ML20080E421
| ML20080E421 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/06/1995 |
| From: | Norrholm L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080E425 | List: |
| References | |
| NUDOCS 9501130206 | |
| Download: ML20080E421 (5) | |
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UNITED STATES 5
NUCLEAR REGULATORY COMMISSION l
WASHINGTON, D.C. 20555-0001'
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t IES UTILITIES INC.
j CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE l
DOCKET NO. 50-331 PUANE ARNOLD ENERGY CENTER AMENOMENT TO FACILITY OPERATING LICENSE Amendment.No. 206 License No. DPR-49' l
l 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by IES Utilities Inc., et al.,
dated July 29, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the '
provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized
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by this amendment can be conducted without endangering the health i
and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the common i'
defense and security or to the health and safety of the public*
and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements -
have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-
= cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
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9501130206 9 106 PDR ADOCK 0 331 P
PDR.
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. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 206, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 120 days of the date of issuance.
i en F0 THE NUCLEAR REG 0LATORY COMMISSION l
\\
Leif J. Norrholm, Director Project Directorate III-3 Division of Reactor Projects III/IV i
Office of Nuclear Reactor Regulation i
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Attachment:
Changes to the Technical I
i Specifications
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Date of issuance: January 6,1995 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 206
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FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the
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l enclosed pages. The revised areas are indicated by vertical lines.
Remove Insert 3.6-11 3.6-11 3.6-28 3.6-28 l
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DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS j
'4 F.'
Jet Pa=a Flow Miamatch F.
Jet Pa=a Flow Mismatch j
f 1.
With core power greater than or 1.
Recirculation pump speed mismatch 9,.
equal to 80% RATED POWER with shall be verified at least once both recirculation pumps at per day.
steady state operation, the speed i
1 of the faster pump may not exceed j
122% of the speed of the slower pump.
i 2.
With core power less than 80%
l RATED POWER with both recirculation pumps at steady state operation, the speed of the
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faster pump may not exceed 1354 of the speed of the slower pump.
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3.
With the recirculation pump j
speeds different by more than the i
specified.11mitst.
a.
restore the recirculation i
pump speeds to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or i
1 b.
one recirculation pump 2.
See Surveillt it Requirement l
shall be tripped. See 4.3.F.4 for SLO requirements.
j Specification 3.3.F.4 for j
SLO requirements.
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G.
Structural Intearity G.
Structural Intearity i
I 1.
At all times, the structural 1.
Inservice inspection of ASME
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integrity of the ASME Section XI Section XI Code Class 1, Class 2, Code class 1, 2, and 3 components and Class 3 components and shall be maintained in accordance inservice testing of ASME Section with Surveillance Requirement XI Code Class 1, Class 2, and j
4.6.G.I.
Class 3 pumps and valves shall be performed in accordence with l
2.
With the structural integrity of Section XI of the ASME Boiler and any ASME Section XI Code Class 1 Pressure Vassel Code and
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l or Class 2 component (s) not applicable Addenda as required by conforming to the above-10CFR50, Section 50.55a.
requirements, restore the-1 structural integrity of the 2.
The augmented inspection program affected component (s) to within for piping identified in NRC its limit or isolate the affected Generic Letter 88-01 shall be component (s) prior to increasing performed in accordance with the i
the Reactor Coolant System staff positions on schedule, t
temperature above 212*F.
methods, personnel, and sample r
expansion included in this Generic l
3.
kith the structural integrity of Letter.
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ar.;r ASME Section XI Code Class 3 component (s) not conforming to 3.
The provisions of Definition 26 the above requirements, restore
~ applicable to the frequencies for
[ SURVEILLANCE FREQUENCY) are the structural integrity of the affected component (s) to within performing inservice inspection its limit or isolate the affected and inservice testing activities.
i l-component (s) from service.
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AMENDMENT NO. JJ3.293,206 3.6-11 i
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s DAEC-1 4
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3.6.G E 4.6.G BASES:
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Structural Integrity
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.A pre-service. inspection of Nuclear Class;I Components.was' conducted to assure-k
. freedom from defects' greater than code allowances-in' addition, this served as J
'a reference base for future, inspections., Prior to operation, the Reactor Coolant System'as described in Article.IS-120 of Section XI of.the ASME Boiler and Pressure. vessel Code was inspected.to provide assurance that the system was-free of gross defects..In addition,'the facility was designed such'that gross defects-should not occur throughout plant life. 'The pre-service-inspection program was based on the,1970 Section XI of the.ASME Code for in :
service inspection. This inspection plan waa designed to reveal' problem areas (should they occur) before a leak inlthe coolant, system could develop. 'Thel program was established to provide reasonable assurance that no LOCA would; occur at the DAEC as a result of leakage or breach of~ pressure-containing.
components.and piping of the Reactor Coolant. System,. portions of the ECCS,. and :
portions of the reactor coolant. associated auxiliary. systems.
A pre-service. inspection was not performed on' Nuclear Class II_ Components because it was not required at that, stage.of DAEC construction when it would have been used. Forithese components, shop and in-plant examination records' of components and welds.will-be used as a basis for comparison with in-service inspection data.
Visual examinations for leaks'will be'made periodically on~ASME Section:XI Class _1, 2 and 3 systems. The inspection program specified encompasses the major areas of-the vessel and piping systems within the'ASME Section XI
-boundaries.
The type of examination planned for each component' depends on location, accessibility,- and type of potential defect._ : Direct visual examination is proposed wherever possible since it is fast and reliable. Surface examinations are planned where practical,.and where added; sensitivity'is required. Ultrasonic examination or radiography shall be used where defects can occur in concealed surfaces. Section 5.2.4 of the Updated FSAR provides; details'of the inservice inspection program.
Starting with the cycle 9/10 Refueling Outage, an augmented inspection program was implemented to address concerns relating to Intergranular. Stress Corrosion Cracking (ICSCC) in reactor coolant piping made of austenitic stainless steel.
The augmented inspection program conforms to the NRC staff's positions set forth in Generic Letter 88-01 and NUREG-0313, Revision-2 for inspection schedule, inspection methods and personnel, and. inspection sample expansion.
The first 10-year interval for inservice testing of pumps'and valves:in accordance with the ASME Code,Section XI commenced ~on February:1, 1975 and ended on January 31, 1985. The second 10-year inservice testing interval' commenced on February 1, 1985 and is scheduled to end on. January 31, 1995.
The second 10-year testing program addresses the requirements of the ASME Code,Section XI, 1980 Editior, with Addenda through Winter 1981,. subject, to the limitations and modifications of 10 CFR 50.55a.
Section 3.9.6 of the-Updated FSAR describes the inservice testing program.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.-
AMENDMENT NO. JJ5,H ),79),206 3.6-28
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