IR 05000263/2010006
| ML101230563 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/03/2010 |
| From: | Robert Daley NRC/RGN-II/DRS/EB3 |
| To: | O'Connor T Northern States Power Co |
| References | |
| IR-10-006 | |
| Download: ML101230563 (20) | |
Text
UNITED STATES May 3, 2010
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000263/2010-006(DRS)
Dear Mr. OConnor:
On March 26, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications inspection at your Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection results, which were discussed on March 26, 2010, with you and other members of your staff.
The inspection 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 this inspection, one NRC-identified finding of very low safety significance was identified. The finding involved a violation of NRC requirements. However, because of its very low safety significance, and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
If you contest the subject or severity of a NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, 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 Monticello Nuclear Generating Plant. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Monticello Nuclear Generating Plant. The information that you provide will be considered in accordance with Inspection Manual Chapter 0305. 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/
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket No.50-263 License No.DPR-22
Enclosure:
Inspection Report 05000263/2010006 w/Attachment: Supplemental Information
REGION III==
Docket No: 50-263 License No: DPR-22 Report No: 05000263/2010006 Licensee: Northern States Power Company, Minnesota Facility: Monticello Nuclear Generating Plant Location: Monticello, MN Dates: March 8 - 26, 2010 Inspectors: D. Szwarc, Senior Reactor Inspector (Lead)
R. Langstaff, Senior Reactor Inspector M. Munir, Reactor Inspector Observer: J. Corujo-Sandin, Reactor Engineer Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
IR 05000263/2010006; 03/08/2010 - 03/26/2010; Monticello Nuclear Generating Plant;
Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.
This report covers a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. One Severity Level IV finding was identified by the inspectors.
The finding was considered a Non-Cited Violation (NCV) of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
NRC-Identified
and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Severity Level IV. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)1 for the licensees failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis which addressed room temperature limitations as to why the isolation of a high pressure coolant injection (HPCI) room cooler did not require prior NRC approval. The licensee entered this issue into their corrective action program and determined that no immediate corrective actions were necessary because administrative controls were in place to ensure that the HPCI room temperature would not exceed the calculated initial room temperature limitation.
The inspectors determined that the finding was more than minor because they could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that the finding was of very low safety significance because the finding did not result in loss of operability or functionality. The finding affected the Mitigating Systems cornerstone attribute of Equipment Performance to ensure the availability and reliability of systems (HPCI) that respond to initiating events to prevent undesirable consequences. This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety in that training of personnel was not sufficient.
H.2(b) (Section 1R17.1.b)
Licensee-Identified Violations
No violations of significance were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
.1 Evaluation of Changes, Tests, or Experiments
a. Inspection Scope
From March 8, 2010 through March 26, 2010, the inspectors reviewed two safety evaluations 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 could not review the minimum sample size of six evaluations because the licensee only performed two evaluations during the triennial sample period. The inspectors also reviewed 18 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
- the changes, tests, or experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;
- the safety issue requiring the change, tests, or experiment was resolved;
- the licensee conclusions for evaluations of changes, tests, or experiments were correct and consistent with 10 CFR 50.59; and
- the design and licensing basis documentation was updated to reflect the change.
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.
This inspection constituted two samples of evaluations and 18 samples of changes as defined in IP 71111.17-04.
b. Findings
Failure to Perform 10 CFR 50.59 Evaluation for Isolation of Room Cooler Which Addressed Temperature Limitations
Introduction:
The inspectors identified a Severity Level IV finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR 50.59, Changes, Tests, and Experiments, for the failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment.
Specifically, the licensee failed to provide a basis which addressed room temperature limitations as to why the isolation of a high pressure coolant injection (HPCI) room cooler did not require prior NRC approval.
Description:
The licensee permanently secured cooling water flow to the B division HPCI room air cooling unit on April 2, 2007, during the 2007 refueling outage. The cooler was isolated to compensate for a lack of emergency service water system flow to the B residual heat removal corner room. Section 6.2.4.1 of the Updated Safety Analysis Report (USAR) identifies one of the design bases of the HPCI system as ensuring that adequate core cooling takes place for all break sizes less than those sizes for which the low pressure coolant injection or core spray can adequately protect the core without assistance from other safeguards systems. Prior to this change associated with isolating the room cooler USAR (Revision 23) Section 6.2.4.2.10 stated:
The HPCI space cooling system is composed of two air cooling units, each capable of maintaining HPCI room temperature below 85°F [degrees Fahrenheit]
during normal operation. Each unit is powered from a separate emergency supplied power source. Cooling water is supplied by the service water system, or the emergency service water system upon loss of off-site power. The HPCI space cooling system is therefore unaffected by a loss of off-site power.
