IR 05000266/2007007
| ML080230532 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/22/2008 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Mccarthy J Florida Power & Light Energy Point Beach |
| References | |
| IR-07-007 | |
| Download: ML080230532 (22) | |
Text
ary 22, 2008
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000266/2007007(DRS); 05000301/2007007(DRS)
Dear Mr. McCarthy:
On December 14, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed a combined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Point Beach Nuclear Plant, Units 1 and 2. The enclosed report documents the results of the inspection, which were discussed with Mr. J. McCarthy, and others of your staff at the completion of the inspection on December 14, 2007.
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of the inspection, one NRC identified finding of very low safety significance was identified, which involved a violation of NRC requirements.
However, because this violation was of very low safety significance and because it was entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of an NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Point Beach Nuclear Plant.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS),
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27 Enclosure: Inspection Report 05000266/2007007(DRS); 05000301/2007007(DRS)
w/Attachment: Supplemental Information cc w/encl: M. Nazar, Senior Vice President and Nuclear Chief Operating Officer J. Stall, Senior Vice President and Chief Nuclear Officer R. Kundalkar, Vice President, Nuclear Technical Services Licensing Manager, Point Beach Nuclear Plant M. Ross, Managing Attorney A. Fernandez, Senior Attorney K. Duveneck, Town Chairman Town of Two Creeks Chairperson Public Service Commission of Wisconsin J. Kitsembel, Electric Division Public Service Commission of Wisconsin State Liaison Officer
SUMMARY OF FINDINGS
IR 05000266/2007007(DRS); 05000301/2007007(DRS); 11/26/2007 through 12/14/2007; Point
Beach Nuclear Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.
The inspection covered a two week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three regional based engineering inspectors. One Green Non-Cited Violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow,
Red), using Inspection Manual Chapter 0609, Significance Determination Process (SDP).
Findings for which the SDP did not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3; dated July 2000.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,
Design Control, that was of very low safety significance involving a calculation that designed the Unit 1 Steam Generator Blowdown (SGBD) Heat Exchanger (HX) Platform to withstand a load from a postulated pipe whip of the condensate return line resulting from a High-Energy Line Break (HELB). The load from a postulated pipe whip applied to the platform was evaluated in calculation PBNP-994-10-S01, SGBD HX Platform Mod.
For Addition of Pipe Rupture Restraint for Condensate Return Line which was approved on April 28, 2007. As a result of this calculation, the design function of the Unit 1 SGBD HX Platform was revised to hold and maintain the steam generator blowdown heat exchangers and condensate return line in position and assure that the platform did not fall onto the safety related Refueling Water Storage Tank (RWST) during a safe shutdown earthquake and a HELB simultaneously. Specifically, the licensee failed to correctly use the original design anchor bolt safety factor in the supporting calculation.
This issue was entered into the licensees corrective action program as condition report CAP 1118144.
The issue was more than minor because the calculation error would be expected to necessitate extensive calculation rework and possibly a modification in order to demonstrate that the platform meets design acceptance limits commensurate with those applied to original design. The finding screened as having very low safety significance (Green) because the inspectors answered yes to question 1 under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. Specifically, the platform remained operable but degraded. The cause of the finding was related to the cross-cutting element in Human Performance, Work Practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (item H.4(c) of IMC 0305). The licensee had failed to correctly use the original design anchor bolt safety factor in all three revisions of the design basis calculation. (Section 1R02.1.b.1)
Licensee-Identified Violations
No findings of significance were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments (71111.02, TI 2515/166)
.1 Review of 10 CFR 50.59 Evaluations and Screenings
a. Inspection Scope
From November 26, 2007 through December 14, 2007, the inspectors reviewed six safety evaluations (SEs) performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate.
The inspectors also reviewed 12 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is included as an attachment to this report.
The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations, and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.
b. Findings
b.1 Incorrect Factor of Safety Specified in Design Evaluation of Unit 1 SGBD HX Platform
Introduction:
The inspectors identified an NCV having very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control. Specifically, the inspectors identified that the design bases calculation for the Unit 1 SGBD HX Platform failed to use the correct anchor bolt safety factor.
Description:
The design function of the Unit 1 SGBD HX Platform from initial construction to 2006 was to hold and maintain the steam generator blowdown heat exchangers in position and assure that the platform did not fall onto the safety related Refueling Water Storage Tank (RWST) during a safe shutdown earthquake. The platform was a braced steel frame with the columns at each corner anchored to the floor at their base. The design function of the RWST was to provide borated water to the safety injection pumps, the residual heat exchangers pumps and the containment spray pumps during a loss-of-coolant accident or a steam line break accident. The RWST was safety related and designed as Seismic Class I. Screening 2007-0066 Installation of Pipe Whip Restraint per Mod EC 10533, applied to the at risk installation and the controls for acceptance and use of a whip restraint for the Unit 1 condensate return line piping. The whip restraint involved installation of steel beams on the existing SGBD HX platform. The whip restraint steel beams were supported from this platform and were attached to the north flange face of two columns. The platform was designed to withstand a load from a postulated pipe whip of the condensate return line resulting from a High-Energy Line Break (HELB). The load from a postulated pipe whip that the whip restraint adds to the platform was evaluated in calculation PBNP-994-10-S01, SGBD HX Platform Mod. For Addition of Pipe Rupture Restraint for Condensate Return Line which was approved on April 28, 2007. As a result of this calculation, the design function of the Unit 1 SGBD HX Platform was revised to hold and maintain the steam generator blowdown heat exchangers and condensate return line in position and assure that the platform did not fall onto the safety related Refueling Water Storage Tank (RWST) during a safe shutdown earthquake and a HELB simultaneously. The inspectors identified a non-conservative technical error in calculation PBNP-994-10-S01.
