Abstract: An advanced small-scale integrated fully natural-circulation pressurized water reactor (PWR), hereafter referred to the heating reactor, is intended for nuclear heating, industrial steam supply, and distributed energy applications. Owing to its integral configuration, full natural-circulation primary system, and passive engineered safety features, its thermal-hydraulic behavior and accident mitigation process differ from those of conventional loop-type PWRs. To support the integral performance verification of this reactor concept, an overall performance test facility was designed and constructed using the hierarchical two-tiered scaling (H2TS) methodology. The facility maintains similarity to the prototype in terms of the main geometric arrangement, system configuration, engineered safety features, and control logic, and is capable of representing key thermal-hydraulic phenomena of the heating reactor under typical transient and accident conditions, including primary-side natural circulation, pressure interaction between the reactor pressure vessel and containment, passive depressurization, gravity-driven injection, and secondary-side passive residual heat removal. Overall performance tests were conducted for two representative design-basis accident scenarios: a pressurizer-top break loss-of-coolant accident (LOCA) and a loss of tertiary heat sink non-LOCA transient. For the pressurizer-top break LOCA, the test conservatively assumes failure of the passive secondary-side residual heat removal system and immediate isolation of the secondary side after break initiation. The transient reproduces the main accident phenomena, including rapid primary depressurization, containment pressurization and steam condensation, actuation of the automatic depressurization system, establishment of direct vessel injection, and long-term passive recirculation cooling. The results show that the passive emergency core cooling (designated as JNG) system is actuated in accordance with the designed logic. The automatic depressurization and direct vessel injection processes are established sequentially, and the collapsed water level in the reactor pressure vessel remains above the top of the active core throughout the transient, indicating that core submergence is maintained under the tested conditions. For the loss of tertiary heat sink non-LOCA transient, the normal heat removal path is isolated, leading to an initial rise in the temperature and pressure of the primary and secondary systems. After the high pressurizer pressure signal is reached, reactor trip is initiated and the core power is reduced according to the decay heat curve. The passive secondary-side residual heat removal (designated as NCR) system is subsequently put into operation and established a natural-circulation heat removal path through the intermediate isolation loop and the external water pool. The test results show that the pressurizer pressure decreases after reaching its peak value, the core inlet and outlet temperatures enter a stable cooling stage after short-term redistribution, and the heat removal power of the NCR system gradually matches the simulated core decay heat. Overall, the experiments verify the response characteristics and accident mitigation logic of the engineered safety systems of the heating reactor, and provide overall performance experimental data for safety analysis code validation and safety review of this reactor type.