I am in my best right here...I am at the apex of doing my job. They were answering in this thorough way because I was representing the New England Coalition. I wish I had an entity whom I could represent to elevated my status to the NRC. They know now i only represent myself.
Along with LaSalle, there was Limerick with their popping open relief valves in this time frame. I got the feeling the NRC was rationalizing away all leaks from the reactors thoughout the nation. This was a group thing with rationalizing away the threat of reactor leaks. SRVs being one, then as a general concern misusing safety systems because of poor maintenance to maintain commercial operation. They could spin a host of rules around so nuclear vessel core leaks didn't matter. Guys, get real, you can spin around any rule to fit any agenda.
These guys were going crazy in the beginning with their pipe dreams of building the new nukes...pushing corporate and plant capacity factor. We were in a nuclear industry tulip mania. I chose the region III because I got wind these guys were in so much trouble. Balls to the walls deregulation fanatic president Bush, who hated all government regulations, just got elected. Clinton was much like president Bush. So I knew the NRC at all through 2001 was spinning out of control.
People who know me think I can see premonitions. I can spin organizations' behaviors in my head and I am good at predicting result. What gets to four srvs leaks on four new srvs valves is deep behavior of a group of peope. I wouldn't have spent so much time on LaSalle or Region III if I din't know they were in big trouble in 2001.
Davis Besse leading fro, 1998 to 2001 was a terribly leaking ship...unbelievably reckless with downplaying a host of reactor vessel leaks to boost capacity factor and support a weakening stock price. Davis Besse leading up to this had very good grades from the NRC...post hole in the head, the NRC admitted they severely inflated the grade of Davis Besse in the lead up to this near miss.
The very serious Davis Besse hole in the reactor head accident discovered in March 2002...
March 12, 2013:
November 29, 2001
(So I added the link to this NRC document.)
Mr. Michael Mulligan
New England Coalition on Nuclear Pollution
5 Woodlawn Lane
Hinsdale, NH 03451
Dear Mr. Mulligan:
Your e-mail dated September 27, 2001, and addressed to Mr. Victor L. Dricks for
Dr. William D. Travers, Executive Director for Operations, has been referred to the Office of Nuclear Reactor Regulation (NRR) pursuant to 10 CFR 2.206 of the Commission’s regulations. A copy of your e-mail and all supplements are enclosed for completeness. Noting your request that the Nuclear Regulatory Commission (NRC) take enforcement action against the LaSalle County Station, Units 1 and 2, (LaSalle), the staff has processed your request following the guidance in Management Directive 8.11, "Review Process for 10 CFR 2.206 Petitions." You requested the following NRC enforcement-related actions:
1. Both units be immediately shutdown for a lengthy maintenance period to replace leaking safety/relief valves (S/RVs).
2. The NRC perform an immediate emergency inspection on the S/RV problems atLaSalle and an assessment of other similar large relief valves at other Exelon Generation Company, LLC, (Exelon, licensee) facilities.
3. The NRC perform a detailed inspection on the suppression pool temperature increases, in-leakage problems, and extended use of the residual heat removal (RHR) system in the suppression pool cooling (SPC) mode during this past summer.As the basis for your request, supplemented by information you provided to Mr. William A. Macon, Jr., on October 3, 2001, you stated the following.
1. NRC Inspection Reports 00-12, 01-02 and 01-03 for LaSalle indicate multiple Unit 1 and Unit 2 S/RVs have excessive internal seat leakage. You further suggest that the NRC displays a "reckless indifference to safety" regarding S/RV leakage and other degraded components which create "unacceptable risks to the surrounding community."
2. NRC Inspection Report 00-11, which references a LaSalle policy directive (LOP-CM-03) that addresses frequent suppression pool cooling and mixing, indicates many years of living with degraded component problems and allowing suppression pool temperature increases up to the 105 Flimit. You further suggest that degraded plant operations are bumping past conservative safety limits and analysis, and that the NRC is being "deceptive" and "amoral" regarding its technical reviews of the industry’s engineering analyses.
3. Operation of the RHR system in the SPC mode is not meant to facilitate normal commercial plant operations. Safety systems are designed to be maintained in a standby state and only run when absolutely necessary. The industry has declared that running these components excessively creates the condition which leads to excessive wear and increasing failures. You further suggest the NRC has become a "one way check valve for the industry" by permitting longer testing timeframes and reduced testing for the nuclear industry’s benefit.You addressed the NRC’s petition review board (PRB) by teleconference on October 12, 2001, to clarify your petition. A transcript of the meeting is enclosed as a supplement to your petition. The results of that discussion have been considered in the PRB's determination regarding your request for immediate action and whether or not the petition meets the criteria for consideration under 10 CFR 2.206. The staff has concluded that your submittal does not meet the criteria for consideration under 10 CFR 2.206 because your petition presents no significant new information and only raises issues that have already been the subject of NRC staff review and evaluation on the LaSalle facility and other similar facilities (e.g., Quad Cities, Fitzpatrick).
