ML063320500
| ML063320500 | |
| Person / Time | |
|---|---|
| Site: | Oregon State University |
| Issue date: | 11/21/2006 |
| From: | Reese S Oregon State University |
| To: | Hughes D NRC/NRR/ADRA/DPR/PRTA |
| References | |
| TAC MC5155 | |
| Download: ML063320500 (14) | |
Text
Radiation Center Oregon State University, 100 Radiation, Corvallis, Oregon 97331-5903 T 541-737-2341 I F 541-737-0480 I http://ne.oregonstate.edu/facilities/radiationcenter Oregon State UIVERSITY November 21, 2006 Mr. Daniel Hughes U. S. Nuclear Regulatory Commission Research and Test Reactors Branch A Office of Nuclear Reactor Regulation Mail Stop 012-G15 One White Flint North 11545 Rockville Pike Rockville, MD 20852-2738
Reference:
Oregon State University TRIGA Reactor (OSTR)Docket No. 50-243, License No. R-106 Request for Additional Information Regarding License Renewal, Oregon State University TRIGA Reactor (TAC No. MC5155), dated October 3, 2006 Request for Extension 60 days to Respond to the RAI, dated November 10, 2006
Subject:
Oregon State University Response to Request for Additional Information Regarding License Renewal, Oregon State University TRIGA Reactor (TAC No. MC5155) dated October 3, 2006 Mr. Hughes: In a letter dated October 3, 2006, the U. S. Nuclear Regulaitory Commission (NRC)requested that Oregon State University (OSU) provide additional information with regards to the OSU license application.
The OSU license application was made pursuant to 10 CFR 50.54 in a letter to the NRC dated October 5, 2004 and was further supplemented in a letter dated August 8, 2005. Enclosed is the OSU response to the request for additional information.
If you have any questions regarding the enclosed responses, please do not hesitate to contact me. I declare under penalty of perjury that the foregoing is true and correct.Executed on: II/CC.Sincerely, Steve Reese Director
Enclosure:
As stated cc: Document Control, NRC Al Adams, NRC Craig Bassett, NRC Jessie Quichocho, NRC John Cassady, OSU Rich Holdren, OSU Todd Palmer, OSU Mike Hartman, OSU Oregon State University
-Response to Request for Additional Information, Rev. -1. In Section 3.2, Meteorological Damage, the licensee states, "The superstructure of the OSTR has been designed for area wind, rain, snow, and ice loads. " Clarify if the roof structure is also designed to withstand any adverse weather condition (e.g., heavy rain or wind) in the area OSU Response:
The superstructure includes the walls and roof which form the confinement structure, so the roof is designed to withstand adverse weather conditions common to the area. The only equipment located on the roof is the secondary cooling tower and HVAC stacks. There is no equipment located on the roof that is used to implement safe shutdown of the plant.2. How is the leak tightness of the aluminum can (Ref SAR Sect 4.2.3) assured? If coolant infiltration were to occur, what assurances are there that it will not interfere with safe operation or prevent a safe shutdown?OSU Response:
The aluminum can surrounding the reflector at the OSTR is known to leak. This leakage was reported to the NRC in a letter from B.Dodd dated 31 August 1987. Flooding of the can does not pose a problem with respect to safe operation of the reactor, nor does it preclude a safe shutdown of the reactor. Based upon all available indications, the reflector can remains in a flooded condition.
The cause of the flooding is believed to be a failed weld. The water in the reactor tank has a very low conductivity, and there have been no indications of accelerated corrosion of the reflector can as a result of the flooding.
In addition, the reflector can and the remaining reactor components in the reactor tank are visually inspected as part of the daily startup and shutdown checklists.
Any gross dimensional distortion of the reflector can would be noted during these inspections.
OSU plans to perform a more thorough inspection of the reflector can during the conversion of the facility from high-enriched uranium to low-enriched uranium. The conversion is expected to happen late in 2008.3. Section 4.5.2.1 is the Core Description.
NRC SRP, Chapter 4 calls for information on excess reactivity and shutdown margin. In this section several different cores were described.
