ML20237C992
| ML20237C992 | |
| Person / Time | |
|---|---|
| Site: | 05000113 |
| Issue date: | 06/30/1998 |
| From: | Williams J ARIZONA, UNIV. OF, TUCSON, AZ |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9808240291 | |
| Download: ML20237C992 (18) | |
Text
.. -
Nuclear Reacto Iaboratory WiE UNIVER5frYOF John G. Williams, Director Engineering Building (20) c-mail: Jgw@bigdog. engr. arizona.edu P. o. Box 210020
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-m August 20,1998 U. S. Nuclear Regulatory Commission Document Control Desk l'
Washington, DC 20555 l
RE: Annual Report for License R-52, Docket 50-113 This is the Annual Report covering the period July 1,1997, through June 30,1998, for j
the activities of the TRIGA Mark l Reactor at the University of Arizona, Tucson, Arizona.
The report is submitted in compliance with Section 6.7e of the Facility Technical l
Specifications and Paragraph 50.59(b) of Title 10, Code of Federal Regulations.
1.
During the reporting period, the reactor was operated for research and education it was used for reactor operator training of replacement operators at this facility. The reactor was used for graduate thesis research and undergraduate Nuclear Engineering course experiments, including approaches to critical, control rod calibrations, measurements of the' dynamic response of the reactor to step and periodic changes in reactivity, and flux mapping. The reactor was also used for neutron activation analysis for l
teaching and research.
l Reactor upgrading or modifications include replacing power supply PS 502 with a supply whose specifications were approved by the Reactor Committee i
and modifying the mode selector switch (see summary of 50.59 review below).
Power channel calibration by the calorimetric method was performed twice during the reporting period. After the first calibration, it was found that improperly seated reactor fuel caused a reactivity change which in turn caused one safety channel to read lower than its proper value. The total worth of the regulating, shim, and transient rods were measured to be
$3.99, $3.06, and $2.41, respectively. The largest change in worth was 5.8% of total worth on the reg rod, which is consistent with the small changes in rod worth due to rotational changes of position of individual fuel elements from fuel movement during approach to critical experiments and fuelinspection.
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r-1 U. S. Nuclear Rngulatory Commission pago 2 August 20,1998
'No fuel elements were measured for length or bend during the reporting period. The number of pulses and elapsed time since the last i
measurement are less than the surveillance requirements in the facility technical specifications.
Maximum reactivity insertion rates of $0.17/sec, $0.10/sec, and $0.16/sec were measured for the regulating, shim, and transient rods, respectively.
All three insertion rates were less than the maximum rate allowed by the facility technical specifications.
The transient rod drive assembly was inspected twice during the reporting period. Both piston seals were found to be in satisfactory condition and no wear or rust accumulation was present in the air cylinder.
Rod drop times from full out to full insertion were measured to be 0.33, 0.31, and 0.53 seconds for the regulating, shim, and transient rods, respectively. There was no appreciable change in the drop times of the control rods since the last rod drop measurements. All three drop times were less than the time required by the facility technical specifications. The l
regulating, shim and transient rods were visually inspected during the 4
l
' reporting period. All rods showed only minimal wear.
l l
The area radiation monitors, the pool activity monitor and the pool conductivity meter were calibrated during the reporting period.
2.
The reactor was critical for a total of 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />, producing 2259 kW-hours (0.094 mW-day) of thermal energy. The cumulative energy output since the facility _was commissioned is 9.506 MW-days. During the reporting period 13 pulses with input reactivity greater than $1.00 were performed. The cumulative number of pulses greater than $1.00 since the time pulsing was initiated is 2129.
The reactor was in operation 69 days during the reporting period, with 124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br /> of operating time, as recorded by the console clock.
3.
