ML20236G730
| ML20236G730 | |
| Person / Time | |
|---|---|
| Site: | Idaho State University |
| Issue date: | 12/31/1997 |
| From: | Bennion J IDAHO STATE UNIV., POCATELLO, ID |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9807060358 | |
| Download: ML20236G730 (12) | |
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l June 29,1998 STATE UNIVERSITY Document Controi Dest U.S. Nuclear Regulatory Commission j Washington, D.C. 20555
Subject:
Transmittal of Annual Facility Operating Report for 1997.
Dear Sir / Madam o.nese of M
Enclosed please find a copy of the Annual Operating' Report for the Idaho State fjj(( University AGN-201M Reactor, License No. R-110, Docket No. 50-284, for calendar .
83m).so60 year 1997. Submission of this report fulfills the requirements of AGN Technical !'
Specification 6.9.1. A copy of this report has also been sent to the Region IV Administrator, as required by the aforementioned technical specification.
If you have any questions concerning the report, please contact me at (208) 236-3351. I 1
Sincerely, John S. Bennion l Reactor Administrator t
cc: Mr. Marvin M. Mendonca, Project Manager Non-Power Reactors and Decommissioning Project Directorate l
Division of Reactor Program Management i Office of Nuclear Reactor Regulation 4he,,- ~
i 9807060358 97123124 236 m PDR ADOCK 050 FAX: R. , ,
(208) 236 4538 - " u v U ISU is An Equal Opportunity Employer a
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Draft Minutes of the Idaho State University Reactor Safety Committee (RSC)
Date: December 29,1997
. Start Time: 1:40 P.M.
End Time: 3:10 P.M.
Location: Lillibridge Engineering L .boratory Rm 218 Members Present: Members Excused.
Mr. Frank Just, Chairman Mr. Rick Cummings Dr. Jay Kunze, Dean ofNnt;ineering
- Dr. John Bennion, Reactor Administrator and Acting Reactor Supervisor Dr. Tom Gesell, Radiation Safety Officer Dr. Frank Harmon Mr. Mike Vaughn Mr. Terry Smith Others Pr.;;dnL Mr. Todd Gansauge
- 1. The meeting was called to order at 1:40 P.M.
- 2. Mr. Todd Gansauge was introduced as the future Reactor Supervisor pending passing the NRC SRO exam.
- 3. The minutes of the October 29,1996 Reactor Safety Committee meeting were reviewed and approved.
- 4. Mr. Mike Vaughn asked the status of missing comments about the console upgrade regarding issues raised by Mr. Jim Larson. Dr. Bennion committed to forwarding clarifying information to the committee.
- 5. Dr. John Bennion reviewed the reportable occurrence of June 29,1997. Failure of a primary fission product barrier caused by a cladding failure on Safety Rod No. 2. The details of the event, and the steps taken to recover were discussed.
- a. Mr. Mike Vaughn asked if metallurgical analysis had been preformed on the cladding failure. Dr. Bennion indicated that he and Mr. Kermit Bunde had taken some photomicrographs of the fracture surface Mr. Vaughn indicated that he would speak with the metallurgists in his section at the INEEL and ask them if they would be willing to take a look at the fracture. The possibility of decontaminating the end cap and sending it to INEEL for analysis was discussed.
Mr. Frank Just asked Dr. Bennion to follow up on this matter and make it an agenda item for the next Reactor Safety Committee meeting.
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- 6. The proposed modifications of the dashpot mounting brackets were discussed.
- a. Mr. Mike Vaughn raised a question about the bearing surface between the proposed dashpot modification and the dashpot shaft. He suggested this bearing surface be made out ofbrass. Dr. Bennion commented that the staff had discussed the possibility of fabrication of this surface from polyethylene.
- b. Mr. Terry Smith suggested that the dashpot mounting modification be validated on ;
the test stand before being installed in the reactor. He raised the issue of dashpot malfunction byjamming in the fully extended position and asked how the resulting position of the elements would effect the shutdown margin. The staff committed to look into this and include it in the 50.59 analysis. i
- c. Dr. Jay Kunze asked about a formal 50.59 analysis for this modification. Dr.
Bennion commented that :he attached handout was the first steps in such an analysis and the analysis was underway.
- d. Mr. Mike Vaughn asked if the manufacture of the dashpot had been contacted ;
about this problem with dashpot mounting. Mr. Todd Gansauge responded stating that discussions had been held with the manufacturer and that the manufacturer ;
indicated that these dashpots were not designed to be bottomed or operated without external shaft suppon.
- 7. Dr. Harmon moved to allow restart of the reactor. Dr. Gesell amended the motion to l include removal of the dashpot between runs. After discussion reactor restart was i approved unanimously with the following conditions: l
- a. The replacement dashpot used during control rod calibration following the reportable occurrence be replaced with one of the unused dashpots until mounting plate modifications can be completed.
- b. The SR-2 dashpot will be removed from the reactor between runs to minimize the time spent with a static load on the graphite piston until modifications to the mounting hardware can be completed. !
i 8. Mr. Terry Smith volunteered to do the two annual audits required of the RSC. He will contact the reactor staff near the end ofJanuary and make arrangements to preform these audits.
- 9. Dr. John Bennion reviewed the results of the latest NRC inspection. He stated that all previous items have been closed. Two non-citatable violations were listed in the inspection report that had been self-identified by the staff.
- 10. Mr. Mike Vaughn brought up the issue of a document control system and storing duplicate records in a fireproof cabinet off site. This issue was discussed and the staff indicated that a lot of effort has gone into organizing and maintaining the reactor records j and that the staffis not aware of any requirements for a document control system or off site storage.
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- 11. Dr. John Bennion updated information on the current status of the reactor console upgrade. Mr. Mike Vaughn mentioned the need for detailed procedures on console changeover.
- 12. Dr. John Bennion briefed the committee on the status of the relicensing effon for the l
reactor. Commenting that he had committed to responding to the NRC's Request for Additional Information as soon as possible.
- 13. Dr. Tom Gesell gave the Radiation Safety Officer report. He reminded the committee that the Technical Safety Office operates under a written agreement for services to the reactor facility. The TSO provided special services, verifying no measurable I-131 uptake, as requested by the reactor staff. He mentioned the fact that Mr. Kermit Bunde underwent a full body assay preformed at the INEEL at Mr. Bunde's request. No evidence of uptake was found. All external doses to facility personnel have been within ALARA limits, based on his reviews ofdosimetry records.
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- 14. Mr. Frank Just stated that Mr. Rick Cummings has asked to step down from the Reactor
~ Safety Committee due to lack of time to fulfill this position. Mr. Ken Waitham and Mr. j Bob Boston were suggested for consideration for RSC appointments. The staff committed to look at the RSC charter and verify staffing requirements for the RSC.
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- 15. Dr. John Bennion reviewed the emergency drill held in January of 1996. He stated that he needs to finish the final repon on this drill and distribute it to the committee.
- 16. The meeting was adjourned at 3:10 P.M.
Submitted by: Todd C. Ganunoe L
Approved:
Reactor Safety Committee Chair Date 3
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Idaho State University AGN-201M Reactor Facility License R 110, Docket No. 50 284
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Annual Operating Report for 1997
- 1. Narrative Summary A. Changes m Facility Design, Performance Characteristics, and Operating Procedures:
No changes in facility design or performance characteristics relating to reactor safety occurred during the reporting period.. An application to amend the facility operating license was submitted to the Nuclear Regulatory Commission to increase the possession limit for urenium-235 from 700 to 993 grams in order to allow the facility to receive spare control elements from Oregon State University (OSU). The spare control elements, complete with fuel material, were from an AGN reactor that was operated at OSU and subsequently decommissioned during the period 1978-1980. The license amendment was issued on August 18,1997.