The licensee evaluated the change under 10 CFR 50.59 Evaluation 07-001, Revise Plant Documents to Reflect Securing Cooling Water Flow to HPCI Room Cooler V-AC-8B by Closing SW-107-2, Revision 0. The evaluation stated:
The proposed activity of isolating cooling water to the B HPCI room air cooling unit does not change the design functions of the HPCI system listed above, nor does it introduce the possibility of a change in the likelihood of a malfunction because it has been analyzed that the HPCI room coolers are not necessary to maintain the proper temperatures in the HPCI room. Operation of the components and equipment of the HPCI system is assured since they will be operating within their design temperature limits, the most limiting of which is 125°F, during a design basis accident. The HPCI system operability is not impacted.
The licensee documented in their analysis that the HPCI room coolers were not necessary to maintain the proper temperatures in the HPCI room in calculation CA-96-020, HPCI Room Transient Temp, Revision 4. The inspectors reviewed calculation CA-96-020 and determined that the calculation assumed a maximum initial room temperature of 100°F. Based on discussions with licensee engineering personnel, the inspectors determined that during high outside temperature conditions, such as during summer, it was necessary to provide cooling to the HPCI room to ensure that temperatures were maintained below 100°F. The licensee relied upon administrative controls to ensure room temperatures were maintained below 100°F. Specifically, the licensee relied upon operator rounds and a non-safety-related temperature alarm, which annunciated in the control room for verifying that the room temperature was below 100°F. However, Evaluation 07-001 did not address the 100°F initial temperature limitation nor did the evaluation discuss the administrative controls in place to ensure that the 100°F limitation was maintained. Based on discussions with licensee engineering personnel and review of procedure B.03.02-5, HPCI System Operation, Revision 36, the inspectors determined that if the 100°F limitation was exceeded, HPCI would be declared inoperable. Evaluation 07-001 did not demonstrate that there would not be a more than minimal increase in the likelihood of occurrence of a malfunction of the HPCI system because the analysis which the evaluation relied upon (that is, calculation CA-96-020) was only valid under certain conditions (that is, the 100°F limitation) and the evaluation did not address those certain conditions. That is, the evaluation did not provide the bases for why the change did not require a license amendment. Title 10 CFR 50.59(d)(1) requires the licensee to include a written evaluation, which provides the bases for the determination that a change, test, or experiment does not require a license amendment.
The inspectors noted that, based on Evaluation 07-001, the licensee revised Section 6.2.4.2.10 of the USAR to read as follows:
The HPCI space cooling system is composed of two air cooling units, which are not required for HPCI operability. The units are powered from separate emergency power sources and cooling water can be supplied from either the service water system or from the emergency service water system upon a loss of off-site power.
The inspectors noted that the statement that the HPCI space cooling system was not required for operability was not fully accurate in that it did not address the 100°F limitation. In actuality, the HPCI space cooling system was required in order to maintain temperature in the room below 100°F otherwise the licensee would declare HPCI inoperable. Therefore, USAR Section 6.2.4.2.10 was inaccurate as written. In addition, the statement that there were two air cooling units and the discussion of water and electrical supplies was only accurate in a literal sense. Though isolated, the B cooler was physically present in the room. In addition, water and electrical supplies could physically be realigned. However, Procedure B.03.02-05 stated that the B cooler could not provide cooling because it was isolated. The annunciator response procedures for HPCI room temperature alarms also had similar language. In addition, system drawings and alignment procedures showed the cooler as being isolated and not in service. The inspectors concluded that the revised USAR statements were misleading and were not reflective of plant operation.
The inspectors reviewed Procedure FG-E-SE-03, 50.59 Resource Manual, Revision 0, and determined that the procedure listed examples which should have prompted the preparer of Evaluation 07-001 to recognize that the administrative controls for maintaining normal room temperatures in lieu of redundant room coolers needed to be addressed in the evaluation. Specifically, Procedure FG-E-SE-03 listed the following as considerations which may be useful in determining if an activity involves more than a minimal increase in likelihood of malfunction:
- Downgrading the support system performance necessary for reliable operation of the important to safety equipment; and
- Reducing system/equipment redundancy, diversity, or independence.
