding) - Shielding equipment, air conditioning equipment, zoning) - Dose measurement, exposure history management, health checks Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 3. 通常運転時の被ばく線量当量の評価 9 (2) 通常運転時の被ばく線量当量評価 ALARA（As Low As Reasonably Achievable） 判断基準（線量目標値）； 0.05 mSv/年 以下 周辺公衆に対する被ばく低減対策；液体，及び気体廃棄物の放出管理（目標値） 周辺公衆に対する被ばく評価 放射性気体 廃棄物の 放出量 放射性希ガスによる 年間実効線量当量を算出する 放射性よう素による 年間実効線量当量を算出する 放射性液体 廃棄物の 放出量 合算 発電所周辺の公衆 が受ける年間実効線 量当量の最大値を算 出する 線量目標値と 比較する よう素を除く放射性物質による 年間実効線量当量を算出する （ミリシーベルト／年） （ミリシーベルト／年） 年間実効線量当量 放射性希ガスによる 年間実効線量当量 0.0030 気体・液体廃棄物中に含 まれる放射性よう素による 年間実効線量当量 0.0006 液体廃棄物中に含まれる 放射性物質（よう素を除く） による年間実効線量当量 発電所周辺の 公衆が受ける 放射線量 年間実効線量 当量の最大値 線量目標値 0.0081 0.05 0.0045 出典；安全審査から見た安全評価 「適切な審査」で「確かな安全」を（通商産業省資源エネルギー庁） The Japan Atomic Power Company Dose target outside the monitoring zone around a LWR plant (annual)･･･0.05 放射線業務従事者に対する被ばく低減対策・評価 ・作業管理（時間・距離・遮へい） ・遮へい設備，空調設備，区域管理 ・被ばく線量測定，被ばく履歴の管理，健康診断 Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 10 (1) Definition of safety evaluation It is the action to evaluate an adequacy of the safety design quantitatively in the process for securing of reactor safety, which comes in pairs of safety design. After that, the process for securing of safety regarding construction and operation follows. Radiation disaster prevention for surrounding public and nuclear plant workers Approval procedures corresponding to them Safety Design Safety Assessment Construction Management of Reactor Facilities Operation Management of Reactor Facilities The Japan Atomic Power Company Installment License (Safety Review) Construction plan approval, preoperation inspection, welding inspection, fuel assembly inspection Approval of operational safety program, regular inspection, safety inspection Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 10 (1) 安全評価とは 原子炉の安全性確保に係るプロセスのうち、安全設計の妥当 性を定量的に評価する活動で、安全設計と対になる。後に建設、 運転に係る安全性確保のプロセスが続く。 放射性物質による周辺公衆ないしは従業員への放射線災害の防止 左記に対応する許認可手続き 安全設計 安全評価 原子炉施設の建設管理 設置許可（安全審査） 工事計画認可、使用前検査、 溶接検査、燃料体検査 原子炉施設の運転管理 保安規定認可、 定期検査、保安検査 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 11 INSAG-3 In the deterministic method, design base events (DBE) are chosen to encompass a range of related possible initiating events which could challenge the safety of the plant. Analysis is used to show that the response of the plant and its safety systems to design basis events satisfies predetermined specifications both for the performance of the plant itself and for meeting safety targets. Safety evaluation in Japan (Reactor Installation Approval Standards＊１) Reactor Installation Approval Standards Article 13 Prevention of abnormal operational transients and escalation of design base accidents Confirmation on compliance with the requirements imposed on facilities subject to the design standards Article 37 Prevention of escalation of severe accidents and others Confirmation on effectiveness of countermeasures implemented at facilities dealing with severe accidents, etc. ＊１：NRA Ordinance Prescribing Standards for the Location, Structure, and Equipment of Commercial Power Reactors and their Auxiliary Facilities Safety evaluation in Japan (existing guidelines) Safety design LWR Facilities guidelines and safety assessment guidelines Confirmation on appropriateness of safety design Siting Evaluation guidelines Confirmation on appropriateness of site conditions (isolation from the public) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 11 INSAG-3/BSP プラントの安全性を脅かし得る起因事象を包絡するように設計基準事象 （DBE）を選定し、DBEに対するプラント及び安全系の応答がプラントの性能 及び安全上の目標への適合性について、あらかじめ定められた仕様を満足す ることを立証する（決定論的手法） 我が国における安全評価（設置許可基準規則） 設置許可基準規則 第十三条 運転時の異常な過渡変化及び設計基準事故の拡大の防止 設計基準対象施設の満たすべき要件を満たしているか確認 第三十七条 重大事故等の拡大の防止等 重大事故等対処施設などによる対策が有効であるか確認 我が国における安全評価（従来の指針類） 安全設計審査指針・安全評価指針 安全設計の妥当性を確認 立地審査指針 立地条件の適否（公衆との離隔）を確認 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation a. 12 Safety design evaluation/effectiveness evaluation Scope of evaluation Abnormal transients during operation Abnormal statuses caused by a singular failure or malfunction of component or singular erroneous operation by an operator that could take place during normal operation, or disturbances of external origin foreseen to occur at a frequency similar to those accidents, which shall be postulated in the safety design process as phenomena that could lead to severe damages to the core or the pressure boundary of reactor coolant if left to continue. Design base accidents Abnormal statuses of a frequency lower than abnormal transients during normal operation, which shall be postulated in the safety design process as phenomena that could lead to release of large amounts of radioactive substances from nuclear power generation facilities if arising. Severe accidents and others Severe accidents: severe damages to the core or fuels and spent fuels stored in nuclear fuel materials storage facilities Severe accidents and others: accidents that could lead to the severe accidents (excluding the abnormal transients during normal operation and design base accidents; hereinafter the same), or the severe accidents themselves The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 12 a. 安全設計評価／有効性評価 評価すべき範囲 「運転時の異常な過渡変化」 「設計基準事故」 通常運転時に予想される機械又は器具の単一の故障若しくはその 誤作動又は運転員の単一の誤操作及びこれらと類似の頻度で発生 すると予想される外乱によって発生する異常な状態であって、当 該状態雅継続した場合には炉心又は原子炉冷却材圧力バウンダリ の著しい損傷が生ずるおそれがあるものとして安全設計上想定す べきもの 発生頻度が運転時の異常な過渡変化より低い異常な状態であって、 当該状態が発生した場合には発電用原子炉施設から多量の放射性 物質が放出するおそれがあるものとして安全設計上想定すべきも の 「重大事故等」 重大事故：炉心の著しい損傷、核燃料物質貯蔵設備に貯蔵する燃 料体又は使用済燃料の著しい損傷 重大事故等：重大事故に至るおそれがある事故（運転時の異常な 過渡変化及び設計基準事故を除く。以下同じ。）又は重大事故 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 13 b. Design base events Events to be evaluated The abnormal transients during operation and design base accidents are subject to evaluation Evaluation is conducted pursuant to Examination Guidelines for Safety Assessment of Light Water Reactor Facilities for Nuclear Power Generation Combinations of the design base events (DBE) and operational conditions mainly of relevant abnormal impact mitigation systems (MS) Abnormal statuses are enveloped appropriately through analyses of a certain number of events The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 13 b. 設計基準事象 評価すべき事象 「運転時の異常な過渡変化」及び「設計基準事故」が対象 「発電用軽水型原子炉施設の安全評価に関する審査指針」等 に基づいて評価を実施 設計基準事象（DBE）と、これに関連する主として異常影響 緩和系（MS）の作動状況の組合せ 異常状態を、ある限られた数の事象の解析で適切に包絡 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 14 b. Design base events Events to be evaluated Abnormal transients during operation Representative events shall be selected for assessment from among the events, which may potentially lead to excessive damage to the core or the reactor coolant pressure boundary if the nuclear reactor facility is left uncontrolled, from the viewpoint of confirming the adequacy of the designed functions of structures, systems and components belonging in general to abnormality mitigation systems, or simply referred to as mitigation systems (MSs), such as the safety protection system and the reactor shutdown system.. Categories of events Abnormal changes in the reactivity or power distribution in the core Abnormal changes in the heat generated or removed in the core Abnormal changes in the pressure or inventory of reactor coolant Other events recognized as necessary based on the design of nuclear reactor facilities If similar events exist, the most severe event with regard to the judgment criteria shall represent others The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 14 b. 設計基準事象 評価すべき事象 「運転時の異常な過渡変化」 原子炉施設が制御されずに放置されると、炉心あるいは原子炉冷 却材圧力バウンダリに過度の損傷をもたらす可能性のある事象に ついて、これらの事象が発生した場合における安全保護系、原 子炉停止系等の主としてMSに属する構築物、系統及び機器の設 計の妥当性を確認する見地から、代表的な事象を選定 事象のカテゴリ 炉心内の反応度又は出力分布の異常な変化 炉心内の熱発生又は熱除去の異常な変化 原子炉冷却材圧力又は原子炉冷却材保有量の異常な変化 その他原子炉施設の設計により必要と認められる事象 類似事象がある場合は、判断基準に照らして最も厳しい事象で代 表 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 15 b. Design base events Events to be evaluated Design base accidents Representative events shall be selected for assessment from among the events, which may potentially lead to undue exposure of the off-site public by the radioactive materials released from the nuclear reactor facility, from the viewpoint of confirming the adequacy of the designed functions of structures, systems and components belonging in general to MSs such as the engineered safety features. Categories of events Loss of reactor coolant or considerable change in core cooling Abnormal insertion of reactivity or rapid changes in reactor power Abnormal release of radioactive materials to the environment Abnormal changes in reactor containment vessel pressure, atmosphere, etc. Other events recognized as necessary based on the design of nuclear reactor facilities If similar events exist, the most severe event with regard to the judgment criteria shall represent others. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 15 b. 設計基準事象 評価すべき事象 「設計基準事故」 原子炉施設から放出される放射性物質による敷地周辺への影響が 大きくなる可能性のある事象について、これらの事象が発生した 場合における工学的安全施設等の主としてMSに属する構築物、 系統及び機器の設計の妥当性を確認する見地から、代表的な事象 を選定 事象のカテゴリ 原子炉冷却材の喪失又は炉心冷却状態の著しい変化 反応度の異常な投入又は原子炉出力の急激な変化 環境への放射性物質の異常な放出 原子炉格納容器内圧力、雰囲気等の異常な変化 その他原子炉施設の設計により必要と認められる事象 類似事象がある場合は、判断基準に照らして最も厳しい事象で代 表 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited The specific events to be evaluated(PWR) 16 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety,” Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 評価すべき具体的な事象（PWR) 16 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited The specific events to be evaluated(PWR) Abnormal operational transients a) Abnormal changes in the reactivity or the output distribution in the core Accidents a) Loss of reactor coolant or significant change in core cooling condition ① Abnormal control rod withdrawal in the start up time of the reactor 1 Loss of reactor coolant ② Abnormal control rod withdrawal during operation 2 Loss of reactor coolant flow ③Control rod fall and imbalance of output 3 Shaft sticking for reactor coolant pump ④ Abnormal dilution of boron in reactor coolant 4 Breakage of main feed water pipe 5 Breakage of main steam pipe b) Abnormal changes in calorification and heat removal in the core ⑤ Partial loss of reactor coolant ⑥ Incorrect operation of stop loop of reactor coolant system ⑦ Loss of external power supply b) Abnormal increase of reactivity or rapid change in reactor output 6 Control rod ejection c) Abnormal release of radioactive material into environment ⑧ Loss of main feed water 7 Damage of radioactive gas waste disposal site ⑨ Abnormal increase of steam load 8 Breakage of steam generator tube ⑩ Abnormal pressure decrease of secondary cooling system 9 Fuel assembly fall ⑪ Excessive feed water to steam generator 1 Loss of reactor coolant 6 Control rod ejection c) Abnormal changes in reactor coolant pressure or reactor coolant holdings 17 d) Abnormal changes in reactor containment vessel and atmosphere, etc. ⑫ Loss of load ⑬ Abnormal pressure decrease of reactor cooling system ⑭ Incorrect operation of ECCS during operation The Japan Atomic Power Company 1 Loss of reactor coolant 10 Combustible gas generation Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 評価すべき具体的な事象（PWR) 運転時の異常な過渡変化（PWR：14） a) 炉心内の反応度又は出力分布の異常な変化 事故（PWR：13） a) 原子炉冷却材の喪失又は炉心冷却状態の著しい変化 ① 原子炉起動時における制御棒の異常な引き抜き 1 原子炉冷却材喪失 ② 出力運転中の制御棒の異常な引き抜き 2 原子炉冷却材流量の喪失 ③ 制御棒の落下及び不整合 3 原子炉冷却材ポンプの軸固着 ④ 原子炉冷却材中のほう素の異常な希釈 4 主給水管破断 b) 炉心内の熱発生又は熱除去の異常な変化 5 主蒸気管破断 ⑤ 原子炉冷却材流量の部分喪失 ⑥ 原子炉冷却材系の停止ループの誤起動 ⑦ 外部電源喪失 b) 反応度の異常な投入又は原子炉出力の急激な変化 6 制御棒飛び出し c) 環境への放射性物質の異常な放出 ⑧ 主給水流量喪失 7 放射性気体廃棄物処理施設の破損 ⑨ 蒸気負荷の異常な増加 8 蒸気発生器伝熱管破損 ⑩ 2次冷却系の異常な減圧 9 燃料集合体の落下 ⑪ 蒸気発生器への過剰給水 1 原子炉冷却材喪失 6 制御棒飛び出し c) 原子炉冷却材圧力又は原子炉冷却材保有量の異常な変化 ⑫ 負荷の喪失 d) 原子炉格納容器内圧力、雰囲気等の異常な変化 ⑬ 原子炉冷却材系の異常な減圧 1 原子炉冷却材喪失 ⑭ 出力運転中の非常用炉心冷却系の誤起動 10 可燃性ガスの発生 The Japan Atomic Power Company 17 Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited The specific events to be evaluated(BWR) 18 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety,” Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 評価すべき具体的な事象（BWR) 18 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited The specific events to be evaluated(BWR) Abnormal operational transients a) Abnormal changes in the reactivity or the output distribution in the core Accidents a) Loss of reactor coolant or significant change in core cooling condition ① Abnormal control rod withdrawal in the start up time of the reactor 1 Loss of reactor coolant ② Abnormal control rod withdrawal during operation 2 Loss of reactor coolant flow b) Abnormal changes in calorification and heat removal in the core 3 Shaft sticking for reactor coolant pump ③ Partial loss of reactor coolant ④ Incorrect operation of stop loop of reactor coolant system ⑤ Loss of external power supply b) Abnormal increase of reactivity or rapid change in reactor output 4 Control rod fall c) Abnormal release of radioactive material into environment ⑥ Loss of feed water heater 5 Damage of radioactive gas waste disposal site ⑦ Incorrect operation of reactor coolant flow control system 6 Breakage of main steam pipe 7 Fuel assembly fall ⑧ Loss of load 1 Loss of reactor coolant ⑨ Incorrect close of main steam isolation valve 4 Control rod fall c) Abnormal changes in reactor coolant pressure or reactor coolant holding ⑩ Failure of feed water control system d) Abnormal changes in reactor containment vessel and atmosphere, etc. ⑪ Failure of reactor pressure control system 1 Loss of reactor coolant ⑫ Overall loss of feed water flow 8 Combustible gas generation 9 Dynamic load generation The Japan Atomic Power Company 19 Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 評価すべき具体的な事象（BWR) 運転時の異常な過渡変化（BWR：12） a) 炉心内の反応度又は出力分布の異常な変化 事故（BWR：12） a) 原子炉冷却材の喪失又は炉心冷却状態の著しい変化 ① 原子炉起動時における制御棒の異常な引き抜き 1 原子炉冷却材喪失 ② 出力運転中の制御棒の異常な引き抜き 2 原子炉冷却材流量の喪失 b) 炉心内の熱発生又は熱除去の異常な変化 3 原子炉冷却材ポンプの軸固着 ③ 原子炉冷却材流量の部分喪失 ④ 原子炉冷却材系の停止ループの誤起動 ⑤ 外部電源喪失 b) 反応度の異常な投入又は原子炉出力の急激な変化 4 制御棒落下 c) 環境への放射性物質の異常な放出 ⑥ 給水加熱喪失 5 放射性気体廃棄物処理施設の破損 ⑦ 原子炉冷却材流量制御系の誤動作 6 主蒸気管破断 7 燃料集合体の落下 ⑧ 負荷の喪失 1 原子炉冷却材喪失 ⑨ 主蒸気隔離弁の誤閉止 4 制御棒落下 c) 原子炉冷却材圧力又は原子炉冷却材保有量の異常な変化 ⑩ 給水制御系の故障 d) 原子炉格納容器内圧力、雰囲気等の異常な変化 ⑪ 原子炉圧力制御系の故障 1 原子炉冷却材喪失 ⑫ 給水流量の全喪失 8 可燃性ガスの発生 9 動荷重の発生 The Japan Atomic Power Company 19 Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 20 b. Design base events Judgment criteria Abnormal transients during operation The design shall allow restoration of the nuclear reactor facilities to normal operation, without damaging the core. The minimum critical heat flux ratio or the minimum critical power ratio shall be larger than the acceptable limit. Fuel cladding shall not be mechanically damaged. Fuel enthalpy shall not exceed the acceptable limit. (The standard for preventing high-temperature rupture, melting, and embrittlement of cladding tubes) Pressure on the reactor coolant pressure boundary shall not exceed 110% of the maximum allowable working pressure. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 20 b. 設計基準事象 判断基準 「運転時の異常な過渡変化」 炉心は損傷に至ることなく、かつ、原子炉施設は通常運転に復帰 できる状態で事象が収束される設計であること 最小限界熱流束比又は最小限界出力比が許容設計限界値以上 燃料被覆材が破損しないこと 燃料エンタルピーが燃料要素の許容損傷限界を超えないこと（被覆管 の高温破裂、溶融及び脆化防止の基準） 原子炉冷却材圧力バウンダリにかかる圧力が最高使用圧力の1.1倍以下 となること The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation b. 21 Design base events Judgment criteria Design base accidents It shall be verified that the nuclear reactor facility is designed such that a postulated event does not lead to melting or considerable damage of the core, that the event does not cause, in its process, a secondary damage which would lead to another abnormal condition, and that the function of the barriers against the release of radioactive materials in the event is adequate. The core shall not be damaged considerably, and adequate coolable state of the core shall be. (The standard related to embrittlement accompanying metal-water reaction (oxidization) of cladding tubes) The maximum temperature of fuel cladding: 1,200ºC or less The amount of oxidization of a fuel cladding tube: 15% of the cladding tube thickness or less The fuel enthalpy shall not exceed the limit values for maintaining integrity of the nuclear core or the pressure boundary of reactor coolant. (The standard for preventing generation of pressure wave due to melting and evaporating of pellets) The pressure exerted on the pressure boundary of reactor coolant shall be 1.2 times the maximum working pressure or less. The pressure on the reactor containment boundary and the temperature at the boundary of the reactor containment shall not exceed the maximum allowable working pressure and temperature, respectively. The facilities subject to the design standard shall not cause radiation damages to the public in factories and other facilities in the vicinity. (5mSv) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 21 b. 