B. Results of Major Surveillance Tests and Inspections:
(1) Channel tests performed on all safety channels and scram interlocks were found to be satisfactory and within specifications.
(2) Power and period calibrations were performed with satisfactory results.
(3) The shield water tank was inspected and no leaks or excessive corrosion were observed.
(4) The seismic displacement interlock was tested satisfactorily.
(5) (a) The control rod drive mechanisms were inspected and tested with satisfactory results.
(b) Ejection times were measured for all scrammable rods and found to be less l
than 130 milliseconds.
i (c) The reactivity worths of all safety and control rods were measured, as well as the time required to drive each rod to its fully inserted position.
Reactivity insertion rates were determined to be less than 0.039% Ak/k s4
($0.052 sd) for all rods.
(d) The shutdown margin was determined to be greater than 1.64% Ak/k ($2.22) with both the most reactive scrammable rod and the fine control rod fully inserted.
(e) All surveillance were within the appropriate Technical Specification requirements.
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c-l Idaho State University AGN-201M Rer.ctor -
1997 Annual Operating Report page 2
- 2. Operating History and Energy Output.
The reactor was operated at power levels up to approximately 4.8 watts for a total 77.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereby generating 1.98 watt-days (47.6 watt-hours) of thermal energy during this reporting period. A summary of monthly operations for 1997 is given in Table I.
Table I. Summary of Monthly Reactor Operations (1 January 1997 through 31 December 1997)
Month Hogi Enerry (W-hr)
January 4.4 0.13 j February 14.3 15.64 March 16.9 0.34 April 8.8 1.98
,. f/ay 10.6 4.10 June 14.6 25.09 Jt.ly 0.0 0.00 August 0.0 0.00 September - 0.0 0.00 October 8.2 0.33 November 0.0 0.00 i
December 0.0 0 tx) i Total 77.8 hr 47.61 W-hr l
! 3. A. Unscheduled Shutdowns and Corrective Actions Taken.
I 3/29/97 The Coarse Control Rod (CCR) drive switch became inoperable during FCR calibration. The reactor was scrammed to allow cleaning of the switch. The switch was thoroughly cleaned and returned to service.
6/25/97 During restart following a planned low level trip on Nuclear Instrument l Channel No.1, Nuclear Instrument Channel No. 3 tripped high due to improper p s J.-
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O Idaho State University AGN-20lM Reactor 1997 Annual Operating Report page 3 range switching by the operator-in-training. Attempts to restart resulted m Safety Rod No. 2 (SR-2) unlatching from the electromagnet as soon as fully l inserted in the core. Investigation revealed a failed dashpot, a component of the SR-2 drive assembly. The run was then terminated and further operations were suspended pending replacement of the dashpot. Later inspection revealed that the aluminum capsule which comprises the cladding of SR-2 had failed. This discovery resulted in the immediate notification of the Nuclear Regulatory Commission as a reponable event, i.e., breach of a primary fission-product ,
berrier (see attached Reportable Occurrence Report). Funher reactor operations '
were suspended until the aluminum capsule could be replaced and the event could be reviewed by the Reactor Safety Committee. In accordance with the recovery plan for the reportable event, the reactor was operated at low power following repair of the SR-2 and replacement of the dashpot during October m order to perform required surveillance necessary to demonstrate that operation of the reactor was within technical specifications. The event was reviewed by the Reactor Safety Committee (RSC) on December 29,1997 (see draft copy of mimttes attached to this report). At that time, the RSC granted permission to resume normal operation of the reactor with the provision that the SR-2 dashpot be removed from the drive assembly between reactor operations. This action will continue until the dashpot mounting assembly can be modified such that the static load of the control element is transferred from the dashpot piston to the mounting assembly. The operations staff is in the process of designing the necessary modifications to the dashpot mounting assembly and preparing the required review to ensure no unreviewed safety questions will result from the proposed modifications. The RSC will then review the proposed modification and staff review of unreviewed safety questions before granting approval to implement the redesigned dashpot mounting assemblies.
B. Inadvenent Scrams and Action Taken.
5/28/97: During ascension to the planned power level, a high trip occurred on Nuclear Instrument Channel No I due to improper switching of by the operator-in-training. The reactor was restarted. After restart, there were 3 spurious high level trips on Nuclear Instrument Channel No. 2 attributed to power fluctuations in the building. The reactor was restarted following verification and readjustment of the Channel No. 2 zero and amplifier balance set points.
6/10/97 While irradiating a sample at 4.5 watts, noise in Nuclear Instrument Channel No. 2 caused a high level trip. The reactor was restarted and taken a power l level of 4.2 watts. Noise in Channel No. 2 resulted in another high level trip I several minutes later. The reactor was restarted and taken to a power level of
) 4.0 watts and irradiation continued.
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4 Idaho State University AGN-20lM Reactor f
1997 Annual Operating Repon (
page 4 I 6/19/97 After maintaining a steady state power of 4.7 watts for 52 minutes, Nuclear Instrument Channel No. I tripped high. The reactor was restaned and taken to the same power level. The problem did not reoccur.
6/25/97 During restan following a planned low level trip on Nuclear Instrument Channel No.1, Nuclear Instrument Channel No. 3 tripped high due to improper range switching by the operator-in-training. Attempts to restan resulted in Safety Rod No. 2 (SR-2) unlatching from the electromagnet as soon as fully inserted in the core. Investigation revealed a failed dashpot, a component of the SR-2 drive assembly. The run was then terminated. Later inspection revealed that the aluminum capsule which comprises the cladding of SR-2 had failed, as reported in item 3.A. above.
10/16/97 During determination of the Fine Control Rod (FCR) reactivity worth, fluctuations in magnet current caused a reactor trip. The reactor was restarted.
- 4. Safety-Related Corrective Maintenance 3/11/97 Two small squares of neoprene rubber were attached to the contact surface of the dashpot for Safety Rod No.1 (SR-1) when it was observed that SR-1 was not making contact with the electromagnet preventing startup.
3/22/97 The alignment of SR-1 was adjusted after to make SR-1 release upon shorting the electromagnet during rod wonh calibrations by the rod drop method.
i 3/29/97 The Coarse Control Rod (CCR) drive switch was thoroughly cleaned using an electronic solvent and returned to service.
5/31/97 The Nuclear Safety Channel No. 3 strip chart recorder was found to operate intermittently. The chart recorder contacts were cleaned and the chart recorder was retumed to service.
7/23/97 The failed SR-2 was dismantled, inspected, and inventoried. Components were placed in sealed plastic bags and stored in an approved location.
l 9/15/97 Disassembled the Oregon State University CCR. Contents were sealed in plastic bags.
9/16/97 Reassembled the ISU SR-2, using the OSU CPR capsule and extension shaft.
9/22/97 to 10/15/97 Installed the repaired SR-2 and drive assembly in the reactor, and installed my g u .e m sf&%272% M
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s Idaho State University AGN-20lM Reactor 1997 Annual Operating Report page5
. replacement dashpot. Performed required Technical Specification surveillance. All Technical Specification requirements were satisfied.
- 5. Modifications.
A. Changes in Facility Design.
There were no changes to the facility design during 1997 to the extent that changed the description of the facility in the application for license and amendments thereto.
