ML070880060
| ML070880060 | |
| Person / Time | |
|---|---|
| Site: | University of Maryland |
| Issue date: | 03/30/2007 |
| From: | Alexander Adams NRC/NRR/ADRA/DPR/PRTA |
| To: | Dan Collins NRC/NRR/ADRA/DPR/PRTA |
| Adams A, NRC/N RR/DRIP/REXB, 415-1127 | |
| References | |
| TAC MB1788 | |
| Download: ML070880060 (18) | |
Text
March 30, 2007MEMORANDUM TO:Daniel S. Collins, Branch ChiefResearch and Test Reactors Branch ADivision of Policy and RulemakingOffice of Nuclear Reactor RegulationFROM:Alexander Adams, Jr., Senior Project Manager/RA/Research and Test Reactors Branch ADivision of Policy and RulemakingOffice of Nuclear Reactor Regulation
SUBJECT:
UNIVERSITY OF MARYLAND - DRAFT REQUEST FOR ADDITIONALINFORMATION (RAI), REGARDING LICENSE RENEWAL FOR THEUNIVERSITY OF MARYLAND TRAINING REACTOR (TAC MB1788)The attached draft RAI was transmitted on March 1, 2007, to Dr. Mohamad Al-Sheikhly of theUniversity of Maryland in preparation for an upcoming conference call. Review of the RAI wouldallow the licensee to identify areas where clarification may be needed, as well as determine andagree upon a schedule for responding to the RAI. This memorandum and its attachment do notconvey a formal request for information or represent a Nuclear Regulatory Commission position.Docket No. 50-166
Attachment:
As stated CONTACT:A. Adams, NRR301-415-1127 March 30, 2007MEMORANDUM TO:Daniel S. Collins, Branch ChiefResearch and Test Reactors Branch ADivision of Policy and RulemakingOffice of Nuclear Reactor RegulationFROM:Alexander Adams, Jr., Senior Project Manager/RA/Research and Test Reactors Branch ADivision of Policy and RulemakingOffice of Nuclear Reactor Regulation
SUBJECT:
UNIVERSITY OF MARYLAND - DRAFT REQUEST FOR ADDITIONALINFORMATION (RAI), REGARDING LICENSE RENEWAL FOR THEUNIVERSITY OF MARYLAND TRAINING REACTOR (TAC MB1788)The attached draft RAI was transmitted on March 1, 2007, to Dr. Mohamad Al-Sheikhly of theUniversity of Maryland in preparation for an upcoming conference call. Review of the RAI wouldallow the licensee to identify areas where clarification may be needed, as well as determine andagree upon a schedule for responding to the RAI. This memorandum and its attachment do notconvey a formal request for information or represent a Nuclear Regulatory Commission position.Docket No. 50-166
Attachment:
As stated CONTACT:A. Adams, NRR301-415-1127DISTRIBUTION:PublicD. CollinsRTR R/FA. Adams Accession Number:ML070880060OFFICEPRTA:PMPRTA:BCPRTA:PMNAMEAAdams:cah DCollins AAdamsDATE03/30/0703/30/0703/30/07 OFFICIAL RECORD COPY Draft RAIRAI #14.
Section 4.3, Reactor Tank. This section describes the design and function of the reactorpool tank. How would you detect leakage from the reactor tank? What is the minimumleakage rate that you can detect and what is the maximum time duration that this rate of leakage could occur before detection? What is the impact on the public health and safety of a pool leak?Univ. of Maryland Response:Leakage or excessive water consumption is determined by monitoring the pool waterlevel and recording the amount of make up water that must be added to keep the biological shield level maintained at or above the required levels. This is accomplished by means of an ultrasonic level detector mounted on the bridge and logging the amount of water added and the total kWh since the last addition. Leakage occurring in the water handling area would be noticed during the initial startup checklist (OP101) where the operator is instructed to check the sump level. An exact minimum detectable leakage rate is unknown. The maximum time that this rate could continue would be not more than ten days, the maximum time that the facility is closed for the winter holidaybreak.BNL Review:The response to this RAI does not fully address the question which asked what theimpact on public health and safety would be. How many possible leakage paths exist?