設計基準事象 判断基準 「設計基準事故」 炉心の溶融あるいは著しい損傷のおそれがなく、かつ、事象の過程におい て他の異常状態の原因となるような2次的損傷が生じなく、さらに放射性 物質の放散に対する障壁の設計が妥当であること 炉心は著しい損傷が発生するおそれがないものであり、かつ、炉心を十分に冷却 できるものであること（被覆管の金属－水反応（酸化）に伴う脆化に係る基準） 燃料被覆管の最高温度は、1,200℃以下 燃料被覆管の酸化量は、被覆管厚さの15%以下 燃料エンタルピーが炉心及び原子炉冷却材圧力バウンダリの健全性を維持するた めの制限値を越えないこと（ペレットの溶融、蒸発による圧力波発生防止の基 準） 原子炉冷却材圧力バウンダリにかかる圧力が最高使用圧力の1.2倍以下となること 原子炉格納容器バウンダリにかかる圧力及び原子炉格納容器バウンダリにおける 温度が最高使用圧力及び最高使用温度以下となること 設計基準対象施設が工場等周辺の公衆に放射線障害を及ぼさないものであること （5mSv） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation b. 22 Design base events Matters to be considered in analyses Scope of consideration The conditions before occurrence of abnormal status constitute the most severe initial conditions with regard to the judgment criteria, taking into consideration the entire scope of normal operation and operation period, changes in core burn-up during the cycle period, and variations due to fuel replacement, etc. All abnormal events that could arise during all the phases of normal operation (startup, shutdown, output operations, hot standby, fuel replacement, and other operational processes of nuclear reactor facilities) are covered. Up to the point where it is reasonable to assume that the situation would resume to a normal state and reach a cold shutdown without difficulty Assumptions regarding safety functions Reliable safety functions Among the safety functions for dealing with events, the functions that may be considered in the analyses are those belonging to MS-1 and 2, as a principle. (Explanations on the appropriateness is needed if MS-3 is to be relied on; reliability equivalent to that for general industrial facilities is not acceptable, as a principle) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 22 b. 設計基準事象 解析に当たって考慮すべき事項 考慮する範囲 異常状態の発生前の状態は、通常運転範囲及び運転期間の全域、 サイクル期間中の炉心燃焼度変化、燃料交換等による変動を考慮 して、判断基準に照らして最も厳しくなる初期状態 通常運転（起動、停止、出力運転、高温待機、燃料取替等の原子炉施 設の運転）の全域にわたって生じ得る異常事象のすべてを包絡 事象が収束して支障なく冷態停止に至ることが合理的に推定でき る時点まで 安全機能に対する仮定 期待できる安全機能 事象に対処するための安全機能のうち、解析で考慮することができる ものは、原則、MS-1、2に属する機能（MS-3に期待する場合は妥当性 の説明要、一般産業施設と同等の信頼性では、原則、期待できない） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation b. 23 Design base events Matters to be considered in analyses Assumptions regarding safety functions Assumption of singular failure Concerning structures, systems, and equipment for dealing with accidents, a singular equipment failure that brings about the most severe analysis result is assumed for each of the basic safety functions; namely, nuclear reactor shutdown, core cooling, and radioactivity containment. Applicable to active equipment for a short period following the occurrence of an event, and to active or passive equipment over a long period Passive equipment is exempted if removal or restoration is possible within a time period that would not affect safety, or if the probability of occurrence is sufficiently low. Equipment that starts operating before the occurrence of an event and continues operating after the occurrence is exempted. Manual operation by operators (the 10-minute rule) An appropriate allowance in time (10 minutes) is considered regarding manual operation by operators necessary for dealing with events. As a principle, the design shall not rely on operations by operators immediately after occurrence of an abnormal status. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 23 b. 設計基準事象 解析に当たって考慮すべき事項 安全機能に対する仮定 単一故障の仮定 「事故」に対処する構築物、系統及び機器について、「原子炉停止」、 「炉心冷却」及び「放射能閉じ込め」の各基本的安全機能別に、解析 の結果を最も厳しくする機器の単一故障を仮定 事象発生後短期間は動的機器、長期間にわたっては動的機器又は静的 機器に適用 静的機器については、安全上支障のない時間内に除去又は修復できる 場合、又は、その発生確率が十分低い場合は適用を除外 事象発生前から動作、かつ、発生後も動作する機器は適用を除外 運転員の手動操作（ 10分ルール） 事象に対処するために必要な運転員の手動操作については、適切な時 間的余裕（10分間）を考慮 異常状態の発生直後は運転員の操作を期待しない設計が原則 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation b. 24 Design base events Matters to be considered in analyses Assumptions regarding safety functions Assumption of loss of external power source If operation of engineered safety features is to be relied on, cases in which an external power source is unavailable shall also be considered when analyzing accidents. However, cases in which external power source remains sound also need to be considered, because effects of availability of external power sources differ by events. Effects of reactor scram Appropriate scram delay time shall be considered. Aside from a singular failure of the nuclear reactor shutdown function, the shutdown effects shall be considered while assuming that a control rod of the maximum reactivity worth is held at fully withdrawn position. (One rod stuck) Computation programs, models, and parameters for analysis Confirm the appropriateness of computation programs, etc. Select models and parameters that bring about more severe evaluation results. Allow for appropriate safety margin when there are uncertain factors regarding parameters. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 24 b. 設計基準事象 解析に当たって考慮すべき事項 安全機能に対する仮定 外部電源喪失の仮定 原子炉スクラムの効果 「事故」の解析に当たって、工学的安全施設の動作を期待する場合は、 外部電源が利用できない場合も想定 ただし、外部電源の有無による影響は事象により異なるため、外部電 源が健全である場合の考察が必要 適切なスクラム遅れ時間を考慮 原子炉停止機能の単一故障とは別に、最大反応度価値の制御棒1本が 全引抜位置にあるものとした（One Rod Stuck）停止効果を考慮 解析に使用する計算プログラム、モデル及びパラメータ 計算プログラム等の妥当性の確認 評価結果を厳しくするモデル、パラメータの選定 パラメータに不確定因子がある場合の適切な安全余裕 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation (ex. Design Base Accident) 25 Examples of Safety Evaluation of Accident Phenomena (PWR) 1) Loss of Nuclear Reactor Coolant [Overview of Phenomenon] This is an assumed phenomenon for confirmation that the design properly ensures coolant for the reactor core in cases where there is a reactor core coolant outflow due to, for example, damage to the piping in the container connected to the nuclear reactor. It is assumed that damage to the piping, for example, that configures the nuclear reactor coolant pressure boundary during reactor drive operation causes an outflow of reactor coolant and a consequent drop in the reactor core cooling capacity. A pipe from the outlet of the primary coolant pump to the nozzle at the inlet of the reactor vessel ruptures at its double end. ( Loss of off-site power is also assumed.) Water level in the reactor falls. Reactor pressure sets to drop. Temperature of the fuel cladding tubes goes up. Bubble grows in the core. Water is automatically injected by the accumulator system, the highpressure injection system, and the low-pressure injection system of the emergency core cooling system (ECCS). Reactor self-control effect (Void effect) makes the reactor power decreases. Temperature of the fuel cladding tubes goes down. Water level in the reactor goes up. Reactor pressure sets to drop. ⇒Given a shutdown signal, all control rods are automatically inserted into the core. Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 25 ( 例 設計基準事故） 事故事象の安全評価の例（PWR) ①原子炉冷却材喪失 【事象の概要】 原子炉に接続されている配管が格納容器内で破損すること等によって，原子炉冷却材が流出する場合について，炉心の冷却が適切に 確保される設計となっていることを確認するために想定する事象です。 原子炉の出力運転中に，原子炉冷却材圧力バウンダリを構成する配管等の破損等により，原子炉冷却材が流出し，炉心の冷却能力が 低下することを想定します。 