What would the excess reactivity and shutdown margins be for the different core configurations?
For excess reactivity and shutdown margin determinations, are the Position A inserts mentioned in Section 13.2.2.2.2 considered as part of the core or as part of an experiment?
OSU Response:
As a point of clarification, the only insert that is put into grid position A-I (the central thimble) is an aluminum slug, which is present in all core configurations.
The inserts mentioned in Section 13.2.2.2.2 go into grid position B-i and consist of:-A FLIP fuel element for the normal core
" A cadmium-lined irradiation tube for the CLICIT core" An irradiation tube for the ICIT core When installed, these various inserts are considered part of the core.Furthermore, the OSTR is fully fueled with FLIP fuel and does not operate with any of the mixed core configurations (Cores #3 -#6 in Table 4-6). The mixed cores were included to provide a historical record of cores that were considered in the conversion to the FLIP fuel and do not represent cores that were ever used operationally at the OSTR.The excess reactivity and shutdown margins associated with the various core configurations from the most recent rod calibrations are given below: Excess Shutdown Reactivity Margin*Core Configuration
[$] [$]Normal 5.79 8.03 ICIT 5.57 7.60 CLICIT 3.59 6.80 Determined with the highest worth non-secured experiment in its most reactive position, the most reactive control rod fully withdrawn, and the reactor in the cold condition with no xenon.4. Some of the items requested in NUREG-153 7, Part 2, Chapter 4 have not been included in the SAR." How does the excess reactivity and shutdown margin change with U andpoison burnup and Pu buildup?OSU Response:
The FLIP fuel is designed to provide long core lifetimes, with the lifetime of the current OSTR core projected to be > 8 MW-yr. It is a characteristic of FLIP cores that early in core life (up to 5 MW/yr) excess reactivity increases as the erbium poison is burned at a rate faster than the 2 3 5 U fuel. Beyond 5 MW-yr, core reactivity decreases as the 2 3 5 U is further consumed and fissile plutonium builds up. The net result of these effects is a core excess reactivity at 8 MW-yr that is approximately the same as the beginning-of-life core excess reactivity." What are the neutronflux densities?
OSU Response:
The peak thermal flux in the OSTR core during steady state operation at 1 MW is approximately Ixl10 3 neutrons/cm 2/sec.
- What is the fuel burnup between reloads or shutdowns?
OSU Response:
Given a FLIP fuel core lifetime of> 8 MW-yr and an average annual operation of the OSTR of approximately 0.125 MW-yr, the current core has a projected lifetime of> 64 yr, hence reloading of the core is not anticipated.
A FLIP fueled core with 8 MW-yr of operation will have a core averaged 2 3 5 U concentration that is -67% of its beginning-of-life value* How were the values for the neutron lifetime and effective delayed neutron ftaction determined, and what are their estimated uncertainties?
OSU Response:
For the OSTR FLIP fuel, the values for the prompt neutron lifetime and the effective delayed neutron fraction, are 20 jisec and 0.0070, respectively.
These values are consistent with those published by General Atomics for the FLIP fuel design (Simnad et. al, Nuclear Technology 28 (1976) 31-56).Simnad et. al calculated the prompt neutron lifetime by analyzing the change in core reactivity associated with homogeneously dispersing a 1/v absorber throughout the core. Peff was calculated by performing two reactivity calculations, one without delayed neutrons and the second with both prompt and delayed neutrons.
No estimated uncertainty is given for either the prompt neutron lifetime or [peff." Have the reactor periods been analyzed?OSU Response:
The OSTR is designed to be pulsed with reactivity insertion up to $2.55. A pulse of $2.55 results in a peak fuel temperature of-400 TC which is well below the 1150 TC temperature limit imposed on the TRIGA FLIP fuel. Further analysis of the reactor periods is not deemed necessary.