Are these leakage paths alarmed? Does all of the leakage go to the sump? Is thesump alarmed? Additionally, the response stated that the max time a pool leak could go undetected would be 10 days. By not being able to know the minimum detectable leak gives rise to evaporative loss masking a leak. Does the licensee estimate the loss due to evaporative loss, or trend pool water makeup to provide an indication of unexpected loss?RAI #18.
Section 4.5.3, Operating Limits. Identify how the value 0.90$ was obtained when ahypothetical 3.50$ excess reactivity is present and the two control rods are introduced.
Is there a common mode failure for multiple experiments that could introduce additional excess reactivity?Univ. of Maryland Response:The MUTR in its current and only, configuration uses three independent control rods tocontrol the reactor. The total reactivity of the control rods is nominally $8.00 with the reactor requiring the nominal removal of $6.60 to bring the reactor to low power critical with the pool water and fuel at room temperature. This results in an excess reactivity of
$1.40 at room temperature with no experiments. The nominal reactivity worths of the control rods are Regulating Rod: $2.30, Shim I: $2.70, and Shim II: $3.00.From the above it is obvious that without experiments the insertion of any two of thethree control rods will bring the reactor subcritical. With Shim II fully withdrawn and theother two rods fully inserted the reactor will be subcritical by more than $3.00. Technical Specifications limits the excess reactivity of the MUTR to $3.50 with reference to the cold critical condition. With this limitation and Shim II fully withdrawn, the reactor would still be subcritical by $0.90. This is almost twice the required shutdown margin of
$0.50. To ensure that the reactor is capable of achieving a safe shutdown, annual calibrations of the control rods followed by computations of excess and shutdownreactivities will be performed. Given the typical MUTR operation history, this timeperiod is sufficient to track any changes that may be caused by fuel burnup.BNL Review:
With excess reactivity of $3.5 present and Shim II is withdrawn, Shim I ($2.7) and the Regulating Rod ($2.3) would allow a total of $5.0 reactivity insertion with an excess of
$1.5. How is the subcriticality value of $0.9 obtained?RAI# 19.
Section 5.2, Primary Coolant System, page 5-1. The MUTR is designed for naturalconvection cooling without forced flow. There is also a heat exchanger (HX) in a forced flow loop present in the primary coolant system, but it is not required for safe operation of the reactor and can be bypassed. The function of this system is to maintain the temperature and chemistry quality of the pool water. The HX is cooled by an open loop of city water which discharges into the city sewer system. The use and safety design of this system is not clear. Please discuss the following:a.When the reactor is operating, what is the normal mode of operation for theprimary coolant system relative to the cleanup system and the HX? Is the operation of this system controlled by a plant operating procedure?Univ. of Maryland Response:When the reactor is operating, the typical alignment of the system is with the primarycoolant pump off and the secondary coolant system is isolated after the heat exchanger.
This alignment provides for the typically 100-psi minimum secondary supply to surpass the primary pressure. This results in a situation where any potential excursion would migrate from the city water supply into the primary system. BNL Review: City water supply appears to maintain the 100-psi minimum pressure in the secondarycoolant system. Is 100 psi the normal city water pressure or is it boosted for the MUTR? What are the normal operating pressures of the primary coolant within the HX for both primary pump running and not running conditions?Are there any operating procedure(s) that control the operation of the primary coolantsystem?Clarify if the cleanup system operates continuously along with the primary coolantsystem operation while in either natural or forced operating mode. b. How often is the HX used and on what conditions would the HX be used andbypassed?Univ. of Maryland Response:The heat exchanger is used in a very limited set of circumstances, the first being todemonstrate the effects of moderator temperature feedback during instructional use of the reactor. This lesson is taught approximately six times per year. The second situation would be in the case of a high-power operation that would result in the primary water reaching a temperature in excess of 40°C. This limit is set to avoid damage to the ion exchange resin that is used to maintain water conductivity within acceptable limits. BNL Review:
On what conditions would the HX be bypassed?c.If there were a reactor coolant piping/component failure outside of the reactorpool, describe what would prevent the pool water from draining or limit the amount of water lost?Univ. of Maryland Response:By design and in practice, any protrusion that exceeds twenty inches below the waterline must be equipped with a siphon break. Therefore the maximum water loss that could possibly occur would be limited to twenty inches.
