ML20210R122
| ML20210R122 | |
| Person / Time | |
|---|---|
| Site: | Dow Chemical Company |
| Issue date: | 08/19/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20210R101 | List: |
| References | |
| NUDOCS 9709020314 | |
| Download: ML20210R122 (10) | |
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}AFETY EVALVATION BY THE QFFICE OF NUCLEAR REACTOR REGULATION
}yPPORTING AMENDHENT NO. 7 TO o,ClllTY OPERATING LICENSE NO. R 108 THE DOW CHEMICAL COMPANY DOCKET NO. 50 264
1.0 INTRODUCTION
By letter dated May 5,1994, as sup)1emented on february 8 and August 7,1995, June 18 and October 8, 1996, and fe)ruary 24 and June 16, 1997, the Dow Chemical Company (the licensee or Dow) applied for an amendment to the operating license for its TRIGA Hark i research reactor located in Midland, Michigan.
The licensee requested that the technical specifications (TSs) section of the license be amended to (1) delete the requirement for a scram i
function on minimum reactor period (2) change the schedule for periodic inspection of reactor fuel, (3) add a defined quarterly surveillance interval, (4) modify the organizational structure for the facility, and (5) make other changes that are administrative in nature to clarify or update the TSs.
The application included technical justification for the proposed changes and a copy of the proposed TSs rewritten in their entirety.
-2.0 EVALUATION The Dow facility houses a Mark I TRIGA reactor using standard uranium-zirconium hydride fuel and is licensed to operate at steady-state thermal power levels up to 300 kW.
The reactor was initially licensed in 1967 to operate at power levels up to 100 kW, and in 1989, during license renewal, an application to operate at 300 kW was evaluated and approved by NRC.
By the licensee's request, the renewed license authorizes the use of stainless steel clad TRIGA fuel except 4
for one aluminum clad element in either the E or F ring of the core. Many TRIGA type reactors operate routinely at power levels at or above 1000 kW with the stainless steel clad type of fuel, and with similar control systems and coolant systems, in addition, many of these TRIGA.jpe reactors are authorized to operate in the pulsed mode, which places additional stresses on fuel elements and components.
However, the Dow reactor is licensed to o)erate only in the steady state mode because pulsing has not been required by tie Dow research and utilization program.
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The Dow reactor was initially installed to operate with the standard analog instrumentation and control system routinely authorized for 1RIGA reactors at that time.
By license amendment in 1993, the licensee was authorized to replace some of the facility's analog instruments with digital computer based systems similar to those already approved and operating on several NRC-licensed TRIGA non power reactors.
2.1 Deletion of Scram function on f.eactor Period TS 3.3 and Table 3.3A, " Minimum Reactor Safety Circuits, Interlocks, and Set Points " currently require an automatic reactor scram if the e folding reactor period for )ower level should decrease to 7 seconds or less.
The licensee requested t1at this scram requirement be deleted from Table 3.3A and be replaced with an interlock requirement that would prevent additional control rod withdrawal if the period decreased to less than 3 seconds.
The licensee also proposed changes to the bases of the 15 to reflect the change from scram to interlock.
The basis for the period interlock in Table 3.3A was proposed by the licensee because it " prevents operation in a regime in which transients could cause the limiting safety system setting to be exceeded." With the addition of the period interlock, the licensee also proposed a change to TS 4.2.3 to refer to surveillance of three interlocks instead of two.
The licensee cited other TRIGA reactors using a period interlock system instead of a scram, such as the Mark I at General Atomics, and referenced discussions of the safety of other licensed TRIGA reactors, for example, the Mark I at Reed College, with maximum excess reactivity limits similar to those in the TSs of the Dow facility.
On the basis of this $1 formation, the licensee stated that rapid insertion of the total available excess reactivity would not cause the maximum fuel tem-perature to reach the safety limit of even the one aluminum clad element in current use.
The NRC staff, in NUREG 1312, " Safety Evaluation Report Related to the Renewal of the facility License for the Research Reactor at the Dow Chemical Company," Section 14.2, also concluded that the rapid insertion of the total available excess reactivity (3.00$) would not cause the maximum fuel temperature to reach the fuel safety limit.