- What is the void coefficient (note: according to NUREG/CR-2387 typical void coefficient for the interstitial water in the core is about -0.2% ACkVkl% void)?OSU Response:
The measured void coefficient for the OSTR varies from--$0.13 / % void to -$0.51 / % void.* What is the xenon and samarium override?OSU Response:
The OSTR does not normally operate for periods of time exceeding 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Therefore, the xenon concentrations throughout the core do not approach their saturated values.
An estimate of the xenon and samarium override for the OSTR, assuming saturated concentrations for both poison species, is $8.* What is the overall power coefficient of reactivity if not accounted for in items listed above?OSU Response:
The reactivity coefficients associated with the OSTR are: Reactor tank water temperature coefficient
--0.05 ¢ / TC Void coefficient
--13 to -51 ¢ / % void Fuel temperature coefficient
-1 ¢ / °C (Note: the fuel temperature coefficient for FLIP fuel is a linear function of fuel temperature.
The value of-i ¢ / TC is the nominal value for the fuel temperature at 1MW.)The overall power coefficient for the OSTR is dominated by the fuel temperature coefficient and has a value of- -0.25 ¢ / kW." Are there any credible situations where aflow instability could occur in a fuel channel?OSU Response:
No known credible situations exist where a flow instability could occur in a fuel channel of a pool-type TRIGA cooled by natural convection.
- 5. Section 4.6 provides the results of the calculations of the maximum heat flux. For which of the cores described in Section 4.5.2.1, and mentioned in Section 13.2.2.2.2, were the calculations performed and will there be significant differences for the other cores?The calculations were performed for the normal core configuration (i.e. a core similar to Core #8 in Table 4-6 with a fuel element in grid position B-i). The conclusions of this analysis are applicable to other core configurations as well since peak fuel element powers do not change appreciably amongst the other core configurations.
- 6. Confusion exists as to the specific power measuring channels that comprise the "power level measuring channel" and/or "power level monitor" referred to in TS 3.2.2. These channels are not defined in the TS or adequately described in the SAR. From the information provided in Sections 7.2.3.1 and 7.4.1, it appears that the Safety and/or the percent power channels could form the "power level measuring channel" referred to in the TS. However, this is not entirely clear, as Section 7.2.3.1 also refers to a "power range monitor" which is apparently composed of the percent power and pulsing channels.
In addition, it is not clear if one or two channels are required to satisfy the power level monitoring capability specified in the TS.OSU Response:
The safety channel and the percent power channel are each separate and individual power level channels.
The displays for each of these channels are separate and unique, as are the detectors the signals originate from. Only one is required by the technical specifications.
Each is distinct from the other. However, as stated in the last paragraph of section 7.2.3.1, the percent power and pulsing channel combined form the power range monitor.7. Reactor power measuring channels are described in Section 7.2.3.1 of the SAR.TS 3.2.2, "Basis, "states. "The power level monitors assure that the reactor power level is adequately monitored for both (sic) steady state, square wave, and pulse modes of operation." TS Table 1, "Minimum Measuring Channels," identifies the "Nvt circuit" as being required while operating in the pulse mode.However, the "Nvt circuit" is not described in Chapter 7 of the SAR. Please clarify.OSU Response:
The pulsing channel is the Nvt circuit.8. SAR Chapter 5 states that cooling of the reactor core is accomplished by natural circulation water flow through the core area combined with a forced-flow circulation ofpool water through a tube-and-shell type heat exchanger.
Chapters 3 and 5 of the SAR indicate that a loss offorced coolantflow will not result in any adverse consequences.
For example, Section 3.1 of the SAR sates: "Natural convection cooling is sufficient to dissipate core heat. " This statement should be clarified as it implies that a loss offorced circulation would have no impact on safe reactor operations which is not consistent with the evaluation presented in SAR Chapter 13 (Section 13.2.4.2.1).
OSU Response:
To be clear, the reactor core cooling is provided exclusively by natural convection.
No forced flow is required.
The circulation of primary water through the heat exchanger removes heat energy from the reactor tank.It is possible to run the reactor without the primary water system functioning.
