ML19241C143
| ML19241C143 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/05/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19241C141 | List: |
| References | |
| SER-790705-1, NUDOCS 7907270364 | |
| Download: ML19241C143 (10) | |
Text
f UNITED STATES Ib, S NUCLE AR REGULATORY COMMISSION 3'
h WASHINGTON, D. C. 20555 h's ~lpE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 52 TO LICENSE N0. DPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER 1.
_ Introduction On April 26, 1976, the Commission sent a generic letter to Iowa Electric Light and Power Company (the licensee) advising them that the inservice inspection and testing requirements for ASME Code Class 1, 2 and 3 com-ponents for nuclear pcwer plants delineated in 10 CFR Part 50.55a were changed by a revision to the regulations published on February 27, 1976.
The revised regulations require inservice inspection and testing to be performed in accordance with the examination and testing requirements set forth in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda thereto.
To avoid potential conflicts between the ASME Code requirements and the Technical Specifications presently in effect for the Duane Arnold Energy Center, we also advised the licensee that he should apply to the Commission for amendment of the Technical Specifi-cations.
Sample language for such Technical Specifications changes was providec as an enclosure to OJr letter of April 26, 1976.
By letter dated November 30, 1977, the licensee requested a change to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC).
The pro-posed amendment and revised Technical Specifications would delete the presen* inspection and testing requirements in Section 3.6.6 of the Technical Specification and substitute therefore - verbatim - the sample language enclosed with our letter of April 26, 1976.
The proposed Technical Specification would require all inspection and testing to be perfonned in accordance with the ASME Code except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Sectior, 50.55a(d)(6)(i).
Our letter of April 26, 1976 also advised the licensee that if he deter-mines that confomance with c.ertain ASME Section XI inservice inspectio'1 and testing requirements is impractical, he should submit information to the Commission to support his determination in accordance with 50.55a(g)
(5)(iii) and (iv).
By letters dated Ncvember 22, 1977 and January 5, 1978, we provided additional guidance in preparing inservice inspection 517 295 7 9 072.70 N
. and testing program descriptions and associated relief requests.
In response to our letters, the licensee submitted a prnposed Inservice Inspection and Testing Program by letter dated March 1, 1978, supplemented by letter dated March 15, 1978.
This submittal also included requests for relief from examining certain components where the licensee deter-mined that it was impossible or impractical to examine or test the specific component because of design, geometry or materials of construc-tion.
As part of our review, a meeting was held with the licensee and his consultants on June 13, 1978.
In response to staff comments, the licensee submitted a revised Inservice Inspection and Testing Program by letter dated October 13, 1978.
This letter also revised the requests for relief where the 'icensee detennined that certain requirements of the ASNE Code cannot be implemented at the Duane Arnold facility because of component or system design, geometry or materials of construction.
2.0 Evaluation 2.1 Technical speci fica tions The changes proposed by the licensee to the Technical Specifications are identical to the sample Technical Specifications enclosed with our letter of April 26, 1976.
The revised Technical Specifications require all inspections and testing to be performed in accordance with the ASME Boiler and Pressure Vessel Code and are acceptable.
2.2 Requests for Relief As required by 10 CFR 50.55a(g), Iowa Electric Light and Power Company has updated the Inservice Inspection Program for the Duane Arnold Nuclear Generating Plant to the requirements of the 1974 Edition through Summer 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code (B&PV Code).
Based or infornation contained in the submittal dated March i, 1978 and the revised submittal dated October 13, 1978, it has been deter-mined that certain requirements of the Code cannot be implemented at the facility because of component or system design, geometry, or materials of construction.
Requested reliefs from those requirements have been reviewed and evaluated by the staf f and our determinations to grant or deny the requests, pursuant to 10 CFR 50.55a(g)(6)(i), are documented below.
2.2.1 Class 1 Copjponents_
A.
Relief is requested from volumetric examination of each neridional weld in the bottom head of the reactor vessel.
(Item B 1.2, Category B-B) 51y 296
. _ Code Requifement The examinations performed during each inspection interval shall cover at least 10% of each meridional head weld.
L_icenset Basis for Reauestina Relief and Alternate Examination Proposed Th se welds are located within the array of control rod driv.es ar d are not accessible for volumetric examination.
Jhese welds will be visually examined for leakage or deposits caused by leakage during the leak testing after each refueling outage and during the hydrostatic test to be performed near the end of the 10-year interval.
