ML052070047

Relief Request for Temporary Non-Code Repair of a Shutdown Cooling Pipe

ML052070047
Person / Time
Site:	FitzPatrick
Issue date:	08/09/2005
From:	Richard Laufer NRC/NRR/DLPM/LPD1
To:	Kansler M Entergy Nuclear Operations
Laufer R, NRR, 415-1373
References
TAC MC7544
Download: ML052070047 (11)

Text

August 9, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE (TAC NO. MC7544)

Dear Mr. Kansler:

By letter dated July 9, 2005, Entergy Nuclear Operations, Inc. (Entergy) submitted a relief request, which proposed a temporary repair to a shutdown cooling suction pipe in the Residual Heat Removal (RHR) system. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. Entergy requested relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, subsection IWC-4000. The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a(a)(3)(ii). On July 10, 2005, the NRC staff granted verbal authorization.

As documented in the enclosed safety evaluation, the NRC staff reviewed your submittal and concluded that an ASME Code repair under the existing plant conditions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the temporary non-code repair was authorized.

The NRC also notes that on July 12, 2005, the temporary non-code repair was replaced by an ASME Code repair.

If you have any questions regarding this matter, please contact John Boska, the NRC project manager for FitzPatrick, at 301-415-2901.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosure:

As stated cc w/encl: See next page

August 9, 2005 Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE (TAC NO. MC7544)

Dear Mr. Kansler:

By letter dated July 9, 2005, Entergy Nuclear Operations, Inc. (Entergy) submitted a relief request, which proposed a temporary repair to a shutdown cooling suction pipe in the Residual Heat Removal (RHR) system. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. Entergy requested relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, subsection IWC-4000. The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a(a)(3)(ii). On July 10, 2005, the NRC staff granted verbal authorization.

As documented in the enclosed safety evaluation, the NRC staff reviewed your submittal and concluded that an ASME Code repair under the existing plant conditions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the temporary non-code repair was authorized.

The NRC also notes that on July 12, 2005, the temporary non-code repair was replaced by an ASME Code repair.

If you have any questions regarding this matter, please contact John Boska, the NRC project manager for FitzPatrick, at 301-415-2901.

Sincerely,

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosure:

Safety Evaluation cc w/encl: See next page DISTRIBUTION:

PUBLIC RLaufer JUhle CHolden PDI-1 R/F JBoska TMarsh OGC ACRS TChan WBateman SLittle MMitchell BMcDermott, RI JTsao Accession Number: ML052070047 *Safety Evaluation provided OFFICE PDI-1\PM PDI-1\LA EMCB\SC* OGC PDI-1\SC NAME JBoska SLittle TChan TColburn for RLaufer DATE 8/01/05 8/01/05 7/22/05 8/05/05 8/09/05 Official Record Copy

FitzPatrick Nuclear Power Plant cc:

Mr. Gary J. Taylor Resident Inspector's Office Chief Executive Officer James A. FitzPatrick Nuclear Power Plant Entergy Operations, Inc. U. S. Nuclear Regulatory Commission 1340 Echelon Parkway P.O. Box 136 Jackson, MS 39213 Lycoming, NY 13093 Mr. John T. Herron Ms. Charlene D. Faison Sr. VP and Chief Operating Officer Manager, Licensing Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.

440 Hamilton Avenue 440 Hamilton Avenue White Plains, NY 10601 White Plains, NY 10601 Mr. Theodore A. Sullivan Mr. Michael J. Colomb Site Vice President Director of Oversight Entergy Nuclear Operations, Inc. Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant 440 Hamilton Avenue P.O. Box 110 White Plains, NY 10601 Lycoming, NY 13093 Mr. David Wallace Mr. Kevin J. Mulligan Director, Nuclear Safety Assurance General Manager, Plant Operations Entergy Nuclear Operations, Inc.

Entergy Nuclear Operations, Inc. James A. FitzPatrick Nuclear Power Plant James A. FitzPatrick Nuclear Power Plant P.O. Box 110 P.O. Box 110 Lycoming, NY 13093 Lycoming, NY 13093 Mr. Richard Plasse Mr. Oscar Limpias Manager, Regulatory Compliance Vice President Engineering Entergy Nuclear Operations, Inc.

Entergy Nuclear Operations, Inc. James A. FitzPatrick Nuclear Power Plant 440 Hamilton Avenue P.O. Box 110 White Plains, NY 10601 Lycoming, NY 13093 Mr. Christopher Schwarz Supervisor Vice President, Operations Support Town of Scriba Entergy Nuclear Operations, Inc. Route 8, Box 382 440 Hamilton Avenue Oswego, NY 13126 White Plains, NY 10601 Mr. Charles Donaldson, Esquire Mr. John F. McCann Assistant Attorney General Director, Licensing New York Department of Law Entergy Nuclear Operations, Inc. 120 Broadway 440 Hamilton Avenue New York, NY 10271 White Plains, NY 10601

FitzPatrick Nuclear Power Plant cc:

Regional Administrator, Region I Ms. Stacey Lousteau U.S. Nuclear Regulatory Commission Treasury Department 475 Allendale Road Entergy Services, Inc.

