IR 05000269/1987023
| ML16161A842 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/07/1987 |
| From: | Bernhard R, Jape F, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16161A841 | List: |
| References | |
| 50-269-87-23, 50-270-87-23, 50-287-87-23, NUDOCS 8707230379 | |
| Download: ML16161A842 (5) | |
Text
2 5. Followup on Service Water Fouling Problems This inspection was performed to verify the actions taken at the Corporate Engineering Offices in response to the service water fouling problems experienced at Ocone Previous site inspections (87-14, 87-17) investi gated the licensee's on-site response and operability determinations for the Low Pressure Injection (LPI)
cooler The Reactor Building Cooling Units (RBCU)
and LPI operability issues were inspected at the Corporate Engineering Office A review was performed of the Duke Power Chronology of Events, dated May 12, 1987, concerning the fouling problems and the utilities response An examination was made of the flow of information from the site through the engineering and compliance staf It was determined that different groups performed the data analysis for the components prior to the Design Engineering Safety Analysis' operability determinatio Test data from the site for the LPI cooler analysis went to the Station Support Grou RBCU information was forwarded to the performance group in Nuclear Technical Services initiall Later, station services analyzed the RBCU dat Compiled data were sent to Design Engineering for review and an oper ability determinatio Licensing and management personnel reviewed the results, and Oconee site personnel were given guidance on the results of the operability determinatio Discussions were held with Design Engineering personnel in the Station Services and Station Support Groups, Performance Group, Safety Analysis Section, and Licensing Sectio A review was made of the decision making process performed in each group for operability determination calculations and evaluation The timing and uncertainties involved in the decisions were compared to the chronology of events. Decisions involving the RBCU's operability were made using data with higher uncertainties due to limitations of measurement accuracy and the lack of information available in industry for evaluating performance of condensing cooler These uncertainties effected the timing of the operability decision The inspectors noted inconsistencies in the value placed on data gathered during testing. Some data from testing was not used for an operability decision due to uncer tainties in measuremen If used, the data would have resulted in the components being less than fully operable and could have resulted in plant shutdow The same data was used later as the basis for an engineering calculation estimating heat removal capacity after cooler cleanin The uncertain value was multiplied by a guess for increase in cooler efficiency due to cleaning to estimate the new heat removal capacity after cleanin The guess was not based upon component testing, and later was found to be inaccurat The estimate of capacity after cleaning was used to justify continued plant operatio PDR ADOCK 05000269 G
PDR The inspectors reviewed the Units FSAR for operability values for RBCU and the LPI cooler The FSAR has heat removal capacities listed for RBCU listed in Section 15.13. The LPI coolers capacities are in figure 6.3-4 of the FSA The inspectors asked what action was initiated prior to allowing operations after determining that heat removal rates were lower than the FSAR value The licensing group stated that 10 CFR 50.59 evaluations were only considered necessary for procedure changes and modifications and were not considered to be required in the case of fouling. Operability evaluations were considered adequate by Duk The findings were consistent with the chronology of event A review was performed on the following procedures:
Design Engineering Department Manual, Section 11.4.6, "Station Operability Determination", 10/31/86 Duke Power Company, Nuclear Production Department Directive 2.8.2(T), "Operability Determination", Revision 0, June 1, 1987 Duke Power Company, Nuclear Production Department Directive 2.8.1(T), "Problem Investigation Process", Revision 1, 10/3/86 V 6. LPSW Inlet Temperature In Excess Of FSAR Value In addition to reviewing information relative to the LPI coolers and RBCUs, the inspectors discussed questions (see NRC Inspection Report 50-269, 270, 287/87-17),
concerning low pressure service water (LPSW)
inlet temperatures recorded in August 1986, during testing of the LPI coolers and the RBCU The inlet temperatures were higher than the maximum design LPSW temperature of 75 degrees F stated in the Oconee FSAR and used in the accident analyse Questions were raised as to whether there have been other instances where the LPSW inlet temperature exceeded 75 degrees and, if so, were evaluations performed to determine the affect on plant operations or the accident analyse The licensee stated that there have been other instances in previous years (during late summer and early fall) where the LPSW inlet temperature exceeded 75 degree The higher temperatures were not evaluated at the time to determine the affect on plant operations or the accident analyse Subsequent to Inspection 269, 270, 287/87-17, the licensee met with NRC Region II management in the Region II office on May 13, 1987, to discuss fouling of the LPI coolers and the RBCUs. The licensee also addressed the question concerning the higher LPSW inlet temperature It was stated during the meeting that evaluations are currently being performed to determine what affect the higher LPSW inlet temperature has on current plant operations and on the accident analyse The licensee made a commitment to complete the evaluation relative to the affect on current
4 plant operations before the LPSW inlet temperature reaches 75 degree Licensee representative restated the above commitment while discussing LPSW questions with the inspectors during this inspectio The question regarding operation with LPSW inlet temperatures greater than the design value, without performing an evaluation to determine the affect on plant operations or the accident analyses will be resolved pending completion of the licensee's current evaluations. This will be tracked as Unresolved Items 50-269, 270, 287/87-23-01, Evaluation of low pressure service water temperatures that exceeded design valu No violations or deviations were identified in the areas inspecte JUL 10 W7 Docket Nos. 50-269, 50-270, 50-287 License Nos. DPR-38, DPR-47, DPR-55 Duke Power Company L.TITN:
Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street Charlotte, NC 28242 Gentlemen:
SUBJECT:
INSPECTION REPORT NOS. 50-269/87-23, 50-270/87-23 AND 50-287/87-23 This refers to the Nuclear Regulatory Commission (NRC) inspection conducted by R. Bernhard on June 3-4, 198 The inspection included a review of activities authorized for your Oconee facilit At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed inspection repor Areas examined during the inspection are identified in the repor Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progres Within the scope of the inspection, no violations or deviations were identifie Your attention is invited to unresolved item identified in the inspection report. This matter will be pursued during future inspection The enclosed Inspection Report documents an oral commitment (paragraph 6)
made by a licensee representative and discussed in the exit interview. If your understanding of this commitment differs from the report statements, please inform this office promptl In accordance with Section 2.790 of the NRC's "Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosure will be placed in the NRC Public Document Roo Should you have any questions concerning this letter, please contact u
Sincerely, Alan R. Herdt, Chief Engineering Branch Division of Reactor Safety Enclosure:
(See page 2)
Duke Power Company
Enclosure:
NRC Inspection Report cc w/encl:
L-M. S. Tuckman, Station Manager bcc w/encl:
1-RC Resident Inspector (,H: Pastis, NRR Sjate of South Carolina
.Hickey, NRR
RII RIIRII RII RHBernhard:er M os ape s
06/
/87 04/9/87 Of/ 2/87
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