Evaluation Because of the design of the vessel, access to these welds is limited from the exterior of the vessel by the control rod drives and from the interior by the core shroud and core plate and preclude volumetric or surface examination with currently available technology.
The licensee has proposed to visually examine these welds for leakage or deposits caused by leakage during leak testing after each refueling outage and during the hydrostatic test performed near the end of the 10-year interval.
In addition, the staff recommended, and the licensee agreed, that the proposed examination be supplemented by a surface examination at areas of the welds accessible for surface examination.
The reactor vessel waa designed, fabricated, examined and tested to the rules of Section III of the ASME Boiler and Pressure Vessel Code.
The inaccessible meridional head welds are in an area of the vessel which is subjected to a lower neutron flux than that of the beltline region and is therefore less susceptible to radiation damage and brittle fracture. Areas of the vessel which are subjected to the higher neutron flux will be examined to Code requirements and the on-going material surveillance program will provide data to determine the condition or change in the properties of materials in the most severe locations.
The Technical Specifications and plant operating procedures require that certain leak detection systems be functioning during operation and inpose limits on the amount of leakage that may be permitted.
Specifically, the plant must be shut down for inspection and corrective 517 297
4-action whenever the leakage system indicates, within a period of four hours or less, an increase in the rate of unidentified leakage in excess of 2 gpm or when the total unidentified leakage exceeds 5 gpm.
The staff concludes that design, leakage monitoring require-ments, examinations being perfonned on other welds on the vessel, and the examination proposed by the licensee wil; provide adequate assurance of the vessel's structural integrity and therefore relief from the volumetric examination requirement for the meridional head welds may be granted.
B.
Relief is req;ested from the volumetric examination of the drain nozzle weld in the bottom hear' of the reactor vessel.
(Item d 1.4, Category B-D).
Co_do Requirement The examination of each nozzle shall cover 100% of the volume saown in Figure IWB-2500 D during each inspection interval.
Licensee Basis fcr Requesting Relief and Alternate Examination Proposed This weld is located within 'he array of control rod drives and is not accessible for volumetric examination.
This weld will be visually examined for leakage or deposits caused by leakage during the leak testing af ter each refueling outage and during hydrostatic test to be performed near the end of the 10 year interval.
Evaluation Complete failure of the weld attaching the 1 3/4 inch 0.D. drain nozzle to the reactor vessel has been demonstrcted by the licensec as a failure which would not cause a net loss of coolant because of the nornal makeup capacity of the reactor coolant makeup system, assuming a simultaneour loss of offsite power. This weld may be exempted from volumetric t., amination as allowed by IWB-1220 (b)(1) and sub-jected to the requirements of Examination Category B-P of Table IWB-2500.
The alternate examination proposed by the licensee exceeds the Code requirements for Category B-P and the staff con-cludes that the proposed alternate examinations will provide adequate assurance of the structural integrity of this weld.
C.
Relief is requested from the visual examination for the internal pressure boundary surfaces of the main recirculation system valves.
(Item B 6.7, Category B-M-2) b\\i 7O@
i
. Code Requiranents One valve in each qroup of valves of the same constructural design (e.g., globe, gate or check valve, manufacturing method and manu-facturer) that performs similar functions in the system shall be visually examined internally during each inspection interval.
Licensee Basis for Requesting Relief and Alternate Examinations Proposed The main recirculation valves are located in piping which penetrates the reactor pressure vessel and cannot be isolated for disassembly and visual examination.
To accomplish the required examination would entail drainage of the reactor vessel well as removal of the core.
These valves will be examined should valve maintenance be required.
For those intervals when valve maintenance does not occur Leak Tests and Pressure Tests will be performed in accordance with Category B-P (a pressure test once in 10 years).
Evaluation Because of the design of the reactor coolant pressure boundary, the interna.1 pressure boundary surface of the main retircula-tion system valves are not accessib!e for visual examination since the valves cannot be isolated f.om the reactor vescel to allow disassembly.
In order to inspect the valves, the fuel must be removed from the reactor core and the reactor vessel must be drained.
The licensee has committed to inspect the valves when valve maintenance is required and to conduct visual examination when the system pressure tests (IWA-5000) are conducted in accordance with the requirements for Category B-P.
For those intervals when the valves are not disassembled and in-spected, the staf f recomnended, and the licensee agreed, that thick-ness measurements (ultrasonic examination) be performed on one valve of the group.