As an example, in order to perform a reactor power calibration the primary water system must be off. The evaluation presented in Section 13.2.4.2.1 relies on a malfunction on each of at least three independent monitoring channels, making the scenario incredible.
- 9. Section 13.2.4.2.1 of the SAR states: "The reactor, however, would shutdown as the water level dropped past the top of the fuel. " There does not appear to be an automatic trip (scram) in response to low pool water level. Therefore, this statement should be clarified to describe the specific mechanisms that would cause the reactor to shutdown (e.g., temperature coefficient of the fuel and/or scram in response to high fuel temperature).
OSU Response:
The use of the term "shut down" should be replaced with"become subcritical due to a lack of moderator" in this context. As stated in that section, water level, water flow, and water temperature alarms would alert the operator and allow them to shut down the reactor long before the conditions described in section 13.2.4.2.1 would occur.10. As described in Section 7.3 of NUREG 153 7 Part 2, the RCS should be designed for reliable operation in the normal range of environmental conditions anticipated within the facility.
Please describe the type of electrical isolation
/physical protection provided for voltage and signal cables associated with the redundant power monitoring channels, as this is not currently in Section 7 of the OSUSAR. For example, are they routed in separate conduits from the reactor to the control room?OSU Response:
To the extent practical, the voltage and signal cables are routed separately.
Additionally, cables are routed in rigid conduits to protect them from environmental conditions.
- 11. Section 7.2.3.1 states that a loss of operating voltage to the percent power channel or pulsing channel (which form the "power range monitor ") will initiate a reactor scram in response to a "non-operable conditions. " Will a loss of operating voltage to the Safety and Percent Power Channels also initiate a "non-operable condition" reactor scram?OSU Response:
The loss of high voltage for the safety channel, percent power channel and the fission chamber, which is the source of signal for the period and wide-range log channels, will result in a reactor scram.12. Section 9.1.2 the ventilation system. The Argon manifold exhaust is part of the reactor bay ventilation system, as shown in Figure 9.1. Section 11.1.5.2 states that the reactor bay ventilation system has HEPA filters on all ducts originating from irradiation or sample handling facilities.
Please clarify the purpose of the HEPA filters and explain why there are no surveillance requirements in TS 4.5for filter testing on the two Argon manifold HEPA filters.OSU Response:
The HEPA filters are included as a good engineering practice and are in the spirit of maintaining doses ALARA, however they are not required.
TS 4.5 is acceptable without testing of the HEPA filters because the ventilation system will function as intended irrespective of the condition of the HEPA filters.13. Section 9.1.2 describes the ventilation system. Section 11.1.1.1.2 notes that the ventilation system keeps occupational building exposures from Ar-41 well below the 10 CFR 20 DAC limits. The basis of TS 3.5for the ventilation system discusses exposure to both the public and occupational dose to workers. It is not clear if the use of the word "ground" in the first sentence of TS 3.5 basis refers to outside the Radiation Center building or inside the building in unrestricted area.Please clarify the purpose and need for the ventilation with respect to both workers and the public for normal operations and for accidents scenarios.
OSU Response:
The word "ground" refers to the unrestricted areas outside of the Reactor and Radiation Center buildings.
The purpose of the ventilation system is to provide complete monitoring of all air leaving the Reactor Building.
To clarify this point, a couple of changes to the SAR should be made. First, the second sentence in the basis for TS 3.5 should be changed to say, "In addition, the worst-case maximum total effective dose equivalent is well below the annual limit for individual members of the public." Second, the second sentence in the basis of TS 5.1 should say, "The facility is designed such that the ventilation system will normally maintain a negative pressure in the Reactor building with respect to the outside atmosphere so that air leaving the Reactor Building is monitored.
Third, the first bullet in Section 11.1.5.2 should be deleted (it is misleading).
Fourth, the first two sentences of Section 9.1 should be replaced to say, "The controlled ventilation system provides a method of monitoring air leaving reactor building.
The objective of the confinement structure is to ensure that provisions are made to reduce the amount of unmonitored radioactivity released into the environment by maintaining a negative pressure within the reactor building during operation." The calculated and measured Ar-41 concentrations for normal operations are given in Section 11.1.1.1.2.