There were two inadvertent reactor scrams during the reporting period. On 7/30/97 a left safety channel scram occurred at 106 kW. This occurred during a three hour irradiation at 100 kW. As the pool temperature increased, the tubulation for the water-tight canister housing the neutron
l U. S. Nucisar R:gulatory Commission pago 3 August 20,1998
' detector for the left safety channel elongated, moving the chamber closer to the reactor core. This caused a higher reading on the left safety channel i
l which led to a reactor scram. No safety limit was exceeded. On 11/4/97 a right safety channel scram occurred during a pulse. When switching to pulse mode, the mode selector switch did not disconnect the right safety channel r.nd the right safety channel scrammed at 106 kW while performing a pulsa. The system failed safe and the reactor did not operate outside its allowed power envelope. The mode switch was repaired, before resuming reactor operations, and later modified to prevent a similar occurrence.
l 4.
Major maintenance included modification of the mode selector switch and replacement of power supply PS 502. Minor maintenance items included servicing the C.A.M. air pump, changing filter cartridges in the water purification system, adding pool water lost by evaporation, replacing burned out light bulbs in the reactor pool, replacing burned out annunciator bulbs in the reactor control console, and making periodic adjustments to the reactor control console circuitry.
i 5.
The Reactor Committee met four times during the reporting period: 9/3/97, 10/22/97,2/18/98, and 5/14/98.
At its meetings and in individual reviews by Committee members, the Committee reviewed operations and operational records of the facility as specified by the Committee charter. This included audit of preliminary check sheets, pulsing check sheets, approach to critical and termination check sheets, operations and maintenance log books, monthly and annual check sheets, irradiation records, and experiments performed with the reactor. The Committee received reports about new operator trainees, laboratory class enrollment, and reviewed the 1997 annual report to the NRC.
The Reactor Committee reviewed a request (attached) from the Reactor Supervisor to accept test specifications for a power supply to replace failed power supply PS 502 in the control console. A direct replacement for the I
failed supply was not available and facility technical specifications require test specifications for a replacement supply to be approved by the Committee. The Committee concurred that the requested test specifications met or exceeded the specifications for the original supply and the acceptance of these specifications did not involve a change in the facility technical specifications or an unreviewed safety question.
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U. S. Nuclaar Regulatory Commission paga 4 l
August 20,1998 The Reactor Committee reviewed a 10CFR50.59 safety evaluation (attached) for a modification of the console mode selector switch.
l Malfunction of this switch caused an unintentional reactor scram on 11/7/97 l
and modification would eliminate a recurrance of situation. The Committee l
concurred that the proposed modification did not involve a change in the i
facility technical specification or an unreviewed safety question and gave j
approval for the modification.
i 6.
No liquid or solid waste was discharged from the facility during the reporting j
L period.
Measurements of the Argon-41 concentration in the reactor pool water have i
demonstrated that the maximum rate of release of Argon-41 from reactor pool water is less than 0.74 pCi per kilowatt-hr of reactor operation. The pneumatic transfer system produces approximately 0.05 Cl of Argon-41 j
per kW-min of reactor operation, some of which is released when the system is operated. Presented below are the calculations of the maximum j
l monthly releases of Argon-41 from the reactor pool surface, the pneumatic j
l transfer system and the totals.
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' Transfer Systerd ^' E:
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July 1997 741.9 0.0 741.9 j
August 1997 0.0 0.0 0.0 j.
September 1997 117.7 0.0 117.7 l-l October 1997 296.5 0.0 296.5 l
November 1997 14.4 0.0 14.4 December 1997 37.6 0.0 37.6 January 1998 285.1 0.0 285.1 February 1998 2.1 0.0 2.1 1
March 1998 0.1 0.0 0.1 April 1998 14.3 0.0 14.3 May 1998 32.8 0.0 32.8 June 1998 129.1 195 324.1 i
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U. S. Nuclect Regulatdry Commission pago 5 August 20,1998 L
The daily calculations for Argon-41 release from the pneumatic transfer l-system did not include decay of the isotope prior to release and, therefore, give an over-estimate of Argon-41 release. The maximum total estimated Argon-41 release from the facility during the reporting period is 1.87 millicuries. There were no other gaseous effluents from the facility during the reporting period.