The staff has determined that measurement of valve body wall thickness to the minimum requirements estr.olished by the ASME B&PV Code,Section III will provide similar.iformation to that obtained in a visual examination, i.e., degradatioa of the wall by generel corrosion, cavitation or erosion, and that this information is adequate in providing assurance of the continued material and structural acceptability of these valves. We conclude that relief from the internal visual examination should be granted.
517 299
_o.
.~ -
2.2.2 Class 2 Components - Pressure Tests A.
Relief is requested from pressure testing the piping from
~
the nain steam relief valves to the torus.
Code Requirement Pressure test near the end of the inspection interval. The system test pressure shall be at least 1.25 times the system design pressure.
Licensee Basis for Requesting Relief and Alternate Testino Proposed T"e pipe is open-ended in the torus.
The pipe and supports will be visually examined.
If there are indicati; _ af structural distress in any component or indications that the component had leaked during operation of tne relief valve, the component will be surface or volumetrically examined.
Cvaluation The relief valve discharge line to the torus is an open-ended line ar.d therefore nay be exempted from the system pressure test requirement as per IWC-5220(d) of Section XI ASME Code.
B.
Reliet Request Relief is requested from Code required distribution of pressure tests for Class 2 components.
Code Requirement The tests required ; hall be distributed as follows:
(a)
Between 25 and '31/3% of the required tests shall be comoleted by the oxpiration of one-third of each inspection interval.
(b)
Between 50 and 66-2/3% of the required tests shall be comoleted by the expiration of two-thirds of each insoection interval.
(c)
The remaining required tests shall be completed by the end of each inspection interval.
5i/
300
. Licensee Basis for Requesting Relief and Proposed Alternate Testing It is impractical to schedule tests at these intervals inasmuch as the systems cannot be isolated at the boundaries created by the NDE exemption.
Redur. dant pressure tests would be performed that would serve no useful purpose.
These systems are normally pressurized for pump or valve,unctional tests which would reveal any degradation of the system.
These components will be pressure tested at or near the end of each 10-year inspection interval.
This proposal is in accordance with Section XI Subcommittee interpretation of Class 2 pressure test requi-ements.
Evaluation The licensee,has proposed that all components be pressure tested at or near the end of each inspection interval (10 years) instead of pressure testing some of the exempted components during the inspection interval.
The staff has evaluated the licensee's basis for requesting relief and concluded that this request should not be granted.
However, the staf f concludes that the following examinations may be conducted.
A system functional test may serve as a system pressure test and at least one visral examination shall be conducted at or near the end of each inspection period coinciding with a system functional test.
In adc' tion a system hydrostatic test shall be conducted at or near the end of each inspection interval.
These requirements are con-sistent with the Winter 77 Addenda Section XI requirements for all Class 2 components and will maintain an acceptable level of quality during the 10-year interval.
The licensee concurs with the staff's recomenda tions.
2.2.3 Class 1, 2 & 3, Components - Pressure Tests Relief is requested from maintainiig pressure four hours for all Class 1, 2 and 3 leakage and hydrostatic tests.
Coae Requirement The test pressure and temperature shall be maintained for at least four hours prior to performance of the examinations.
Licensee Basis for Relief and Proposed Alternate Testing The intent of the Code is to hold the pressure for four hours during testing of components that are covered by insulation.
No useful purpose would be achieved in holding the pressure for four hours where the components to be examined are exposed.
517 301
. The test pressure and temperature will be maintained for a minimum of four hours as required by IWA-5210 (a) where areas of examination are not exposed and accessible for visual examination.
The test pressure and temperature will be maintained for a minimum of 10 minutes where areas are exposed and accessible for visual examination.
Evaluation The staff concludes that this relief should be granted with the following conditions.
(a) When performing a system pre.sure test the entire system must be directly visible.
This includes the welds and all base materials.
(b)
Following a repair the repaired area must be accessible for a direct visual examination.
(c) When the areas are exposed, the pressure and temperature shall be maintained for a minimun time of 10 minutes and for such additional time as may be necessary to conduct the examinations.
The above conditions are consistent with the rules of Section XI Winter Addenda which the staff finds acceptable and which will not decrease the quality or safety of the facility.
The licensee has accepted the above conditions and will revise the inspection procedures accordingly.
