ML20217E907
| ML20217E907 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/02/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217E896 | List: |
| References | |
| NUDOCS 9710070300 | |
| Download: ML20217E907 (5) | |
Text
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- UNITED STATES
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST (RR-A15) FOR SYSTEM LEAKAGE TEST REQUIRED BY AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI l
DETROIT EDISON COMPANY FERMI 2
, DOCKET NO. 50-341 l
1.0 INTRODUCTION
By letter dated September 9,1997, as supplemented by a letter dated September 16,1997, the Detroit Edison Company (DECO or the licensee) requested relief from certain inservice inspection requirements of the ASME Boiler and Pressure Vessel Code (Code),Section XI for the Fermi 2 plant. The licensee requested relief from performing a system leakage test at the nominal reactor coolant system pressure due to adverse plant conditions.
The technical specifications for Fermi 2 state that the inservice inspection of the ASME Code Class 1,2 and 3 components shall be performed in accordance with Section XI of the Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). In addition, 10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed altematives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2 and 3 components (including supports) shall meet the requirements, sxcept the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten year. Interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference h 10 CFR 50.55a(b) tws!ve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable ASME Code,Section XI, for Ferm! 2 first ten-year inservice inspection (ISI) interval is the 1980 Edition through Winter 1981 addendum. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated 9710070300 971002 PDR ADOCK 05000341 O PDR S
4-by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval, in letters dated September g, and September 16,1997. DECO submitted to the NRC its Relief Request No. RR-A15 for *he Fall igg 7 Midcycle Outage of the current inservice inspection interval, requesting relief from performing a system leakage test at the nominal reactor coolant system pressure due to adverse plant conditions. The staff has reviewed and evaluated the licensee's request to perform the system leakage test and coincident VT-2 visual examination below the Code required pressure, pursuant to 10 CFR 50.55a(s)(3)(ii) for Fermi 2, 2.0 SUPPORTING INFORMATION SUBMITTED BY THE LICENSEE
, COMPONENT FUNCTION / DESCRIPTION:
Reactor Vessel Pressure Retaining Boundary / Bolted Flange Connections Including Reactor Pressure Vessel (RPV) Head Flange, Vent Line Flanges, Main Steam Safety Relief Valve (SRV) Pilot Valve Flanges and other Class 1 Flanges resulting from emergent work.
SYSTEM: B11, RPV and B21, Nuclear Boiler System ASME CODE CLASS: Class 1 ASME SECTION XI REQUIREMENTS:
The American Society of Mechanical Engineers (ASME)Section XI,1980 Edition including the Winter 1981 Addenda, Subsection IWB, Table IWB-2500-1, Category B-P, Item B15.10, requires a VT-2 examination be performed during a system leakage test each refueling outage.- Subsection IWA-5214(e) also requires the performance of a system leakage test and associated VT-2 examination after disassembly and reassembly of mechanical joints for repair or replacement.
Detroit Edison believes these paragraphs indicate that it is the intent of ASME Section XI to require a system leakage test anytime an ASME Class 1 mechanical joint is disassembled and reassembled to assure that the bolted connections disassembled during the outage are leak tight following the maintenance activity. Detroit Edison further believes this is a conservative interprets"on of code requirements.
BASIS FOR RELIEF:
Plant conditions expected during a Midcycle Outage in the Fall of 1997, to replace leaking fuel bundle (s) and replace the Main Steam SRV. Pilot Valves, are different than normally encountered during a scheduled refueling outage. Decay heat from a partially bumed core with a shut down of only 11 days will be significantly higher than it would be following a normal refuel Outage where approximately one-third of the 764 fuel assemblies are replaced.
The decay hesi rate is predicted to be greater than three times that previously experienced during the shortest refuel outage at fermi 2. The anticipated excess decay heat present
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. during this Mideycle Outage poses an undue hardship to the safe and controlled performance of a code required system leakage test and the associated VT-2 examinations.