The inspectors identified that the calculation evaluated acceptability of the anchor bolts based on a factor of safety of 2 under the faulted load condition. The acceptance criteria for anchor bolt design established in the design basis calculation was in accordance with Specification 10447-P500(Q) Technical Specification for Inspection and Testing Concrete Expansion Bolts for Seismic Category 1 Pipe Supports for the Point Beach Generating Station. The requirement in the specification was that shell or wedge type anchor bolts had a factor of safety greater than or equal to four based on the ultimate bolt capacity. The licensee determined that the calculation PBNP-994-10-S01 specified the operability limit for the anchor bolt allowable instead of the design limit. The operability limit only required that shell or wedge type anchor bolts had a safety factor equal to two based on the ultimate bolt capacity. The use of the operability limit did not meet design requirements.
The inspectors also identified non-typical computer modeling used in the calculation and the as built design of the welded connection for attachment of the pipe whip restraint steel to the platform steel columns. Specifically, the inspectors identified that the whip restraint base plates were welded to the platform columns using a single line weld along one vertical edge of the baseplate. This condition was not reflected in the evaluation which models these attachments as simple pin connections without any eccentricity.
The line of action of the loading was offset or eccentric from the center of gravity of the single line fillet weld connection. When the applied forces do not pass through the center of gravity of a weld configuration, a moment is created. The moment was equal to the force times the eccentric length. The computer model of this pin connection did not consider these induced moments. The offsets or eccentricities also produce potential additional torsional stresses on the existing column which were not considered in the calculation. The calculation therefore did not account for the effects of the connection eccentricity on the welds, the whip restraint steel and the platform column steel. In addition, the calculation did not provide the basis for treating welded connection as pin connection.
The licensee agreed that the connection was not typical of code design. In addition, the design basis calculation did not consider the effects of base plate flexibility and prying action in the calculation of anchor bolt loads. The baseplate was analyzed as a rigid plate but this assumption was never validated in the analysis. As a result, the inspectors could not be sure the modeling was adequate without the licensee performing confirmatory calculations that addressed these concerns. The licensee planned to revise this calculation to address the anchor bolt factor of safety error and as part of that effort to re-evaluate with respect to these other calculational concerns to verify whether the current design is adequate. This issue was entered in the licensees corrective action program as CAP 1118200.
Analysis:
The inspectors determined that the failure to use correct anchor bolt safety factor in the design was a performance deficiency and a finding. The mitigating systems cornerstone was affected as the failure of the Unit 1 SGBD HX Platform could result in the failure of the RWST to fulfill its design function. No other cornerstones were determined to be degraded as a result of this issue. The finding was determined to be greater than minor because the calculation error would be expected to necessitate extensive calculational rework and possibly a modification to ensure that the platform met design acceptance limits.
The finding screened as having very low significance (Green) using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for the AtPower Situations, because the inspectors answered yes to question 1 under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. The inspectors agreed with the licensees position that the platform was operable but degraded.
Therefore, the inspectors concluded that the finding did not represent an actual loss of safety function, and the issue screened out as having a very low safety significance or Green.
The cause of the finding is related to the cross-cutting element in Human Performance, Work Practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (item H.4(c) of IMC 0305). Specifically, the licensee had failed to correctly use the original design anchor bolt safety factor in all three revisions of the design basis calculation. Each calculation revision was performed by a different contractor and failed to identify this non-conservative technical error.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Contrary to the above, the performance of design reviews was inadequate, in that design calculation PBNP-994-10-S01, SGBD HX Platform Mod. For Addition of Pipe Rupture Restraint for Condensate Return Line did not demonstrate the Unit 1 SGBD HX Platform will maintain structural integrity and assure that the platform will not fall onto the safety related Refueling Water Storage Tank (RWST) during a safe shutdown earthquake and a HELB simultaneously. Because the violation was of very low safety significance and the licensee entered the violation into their corrective action system as condition report CAP 1118144, this violation is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000266/2007007-01(DRS); NCV 05000331/2007007-01(DRS))
1R17 Permanent Plant Modifications
.1 Review of Permanent Plant Modifications
a. Inspection Scope
From November 26, 2007 through December 14, 2007, the inspectors reviewed six permanent plant modifications (six samples) that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. As per inspection procedure 71111.17B, one modification was chosen that affected the barrier integrity cornerstone. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements, and the licensing bases, and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.