Additionally, Procedure FG-E-SE-03 described changes involving new or modified operator actions that support a design function as being potentially acceptable provided certain criteria were met. The inspectors considered administrative controls to be a form of operator actions. Based on review of Procedure FG-E-SE-03, the inspectors determined that the preparer of Evaluation 007-01 had been provided adequate procedural guidance. During the inspection, the inspectors observed weaknesses in the understanding of 10 CFR 50.59 requirements for documenting evaluation bases among senior engineering staff, engineering management, and regulatory assurance staff interacting with the team. Consequently, the inspectors concluded that training for preparers of evaluations was inadequate.
The licensee entered this issue into their corrective action program as AR 01223907, Eval 07-001 - insufficient info. w/regard to initial temp, dated March 24, 2010. The licensee determined that no immediate corrective actions were necessary because administrative controls were in place to ensure that the HPCI room temperature would not exceed the 100°F initial room temperature limitation. The licensee planned to review their training and processes for performing 10 CFR 50.59 evaluations.
Analysis:
The inspectors determined that the failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency. The inspectors determined that this issue was a performance deficiency warranting a significance evaluation because the change the licensee made to USAR Section 6.2.4.2.10 was not equivalent to the previous revision of that section and was no longer accurate. The finding was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval.
Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 4a for the Mitigating Systems cornerstone. The finding affected the Mitigating Systems cornerstone attribute of Equipment Performance to ensure the availability and reliability of systems (HPCI) that respond to initiating events to prevent undesirable consequences. The finding screened to Green because the finding did not result in loss of operability or functionality. Specifically, the inspectors determined that the licensee had adequate administrative controls in place to ensure that the HPCI room temperature would not exceed the 100°F initial room temperature limitation. In accordance with Section XIII Supplement I.D.5 of the NRCs Enforcement Policy this finding is being dispositioned as a Severity Level IV finding because it was evaluated as having very low safety significance (Green) using the SDP.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance within the resources component because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety in that training of personnel was not sufficient.
Specifically, the inspectors determined that the training of 10 CFR 50.59 evaluation preparers was not adequate to ensure existing guidance was correctly applied. H.2(b)
Enforcement:
Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)requires the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). Title 10 CFR 50.59(d)(1) requires that these records include a written evaluation, which provides the bases for the determination that the change, test, or experiment does not require a license amendment. Title 10 CFR 50.59(c)(2) requires a licensee to obtain a license amendment prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the final safety analysis report (as updated). The HPCI system was an SSC important to safety previously evaluated in the final safety analysis report (as updated). Section 6.2.4.2.10 of the final safety analysis report (as updated)
(Revision 23) stated the HPCI space cooling system is composed of two air cooling units, each capable of maintaining HPCI room temperature below 85°F during normal operation.
Contrary to the above, from April 2, 2007 to March 26, 2010, the licensee failed to provide a written evaluation, which provided the bases for determining that a change, test or experiment made pursuant to 10 CFR 50.59(c) did not require a license amendment. Specifically, on April 2, 2007, the licensee made a change pursuant to 10 CFR 50.59(c) in that the licensee permanently isolated cooling water to one HPCI room space cooler and relied upon administrative controls to maintain normal HPCI room temperatures below limits assumed in analyses for supporting operation of HPCI under accident conditions. The written evaluation documented by 10 CFR 50.59 Evaluation 07-001 did not provide a basis for determination that the change, test, or experiment would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report (as updated). The evaluation did not determine that changing the USAR from requiring HPCI space cooling to maintain temperature below 85°F to stating that it was not required for HPCI operability was equivalent. Because this violation was of very low safety significance, was not repetitive or willful, and it was entered into the licensees corrective action program as AR 01223907, Eval 07-001 -
insufficient info. w/regard to initial temp, dated March 24, 2010, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV05000263/2010006-01, Failure to Perform 10 CFR 50.59 Evaluation For Isolation of Room Cooler Which Addressed Temperature Limitations).
.2 Permanent Plant Modifications
a. Inspection Scope
From March 8, 2010 through March 26, 2010, the inspectors reviewed nine permanent plant modifications, including two equivalency evaluations that had been installed in the plant during the last three years. This review included in-plant walkdowns for portions of the modified division 1 125 volts direct current (VDC) battery charger number 10; 125 VDC battery number 11 room; emergency diesel generator (EDG) oil storage tank level indicating switch; residual heat removal (RHR) service water strainers; RHR piping insulation; control rod drive (CRD) scram solenoid pilot valve; HPCI; and the HPCI auxiliary lube oil pump.