1次冷却材ポンプ出口から原子炉容器の 入口ノズルまでの間の配管の両端破断 （外部電源の喪失も想定する） 原子炉の水位が低下する 原子炉の圧力が低下する 燃料被覆管の温度が上昇する 炉心内で急激に気泡が発生する 原子炉の自己制御効果(ボイド効果）により 原子炉出力が低下する 非常用炉心冷却装置 （ECCS)の蓄圧注入系・高 圧注入系・余熱除去系によ り自動注水する 原子炉の圧力が低下する ⇒ 原子炉停止系が作動する 燃料被覆管の温度が低下する 原子炉の水位が上昇する Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 26 It is assumed the pipe from the outlet of the primary coolant pump to the nozzle at the inlet of the reactor vessel ruptures at its double end. Examples of Safety Evaluation of Accident Phenomena (PWR) 1) Loss of Nuclear Reactor Coolant Fuel cladding tubes temp (℃) Decision criteria (1,200℃) Fuels start submerged ECCS works Rupture occurs Time(sec) Changes in temperature of fuel cladding tubes Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価( 例 設計基準事故） 26 事故事象の安全評価の例（PWR) ①原子炉冷却材喪失 1次冷却材ポンプ出口から原子 炉容器入口ノズルまでの間の 配管が両端破断したと想定す る 燃料被覆管の温度(℃) 判断基準(1,200℃) 燃料が浸水開始(再冠水開始） ECCS 作動 ブローﾀﾞｳﾝ終了 炉心フラッシングによる除熱量の減少及びそ れに続く炉心流量停滞による温度上昇 破断発生 Time(sec) 燃料被覆管の温度の変化 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 27 Examples of Safety Evaluation of Accident Phenomena (BWR) 1) Loss of Reactor Coolant [Overview of Phenomenon] This is an assumed phenomenon for confirmation that the design properly ensures coolant for the reactor core in cases where there is a reactor core coolant outflow due, for example, to damage to the piping in the container connected to the nuclear reactor. Fuel rods Reactor water level(m) With water around the fuel rods evaporating due to the still high hiea valve,cooling temporarily runs short Core starts exposed due to the lowering water level Submerged again due to ECCS Cooling by steam is operation taking place Water level is restored by the temporarily growing core flow Automatic insertion of control rods Rupture occurs ECCS works Fuel cladding tubes temperature (℃) It is assumed that damage to the piping, for example, that configures the nuclear reactor coolant pressure boundary during reactor power operation causes an outflow of coolant and a consequent drop in the reactor core cooling capacity. Decision criteria (1,200℃) Temperature briefly rises with the flow of coolant temporarily running short With top of the fuels Submerged again,and Exposed temperature rises Temperature falls Automatic insertion of control rods Rupture occurs ECCS works Time(sec) Change in reactor water level Time(sec) Change in temperature of fuel cladding tubes Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety,” Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価( 例 設計基準事故） 27 事故事象の安全評価の例（BWR) ①原子炉冷却材喪失 【事象の概要】 原子炉に接続されている配管が格納容器内で破損すること等に よって，原子炉冷却材が流出する場合について，炉心の冷却が適切 に確保される設計となっていることを確認するために想定する事象で す。 原子炉の出力運転中に，原子炉冷却材圧力バウンダリを構成する 配管等の破損等により，原子炉冷却材が流出し，炉心の冷却能力が 低下することを想定します。 Source: Safety Evaluation Affirmed by Safety Examinations, “Adequate Examinations” Ensure “Solid Safety”, Agency of Natural Resources and Energy Ministry of International Trade and Industry JAPAN The Japan Atomic Power Company Handle with care; restricted to authorized persons; 取扱注意 関係者限り 目的外使用・複製・開示等禁止 use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation c. 28 Site evaluation (reference) Events to be evaluated Conceptual accidents that are not likely to take place are considered, for confirming isolation from the public. Severe accidents Serious accidents that could take place in the worst case, when seen from technical viewpoints Loss-of-coolant accident (LOCA) (BWR/PWR), rupture of main steam pipe (BWR), rupture of heat transfer pipe from steam generator (PWR) Hypothetical accidents Accidents that are worse than the severe accidents and are not conceivable to take place when seen from technical viewpoints (An example assumed is diffusion of radioactive materials due to malfunctioning of some of the safety protection facilities that were expected to function in a severe accident) Loss-of-coolant accident (LOCA) (BWR/PWR), rupture of main steam pipe (BWR), rupture of heat transfer pipe from steam generator (PWR) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 28 c. 立地評価（参考） 評価すべき事象 公衆からの離隔を確認するため、現実には起こる蓋然性のな い観念的な事故を対象 「重大事故」 技術的見地からみて、最悪の場合には起こるかもしれないと考え られる重大な事故 LOCA（BWR/PWR）、主蒸気管破断事故（BWR）、蒸気発生器伝熱管破 損事故（PWR） 「仮想事故」 重大事故を超えるような技術的見地からは起こるとは考えられな い事故（例えば、重大事故で期待した安全防護施設のいくつかが 動作しないと仮想し、放射性物質の放散を仮想） LOCA（BWR/PWR）、主蒸気管破断事故（BWR）、蒸気発生器伝熱管破 損事故（PWR） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 29 Severe accidents and others Events to be evaluated Prevention of substantial core damage Evaluation is conducted pursuant to Examination Guidelines for Effectiveness Evaluation of Countermeasures to Prevent Damages to Reactor Cores and Reactor Containment Vessels of Commercial Nuclear Power Reactors. Prevention of damages to reactor containment vessels Evaluation is conducted pursuant to Examination Guidelines for Effectiveness Evaluation of Countermeasures to Prevent Damages to Reactor Cores and Reactor Containment Vessels of Commercial Nuclear Power Reactors. Prevention of damages to fuel assemblies in spent fuel storage pools Evaluation is conducted pursuant to Examination Guidelines for Effectiveness Evaluation of Countermeasures to Prevent Damages to Fuel Assemblies in Spent Fuel Storage Pools of Commercial Nuclear Power Reactors. Prevention of damages to fuel assemblies in nuclear reactors that have been shut down Evaluation is conducted pursuant to Examination Guidelines for Effectiveness Evaluation of Countermeasures to Prevent Damages to Fuel Assemblies in ShutDown Nuclear Reactors of Commercial Nuclear Power Reactors. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 29 d. 重大事故等 評価すべき事象 炉心の著しい損傷の防止 原子炉格納容器の破損防止 「実用発電用原子炉に係る炉心損傷防止対策及び格納容器破損防 止対策の有効性評価に関する審査ガイド」に基づいて評価を実施 使用済燃料貯蔵槽内の燃料損傷の防止 「実用発電用原子炉に係る炉心損傷防止対策及び格納容器破損防 止対策の有効性評価に関する審査ガイド」に基づいて評価を実施 「実用発電用原子炉に係る使用済燃料貯蔵槽における燃料損傷防 止対策の有効性評価に関する審査ガイド」に基づいて評価を実施 運転停止中原子炉内の燃料損傷の防止 「実用発電用原子炉に係る運転停止中原子炉における燃料損傷防 止対策の有効性評価に関する審査ガイド」に基づいて評価を実施 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 30 Severe accidents and others Events to be evaluated Prevention of substantial core damage Accident sequences to be evaluated are those assuming the possibility of severe reactor core damages when the safety functions have been lost with structures, systems, and equipment that are required to be designed not to allow impairment of the reactor safety in cases of abnormal transients during operation and design base accidents. Accident sequence group to be assumed in all cases (see the table below) Accident sequence group extracted by individual plant evaluations (probabilistic risk assessment; PRA) BWR PWR Loss of high-pressure and low-pressure water injection functions Loss of function to remove heat from the secondary system Loss of high-pressure water injection and depressurization functions Loss of all AC power sources (with or without seal LOCA) Loss of all AC power sources Loss of component cooling water system Loss of decay heat removal function (loss of water intake function, defect in RHR) Loss of heat removal function of nuclear reactor containment vessel Loss of nuclear reactor shutdown function Loss of nuclear reactor shutdown function Loss of water injection function at times of LOCA (large-scale rupture, small- and medium-scale rupture) Loss of water injection function of ECCS (LOCA with large-scale rupture or small-, medium-scale rupture) Containment vessel bypass (IS-LOCA) Loss of recirculation function of ECCS (LOCA with large-scale or small-, medium-scale rupture) Containment vessel bypass (IS-LOCA, SGTR) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 30 d. 重大事故等 評価すべき事象 炉心の著しい損傷の防止 運転時の異常な過渡変化及び設計基準事故に対して原子炉の安全性を損な うことがないよう設計することを求められている構築物、系統及び機器が その安全機能を喪失した場合であって、炉心の著しい損傷に至る可能性が あると想定する事故シーケンスが対象 必ず想定する事故シーケンスグループ（下表） 個別プラント評価（確率論的リスク評価、PRA）により抽出した事故シーケンス グループ BWR PWR 高圧・低圧注水機能喪失 2次系からの除熱機能喪失 高圧注水・減圧機能喪失 全交流動力電源喪失（シールLOCAあり、無し） 全交流動力電源喪失 原子炉補機冷却機能喪失 崩壊熱除去機能喪失（取水機能喪失、RHR故障） 原子炉格納容器の除熱機能喪失 原子炉停止機能喪失 原子炉停止機能喪失 LOCA時注水機能喪失（大破断、中小破断） ECCS注水機能喪失（大破断LOCA、中小破断LOCA） 格納容器バイパス（IS-LOCA） ECCS再循環機能喪失（大破断LOCA、中小破断LOCA） 格納容器バイパス（IS-LOCA、SGTR） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 31 Severe accidents and others Events to be evaluated Prevention of damages to nuclear reactor containment vessels The containment vessel damage modes to be evaluated are those assumed at times of severe accidents Containment vessel damage modes to be assumed in all cases (see the table below) Containment vessel damage modes extracted by individual plant evaluations (probabilistic risk assessment; PRA) Static loads deriving from atmospheric pressure and temperature (containment vessel damages due to overpressure and hyperthermia) Discharge of high-pressure molten materials / direct heating of atmosphere in containment vessels Interaction between molten fuel and coolant outside of nuclear reactor pressure vessels Hydrogen combustion Direct contact with containment vessels (shell attack) Interaction between molten reactor core and concrete The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 31 d. 