While there are no measured values for the Ar-41 concentration in the Reactor Building while at full power with the ventilation system off (this would be TS violation), we know that the measured concentrations very near the grate on the reactor top were below the DAC for Ar-41 with the ventilation system on. If the ventilation system is off while at full power and equilibrium, it is reasonable to assume that the concentration on the Reactor Building floor would not be greater than this because the dilution that takes place going from the small (effective) volume on the reactor top to the large volume available on the Reactor Building floor.Automatic shutdown of the ventilation system confines the free air volume of the reactor building during emergency conditions.
However, as shown in Chapter 13, both occupational and general public doses are below the applicable annual limits found in 10 CFR 20 whether or not the ventilation system is running during the emergency.
- 14. The SAR describes a specialized chamber inside Beamport #4 to produce Ar-41.Please provide information on what individual components of the argon production facility equipment are physically located inside Beamport #4.Additionally, please provide the operating pressure that the transfer lines or equipment located inside the beamport will be exposed to and state whether periodic inspections are performed on equipment inside the beam port for material degradation.
OSU Response:
The chamber inside beam port #4 is a 2 liter, stainless steel container.
The container is connected via 1/8-inch lines to the transfer system on the exterior of the biological shield. The 1/8-inch lines allow the transfer of the argon past both shield plugs which are also located within the beam port. Typical operating pressure for the system is 40 psia. Periodic inspections of the system are not performed because doing so would cause unneeded personnel dose. However, the entire system must be evacuated prior to every use such that any loss of system-wide integrity would be immediately observed.15. ANS 15.1, item 6. 5 also would require ROC review of substantive changes to approved experiments.
Please explain why this requirement is not part of the OSTR Technical specifications.
OSU Response:
ROC review of substantive changes to approved experiments is completely appropriate.
This requirement should be added to the technical specifications.
- 16. TS 4.8 is basically a restatement of the limiting conditions of operation (TS 3.8.2).TS 4.8 does not mention surveillance or inspection requirements for any experimental apparatus.
TS 4.8 should contain requirements for inspecting the appropriate experiments periodically for material degradation to prevent failures, if the experiment is installed in the reactor beyond some pre-determined time frame, as per the guidance in Regulatory Guide 2.2, paragraph C. 2.f Radiation Sensitive Materials.
OSU Response:
Paragraph C.2.f or Regulatory Guide 2.2 calls for an assessment of the consequences of physical or chemical changes in the material content of an experiment as a result of its presence in a radiation field. This evaluation is performed and is required by our procedures.
Additionally, technical specification 3.8.3 assumes complete failure of an experiment and requires limits on the total radioactivity produced such that the applicable dose limits in 10 CFR 20 are not exceeded.17. The NRC SRP, App. 14.1, item 6.3, Radiation Safety states that the TSs should state management's commitment to practice an effective ALARA program. This is currently not in TS 6. 3. Please address.OSU Response:
Technical specification 6.3 states that the guidelines of ANSI/ANS 15.11, "Radiation Protection at Research Reactor Facilities" should be followed.
The entire point of ANSI/ANS 15.11 is to maintain doses ALARA. Additionally, ALARA is already a regulatory requirement (e.g., 10 CFR 20.1101 (b)) and a technical specification addressing something that is already required by regulation is redundant.
Technical Specification 6.3 should be changed to state that "The program should use the guidelines of ANSI/ANS 15.11-1993, ..." The individual responsible for implementing the Radiation Safety Program at the OSTR is the Senior Health Physicist (Level 2), as indicated in Technical Specification 6.1.2.c.18. The NRC SRP and ANS 15.1 specify that the purpose of the review committee is to provide independent oversight.
The SRP also states that it is desirable to have members on the committee who are not employed by the reactor owner, and that the operating staff should not constitute the majority of a quorum. The ROC as descried in the SAR has all of the reactor line management on the ROC, including Level 1, 2, & 3. This does not appear to provide the desired independence.