B. Changes to Procedures.
None.
' C. Experiments.
No new or untried experiments or tests were performed during 1997.
D.- Reactor Safety Committee.
- As' of the end of the reporting period, membership of the Reactor Safety Committee (RSC) consisted of the following individuals:
Frank H. Just - Chair Jay F. Kunze - Dean, College of Engineering John S. Bennion - Reactor Administrator and Acting Reactor Supervisor l Thomas F. Gesell - Radiation Safety Officer '
J. Frank Harmon Terry W. Smith Michael E. Vaughan
- 6. Summary of Changes Reportable under 10 CFR 50.59. d L
~ None. '
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- 7. Radioactive Effluents.
A. Liquid Waste-Total Activity Released: None.
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- Idaho State University AGN-20lM Reactor 1997 Annual Operating Report page 6 B. Gaseous Waste - Total Estimated Activity Released: 1.04 uCi.
The AGN-201 Reactor was operated for 77.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at power levels up_ to approximately 4.8 watts. At this power level argon-41 production is negligible and substantially below the effluent concentration limit given in 10 CFR 20 Appendix B, Table 2. The total activity of Ar-41 released to the environment was conservatively estimated at 1.04 pCi. This activity corresponds to the total activity of all gaseous radioactive effluent from the facility. A monthly summary of gaseous releases is given in Table II.
Table 11. Surnmary of Monthly Gaseous Radioactive Effluent Releases (1 January 1997 through 31 December 1997)
Month Ar-41 (uCi)
January 0.003 February 0.341 March 0.007 April 0.043 May 0.089 June 0.548 July - 0.000 August 0.000 September 0.000 October 0.000 November 0.000 Decembe_r 0.007 Total activity: 1.038 Ci C. Solid Waste - Total Activity: None.
- 8. Environmental radiation surveys, performed at the facility boundary while the reactor was l operating at 94% of full licensed power (4.7 watts), measured a maximum combined L neutron and gamma dose equivalent rate ofless than 1.6 mrem hr' at the outside walls of the building proximal to the reactor. j l
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- 9. Radiation Exposures.
Personnel radiation exposures are reviewed quarterly by the Radiation Safety Officer.
Annual reports of ionizing radiation doses are provided by the Radiation Safety Officer to all monitored personnel as required under the provisions of 10 CFR 19.
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9 Idaho State University AGN-20lM Reactor 1997 Annual Operating Report page 7 Personnel with duties in the reactor laboratory on either a regular or occasional basis have been issued radiation dosimeters by the Idaho State University Technical Safety Office.
These dosimeters are changed out on a quarterly basis. The duty category and monitoring period of personnel are summarized in Table III:
Table III. Personnel Monitored for Exposure to Ionizing Radiation Name Monitorine Period Duty Catecorv Kazi Ahmed 1/1/97 12/31/97 Regular Rick Baker 1/1/97- 12/31/97 Occasional John S. Bennion 1/1/97 12/31/97 Regular Robert D. Boston 1/1/97- 12/31/97 Terminated 8/15/97 Kermit A. Bunde 1/1/97- 12/31/97 Regular Todd C. Gansauge 5/1/97 - 12/31/97 Regular Dirk Howlett 1/1/97 - 12/31/97 Occasional Raed Jaber 1/1/97- 12/31/97 Occasional Michael F. Jolley 1/1/97 - 12/31/97 Occasional Jay F. Kunze I/1/97-12/31/97 Regular Sad Jarall 1/1/97 - 12/31/97 Occasional James K. Sample 1/1/97 12/31/97 Occasional Alan G. Stephens 1/1/97 - 12/31/97 Occasional William Taylor 1/1/97- 12/31/97 Occasional Miles Whiting 1/1/97 - 12/31/97 Occasional Dose Equivalent summary for Reporting Period:
Measured Doses 1/1/97 - 12/31/97 Whole-body dose equivalents: s 10 mrem for most personnel.
Maximum measured quarterly dose = 50 mrem shallow dose from a ring badge.
Minimum detectable dose equivalent per quarterly badge = 10 mrem.
None of the 122 visitors to the facility during 1997 received a measurable dose. Therefore, the average and maximum doses are all within NRC guidelines. A summary of whole-body exposures for facility personnel is presented in Table IV.
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Idaho State University AGN-20lM Reactor 1997 Annual Operating Report-page 8.
Table IV. Summary of Whole-Body Exposures (1 January 1997 through 31 December 1997)
Estimated whole-body exposure Number ofindividuals in range (rem): each range:
No Measurable Dose 7 less than 0.10 8 0.10 to 0.25 0 0.25 to 0.50 0 0.50 to 0.75 0 0.75 to 1.00 0 1.00 to 2.00 0 2.00 to 3.00 0 3.00 to 4.00 -0 4.00 to 5.00 0 Greater than 5 rem 0 Total number of individuals reported: 15 Report submitted by:
Todd C. Gansuage, Reactor Supervisor (April 1998 to present)
' John S. Bennion, Reactor Administrator and Acting Reactor Supervisor (to April 1998) l 1 l
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August 14,1997 Mr. Marvin M. Mendonca U.S. Nuclear Regulatory Commission STA'l,E eDNe UNIVERSITY M.S. Il-B-20 ,
Washingtcn, D.C. 20555 l
Subject:
Transmittal of facility status report following the 7/97 reponable occurrence at Idaho State University AGN-201 reactor f acility.
Co8ege of
Dear Mr. Mendonca:
]'*[b' 8060 Pcatello, Idaho Enclosed please find a copy of the facility status report regarding the July 1997 8320Sso60 reportable occurrence at the ISU AGN-201 nuclear reactor, License No. R-110, Docket No. 50-284, involving the abnormal degradation of a primary fission-product barrier. As requested, the report discusses the current status of the facility and actions taken and planned to recover from the event and allow recommencement of normal reactor operations. I apologize for any inconvenience that the delay in submitting this report may have caused you or your staff.
A copy of this report was transmitted to your office by facsimile this afternoon, August 14th, at approximately 2:10 p.m. MST. As we discussed during our telephone conversation earlier today, I understand that you will deliver a copy of the report to the NRC Document Control Center. Your assistance is greatly appreciated.
elease call me at (208) 236-3351 regarding any questions you may have conceming this matter.
Sincerely yours, John S. Bennion Reactor Administrator
Enclosures:
(1) Facility Status Report to the US NRC Regarding the Control Element Cladding Failure at the ISU AGN-201 Nuclear Reactor (2) Envelope containing original photographs.
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(208) 236-2902 FAX:
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- veg August 14,1997 Mr. Marvin M. Mendonca U.S. Nuclear Regulatory Commission STA'I,E eDNe UNIVERSITY M.S. Il-B-20 Washington, D.C. 20555 FAX:(301) 415-2279 ,
Subject:
Transmittal of facility status repon following the 7/97 reportable occurrence at Idaho State University AGN-201 reactor facility.
Engineering Campus Box 8060 Pmeuo. No
Dear Mr. Mendonca:
83209-8060 Attached please find a copy of the facility status report regarding the July 1997 reponable occu Tence at the ISU AGN-201 nuclear reactor, License No. R-110, Docket No. 50-284, involving the abnormal degradation of a primary fission-product barrier. As requested, the report discusses the current status of the facility and actions taken and planned to recover from the event and allow recommencement of normal reactor operations.
The original copy of the report including original photographs is being sent under separate cover via priority U.S. Mail. Please call me at (208) 236-3351 regarding any questions you may have concerning this matter.