重大事故等 評価すべき事象 原子炉格納容器の破損の防止 重大事故が発生した場合において想定する格納容器破損モードが 対象 必ず想定する格納容器破損モード（下表） 個別プラント評価（確率論的リスク評価、PRA）により抽出した格納容 器破損モード 雰囲気圧力・温度による静的負荷（格納容器過圧・過温破損） 高圧溶融物放出／格納容器雰囲気直接加熱 原子炉圧力容器外の溶融燃料－冷却材相互作用 水素燃焼 格納容器直接接触（シェルアタック） 溶融炉心・コンクリート相互作用 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 32 Phenomena in containment vessels Hydrogen burning PCV damage due to semistatic excessive pressure FP behavior in PCV FP behavior in the primary system High-pressure melted core exposure (direct contact with PCV ambient atmosphere) PCV leak due to excessive temperature of penetration portion High-pressure melted core exposure (direct contact with PCV) Vapor explosion in RPV Damage in RPV Vapor explosion in RPV Core concrete reaction The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 32 （格納容器内の事象） 水素燃焼 準静的過圧による PCV破損 PCV内のFP挙動 1次系内のFP挙動 貫通部過温による PCV漏えい 高圧溶融炉心噴出 （PCV雰囲気直接接触） 高圧溶融炉心噴出 （PCV直接接触） RPV内での水蒸気爆発 RPV破損 PCV内での水蒸気爆発 コア－コンク リート反応 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 33 Severe accidents and others Events to be evaluated Prevention of damages to fuel assemblies in spent fuel storage pools Assumed accident 1 An accident in which the water temperature inside a spent fuel storage pool rises due to loss of the cooling function and water injection function in the pool, and the water level lowers due to evaporation Assumed accident 2 An accident in which a small amount of water in a spent fuel storage pool is lost due to a siphoning phenomenon and other reasons, and the water level of the pool lowers Prevention of damages to fuel assemblies in nuclear reactors that have been shut down The accident sequence group during shutdown to be assumed in all cases (see the table below) Accident sequence group during shutdown extracted by individual plant evaluation (probabilistic risk assessment; PRA) Loss of the decay heat removal function (loss of the cooling function during shutdown due to defect in RHR) Loss of all the AC power sources Flow out of nuclear reactor coolant Erroneous insertion of reactivity The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 33 d. 重大事故等 評価すべき事象 使用済燃料貯蔵槽内の燃料損傷の防止 想定事故１ 使用済燃料貯蔵槽の冷却機能又は注水機能が喪失することにより、使用 済燃料貯蔵槽内の水の温度が上昇し、蒸発により水位が低下する事故 想定事故２ サイフォン現象等により使用済燃料貯蔵槽内の水の小規模な喪失が発生 し、使用済燃料貯蔵槽の水位が低下する事故 運転停止中原子炉内の燃料損傷の防止 必ず想定する運転停止中事故シーケンスグループ（下表） 個別プラント評価（確率論的リスク評価、PRA）により抽出した運転停 止中事故シーケンスグループ 崩壊熱除去機能喪失（RHR故障による停止時冷却機能喪失） 全交流動力電源喪失 原子炉冷却材の流出 反応度誤投入 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 34 Severe accidents and others Judgment criteria Prevention of substantial core damage Concerning cases among the accident sequence group assumed in which the nuclear reactor containment vessel is expected to continue functioning after suffering severe damages to the core, sufficient countermeasures for preventing severe damages to the core shall have been planned and furthermore the countermeasures shall prove effective within the assumed scope. Concerning cases among the accident sequence group assumed in which it is difficult to expect the nuclear reactor containment vessel to continue functioning after suffering severe damages to the core, the countermeasures for preventing severe damages to cores shall be effective. The core shall be free of possibilities of serious damage, and furthermore sufficient cooling of the core shall be possible. (The standard related to embrittlement accompanying metal-water reaction (oxidization) of cladding tubes) Maximum temperature of the fuel cladding: 1,200ºC or less Amount of oxidization of a fuel cladding tube: 15% of the cladding tube thickness before severe oxidization or less The pressure exerted on the pressure boundary of reactor coolant shall be less than 1.2 times the maximum working pressure or the critical pressure. The pressure exerted on the boundary of nuclear reactor containment vessel shall be less than the maximum working pressure or the critical pressure. The temperature at the boundary of nuclear reactor containment vessel shall be less than the maximum working temperature or the critical temperature. In an effective evaluation of accident sequence groups using pressure release apparatus in the containment vessel, no serious risk of radiation exposure shall be posed to the public in the vicinity. (Roughly up to 5mSv per accident) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 34 d. 重大事故等 判断基準 「炉心の著しい損傷の防止」 想定する事故シーケンスグループのうち炉心の著しい損傷後に原子炉格納 容器の機能に期待できるものにあっては、炉心の著しい損傷を防止するた めの十分な対策が計画されており、かつ、その対策が想定する範囲内で有 効性があること。 想定する事故シーケンスグループのうち炉心の著しい損傷後の原子炉格納 容器の機能に期待することが困難なものにあっては、炉心の著しい損傷を 防止する対策に有効性があること。 炉心は著しい損傷が発生するおそれがないものであり、かつ、炉心を十分に冷却 できるものであること（被覆管の金属－水反応（酸化）に伴う脆化に係る基準） 燃料被覆管の最高温度は、1,200℃以下 燃料被覆管の酸化量は、酸化が著しくなる前の被覆管厚さの15%以下 原子炉冷却材圧力バウンダリにかかる圧力が最高使用圧力の1.2倍又は限界圧力を 下回ること 原子炉格納容器バウンダリにかかる圧力が最高使用圧力又は限界圧力を下回るこ と 原子炉格納容器バウンダリにかかる温度が最高使用温度又は限界温度を下回るこ と 格納容器圧力逃がし装置を使用する事故シーケンスグループの有効性評価では、 周辺の公衆に対して著しい放射線被ばくのリスクを与えないこと（発生事故当た り概ね5mSv以下） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 35 Severe accidents and others Judgment criteria Prevention of damages to nuclear reactor containment vessels Concerning the containment vessel damage modes assumed, measures to prevent damages to nuclear reactor containment vessels and furthermore measures to prevent release of radioactive substances outside of the site at an abnormal level shall be effective. The pressure exerted on the boundary of nuclear reactor containment vessel shall be less than the maximum working pressure or the critical pressure. The temperature at the boundary of nuclear reactor containment vessel shall be less than the maximum working temperature or the critical temperature. The total amount of radioactive substances released shall pose as small an impact on the environment as possible, also taking into account the viewpoints on environmental pollution by radioactive substances. (The amount of Cs-137 released shall be no more than 100TBq) The pressure of nuclear reactor coolant shall have been lowered to 2.0MPa or less, by the time the nuclear reactor pressure vessel suffers damage. The functions of the boundary of nuclear reactor containment vessel shall not be lost due to the thermal and mechanical loads caused by rapid interactions between the molten fuel and coolant outside of the nuclear reactor pressure vessel. Hydrogen detonation that could damage the nuclear reactor containment vessel shall be prevented. The hydrogen concentration shall be 13 vol% or less when converted to dry conditions, or the oxygen concentration shall be 5 vol% or less. The standards for pressure exerted on the boundary of nuclear reactor containment vessel shall be met even when a flammable gas has accumulated or undergone combustion. In case the molten core that falls on the floor of nuclear reactor containment vessel spreads, it shall not come into contact with the boundary of the nuclear reactor containment vessel, and the molten core shall be cooled appropriately. The supporting function of structural members of the nuclear reactor containment vessel shall not be lost due to corrosion by the molten core, and the molten core shall be cooled appropriately. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 35 d. 重大事故等 判断基準 「原子炉格納容器の破損の防止」 想定する格納容器破損モードに対して、原子炉格納容器の破損を防止し、 かつ、放射性物質が異常な水準で敷地外へ放出されることを防止する対策 に有効性があること。 原子炉格納容器バウンダリにかかる圧力が最高使用圧力又は限界圧力を下回るこ と 原子炉格納容器バウンダリにかかる温度が最高使用温度又は限界温度を下回るこ と 放射性物質の総放出量は、放射性物質による環境への汚染の視点も含め、環境へ の影響をできるだけ小さくとどめるものであること（Cs-137の放出量が100TBq以 下） 原子炉圧力容器の破損までに原子炉冷却材圧力は2.0MPa以下に低減されているこ と 急速な原子炉圧力容器外の溶融燃料－冷却材相互作用による熱的・機械的荷重に よって原子炉格納容器バウンダリの機能が喪失しないこと 原子炉格納容器が破損する可能性のある水素の爆轟を防止すること 水素濃度がドライ条件に換算して13vol%以下又は酸素濃度が5vol%以下で あること 可燃性ガスの蓄積、燃焼が生じた場合においても、原子炉格納容器バウンダリに かかる圧力の基準を満足すること 原子炉格納容器の床上に落下した溶融炉心が床面を拡がり原子炉格納容器バウン ダリと接触しないこと及び溶融炉心が適切に冷却されること Handle with care; restricted to authorized persons; The Japan Atomic Power Company use outside of the purposes of use, duplication, and disclosure are prohibited 溶融炉心による侵食によって、原子炉格納容器の構造部材の支持機能が喪失しな 4. Safety Design Evaluation d. 36 Severe accidents and others Judgment criteria Prevention of damages to fuel assemblies in spent fuel storage pools The following evaluation items shall be satisfied with regard to the assumed accidents 1 and 2. The top of the effective length of the fuels shall be underwater. A water level that maintains the radiation shielding properties shall be secured. The subcritical state shall be maintained. Prevention of damages to fuel assemblies in nuclear reactors that have been shut down The following evaluation items shall be satisfied with regard to the assumed accident sequence group during shutdown. The top of the effective length of the fuels shall be underwater. A water level that maintains the radiation shielding properties shall be secured. The subcritical state shall be maintained. (Excluding the criticality during normal operation and temporary criticality with minute rise in output that would not affect the soundness of the fuel) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 36 d. 