BNL Review: Response acceptable.d. If a primary to secondary leak were to develop in the HX, which way would theleakage flow, and how would your design prevent the escape of primary water into the city water system? Please consider this question with the primary pump both running and shut down. If you cannot show that pressure on the secondary side is higher than pressure on the primary side of the heat exchanger at all times, analyze the impact of a HX leak from a radiological standpoint. Is there any radiological monitoring of the discharge of the city water before it goes into the city sewer system?Univ. of Maryland Response:By design the secondary water pressure is kept higher than the primary system. This isthe situation regardless of whether the primary coolant pump is operating. There is no monitoring of the secondary discharge for the previously stated reason.
BNL Review: Response acceptable.RAI #20.
Section 5.6, Nitrogen-16 Control System, page 5-4. In section 5.6 discussing the N-16control system, it is stated that the outlet pipe is equipped with a siphon break to preclude a significant loss of primary coolant in the event of a piping failure outside of the pool tank. Figure 5.1 does not show the location of the siphon break. Indicate where the siphon break is located in this figure. Also, explain how this siphon break precludes a significant loss of pool water in the event of a piping failure outside the pool tank.Univ. of Maryland Response:By design and in practice, any protrusion that exceeds twenty inches below the waterline must be equipped with a siphon break. Therefore the maximum water loss that could possibly occur would be limited to twenty inches. The siphon break is illustratedin Fig. 5.1 on the discharge side of the pump at the level of the top of the tank. The siphon break works on the principal that in the event of a piping failure, the break would allow air to enter the system thus preventing the siphoning of the biological shield. BNL Review:
Please identify the location of the siphon break in Figure 5.1 on the discharge side of the primary pump. Are there any other "protrusions" that exceed 20" below the water line (including the suction side of the pump)?RAI# 21.
Section 5.7, Reactor Sump, page 5-7. Figure 5.4 presents the reactor sump waterhandling system. In this figure the well and the sump structure are shown as separate structures. Describe the physical connection between the well and the concrete sump pit. Does the spring check valve on the city water line in Figure 5.4 provide assurance that no backflow into the city water system occurs? If so, what testing or inspection is performed on this valve to ensure no degradation and that it is operating as designed?Univ. of Maryland Response:The well is shown twice, once to illustrate the well position in the sump and then againto illustrate the pump pick up system. The backflow prevention preventing check valve does indeed prevent migration of the primary water into the city system. The valve immediately preceding the check valve is closed when the pump valves are aligned to allow sump water to enter the pump. The check valve is a redundant system and is not inspected or tested.BNL Review:
Check valves in nuclear plants frequently develop leaks. Without being inspected or tested, how would one assure that the check valve and the preceding ball valve are performing their safety functions? RAI # 32.
Section 9.2.3, Fuel-Rod Transfer Cask. The SAR states that the fuel rod transfer caskweighs approximately 5700 lbs. Provide a discussion on the procedures, equipment, and lifting capacities associated with this load handling at MUTR.Univ. of Maryland Response:See Fig 3.2 of the SAR regarding the crane position. This crane is certified to 6000pounds. The procedures w ould be considered a "Special Experiment" and will bewritten in conjunction with the Radiation Safety Officer and the Reactor Safety Committee.BNL Review:The licensee responded that the fuel transfer cask weighs 5700 lbs and the crane has acapacity of 6000 lbs. The intent of the question was to get an understanding of the precautions, inspections, procedures used to insure the crane can handle this heavyload. Their response is that procedures would be considered a 'special experiment' andwould be written in conjunction with the radiation safety officer and reactor safety committee. The response does not provide the information required to answer the RAI.
From this answer it can not be discerned whether no such lift has ever occurred and therefore no such procedures exist, or if the licensee performs these lifts with no formal procedures and precautions.RAI #38.Section 10.2, Experimental Facilities. Provide a more detailed description of thefunctional design of the Thermal Column, Beam Ports and Through Tube experimental facilities. For example, from the description provided it is not clear if these facilitiesrequire an air exhaust system or if beam tubes must be filled with demineralized waterto provide shielding. Additionally, Figures 10.3 "Beam Tube," 10.4, "Beam Tube Plugs,"
and 10.5, "Through Tube" are not legible (apparently due to their reduced size). Please provide clearer illustrations that are of a larger size. Enlarging each figure to 8 1/2 x 11 inches should be sufficient.