Sincerely yours, John S. Bennion Reactor Administrator Attachments: (1) Facility Status Report to the US NRC Regarding the Control Element Cladding Failure at the ISU AGN-201 Nuclear Reactor Phone:
(2c3) 2362902 FAX: _j (208) 2 % 4538 _
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.. .n FACILITY STATUS REPORT TO TIIE U.S. NUCLEAR REGULATORY COMMISSION REGARDING THE CONTROL ELEMENT CLADDING FAILURE AT TIIE IDAIIO STATE UNIVERSITY AGN-201 NUCLEAR REACTOR Introduction This report provides a brief summary of the current status of the Idaho State University (ISU)
AGN-201 reactor, License No. R-110, Docket No. 50-284, following the discovery of a breach in the control element capsule on July 3,1997. The failed element, Safety Rod No. 2 (SR-2), contains fuel material in the form of 20Eenriched UO 2homogeneously dispersed in a radiation stabilized polyethylene matrix. The failure of the aluminum capsule, which is considered a primary fission-product barrier, exposed a small portion of the fuel material. As required by AGN Technical Specification 6.9.2.a.(3) of the operating license, this event was classified as a reponable occurrence and was promptly reported to the U.S. Nuclear Regulatory l Commission with a written follow-up report, dated July 21,1997, which was transmitted to the 1
NRC by facsimile July 18,1997.
Immediately following discovery of the cladding failure, facility personnel with the assistance of the personnel from the ISU Technical Safety Office performed surveillance for radioactive contamination, radiation fields, and airbome radioactive particulate. Surveyed areas included l I
the region beneath the reactor core near the control element access, the general vicinity of the scactor, affected components, and the reactor laborcory room. Contamination surveys of the 2
affected components showed only low levels (barely above background, net <100 cpm /100 cm except for the inside of the capsule) of removable contamination on surfaces that were in direct contact with the fuel material; specifically, on the external surface of the SR-2 capsule adjacent to the failed weld and on the interior surface of the detached end cap. Analysis of the filter media from the air paniculate sampler showed no evidence of airborne radioactive paniculates beyond what would be expected from, and consistent with, shon-lived radon progeny.
Subsequent measurements 5 weeks later showed the same radon results. In addition, all personnel present at the time of the initial dashpot failure and during the later removal of SR-2 underwent in vivo counting of the thyroid gland for uptake of I-131. Results from the thyroid counting did not show any [x>sitive indication of elevated emissions from the thyroid glands of any of the personnel surveyed.
ISU AGN-201 Reactor Facihty status Report August 14.1997 page 1 m : .. - v:xww ~ . m a x :~;;;; ; ; w ~ , -
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The Intemal Incident Assessment Committee (IIAC), which was summarily formed to determine the cause and assess the consequences of the event, concluded that the failure of the SR-2 capsule was caused directly by the failure of the dashpot mounted beneath the control element. This dashpot serves to decelerate the element following ejection from the core as a result of either a scram or the nonnal shutdown sequence. Without the damping action of the dashpot to absorb the energy of the ejected element, the abrupt and unattenuated impact of SR-2 against the steel frame of the drive assembly was sufficient to fracture the weldjoining the end cap to the capsule tube. It is also possible that normal aging of the aluminum components may have contributed to the capsule failu.e. The IIAC alsc concluded that the radiological consequences of event were negligible and had no adverse impact on the health and safety of facility personnel, the public, or the environment.
Enclosed with this report are eight photographs of the SR-2 components that were taken to document the extent of damage incurred by the capsule as a result of this event. j i
Figure i shows the SR-2 drive assembly and the complete SR-2 control element assembly subsequent to removal from the reactor. Note the black tip of the control element assembly at the left of the photograph. As may be seen in more detail in Figure 2, this black tip is the distal UO 2-polyethylene fuel disk / cylinder protruding approximately 2 cm out of the element capsule as a result of the complete fracture of the weldjoining the end cap to the capsule tube and the consequent loss of the end cap. The undamaged CCR (Coarse Control Rod) isjuxtaposed with I
SR-2 for comparison.
i Figure 3 shows a close-up view of the end portion of SR-2 with the protruding fuel l disk / cylinder. Also shown in this figure are the detached SR-2 end cap and the CCR. Note the jagged edge of the capsule tube where the weldjoining the end cap failed. In Figure 4, the SR-2 has been disassembled and its component pans have been arranged for display. Starting from the bottom of the photograph is the SR-2 capsule. Above the capsule, from left to right, are the compression spring, washer,9-cm-high graphite reflector disk / cylinder, and four 4-cm-high fuel disks / cylinders. Above the intemal SR-2 components is the extension shaft which connects the fueled portion of the element to the drive carriage. The entire control element assembly is created by loading into the capsule, in sequence, the four fuel disks, the graphite 1
disk, the compression spring, and the washer, which is then screwed onto the threaded extension shaft. A gas-tight seal for the control element is obtained by tightening the capsule to 1511 AGN.201 Reactor Facihty Status Report August 14.1997 page 2
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compress the seated o-ring that is visible in the photograph at the right end of the extension
! shaft,just to the left of the screw threads.
Figure 5 gives a close-up view of the end of the SR-2 capsule and the detached end cap showing the fractured weld. Figure 6 shows the failed dashpot next to the threaded flange with attached mounting studs which are used to secure the dashpot to the control element drive assembly. The intemal components of the dashpot are visible through the transparent cylinder tube. Figure 7 gives a close-up view of the broken dashpot. The broken pieces of the graphite piston are visible within the dashpot cylinder.
Figure 8 shows one of the new replacement dashpots obtained from Airpot,Inc. The cylinder
} of the new dashpot is made of a black, opaque plastic instead of a transparent material and conceals the internal graphite piston from view.
I As discussed in the July 18,1997, report to the NRC, three options are available for replacing the SR-2 capsule to make the reactor operable so that the facility may resume normal operations.
The first option involves attempting to repair the failed capstile. This option would require decontamination of interior surfaces, expen welding of the delicate components, and finally, pressure testing to ensure that the capsule is air-tight. The second option entails fabrication of a new capsule. The third option is to locate and transfer to ISU a replacement capsule from a decommissioned AGN-201 reactor. After a careful consideration of these options, we concluded that the only viable option is to acquire salvaged units from a decommissioned facility.
! The proposed plan for recovery, therefore, is as follows:
. Submit a comprehensive repon to the RSC for review.
. Transfer replacement control elements to ISU from the Oregon State University nuclear I facility,
. Install replacement SR-2 control element in the ISU reactor. j Replace the SR-2 dashpot on the drive assembly. j
- Perfomi all required surveillance procedums, e.g., measurement of scram time and rod f
worth. I
- Submit a final recovery repon to the RSC for review and seek approval to resume normal reactor operations I I
. Submit a counesy copy of the final recovery report to the NRC.
ISU AGN.201 Reactor Facility Status Report August 14. 1997 page 3
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Current Status of the Facility The reactor facility remains shutdown, as it has since discovery of the failed dashpot and the subsequent discovery of the failed SR-2 capsule.
The Reactor Safety Committee (RSC) has been notified of the event and is expecting a repon for review at the next scheduled meeting. The RSC will review the event, including actions
- taken to correct the problem and to prevent its recurrence, and may advise on any further actions deemed appropriate. Once the required repairs have been made to the affected control element (which will effectively entail the replacement of the control element), the facility will submit a final recovery report to the RSC and request approval to resume reactor operations.
