ML13234A360
| ML13234A360 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/22/2013 |
| From: | Bill Dean Region 1 Administrator |
| To: | Springer D State of VT, Dept of Public Service |
| Dean W | |
| References | |
| EDATS: RegionI-2013-0198 | |
| Download: ML13234A360 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PENNSYLVANIA 19406-2713 August 22, 2013 Mr. Darren Springer Deputy Commissioner Department of Public Service State of Vermont 112 State Street Montpelier, VT 05620-2601
Dear Mr. Springer:
On behalf of the U.S. Nuclear Regulatory Commission (NRC), I am responding to your July 26, 2013, letter regarding an issue involving spurious spikes of certain radiation monitors at Vermont Yankee. Specifically, you requested: (1) an understanding of the information Vermont Yankee provided to the NRC with respect to these incidents; (2) the age and functionality of the radiation monitors; and (3) an understanding of why Vermont Yankee made a 60-day report under NRC regulations versus an immediate report. As you are aware, the NRC briefed members of your staff on the details related to this issue via a teleconference call on August 1, 2013.
As we discussed on the call, although the radiation monitoring system remained operable, the system did produce a false high (spurious spiking) reading. The NRCs review of these issues is currently ongoing. To date, we have determined that upon receipt of the actuation signal, plant operators followed procedures to verify that safety equipment responded as designed, and monitored plant parameters for adverse trends. Although these monitors produced an invalid safety system actuation, the plant responded as designed, the safety system remained operable, no increased radiation levels were detected, and the equipment defaulted to a safe condition to protect the public health and safety. We note that on August 19, 2013, the A refuel floor radiation monitor again experienced spurious spiking similar to the previous incidents on June 14 and July 11, 2013, that resulted in an invalid primary containment isolation system (PCIS) Group 3 actuation signal. Again, the plant operators responded appropriately and the resident inspectors were notified of the occurrence. As required by our regulations, this latest incident along with two additional incidents in July will result in 60-day reports similar to the one referenced in your letter and enclosed with this letter. The licensee has captured this most recent event in their corrective action program and we will evaluate their efforts to assess and resolve this matter, as well.
Regarding the age of these radiation monitors, our inspections to date have determined that three of the four radiation detectors involved were installed in 2012 and the fourth was installed in 2010. The most recent spiking incident occurred on a detector that was replaced in the past month. The licensee has a preventive maintenance program that replaces the detectors on a seven year frequency. Part of the licensees evaluation of this matter will be focused on why the monitors are experiencing spurious spiking. Since these monitors are assessed through the NRCs reactor oversight process, which is the NRCs ongoing inspection and oversight program
D. Springer 2
that verifies the licensees compliance with NRC regulations, they would not be part of our review of Vermont Yankees application for license renewal.
Finally, the reporting requirements applicable to this incident are located in Title 10 of the Code of Federal Regulations (10 CFR) 50.72 and 10 CFR 50.73. Our regulations are constructed so that immediate reporting requirements (10 CFR 50.72) are reserved for issues where immediate NRC action may be required to protect the public health and safety. This incident was properly reported within the 60-day reporting requirements as defined in 10 CFR 50.73. We have enclosed the licensees official 60-day notification (Event Notification 42911) for the first two invalid PCIS Group 3 actuations and a copy of our reporting regulations (10 CFR 50.72 and 10 CFR 50.73) for your information. Note that prior to that official notification, our resident inspectors were aware of the invalid PCIS Group 3 actuations through their ongoing reviews of the licensees condition reporting system and the main control room narrative logs which the residents review on a daily basis in accordance with the reactor oversight process.
Thank you for expressing your concern and should you have any further questions, please do not hesitate to contact me or Ms. Nancy McNamara, Region I State Liaison Officer, at (610) 337-5337.
