ML20136F604
| ML20136F604 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 06/15/1995 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20136C539 | List:
|
| References | |
| FOIA-96-485 NUDOCS 9703170012 | |
| Download: ML20136F604 (7) | |
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ATTACHMENT 1
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SAFETY EVALUATION BY THE OFFICE OF NL' CLEAR REACTOR REGULATION j
OF THE SECOND TEN YEAR INTERVAL INSERVICE INSPECTION PRocP_AM PLAN REQUEST FOR RELIEF NO. 19 i
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FLORIDA POWER AND LIGHT COMPANY j3 ST. LUCIE NUCLEAR POWER PLANT. UNIT 2 DOCKET NUMBER: 50-389
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1.0 INTRODUCTION
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The Technical Specifications for St. Lucie Nuclear Power Plant, Unit 2 state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be performed in accordance with
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Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda if as required by 10 CFR 50.55a(g), except where specific written relief has been l
'1 granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1).
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10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (1) the proposed alternatives j
would provide an acceptable level of quality and safety or (ii) compliance j
i with the specified requirements would result in hardship or unusual 1
i 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, except the design and access j
provisions and the preservice examination requirements, set forth in the ASME 1
Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant
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Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations 1
require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply j
with the requirements in the latest edition and addenda of Section XI of the l
ASME Code incorporated by reference in 10 CFR'50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the j
ASME Code for the St. Lucie Nuclear Plant, Unit 2 second 10-year inservice
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inspection (ISI) interval is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission 1
approval.
i Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission I
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in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to i
10 CFR 50.55a(g)(6)(1), the Commission may grant relief and may impose jl1 alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise j
in the public interest, giving due consideration to the burden upon the J
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j 9703170012 970301 4
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2 licensee that could result if the requirements were imposed.
In a letter dated April 3. 1995, Florida Power and Light Company submitted to the NRC its second ten-year interval inservice inspection program plan Request for Relief No.18 regarding visual examination of insulation components for St. Lucie Nuclear Plant, Unit 2.
I 2.0 EVALUATION AND CONCLUSIONS The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the information provided by the licensee in support of its second ten-year interval inservice inspection j
program plan Request for Relief No.19 regarding visual examination of insulation components for St. Lucie Nuclear Plant, Unit 2.
Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report I
attached.
.The staff reviewed the licensee's proposed alternatives to removal of
-insulation at bolted connections in systems borated for the purpose of controlling reactivity and concluded that requiring the licensee to remove insulation on all bolted connections in systems borated for the purpose of controlling reactivity during the upcoming refueling outage would result.in a hardship without a compensating increase in safety. The licensee's proposed alternatives for the interim request for relief should provide reasonable assurance of operational readiness. Therefore, interia relief for the currently scheduled refueling outage in September,1995, is authorized pursuant to 10 CFR 50.55a(a)(3)(11) provided that the licensee follows its commitment to remove all removable insulation at bolted connections in borated systems and perform a VT-2 visual examination.
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ATTACHMENT 2 TECMICAL EVALUATI0ll LETTER l
ON THE SEC05 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF IRNIBER 19 1
FOR FLORIDA PONER AW LIGHT COMPANY ST. LUCIE BRICLEAR PLANT, UNIT 2 j.
D0CKET ORNSER: 50-389 1.0 HTRODUCTION t
l The licensee, Florida Power and Light Company (FPL), submitted Request for i
Relief Number 19, for the second 10-year inservice inspection (ISI) interval, j
in a letter dated April 3, 1995. The Idaho National Engineering Laboratory
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(INEL) has evaluated the subject request for relief in the following section..
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I 2.0 EVALUATION I
The Code of record for the St. Lucie Nuclear Plant, Unit 2, second 10-year j
inservice inspection interval, which began August 8, 1993, is the 1989 Edition i
of the American Society of Nechanical Engineers,Section XI. The information provided by the licensee in support of the request for relief from the i
impractical requirement has been evaluated and the basis for granting relief from that requirement is documented below.
l Recuest for Relief 19. Paraaraoh IWA-5242(a). Visual Examination of Insulated Connonents i
Code Raouirement: Paragraph IWA-5242(a) requires that insulation be removed from pressure-retaining bolted connections for VT-2 visual examination on systems borated for the purpose of controlling reactivity.
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Licensee's Code Relief Reauest: The licensee requested interim relief from the removal of insulation on pressure-retaining bolted connections in borated systems during VT-2 visual examination for the upcoming refueling outage scheduled for September 1995.
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Licensee's Basis for Reauestina Relief (as stated):
"For systems borated for the purpose of controlling reactivity, removal of insulation from bolted connections for the purpose of performing a visual examination for corrosion will involve a significant increase in man hours, radiation exposure, and material.
i "The quantity of bolted connections which will require insulation removal and restoration, as determined by an initial review of drawings and other J
design documents, involves a significant increase in the amount of man hours and material. This hardship in turn, results in escalated operations maintenance costs, and radiation exposure, without a compensating increase in the level of quality and safety.