Following the initial surveillance of the SR-2 components, the element was disassembled and the various components were packaged to prevent the dispersal of radioactive material. The fuel disks / cylinders were sealed in a polyethylene package and stored in the appropriate location.
The aluminum capsule and end cap have been surveyed and sealed in polyethylene pending final disposition. All other intemal compon'ents contaminated by fuel and/or fission product radionuclides have been surveyed and sealed in polyethylene pending final disposition. These materials are currently being stored within the facility boundaries.
All control rods have been removed from the reactor, surveyed, and inspected for signs of possible abnormal degradation around the end cap of the fuel capsule. All of the capsules were found to be in good condition with no evidence of weld deterioration. Contamination surveys of the capsules showed no evidence of leakage of fuel material or fission-product radionuclides.
An application for an amendment to the operating license has been submitted to the NRC which would allow the facility to possess additional fissile material and enable the transfer of salvaged control elements to ISU from the Oregon State University AGN-201 reactor. This reactor (License No. R-51, Docket No. 50-106) was decommissioned from 1978-1980. The NRC terminated the operating license November 10,1981. OSU has retained all of the AGN fuel, including the complete core and the three fueled control elements, having transferred the fuel material to OSU's TRIGA NRC license. Discussions for transferring these control elements to ISU have been initiated with Dr. Brian Dodd, Director of the OSU Radiation Center, and with representatives of the U.S. Department of Energy, which owns the fuel. Dr. Dodd fully supports the transfer of the three intact control elements and the DOE has indicated that they will
.!SU AGN 201 Reactor Facility Status Repon August 14.1997 page 4
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provide financial assistance and transfer carks as necessary for transporting the control elements '
to ISU. OSU personnel have initiated the transfer process and are awaiting notification of approval of the ISU license amendment application to proceed with the transfer.
Three new replacement dashpots have been obtained from Airpot, Inc., manufacturers of the failed SR-2 dashpot. A new dashpot will be installed on the drive assembly for each of the scrammable elements as discussed in the next section of this report. A report is being prepared as a coprtesy for submittal to Airpot, Inc., describing the dashpot failure complete with some of the photographs included in this report that may help them determine the cause of disintegration of the graphite piston for quality control purposes.
Future Actions The following actions will t$e proposed to the RSC to prevent recurrence of this event.
. All existing dashpots will be replaced with new units. The existing dash pots that are in good working condition will be retained as spare components in case of failure or deterioration of the new units. Once installed, should one of the new units fail or otherwise deteriorate to the extent that failure is probable,it shall be replaced temporarily by one of the existing units and another equivalent unit shall be purchased from the manufacturer for immediate installation.
Future annual inspections of the control elements will be aggressively performed. In particular, inspections will focus on the end region of the capsule for any evidence of weld cracking or other signs of deterioration and on the dashpot for evidence of excessive wear of the seal or excessive play in the piston which might indicate impending failure. Any evidence of degradation of either the capsule or the dashpot will be sufficient reason for immediate replacement.
We plan to modify the safety rod drive logic circuits to allow the safety rods to be manually withdrawn at the conclusion of reactor operation instead of scramming the reactor. This modification will reduce the number of scram cycles on the scrammable control elements.
Currently both safety rods must be fully inserted or " cocked" before either of the two control rods can be driven for reactor startup. .a the safety rods are m:ked, the only method for l
! lowering the safety rods and hence to terminate reactor operation is to scram the control ISU AGN 201 Reactor Facility Status Reimn August 14. 1997 page 5 j i
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elements. The control element drive logic will be modified to permit manual withdrawal of all )
control elements while retaining the original design feature requiring that the safety rods must be cocked before insertion of the remaining control rods can occur. This modification would not, of course, interfere with any of the reactor safety systems, nor the scramming function of the control elements. 'Ihe primary benefit from the proposed modification would be a significant {
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reduction in the number of scram cycles and subsequent stresses exetted on the scrammable control elements and dashpots and would virtually eliminate the possibility of another capsule failureJ Additional actions to be determined by the RSC may also be implemented.
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ISU AGN.201 Reactor Facility Status Repon August 14,1997 page 6
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t July 21,1997 Mr. Marvin M. Mendonca U.S. Nuclear Regulatory Commission PDNP
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STATE UNIVERSITY M.S. ii-B-20 Washington, D.C. 20555 FAX: (3081) 415-2279
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Subject:
Transmittal of written report regarding reportable occurrence at Idaho State University AGN-201 reactor.
Dear Mr. Mendonca:
Conege of We Attached please find a copy of the written follow-up repon regarding the reponable D"u'o, ccurrence at the ISU AGN-201 nuclear reactor, License No. R-110, Docket No. 50-83209-8 e 284, which involved the abnomial degradation of a fission-product barrier. The report describes the event, assesses the probable cause and consequences, and ;
discusses corrective actions and measures taken to prevent recurrence. It is being i submitted in compliance of Technical Specification 6.9.2(a). A copy of this report was sent to your office Friday afternoon, July 18th, at 5:30 MDT.
As discussed in the report, this incident was assessed to have no adverse impact on the health and safety of the public or the environment. None of the operations staff received elevated dose equivalent as a result of the event. I Please feel free to contact me at (208) 236-3351 regarding any questions you may have concerning this matter.
Sincerely yours, D D _- -
John S. Bennion Reactor Administrator Attachments: (1) Report to the US NRC Regarding the Control Element Failure at the ISU AGN-201 Nuclear Reactor
, (2) Memorandum Dated July 7,1997 from J. Bennion to T. Baccus l (3) Memorandum Dated July 9,1997 from T. Gesell to J. Bennion l (4) Memorandum Dated July 15,1997 from T. Baccus to J. Bennion (5) Memorandum Dated July 18,1997 from T. Gansauge to J. Bennion l
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i REPORT TO THE U.S. NUCLEAR REGULATORY COMMISSION REGARDING THE CONTkOL ELEMENT CLADDING FAILURE AT THE !
IDAHO STATE UNIVERSITY AGN-201 NUCLEAR REACTOR !
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Introduction ;
1 i l This document provides a written report of the sequence of events leading to the discovery of !
the failure of a primary fission-product barrier (fuel element cladding) of the Idaho State University (ISU) AGN-201 nuclear reactor, US NRC License No. R-110, Docket No. 284.
Such an event, i.e., the abnormal degradation of a fission-product barrier, is defined by the Technical Specification 6.9.2(a)(3) of the facility operating license as a reportable occurrence requiring prompt notification of the NRC with a follow-up written report. As required by the Technical Specifications, the NRC was promptly notified of the incident by telephone on the day of discovery. Additional calls were placed to NRC during the following week to apprise j the Project Manager of the current status of the facility and progress made towards recovery. j This report describes the event, assesses the probable cause and consequences, and discusses corrective actions and measures taken to prevent recurrence.
Description of Relevant Reactor Components The AGN-201 is a self-contained, graphite-moderated training reactor with a maximum thermal I power output of 5 watts. It consists of two basic units, the reactor unit and the control !
console. The reactor unit is composed of a central sealed cylindrical core can containing the I nuclear fuel material enclosed in a 20-cm-thick graphite reflector, which is enclosed in a 10-cm- I thick lead shield, which is enclosed by a 55-cm-thick water shield for shielding against fast neutrons. Figure I shows a vertical view of the AGN-201 reactor unit.
