4.0 Summary of Major DCISC Review Topics, 17th Annual Report - July 1, 2006 thru June 30, 2007
4.12 Nuclear Fuel Performance
4.12.1 Overview and Previous Activities
The DCISC has been following performance of nuclear fuel and fuel-related matters at DCPP since its beginning in 1990. The Committee receives regular reports on nuclear fuel performance and any problems from PG&E both in Fact-finding and public meetings and as input to the annual report. DCISC follows-up on problems and activities in its Fact-finding meetings at DCPP and PG&E Headquarters.
DCPP fuel reliability is the most important fuel attribute monitored during operation. It is important to assure that the fuel integrity is preserved to avoid fission product leakage into the reactor coolant system (RCS) and ultimately into RCS cleanup and support systems resulting in increased personnel dose, radioactive waste and potential off-site releases.
Since DCISC was formed in 1990, fuel reliability had been excellent until November 1994 when Unit 2 fuel began to show signs of leakage and experienced localized fuel damage. Unit 2 has had several additional fuel leaks since then. Leakage is measured by the amount of radioactivity in RCS samples, with a current goal of less than 5.0 x 10-4 microCuries (Ci) of Iodine-131 per gram of coolant. The following depicts the RCS radioactivity trend for a five-year period:
Reactor Coolant System (RCS) Radioactivity (microCuries/gram of coolant Iodine-131)
| Period | Goal | Unit 1 Actual (Ci/gm) | Unit 2 Actual (Ci/gm) |
|---|---|---|---|
| 02-03 | 5.0 x 10-4 | 1.0 x 10-6 | 1.48 x 10-3 |
| 03-04 | 5.0 x 10-4 | 1.0 x 10-6 | 1.0 x 10-6 |
| 04-05 | 5.0 x 10-4 | 1.0 x 10-6 | 1.0 x 10-6 |
| 05-06 | 5.0 x 10-4 | 1.0 x 10-6 | 1.0 x 10-6 |
| 06-07* | 5.0 x 10-4 | 1.0 x 10-6 | 1.0 x 10-6 |
* Through June 2007
In addition to regular fuel performance updates, DCISC has investigated the following fuel-related topics:
- Spent Fuel Pool (SFP) Safety Issues
- Boraflex Degradation
- Stuck Control Rods
- Axial Flux Axial Offset Anomaly
- ATWS Moderator Core Temperature Coefficient
- Unit 2 Fuel Failures
- Top Nozzle Spring Bolting Failures
- Spent Fuel Storage Status & ISFSI
- Nuclear Fuel Gap Re-Opening
- Extended Fuel Cycle
- Reload Safety Evaluation and Process
- North Anna Fuel Break & Implications for DCPP
- Movable In Core Detector System Operator Burdens/Workarounds
- RCCA Tip Swelling
- Unit 2 Baffle Jetting
- Reactor Coolant System pH Program
- Effect of Zinc in RCS
- 2R11 - Fuel Performance Results
- Status of Spent Fuel Storage Backup Options
- Spent Fuel Pool Thermal Management
- December 23, 2004 Spent Fuel Partial Water Loss Event
The DCISC has concluded in the past that PG&E appeared to be handling fuel or fuel-related problems appropriately. The DCPP Unit 1 core has been reliable and clean; however, Unit 2 had experienced four instances of fuel leakage due to manufacturing defects, debris and some problems with baffle jetting; however, the last three fuel cycles have been clean with no failures. PG&E appears to be handling those problems appropriately.
4.12.2 Current Period Activities
The DCISC reviewed the following nuclear fuel topics during this reporting period:
- Use of Best Estimate Analyzer for Core Operations - Nuclear (BEACON)
- New Control Rod Worth Test
Use of Best Estimate Analyzer for Core Operations - Nuclear (BEACON)
The DCISC Fact-finding Team met with Shane Guess, an Engineer in the Reactor Engineering Group, at the October 25-26, 2006 Fact-finding Meeting (Volume II, Exhibit D.3, Section 3.2) to discuss BEACON (Best Estimate Analyzer for Core operations – Nuclear).
BEACON is a Westinghouse-developed analytical tool for predicting the power distribution in a reactor core. It is significantly more accurate than previous modeling tools because it includes a detailed model for fuel burn up that predicts with high accuracy the changing isotopic and elemental composition of the fuel due to fission and neutron capture reactions. Confirming that the reactor core power distribution remains inside design limits is important for safety, because during a design basis accident initial fuel damage would occur at the locations in the core with higher than average power. DCPP uses BEACON to augment the functional capability of the incore flux mapping system above 25 full power for the purpose of Technical Specification required power distribution surveillances.
DCPP uses BEACON for the following specific purposes:
- Reactor Core Power Distribution Measurement vs. Flux Map – a calibrated BEACON power distribution monitoring system calculates peaking factors (required by Technical Specifications every 31 Effective Full Power Days [EFPD]) without the need for an incore flux map.
- Reactor Ramp Planning: BEACON’s load swing calculation allows for the Reactor Engineer to calculate predicted volumes of boric acid and primary water during a power ramp and also to predict delta-I behavior for Operations.
- Other: Reactivity Balances, Reactor Shutdown Margin calculation, Estimated critical Condition calculation, and Single-Point calibration.
BEACON is initially calibrated using incore flux measurements. These are obtained from the plant movable Incore Detector System. It includes an online three dimensional nodal model that is continuously updated to reflect the current reactor operating conditions. The power distribution calculation used in core monitoring is updated with core exit thermocouple measurements each minute. BEACON utilizes a surface spline fitting technique to interpolate/extrapolate the incore flux detector measurements and thermocouple measurements to unmeasured fuel assemblies. The neutronics calculations used by BEACON solve the classic diffusion equation which describes the nuclear behavior of the reactor core. Calibration with the incore detectors is required every 180 EFPD.