King of Prussia, PA 19406 639 Loyola Avenue Mail Stop L-ENT-15E Oswego County Administrator New Orleans, LA 70113 Mr. Steven Lyman 46 East Bridge Street Ms. Deb Katz, Executive Director Oswego, NY 13126 Nuclear Security Coalition c/o Citizens Awareness Network Mr. Peter R. Smith, President P.O. Box 83 New York State Energy, Research, Shelburne Falls, MA 01370 and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Mr. Travis C. McCullough Assistant General Counsel Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Mr. James H. Sniezek BWR SRC Consultant 5486 Nithsdale Drive Salisbury, MD 21801-2490 Mr. Michael D. Lyster BWR SRC Consultant 5931 Barclay Lane Naples, FL 34110-7306

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST FOR TEMPORARY NON-CODE REPAIR OF A SHUTDOWN COOLING PIPE ENTERGY NUCLEAR OPERATIONS, INC.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

1.0 INTRODUCTION

Entergy Nuclear Operations, Inc. (Entergy or the licensee), the licensee for the James A.

FitzPatrick Nuclear Power Plant (JAF), was planning a repair to a through-wall crack in the shutdown cooling suction pipe in the Residual Heat Removal (RHR) system in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI, subsection IWC-4000. Entergy determined that an ASME Code repair was not possible unless the pipe was drained of water. JAF was in cold shutdown and the shutdown cooling suction pipe was being used for decay heat removal from the reactor fuel. The licensee stated that implementing an alternative method of decay heat removal in cold shutdown to allow draining the pipe would present significant hardship and a challenge to safe plant operation without a compensating increase in the level of quality and safety.

By letter dated July 9, 2005, Entergy submitted a relief request which proposed a temporary repair, which did not fully conform to the ASME Code, to the shutdown cooling suction pipe.

The request was made pursuant to Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.55a(a)(3)(ii). Entergy requested relief from the ASME Code,Section XI, subsection IWC-4000. By letter dated July 10, 2005, Entergy submitted additional information as requested by the Nuclear Regulatory Commission (NRC) staff during a telephone conference on July 10, 2005. On July 10, 2005, the NRC staff granted verbal authorization for the temporary non-code repair. This evaluation addresses the merits of the requested relief pursuant to 10 CFR 50.55a(a)(3)(ii).

2.0 REGULATORY EVALUATION

The inservice inspection of the ASME Code, Class 1, Class 2, and Class 3 components in nuclear plants is to be performed in accordance with the ASME Code,Section XI, and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Paragraph 10 CFR 50.55a(a)(3) states: "Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f),

(g), and (h) of this section or portions thereof may be used when authorized by the Director of Enclosure

the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that: (i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty 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) will meet the requirements, except 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 10-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 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 inservice inspection ASME Code of record for FitzPatricks current 10-year inservice inspection interval is the 1989 edition of the ASME Code,Section XI, no addenda.

3.0 TECHNICAL EVALUATION

3.1 ASME Code Component Affected The licensee requested relief for the common suction header portion of the shutdown cooling piping within the RHR system. It is classified as ASME Class 2, Examination Category C-H, Item C7.30 and C7.40, as shown in Table IWC-2500-1 of the 1989 edition of the ASME Code,Section XI.

3.2 Applicable ASME Code Edition and Addenda The ASME Code of Record for the FitzPatrick Repair/Replacement Program is the 1989 edition, no addenda, of the ASME Code,Section XI.

3.3 Applicable ASME Code Requirement The licensee requested relief from the requirements of Article IWC-4000 of the 1989 edition of the ASME Code,Section XI, which states that the rules of IWA-4000 apply. Article IWA-4000 specifies ASME Code-acceptable repair methods for service-related piping flaws that exceed ASME Code acceptable limits. An ASME Code repair is required to restore the structural integrity of the flawed ASME Code piping, independent of the operational mode of the plant when the flaw is detected.

3.4 Licensees Reason for Request The licensee stated that the plant is in Mode 4, Cold Shutdown, with the RHR system lined up in the shutdown cooling mode. Although the shutdown cooling suction line is degraded and the system has been declared inoperable, it is currently performing the required decay heat removal function. A temporary support has been installed nearby the crack area to reduce the static loading on the affected pipe support. In order to perform an ASME Code repair, the

shutdown cooling suction line would need to be removed from service, thus requiring use of an alternative means of decay heat removal.