The licensee stated that it was desirable to maintain a limit on the minimum reactor period to ensure safety and to alloy for more reliable operational control of the reactor at all times.
This control can be achieved by means of the proposed interlock.
Therefore, the reactor scram, although it seems a more limiting restriction, is not necessary for a TRIGA reactor with the core size and reactivity limits of the Dow reactor.
The licensee also pointed out that the software and hardware capability to implement this change already exists in the computer-based instrument and control system because such an interlock had been approved by NRC for use on a reactor for the designer and
L vendor (Genera' Atomics) of that system.
Therefore, verification and valida-tion _of the necessary software have already been performed and have been evaluated by NRd.
The staff finds that the licensee's safety considerations and the proposed substitution of the-minimum-period interlock for the minimum-perica scram for this reactor is acceptable. This automatic protective interlock gives reasonable assurance that the facility can continue to be operated safely.
Therefore, this change in operating conditions and the associated change in the Dow facilit) TSs is acceptable.
2.2 Inspection of Reactor fuel The staff requires that licensees of non-power reactors establish procedures to ensure the.t the reactor is not knowingly operated with damaged fuel.
Accordingly. TS 4.5 of the Dow reactor license currently reads:
Each fuel element shall be visually examined annually.
The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal.
A TRIGA fuel element shall be removed from the core ift a)
The transverse bend exceeds 0.125 inch over the length of the
- cladding, b)
The length exceeds the original length by 0.125 inch, c) A clad defect exists as indicated by release of fission products.
The licensee-has proposed that TS 4.5 be amended to read:
Each fuel element shall be examined visually and for changes in transverse bend and length at least once each five years, with at least 20 percent of the fuel elements examined each year.
If a damaged fuel element is identified, the entire inventory of fuel elements will be inspected prior to subsequent operations.
The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal.
A TRIGA fuel element shall be considered damaged and removed from the core ift a)' The transverse bend exceeds 0.125 inch over the length of the cladding.
-b) -The length exceeds the original length by 0.125 inch, c) A clad defect exists as indicated by release of fission products.
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.o The licensee has proposed additional wording for the basis of the TS that reads:
Experience at the Dow TRIGA Research reactor and at other TRIGA reactors indicates that examination of a five-year cycle is adequate to detect problems, especially in TRIGA reactors that are not pulsed.
A five-year cycle reduces the handling of the fuel elements and thus reduces the risk of accident or damage due to handling.
The Itcensee has requested that the average interval between inspections be extended by allowing a different one-fif th of the elements to be inspected annually.
This measure would extend the time between insoections of any one fuel element to 5 years.
This change is justified on the following bases:
when the facility was initially licensed, it was prudently conservative to inspect fuel elements frequently.
The inspection could reveal defects in fabrication or early deterioration.
in fact, one fuel rod released fission products shortly _after reactor startup. However, after more than two decades of o>eration with no additional fuel failures, there is reasonable assue*ance of tie high integrity of the fuel in the Dow reactor.
In adriition, (1) experience with many hundreds of TRIGA fuel rods fabricated to the same standards and operated in other reactors for at least as long and at thermal power levels several times those licenaed at Dow has confirmed both the initial and the continuing integrity of TRIGA fuel operated at steady-state power levels; (2) the Dow TRIGA reactor is authorized to be operated only in the steady-state mode, and it is believed that pulse operation would subject TRIGA fuel to more severe o>erational stresses; (3) the proposed decrease in inspection frequency would 3e consistent with the frequency allowed at several other licensed TRIGA reactors; and (4) the proposed decrease in inspection frequency would decrease the likelihood of fuel damage due just to fuel handling, and could decrease radiation exposure of the staff, in accordance with the as low as reasonable achievable (ALARA) principle, in addition to the annual visual inspection of the fuel elements required by the TSs, the licensee currently implements an internal procedure that providos objective measurements of fuel element length and bend on a 5-year schedule.
This procedure is similar to recommendations of the TRIGA fuel vendor and to requirements included in the TSs of other TRIGA reactors.
The proposed TSs include these additional procedures in order to improve clarity and spect-ficity and to compensate for deleting the annual visual inspection.
