scholarly journals The Optimization of Radiation Protection in the Design of the High Temperature Reactor-Pebble-Bed Module

2017 ◽  
Vol 2017 ◽  
pp. 1-15
Author(s):  
Sida Sun ◽  
Hong Li ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Radiation Protection ◽  
Nuclear Power ◽  
Pebble Bed ◽  
Pressurized Water ◽  
Pebble Bed Reactors ◽  
Dose Rates ◽  
Optimization Routine ◽  
High Temperature Reactor ◽  
Water Reactors

The optimization of radiation protection is an important task in both the design and operation of a nuclear power plant. Although this topic has been considerably investigated for pressurized water reactors, there are very few public reports on it for pebble-bed reactors. This paper proposes a routine that jointly optimizes the system design and radiation protection of High Temperature Reactor-Pebble-Bed Module (HTR-PM) towards the As Low As Reasonably Achievable (ALARA) principle. A systematic framework is also established for the optimization of radiation protection for pebble-bed reactors. Typical calculations for the radiation protection of radioactivity-related systems are presented to quantitatively evaluate the efficiency of the optimization routine, which achieve 23.3%~90.6% reduction of either dose rate or shielding or both of them. The annual collective doses of different systems are reduced through iterative optimization of the dose rates, designs, maintenance procedures, and work durations and compared against the previous estimates. The comparison demonstrates that the annual collective dose of HTR-PM is reduced from 0.490 man-Sv/a before optimization to 0.445 man-Sv/a after optimization, which complies with the requirements of the Chinese regulatory guide and proves the effectiveness of the proposed routine and framework.

Assessment of Control Room Radiological Habitability of High-Temperature Reactor Pebble-Bed Module in Shidao Bay Multi-Reactor Nuclear Power Site

2018 ◽  
Author(s):  
Xinpeng Li ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Source Term ◽  
Meteorological Data ◽  
Dose Calculation ◽  
Dose Assessment ◽  
Specific Information ◽  
Control Room ◽  
Pebble Bed ◽  
High Temperature Reactor

The control room radiological habitability (CRRH) is important for staff safety in a nuclear power plant, which is also a licensing requirement of the High-temperature Reactor Pebble-bed Module (HTR-PM) in China. Meanwhile, the complexity of the dose assessment increases for the multi-reactor site, which put forward higher requirements for building layout. The CRRH is investigated comprehensively for the multi-reactor site at Shidao Bay in this study. For a large-break loss of coolant accident of HTR-PM and CAP1000 in Shidao Bay nuclear power site, this study estimates doses of body, thyroid and skin due to three exposure pathways using NRC-recommended ARCON96 and dose calculation method in RG 1.195. To perform a realistic evaluation, the latest design and site-specific information are utilized as the input parameters, including the unique accidental source term of HTR-PM and the RG1.183-recommended source term of CAP1000, the release and ventilation parameters, the final layout and the meteorological data in a whole year. The evaluation results demonstrate that the individual dose level of staff in the control room is far below the requirement of the regulatory guide, which guarantees the CRRH of HTR-PM. The contribution of primary radionuclides suggests that tellurium and iodine are primary contributors of the inhalation dose of body and thyroid, which is worthy of paying particular attention to the CRRH design in HTR-PM.

Radiation Shielding Design of High Temperature Reactor Pebble-Bed Module (HTR-PM)

2016 ◽  
Author(s):  
Sida Sun ◽  
Sheng Fang ◽  
Hong Li
Keyword(s):  
High Temperature ◽  
Dose Calculation ◽  
Radiation Shielding ◽  
Source Analysis ◽  
Pebble Bed ◽  
Pebble Bed Reactors ◽  
Protection Strategies ◽  
Reduction Methods ◽  
High Temperature Reactor ◽  
Shielding Design

Radiation safety is an important concern in the design and licensing of the 200MWe High Temperature Reactor Pebble-bed Module (HTR-PM) demonstration power plant in China. To meet the requirement of the regulatory, various radiation protection strategies and methods are applied in the design process of systems and components of HTR-PM. In this study, the radiation shielding design of HTR-PM is reviewed, which includes the radiation source analysis, in-house dose calculation tool, shielding and dose reduction methods used for primary systems. The underlying conservative assumption is also discussed for correctly evaluating the dose calculation result. This summary provides a relatively systematic review of the radiation shielding methods in the design phase of HTR-PM, which may provide useful information and experiences for the radiation shielding design of future pebble-bed reactors.