The modifications were selected based upon risk-significance, safety-significance, and complexity. The inspectors reviewed the modifications selected to determine if:
- the supporting design and licensing basis documentation was updated;
- the changes were in accordance with the specified design requirements;
- the procedures and training plans affected by the modification have been adequately updated;
- the test documentation as required by the applicable test programs has been updated; and
- post-modification testing adequately verified system operability and/or functionality.
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.
This inspection constituted nine permanent plant modification samples as defined in IP 71111.17-04.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From March 8, 2010 through March 26, 2010, the inspectors reviewed 14 corrective action process 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 related to permanent pant 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 inspectors are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Power Uprate Related Inspection Activities
a. Inspection Scope
The inspectors reviewed three plant modifications implemented for the extended power uprate. The review included the replacement of inboard main steam isolation valves and the removal of bricks from the bio-shield because they posed a missile threat. The inspectors also reviewed the electrical portions of a modification allowing operations the ability to throttle the safety-related outboard containment spray valves.
- EC 12463, Remove Bricks from the Bioshield to Improve Margin for Potential Missiles in the Drywell.
b. Findings
No findings of significance were identified.
.2 Inadequate Missile Protection for the EDG System Components
The inspectors identified issues with the licensees protection of EDG components from the effects of a design basis tornado. During the 2009 component design bases inspection (CDBI) the inspectors identified an unresolved item (URI 05000263/2009007-06) regarding the design and licensing basis of the EDG building ventilation system.
The concern was whether the ventilation system had to be protected from the effects of a design basis tornado.
During this inspection, while conducting a walkdown of the roof of the EDG building, the inspectors questioned the licensee on whether the air intakes for the EDGs were designed to withstand a tornado missile strike. The licensee stated that they were not designed for this, but that the EDGs could operate by drawing only inside air should the outside air intakes become clogged or crimped. The inspectors questioned whether the licensee had an analysis to support the operation of the EDGs using inside air to which the licensee replied that they did not. The inspectors were concerned that there would not be an adequate supply of air for combustion in the EDGs and to provide room cooling.
The inspectors identified the following statement in USAR Section E.2.1, which was the plant comparative evaluation with the proposed Atomic Energy Commission (AEC) 70 design criteria:
Those systems and components of reactor facilities, which are essential to prevention of accidents, which could affect the public health and safety or to mitigation to their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect:
- (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and surrounding area; and
- (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.
The inspectors believed that based on this statement, and other sections of the USAR mentioned in the original description of the URI, the licensee had to consider the effects of multiple missile strikes on redundant components of a system or components (EDG)required to mitigate the consequences of an accident. The inspectors believed that a design basis tornado includes a tornado event where multiple missiles are generated that could impact multiple areas of the plant. In response to inspectors concerns the licensee stated that a missile strike that would disable both EDGs was not considered a credible event. The specific concern was whether multiple missile strikes on redundant components of an SSC was part of the licensees licensing and design basis and whether all components of the standby diesel generator system had to be protected from the effects of a design basis tornado. These additional concerns will be added to URI 05000263/2009007-06 for resolution pending discussion with the Office of Nuclear Reactor Regulation.
4OA6 Meetings
.1 Exit Meeting Summary
On March 26, 2010, the inspectors presented the inspection results to Mr. OConnor and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- T. OConnor, Site Vice President
- R. Baumer, Compliance Engineer
- T. Erickson, System Engineering Supervisor
- J. Grubb, Plant Manager
- B. Halverson, Engineering Supervisor
- N. Haskell, Site Engineering Director
- V. Karls, Project Engineer
- J. Ohotto, Engineering Design Manager
- S. Speight, Regulatory Affairs Manager
- E. Watzl, Engineering Supervisor
Nuclear Regulatory Commission
- R. Daley, Chief, Engineering Branch 3, Division of Reactor Safety
- L. Haeg, Resident Inspector
- S. Thomas, Senior Resident Inspector
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened and Closed
- 05000263/2010006-01 NCV Failure to Perform 10 CFR 50.59 Evaluation For Isolation of Room Cooler Which Addressed Temperature Limitations
Discussed
- 05000263/2009007-06 URI Inadequate Missile Protection for the EDG System Components Attachment