Sincerely,
/RA/
William M. Dean Regional Administrator
Enclosures:
- 1. Event Notification Report
- 2. 10 CFR 50.72
SUNSI Review x
Non-Sensitive Sensitive x
Publicly Available
Non-Publicly Available OFFICE klm RI/DRP RI/DRP RI/DRP RI/SLO RI/ORA NAME SShaffer/SWS FBower/FLB DRoberts/DJR NMcNamara/DT for WDean/WMD DATE 8/14/13 8/14/13 8/20/13 8/14/13 8/22/13
Event Text Power Reactor Event Number: 49211 Facility: VERMONT YANKEE Region: 1 State: VT Unit: [1] [ ] [ ]
RX Type: [1] GE-4 NRC Notified By: BENJAMIN EGNEW HQ OPS Officer: DONG HWA PARK Notification Date: 07/24/2013 Notification Time: 07:55 [ET]
Event Date: 06/14/2013 Event Time: 18:06 [EDT]
Last Update Date: 07/24/2013 Emergency Class: NON EMERGENCY 10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION Person (Organization):
ANTHONY DIMITRIADIS (R1DO)
Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode 1
N Y
100 Power Operation 100 Power Operation INVALID PRIMARY CONTAINMENT ISOLATION SYSTEM ACTUATION DUE TO A RADIATION MONITOR SPIKE "This notification is being made in accordance with 10CFR50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Primary Containment Isolation System (PCIS) Group 3 actuation signal that affected containment valves in more than one system.
"On 6/14/2013, and again on 7/11/2013, with the reactor at 100% power, an invalid PCIS Group 3 actuation occurred from a momentary spike of the 'A' Refuel Floor radiation monitor which reached the instrument's high radiation trip setpoint. A radiation protection technician was dispatched to the refuel floor and dose rates in the vicinity of the 'A' radiation monitor detector were verified to be normal and below the alarm setpoints. The radiation monitor was verified to be indicating normal expected radiation levels. Subsequent visual inspection and functional checks of the radiation monitors were completed satisfactory and the instrument channel was returned to service. The cause of the spurious spikes is attributed to an unknown source of electrical noise. The issue with spurious spiking has been entered into the station's corrective action program.
"Both trains of Standby Gas Treatment System started as designed and Reactor Building ventilation isolated. The train actuation was complete.
"The PCIS functioned successfully providing a complete Group 3 isolation. The PCIS Group 3 isolation involves the following systems.
"Drywell and Suppression Chamber air and vent: V16-19-6A, 6B, 7, 7A, 7B, 8, 9, 10, 23 "Containment Makeup: V-16-20-20, 22A, 22B "Containment Air Sampling: VG-23, 26, V109-76A, 76B "Containment Air compressor suction: V72-38A, 38B "Containment Air Dilution: VG-9A, 9B, 22A, 22B, NG-11A, 11B, 12A, 12B, 13A, 13B "In accordance with 10CFR50.73(a)(1) a telephone notification is being made instead of submitting a written Licensee Event Report."
The licensee has notified the NRC Resident Inspector and will notify the State and local agencies.
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Home > NRC Library > Document Collections > NRC Regulations (10 CFR) > Part Index > § 50.72 Immediate notification requirements for operating nuclear power reactors.
§ 50.72 Immediate notification requirements for operating nuclear power reactors.
(a) General requirements.1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part or a combined license under part 52 of this chapter after the Commission makes the finding under § 52.103(g), shall notify the NRC Operations Center via the Emergency Notification System of:
(i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan; 2 or (ii) Those non-emergency events specified in paragraph (b) of this section that occurred within three years of the date of discovery.
(2) If the Emergency Notification System is inoperative, the licensee shall make the required notifications via commercial telephone service, other dedicated telephone system, or any other method which will ensure that a report is made as soon as practical to the NRC Operations Center.3 (3) The licensee shall notify the NRC immediately after notification of the appropriate State or local agencies and not later than one hour after the time the licensee declares one of the Emergency Classes.
(4) The licensee shall activate the Emergency Response Data System (ERDS) 4 as soon as possible but not later than one hour after declaring an Emergency Class of alert, site area emergency, or general emergency. The ERDS may also be activated by the licensee during emergency drills or exercises if the licensee's computer system has the capability to transmit the exercise data.
(5) When making a report under paragraph (a)(1) of this section, the licensee shall identify:
(i) The Emergency Class declared; or (ii) Paragraph (b)(1), One-hour reports, paragraph (b)(2), Four-hour reports, or paragraph (b)(3), Eight-hour reports, as the paragraph of this section requiring notification of the non-emergency event.
(b) Non-emergency events--(1) One-hour reports. If not reported as a declaration of an Emergency Class under paragraph (a) of this section, the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plant's Technical Specifications authorized pursuant to Sec. 50.54(x) of this part.