The AGN-201 reactor has four active control elements containing the same nuclear fuel material as the reactor core proper. Fuel consists of 100- m diameter UO2 particles, enriched to 20% in i U-235, dispersed homogeneously throughout a matrix of high-density polyethylene. Fuel I disks were made by pressing weighed quantities of UO2P / polyethylene powder in a mold under high pressure. The control elements, each containing 4 fuel disks (cylinders) with a total active length of about 16 cm, are inse ted vertically upward into the reactor core from the bottom of I the reactor unit to increase reactivity.
l l Table I summarizes the physical properties of the AGN-201 control elements. Three of the j four control elements, Safety Rod No.1 (SR-1), Safety Rod No. 2 (SR-2), and the Coarse Control Rod (CCR), are identical, having the same physical dimensions and the same approximate reactivity worth. The fourth control element, the Fine Control Rod (FCR), is smaller (about one-half the diameter) and has about one-fourth of the reactivity of each of the three large control elements. All large control elements are electromagnetically coupled to a drive carriage which moves vertically along a lead screw connected by a chain linkage to a reversible DC motor. The FCR is coupled directly to the drive carriage and has no scramming capability.
A control element assembly is comprised of the capsule, which provides the primary fission-product barrier, four fuel disks, one graphite reflector disk at the bottom, a ferrous compression spring, and the extension tube or shaft. The capsule appears to be fabricated from 1 0.065-inch-thick aluminum (6061T6) tubing by welding a flat end cap to the capsule tubing.
l The welded joint was then mechanically ground to make a smooth and slightly rounded l
cylindrical surface. The capsule is loaded with the four fuel disks followed by the graphite disk and compression spring. The open end of the capsule is threaded and screws onto the l
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.1 extension shaft. An o-ring allows the capsule to be hermetically scaled when the capsule is tightly screwed onto the extension shaft. Within the capsule, fuel is held against the distal end cap under spring loading. The control rod assembly is connected to the armature plate by means of a threadedjoint thus forming the complete control rod drive assembly, as shown in Figure 2. This latter assembly is suspended from the reactor tank by threaded studs below the sealed core can and is covered by the control element access cover which serves as a secondary barrier against the release of fission products.
Table 1. Summary of physical properties of AGN-201 control elements.
Control Element Fuel Disk Dimensions Nominal Fissile Reactivity 2 (4 Disks per Element) Contenti (gm) (Ek/k, [$])
Safety Rod No. I 4.7-cm diameter 14.4 1.15% [$1.56]
4.0-cm height Safety Rod No. 2 4.7-cm diameter 14.4 1.14% [$1.54]
4.0-cm height j Coarse Control Rod 4.7-cm diameter 14.4 1.18% [$l.59] i 4.0-cm height l Fine Control Rod 2.3-cm diameter 3.6 0.31% [$0.42]
4.0-cm height i Total fissile mass per control element (4 fuel disk-cylinders in each).
2 Most recent reactivity measurements, completed 3/11/97.
The AGN reactor is brought to operating power by inserting, in sequence, the two safety rods, which must be latched, or " cocked," into their fully inserted positions before the coarse and fine control rods may be driven. The coarse and fine control rods are then inserted to make the reactor slightly supercritical to allow the power to increase to the desired level. Once the desired operating power is reached, one or both of the moveable control rods are withdrawn to stabilize the power level. The reactor may then be operated at a steady power level as necessary until the operation is to be terminated. Normal shut down of the reactor is accomplished by scramming the safety and coarse control rods. This usually occurs by pressing the manual scram button which deenergizes the electromagnets and causes the three scrammable control elements to be ejected rapidly from the core to their safe positions.
Ejection occurs within 120 ms under the combined action of gravity and spring loading giving an initial acceleration of approximately 5 g. Each scrammable element is equipped with a shock-absorbing dashpot to gradually decelerate the element during the last 10 cm of travel.
SR-1 is equipped with the original hydraulic (oil-damped) dashpot, whereas the SR-2 and CCR elements are equipped with newer pneumatic (air-damped) dashpots. Once the control element reaches the safe or fully-withdrawn position it activates a proximity switch which causes the carriage to drive down so that the electromagnet engages the control element armature plate thereby allowing the reactor to be restarted.
Description of Incident and Immediate Actions Taken On June 25,1997, two members of the reactor operating staff, a Senior Reactor Operator (SRO) and an SRO trainee, were operating the AGN-201 nuclear reactor during a routine, after-hours training mn. The purpose of the operation was to provide supplemental operating experience for the SRO trainee, who was preparing for an imminent NRC SRO examination, l
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1 and an opponunity for the SRO to meet quanerly requalification operating requirements by supervismg the activities of the trainee. In addition, the CCR had been removed from the reactor two days before as pan of a training activity for the SRO trainee, and the operators had been asked to verify that the CCR was reinstalled properly and was operating correctly in order to complete the control element maintenance procedure.
During the first two hours of the run, the operators verified that the CCR was installed correctly and was indeed operational. They had successfully taken the reactor tc, a power level of 4 W,80% of the maximum licensed power. Reactor power was stabilized at 4 W at 20:41 MDT. They maintained the power at 4 W for 2 minutes and then reduced the power to 0.1 W to provide the trainee with additional experience in reactivity manipulation. At 20:53 the operators reduced power funher to observe the power level at which the low-level trip would actuate on Nuclear Instrument Channel No.1. The low-level scram occurred at 20:58.
The reactor operator attempted to restan the reactor at 20:59 and intended to take the reactor power to I W on a positive period of approximately 25 seconds. During power ascension, however, the operator made a switching error on Nuclear Instrument Channel No. 3, switching to a more sensitive power range rather than to a higher scale, and induced a high-level scram.
The time of this scram was logged at 21:05. The operator then attempted a second restart. As SR-2 approached its fully inserted position (approximately 24.5 cm), it dropped unexpectedly, )
i.e., disengaging from the electromagnet. After scramming the reactor to drive the safety rod j carriages down to engage the control elements, a second attempt was made to restart the reactor, again resulting in the SR-2 disengaging from the electromagnet as it reached its fully inserted position. Both operators noticed an abnormal sound, described as somewhat " louder than usual and more metallic in nature" than is normally heard when a control element drops.
l At this time SR-1 was scrammed and, after making the necessary radiological survey, the SRO entered the pedestal area to investigate the cause for the disengagement of SR-2. Exposure levels undemeath the reactor were normal and less than 0.1 mrem hrd. The SRO removed the control element access cover and unscrewed the dashpot. Examination of the dashpot intemal components through the transparent cylinder revealed that the graphite piston had disintegrated thereby rendering the dashpot useless. The SRO then called the Reactor Administrator and Acting Reactor Supervisor, who was in his office, and notified him of the failure of the dashpot.
The dashpot was surveyed for induced radioactivity and contamination and inspected.
A more detailed description of the events that transpired the evening of the June 25th is given in the attached memorandum prepared by K. Bunde and T. Gansauge, the reactor operators that l night.
The next morning,i cility staff made a concened effon to locate an equivalent dashpot to replace the one that had failed. Airpot Inc., the company that had manufactured the broken dashpot, was contacted. According to company records, the dashpot was a special order that had been placed about fifteen years earlier. However, the Vice President for Research indicated that they could manufacture replacement dashpots with a 2- or 3-day tumaround.
With this infomiation, an order was placed for three new dashpots with instructions to expedite shipment. Also, the NRC was contacted to cancel the SRO examination which had been scheduled for Wednesday, July 2nd. Instead, the examination was rescheduled one week later for Tuesday, July 8th. Airpot representatives said that they would ship one of the replacement dashpots to ISU no later than Tuesday, July 1st, which should have allowed enough time to install the dashpot and ensure that the reactor was operating properly before the NRC examiner was rescheduled to arrive at ISU to administer the SRO examination.