Test Considerations: The system leakage test is performed with the vessel in a near water solid condition such that any leakage would be in the form of water. The reactor coolant boundary is pressurized by use of the Control Rod Drive Pumps with pressure control achieved by regulating the Reactor Water Cleanup Blowdown path to the Ma!n Condenser.
The limiting test initiation temperature required to meet Technical Specification (TS) fracture prevention criteria is 163.5 'F (TS 3/4.4.6) at the reactor vessel flange with an associated system leakage test pressure of 1025 psig (nominal operating pressu .). Historically, the reactor coolant temperature required to achieve this metal temperature has been approximately 185 'F. This is very close to the bounding limit of 200 'F for OPERATIONAL CONDITION 4 (Cold Shutdown).
Decay Heat Considerations: With the anticipated decay heat, the vessel coolant heat-up rate is expected to be between 9.8 'F and 10.3 'F/hr. This results in a coolant temperature limited test duration of approximately 11/2 hours. These conditions preclude safe performance of the test prior to exceeding the TS OPERATIONAL CONDITION 4 temperature limit of 200 'F beyond which OPERATIONAL CONDITION 3 would be entered. The
- operations staff has concluded through use of the simulator that controlled pressurization, performance of visual examinations, and system restoration is expected to take at least 21/2 hours.
Personnel Safety Considerations: The Fermi 2 plant was designed to allow performance of required system leakage tests in OPERATIONAL CONDITION 4, without the drpell head installed. Performing the test in OPERATIONAL CONDITION 2 or 3 would require installation j of the drywell head to establish primary containment. Installing the drywell head creates a confined space above the drywell bulkhead. The confined space presents the following safety challenges to inspection personnel:
Accessibility - Access to this area is provided by a temporary 14 foot ladder installed on top of the biological shieM through one of two drywell bulkhead access hatches.- This also requires a s, tenuous climb from the bottom of the drywell rather than descending from the refueling floor into the cavity.
Inspectsbility - Clearances between the vessel mirror insulation on the RPV head and the drywell head (approximately 3 feet) restricts the visual examiners mobility for performance of inspections.
Temperature - The ambient temperature with the drywell head installed, in the area of inspection, is expected to be in excess of 130 'F.
The confined space air temperatures combined with limited accessibility and restricted inspection clearances create a hazardous work cor tition for inspection personnel performing the subject inspections.
Technical Considerations: . Plant Technical Specifications require primary containment to be OPERABLE (drywell head installed and containment isolation valves functioning) if the reactor coolant temperature exceeds 200 *F. Allowing the coolant terrperature to exceed 200 'F would require that other plant systems such as High Pressure Coolant injection
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.. (HPCI) arvi Reactor Core Isolation Cooling (RCIC) to be operable. Performing the test in this manner is not possible because the steam supply lines would be flooded with water during the test, rendering these systems inoperable, if the test were to be performed in OPERATIONAL CONDITION 2 or 3 the ambient <
temperature under the drywell head would be in excess of 140 'F. The vessel metal temperature would exceed 220 'F and any leakage would be in the form of steam.
ALTERNATIVE EXAMINATION:
Detroit Edison proposes perf0rming the pressure test and coincident VT-2 examination at a pressure of at least 750 psig. Performing the test at this pressurs without the drywell head installed allows inspections to be performed in OPERATIONAL CONDITION 4 as designed and eliminates the personnel safety issues asr,ociated with performing the inspections with the drywell head installed. This also provides for a much lower test initiation temperature of 132.2 'F on the RPV flange, with an associated coolant temperature of approximately 153 'F. The reduced reactor coolant test initiation temperature would increase the temperature margin to OPERATIONAL CONDITION 3 to approximately 47 'F. As a result,-
i the system leakage test (pressurization, inspection, and recovery) can be completed in a safe I and controlled manner and eliminate the operator control challenges present in performing a 1025 psig pressure test. Testing can be completed without challenging the OPERATIONAL CONDITION 4 limiting coolant temperature of 200 'F. This precludes the need to establish primary containment (having the drywell head installed) and would alleviate HPCI and RCIC OPERABILITY concems.