(2) Four-hour reports. If not reported under paragraphs (a) or (b)(1) of this section, the licensee shall notify the NRC as soon as practical and in all cases, within four hours of the occurrence of any of the following:
(i) The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.
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(ii)-(iii) [Reserved]
(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
(v)-(x) [Reserved]
(xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.
(3) Eight-hour reports. If not reported under paragraphs (a), (b)(1) or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following:
(i) [Reserved]
(ii) Any event or condition that results in:
(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
(iii) [Reserved]
(iv)(A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.
(B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are:
(1) Reactor protection system (RPS) including: Reactor scram and reactor trip. 5 (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).
(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: High-head, intermediate
-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.
(4) ECCS for boiling water reactors (BWRs) including: High-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.
(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.
(6) PWR auxiliary or emergency feedwater system.
(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.
(8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.
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(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:
(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.
(vi) Events covered in paragraph (b)(3)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function.
(vii)-(xi) [Reserved]
(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.
(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).
(c) Followup notification. With respect to the telephone notifications made under paragraphs (a) and (b) of this section, in addition to making the required initial notification, each licensee, shall during the course of the event:
(1) Immediately report (i) any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the Emergency Classes, if such a declaration has not been previously made, or (ii) any change from one Emergency Class to another, or (iii) a termination of the Emergency Class.
(2) Immediately report (i) the results of ensuing evaluations or assessments of plant conditions, (ii) the effectiveness of response or protective measures taken, and (iii) information related to plant behavior that is not understood.
(3) Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC.
[48 FR 39046, Aug. 29, 1983; 48 FR 40882, Sept. 12, 1983; 55 FR 29194, July 18, 1990, as amended at 56 FR 944, Jan. 10, 1991; 56 FR 23473, May 21, 1991; 56 FR 40184, Aug. 13, 1991; 57 FR 41381, Sept. 10, 1992; 58 FR 67661, Dec. 22, 1993; 59 FR 14087, Mar. 25, 1994; 65 FR 63786, Oct. 25, 2000; 72 FR 49502, Aug. 28, 2007]
- 1. Other requirements for immediate notification of the NRC by licensed operating nuclear power rectors are contained elsewhere in this chapter, in particular Secs. 20.1906, 20.2202, 50.36, 72.216, and 73.71.
- 2. These Emergency Classes are addressed in Appendix E of this part.
- 3. Commercial telephone number of the NRC Operations Center is (301) 816-5100.
- 4. Requirements for ERDS are addressed in Appendix E,Section VI.
- 5. Actuation of the RPS when the reactor is critical is reportable under paragraph (b)(2)(iv)(B) of this section.
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Home > NRC Library > Document Collections > NRC Regulations (10 CFR) > Part Index > § 50.73 Licensee event report system.
§ 50.73 Licensee event report system.
(a) Reportable events. (1) The holder of an operating license under this part or a combined license under part 52 of this chapter (after the Commission has made the finding under § 52.103(g) of this chapter) for a nuclear power plant (licensee) shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. In the case of an invalid actuation reported under § 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. Unless otherwise specified in this section, the licensee shall report an event if it occurred within 3 years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.
(2) The licensee shall report:
(i)(A) The completion of any nuclear plant shutdown required by the plant's Technical Specifications.
(B) Any operation or condition which was prohibited by the plant's Technical Specifications except when:
(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.
(C) Any deviation from the plant's Technical Specifications authorized pursuant to Sec. 50.54(x) of this part.
(ii) Any event or condition that resulted in:
(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.
(iv)(A) Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section, except when:
(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or Page 1 of 5 NRC: 10 CFR 50.73 Licensee event report system.
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(2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.
(B) The systems to which the requirements of paragraph (a)(2)(iv)(A) of this section apply are:
(1) Reactor protection system (RPS) including: reactor scram or reactor trip.
(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).
(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.
(4) ECCS for boiling water reactors (BWRs) including: high-pressure and low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.
(5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system.
(6) PWR auxiliary or emergency feedwater system.
(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.
(8) Emergency ac electrical power systems, including: emergency diesel generators (EDGs); hydroelectric facilities used in lieu of EDGs at the Oconee Station; and BWR dedicated Division 3 EDGs.
(9) Emergency service water systems that do not normally run and that serve as ultimate heat sinks.