Because the licensee was already performing the objective measunements under internal procedures, this change in the TSs does not increase the licensee't workload.
The staff finds that the information and considerations provided by the licensee and the proposed changes in the fuel inspection specification are acceptable.
2.3 Adding a Defined Surveillance Interval
-The licensee requested that a quarterly interval be added to the set of defined intervals at the beginning of Section 4 of the TS, " Surveillance Requirements," for actions such as holding meetings or performing surveillance functions.
The proposed definition is " quarterly - not to exceed four months." The proposed added interval requires the action to be performed at least four times during a calendar year of normal o)eration.
Some meetings are required to be held quarterly by Section 6 of tie TS. Adding this quarterly interval is consistent with other non-power reactor TSs acceptable to NRC, and with recommendations of American Itational-Standards Institute /
American Nuclear Society ANSI /ANS-15.1, ' Development of Technical Speci-fications for Research Reactors." The staff finds this change acceptable.
2.4' Changes to the Description of Nuclear Instrumentation The licensee proposed a number of changes to the TSs to better describe the nuclear instrumentation. -The changes to the TSs address the NM1000 and NPP1000 nuclear instrumentation.
The NPP1000 is a channel consisting of an ton chamber neutron detector, a single linear reactor power readout, and one linear input to the rea;; tor protection circuit.
This channel is entirely analog in operation.
The NM1000 is a channel consisting of a fission neutron detector, signal processing circuitry, and the following outputs: wide-range linear reactor power readout, linear input to the reactor protection circuit, logarithm of reactor power readout, reactor period.eadout, and reactor period input to the control rod interlock circuitry.
Because the signal processing circuitry is at least partially digital, the NM1000 channel is non-analog.
.The TSs pertaining to these channels-were not internally consistent but have now been changed so that they are. The revised TSs still require these two types of channels for reactor operation, but the componants are described in the safety analysis report rather than in the TSs themselves.
The specific changes to the TSs are described below.
The licensee proposed changes to Table 3.3A for the descriptions of some of the scram channels.
The licensee proposed to change " Reactor Power Level" to
" Reactor Power Level NM1000 & NPP1000"; the Wide-Range Linear / Log Channel Detector Power Supply" to "NPP1000 Detector High-Voltage Power Supply"; and the " Percent Power Channel Detector Power Supply" to "NM1000 Detector High-Voltage Power Supply." The licensee also proposed to change the bases for this table to replace reference to the wide-range linear-power channel and the wide-range log power channel with references to NPP1000 and NM1000 power-channels.-
The licensee proposed changes to Table 3.38, " Measuring Channels," and the bases for the table that changed the description of the measuring channels and reduced-the number of channels listed. The licensee replaced " Wide-range Log
-N and Period Channel" with "NM1000" and combined " Power-Level Channel (Linear)" and " Power-Level Channel (Percent Power)" into one channel and replaced them with 'NPP1000." The new description more accurately describes the measuring channels.
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6-The licensee also proposed changes to TS 4.2, " Reactor Control and Safety i
Systems," to consistently use the channel descriptions NPP1000 and NM1000.
In TS 4.2.2, the licensee replaced " wide range linear power channel" with 'NM1000 power level channel."
In TS 4.2.3, the licensee removed the reference to the log power channel because it is already referred to in the TSs and further reference is redundant.
The licensee changed the basis of TS 4.2.2 by replacing reference to the " wide range linear channel" with reference to the "NM1000 channel."
Reference is made in the basis to three neutron detectors.
This refereace is in error and has been replaced with a reference to two neutron detectors.
The staff has determined that these changes are administrative in nature and improve the description of the nuclear instrumentation in the TSs.
Therefore, these changes are acceptable to the staff.
2.5 Organizational Changes The licensee has proposed changes to the organization for the reactor.
The current TSs referred inconsistently to the Facility Director, whose functions were not defined, in response to a request for additional information from the staff, the licensee proposed revisions to Figure 6.1 and TSs 6.1.1 and 6.6.2 to include the functions of the Facility Director and to consistently refer to the position in the management of reactor operations and the operating license.