7.
Twelve (12) persons were issued film badges on a monthly basis for all or i
part of the reporting period in the Nuclear Reactor Laboratory. The persons l
l receiving badges included all reactor operators, faculty and staff members l
using the reactor laboratory, researchers, and all students in laboratory j
courses. Five badged individuals were reported as having received L
exposures of non-penetrating radiation from 10-20 mrem on body badges.
l l
One hundred seventy-one (171) non-badged persons were admitted to the I
Reactor Laboratory in classes, tours, or on official business during the i
reporting period. All groups were issued pocket dosimeters. Pocket i
dosimeters issued to visitors indicated that no exposure was received.
l 8.
Radiation surveys of the reactor room, control room, and experiment set-up room were conducted monthly during the reporting period by members of the University of Arizona Radiation Control Office using direct measurement l
l and wipe tests. The results show little detectable activity except where j
expected (i.e., irradiated samples in storage areas and internal wall i
surfaces of the irradiation facilities). Other radiation surveys were performed by members of the reactor laboratory staff when necessary. No L
radiation exposure which can be attributed to reactor operations has been l
detected outside the reactor laboratory.
l 9.
Environmental TLD monitors at 3 locations on the building housing the reactor and at 10 other locations on the University campus were replaced l
and read quarterly during the reporting period. For the 12-month period from July 1,1997, through June 30,1998, the average yearly total of the 3 TLDs located on the reactor building roof was 92.3 mrem and the average i
of the 10 other TLDs was 111.5 mrem, after subtraction of the average reading of two control TLDs, which were kept in a shielded container in a non-radiation area. This is consistent with similar measurements for these locations in previous years. Thus, there is no evidence that radiation exposures in the vicinity of the reactor are higher than normal. In January, 1994, eight TLD monitors were placed at the periphery of the restricted area. In April,1994, two TLD monitors were placed in an office area far L-
l U. S. Nucirr R:gulatory Commission page 6 i
l August 20,1998 removed from the restricted area to provide a baseline reference for building background. The lowest total exposure for the reporting period at i
the restricted area periphery was 5 mrem, while the highest was 99 mrem.
Exposure to the public was less than 100 mrem / year. Surveys performed at the periphery of the restricted area with the reactor operating at full power showed the dose rate to be much less than 2 mrem / hour.
In writing this report, I have tried to be both complete and as brief as is rea-sonable, and still satisfy the requirements of 10CFR50.59, the Facility l
Technical Specifications, and the needs of the Commission. If other or more l
detailed information is needed, please contact me at your earliest convenience.
Sincerely,fW John G. Williams, Director Nuclear Reactor Laboratory JGW:HD/dg l
cc:
Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation One White Flint North 11555 Rockville Pike l
Rockville, MD 20852-2738 l
Mr. Marvin Mendonca Project Manager USNRC Dr. Michael Cusanovich Vice President for Research University of Arizona l
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i Date: 5/20/98 t
Merho To:
Reactor Committee.
From: H. Doane, Reactor Supervisor Re:
Failed Power Supply in TRIGA console i
l On May 14,1998 during the performance of Preliminary Checklist No. 4879,it was f
discovered that power supply 502 had failed.- This is a 12 volt,3 amp DC supply (Technipower Power Supply PL-12.0-3.0) that provides power to the vertical light panels on the reactor console and to the relay that operates the exhaust fan.
l Technical Specification 4.5.b (Attachment A) requires the testing of any addition, q
modification, or maintenance in accordance with the specifications to which the systems L
(PS502) were originally designed and fabricated pf to specifications approved by the Rextor Committee.. Unfortunately, General Atomics (original manufacturer of our TRIGA and now called Sorrento Electronics) is not able to supply an exact replacement l
nor the original specifications for PS502 (Attachment B).
l Reference to Technipower modular, linear, D C power supplies was found in a 1980/1981 Electronic Design's GOLD BOOK (Attachment C). Most current manufactures of linear l
D C power supplies have specifications equal to or exceeding the specifications.given for l
Technipower supplies in the GOLD BOOK.