"In an effort to minimize the impact of these examinations in the future, i
FPL will evaluate the feasibility and cost benefit of an insulation modification at applicable locations, such that an examination may be performed without the need to remove insulation each time. The j
evaluation of the feasibility of this modification, however, cannot be completed until a walkdown of the piping is performed. This walkdown cannot be performed at power.
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"This interim relief will provide time to resolve the scope of these examinations through the ASME Code process and evaluate the results of system walkdowns conducted inside containment to determine the feasibility of permanent design changes."
Licensee's Proposed Alternative Examination (as stated):
" Florida Power Light will check bolted connections for leakage when performing system examinations as follows:
"As soon as possible, after coming off line for a refueling outage, a leak test is coordinated by the system engineers inside the containment per the plant surveillance program.
"Dr ritg the outage, suitcase style insulation will be removed from the b actor Coolant and Charging systems inside containment, and the con'e :tions visually examined (VT-2) for evidence of leakage when the plant is depressurized. When evidence of leakage is identified, repairs will be performed in accordance with the current maintenance work practices.
"During the outage, any Class 1 or Class 2 insulated connections in the Reactor Coolant and Charging systems inside containment that are disassembled will be examined for evidence of leakage by maintenance personnel. Repairs will be performed in accordance with the current maintenance work practices.
" Prior to reactor criticality, following a refueling outage, a system leakage test is performed at normal operating pressure and i
temperature with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time.
l "These Leakage Tests will include looking for the following conditions:
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" pooling of water directly under the bolted connections;
" water leaking from the lowest elevation section of vertical lines containing bolted connections; and l
"discolorization or residue on surfaces examined shall be given particular attention to detect evidence of boric acid accumulations from borated reactor coolant leakage."
Evaluation: Paragraph IWA-5242(a) requires the removal of insulation from pressure-retaining bolted connections in borated systems for direct VT-2 visual examination during system pressure testing. The licensee noted that, based on an initial review of drawings and other design documents, the number of bolted connections requiring insulation removal will result in 4 significant increase in man hours, radiation exposure, and materials.
As an alternative to the removal of all insulation for the upcoming refueling outage, the licensee has proposed to visually examine bolted connections, inside and outside of containment, during leakage tests as part of the plant surveillance program. During the outage, removable insulation (suitcase style insulation) will be removed from the reactor coolant and charging systems inside containment, and a VT-2 visual examination of the bolted connection will be performed. Where maintenance requires the removal of j
insulation on the reactor coolant and charging systems inside I
containment, maintenance personnel will look for evidence of leakage.
In addition, after refueling and prior to reactor criticality, the licensee will perform a VT-2 visual examination, following a 4-hour hold time, in conjunction with the'1eakage test
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at nonnal operating pressure. During this walkdown the licensee will look for evidence of leakage under bolted connections, water leaking from the lowest elevation of vertical piping containing bolted connections, and pay particular attention to discolorization or residue on surfaces from borated reactor coolant leakage.
The INEL staff has reviewed the licensee's p:oposed alternatives to removal of insulation at bolted connections in systems borated for the purpose of controlling reactivity. Based on this evaluation, it 5
l is concluded that requiring the licensee to remove insulation on all bolted connections in systems borated for the purpose of controlling reactivity during the upcoming refueling outage would result in a hardship without a compensating increase in safety. The INEL staff believes that the licensee's proposed alternatives for the interim request for relief will provide reasonable assurance of operational 3
readiness. Therefore, it is recommended that interim relief for the currently scheduled refueling outage in September,1995, be authorized pursuant to 10 CFR 50.55a(a)(3)(ii) provided that for the upcoming September,1995, refueling outage, all removable insulation at bolted connections in borated systems is removed and a VT-2 visual examination is performed.
3.0 CONCLUSION
The INEL staff has reviewed the licensee's Request for Relief Number 19 and recommends that interim relief be authorized, pursuant to 10 CFR 50.55a(a)(3)(ii), for the upcoming St. Lucie, Unit 2, refueling outage, currently scheduled for September,1995, provided that all removable insulation is removed from bolted connections in borated systems and a VT-2 visual examination is performed.
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ATTACHNENT 3 1
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SALP INPUT I
i LICENSEE:
Florida Power and L.ight Company FACILITY:
St. Lucie Nuclear Power Plant l
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DOCKET No.:
50-389 TAC NO.:
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LICENSING ACTIVITY: Review of the Second Ten-Year Interval Inservice Inspection Program Plan Request for Relief No. 19 for St. Lucie Nuclear Plant, Unit 2 SulmARY OF ACTIVITY: The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has reviewed and evaluated the information provided by Florida Power and Light Company in its letter dated April 3, 1995, related to the second ten-year interval inservice inspection program plan Request for Relief No. 19 for the St. Lucie Nuclear i
Plant, Unit 2.
I NARRATIVE DISCUSSION OF LICENSEE PERF0=_^.MCE
. FUNCTION AREA - ENGINEERING AND TECHNICAL SUPPORT: The licensee's Engineering and Technical Staff provided adequate technical information regarding its second ten-year interval inservice inspection program plan Request for Relief No. 19. The staff was able to authorize the licensee's alternative on an interim basis based on the information provided by the licensee's staff.
REVIEWER:
T.K. McLellan, 415-2716 DATE:
06/06/95 i
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