Your petition, supplemented by information you provided to Mr. Macon on October 9, 2001, raises additional concerns about potential NRC misconduct. You state that the agency has ignored known problems with leaking S/RVs and RHR reliability at LaSalle and throughout the industry. The staff is treating these concerns as assertions of impropriety by NRC staff and has referred them to the Office of the Inspector General. Although the staff has concluded that your submittal does not meet the criteria for consideration under 10 CFR 2.206, the staff has reviewed the relevant technical issues and has developed the following response to your concerns:
EVALUATION
LaSalle County Station, Units 1 and 2, currently operates in accordance with a set of improved technical specifications (TSs) based on NUREG 1433, Revision 1, "Standard Technical Specifications, General Electric Plants BWR/4," dated April 1995, NUREG-1434, Revision 1, "Standard Technical Specifications, General Electric Plants BWR/6," dated April 1995, and on guidance provided in the Commission’s "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published on July 22, 1993 (58 FR 39132). These technical references, prepared by the NRC staff, have been extensively reviewed by the industry, professional organizations, academic institutions, and the public. The staff prepared the Safety Evaluation (SE) for the LaSalle improved TS conversion in accordance with these references and NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," dated July 1981. On March 30, 2001, the Commission issued Amendment 147 to Facility Operating License No. NPF-11 and Amendment No. 133 to Facility Operating License No. NPF-18 for the LaSalle County Station, Units 1 and 2, respectively. The licensee implemented the improved TS at LaSalle on May 1, 2001, which are currently the licensing basis under which NRC inspectors monitor plant activities, evaluate if NRC requirements are violated, and, if a violation is found, determine its effect on plant safety and risk. Excerpts from the LaSalle TS and bases are enclosed for reference to clarify the staff’s evaluation of your safety concerns.
Safety/Relief Valves (S/RVs)
In your 10 CFR 2.206 petition and in previous communications with the staff regarding LaSalle and other facilities (e.g., Limerick, Prairie Island, Susquehanna), you have expressed concerns regarding the degradation of S/RVs, the amount of leakage which is considered to be acceptable for S/RVs and other large relief valves, and the surveillance testing interval which is acceptable to satisfy ASME Code requirements. The staff has responded to you on several occasions, most recently on September 14, 2001, in response to your June 21, 2001, concerns regarding leakage of large remotely controlled relief valves. The NRC staff considers some leakage to be acceptable without affecting plant operation or safety, and, in fact, all S/RVs may leak without necessarily rendering them mechanically inoperable and incapable of performing their safety functions in the event of a reactor overpressurization event. As long as TS Limiting Condition for Operation (LCO) 3.4.4 and Surveillance Requirement (SR) 3.4.4.1 are satisfied (see enclosed excerpts), no NRC enforcement-related action is warranted. The staff further considers that NRC requirements are being met to ensure S/RVs are monitored and maintained in a condition that ensures they will perform their safety functions.
Safety-related components such as S/RVs are within the scope of the Maintenance Rule (10 CFR 50.65). The Maintenance Rule requires that licensees monitor the performance or condition of components, such as S/RVs and other large relief valves, against licensee established goals commensurate with safety, taking into account industry-wide operating experience. Licensees must take corrective action when these goals are not met. The NRC has determined that the Boiling Water Reactor (BWR) Owners Group and individual licensees at LaSalle and other facilities have significantly improved the performance of three-stage S/RVs and two-stage S/RVs as demonstrated by plant-specific operational experience and test data. The NRC staff does not believe there is a generic problem regarding the operability of S/RVs, despite known component degradation and leakage problems. These are maintenance issues which fall within the scope of the licensee’s maintenance programs and corrective action programs, and not within the scope of direct NRC enforcement. The licensee has indicated that it plans to replace the current hard-seat S/RVs with soft-seat S/RVs during upcoming outages to fix the leakage problems, and the staff is satisfied that these planned corrective actions will be sufficient. The staff continues to monitor the S/RV leakage problems, but neither the NRC staff’s evaluation, nor industry operational data, indicates that the currently installed S/RVs pose a risk-significant safety concern.