It is unusual to have the Level ] manager on the committee, since that is the one appointing the committee and the one to whom the committee typically reports its independent oversight result. Also, neither the SAR nor the TS address the quorum requirements or the membership ofpersons not employed by OSU It would appear from the membership that operations line management would usually be majority of the committee.
Please address.OSU Response:
The charter of the ROC should be amended to remove voting privileges from the Director (level 1) and the Reactor Supervisor (level 3).Additionally, a quorum requirement of five (5) voting members should be put in place. This would guarantee that the operations management could never have a majority vote on the committee.
However, we feel it is very important that the Reactor Administrator (level 2) and Senior Health Physicist (level 2)maintain voting privileges.
In addition, changes to the ROC charter should be made to ensure that there is representation by members employed outside of the Radiation Center. The qualifications for the individuals filling the roles described below: " At least one person whose field of expertise is another branch of engineering
- At least one person whose field of expertise is radiation chemistry, nuclear chemistry, health physics, or radiation biology should be expanded to include employment in an OSU academic unit outside of the Radiation Center or employment by an outside agency.19. Section 12.2.24 of the SAR discusses the audit function at the OSTR and includes audits of reactor operating areas and record, and reportable occurrences.
TS 6. 2 requires the performance of an annual audit. However, a number of areas delineated in the SAR and ANS 15. 1 for audit are not specified in the TS or the SAR, namely: conformance with TSs, and actions taken to correct deficiencies, emergency plan and implementing procedures, security plan, facility procedures, experiments, surveillances, and the training program. The annual audit of the health physics is listed in Chap. 11 of the SAR but should also be in the TSs. Also, the SAR should specify that no individual responsible for an area may conduct the audit. Please address.OSU Response:
We would prefer not to have a technical specification requirement to audit the emergency plan or the physical security plan due to restrictions on access to safeguards information created by the Energy Policy Act of 2005. The other items listed (i.e., procedures, experiments, surveillances and the training program) are considered operating records. The requirement for a review of the radiation protection program already exists in 10 CFR 20.1101 (c) such that the same requirement in the technical specifications would be redundant.
Finally, we would prefer to leave specifics on who and how audits are performed in the ROC charter. However, OSTROP Section IV.C. 1 .d should be revised to state "The person performing the audit will not be the person normally performing the function stated." 20. Section 12.3 of the SAR discusses procedures.
The process in TS 6.4 for minor changes to reactor safety procedures is reasonable, but the separate process for minor ("unsubstantive" is used in the TS) changes to radiation protection procedures is not. Guidance in ANS 15.1 is for such changes to be made by Level 3 management with approval by Level 2 in 14 days. TS 6.4 allow these changes with no prior approval and final approval by the Senior Health Physicist (Level 2) in 120 days.OSU Response:
Section 6.4 of the technical specifications should be changed to say, "Unsubstantive changes to radiation protection procedures shall be approved prior to implementation by the Senior Health Physicist and documented by the Senior Health Physicist within 120 days of implementation." 21. The unusual need to deviate from procedures to deal with special circumstances is addressed in ANS 15.1, which states that such actions should receive approval of the SRO and be subsequently documented and reported to Level 2 management.
SAR Section 12.3 addresses this need but does not mention any controls such as those stated in ANS 15.1. Please address.OSU Response:
Section 6.4 of the technical specification should include the following sentence, "Temporary deviations for the procedures may be made by the responsible senior reactor operator in order to deal with special or unusual circumstances or conditions.
Such deviations shall be documented and reported to the Reactor Administrator." 22. TS 6. 5.1. c gives the required content for the report to the NRC in the event of a Safety Limit violation.
This area is not discussed in Chapter 12 of the SAR. The TS agrees with the recommendations of ANS 15.1 with the exception that it does not specify that the report contain the effects (if any) of the violation on the health and safety ofpersonnel and the public. Also in TS 6. 6. 2. a. ] the word"accidental" should be deleted And TS 6. 6 should note that duplicate telephone and written reports should also be made to the NRC Regional Office. Please Address.OSU Response:
Section 6.5.1 .c of the technical specifications should be changed to read, "...of the causes, effects upon structures, systems or components and on the health and safety of personnel and the public,...".