The licensee stated that the most preferable alternative means of heat removal in cold shutdown is to use the main steam line drains and the main condenser. Using this method, reactor vessel level is raised to the main steam line elevation and reactor coolant water flows to the main condenser via the main steam line drains. The reactor coolant is cooled by circulating water pumped through the main condenser tubes and returned to the reactor vessel via the condensate and feed water systems. This alternative method presents the fewest challenges because it uses the normal feedwater connections and has little impact on reactor water chemistry. However, this method will not be effective because based on the licensees calculations, it can only remove about half the current decay heat being produced by the reactor fuel assemblies.

Besides the above cooling method, the licensee stated that as required by the plant Technical Specifications (TSs) alternate means of decay heat removal are available if needed to maintain the plant in the safe shutdown condition. However, the use of these alternate methods of decay heat removal presents various levels of operational challenges and hardships.

The first alternative method is to use the core spray system and the low pressure coolant injection system in conjunction with the safety relief valves. This alternative method would effectively remove the decay heat, and is credited in the accident analysis of the Updated Final Safety Analysis Report for this purpose. In this alternative method, the RHR system is lined up in the low pressure coolant injection mode and suction is taken from the torus. Torus water is injected into the reactor vessel and returned to the torus for cooling via the safety relief valves.

However, this method is undesirable due to the upset in normal reactor water chemistry caused by the introduction of torus water into the reactor vessel. This flow path may also introduce particulates from corrosion (commonly known as crud) into the safety relief valves which may result in damage to the valve seats. Industry operating experience demonstrates that the service life of the safety relief valves may be negatively affected with this method. Due to the high flow rate associated with the low pressure coolant injection system, this method also presents a challenge to the plant operators in controlling the heat-up and cooldown rates.

The second alternative is to use the feedwater and condensate systems in conjunction with the safety relief valves to provide a flow of cool water through the reactor vessel. This method has less impact on reactor water chemistry than the previously described method using torus water, but the potential impact on the safety relief valves is the same. In addition, use of this method would require processing and discharging thousands of gallons of liquid radwaste.

The third alternative is to use the decay heat removal system connected to the spent fuel pool.

This method requires removing the reactor vessel head and flooding the refueling cavity, which provides a path for water to flow from the vessel to the spent fuel pool. Use of this method would require extended use of the shutdown cooling system in its degraded condition until the reactor vessel could be flooded up, the vessel head, vessel steam dryer and steam separator removed, the refueling cavity flooded, and conditions for the start-up of decay heat removal established. These conditions include installation of a non-safety related diesel. The disadvantage of this alternative is that it would incur radioactive dose of about 10 to 15 person-rem for reactor disassembly and reassembly.

The licensee concluded that the above alternatives present challenges and hardships due to the impact on equipment and plant operations without any gain in safety or quality, whereas the proposed non-ASME Code repair allows for expedited restoration of the piping integrity with no operational hardships.

3.5 Proposed Alternative and Basis for Use The proposed non-ASME Code repair consists of stress relieving the crack and applying a seal weld to the crack surface. In addition, a temporary pipe support has been installed within 5 feet of the affected pipe support to reduce the static loading on the crack. The licensee has evaluated the adequacy of the affected pipe support, assuming no weld in the repair area, to continue to act as a seismic support and found it to be acceptable for all loading conditions.

After the repair, the licensee will visually inspect the temporary non-ASME Code repair area once a day while the shutdown cooling line is in operation to ensure the leak tightness of the weld and piping is maintained.

The licensee has analyzed the non-ASME Code weld repair, including all loading conditions to demonstrate the operability of the piping in a degraded condition. This will support plant mode changes upon completion of the torus repair and pressure testing. The licensee will then proceed with plant heatup into Mode 3 (Hot Shutdown), thus allowing the shutdown cooling system to be removed from service. Decay heat removal will be accomplished by steaming through the main steam lines to the main condenser. An ASME Code repair in accordance with the requirements of IWC-4000 will be performed when this section of the shutdown cooling line can be removed from service and prior to entering Mode 2 (Startup). Pressure test of the ASME Code repair will be performed in accordance with IWC-5000 requirements.

3.6 Duration of Proposed Alternative The licensee requested this non-ASME Code repair to be a one-time, temporary relief from the requirements of IWC-4000. The non-ASME Code repair would be immaterial as soon as the ASME Code repair is performed on the crack. It is anticipated that the ASME Code repair would be completed a few days after the non-ASME Code repair.

3.7 Staffs Evaluation Based on the above discussion, the NRC staff agrees with the licensee that performing an ASME Code repair of the degraded shutdown cooling piping when the piping is needed for decay heat removal would result in hardship or unusual difficulty.