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t However, the dashpot was not shipped by the date as promised and the facility administration became concerned that the dashpot might not be installed in time for the NRC examination. On Thursday, July 3rd, in preparation for installing the new dashpot, the SR-2 assembly was removed from the drive assembly for inspection to ensure that the element was not damaged when the dashpot failed. When the element was transferred from the pedestal area to the Reactor Supervisor it became apparent that the end cap of the capsule had been punched through and the distal fuel disk was protruding about 2 cm out of the end of the capsule.
The discovery of the failure of a primary fission-product barrier prompted the following actions. First, the control element was placed on a plastic sheet to prevent any spread of radioactive material. Next, the element was thoroughly surveyed for direct radiation exposure levels and for removable contamination. The Dean of the College of Engineering, a Certified Health Physicist, was notified of the incident and came to the reactor laboratory to inspect the breached control element. The ISU Technical Safety Office (TSO) was also notified. A TSO staff member came to the facility and provided assistance in completing the radiological surveys. An air paniculate sampler was set up next to SR-2 near the end of the capsule and sampled airbome material for 78 min. All contamination wipes and the air particulate sample were counted in the facility and then given to the TSO for funher analysis using a liquid scintillation counter.
An Intemal Incident Assessment Committee consisting of the Dean (Dr. Jay Kunze), the Director of the Nuclear Engineering Graduate Program (Dr. Alan Stephens), and the Reactor Administrator (Dr. John Bennion), was formed to review the incident, determine the cause, and review an initial plan for recovery. The committee met that aftemoon examined the failed components and interviewed the personnel present at the time of apparent failure, i.e., the evening of June 25th.
The following Monday, July 7th, the incident was reported to Dr. Tom Gesell, the ISU Radiation Safety Officer, who had been absent from campus when the capsule breach was discovered. Dr. Gesell ordered in vivo thyroid counting of all personnel present during the incident. In addition, the wipe samples were analyzed with a high-purity germanium spectrometer to identify gamma-emitting contaminants present in the samples. The results of various radiological surveys were consistently negative and are included in the attached memoranda.
Assessment of Probable Cause and Consequences As a result of personal interviews with the reactor operating staff and inspection of the failed control element capsule, the Internal Incident Assessment Committee concluded that the capsule failure was caused directly by the failure of the dashpot. The Npact of.SR-2 at the end of travel, without benefit of the damping action of the dashpot, foilowing ejection from its fully inserted position, was sufficient fracture the weld. The final break of the weld may not have occurred until the start of the next scram. l A conservative estimate of the inventory of I-131 in SR-2 at the time of the cladding failure gives 28 pCi. Assuming that 1% of the total radiciodine content was released at the time of the l
breach of the primary fission-product barrier, a very conservative assumption since the l polyethylene matrix retains nearly all of the fission products, gives 280 nCi as the amount of 1-131 that was released to the environment. This quantity, divided by the building exhaust rate and averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period following the incident, is well below federal effluent I
concentration limits published in 10 CFR 20, Appendix B, Table 2: 2E-10 pCi/ml. I Furthermore, results of the thyroid counting by the TSO showed that none of the facility personnel approached the verification level of 9.4 nCi for uptake by the thyroid gland.
l 6
The overall assessment of the radiological consequences is that this incident has no adverse '
impact to the health and safety of employees or the public. Such a conclusion is justified because of the low power output of the reactor, the limited operating history at the time of failure, and the small fraction of fuel material that was contamed in the control capsule, about 2% of the total fissile mass of the core.
1 Plan for Recovery The proposed plan for recovery is as follows:
- Replace the SR-2 capsule with its fuel content of 4 disks.
- Submit a comprehensive repon to the RSC for review.
- Replace the SR-2 dashpot on the drive assembly.
- Transfer replacement control elements to ISU.
- Install replacement SR-2 control elements in the ISU reactor.
- Perform requisite surveillance, e.g., measurement of scram time and rod worth.
- Submit a recovery report to the RSC for review and seek approval to resume nonnal reactor l operations
- Submit a courtesy recovery report to the NRC.
Three options are available for replacing the capsule. First, the existing capsule could be repaired. This option would require decontamination of interior surface contamination and j expen welding of the delicate components with subsequent pressure testing to ensure the capsule is air-tight. Second, a new capsule could be fabricated, a process that would be expected to be difficult at best. Third, a replacement capsule from a decommissioned AGN- !'
201 reactor could be located and transferred to ISU for installation in the ISU reactor. This last option is preferable and will be pursued.
Conclusion The incident derribed in this report resulted in negligible exposure of the facility operating staff or others pd ent in the building. A negligible amount of radioactive material may have been released to the environment as a result of the breach of the SR-2 capsule. The amount of material released ivas far below effluent limits and posed no risk to the health and safety of the public or to the environment.
The reactor facility is shutdown. Operations are expected to resume when replacement control elements can be transferred from Oregon State University to ISU. Before transfer can take place, however, the AGN operating license must be amended to permit the facility to possess additional fuel. An amendment will be immediately requested to allow ISU to increase the possession limit from 700 gm of U-235 to 730 gm which will enable the OSU control elements to be transferred to the AGN license. When transfer is complete, the SR-2 will be replaced and all necessary surveillance will be performed to ensure that the reactor is fully operational and meets all pertinent technical specifications. Normal operations will resume following complete review of the repair by the ISU Reactor Safety Committee and approval to restart the reactor.
The following actions and measures will be taken in order to prevent recurrence of this type of failure of a primary fission-product barrier. First, the control element capsules will be examined more closely for signs of wear or degradation during the annual control element maintenance program. Second, all dashpots will be inspected carefully for evidence of degradation of the seal around the plunger rod which might indicate excessive wear and might contribute to the catastrophic failure of the damping piston. In addition, as a possible long-term remedy, the facility will investigate the practicality of modifying the control element drive 7
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4 logic to allow both safety rods to be driven down manually rather than having to scram the rods to shut down the reac'or. Such a modification would help to reduce impact frequency on both the control elements and the dashpots.
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MEMORANDUM l
DATE: July 7,1997 l l
TO: Tom Baccus, Technical Safety Office FROM: John S. Bennion, Assistant Professor and Reactor Administrator STATE UNIVERSITY
SUBJECT:
Locations of wipe samples from the AGN safety rod cladding failure submitted to the TSO for analysis.
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The following are the locations and results of the wipe samples taken July 3,1997, upon discovery of the Safety Rod No. 2 (SR-2) cladding failure. The wipes were coun' J using Cdece of a pancake G-M detector connected to a Ludlum Model 2A Survey Meter, Serial No. 8266, Engineenng calibration due December 1997. Background for the detector was 60 20 cpm.
caron Box 8060 Information for the air particulate sam as follows: sampler on @ l1:44:30; sampler
[ Posello. Idah off @ 13:02:30; flow rate meter read 2 ccfm. The sampler was set on the work table 83 " located against the south face of the react concrete-block shield, about 30 cm away from the end of the failed SR-2. 2fy Vial 1: 4-inch diameter air sample filter.
Gross count rate at center of filter: 460 60 cpm. (Iligh count rate was suspected to be caused by short-lived radon progeny.)
Vial 2: SR-2 Dashpot.
Gross count rate: 80 40 cpm.