2.2.4 Overall Evaluation We have evaluated requests for relief from Code requirements which the licensee has determined to be impractical for implementation at the facility and granted relief from the requirements in those cases where our evaluation deems that such relief will not endanger life or property and is in the public interest giving due consideration to the burden placed on the licensee if the requirements were imposed.
We conclude that the Inservice Inspection Program meets the requirements of the 1974 Edition through Summer 1975 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, to the extent practical and thus meets the requirements set forth in 10 CFR 50.55a(g).
2.3 Recirculation Inlet Safe Ends _
On June 17, 1978, the licensee discovered a through-wall crack in a safe-end on one of the recirculation inlet lines.
The safe-end is a short transition piece (approximately 8 inches long) joining the recirculation inlet line to the nozzle on the reactor vessel and to the internal inlet line to the jet pumps.
Nondestructive testing of the other seven identical safe-ends revealed that all had indications of cracks or weld irregularities; however, these 5i7 302
. flaws did not penetrate to the surface of the safe-ends. The licensee rem 0ved all eight safe-ends and replaced them with safe-ends of an improved design. The new design minimizes the tight crevice formed by the fit up of the safe-end and ar internal thermal sleeve; such crevices are known to enhance the possibility of : tress corrosion cracking in an adverse chemical environment.
On March 5,1979, we issued Amendment No. 49 to Facility License No. DPR-49 authorizing the Duane Arnold Energy Center to resume normal power operations following installation of the eight new replacement recirculation inlet safe-ends.
Amendment No. 49 also changed the Technical Specifications to incor-porate augmented inservice inspection of the modified safe-ends.
All pressure boundary welds (designated as numbers 2, 6 and 7) were subjected to an ultrasonic examination to provide a base line for future examinations. Com-plete recordings were made of these examinations to ensure that any changes in ultrasonic results indicative of cracking during service will be identified.
The specific program that will be followed will be to ultrasonically examine all three welds in one half (four) of the eight safe end assemblies every refueling outage.
This program will continue at least until every weld involved will have been inspected twice.
As the Duane Arnold plant is on a yearly refueling schedule, this means that welds in four of the safe-end assemblies will be inspected af ter one and three years of operation, the remaining four will be inspected af ter two and four years of operation.
The requirements in Amendment No. 49 are not changed by this amendment and supersede the less frequent examinations that would be required by the ASME Boiler and Pressure Vessel Code.
During the outage associated with replacement of the safe ends on the recir-culation inlet lines, Iowa Electric performed ultrasonic examinations of fle core spray, the feedwater (FW) and the control rod drive hydraulic ystem return (CRD HSR) nozzle safe ends. These are the only locations on thc D4EC reactcr vessel where a thermal sleeve is welded to a safe end and that contain potentially creviced Alloy 600 material on the primary pressure.
There are significant differences, however, between the above safe ends and the recirculation system inlet nozzle safe ends. 'The maximum calculated stress index values for the core spray, feedwater and CR0 HSR safe ends is 0.71,1.19 and 0.88, respectively, compared to 2.24 for the recirculation inlet safe ends.
The of essure boundary material for the FW safe end is carbon steel with Ni-Cr-Fe weld metal inlay at the crevice location, whereas, the material in t'le heat affected zone of the recirculation inlet safe end (where the crack occurred) is wrought Alloy 600.
The crevice in the FW safe end is 0.125 inchos maximum as opposed to the 0.50 inch length in the former recirculatio" inlet safe ends.
As documented in Report No. 50-331/79-10 n 27, 1979 issued by NRC's Office of Inspection and Enforcement dated Mar Regica III, a fuli volumetric examination of the above safe ends was performed, with emphasis in the ultrasonic examination on the areas of the safe ends 517 303
. where thermal sleeves are welded to the safe ends.
No significant inidcations of cracking or reflections were noted in the areas of interest on any of the safe ends. Adequate base lines for future ultrasonic exa:ainations were established for the core spray, feedwater and CRD HSR nozzle safe ends.
These safe ends will be examined by ultrasonic inspection techniques during subs?quent refueling outages.
3.0 Environmental Considerations We have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any signi ficant environmental impact.
Having made this determination, we have further concluded that this amendment involves an action which is insig-nificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmeTtal impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusion We have concluded:
(1) because the amendment does not invohe a significant increase ir. the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted ir, compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the cormon defense and security or to the health and safety of the public.
Deted: July 5, 1979 517 304