Being used for safety-related applications, BEACON falls under the Software Quality Assurance Program (SQAP). A cycle-specific version is verified each restart following a refueling outage. DCPP considers BEACON to be very accurate; however, it has exhibited a slight delay in reaction to boration and slight additional amounts of boration than predicted. This is likely caused by an approximate 13 core bypass of injected boric acid due to injection and letdown point geometry.
The FFT reviewed a July 1, 2006 100 - 90 - 100 Power Unit 2 Cycle 14 Ramp Plan created by DCPP Reactor Engineering using BEACON. The plant was being ramped down for Digital Feedwater System testing at 90 power. The plan had been requested by Operations. The plan stated clearly that it was to be used as a guidance tool, not to be considered operating limits nor used a script. The plan was used in a pre-job Operations brief prior to beginning the ramp down operation. The plan provided a suggested timeline, ramp rates, process hold points, control rod operation, and predicted boric acid additions and primary water additions. Included were timeline graphs. The Ramp Plan appeared to be a useful tool for Operations.
DCPP’s use of the Best Estimate Analyzer for Core Operations – Nuclear (BEACON) appears to be an effective and safe alternative for reducing the need to use of the Movable Incore Detector System (MIDS) to perform the Technical Specification required reactor core power distribution measurement every 31 Effective Full Power Days. Use of BEACON in the industry and at DCPP has been approved by the Nuclear Regulatory Commission.
New Control Rod Worth Test
The DCISC Fact-finding Team met with Ken Kargol, Engineer in Reactor Engineering, at the October 25-26, 2006 Fact-finding Meeting (Volume II, Exhibit D.3, Section 3.6) to discuss the new Control Rod Worth Test.
Technical Specification-required control rod worth, subcritical negative reactivity, and Moderator Temperature Coefficient (MTC) testing is performed during each start-up from a Refueling Outage in which fuel changes are routinely made, i.e., approximately one-third of the core is replaced with new fuel. Typically, testing has been performed at hold points during power ascension requiring approximately 24 hours of outage time. A new process developed by Westinghouse shortens the required outage time to approximately 7.5 hours or less for an outage savings of about 17.5 hours. Control rod drop testing is still required.
The new process is Subcritical Physics Testing (SPT) based on Westinghouse Subcritical Rod Worth Measurement (SRWM) methodology. SPT is performed when the reactor is subcritical rather than at power in conventional testing. SPT does not require removing a protection system input from service, saving time and avoiding a degraded reactor protection system.
DCPP and Westinghouse considered test results to be excellent and sufficient to provide enough confidence to allow power operation to commence. DCPP reported excellent agreement between the predicted and measured Inverse Count Rate Ratio (ICRR) values.
DCPP’s new reactor physics testing procedure accurately measures core parameters, saves over 17 hours of outage time, and permits measuring Technical Specification-required reactor core parameters without removing any reactor protection inputs, resulting in a safer configuration.
Fuel Performance for Unit 1 and Unit 2 Including Review of Fuel Leakage Issues
The DCISC received an update on the DCPP nuclear fuel performance at the June 13-14, 2007 Public Meeting (Volume II, Exhibit B.9, Page B.9-26).
DCPP, U-1 has had a total of 1,349 fuel assemblies loaded in its core and U-2 a total of 1,289 fuel assemblies. U-1 has not experienced a fuel defect since 1991. U-2 is currently estimated to have one leaking fuel rod, which was identified approximately 200 days into its current operating cycle. DCPP has the capability to remove and replace the pin from a fuel assembly with a defective rod on site and to replace a defective rod within an assembly with a stainless steel rod. The defective pins are stored in a dedicated canister and he stated that DCPP has never sent spent fuel off site for further testing.
DCPP’s comprehensive Fuel Management Program includes the following key preventive features:
- Cooperative industry effort to oversee the fuel supplier’s quality control program.
- Onsite detailed fuel inspection.
- Careful control of fuel handling by Westinghouse.
- FME Control Program for all plant systems in contact with fuel.
- Strict chemistry control in the Reactor Coolant System to prevent corrosion.
- Conservative limits on core power distribution to prevent hot spots in the core.
- Fuel design features which protect against potential debris fretting of fuel cladding,
- usually occurring in the bottom of the fuel assembly, including adding a hardened oxide coating, beginning at the bottom, solid section of the rods.
- Foreign object search and retrieval during outages using an articulated arm and camera with capability to go below the core plate to look for debris.
Daily chemistry sampling is being done on U-2 to determine the nature of the suspected fuel defect. To date, because there have been no rapid swings in U-2 power level, it is undetermined if the defect is due to debris or manufacturing and whether it is within first or second generation fuel.
Key leak mitigation features of the Fuel Management Program include:
- Imposition of tighter restrictions on how fast power can be increased.
- Inspection of all fuel during refueling outages, to ensure leaking fuel is not reused (and to identify the cause of failure).
- Repair or replacement of leaking fuel.
- All fuel will be inspected with special test equipment by fuel sipping or canister sipping, another Westinghouse technology.
4.12.3 Conclusions and Recommendations
- Conclusion:
- DCPP has an effective fuel management program as evidenced by trouble-free operating cycles between refueling outages. DCPP has had no Unit 1 fuel failures in many years and no Unit 2 fuel failure problems for almost three cycles (except for a recent Unit 2 single fuel rod leak indication) due to an aggressive program of prevention and mitigation techniques. DCPP appears to be resolving its nuclear fuel issues effectively.
- Recommendations:
- None