Suppression Pool Average Temperature
In your 10 CFR 2.206 petition you have expressed concerns regarding the licensee approaching TS limits on suppression pool temperatures. This is an operating issue which falls within the scope of the licensee’s operating procedures, and not within the scope of direct NRC enforcement. As long as LCO 3.6.2.1 and SR 3.6.2.1.1 are satisfied, no NRC enforcement related action is warranted. TS limits are not safety analysis limits, and approaching a TS limit is within the bounds of acceptable plant operation as long as the limit is not exceeded. NRC inspectors continue to monitor plant activities to ensure NRC requirements are met and plant procedures are followed. There has been no indication that these TS limits have been exceeded or plant procedures violated, nor has there been any other indication to suggest that the current safety analyses are non conservative.
Residual Heat Removal (RHR) Suppression Pool Cooling
In your 10 CFR 2.206 petition, you express concerns regarding the degradation of the RHR system, excessive run times of the RHR subsystems, and the operation of safety systems such as RHR during regular plant operation rather than maintaining them in a standby state. Although there is no TS limit or other licensing restriction on run times for the RHR pumps at LaSalle, there are starting limitations on the pumps and they are required to be run quarterly (every 92 days) in accordance with the inservice testing (IST) program. The pumps are, in fact, designed for extended operation for use during the long term core cooling mode of operation.
As long as LCO 3.6.2.3, SR 3.6.2.3.1 and SR 3.6.2.3.2 are satisfied, no NRC enforcementrelated action is warranted. The staff considers that NRC requirements are being met to ensure RHR suppression pool cooling subsystems are monitored and maintained in a condition that ensures they will perform their safety functions. Safety-related components such as RHR pumps and valves are within the scope of the Maintenance Rule (10 CFR 50.65). Licensees must take corrective action when licensee established goals are not met. The NRC staff does not believe there is a generic problem regarding extended use of the RHR system in the SPC mode, as far as normal system reliability and operability are concerned. Concerns about excessive wear and increased risk of failures of RHR system components are maintenance issues which fall within the scope of the licensee’s maintenance programs and corrective action programs, and not within the scope of direct NRC enforcement. The staff continues to monitor the safety system performance of RHR and other systems, but neither the NRC staff’s evaluation, nor industry operational data, indicate that the currently demonstrated level of performance at LaSalle poses a risk-significant safety concern.
However, the NRC staff shares your concerns about extended use of the RHR system in the SPC mode and the potential for water hammer in the RHR system during a design basis loss of coolant accident (LOCA) coincident with a loss of offsite power (LOOP) while the system is aligned in this mode. This issue has been previously identified in NRC Information Notice (IN) 87-10, "Potential for Water Hammer During Restart of Residual Heat Removal Pumps," dated February 11, 1987, and Supplement 1, dated May 15, 1997. This supplement specifically addresses the increased use of RHR pumps in the SPC mode due to leaking S/RVs. The concern is that during a design basis LOCA coincident with a LOOP, the LOOP, subsequent valve realignment, and large elevation differences may allow portions of the RHR system to drain down to the suppression pool, leaving voids in the RHR piping. When the emergency diesel generators reenergize the emergency buses in response to the LOOP, the RHR pumps will start and possibly cause water hammer damage in the voided RHR loop.
In 1993, NRC inspectors expressed concerns that the licensee had not adequately addressed IN 87-10, and the licensee subsequently performed additional analysis and testing and concluded that the potential for severe water hammer was possible. As a result, Sargent and Lundy performed water hammer analysis EMD-067982, "Evaluation of Potential Water Hammer In Residual Heat Removal System," Revision 0, dated February 18, 1994. This report concluded that although a water hammer would occur, the RHR system would maintain its pressure boundary integrity, structural stability, and functional capability during the water hammer event. NRC inspectors noted that plastic deformation and ovalization of system piping as well as snubber failure were also predicted. These results were subsequently documented in the LaSalle updated final safety analysis report (UFSAR).
In December 1995, General Electric Report NEDC-32513, "Suppression Pool Cooling and Water Hammer," was issued to document the conclusion of a General Electric review of the generic water hammer issue. In that report, the following conclusions were documented:
Operation of the RHR system in the SPC mode has been expected to be an infrequent occurrence during normal operation. As a result, the original LOCA design basis and supporting analysis only assumed initiation of the ECCS/LPCI [Emergency Core Cooling Systems/Low Pressure Coolant Injection] mode to be from a standby configuration.