On May 20,1998, all interlock and SCRAM circuits on the TRIGA console were tested l
with the optput of PS502 open and shorted together (Operations Log 44, page 249). All l
interlocks and SCRAMS tested operational with an open or short circuit at the output of PS 502. Therefore, the state of PS 502 does not affect the safe operation of the reactor and PS 502 need not require a ridgid set of manufacturing specifications.
Section 3 of the U of A Reactor Committee Charter (Attachment D) allows for i
Committee decisions to be made without a meeting by the written affirmative vote of a majority of Committee members, who shall have reviewed the proposed. This decision shall be reported to all Committee members (at the next scheduled meeting).
a I am enclosing a voting form for establishing Committee specifications for PS.iO2 (Attachment E). Please review the attachments and execute your vote on Attachment E.
To save time, this memo will be hand carried to each committee member.
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A T rA C HNcN T A page 21 4.5 Maintenance Acolicability This specification applies to the surveillance requirements following maintenance of a i
l control or safety system.
Ot.lc.ttva The objective is to assure that a system is operable before being used l
after maintenance has been performed.
Specification Following maintenance or modification of a control or safety system or component, l
a.
l it shall be verified that the system is operable prior to its retum to service. A l
system shall not be considered operable until after it is successfully tested.
l
- b. Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and l
tested in accordance with the specifications to which the systems were originally l
designed and fabricated or to specifications approved by the Reactor Committee.
l A licensed reactor operator shall be present during maintenance of the reactor c.
i control and safety system.
Hasis This specification relates to changes in reactor systems which could directly affect the l
safety of the reactor. Changes or replacements to these systems which meet the original design specifications are considered to meet the presently accepted operating criteria.
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v Tetephone: 1800 TRioA-55 Fax 619-457-8786 Via Fax:(520) 621-8096 5/18/1998 University of Arizona Department of Nuclear & Energy Engineering Tucson, AZ 87521
Subject:
Failed Technipower Power Supply PL-12.0-3.0
Reference:
Telephone conversations 5-15 98 and 5-18-98 Attachments: None
Dear Sir or Madam:
Sorrento Electronics (SE), an affiliate company of General Atomics, in an effort to assist the University of Arizona TRIGA* reactor recommends the following action.
The original manufacturer no longer supports the power supply in question. As there is no available specification for this supply, we have found this supply to be a commercial grade supply of 12VDC at a rated current of 3.0 amps. To the best of our knowledge there is no ripple or noise specification that will preclude using any commercial equivalent. It must be noted that no testing was performed to verify this conclusion. This verification is the University of Arizona TRIGA* reactor responsibility.
We value our continuing relationship and look forward to continuing to be of service to the University of Arizona TRIGA* reactor. If there are any questions, I may be reached by fax at (619) 457-8786 or e-mail at John.faircloth2 gat.com. You may also contact Dr. Junaid Razvi at razvi@ gat.com.
Best regards, Johf R. F ircloth S,ctuor gineer 8
TRIGA Products and Services AN AFFILIATE OF GENERhl ATOMICS 10140 PLANCEP$ COURT. SAN 01G0 OA 92'21-M90 I
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Date: 5/18/98 I
To: File
+
Subject:
Reactor Power Supply 502 Replacement.
The Gulf TRIGA Reactors - Gulf Energy & Environmental Systems Manual identifies the power supply manufacture as TECH (Technipower) with Model # PL12.0-3.0. The supply is ra'ed at 12 V and 3 A. Gulf does not stock any replacement power supplies nor can they supply us with the technical specifications.
A 1980/1981 Electronic Design's GOLD BOOK describes Technipower - Mil-Qualified, High Reliability.
Modular Power Sources with the following specifications:
base temperature 95 *C source and load effect 0.05 %
PARD (ripple) i i my short circuit, overvoltage protection We are replacing the supply with a SOLA model SLS-12-034 which is rated at 12 V and 3.4 A.
