IR 05000219/1986036
| ML20215G148 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/12/1986 |
| From: | Pullani S, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215G145 | List: |
| References | |
| 50-219-86-36, NUDOCS 8612240341 | |
| Download: ML20215G148 (6) | |
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I U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-219/86-36 Docket No. 50-219 License No. DPR-16 Licensee: GPU Nuclear Corporation P. O. Box 388 Forked River, NJ 08731 Facility Name: Oyster Creek Nuclear Generating Station Inspection At:
Forked River, New Jersey Inspection Conducted: November 17-21, 1986 Inspectors:
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,/P.C.Wey,Reacto Engineer date Approved by:
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'S.V.Pu/lani, ing Chief, date Test Prog ection, 08, DRS Inspection Summary:
Inspection on November 17-21, 1986 (Inspection Report No. 50-219/86-36)
Areas Inspected:
Cycle 11 Startup Physics Testing and Cycle 10 Fuel Failure Followup.
Results:
No violations were identified.
NOTE:
For acronyms not identified, refer to NUREG-0544, " Handbook of Acronyms and Initialisms".
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DETAILS 1.
Persons Contacted GPU Nuclear Corporation
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J. D. Dougher, Nuclear Engineer, Oyster Creek Fuel Project
- W. J. Emrich, Jr., Core Engineer
- R. F. Fenti, QA Mod / Ops Manager
- P. B. Fiedler, Vice President and Director, Qyster Creek.
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J. R. Molnar, Core Engineering Manager W. Pelenski, Manager, Computer Applications
- M. S. Radvansky, Manager, Technical Functions, Oyster Creek
- A. H. Rone, Plant Operations Engineering Manager i.
- W. Smith, Plant Engineering Director
- J. L. Sullivan, Jr., Plant Operations Director i,-
- W Bateman, Senior Resident Inspector
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- R. W. Borchardt, Senior Resident Inspector, Hope Creek
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- J. Wechselberger, Resident Inspector S
The inspector also contacted other licensee personnel during the inspection.
- Denotes those present at the exit meeting on November 21, 1986.
l 2.
Cycle 11 Reload Safety Evaluation and Core Verification j.
The Cycle 11 core contains 188 fresh fuel assemblies (GE Types P8DRB265H, P80RB299ZA and P80RB299Z) and 372 irradiated fuel assemblies (Exxon Type l
VB and GE Types. P80RB239 and P80RB265H).
The reload design, safety
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analyses and the required Technical Specifications (TS) change were i
submitted to the NRC for review.
This reload submittal was found
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acceptable (Letter from J. N. Donohew, Jr. (NRC) to P.B. Fiedler (GPU-Oyster Creek), dated October 27,1986).
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While reviewing the licensee's loading pattern, the inspector noted that the Cycle 11 loading is a Control Cell Core (CCC) configuration.
In the CCC, control rod movement is limited to a fixed group of control rods.
High burnup (low reactivity) Exxon Type VB fuel bundles are loaded around each of these control rods. This loading pattern is consistent with the j
licensee's safety evaluation report (TDR-768), " Reload Information and l
Safety Evaluation Report for Oyster Creek Cycle 11," Revision 1. The NRR's Safety Evaluation (TS Amendment No. 111 review), however, called the load-ing pattern a " conventional scatter loading pattern".
This matter was discussed with the NRR project manager and technical reviewer. Apparent-ly, a different terminology with regard to the loading pattern was used but the conclusion of the licensee's reload submittal remains acceptable.
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The Core Engineering group performed core post-alteration inspection and verification in accordance with procedure 1001.24, Revision 8.
The inspector selectively compared the core verification videotapes and independently verified that the core loading agreed with the intended core loading pattern. No ir. adequacies were identified.
3.
Cycle 11 Startup Testir.;
The licensee plans to start the Cycle 11 at end of November,1986 after a seven-month maintenance and refueling outage.
During this inspection period, November 17-21, 1986, the unit was still in preparation for startup.
The inspector reviewed selected test programs and available results to verify the following, wl:ere applicable:
Procedures were provided with the detailed stepwise instructions,
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including Precautions, Limitations, and Acceptance Criteria;
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Technical content of the procedures was sufficient to result in satisfactory calibration and test; Provisions for recovering from anomalous conditions were provided;
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Methods and calculations were clearly specified and tests were
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conducted accordingly; Review, approval, and documentation of the results were in
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accordance with the requirements of the TS and the licensee's administrative controls.
The following tests were reviewed:
3.1 Control Rod Drive Scram Time Test The control rod drive (CRD) scram time test was performed in accordance with Procedure 617.4.003, Control Rod Scram Insertion Time Test and Valve IST Test, Revision 8.
The inspector verified by review of the test data obtained on October 12, 1986 that the average scram times at various insertion levels and mean 90% insertion times were all within the TS limits.
No unacceptable conditions were identified.
3.2 Shutdown Margin (SDM)
The SDH Measurement was performed in accordance with Procedure 1001.27, Shutdown Margin Measurement Test, Revision 11.
The test
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was performed on September 22, 1986, with a moderator temperature of 86*F.
A shutdown margin of at least 2.7% AK was demonstrated with the strongest rod (46-23) fully withdrawn. This result meets the TS SDM requirement (0.25% AK + R) with an additional margin of 2.1% AK.
A subsequent SDM demenstration was performed to verify that the core remained in subcritical conditions during withdrawal of the control rod diagonally adjacent to any specified fully withdrawn rod. This test was performed in accordance with Procedure 1001.26, Shutdown Margin Demonstration, Revision 5, on September 24, 1986, without the reactor going critical and thus demonstrated the required SDM.