The frequency of occurrence of a LOOP/LOCA coincident with the RHR system being in the SPC mode is less than the probability of events considered in the design of BWRs (< 1.0x10-6 per year, per ANSI/ANS-52.1, "Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants").Although LOOP/LOCA occurrence during secondary modes of operation (such as SPC mode) may not have been included in the original design basis, the staff has determined that the increased use of SPC mode, possibly beyond the frequency defined as "short operational periods," would require analysis of the event and the corresponding draindown and water hammer. The LaSalle licensing basis contains no specific restrictions regarding the time in which RHR may be operated in the SPC mode and the 1994 water hammer analysis concludes that the RHR safety function will be maintained despite the potential for severe water hammer, which indicates that LaSalle has been operating within currently acceptable limits and analyses.
(The below talking about, I was banging on the NRC before April/May 2001. This got the whole industry to reevaluate running invaluable safety systems to maintain commercial operation)
Due to the number of S/RVs leaking at both LaSalle units during the current operating cycles, and the expected increased use of RHR in the SPC mode during the summer months with elevated ultimate heat sink (UHS) temperatures, NRC inspectors began reviewing the LaSalle water hammer analysis during April/May 2001 and began an iterative series of discussions with the licensee. As late as September 17, 2001, prior to your petition on September 27, 2001, NRC Inspection Report 01-10 notes that the inspectors reviewed selected Operability Evaluations and Condition Reports related to the leaking S/RVs and a licensee management decision to operate one train of the Unit 1 RHR system continuously in the SPC mode, and identified this issue as an Unresolved Item (URI 50-373/2001010-02).
This unresolved item involves regulatory interpretations unrelated to the technical concerns raised in your petition. For example, the staff currently has a concern that the 1994 RHR water hammer analysis does not meet the criteria specified in Appendix F of Section III of the ASME Code. The licensee has commissioned an independent contractor to review the analysis and determine whether the analysis is reasonable to demonstrate system functionality. Additionally, the staff is reviewing the overall adequacy of the LaSalle water hammer analysis and the applicability of the recently revised 10 CFR 50.59 change control process to this issue.
These ongoing discussions primarily involve regulatory interpretations and do not involve any new technical issues which have not already been the subject of NRC staff review and evaluation (e.g., Fitzpatrick in 1996, Quad Cities in 1997). Your petition does not present any significant new information which may be relevant to these discussions. The staff continues to monitor the LaSalle water hammer analysis issue and will employ whatever regulatory actions are appropriate, including enforcement action if warranted.
CONCLUSION
Based on the above, the NRC staff has concluded that your submittal dated September 27, 2001, supplemented by information provided on October 3, 2001, does not meet the criteria for consideration under 10 CFR 2.206 because your petition presents no significant new information and only raises issues that have already been the subject of NRC staff review and evaluation on the LaSalle facility, other similar facilities (e.g., Prairie Island, Susquehanna, Limerick, Fitzpatrick, Quad Cities), and on a generic basis, for which the issues have been resolved and the resolutions are applicable to LaSalle. No NRC enforcement-related action is warranted based upon the information you have presented.
Mr. Mulligan, please understand that if a violation of NRC requirements is found during NRC inspections or brought to the attention of the NRC by either plant personnel or other individuals, there are basically two mechanisms used by the NRC to address the problem based upon its effect on plant safety and risk. If the violation is of very low safety significance, it will be discussed in an inspection report with no formal enforcement action. The utility is expected to deal with the violation through its corrective action program, correcting the violation and taking
steps to prevent a recurrence. If the NRC risk evaluation finds that the violation has a higher risk significance, a Notice of Violation will be issued to the licensee which may or may not involve a civil penalty. A Notice of Violation requires the licensee to respond formally to the NRC with its actions to correct the violation and what steps it will take to prevent the violation from occurring again. Both mechanisms involve a public process and all documentation is available for public review.
In summary, the NRC staff concludes that no violation of NRC requirements exists at the LaSalle County Station, Units 1 and 2, which warrants NRC enforcement-related action.
Your concerns related to excessive leakage of S/RVs, suppression pool temperatures approaching operational limits, and extended use of the RHR system in the SPC mode have been previously addressed and evaluated by the staff. The NRC, therefore, does not intend to review your concerns under the 10 CFR 2.206 petition process for the aforementioned reasons.
Thank you for bringing these issues to the attention of the NRC.
Sincerely,
/RA/
John A. Zwolinski, Director
Division of Licensing Project Management
Office of Nuclear Reactor Regulation
Docket Nos. 50-373 and 50-374
Enclosures: As stated
cc w/Enclosures: See next page
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