. The Specifications are:
Operation Temperature Range:
0 to 50 *C Line and Load Regulation:
Combined 0.1 %
Ripple:
3.0 mV maximum peak to peak Overload Protection:
Automatic Current Limiting I
- Overvoltage Protection:
built in Because the GOLD BOOK does not identify the power supplies by model, we do not know if the numbers apply to our model or are the "best" for that family of supplies.
For the application, which is powering console panel lamps and picking a relay, the replacement supply is adequate. Any specification differences are minor and will not affect the operation of the reactor.
Both supplies use linear series regulators with state of the art electronic components.
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3.
Meetino Frecuency and Quorym A quorum of the RC shall consist of at least three voting members and shall include the chairman or his designee. With the approval cf the t
chairman, decisions of the Committee may be made without a meeting b the written affirmative vote of a majority of Committee members, who shall
)
i have reviewed the proposal. The decision shall be reported to all Committee members.
The RC shall meet at least quarterly. This shall mean: At least once l
each calendar quarter, at intervals not to exceed four months.
4 I
4.
Minutes of Meetings Minutos of all RC meetings shall be taken by the chairman or his designee and shall be distributed to all RC members prior to the next scheduled meeting. These minutes shali be kept as a permanent file in the Nuclear Reactor Laboratory.
5.
Use of Subcommittees The Chairman, at his discretion, may appoint subcommittees to evaluate the specific items. A subcommittee shall consist of one or more RC members.
1 The RC may vote to delegate authority to a subecmmittee to take action on a specific issue. When such authority is given, the actions of the subcommittee have the same authority as action taken by the RC. Any such actions taken by any subcommittee must be reviewed and approved by the RC at the next scheduled meeting.
6.
Presentation of Information to the RC Information pertinent to the operation of the NRL will normally be presented to the RC by the Director or Supervisor of the NRL. Any non-members desiring to present such information to the RC may do so with the prior approval of the RC chairman. The RC chairman will determine which items of information will be addressed during any meeting.
Ill. RESPONSIBILITIES 1.
Audit The chairman of the RC shall appoint incividual members er subcommittees to audit operaticns cf the NRL fcr cperating periccs net ::
Uthersity cf AnxM,%5&3rcn.%3clcr wmna e 1
l.
ATTecRMcNT C Manufacturer's Performance Specification for PS 502
_12 VDC @ 3 amps j
Operation Temperature Range:
0 to 50 'C i
. Line and Load Regulation:
Combined 0.1 % -
4 Ripple:
3.0 mV maximum peak to peak Overload Protection:
Automatic Current Limiting.
. Overvoltage Protection:
built in i
. Proposed Test Specification for Acceptance at the University of Arizona Research.
Reactor L
Adjust voltage at no load to 12.00 VDC Load supply to 3 amps Measure voltage
. Accept i0.5 % of no load voltage
{
Substitution of power supply 502 as described in H. Doane's memo of 5/20/98, with a l
t power suont meeting the above performance and test specifications does not constitute a f
change m the technical specifications or an unreviewed safety question for the University(
of Arizona Research Reactor.
f Specification Approved
- Specification Not Approved Signed [
_/
Date _
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SAFETY EVALUATION L
MODIFICATION OF THE CONSOLE MODE SELECTOR SWITCH l
for the University of Arizona Nuclear Reactor Laboratory TRIGA Reactor t
l l INTRODUCTION AND PURPOSE l
Description of a Recent Occurrence On November 4,1997, a pulse was performed with the mode switch in its Pulse High setting and with the transient rod anvil set for a $1.75 pulse. The transient rod operated normally, fuel temperature rise was recorded showing energy deposition in the range i
expected for a normal pulse and the timer scram operated normally. No valid indication appeared, however, for the peak reactor power or the integrated pulse energy. The right safety channel indicated a trip on reactor power.