No unacceptable conditions were identified.
4.
Fuel Performance Followup The licensee performed fuel inspection during the current Cycle 10/11 refueling outage, and identified 47 failed fuel assemblies. The causes of fuel failure were described previcusly in the NRC inspection report 50-219/86-34.
During this inspection peried, the inspector reviewed the adequacy of the licensee's corrective actions as follows:
Cycle 11 Core Redesign The redesigned core consists of 188 fresh GE fuel assemblies instead of 164 fresh fuel assemblies of the original design.
This redesign was required to accommodate a new loading scheme.
This includes:
(1) all failed fuel assemblies (37 of which were originally scheduled for reload-ing) and 3 non-failed symmetric partners to a failed once burnt fuel assembly. All of these elements were removed and were not to be reused in the Cycle 11 operation; (ii) all incipients (fuel assemblies which were located at symmetric locations to the failed assemblies) were loaded in core peripheral locations to reduce fuel duty.
The redesigned Cycle 11 core along with associated reload submittal were reviewed by the NRR, and were found acceptable (Section 2).
Operation Procedures / Fuel Performance Monitoring Program The inspector reviewed a draft copy of revised plant procedure 1001.22,
" Power Distribution Control During Power Operation".
This procedure establishes guidelines for the operation and monitoring of the reactor core during power operation in the RUN mode.
The inspector noted that more guidance on PCIOMR monitoring has been incorporated in the revised procedure.
In addition, a specific responsibility was delegated to the Core Engineer to advise the operation staff with respect to power distri-bution control during power operatio i
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Plant Procedure 1001.33, " Core Follow", was revised to incorporate a fuel performance monitoring program. A computer code "0FFGAS" was written to analyze fission product data from the offgas release.
This information can be used to assess the possibility of fuel failure.
Following all startups or significant power maneuvers, this procedure (1001.33) also requires that a critique be initiated to review any sig-nificant event or unexpected occurrences.
Extended Load Line Limit Analysis Cycle 10 was the first reload cycle which used mixed vendor fuel, i.e. GE P8X8R and Exxon Type VB fuel. The safety analysis performed to support Cycle 10 operation was based on the GE supplied rod line which required more rod manipulation at high power to obtain and maintain the target rod pattern. These rod manipulations sometimes resulted in large local power increases when high rod worth rods were withdrawn at high power. The lic-ensee has since acquired an extended load line limit analysis (ELLLA) from GE for the upcoming Cycle 11 operation. The ELLLA and the associated re-vision of the APRM Scram and Rod Block Lines provide more operating flexi-bility which allows plant operation at high power by adjusting the recir-culation pump speed rather than rod position.
These items were reviewed ano approved by the NRR, and have been incorporated in the appropriate plant operating procedures.
Power Shape Monitoring System (PSMS) Testing The licensee's core monitoring code, PSMS, Revision 1, was used during Cycle 10 operation. Although this code provided adequate monitoring for the TS required parameters, such as MCPR, MLHGR, and MAPLHGR, it contained errors and shortcomings in monitoring PCIOMR envelopes which were recommended by the fuel vendor to prevent pellet / clad interaction failures.
PSMS problems identified during Cycle 10 operation were described in the NRC inspection report 50-219/86-34.
The licensee plans to use PSMS, Revision 2, for the Cycle 11 operation.
The software package includes an official EPRI PSMS, Revision 2,
Expert-Ease System developed PCIOMR sub program, and many of the licensee's own enhancements and modification sub programs.
The PSMS Revision 2 has been tested extensively in both static and dynamic modes.
During the inspection period (November 17-21, 1986), the final readiness of PSMS Revision 2 for Cycle 11 operation was being reviewed jointly by the Fuel Projects, Core Engineering and Computer Applications groups.
Discussion with the cognizant nuclear fuel and core engineers both indicated that no open items would hinder the PSMS Revision 2 to function properly. The inspector reviewed the PSMS integrated system test results and noted its completio ;
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The inspector also noted that additional training for the Core Engineer
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and Plant: Operations staff in PSMS 'and PCIOMR guidelines were conducted-during the refueling outage.
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Conclusions The licensee is acting prudent 1y' as ' discussed above to correct 'the occurrence of fuel failures in Cycle _10 operation. The implementation of these steps would minimize the recurrence of ' PCI related. failures and closely monitor the fuel performance during Cycle 11 operation.
In addition,'the licensee stated that starting with Cycle 12, barrier fuel which is designed to reduce failures due to Pellet / Clad Interaction will be utilized for future cycle operation.
The. licensee's corrective' action implementation is therefore considered adequate.
5.
Independent Calculations / Verifications The inspector performed independent calculations / verifications _of Cycle 11 startup physics testing related activities. These included the following:
Core loading verification as described in Section 2.
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Independent verification of the licensee's SDM calculation and it's
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compliance with the TS requirement.
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QA/QC Interface The licensee QA group performed surveillance coverage on the Core Loading Verification and SDM Testing.
The activities included the direct field observation and verification of procedure compliance.
Review of these surveillance reports indicated that QA's surveillance coverage was thorough and comprehensive.
7.
Management Meeting Licensee management was informed of the purpose and scope of the inspec-tion at the entrance interview.
The findings of the inspection were periodically discussed and were summarized at the conclusion of the inspection on November 21, 1985.
Attendees at the exit interview are denoted in paragraph 1.
At no time during this inspection was written material provided to the licensee. Based on the NRC Region I review of this report and discussions held with the licensee representatives at the exit, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.