TOMOFUMI YAMAMOTO et al.[Left hand page running head is author’s name in Times New Roman 8 point bold capitals, centred. For more than two authors, write

AUTHOR et al.]DEVELOPMENTS FOR NUCLEAR POWER PLANT SAFETYOverview of Technology Developments for Continuous Improvements of Nuclear Safety

TOMOFUMI YAMAMOTONuclear Energy Systems Division, Mitsubishi Heavy Industries, Ltd.Tokyo, JapanEmail: tomofumi_yamamoto@mhi.co.jp

AKIRA OHNUKINuclear Energy Systems Division, Mitsubishi Heavy Industries, Ltd.Kobe, Japan

HIROHI SHIMIZUResearch & Innovation Center, Mitsubishi Heavy Industries, Ltd.Hiroshima, Japan

Abstract

Since Fukushima Daiichi Accident in 2011, improvements and strengthening of nuclear safety have been discussed and implemented in Japan. To strengthen nuclear safety with the concept of Defence-in-Depth, prevention of accidents is essential as well as mitigation of severe accidents. Several research and development (R&D) programs have been conducted to improve the safety of nuclear power plants (NPP) under the government support program in Japan. For instance, core cooling properties using steam generators under station blackout condition were verified. For instance, an evaluation method for a seismic isolation system considering beyond the design conditions has been established. This paper reports an outline of the results of typical R&D programs and discusses the direction of R&D to improve NPP safety.

1. INTRODUCTION

Since Fukushima Daiichi Accident in 2011, strengthening of nuclear safety has been discussed and implemented in Japan. To strengthen nuclear safety with the concept of Defence-in-Depth, prevention of accidents is essential as well as mitigation of the consequences of accidents. Several research and development (R&D) programs have been conducted to improve the safety of nuclear power plants (NPP) under the government support program in Japan since 2011.

2. STRENGTHENING OF OVERALL NUCLEAR SAFETY

Based on the lessons learned from Fukushima Daiichi Accident, the needs for technology development were discussed to strengthen NPP safety. According to the concept of Defence in Depth, the middle and long term direction and the technology developments were surveyed. This survey excluded near term countermeasures for NPP. The results of the survey for PWR plants are summarized in Table 2.1. As typical results, the core cooling measures using steam generators and the seismic isolation system for nuclear installations are shown in the following chapters.

TABLE 2.1 Technology Developments Based on the Concept of Defence-in-Depth (PWR)

levels of DiD

ObjectiveDirection to strengthen

safety functionsTechnology developments

Level 1 Prevention of abnormal operation and failures

Earthquake-resistance - Seismic isolation system- Enhancement of seismic evaluation method for steam generator

Level 2 Control of abnormal operation and detection of failures

Maintain subcriticality to cold shutdown only with control rods

- New core internals with many reactor control clusters- Enhancement of CFD analysis for core internals

Level 3 Control of accident within the design

Diversity for reactor core cooling

- Enhancement of core cooling capability by steam generator

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basis - Air cooling system/equipmentLevel 4 Control of severe

plant conditionsCooling of melting core - In-vessel retention for large reactor

3. ADVANCES FOR CORE COOLING MEASURE USING SG SECONDARY-SIDE DEPRESSURIZATION

3.1. Background and R&D Activities

In light of the lessons learned from station blackout (SBO) accidents of Fukushima Daiichi reactors, it is important to line up various cooling measures for the reactor core and containment. A reliable alternative safety measure has been developed to cool the reactor core under a small break loss-of-coolant accident (SBLOCA) of PWR using SG secondary-side depressurization, as shown in Fig. 3.1. This safety measure adopts an early SG secondary-side depressurization to promote an early activation of accumulators (ACCs) and low-pressure injection (LPI) system.