The affected piping is ASME Code, Class 2, which was designed to the 1967 edition through 1969 addenda of the United States of America Standards Institute (now American National Standards Institute) B31.1 Code. The piping and associated supports are designed for pressure, deadweight, thermal, operating and design-basis earthquake loading, and torus-attached piping dynamic loads. The piping is fabricated with carbon steel. The pipe is 20-inch in nominal diameter with a wall thickness of 0.375 inches. The system operating pressure is 50 psig and design pressure is 150 psig with an operating temperature of 90 to 280 degrees F.

The crack is 6.5 inches in length, and follows the toe of the pipe support trunnion-to-pipe fillet weld. The integral attachment weld is a 3/8-inch fillet weld, continuous around the 6-inch diameter (with 6.625-inch outside diameter) pipe trunnion for a total length of 21 inches with the crack comprising about 31 percent of the total weld length.

The crack extent was determined based on a VT-1 visual examination and confirmed by a fluorescent wet-magnetic surface examination for about 60 percent of the weld based on accessibility. The remaining third of the weld (back side of the trunnion) opposite of the crack was verified to have no indications based on a penetrant examination.

The licensee will drill a hole at both ends of the crack to relieve the stresses in the crack. The NRC staff finds that this is a generally accepted method of crack arrest to prevent the crack from further propagation.

The licensee will use the shielded metal arc welding process with a 3/32-inch E7018 electrode to deposit weld metal on the crack surface. The seal weld deposit is anticipated to require three beads laid adjacent to each other which will produce 1 layer of weld about 1/8-inch thick.

The licensee stated that the purpose of the temporary seal weld is to prevent further leakage until an ASME Code repair can be performed. The NRC staff agrees with the licensee that the seal weld is to prevent leakage. However, it also provides certain bonding on the crack face to minimize any potential crack propagation. The staff noted that the seal weld would not provide the same structural integrity to the affected piping as the ASME Code repair. Nevertheless, the seal weld would provide a certain level of structural integrity to the affected piping to minimize crack propagation.

As part of the non-ASME Code repair, the licensee has modified the existing pipe supports in the affected area of the pipe. The trunnion of the affected pipe support (PFSK-2285) is designed as a two way vertical support. About 2 feet away is a deadweight support (PFSK-2084) that was found to be misadjusted. This resulted in the PFSK-2285 pipe support carrying additional deadweight load for which it was not designed. PFSK-2285 also supports vertical dynamic loads (seismic and torus-attached piping loads) and a small downward thermal load.

The normal piping vibration was attributed as the primary factor in the apparent high cycle fatigue failure near the trunnion-to-pipe attachment weld. The localized vibration at the PFSK-2285 pipe support has been reduced by restoring the adjacent pipe support (PFSK-2084) by installing a shim plate where the support was not load-bearing as designed. The licensee stated that vibration is no longer a concern for the short term at the degraded pipe condition.

The NRC staff agrees with this assessment.

In addition, the properly shimmed PFSK-2084 pipe support will reduce the deadweight load that the PFSK-2855 pipe support has been carrying. As a preventive measure, the licensee installed a temporary deadweight support under the pipe about 5 feet away from the PFSK-2285 support and on the opposite side of the PFSK-2084 support. This means that two pipe supports are straddling the degraded PFSK-2285 support which will reduce the loading on the crack and thus minimize crack propagation.

The licensee stated that its analysis showed that the fillet weld could be reduced 50 percent in length and still provide adequate load capacity for the design conditions. With the additional temporary support, the loading on the fillet weld would be reduced which would provide a favorable condition for the affected piping prior to the ASME Code repair.

The NRC staff finds that the temporary non-ASME Code repair is acceptable because (1) the licensee has performed appropriate crack arrest; (2) the seal weld applied to the crack opening will minimize further crack propagation; and (3) the crack will be stabilized and will not experience much loading during the non-ASME Code repair condition because a temporary support is installed adjacent to the PFSK-2285 support and the existing PFSK-2084 support has been properly shimmed.

In addition, in the July 10, 2005, letter, the licensee provided a regulatory commitment, which states that ENO [Entergy Nuclear Operations, Inc.] will complete the ASME Code repairs and required inspections to the RHR SDC [shutdown cooling] piping prior to startup (entry into Modes 1 or 2) from the current forced outage. This commitment was completed prior to plant startup.

4.0 CONCLUSION

Based on the review of information submitted, the NRC staff has determined that the licensee has demonstrated that compliance with the applicable ASME Code repair/replacement requirements for the degraded shutdown cooling piping of the RHR system would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The staff finds that the licensees non-ASME Code repair as discussed in relief request No.

RR-38, will provide an acceptable level of structural integrity to the degraded shutdown cooling piping prior to the ASME Code repair. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the licensees proposed non-ASME Code repair of the common suction header of the shutdown cooling piping of the RHR system at JAF.

All other requirements of Section XI of the ASME Code for which relief has not been specifically requested remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: John Tsao, John Boska Date: August 9, 2005