重大事故等 判断基準 「使用済燃料貯蔵槽内の燃料損傷の防止」 想定事故１及び想定事故２に対して、以下の評価項目を満足する こと。 燃料有効長頂部が冠水していること 放射線の遮蔽が維持される水位を確保すること 未臨界が維持されること 「運転停止中原子炉内の燃料損傷の防止」 想定する運転停止中事故シーケンスグループに対して、以下の評 価項目を満足すること。 燃料有効長頂部が冠水していること 放射線の遮蔽が維持される水位を確保すること 未臨界が維持されること（通常の運転操作における臨界、又は燃料の 健全性に影響を与えない一時的かつ僅かな出力上昇を伴う臨界は除 く） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 37 Severe accidents and others Matters to be considered in analyses (Prevention of substantial core damage) Method and scope of effective evaluation Best estimate method is applied. (Application of conservative assumptions and conditions is not denied) Models that have been verified through experiments and others and are with an appropriate scope of application shall be used. When a model with high uncertainty is used or if the scope of application of a verified model is exceeded, its impacts shall be taken into consideration appropriately based on the results of sensitivity analysis, etc. As a principle, the analysis shall cover up to the time point when the accident has been cleared and the nuclear reactor has reached a stable shutdown (hightemperature or low-temperature shutdown). (At least seven days, assuming that no outside support is available. If the stable condition was reached in a period less than seven days, it must be demonstrated that the stable conditions may be maintained.) If there are several countermeasures, basically the effectiveness of each shall be evaluated. (If the envelope coverage of evaluation conditions may be demonstrated, an effectiveness evaluation based on the envelope conditions may be conducted to represent the others) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 37 d. 重大事故等 解析に当たって考慮すべき事項（炉心の著しい損傷の防止） 有効性評価の手法及び範囲 最適評価手法を適用（保守的な仮定及び条件の適用は否定しな い） 実験等を基に検証され、適用範囲が適切なモデルを使用 不確かさが大きいモデルを使用する場合又は検証されたモデル の適用範囲を超える場合には、感度解析結果等を基にその影響 を適切に考慮 原則として事故が収束し、原子炉が安定停止状態（高温停止状 態又は低温停止状態）に導かれる時点まで評価（少なくとも外 部支援がないものとして7日間。7日間より短い期間で安定状態 に至った場合は、その状態を維持できることを示す。） 複数の対策がある場合には、各々の対策について有効性を評価 することを基本とする（評価条件の包絡性を示すことができれ ば、包絡条件による有効性評価で代表させることも可） The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 38 Severe accidents and others Matters to be considered in analyses (Prevention of substantial core damage) Common conditions for effectiveness evaluation The nuclear reactor is operated at the rated thermal output. The output distribution at the core, core flow rate, decay heat, and other parameters shall be of realistic values supported by design values, etc. Conditions applicable to facilities for dealing with design base accidents The design values are used for the capacity of facilities. (If values other than the design values are used, the grounds and appropriateness for using the value must be demonstrated.) Instrumental errors are not taken into consideration for operation set points and other parameters. Except for facilities assumed to fail, facilities may be expected to function if appropriateness of the expectation has been demonstrated. (The pressure, temperature, water level, etc. of the nuclear reactor) Facilities assumed to fail are not expected to recover. Effects of availability of external power sources shall be taken into consideration. Operating conditions of facilities for dealing with severe accidents and others The time to implement the countermeasures is set in consideration of the training records, etc. The operating conditions, capacity, and delay time are set based on the design specifications. If there is uncertainty about the operating environment, its effects shall be considered. Singular failures are not assumed. The appropriateness of operating procedures related to the countermeasures shall be demonstrated. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 38 d. 重大事故等 解析に当たって考慮すべき事項（炉心の著しい損傷の防止） 有効性評価の共通条件 原子炉は定格熱出力運転 炉心の出力分布、炉心流量及び崩壊熱等は設計値等に基づく現実的な値 設計基準事故対処設備の適用条件 設備の容量は設計値を使用（設計値と異なる値を使用する場合は、その根拠と 妥当性を示すこと）し、作動設定点等については計装上の誤差は考慮しない 故障した想定を設備を除き、設備の機能を期待することの妥当性（原子炉の圧 力、温度及び水位等）が示された場合には、その機能を期待できる 故障を想定した設備の復旧には期待しない 外部電源の有無の影響を考慮 重大事故等対処設備の作動条件 対策実施時間は訓練実績等に基づき設定 作動条件、容量及び時間遅れを、設計仕様に基づき設定する 作動環境の不確かさがある場合は、その影響を考慮 単一故障は仮定しない 対策に関連する操作手順の妥当性を示す The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 39 Severe accidents and others Matters to be considered in analyses (Prevention of damages to nuclear reactor containment vessels) Method and scope of effectiveness evaluation Best estimate method is applied. (Application of conservative assumptions and conditions is not denied) Codes that have been verified through experiments and others and are with the appropriate scope of application shall be used. When a model with high uncertainty is used or if the scope of application of a verified model is exceeded, its impacts shall be taken into consideration appropriately based on the results of sensitivity analysis, etc. As a principle, the analysis shall cover up to the time point when the accident has been cleared and the nuclear reactor and the nuclear reactor containment vessel have reached a stable condition. (At least seven days, assuming that no outside support is available. If the stable condition is reached in a period less than seven days, it must be demonstrated that the stable condition may be maintained.) If there are several countermeasures being implemented, the effectiveness of each shall be evaluated. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 39 d. 重大事故等 解析に当たって考慮すべき事項（原子炉格納容器の破損の防 止） 有効性評価の手法及び範囲 最適評価手法を適用（保守的な仮定及び条件の適用は否定しな い） 実験等に基に検証され、適用範囲が適切なコードを使用 不確かさが大きいモデルを使用する場合又は検証されたモデルの 適用範囲を超える場合には、感度解析結果等を基にその影響を適 切に考慮 原則として事故が収束し、原子炉及び原子炉格納容器が安定状態 に導かれる時点まで評価（少なくとも外部支援がないものとして 7日間。7日間より短い期間で安定状態に至った場合は、その状態 を維持できることを示す。） 複数の対策がとられている場合には、各々の対策について有効性 を評価する The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 40 Severe accidents and others Matters to be considered in analyses (Prevention of damages to nuclear reactor containment vessels) Common conditions for effectiveness evaluation The nuclear reactor is operated at the rated thermal output. The output distribution at the core, core flow rate, decay heat, and other parameters shall be of realistic values supported by design values, etc. Conditions applicable to facilities for dealing with design base accidents The design values are used for the capacity of facilities. (If values other than the design values are used, the grounds and appropriateness for using the value must be demonstrated.) Instrumental errors are not taken into consideration for operation set points and other parameters. Except for facilities assumed to fail, facilities may be expected to function if appropriateness of the expectation has been demonstrated. (The pressure, temperature, water level, etc. of the nuclear reactor containment vessel) Facilities assumed to fail are not expected to recover. Effects of availability of external power sources shall be taken into consideration. Operating conditions of facilities for dealing with severe accidents and others The time to implement the countermeasures is set in consideration of the training records, etc. The operating conditions, capacity, and delay time are set based on the design specifications. If there is uncertainty about the operating environment, its effects shall be considered. Singular failures are not assumed. The appropriateness of operating procedures related to countermeasures shall be demonstrated. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 40 d. 