A digital data acquisition system connected to the right safety channel showed that the signal on that channel displayed a normal reactor period of approximately 11 ms, but then saturated. This showed that the channel was connected to a valid indication of reactor power level from the ion chamber that is normally connected in steady state mode, not the high range ion chamber which should be selected in pulse mode. The trip on the right safety channel was attributed to power in excess of 110 kW, which is the steady state limit.
Testing of the mode switch revealed that changing the switch from its manual to pulse positions did not reliable switch between the normal and high range signals. The problem was corrected after removal of the rear casing of the mode switch and application of switch cleaning fluid. After reassembly, the system appeared to behave normally. The reactor was then operated at 95 KW steady state while the high range channel was calibrated and shown to behave normally. A pulse was then performed j
and normal indications were found on all channels.
Analysis and Diaanosis of the Cause I
A malfunction of the mode switch was the cause. It resulted in failure to pick a relay which switches from the normal to the high range channel when pulse mode is selected, and which also switches in the nyt circuitry that indicates integrated pulse energy.
The following symptoms demonstrate that the high range channel was not engaged, but i
l instead the normal range uncompensated ion chamber was operating: (a) the reactor i
tripped on a high power scram from the right safety channel, even though the calculated peak power was less than 100 MW, and the trip was set at 1000 MW for pulse mode; (b) fuel temperature indications showed a peak temperature of 186 C, in the normal range for a $1.75 pulse, corresponding to energy deposition of 9.3 MW and peak power of 92.4 MW; (c) a digital data collection system connected to the channel UnNorsity of Arinone Research Reector muca.es.e-se as
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pags 2 (after the switching relay) saturated, though it should not have done so below 1100 MW; (d) prior to saturation, the same data showed an exponential increase with a period of 11 ms, normal for a $1.75 pulse, an indication that the channel was connected to a true power indication, though not the high range one.
~
On November 10,1997, the mode switch was removed from the console for inspection and cleaning (see attached memo and part drawing). As a result of this inspection, it was determined that the problem had actually been caused by a mechanical misalignment of the rear wafer on the switch, not by dirt. This was attributed to wear and aging that had loosened the nuts holding the rear case. This also loosened the wafer assembly which is held together by the same nuts. The problem was corrected not by the cleaning fluid, as initially supposed, but by the re-tichtening of these nuts that occurred on reassembly.
l
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L Corrective and Remedial Action l
I The immediate problem was corrected before the reactor was placed back in service on November 4. Tightening of the nuts holding the casing of the mode switch also secured the switch wafers in their correct operating alignment.
To prevent a recurrence, the Electronics Technician and the Reactor Supervisor have proposed a small modification in the switch assembiy, so that the wafer assembly is held together independently of the nuts which secure the rear casing.
I Definition of 10CFR50.59 l
10CFR50.59, Changes, Tests and Experiments, permits the license holder to "make changes in the facility as described in the safety analysis report... without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question." The regulation further states that:
A proposed change... shall be deemed to involve an unreviewed safety question (i)if the probability of occurrence or the consequences of an' accident of malfunction of equipment important to safety previously evaluated in the safety analysis report may increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the ll basis for any technical specification is reduced.
The licensee shall maintain records of changes in the facility... to the extent that these changes constitute changes in the facility as described in the safety analysis report.... These records must include a written safety evaluation which UnkersityofMtone Resentch Reactor
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.provides the bases for the determination that the change... does not involve an unreviewed safety question.
The licensee shall submit, as specified in $50.4, a repcrt containing a brief description of any changes, tests, and experiments, including a summary of the i
safety evaluation of each. The report may be submitted annually or along with j
the FSAR updates as required by $50.71(e), or at such shorter intervals as may be specified in the license.