Univ. of Maryland Response:Univ. of Maryland provided 4 attached figures.BNL Review: Response Incomplete. The response only provided enlarged drawings and does notinclude any additional information (narrative description) needed to facilitate the reviewof the design of the Thermal Column, Beam Ports and Through Tube experimental facilities. RAI # 39.
Section 10.2.4, Pneumatic Transfer System. Provide a more detailed description of thepneumatic transfer system design and operation and the administrative controls governing its use. Specific topics to be addressed include the source(s) of CO2, potential consequences of a stuck / immovable rabbit assembly and design features and/or administrative controls provided to preclude or mitigate this occurrence, manual and automatic timing modes of operation, system venting, and design features which preclude the potential for a failure within this system to result in a loss of pool water inventory.Univ. of Maryland Response:The Pneumatic Transfer System is supplied with CO2 by means of a standard cylinderthat is secured to the wall in the hot room. The resulting circumstances that would occur from a "stuck" rabbit capsule are of an extremely low risk of causing either core damage or loss of pool water due to the fact that the rabbit experiments are limited to a maximum of $1.00 reactivity and the CO2 piping is routed to exit the source tank vertically on a path that routes to the top of the security cage. Flooding of the rabbit receiver could reach a level equal to the height of the tank water and not above. Timing of the insertion is controlled from the auxiliary console with the operator controlling theinsertion. Capsule ejection may be set to automatically occur at the end of a preset time with the operator having a manual return switch mounted in the same location as the insertion. The capsule is then returned to a receiver in a sealed glove box in the hot room, which is the same location as the sample started its path into the core. The glove box is equipped with a fan which vents the box into the bay area. The glove box is also equipped with a radiation monitor which provides a display at the operators console and the door which allows entry into the hot room. Procedures governing the operation of the system are contained in OP 105R12 Installation of Experiments. BNL Review:From this response, it appears that the glove box vents directly to an area wherepersonnel may be located ("The glove box is equipped with a fan which vents the box into the bay area." ). Provide a more detailed description of the design of the glove box ventilation system including a technical basis for the current design. In addition to a narrative description, the re sponse should include illustrations (drawings), which depictthe current configuration of the glove box ventilation system.RAI # 40.
Section 10.2.5, Other Locations. This section states that the reactor grid plate andreactor pool tank may be utilized to conduct experiments. Provide a more detaileddescription of the functional design of these facilities, the type of experiments that aretypically conducted at these locations, and if any special precautions or limitations areneeded to ensure their safe use.
Univ. of Maryland Response:The standard review format must be followed before conducting any experiment. Inaddition to the standard review, all experiments must be evaluated against the experimental maximum for reactivity worth.BNL Review:Response Incomplete - Response does not address requested information regarding amore detailed description of the functional design and the type of experiments that are typically conducted at the reactor grid plat e and reactor pool tank experimental facilities.RAI# 54 Section 13.2.1, Maximum Hypothetical Accident, page 13-2. What is the reference forthe isotopic loading in one fuel element of the MUTR after an infinite operation at 250 kW?BNL ReviewThe NUREG/CR2387 provided isotopic loading for a finite burnup of 365 Mwd. TheNUREG inventory was then converted for the MUTR. Why was the MUTR inventoryreferred to as for infinite operation?RAI# 55 Section 13.2.1, Maximum Hypothetical Accident, page 13-3. What is the basis forassuming a value of 0.01 for the atmospheric dispersion factor (x/Q)? What are the release pathways to the environment for the HMA? If the release point is elevated, has the possibility been examined that the highest dose may be from overhead cloud shineinstead of cloud immersion?BNL Review Response did not address the second part of the RAI that questioned the possibility of ahigher dose from overhead cloud shine instead of cloud immersion.RAI# 58 Section 13.2.1, Maximum Hypothetical Accident, page 13-4. The analysis only providesdose consequences for downwind locations (unrestricted areas). What is the projecteddose for facility staff in the reactor bay (restricted area)? Doses in the unrestricted areas should be given for the maximum exposed person, the nearest residence, andother locations of interest such as the nearest dormitory.BNL ReviewStill need to review LOCADOSE.xls. Please provide.RAI# 61Section 13.2.2.3, Insertion of Fuel, page 13-7. Table 13.7 gives the calculated peak fueltemperatures for a 3.70$ reactivity pulse, at initial powers of 0.01 kW and 250 kW respectively. What is the basis for choosing a pulse of 3.70$? What is the locationwhere a fuel cluster is added that results in a 3.70$ excess reactivity? Does this analysis form the technical basis for limiting the excess reactivity to 3.50$?BNL ReviewThe RAI pertains to the discussion in the last paragraph on page 13-6 of the SAR. Theinsertion of $3.70 of prompt reactivity refers to a situation where the reactor is already critical and an additional fuel element is inserted to the core. The response did not address this particular scenario.RAI# 62 Section 13.2.3, Loss of Coolant, page 13-7. The discussion on a loss of coolantaccident (LOCA) noted that audible signals in the main reactor room, or on the west balcony, would warn persons entering those areas of high radiation conditions. Whenthe building is unoccupied how would the high radiation condition be communicated to emergency response personnel? Is there an outside alarm to alert people to keep awayfrom the facility? What is the projected dose for a person standing outside the reactorbuilding? Please provide a copy of your calculations showing the dose rates from the LOCA. What are dose rates immediately following uncovering of the core? BNL ReviewStill need to review LOCADOSE.xls. Please provide.
RAI# 63 Section 13.2.4.1, Fission Product Inventory, page 13-8. How does the fission productinventory listed in Table 13.8 compare with the source terms assumed for the Maximum Hypothetical Accident? Please calculate the fission product inventory for your fuel element.BNL ReviewThe response provided a list of fission product inventory for the MUTR fuel element. However the response did not demonstrate the consistency of the listed inventory withthe values given in Table 13.8 of the SAR. The SAR listed bromine as one of the elements to be released but the response did not have isotopes of bromine listed as part of the inventory.RAI# 64Section 13.2.4.2, Contamination of the Pool Water with Radioactivity, page 13-8. Isthere a reference for the maximum water activity of 6.687x10
-4 mCi/ml in a fuel claddingfailure?BNL ReviewThe response did not demonstrate how the maximum water activity (from claddingfailure) of 6.687x10
-4 Ci/ml was calculated.RAI #72.
Section V of the 1999-2000 Annual Operating Report states that continuous monitoringfor the year was accomplished using fixed-mounted film badges throughout the interior of the reactor building. Facility Technical Specification 3.6.4 specifies that the campus radiation safety organization maintain an environmental monitor at the site boundary as well. Explain how compliance is demonstrated, and if any abnormal radiation levels were ever detected.Univ. of Maryland Response:These environmental monitors are mounted on the east and west exterior walls of thebuilding and have been evaluated for more than a decade. A synopsis of the readings is shown in the following table for the period of January 15, 1999 to present. The results are gross readings and are not background adjusted.These results are typical of what is observed during a review of all environmentalmonitors. Considering the naturally occurring background, these results are statistically indistinguishable from other areas that are not in the vicinity of a nuclear facility. Allresults of the area monitors are delivered to the University of Maryland Radiation Safety Office where they are reviewed and approved by the Radiation Safety Officer. Any elevated readings will be brought to the attention of the Reactor Director and ifwarranted, to the Radiation and Reactor Safety Committees.
BNL Review:Response Incomplete: UMD provided annual doses from two locations (one on the east wall and the other the west wall) for the time period 1999 to 2004. This data is summarized below:YearWest Wall AnnualDose (mr)East Wall AnnualDose (mr)2004 (1 st Qtr.Only)21.023.02003112.0114.02002108.0116.4 2001109.685.4 200096.5106.7 199988.587.510 CFR Part 20.1301(a)(1) establishes a limit of 100 mrem per year TEDE for membersof the public. For several of the above years, the cumulative annual doses exceed this limit. However, as discussed, these average values are not corrected for background.
The licensee does not specify what the level of background radiation is in this location.