For PWR plants, the SG secondary-side depressurization was investigated as a core cooling measure by Asaka et al. [1][2][3]. The study focused on the effectiveness of accident management (AM) by an operator action under total failure of a high-pressure injection system during SBLOCA, and the timing to open SG depressurization valves was relatively late such as 10 min after the event initiation. In contrast to the AM measure, the activation timing of this safety system adopts the period just after transmission of the Safety Injection (SI) signal.

There is no systematic validation database to assure the feasibility of the safety measure and therefore we planned to perform several tests using the ROSA/large-scale test facility (LSTF) [4]. The test parameters are (a) break size, (b) cooling capacity at each loop, (c) effect of dissolved nitrogen gas in ACC water and (d) onset timing of SG secondary-side depressurization.

The safety of an actual reactor should be checked by an analytical method that is validated using several appropriate databases. Therefore, the applicability of safety assessment code M-RELAP5 has been investigated using the test data. M-RELAP5 has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) to analyze SBLOCA and SBO of PWR for licensing safety calculations. MHI specifically selected the best-estimate thermal-hydraulics code, RELAP5-3D [5], as the base code of M-RELAP5, and modified it by incorporating conservative models such that the code is applicable to licensing safety calculations [6].

In this chapter, we show some typical results for SBLOCA obtained in this project and the applicability of M-RELAP5. Then the impact for safety advances from this project is described.

Hot LegCold Leg

Pressurizer

Accumulator

Low Pressure Injection Pump

Steam Generator

Small Break LOCA

Cross-over Leg

Core

Main Steam Relief Valve

Time

Prim

ary

Pre

ssur

e

Accumulator Injection Start

Small Break LOCA occur

SI signalSG depressurization initiation

Reactor trip

Low-pressure injection start

FIG. 3.1 Schematic of the safety measure against SBLOCA using SG secondary-side depressurization

3.2. Typical Results

The test conditions were selected to cover typical phenomena encountered during SBLOCA. The tests were performed at several break sizes, i.e. 2-in., 4-in., 6-in, 8-in. and 10-in, which show the beak diameter in an actual reactor. In this section, show the results for the 8-in. break test case where the highest PCT (Peak Clad Temperature) was recorded.

Figure 3.2 and Fig. 3.3 show results up to 500s for the 8-in SBLOCA test. The break valve opened at 0s and then primary pressure reduced shortly thereafter due to the loss of primary coolant. After that, scram signal and safety injection signal initiated at 13s and 16s, respectively. SG secondary pressure increased due to the main steam isolation valve closing upon receiving the scram signal and then dropped rapidly after SG depressurization valve opened. Primary system remained cooled by the SG secondary side and the primary

Table 4.1 　 　 Development process

Fig.4.1 Base isolation concept for NPP

PWR Plant BWR Plant

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pressure decreased along the SG secondary pressure decreasing thereafter. The primary pressure decreases to the actuating level of ACC at about 350s which is earlier than the case without secondary-side depressurization.

After the break initiation, the core inventory was decreasing due to the break flow and the core began to uncover and heat-up from about 280s as shown in Fig. 3.3. Coolant injection from ACC initiated when the primary pressure became lower than the ACC pressure. Clad surface temperature shifted to decline and the heat-up was ended when the core was completely recovered.

M-RELAP5 code predicted well the overall trend of thermal-hydraulic response observed in the test. On the other hand, the code overestimated the clad surface temperature including PCT. In the calculation, some amount of condensed water accumulated along SG U-tube hot leg side and maintained due to so-called counter current flow limitation (CCFL) along the SG U-tubes. This condensed water increased the static head along the U-tube and contributed to push down the core liquid level. As the result, the code evaluated the higher PCT than the test. From the point of safety evaluation, M-RELAP5 code can give a conservative evaluation.

3.3. Impact for Safety Advances

As described in Section 3.1, several sensitivity tests for break sizes, cooling capacity at each loop, effect of dissolved N2 gas in ACC water and onset timing of SG secondary-side depressurization were performed. It was confirmed that M-RELAP5 is applicable to those several test parameters and is revealed to keep conservative predictions.