The purpose of th'.s avaluation is to document sufficient facts to justify identifying the modification of the mode selector switch as not being an unreviewed safety question.
l Procosed Modification The mode selector switch (General Atomics part number ELD 239-4300C) is a five position, four wafer switch assembly. It is covered by a then aluminum housing that is l
made in two pieces. The kny,b shaft, the wafer assemblies and the aluminum cover are all held together by two th'ae-inch long bolts (see figure A). The two nuts that secure these bolts in place are located outside the rear section of the cover. The loosening of thue nuts (over time) is what caused the occurrence described above.
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Figure A. Mode Selector Switch Before Proposed Modification The proposed modification would shorten the two insulated spacers (between the rear wafer 9nd the inside of rear section of the cover) by the thickness of one nut plus the thickness of one lock washer (compressed). The locit washer and the nut would be placed on the bolt directly behind the rear wafer. The switch wafers would now be aligned and supported independent of the rear cover section. The two shortened insulating spacers would be placed over the bolts and the rear cover section would also UNversty of Arkone Research Reactor a,u u
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l paga 4 be placed over the bolts. Two nuts would then secure the rear cover section to the already positioned and secured switch assembly (see figure B).
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Figure B.
Mode Selector Switch After Proposed Modification The modification does not change any physical dimensions of the mode selector switch or require any changes in the mounting of the switch in the control console. The i
modification does not reduce electrical isolation of the wires attached to the wafers of the switch. The modification provides for a more secure alignment of the switch wafers l
relative to the position of the knob shaft. By having structural switch support independent of the rear cover section, it is believed that the probability of the occurrence described above would be greatly reduced.
dVALUATION DOES THE PROPOSED MODIFICATION INVOLVE A CHANGE IN THE TECHNICAL SPECIFICATIONS INCORPORATED IN THE FACILITY LICENSE 7 l
The mode selector switch (MSS) is not mentioned in the facility technical specifications.
l Therefore, any modification to the MSS does not involve a change in the technical l
specifications.
l l
l LTRIGAConsowsstSw SE #8 M
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IS THE PROBABILITY OF OCCURRENCE OR ARE THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAFETY ANALYSIS REPORT INCREASED BY THE CHANGE 7 The Safety Analysis Report describes the function of the MSS on page 44. It states that 1
the MSS is divided into three isolated sections to assure the independence of the operation system and the two redundant portions of the protection system. As
- demonstrated above, the preposed modification to the MSS does not reduce the electrical isolation of these three switch sections, nor does the proposed modification l
change in any way the individual circuit paths defined by the unique positions of the switch. The proposed modification does not affect the probability of occurrence of an l
. accident or malfunction of equipment important to safety which has been previously l
evaluated.
L DOES THE CHANGE CREATE THE POSSIBILITY FOR AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE SAFETY ANALYSIS REPORT 7 The proposed modification of the MSS does not create the possibility for an accident or l
malfunction of a different type than any evaluated previously in the safety analysis j
report. As described above, the proposed modification does not alter the physical size, console mounting, wiring positioning or section isolation. The "non-physical" proposed l
modification does not contribute to the creation of any new accident modes.
IS THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION REDUCED 7 l
The proposed modification does not affect the basis for any technical specification. The l
proposed modification will assure secure alignment of the switch wafers relative to the knob shaft, thus reducing the probability of the bolts securing the MSS loosening over f
time.
CONCLUSION Based on the information provided, the proposed modification to the MSS does not involve a change in the facility technical specifications nor does it constitute an Unreviewed Safety Question as defined in 10CFR50.59. Having met the requirements of 10CFR50.59, the proposed modification to the MSS may be implemented without prior U.S. Nuclear Regulatory Commission approval.
Unkersity of Annone Research Reector ew-se ao
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r 1%ur 1 An A//c /9 9 Prepared b poane Date Reactor S rVisor I
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Reviewed by W. Lohmeier Date Reactor Electronics Technician i
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/$ proved by John G. Williams, Ph.D.
Date Director, Nuclear Reactor Laboratory A
df/S 98
~
pproved by P. A. Secky,r(fr.,Th.D.
Date Chairman, Reactor Committee for the Reactor Committee l
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