The Itcensee has added a definition to the TSs for the facility Director that reads:
1.10.
Facility Director -
Person with line management responsibility to whom the Reactor Supervisor reports.
The person also serves as chairperson of the Reactor Operations Committee.
This title of " Facility Director" replaces the title of " Analytical Laboratory Research Manager." The licensee also changed Section 6. " Administrative Controls " of the TSs to reflect the addition of this definition.
This definition 'larifies the TSs and is therefore acceptable to the staff.
The licensee has proposed to change the name of the " Analytical Laboratory" to the " Analytical Sciences Laboratory" in TS 6.1, " Organization." This change is administrative in nature and is acceptable to the staff.
The licensee changed Figure 6.1, " Administration," by updating titles and clarifying the organizational relationships for the radiation safety and reactor o)erations functions.
Compared to the current TSs, the inter-relations 11ps between the radiation safety and reactor operations functions are better shown in the proposed figure.
The Radiation Safety Committee (RSC) holds the reactor license for the Dow Chemical Company.
The RSC also oversees the radiation safety program and is responsible for its implementation.
The Manager of Industrial Hygiene
___.m 1
Research and Technology is responsible for the radiation safety function and reports to the RSC in matters of radiation safety.
The Radiation Safety Officer is responsible for radiation safety and has lines of comunication with the RSC, the Reactor Operations Comittee (ROC), and the Reactor Supervisor.
The Radiation Safety Officer reports to the Supervisor, Industrial Hygiene, who, in turn, reports to the Managar of Industrial Hygiene Research and Technology.
A Senior Research Manager is responsible for the reactor and reports to the RSC in matters of reactor operations.
The licensed reactor operators and senior reactor operators report to the Reactor Supervisor, who reports to the Facility Director, who reports to the Senior Research Manager.
The ROC is responsible for the retien and audit function for the reactor.
The ROC has lines of comunication with the facility Director, the Reactor Super-visor, the RSC, and the Radiation Safety Officer.
The ROC reports to the Senior Research Manager, who is the comittee chair.
The licensee has proposed changes to TSs 6.1.1, 6.1.?, and 6.2.1 consistent with the proposed organizational chart and to allow the Facility Director end the Reactor Supervisor positions to be held by the same person.
Although more complex than most non-power recctor organizations because of its corporate nature, the organization for the Dow TRIGA Research Reactor has the organizational attributes suggested '
- ANSI /ANS-15.1, which is supported by the NRC staff.
Four levels of orga.
$ tion are pretent:
the review and audit function reports to upper (Level 1) management and comunicates with the reactor (Level 2) management, and the radiation safety function reports to upper managenent and comunicates with Level 3 management in the person of the Reactor Supervisor. At the working level, the radiation safety function has a reporting chain that is separate from the reporting chain of reactor opera-tions management.
The two reporting chains meet at the RSC.
Because these organizational attributes are present, the propnsed organization is acceptable to the staff.
In TS 6.6.2, "Special Reports," the licensee has proposed changing the description of the staff members whose permanent change must be reported to NRC within 30 days. The TSs currently state that changes in Level 1 and Level 2 personnet must be reported (these are generic terms from ANS-15.1).
The licensee has proposed changing this description to that for the Reactor Supervisor or the Facility Directer.
This tNange is acceptable to the staff because it clarifies the TS.
The staff has considered each of these changes and has determined that in no case would the effectiveness of the Dow TSs or reactor safety be decreased by the changes.
Therefore, these changes are acceptable to the staff.
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! 2.6 Other Changer to the Technical Specifications The licensee also requested changes in the Dow TSs that are considered by the staff to be administrative changes and that are non-technical and non-safety related. These proposed changes are intended to clarify some issues, to update some details, to increase the specificity of some of the contents, or to correct errors or inconsistencies between some TSs and their bases.
The licensee has also retyped the entire TSs to improve the presentation of the TSs.
TS definition 1.9 currently reads:
1.9.
Exoerimental Facilities include the rotary specimen rack, vertical tubes, pneumatic transfer systems, the central thimble, and the area surrounding the core.
The licensee has proposed changing this TS to read:
1.9.