The results of this project provide technical evidence that the AM measure can be activated without any concerns regarding several uncertainties. This contributes to enhance the reliability of the AM measure and is useful for refining the time-margin for operator action in future.

4. DEVELOPMENT OF SEISMIC ISOLATION SYSTEM FOR NUCLEAR FACILITIES

4.1. Purpose of developing the seismic isolation system

In order to secure the integrity of reactor buildings against huge earthquakes in the future, base isolation systems are effective approaches and could also realize the standard design not to depend on site conditions.

At the aim of adopting a seismic isolation system to nuclear facilities, this project studied the following items; (1) obtaining highly aseismic performance by installing base-isolation, (2) grasping the ultimate strength of isolator based on the full-scale breaking tests, and (3) establishing the evaluation of “a residual risk” for phenomena exceeding the design conditions.

4.2. Contents of developing the seismic isolation system

In this project, the ground motion for seismic study used an artificial wave enveloping general Japanese NPP sites, which of the maximum acceleration was 800 cm/sec2 and the maximum velocity was 200 cm/s. The study isolators adopted a lead rubber bearing (LRB) of 1600 mm diameter, which was one of the largest scale in Japan. As shown in Table 4.1, this project studied characteristic tests of full-scale isolators, design seismic evaluation of base-isolated building, verification tests of crossover piping between base-isolated and non-base-isolated buildings, and residual risk evaluation against huge earthquakes in the future.

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FIG. 3.2 Comparison of primary and secondary pressures between test and M-RELAP5

FIG. 3.3 Comparison of PCT between test and M-RELAP5

: FEM Model: Beam Model

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4.3. Overview of development results

It was confirmed that the seismic isolation system is applicable to actual NPP’s facilities such as the base-isolated buildings, and this project completed infrastructure improvement for deployment to actual NPP. The development results are overviewed as follows.(a) Characteristic tests of full-scale isolators

As shown in Fig. 4.2, static breaking tests using the 1600 mm-dia. LRBs were performed for the first time in the world. These tests determined various characteristics such as the breaking capacity, and this project created a design restore model combining horizontal and vertical, the schematic diagram for full-scale breaking capacity and so forth. (b) Design seismic evaluation of base-isolated building

Actual isolator specifications were decided by seismic design after studying ground motion, and seismic integrity was investigated based on the beam model and 3D-FEM model of base-isolated building as shown in Fig. 4.3. As a result, this project conducted a combined evaluation between horizontal and vertical for isolator, the rational design of isolated pedestal and so forth based on the testing results conducted by this project.(c) Verification tests of the crossover piping between base-isolated and non-base-isolated buildings

The seismic relative displacement of crossover piping between these buildings was absorbed by routing design. This project verified the integrity of crossover piping by shaking tests using 1/10 scale routing as shown in Fig.4.4 and the static-loading repeated tests using 1/4 scale piping, and reflected these test results in the crossover piping design. (d) Residual risk evaluation

As the PRA method, this project studied the fragilities of base-isolated buildings based on various failure modes such as the failure probability of seismic isolators as shown in Fig. 4.5, and evaluated the validity of these fragilities.(e) Conclusion and further work

Developing the base of the seismic isolation system in this project expanded the flexibility of the aseismic design to NPP’s facilities. In future, further work is required to improve high damping isolators in preparation for further huge earthquakes, and examine fail-safe devices against earthquakes beyond the design basis.

Schematic diagram for breaking capacityPhoto under breakingBreaking test equipment(Max horizontal load 25.1×103kN)

Fig. 4.2 Breaking tests of full-scale isolators

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5. SUMMARY

Based on the lessons learned from Fukushima Daiichi Accident, several technology developments to strengthen NPP safety have been completed in March, 2017. The results will be considered for continuous improvement of NPP safety.

REFERENCES

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