We agree that the word "accidental" should be removed from section 6.6.2.a. 1.We strongly feel that only one line of communication should be maintained with the NRC and that should be with the Research and Test Reactor group.23. SAR Section 13.2.3 discusses a loss of coolant accident.
What are the expected integrated doses that could be received by the staff and the general public from such an event?OSU Response:
It should be noted that an instantaneous complete loss of water is an incredible scenario.
It is also unrealistic to assume that an individual on the reactor top would not realize that a LOCA was in progress and leave the reactor top prior to a complete loss of water. That being said, it is not unreasonable to expect that an individual should exit the reactor top within 10 seconds. Based on this 10 second stay time and the data presented in Table 13-14, a time after shutdown of 10 seconds would result in an integrated dose of 27.8 rem. All other times after shutdown would result in doses below the occupational limit.It is reasonable to expect that it will take no more than an hour to perform any necessary initial corrective actions in the reactor bay and at the site boundary for a LOCA. Based on a one hour stay time, all integrated doses from the data presented in Tables 13-15 and 13-16 would be below the applicable limits given in 10 CFR 20.24. The licensee did not specifically discuss prior use of reactor components for the OSTR in the SAR. SAR Chapter 16.1 discusses prior use of components and that fact that degradation mechanisms should be addressed.
This is clearly applicable to a license renewal situation.
Relevant information on prior component use and how deterioration of existing components is judged not to be a problem during the license renewal period. Please discuss reactor components that are continuing to be used for the OSTR that perform a safety function and address the concerns identified in SRP Section 16.1.OSU Response:
The lack of information presented for Chapter 16 was an oversight on our part. The following material should be included in Chapter 16:
16.1 Prior Use of Reactor Components The primary method to detect deterioration of system or components important to safety is through a surveillance and maintenance program. Each structure, system or component may have its own unique failure mechanism (i.e., primary tank corrosion, gear wear for control rod drives, etc). The OSTR has an extensive program for surveillance and maintenance on monthly, quarterly, semi-annual, and annual cycles. The individuals responsible for this program are also aware that each form of degradation may have a different point in time in which the degradation may interfere with the function of the structure, system or component.
Examples of monthly surveillance and maintenance checks include, but not limited to, primary pump bearings, primary tank water level alarms, primary tank water temperature alarms. Examples of quarterly surveillance and maintenance checks include emergency response plan equipment inventories, stack airborne monitor tape speed and alarm functional checks, and area radiation monitor functional checks. Examples of semi-annual surveillance and maintenance checks include functional checks of console interlocks, control rod withdrawal, insertion, and scram times, ventilation shutdown functional test, fuel element temperature channel calibrations, cleaning and lubrication of the transient rod carrier barrel, lubrication of the ball-nut drive on the transient rod carrier, console voltage checks, standard control rod motor checks, ion chamber resistance measurements, and fission chamber resistance measurements.
Annual surveillance and maintenance includes such things as control rod calibrations, reactor power calibration, primary tank water temperature calibrations, continuous and stack air monitor calibrations, area radiation monitor calibrations, control rod drive inspections, reactor tank and core component inspections, fuel element inspections for selected elements, and primary tank water low level alarm functional check.The frequency of these items is determined by identifying the appropriate amount of time between checks to observe a potentially degraded condition.
As systems are upgraded it is sometimes appropriate that the frequency of a surveillance and maintenance item may be relaxed from, as an example, monthly to quarterly.
Conversely, as the age of some components increases, it is sometimes necessary to increase the frequency of the surveillance and maintenance item. The frequency of a long established surveillance and maintenance item is ultimately driven by experience with the component in question.
For new structures, systems, and components, the frequency is usually that recommended by the manufacturer.
16.2 Medical Use of Non-Power Reactors At this time, there are no current or planned medical uses of the OSTR.