Performing the tett at this reduced pressure would not be expected to significantly influence visual leakage detection capabilities on flanged joints. When flanged joints do leak, it is normally evident at pressures well below nominal operating pressure. Typically, bolted flange joints are leak tight providi..g the mating surfaces and gasket are clean and free of damage
- and torquing is done in the proper sequence to a consistent torque value. Gasket installation and torquing activities are procedurally controlled at Fermi 2 to ensure satisfactory results.
- In the unlikely event that any leakage occurs at increased pressure during startup or operation, other detection capabilities exist to identify any such leakage. Technical Specifications 3.4.3.1 and 3.4.3.2, require the vessel flange leak detection system to be operable and that total reactor coolant system leakage must still meet established acceptance limits during plant operation. Additionally, temperature monitoring and gaseous radioactivity monitoring capabilities are available to detect abnorinal conditions indicative of leakage. Unidentified leakage has been less than 0.5 gallons per minute throughout the present operating cycle and would not be expected to incmase significantly following the Midcycle Outage.
APPLICABLE TIME PERIOD:
Relief is requested for one-time performance during the Fall 1997 Midcycle Outage.
3.0 EVALUATION
' The ASME Code,Section XI,1980 Edition including the Winter '81 Addendum requires that the system leakage test be performed at the nominal reactor coolant pressure of 1025 psig.
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. The staff has determined that the constraint resulting from high decay heat in the reactor core contributes to a hazardous working condition inside the drywell due to high ambient temperature, and imposes limiting conditions to perform the system leakage test below 200_'F (Shutdown Condition) during the short midcycle outage. Therefore, a reduced pressure system leakage test of the vessel and the reactor coolant system is necessary in order to maintain coolant temperature low enough during the test to preclude entering into higher modes of operation inadvertently with the drywell head removed. Hence, the licensee has proposed to conduct the system leakage test at 750 psig instead of the required test pressure of 1025 psig. The VT-2 visual examination of the flanged connections within the pressure boundary should allow detection of leakage, even at a pressure of 750 psig, should there be a leakage path. The industry experience further demonstrates that only a very small percentage of pressure boundary leakages are generally attributable to leakage from bolted flange joints that are procedurally controlled in regart to torquing of bolts and gasket installation. Nevertheless, in the unlikely event of an extremely small leak escaping detection during the system leakage test, it can still be detected during startup or operation by other detection capabilities while the leakage is small, without affecting the health and safety of the public. Therefore, the proposed testing provides a reasonable assurance of operational
' readiness. If the licensee were to comply with the requirement of the Code in conducting the test at a pressure of 1025 psig, it will cause significant hardship to the licensee in terms of personnel hazard without a compensating increase in quality and safety.
The Component Function / Description Section of the request for relief listed the specific components in the upper portions of the drywell (i.e., at the elevation that incudes the SRVs and above) for which relief was requested. However, the description also included a more generic statement for other Class i flanges that required inspection as a result of emergent work. This statement could be read as including flanges in accessible areas of the plant.
Based on a telephone conversation with the licensee staff, this statement is clarifed to only inc!ude Class i flanges that require inspection as a result of emergent work that are located in the upper portions of the drywell as defined above. This relief request does not cover-Class 1 flanges in other areas of the plant.
4.0 CONCLUSION
i The staff has reviewed the licensee's submittal of Relief Request No. RR-A15 and has concluded, pursuant to 10 CFR 50.55a(a)(3)(ii), that the required system leakage test at nominal reactor coolant pressure and associated VT-2 examinations result in hardship for the licensee without a compensating increase in the level of quality and safety. The licensee has demonstrated that the proposed attemative inspection provides reasonable assurance of necessary quality and safety. Therefore, the staff concludes that the relief from the code requirement may be granted for relief request RR-A15 for one-time performance during the Fall 1997 Midcycle Outage. Such relief is authorized by law and will not endanger public health and safety or the common defense and security and is otherwise in the public interest.
Principal Contributor: P, Patnaik Date: October 2, 1997
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