重大事故等 解析に当たって考慮すべき事項（原子炉格納容器の破損の防 止） 有効性評価の共通条件 原子炉は定格熱出力運転 炉心の出力分布、炉心流量及び崩壊熱等は設計値等に基づく現実的な値 設計基準事故対処設備の適用条件 設備の容量は設計値を使用（設計値と異なる値を使用する場合は、その根拠と 妥当性を示すこと）し、作動設定点等については計装上の誤差は考慮しない 故障した想定を設備を除き、設備の機能を期待することの妥当性（原子炉格納 容器内の圧力、温度及び水位等）が示された場合には、その機能を期待できる 故障を想定した設備の復旧には期待しない 外部電源の有無の影響を考慮 重大事故等対処設備の作動条件 対策実施時間は訓練実績等に基づき設定 作動条件、容量及び時間遅れを、設計仕様に基づき設定する 作動環境の不確かさがある場合は、その影響を考慮 単一故障は仮定しない 対策に関連する手順の妥当性を示す The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 41 Severe accidents and others Matters to be considered in analyses (Prevention of damages to fuel assemblies in spent fuel storage pools) Method and scope of effectiveness evaluation Best estimate method is applied. (Application of conservative assumptions and conditions is not denied) As a principle, the analysis shall cover up to the time point when the water level of spent fuel storage pool has resumed to the original level and the water level and temperature have reached a stable condition. (At least seven days, assuming that no outside support is available. If the stable condition is reached in a period less than seven days, it must be demonstrated that the stable condition may be maintained.) Common conditions for effectiveness evaluation Conditions inside of the spent fuel storage pool It is assumed that fuels for the entire core that have been taken out in the shortest period possible after the nuclear reactor shutdown are being temporarily stored, separately from stored fuels. Decay heat shall be evaluated appropriately taking into consideration the fuel composition, burn-up, etc. and based on the design. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 41 d. 重大事故等 解析に当たって考慮すべき事項（使用済燃料貯蔵槽内の燃料 損傷の防止） 有効性評価の手法及び範囲 最適評価手法を適用（保守的な仮定及び条件の適用は否定しな い） 使用済燃料貯蔵槽の水位が回復し、水位及び温度が安定した状態 に導かれる時点まで評価（少なくとも外部支援がないものとして 7日間。7日間より短い期間で安定状態に至った場合は、その状態 を維持できることを示す。） 有効性評価の共通条件 使用済燃料貯蔵槽内の状態等 貯蔵燃料の他に、原子炉停止後に最短で取り出された全炉心分の燃料 が一時保管されていることとする 崩壊熱は、燃料組成及び燃焼度等を考慮し設計に基づき適正に評価 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 42 Severe accidents and others Matters to be considered in analyses (Prevention of damages to fuel assemblies in spent fuel storage pools) Common conditions for effectiveness evaluation (cont’d.) Conditions applicable to safety facilities The design values are used for the capacity of facilities. (If values other than the design values are used, the grounds and appropriateness for using the value must be demonstrated.) Instrumental errors are not taken into consideration for operation set points and others. Except for facilities assumed to fail, facilities may be expected to function if appropriateness of the expectation has been demonstrated. (The stand-by conditions of facilities and effects of changes of temperature, water level, etc. of the spent fuel storage pool) Facilities assumed to fail are not expected to recover. Effects of availability of external power sources shall be taken into consideration. Operating conditions of facilities for dealing with severe accidents and others The time to implement the countermeasures is set in consideration of the training records, etc. The operating conditions, capacity, and delay time are set based on the design specifications. If there is uncertainty about the operating environment, its effects shall be considered. Singular failures are not assumed. The appropriateness of operating procedures related to the countermeasures shall be demonstrated. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 42 d. 重大事故等 解析に当たって考慮すべき事項（使用済燃料貯蔵槽内の燃料 損傷の防止） 有効性評価の共通条件（続き） 安全施設の適用条件 設備の容量等は設計値を使用（設計値と異なる値を使用する場合は、その根拠 と妥当性を示すこと）し、作動設定点等については計装上の誤差は考慮しない 故障した想定を設備を除き、設備の機能を期待することの妥当性（設備の待機 状態及び使用済燃料貯蔵槽の温度、水位等の変化の影響等）が示された場合に は、その機能を期待できる 故障を想定した設備の復旧には期待しない 外部電源の有無の影響を考慮 重大事故等対処設備の作動条件 対策実施時間は訓練実績等に基づき設定 作動条件、容量及び時間遅れを、設計仕様に基づき設定する 作動環境の不確かさがある場合は、その影響を考慮 単一故障は仮定しない 対策に関連する手順の妥当性を示す The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 43 Severe accidents and others Matters to be considered in analyses (Prevention of damages to fuel assemblies in nuclear reactors that have been shut down) Method and scope of effectiveness evaluation Best estimate method is applied. (Application assumptions and conditions is not denied) of conservative As a principle, the analysis shall cover up to the time point when the accident has been cleared and the nuclear reactor has reached a stable condition. Common conditions for effectiveness evaluation The period of nuclear reactor shutdown The period from the parallel-off of major power generators in the nuclear reactor shutdown process to the parallel of major power generators in the nuclear reactor startup process (except when all the fuels have been taken out into a spent fuel storage pool) The nuclear reactor shutdown period shall be segmented appropriately to suit the pressure, temperature, and water level of the nuclear reactor and the work progress situations, etc. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 43 d. 重大事故等 解析に当たって考慮すべき事項（運転停止中原子炉内の燃料 損傷の防止） 有効性評価の手法及び範囲 最適評価手法を適用（保守的な仮定及び条件の適用は否定しな い） 原則として事故が収束し、原子炉が安定した状態に導かれる時点 まで評価 有効性評価の共通条件 原子炉の運転停止中の期間 原子炉運転停止の過程における主発電機の解列から、原子炉起動の過 程における主発電機の並列までの期間（全燃料が使用済燃料貯蔵槽に 取り出された場合は除く） 原子炉の運転停止中の期間を、原子炉の圧力、温度、水位及び作業状 況等に応じて適切に区分 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 44 Severe accidents and others Matters to be considered in analyses (Prevention of damages to fuel assemblies in nuclear reactors that have been shut down) Common conditions for effectiveness evaluation (cont’d.) The flow rate at the nuclear reactor core, decay heat, and other parameters shall be of realistic values supported by design values, etc. Conditions applicable to safety facilities The design values are used for the capacity of facilities. (If values other than the design values are used, the grounds and appropriateness for using the value must be demonstrated.) Instrumental errors are not taken into consideration for operation set points and others. Except for facilities assumed to fail or that are exempted from stand-by, facilities may be expected to function if appropriateness of the expectation has been demonstrated. (The pressure, temperature, water level, etc. of the nuclear reactor) Facilities assumed to fail or that are exempted from stand-by are not expected to recover. Effects of availability of external power sources shall be taken into consideration. Operating conditions of facilities for dealing with severe accidents and others The time to implement the countermeasures is set in consideration of the training records, etc. The operating conditions, capacity, and delay time are set based on the design specifications. If there is uncertainty about the operating environment, its effects shall be considered. Singular failures are not assumed. The appropriateness of operating procedures related to the countermeasures shall be demonstrated. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 44 d. 重大事故等 解析に当たって考慮すべき事項（運転停止中原子炉内の燃料 損傷の防止） 有効性評価の共通条件（続き） 原子炉内の炉心流量及び崩壊熱等は設計値等に基づく現実的な値 安全施設の適用条件 設備の容量は設計値を使用（設計値と異なる値を使用する場合は、その根拠と 妥当性を示すこと）し、作動設定点等については計装上の誤差は考慮しない 故障又は待機除外を仮定した設備を除き、設備の機能を期待することの妥当性 （原子炉の圧力、温度及び水位等）が示された場合には、その機能を期待でき る 故障又は待機除外を仮定した設備の復旧には期待しない 外部電源の有無の影響を考慮 重大事故等対処設備の作動条件 対策実施時間は訓練実績等に基づき設定 作動条件、容量及び時間遅れを、設計仕様に基づき設定する 作動環境の不確かさがある場合は、その影響を考慮 単一故障は仮定しない 対策に関連する操作手順の妥当性を示す The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. 45 Severe accidents and others Probabilistic risk assessment (PRA) Used for extracting accident sequence groups of individual plants For nuclear power stations and other large-scale, complicated systems, the system safety (or risks) may be evaluated comprehensively and quantitatively with regard to all possible accidents that could occur, by estimating and assessing the probability of occurrence of the accidents and their impacts. WASH-1400 (1975) Analytically evaluates comprehensive risks of nuclear power stations Comprehensively applies the probabilistic approach Foresaw events similar to the Three Mile Island accident (1979) (Drew attention to probabilistic safety assessment technologies) NUREG-1150 (1990) A revised version of WASH-1400 used for risk assessment at five plants in the United States Latest findings on severe accidents Uncertainty evaluation The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 45 d. 