Vial 3: Cap retrieval rod.
Gross count rate: 80 40 cpm.
Vial 4: SR-2 Dashpot.
Gross count rate: 60120 cpm. 4 i
Vial 5: Top ponion of SR-2 drive mechanism.
Gross count rate: 60120 cpm. )
i Vial 6: Top portion of SR-2 capsule near failure (within -10 cm failure).
Gross count rate: 120160 cpm.
Vial 7: CCR fuel capsule.
Gross count rate: 60120 cpm.
Vial 8: SR-2 capsule - detached end cap.
Gross count rate: 100 i 60 cpm.
Vial 9: SR-2 interior thimble.
Gross count rate: 160 i 60 cpm.
Vial 10: SR-2 entire rod -10 cm below cladding failure.
Phone: Gross ccunt rate: 80 i 60 cpm.
. (208) 236-2902 FAX:
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IDAHO memonmous STATE UNIVERSITY Date: July 9,1997 To: TSO Files From: Tom Gesell
Subject:
TSO response to broken fueled control rod incident at the ISU College of Engineering AGN 201 reactor.
Following discovery of the broken control rod on 7/3/97, reactor personnel made appropriate direct radiation and removable contamination surveys and notified the TSO. No unusual direct radiation fields were noted by reactor personnel. The removable contamination measurements (wipes) are listed on J,*haL'*{agg
, the attached memorandum from John Bennion dated 7/7/97. The wipes were 785 South Eighth Avenue recounted by TSO in a liquid scintillation counter; the results are also attached.
{'"a[joI70s370, The wipes were then counted on an intrinsic germanium detector; a small amount of "'Cs was identified but not quantified because the laboratory did not ra pos> 23523n have a calibration standard in geometry equivalent to the wipes, which were in liquid scintillation vials.
Thomas F. Gesell Director The instrument normally used by TSO to measure '2'I in thyroid (Ludlum 2200 scaler equipped with a 44-3 probe) was readjusted to improve response to "'I 2'
[" $)3 23%', and used to measure the thyroids of reactor personnel who were in the vicinity gesell@ physics.isu edu of the reactor following the incident. The settings used were:
HV: 195 on potentiometer THR: 50 on potentiometer WIN: 900 on potentiometer 0.1/1 toggle: 1 IN/OUT toggle: IN Calibration in approximate thyroid geometry was made with a 3Ba button source that had 50 nCi of activity remaining as of July 8,1997. Efficiency was determined with the following equation.
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' Thyroid counts'were made on three individuals from the reactor program, Kermit Bunde, John Bennion and Todd Gansauge. None approached the verification level of 9.4 nCi for 1. The results were recorded on bioassay forms and placed in the individuals' files.
enc: as stated cc: John Bennion, Tom Baccus l
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- To: Dr. Bennion ,
Reactor Supervisor
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l From: Tom Baccus MHO neaitn ehysicist, TSo STATE UNIVERSITY
Subject:
Analysis results for wipe samples from AGN safety rod cladding failm Technical safety omce Dr. Bennion:
Idaho stare Unwersny PO flox 8060 Pocueno. id The ten wipe samples from the AGN safety rod cladding failure, provided by you, 32 n so60 were counted and all were found to be less than the regulatory limit for removable Phone Contamination. Although all wipes were below the regulatory limits, some (20:3236-23 " showed small amounts of '"Cs contamination. It is therefore recommended that Fax the failed safety rod be controlled as radioactively contaminated material and all o >236a6e appropriate safety precautions be observed.
Sample counting was performed with a Beckman LS7500 Liquid Scintillation counter, serial number 101295.
Radiaton Safety
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MEMO To: Dr. John Bennion Reactor Administrator and Acting Reactor Supervisor From: Todd Gansauge and Kermit Bunde
Subject:
Failure of SR2 Date: July 18,1997 Statement of events surrounding SR2 control rod failure which appears to have p . occurred June 25,1997.
f On Monday June 23,1997 Dr. John Bennion and Mr. Todd Gansauge pulled the coarse control rod and rod drive from the reactor. The purpose of this inspection was L
to familiarize Mr. Gansauge with the rod drive mechanism in preparation for an NRC L licensing exam schedule for July 2,1997. Procedure MP-1 was started, the rod and drive were exanuned, swiped for contamination, and replaced in the reactor that aflemoon The MP-1 procedure requires that reassembly of control rod drives and control rods be verified by a second licensed reactor operator. Arrangements had already been made for Mr. Kermit Bunde to run the reactor on the evening of Wednesday June 25th for the purpose of meetmg quarterly requalification requirements. It was decided that Mr. Bunde and Mr. Gansauge would complete the MP-1 procedure and Mr. Bunde would verify the control rod reinstallation before running Wednesday evening.
L The completion of MP-1 for the course control rod went without incident. The reactor was brought to initial criticality for the day at a power level of 0.01 watts as per standard startup procedure. This criticality was achieved at 20:29 hours the evening ofJune 25th.
The power level was then raised to 4.0 watis and stabilized by 20.41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br />. At 20:43 hours power was reduced from 4.0 watts to 0.1 watt. This level was reached at 20:51 hours. At 20:53 hours it was decided to reduce the power further. The log
- entry indicates intention to reduce power to 10 microwatts.
The operators knew that they would not achieve this low power level, because the J detector for channel I was in the raised position. .The decision was made to follow the power down and see at what point channel I would scram low. This scram occurred at 20:58. Channel 3 was readmg approximately 1.0 E-l 1 amps at that time, correspondmg to a power level around 220 microwatts.
The reactor was restarted at 20:59. The operator had planned to bring the reactor to L
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w a power level of I watt. The reactor was increasing power on a period of approximately 25 seconds Several decades before reaching I watt an operator error caused a high level scram of channel 3. The channel 3 meter was crossing the 70%
mark when Mr. Gansauge reached up to switch ranges on the channel 3 power level range selector switch. Mr. Gansauge nustakenly rotated the switch in the wrong direction and switched channel 3 to a more sensitive setting. This resulted in a high scram of channel 3, The time of this event was 21:05.
Restart was attempted. As soon as SR2 was fully driven into the core, the control rod dropped away from the electromagnet. The manual scram button was then pressed allowing the SR2 rod drive to descend and reacquire the control rod Restart was attempted, with the same results. SR2 dropped away from the magnet as
. soon as fully inserted. This was accompanied by an abnormal sound. The sound
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was louder than usual and more metallic in nature. The reactor was scrammed via the manual scram button to reposition the rod drive mechanisms before Mr. Bunde opened the reactor skirt door and removed the access cover to investigate the unusual noise. Mr. Bunde brought a portable survey meter with him and noted no unusual radiation levels inside the reactor skirt (< 0.1 mr/hr). Removal of the dashpot by Mr..
Bunde showed that the dashpot for SR2 had failed. The graphite piston within the air driven dashpot had crumbled into many pieces.
The Reactor. Administrator and Acting Reactor Super isor (Dr. .lohn Bennion) was in the building and contacted. The dashpot was examined and surveyed for contamination.
The company who had manufactured this dashpot was contacted the following day By Friday of that week an order was placed for a replacement dashpot. 1 On July 3rd 1997 Dr. Bennion and Mr. Gansauge pulled the SR2 control rod. At that time they(ound that the weld along the top of the control rod had broken j Aposing fuel.
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Mr. Todd Gansauge, Senior R etBr'Operasar in trairtirrg- l MI. Kennit Buode, Senior Reactor Operator SOP-70094
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