In order to verify compliance, please provide the background radiation levels so these readings can be adjusted. The licensee has also not provided an explanation for the readings outside the reactorbuilding or the variations between readings. A typical expected background dose might be in the 50 mrem to 70 mrem range. If so, there is a measurable dose above background that needs to be explained. Further, UMD states that film badges are used for monitoring radiation levels inside the reactor building. Are the same types of detectors used outside as well? If so, UMD should explain why they are using old technology film badges rather than TLDs. Also, the monitors are numbered #2 and #7.
Are there other external monitors located outside the facility? If so, where are they located and what readings were obtained from them.RAI #73.
Discuss actual releases of airborne, liquid and solid waste from the facility for the past10 years and if these trends are expected to continue in the future. Univ. of Maryland Response:The manufacture and release of liquid and solid waste is typically limited to one cubicfoot of low-level resin from the ion exchange column and one sock filter per year. The levels are typically not more than three times the background levels of the area. The sock filter is approximately 500 grams and consists of a fibrous synthetic tube that is closed on one end. This is the coarse filter for the primary coolant system as described in the FSAR.There has been no liquid waste released other than the overflow from the vesseloverflow. This waste was surveyed by the Radiation Safety Office and approved for discharge into the environment. The sump is emptied approximately every three to five years with a total volume of approximately 50 gallons. This pattern of minimal release is expected to continue, as the MUTR is predominantly a training facility with occasionalNAA that is typically performed with extremely small quantities of samples that are held for decay and not disposed of as radioactive waste.BNL Review:Response Incomplete: The applicant has addressed the liquid and solid waste releasedfrom the facility. Both appear within regulatory guidelines and do not pose a public health and safety risk or environmental hazard. Both releases are controlled by the applicant and assessed prior to release from the facility either as LLW or to the sanitarysystem. The applicant has not addressed airborne release in a similar manner and shouldsupply that information.TECHNICAL SPECIFICATIONSRAI# 77.
TS 2.1, Safety Limit. By stipulating the safety limit for the fuel fully immersed inwater, are you ensuring that the cladding temperature will be less than 500 C at alltimes? Please give a more detailed explanation and specific references to support yourproposed safety limit.Univ. of Maryland Response: Revised TS submittal.
BNL Review: Licensee still has provided no specific reference for the 100 degree C fuel temperaturelimit.RAI# 78.
TS 2.2, Limiting Safety System Setting. Please provide the calculations referenced inthe basis for this TS that shows that the LSSS is sufficient to protect the SL with the instrumented fuel element at any position in the reactor core and the calculations that support the statement that sufficient margin is present to account for uncertainty in the accuracy of the fuel temperature measurement channel and any overshoot in reactor power resulting from a reactor transient during steady state mode operation. Section4.5.3 of the SAR discusses a LSSS of 175°C (however, Table 3.1 contains a scram set-point of 175°C) while TS 2.2 has a value of 350°C. Please explain the difference in the values. Univ. of Maryland Response:
See revised section 2.2 BNL Review:
MUTR has revised LSSS for fuel temp. from 350C to 175C. However, there is still nospecific reference for calculations mentioned in the TS.RAI# 79.
TS 3.1.3.b, Reactor Core Parameters. Your TS uses the terms "fuel elements" and "fuelbundles." Please define a fuel bundle. Is fuel normally handled as elements or bundles?
If fuel is handled in bundles, explain how the reactor will remain sub-critical if the core issub-critical by the worth of the most reactive fuel element and a fuel bundle is added to the reactor. Univ. of Maryland Response: See TS definition 1.8 BNL Review: MUTR has defined fuel bundle in TS 1.18. However, they still have not responded tothe original questions in the RAI.RAI# 80.
TS 3.1.3.c, REACTOR CORE PARAMETERS. Please provide a calculation that showsthat the reactor will remain sub-critical if the most reactive control rod is removed from the core if the four least reactive fuel bundles are removed. Would a requirement that enough fuel bundles are to be removed from the core prior to control rod removal such that the reactor remains at some minimal sub-critical level after removal of the control rod be simpler?Univ. of Maryland Response:Assuming that the four bundles are not removed from the core, the MUTR is required tohave a minimum of $0.50 shutdown margin. This margin is calculated with the most reactive rod stuck in the fully withdrawn position and the maximum permissible experimental reactivity inserted. Therefore, the MUTR cannot achieve criticality or maintain criticality even with the four bundles in the core. By removing the four referenced bundles, regardless of their position, the reactor will remain Subcritical. BNL Review: Response acceptable, but needs to be referenced in the TS Bases.
