ML20213G589
| ML20213G589 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/10/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213G577 | List: |
| References | |
| TAC-61724, NUDOCS 8611180212 | |
| Download: ML20213G589 (3) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION UNIT NO. 1 DOCKET NO. 50-285 RELIEF FROM CERTAIN REQUIREMENTS OF SECTION XI 0F THE ASME CODE I. INTRODUCTION The Technical Specifications for the Fort Calhoun Station require that in-service examination of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Code as required by 10 CFR 50.55a(g)(4) except where specific written relief has been granted by the Commission. Some plants were designed in conformance to early editions of this Code Section, consequently certain requirements of later editions and addenda of Section XI are impractical to perform because of the plant's design, component geometry, and material of construction. Para-graph 10 CFR 50.55a(g)(6)(i) authorizes the Commission to grant relief from those requirements upon making the necessary findings.
In a letter dated June 10, 1986, the Omaha Public Power District (OPPD), the licensee, identified specific ASME Code requirements that OPPD determined to be impractical to perform at the Fort Calhoun Station and requested relief from these requirements. The staff has evaluated the licensee's supporting technical justification for these relief requests and finds them to be acceptable.
II. DISCUSSION The 1980 Edition including Addenda through Winter 1980 of ASME Section XI requires a volumetric and surface examination of piping greater than 1/2 inch thickness in order to detect a flaw in the lower 1/3 of the weld by volumetric examination techniques and the outside surface of the weld by surface examination techniques.
The licensee submitted inspection isometric drawings that identify the inaccessible welds, materials of construction, nominal diameter and nominal wall thickness or pipe schedule. The subject welds are in the main steam and feedwater system outside of containment in Room 81 of the Fort Calhoun facility.
A total of 29 welds (10 circumferential welds and 11 branch connection welds in the main steam system and 8 circumferential welds in the feedwater system) are inaccessible. The licensee requests Commission approval to exclude the inaccessible welds from the Inservice Inspection Program.
8611180212 861110 PDR ADOCK 05000285 O. PDR
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9 The protective enclosures in Room 81 were designed to contain the steam or
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fluid from a large break to the extent that severe impingement of jets on safety-related equipment or structures cannot occur. The protective en-closures consist of a series of longitudinal flat steel bars placed around the outside of the pipe insulation and held in position by a series of independent loops of wire rope. At the main steam safety valves, the spacing of the wire rope is increased to clear the valve nozzles and their reinforcing plates. At these locations the longitudinal enclosure bars are thickened and reinforced with stiffener plates. The wire rope size is increased and the spacing immediately adjacent to the nozzles is decreased to compensate for the increased spacing between wire rope loops in this area. In order to main-tain the configuration of the bars following pipe rupture, the individual bars are tightly fastened to the wire rope by means of wire rope clips. Each bar is fastened by clips at a spacing of approximately 1 foot. The protective enclosures around the feedwater lines are similar to those for the main steam line although the loads involved are far smaller.
The design criteria for the main steam and feedwater piping protective en-closures is in Attachment C to Appendix M of the updated Safety Analysis Report. These documents reflect the results of the " Fort Calhoun Unit 1 Pipe Rupture Analysis of Feedwater and Main Steam Outside of Containment" by Nuclear Services Corporation (Campbell, California) done in 1973, and on file in the OPPD Nuclear QA/ Records Management System. Basically, the con-clusion of this report is that the design of the protective enclosures is sufficient to assure the mitigation of a main steam line rupture or feedwater rupture in Room 81 so that all vital control capabilities will be maintained.
Updated Safety Analysis Report Section 14.12, " Main Steam Line Break" assures us that the radiological consequences of a main steam line break, concurrent with steam generator tube rupture, will not exceed the exposure limits of 10 CFR Part 100.
Two of the welds in the feedwater system, weld #7 on Figure B-7 and #11 on Figure B-8 were examined during the 1985 outage. However, these welds are located slightly inside of the bio-shield penetrations for the feedwater lines and only a limited examination of the welds was performed (approxi-mately 40% of the total weld length). The interference caused by their physical location does not prevent a 100% one-sided ultrasonic examination.
Because 100% of the weld volume can be covered with the UT examination, OPPD believes that the ultrasonic examinations will provide an adequate examination of the weld in the upcoming years. OPPD will, therefore, continue the surface examination to the extent possible, and do 100% of the required volumetric examination as scheduled in the normal ISI schedule. The feedwater welds are monitored by the required IWA-5000 system leakage test.
OPPD believes that between the ultrasonic examinations that are required by the Code and can be performed and the routine performance of the system leak checks, there is sufficient alternate examination to assure the integrity of weld 7 (Drawing B-7) and weld 11 (Drawing B-8) and, therefore these welds can be exempted from the surface examination requirements of Section XI.
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III. EVALUATION AND CONCLUSION The staff has completed the review of the licensee's submittal based on the provisions of 10 CFR 50.55a(g)(6)(i) and has determined that the protective enclosures make inservice inspection impractical to the extent required by ASME Section XI.
A partial examination can be performed on two of the welds. Since welds #7 and #11 are fabricated from carbon steel material, the staff agrees with the licensee that an ultrasonic examination of 100% of the weld volume and the limited surface examination performed during the 1985 outage should have detected any significant flaws.
ASME Section XI requires a volumetric and surface examination of a representa-tive sample of welds in order to detect unanticipated generic service-induced degradation in a piping system. The regulation and the ASME Code have pro-visions for the substitution of an alternative program that is demonstrated to be equivalent or superior to the specified requirement. The staff considers an Inservice Inspection Program that maintains the size of the examination sample to be equivalent.
Based on the foregoing, and pursuant to 10 CFR 50.55a(g)(6)(i), the staff has determined that certain ASME Section XI required inservice examinations are impractical to perform. The design of the subject welds incorporates pro-tective enclosures to mitigate the effects of a postulated large pipe break.
For subject welds #7 and #11 in the feedwater system, the staff concludes that the ultrasonic examination, the limited surface examination, and the Code-required pressure tests provide an acceptable level of quality and safety.
Therefore, relief from the balance of the Code-required surface examination may be granted.
The staff concludes that relief may be granted from the required inservice inspection of the other welds made inaccessible by the protective enclosures provided that the licensee selects for inservice inspection at least an equivalent number of accessible welds in these systems that are of similar material, configuration, nominal diameter, nominal wall thickness, and operating conditions.
Dated: November 10, 1986 Principal Contributor:
M. Hum, EB
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