The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA2 Other Activities
.1 (Open) Temporary Instruction (TI) 2515/166, Pressurized Water Reactor Containment
Sump Blockage (NRC Generic Letter 2004-02)
Generic Letter (GL) 2004-02 requested that licensees to perform a mechanistic evaluation of the potential for the adverse effects of post-accident debris blockage and operation with debris-laden fluids to impede or prevent the recirculation functions of the Emergency Core Cooling (ECCS) systems and Containment Spray (CSS) systems following all postulated accidents for which the recirculation of these systems is required.
The generic letter also requested that addressees implement any needed plant modifications. The purpose of the TI 2515/166 is to verify the licensee has implemented plant modifications and procedure changes committed to by the licensee in their GL 2004-02 responses. As discussed in Section 1R02 and 1R17 of this report, the inspectors reviewed the licensees modifications and 10 CFR 50.59 safety evaluations.
Related documents were reviewed in IR 2006013 and IR 2007003. The inspectors did not identify any violations with NRC regulations and as required by the TI, addressed the questions below:
a. Did the licensee implement the plant modifications and procedure changes committed to in their GL 2004-02 responses?
The inspectors reviewed the licensees responses and identified the following commitments in Nuclear Management Company, LLC (NMC) letter L-HU-05-004 dated March 7, 2005:
NMC will perform latent debris sampling at Point Beach Nuclear Plant (PBNP),
Unit 1, during the Fall 2005 refueling outage and at Unit 2 during the Spring 2005 refueling outage.
The licensee completed these commitments in October, and April 2005 respectively.
The inspectors reviewed the licensees responses and identified the following commitments in NRC 2005-0109 dated September 1, 2005:
NMC will evaluate and modify as appropriate the PBNP Unit 1 and Unit 2 Emergency Core Cooling (ECCS) systems to support long-term decay heat removal and resolve the issues identified in GL 2004-02 by December 31, 2007.
The inspectors determined that the licensee completed the sump modifications for Unit 1 and Unit 2 in April, 2007 and November, 2006 respectively. The licensee has requested to extend the final submittal date to June 30, 2008 in FPL Energy letter NRC 2007-0085.
NMC will update the PBNP licensing basis to reflect the results of the analyses and modifications performed to demonstrate compliance with the regulatory requirements of GL 2004-02. This update will be performed in accordance with 10 CFR 50.71.
The inspectors verified that the June 2007 Updated Final Safety Analysis Report (UFSAR) has been updated to reflect the new sump strainer design and containment coating program requirements.
NMC will establish administrative controls at PBNP to have proposed insulation changes inside containment reviewed and approved by engineering. This will ensure insulation upgrades, repairs, and replacements do not result in an unanalyzed debris mix or quantity.
These controls will be established prior to the beginning of the spring 2007 (Unit 1) refueling outage.
The inspectors reviewed procedure NP 7.2.28, Containment Debris Control Program Revision 2 and verified that it implements this commitment.
4. NMC will provide a separate submittal to update the responses to
requests (d)(i) through (d)(iii), and (d)(v) through (d)(vii) within 60 days of acceptance of the final screen design. Acceptance of the final screen designs by the PBNP Plant Oversight Review Committee (PORC) is scheduled for February 15, 2006 (Unit 2) and June 22, 2006 (Unit 1).
The licensee updated the commitment date for this item to Dec 31, 2007 in NMC letters NRC 2006-0077 and NRC 2006-0092. This item had not been completed by the end of the inspection.
b. Has the licensee updated its licensing bases to reflect the corrective actions taken in response to GL 2004-02?
The inspectors verified that the June 2007 Updated Final Safety Analysis Report (UFSAR) has been updated to reflect the new sump strainer design and containment coating program requirements. The licensee stated that any further changes associated with the final submittal will be included in the June 2008 UFSAR update.
c. If the licensee or plant has obtained an extension past the completion date of this TI, document what actions have been completed, what actions are outstanding, and close the TI for the plant that has the extension.
The licensee has requested to extend the final submittal date to June 30, 2008 in FPL Energy letter NRC 2007-0085, dated November 16, 2007. This extension was to allow for developing, performing and documenting additional chemical effect testing.
This TI is open pending further NRC review.
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From November 26 through December 14, 2007, the inspectors reviewed six Corrective Action Process (CAP) documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions (CA) related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. J. McCarthy and others of the licensees staff, on December 14, 2007. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary.
No proprietary information was identified.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- B. Scherwinski, Regulatory Affairs
- F. Flentje, Licensing Supervisor
- K. Locke, Regulatory Assurance
- L. Peterson, Design Engineer Manager
- P. Wild, Design Engineering Projects Supervisor
- B. Woyak, Design Engineering Supervisor
- R. Chapman, Design Engineering
Nuclear Regulatory Commission
- R. Ruiz, Resident Inspector
Attachment
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000266/2007007-01 NCV Incorrect Factor of Safety Specified in Design Evaluation of Unit 1 SGBD HX Platform
Discussed
None Attachment