Preferred core conceptual design of pebble bed advanced high temperature reactor

2021 ◽  
Vol 151 ◽  
pp. 107983
Author(s):  
Lianjie Wang ◽  
Wei Sun ◽  
Bangyang Xia ◽  
Yang Zou ◽  
Rui Yan
Keyword(s):  
High Temperature ◽  
Conceptual Design ◽  
Pebble Bed ◽  
High Temperature Reactor

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

The Modelling and Coupling Methodology of ANSYS CFX Using Porous Media for PB-AHTR

2013 ◽  
Author(s):  
Linsen Li ◽  
Haomin Yuan ◽  
Kan Wang
Keyword(s):  
Porous Media ◽  
Monte Carlo ◽  
Steady State ◽  
High Temperature ◽  
Monte Carlo Code ◽  
First Principle ◽  
Pebble Bed ◽  
Coupled Calculation ◽  
Ansys Cfx ◽  
High Temperature Reactor

This paper introduces a first-principle steady-state coupling methodology using the Monte Carlo Code RMC and the CFD code CFX which can be used for the analysis of small and medium reactors. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. A Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core is modeled using this method. The porous media is used in the CFX model to simulate the pebble bed structure in PB-AHTR. This research concludes that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Comparison of US and UK Practices and Philosophies for Nuclear Plant Control Room Design

1982 ◽  
Vol 26 (7) ◽  
pp. 659-663
Author(s):  
Les Ainsworth
Keyword(s):  
Nuclear Power ◽  
Plant Design ◽  
Control Room ◽  
Pressurized Water ◽  
Basic Task ◽  
Room Design ◽  
Power Plant Control ◽  
Gas Cooling ◽  
Water Reactors ◽  
Plant Control

The American and British nuclear power programmes have in the past taken divergent routes, with the Americans choosing pressurized water reactors, whilst the British have opted for gas cooling. Although the technology and plant design for these two systems encompasses many fundamental differences, some of which have ramifications for the controllers, the basic task of monitoring and controlling a reactor holds many similarities in both countries. It is therefore instructive to compare and contrast the approaches which have been taken to human factors in nuclear power plant control room design on both sides of the Atlantic.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Long Term Stability Evaluation of Stress Improvement Effect on Ni-Based Alloy Welds by Various Residual Stress Improvement Methods

2021 ◽  
Vol 1016 ◽  
pp. 819-825
Author(s):  
Li Na Yu ◽  
Kazuyoshi Saida ◽  
Masahito Mochizuki ◽  
Kazutoshi Nishimoto ◽  
Naoki Chigusa
Keyword(s):  
Residual Stress ◽  
High Temperature ◽  
Stress Relaxation ◽  
Compressive Residual Stress ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Residual Stress Relaxation ◽  
Temperature Environment ◽  
Water Reactors ◽  
High Temperature Environment

Stress corrosion cracking (SCC) is one of serious aging degradation problems for the Alloy 600 components of pressurized water reactors (PWRs). In order to prevent SCC, various methods such as water jet peening (WJP), laser peening (LP), surface polishing have been used to introduce compressive stresses at the surfaces of the PWR components. However, it has been reported that such compressive residual stress introduced by these methods might be relaxed during the practical operation, because of high temperature environment. In this study, the hardness reduction behavior of the Alloy 600 processed by LP, Buff and WJP in the thermal aging process has been investigated to estimate the stability of the residual stress improving effect by each method, based on the fact that there is a correlation between the compressive residual stress relaxation and the decrease of hardness. The behavior of the residual stress relaxation in the processed materials in the high temperature environment has been discussed with kinetic analysis.

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