Experimental Facilities include the rotary specimen rack, sample containers replacing fuel elements or dummy fuel elements in the core, pneumatic transfer systems, the central thimble, and the area surrounding the core.
The licensee states that the proposed wording better describes the experimental facilities.
The staff finds this clarifying change acceptable.
TS definition 1.17 currently reads:
1.17.
Radiation Safety Committee (RSC) - The RSC is chartered by the Dow Chemical Company to be responsible for the license for the Dow TRIGA Research Reactor facility.
Because of other changes to the definitions section of the TSs, the licensee has renumbered the TS and has proposed changing this TS to read:
1.18.
Radiation Safety Committee (RSC1 - The RSC is responsible for the administration of all Dow Midland location activities involving the use of radioactive materials and radiation sources including assuring compliance with U.S. NRC regulations.
The licensee states that this proposed definition better reflects the responsibilities of the RSC and was proposed after consultation with the RSC.
The staff finds this change acceptable.
The licensee has-proposed changing " safety-system" to " safety system" in TS definition l.28 (now number TS 1.29) to make the use of the term " safety system" consistent throughout the TSs.
This change is acceptable to the staff because it is editorial in nature and does not change the meaning of the TS.
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l 9-i The licensee has proposed changes in the bases of TS 3.4, " Coolant System,'
i and 4.2.4, " Reactor control and Safety Systems," to replace a reference to a s)ecific number of ywrs of experience (TS 4.2.4 currently say(19 years) with j
t1e systems with a refwrence to the initial date of operation 1967) to more accurately convey the extent of experience with the systems.
This change is administrative in nature and acceptable to the staff.
4 The licensee has proposed a change to TS 3.5, Confinement, to add a reference to ' Door 10" in order to be more-specific about the external dcor referred to in the TS. This change is acceptable to the staff because it clarifles the TS.
- The licensee has proposed a change to TS 3.6, " Radiation Monitoring Systems,"
i to more completely describe the function of the systems.
The TS currantly reads, in part:
A Continuous Air Monitor (CAM) (with readout meter and audible alarm) in the reactor room must be operating during operation of the reactor.
The licensee has proposed this entry be changed to -
A Continuous Air Monitor (CAM)lates in the reactor room must be(wit to measure radioactive particu operating during operation of the reactnr.
This enange is acceptable to the sts'f because it clarifies the 15.
The licensee has proposed changes to TS 6.5. " Required Actions," that would add the word "and" between the last two items of the lists in TSs 6.5.1.a-and 6.5.1.b of required actions in case of-a safety limit violation and to TS 6.5.2.a of required actions in case of a reportable occurrence to clearly indicate. that. all of the actions on the list must be performed.f a safety limit violation or a reportable occurrence takes place.
The staff finds these changes acceptable'because they clarify the TS.
-The licensce proposed changes to TSs 6.6.1 and 6.6.2 to update points of submission of information tc NRC.
This change is editorial in nature and is acceptable to the staff.
The licensee proposed changing the reference to TS 1.28-to TS 1.29 in TSs 6.5.2 and 6.6.2 to reflect the change in the numbering of definitions in Section 1 because of the addition of a new definition.
This_ change is editorial in nature and is acceptable to the staff.
3.0 ENVIRONMENTAL CONSIDERATION
The portions of this amendment discussed in Sections 2.1, 2.2, 2.3, and 2.4 involve changes in the installatica or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes in
y *o inspaction and surveillance requirements.
The staff has determined that these portions of the amendment involve no significant increase in the amounts. and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure.
The portions of the amendment discussed in Sections 2.5 and 2.6 involve changes in recordkeeping, reporting, or aduinistrative pi:,cedures or requirements. Accordingly, these portions of the amendment meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10).
Pursuant to 10 CFR 51.22 statement or environmental assessment need be prepa(b), no environmental impact red in connection with the issuance of these portions of the amendment.
4.0 CONCLUSION
The staff has concluded, on the basis of the considerations discussed above, that (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from any accident pre-viously evaluated, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards cons.deration; t
(2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public.
Principal Contribetors:
A. Adams, Jr.
R. Carter, INEEL Date: ' Aug,ust 19, 1997 j
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