重大事故等 確率論的リスク評価（PRA） 個別プラントの事故シーケンスグループの抽出で使用 原子力発電所など、大規模で複雑なシステムについて、発生 し得るあらゆる事故を対象として、その事故の発生確率と発 生したときの影響を推定・評価し、システムの安全性（又は リスク）を総合的、定量的に評価することが可能 WASH-1400（1975） 原子力発電所の総合的なリスクを解析的に評価 確率論的アプローチを包括的に適用 TMI事故（1979）の類似事象を予見（PSA技術の注目） NUREG-1150（1990） WASH-1400の改訂版、米国5プラントのリスク評価 シビアアクシデントの最新知見 不確かさ評価 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 46 d. Severe accidents and others PRA Item Deterministic safety assessment Probabilistic safety assessment (1) Object of evaluation The most severe process is evaluated among diverse event processes that may be anticipated, by gathering events into a relatively small number of typical events and assigning conservative conditions. Events that are not conceivable in terms of technical viewpoints are excluded, but the grounds of exclusion are not clear. As a principle, all the initiating events are evaluated; in reality, however, a small number of groups of typical initiating events are evaluated. The entire spectra of event processes (all the conceivable processes) that follow the occurrence of initiating events are evaluated. (2) Handling of failures Handled pursuant to the singular failure standards. Dependent failures that arise as a result of singular failure are considered, but failures of common causes are excluded from evaluation, in view of independence of design. Multiple failure is considered, as are failures of common causes. (3) Analysis conditions Conservative conditions that bring about more severe results Nominal (realistic) conditions, as a principle Conservative conditions may be set when uncertainty is significant. (4) Analysis results Analysis values derive from representative physical quantities (parameters) for judging appropriateness of safety design, etc. Various analysis values related to risks are obtained, such as risks to the reliability of systems and equipment and the level of contribution to the risks, in addition to the risk values for evaluating comprehensive safety. The risk values, reliability, and other results are indicated by the most probable values and their uncertainty levels. The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 46 d. 重大事故等 PRA 項目 決定論的安全評価 確率論的安全評価 （1）評価対象 比較的少数の代表事象に集約し、保守的な 条件を与えることにより、予想される多様な 事象推移のうち最も厳しい推移を評価する。 工学的見地から起こるとは考えられない事 象は排除されるが、その根拠は明確でない。 原則として全起因事象を対象とするが現実には 少数群の代表起因事象を対象とする。 起因事象発生後の事象推移は全スペクトル（考 えられる推移の全て）を評価する。 （2）故障の扱い 単一故障基準による。 単一故障の結果生じる従属故障は考慮され るが、共通原因故障は設計上の独立性の 配慮等から評価では排除される。 多重故障、さらに共通原因故障についても考慮 される。 （3）解析条件 結果が厳しくなるよう保守的な条件。 原則としてノミナル（現実的）条件。 不確かさが大きい場合は保守的な設定もあり得 る。 （4）解析結果 安全設計の妥当性等を判断するための代 表的な物理量（パラメータ）の解析値が得ら れる。 総合的な安全性を評価するためのリスク値のほ か、システム、機器の信頼度やリスクへの寄与 度等のリスクに関連した種々の解析値が得られ る。 リスク値、信頼度等の結果は、最確値とその不 確かさで表現される。 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 47 d. Severe accidents and others PRA Item Deterministic safety assessment Probabilistic safety assessment (5) Judgment on results The appropriateness of safety design and other aspects are judged based on whether the analysis values of representative physical quantities (parameters) satisfy the prescribed criteria. Judgment is made based on whether the analysis values satisfy the judgment criteria prescribed for various uses. (6) Uses/ utilization The suitability of site conditions and appropriateness of safety design are confirmed in the basic design phase. Utilized for decision-making not only in the basic design phase, but also for detailed design, operation management, maintenance, and management in the construction and operation phases. (7) Others Unable to respond with regard to hypothetical appropriateness of analyses, such as elimination of multiple failures. Difficult to reach agreement on the sufficiency of conservativeness of evaluation. The reliability of assessment results is affected by inadequacy of knowledge (on severe accidents and others) and uncertainty in probabilistic estimation (frequency of initiating events, equipment failure ratio, etc.) The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 47 d. 重大事故等 PRA 項目 決定論的安全評価 確率論的安全評価 （5）結果の判断 代表的な物理量（パラメータ）の解析値が、定 められた基準を満たすことにより安全設計の 妥当性等が判断される。 解析値が、種々の用途で定められた判断基準 を満たすことにより判断される。 （6）用途/活用 基本設計段階における立地条件の適否、安 全設計の妥当性の確認 基本設計段階のみならず、建設、運転段階に おける詳細設計、運転管理、保守管理におけ る意思決定に活用 （7）その他 多重故障の排除等、解析上の仮定の適切性 については答えられない。 評価の保守性の十分性についての合意が得 られ難い。 評価結果の信頼性は、知識の不完全さ（シビ アアクシデントの知見等）、確率推定の不確か さ（起因事象の発生頻度、機器故障率等）に左 右される。 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 48 d. Severe accidents and others PRA PRA levels and processes Categories and evaluation levels of PRA Causal events Level Earthquake Fire Flood Flying objects Danger level evaluation Level Failure evaluation Studies on external cause-initiated events Classification of core damage accident sequences Reliability analysis on safety systems Core damage frequency Classification of containment vessel damage sequences External causes Internal causes Studies on internal cause-initiated events Response and damage evaluation Level Studies on occurrence probability of phenomena and events Containment vessel damage frequency Core damage accident analysis Containment vessel damage frequency Source term Accident occurrence frequency evaluation Accident development evaluation Analysis on fission products transfer in the environment and public exposure dose Level Level Source term evaluation at time of accident Risk Level Analysis on fission products transfer in the environment Level Public risk assessment PRA process Source: ATOMICA, an atomic energy encyclopedia on the Internet The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 48 d. 重大事故等 PRA PRAのレベルと手順 PRAの分類と評価レベル PRAの手順 出典：原子力百科事典ATOMICA The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation 49 d. Severe accidents and others Pipe rupture PRA Power source Shutdown system Emergency core cooling Radiation removal Containment vessel Loss of safety facility power source Occurrence probability Sequence Damaged Loss of DC power source Loss of AC power source Unsuccessful Damaged Actuated Unsuccessful Unsuccessful Actuated Loss of AC power source in site Loss of AC power source outside of site Unsuccessful Broke down Initial abnormality Unsuccessful Unsuccessful Loss of power transmission system A Failure of common cause Loss of power transmission system B Broke down Example of events tree when responding to LOCA Loss of diesel generator A Loss of diesel generator B Loss of diesel generator C Example of fault tree Source: Shunsuke Kondo, Nuclear Power Safety, Dobunshoin, 1990, P. 196 Source: Shunsuke Kondo, Nuclear Power Safety, Dobunshoin, 1990, P. 199 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 49 d. 重大事故等 PRA LOCAに対するイベントツリーの例 フォールトツリーの例 The Japan Atomic Power Company Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. Safety Design Evaluation d. Severe accidents and others PRA Transients 50 Reactor shutdown Example of confirmation on whether there is an accident sequence group to be extracted through PRA on individual plant Core Integrity of HighLowAccident Decay heat pressure pressure depressuriz pressure removal sequence group boundary core cooling core cooling ation Successful No core damage Successful Failed Successful Successful Successful Successful Failed Successful Failed Failed No core damage (d) (a) Failed (b) Successful Successful Failed Successful Failed Successful Successful Failed (d) Make certain to confirm the existence of only the accident sequence groups assumed. (Conduct PRA on all the initiating events assumed, and if an accident sequence group other than those assumed has been extracted, make certain to add it to the accident sequence groups assumed.) Failed Failed No core damage (d) No core damage (d) (a) Failed (b) Failed (e) The Japan Atomic Power Company Accident sequence groups (a) Loss of high-pressure and low-pressure water injection function (b) Loss of high-pressure water injection and depressurization functions (c) Loss of all AC power sources (d) Loss of decay heat removal function (e) Loss of nuclear reactor shutdown function (f) Loss of water injection function at times of LOCA (g) Containment vessel bypassing (LOCA at interface system) Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited 4. 安全設計評価 50 d. 重大事故等 PRA 個別プラントのPRAにより抽出すべき事故シーケンスグループの有無の 確認例 圧力 過渡事象 原子炉停止 バウンダリ 健全性 高圧炉心 低圧炉心 事故シーケンス 原子炉減圧 崩壊熱除去 冷却 冷却 グループ 成功 成功 失敗 成功 成功 成功 成功 失敗 成功 失敗 失敗 （ｄ） 炉心損傷なし （ｄ） （ａ） 失敗 （ｂ） 成功 成功 失敗 失敗 成功 成功 成功 失敗 炉心損傷なし 失敗 失敗 失敗 失敗 The Japan Atomic Power Company 炉心損傷なし （ｄ） 炉心損傷なし （ｄ） （ａ） （ｂ） （ｅ） 必ず想定する事故シーケン スグループのみであること を確認（想定される起因事 象すべてでPRAを実施し、 必ず想定する事故シーケン スグループ以外の事故 シーケンスグループが抽出 された場合には、想定する 事故シーケンスグループに 追加） 事故シーケンスグループ (a)高圧・低圧注水機能喪失 (b)高圧注水・減圧機能喪失 (c)全交流動力電源喪失 (d)崩壊熱除去機能喪失 (e)原子炉停止機能喪失 (f)LOCA時注水機能喪失 (g)格納容器バイパス（ｲﾝﾀｰﾌｪｲｽｼｽﾃﾑLOCA） Handle with care; restricted to authorized persons; use outside of the purposes of use, duplication, and disclosure are prohibited
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