Darlington NGS reactor operation with flux tilt impact on neutron overpower protection
Description
This paper discusses the analysis performed to assess the Neutron Overpower (NOP) system trip coverage for Darlington NGS operation with initial side-to-side and bottom-to-top flux/power tilts. The required trip setpoints of the design basis set of flux shapes were calculated using the Monte Carlo statistical method. Initial flux tilt values of up to 20% were considered in the analyses. The analysis shows that the NOP system with the nominal trip setpoint of 126.0 percent provides adequate protection as long as the NOP detector signals are not re-calibrated after the occurrence of the tilt, or are re-calibrated using the CPPF (channel power peaking factor) corresponding to the tilted flux shape. The analysis also shows that for the case of re-calibrating the detector signals using the CPPF corresponding to the nominal un-tilted flux shape, the NOP system provides the required coverage for tilts of up to 6.8 percent. (author). 9 refs., 2 figs
Files
29000329.pdf
Files
(237.9 kB)
| Name | Size | Download all |
|---|---|---|
|
md5:3e37247e38c6da1d7b9dce12646c19ab
|
237.9 kB | Preview Download |
Additional details
Additional titles
- Augmented title (English)
- Nuclear Generating Station
Publishing Information
- Imprint Title
- Proceedings of the 19. Canadian Nuclear Society simulation symposium
- Imprint Pagination
- [350 p.].
- Journal Page Range
- [9 p.].
- Report number
- INIS-CA--0010
Conference
- Title
- 19. CNS simulation symposium.
- Dates
- 16-17 Oct 1995.
- Place
- Hamilton, ON (Canada).
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 29000329
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CALIBRATION; COMPUTERIZED CONTROL SYSTEMS; COMPUTERIZED SIMULATION; DARLINGTON-1 REACTOR; MONTE CARLO METHOD; NEUTRON FLUX TILTING; REACTOR OPERATION; REACTOR PROTECTION SYSTEMS; REACTOR SAFETY FUSES; SCRAM
- Descriptors DEC
- CALCULATION METHODS; CANDU TYPE REACTORS; CONTROL SYSTEMS; ENGINEERED SAFETY SYSTEMS; HEAVY WATER COOLED REACTORS; HEAVY WATER MODERATED REACTORS; NATURAL URANIUM REACTORS; ON-LINE CONTROL SYSTEMS; ON-LINE SYSTEMS; OPERATION; PHWR TYPE REACTORS; POWER REACTORS; PRESSURE TUBE REACTORS; REACTOR COMPONENTS; REACTOR SHUTDOWN; REACTORS; SHUTDOWN; SIMULATION; THERMAL REACTORS