RAI# 84.
Ts 3.2.2, Reactor Control and Safety Systems. Please discuss the maximum powerramp that would result from adding $0.30 per second of reactivity to the reactor, starting from a low power condition. Also, discuss the reactor safety system response to the reactivity addition, including power overshoot.
Univ. of Maryland Response: See attachment 1 (2 pages): BNL Review:
Response acceptable, but needs to be referenced in the TS Bases. RAI# 103.
TS Figure 6.1 and 6.2. Please clarify the meaning of solid and dotted lines on thestructure diagrams. The solid line shown on Figure 6.2 between the Chairman of the Department of Materials and Nuclear Engineering and the Reactor Safety Committee is not on Figure 6.1 (similar comments on SAR Figures 12.1 and 12.2). Please explain.Univ. of Maryland Response: Following the example of ANSI/ANS 15.1, the dotted lines signify communication lineswhere the solid lines indicate reporting lines.BNL Review: As noted in RAI, figures are not consistent. Fig. 6.1 still does not show solid linebetween Chairman and RSC, as on Fig. 6.2.RAI# 107.
TS 6.2.3, Reactor Safety Committee Review Function and Sar Section 12.2.3, ReviewFunction. There are a number of items, specified for review by the safety review committee in ANS 15.1, that are not included in the responsibility of the review committee (RSC) or are significantly different from the items given in ANS-15.1. Some examples are: (1) all new procedures and major revisions thereto having safety significance, (2) proposed changes to reactor facility equipment, or systems havingsafety significance, (3) new experiments that could affect reactivity or result in the release of radioactivity, and (4) violations of internal procedures or instructions having safety significance. Please modify the committee review functions to match those inANS-15.1 or justify your proposed differences. Univ. of Maryland Response: TS modified to conform to ANSI 15.1 BNL Review:It is noted that TS 6.2.3.1 mentions an unreviewed safety question (USQ). 10 CFR50.59 no longer uses the USQ terminology, thus plant practice and procedures shouldconform to the updated regulation (Note- this problem was introduced in the new draft set of TSs).RAI# 109.
TS 6.4, Operating Procedures. TS 6.4 addresses most of the required procedure typesof ANS 15.1, but a few were not covered by the TS, specifically: administrative controls for conduct of irradiations and experiments that could affect reactor safety or core reactivity, implementation of the emergency and security plans, and personnel radiation protection (including commitment to ALARA per ANSI/ANS-15.11). Please justify the reason these are not addressed or add them to the TS. Univ. of Maryland Response: See TS as revised109.
BNL Review:Please confirm that TS 6.4.2 includes procedures or administrative controls for theconduct of irradiations and experiments. RAI# 117.
TS 6.7.2, Special Reports. NRC has changed administrative policy in a few areasrelated to this TS as follows. Provide a telephone report, confirmed in writing by fax (no telegraph), within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC operat ions center or the MUTR NRC projectmanager. Provide the 14 day written report to the NRC document control desk (no needfor copies to director of NRR or Region I). Please revise the TS accordingly. Univ. of Maryland Response:See TS as modified BNL Review:Question for NRC staff - Confirm with NRC current administrative reporting guidelinesfor reportable occurrences.New RAIs:RAI# 120.
(TS) Two of the scram setpoints listed in TS Table 3.1, 120% power and period lessthan 5 sec, are specified as LCOs. They should be listed as LSSS, similar to the scram on fuel temp. at 175C.RAI# 121.
(Chapter 2) What is the average wind speed and direction for the MUTR site?
RAI# 122.
(Chapter 16.2) Has the MUTR been used for medical therapy and are there any plans touse the facility for medical purposes during the period of extended operation?RAI# 123.
(TS) Chapter 7 of SAR discusses the two scrams that come from the rad monitoringsystem. Why are these not mentioned in TS Section 3.3 and TS Tables 3.1 & 3.5?