(v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:
(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.
(vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function.
(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:
(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or Page 2 of 5 NRC: 10 CFR 50.73 Licensee event report system.
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(D) Mitigate the consequences of an accident.
(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.
(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e.,
unrestricted area) for all radionuclides except tritium and dissolved noble gases.
(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:
(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.
(B) Events covered in paragraph (a)(2)(ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (a)(2)(ix)(A) of this section if the event results from:
(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.
(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.
(b) Contents. The Licensee Event Report shall contain:
(1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence.
(2)(i) A clear, specific, narrative description of what occurred so that knowledgeable readers conversant with the design of commercial nuclear power plants, but not familiar with the details of a particular plant, can understand the complete event.
(ii) The narrative description must include the following specific information as appropriate for the particular event:
(A) Plant operating conditions before the event.
(B) Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event.
(C) Dates and approximate times of occurrences.
(D) The cause of each component or system failure or personnel error, if known.
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(E) The failure mode, mechanism, and effect of each failed component, if known.
(F) The Energy Industry Identification System component function identifier and system name of each component or system referred to in the LER.
(1) The Energy Industry Identification System is defined in: IEEE Std 803-1983 (May 16, 1983) Recommended Practice for Unique Identification in Power Plants and Related Facilities--Principles and Definitions.
(2) IEEE Std 803-1983 has been approved for incorporation by reference by the Director of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR part 51.
(3) A notice of any changes made to the material incorporated by reference will be published in the Federal Register.
Copies may be obtained from the Institute of Electrical and Electronics Engineers, 445 Hoes Lane, P.O. Box 1331, Piscataway, NJ 08855-1331. IEEE Std 803-1983 is available for inspection at the NRC's Technical Library, which is located in the Two White Flint North Building, 11545 Rockville Pike, Rockville, Maryland 20852-2738; or at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202-741-6030, or go to: http://www.archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html.
(G) For failures of components with multiple functions, include a list of systems or secondary functions that were also affected.
(H) For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service.
(I) The method of discovery of each component or system failure or procedural error.
(J) For each human performance related root cause, the licensee shall discuss the cause(s) and circumstances.
(K) Automatically and manually initiated safety system responses.
(L) The manufacturer and model number (or other identification) of each component that failed during the event.
(3) An assessment of the safety consequences and implications of the event. This assessment must include:
(i) The availability of systems or components that could have performed the same function as the components and systems that failed during the event, and (ii) For events that occurred when the reactor was shutdown, the availability of systems or components that are needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
(4) A description of any corrective actions planned as a result of the event, including those to reduce the probability of similar events occurring in the future.
(5) Reference to any previous similar events at the same plant that are known to the licensee.
(6) The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information concerning the event and the plant's characteristics.
(c) Supplemental information. The Commission may require the licensee to submit specific additional information beyond that required by paragraph (b) of this section if the Commission finds that supplemental material is necessary for complete understanding of an unusually complex or significant event. These requests for supplemental information Page 4 of 5 NRC: 10 CFR 50.73 Licensee event report system.
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will be made in writing and the licensee shall submit, as specified in § 50.4, the requested information as a supplement to the initial LER.
(d) Submission of reports. Licensee Event Reports must be prepared on Form NRC 366 and submitted to the U.S.
Nuclear Regulatory Commission, as specified in § 50.4.
(e) Report legibility. The reports and copies that licensees are required to submit to the Commission under the provisions of this section must be of sufficient quality to permit legible reproduction and micrographic processing.
(f) [Reserved]
(g) Reportable occurrences. The requirements contained in this section replace all existing requirements for licensees to report "Reportable Occurrences" as defined in individual plant Technical Specifications.
[48 FR 33858, July 26, 1983, as amended at 49 FR 47824, Dec. 7, 1984; 51 FR 40310, Nov. 6, 1986; 56 FR 23473, May 21, 1991; 56 FR 61352, Dec. 3, 1991; 57 FR 41381, Sept. 10, 1992; 58 FR 67661, Dec. 22, 1993; 59 FR 50689, Oct. 5, 1994; 63 FR 50480, Sept. 22, 1998; 65 FR 63787, Oct. 25, 2000; 69 FR 18803, Apr. 9, 2004; 72 FR 49502, Aug. 28, 2007]
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