ML110320416
| ML110320416 | |
| Person / Time | |
|---|---|
| Site: | Rhode Island Atomic Energy Commission |
| Issue date: | 01/24/2011 |
| From: | Tehan T State of RI, Atomic Energy Comm |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML110320416 (38) | |
Text
STATE OF RHODE ISLAND AND PROVIDENCE PLANTATIONS i RHODE ISLAND ATOMIC ENERGY COMMISSION Rhode Island Nuclear Science Center 16 Reactor Road Narragansett, RI 02882-1165 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 January 24, 2010 Re: Letter dated 13 April 2010 Docket No. 50-193
Dear Mr. Kennedy:
Enclosure one is attached in reply to your Request for Additional Information (RAI) dated April 13, 2010 regarding license renewal for the Rhode Island Nuclear Science Center Reactor (RINSC). The enclosure contains the eighth set of answers to the questions specified in your letter. Tl'e RINSC staff is continuing to work on the RAI questions that remain outstanding.
Very trulyyo s, Terry Ten Director Rho s nd Nuclear Science Center I certify under penalty of perjury that the representations mad ae et and corre t.
Executed on: /,,- I By:__________
Docket No. 50-193 Enclosures 1 Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 14.39 TS 2.2.1 gives LSSS for reactor thermal power, primary coolant flow through the core, height of water above the top of the core, and reactor coolant outlet temperature. TS 2.1.1 establishes SLs for these variables. 10 CFR 50.36(c)(1)(ii)(A) requires that, "where a limiting safety system. setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded." Explain how the LSSS satisfy the requirement of 10 CFR 50.36. Include analyses, with fully justified assumptions, that show the LSSS prevent exceeding a SL for all operations allowed by the proposed TS and all credible accidents. Per 10 CFR 50.36(a)(1), these analyses shall be summarized and/or referenced in the bases for the LSSS. (See RAI 14.32)
See rewritten TS 2.2.1 in response to 14.36. A summary of the new transient analyses is included in the Bases.
14.43 The bases for TS 2.2.1 state, "flow and temperature limits were chosen to prevent incipient boiling even if transient power rises to the 2 MW trip limit of 2.4 MW."
However, Section 4.6.4 of the SAR states that during a rising power transient, the calculated fuel surface temperature would be above the onset of nucleate boiling temperature. Explain this inconsistency between the bases of TS 2.2.1 and the analysis in the SAR.
See rewritten TS 2.2.1 in response to 14.36. The revised thermal hydraulic analysis for the SAR is consistent with the revised TS.
14.51 The bases for TS 2.2.2 state, "with a 15% overpower trip, 115 kW will be the LSSS." This seems to'be an arbitrary value With no supporting analysis or justification. Provide an analysis, with fully justified assumptions, that demonstrates the LSSS on reactor power will prevent a SL from being exceeded for all operations allowed by the proposed TS and all credible accidents.
See rewritten TS 2.2.2 in response to 14.52. A summary of the revised thermal/transient analyses is included in the response to this RAI.
14.65 TS 3.1.7 states, "Experiments which could increase reactivity by flooding, shall not remain in or adjacent to the core unless the shutdown margin required in Specification 3.1.1 would be satisfied after flooding." Explain why experiments that could reduce the shutdown margin below 1.0 %Ak/k by flooding would ever be allowed in or adjacent to the core, and revise the proposed TS as appropriate.
(See RAI 4.14)
Technical Specification 1.16 takes into consideration credible malfunction in the definition of the reactivity worth of experiments. See the answer to RAI question 14.17. Technical Specification 3.1.7 makes clear that flooding is a credible malfunction.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 As discussed in the answer to RAI question 14.137, in order to determine the reactivity worth of a new experiment for which there is no data based on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core. In that case, it is possible that an experiment could be found to have enough positive reactivity that if additional positive reactivity were added due to flooding, the shutdown margin would be less than 1.0 % dK/K. In that event, Technical Specification 3.1.7 requires that the experiment be removed immediately.
14.68 TS 3.2.1 specifies reactor safety systems and safety-related instrumentation that are required for critical reactor operation. However, the proposed TS do not contain any requirements for reactor safety systems and safety-related instrumentation that must be operable when the reactor is subcritical, but not secured. Explain why the proposed TS do not require any operable safety systems or safety-related instrumentation when the reactor is subcritical, but not secured (e.g., movement of fuel in the reactor core). Explain why the radiation monitors listed in Table 3.2 are not required during work of the types specified in TS 1.19.1.c and TS 1.19.1 .d. Revise the proposed TS as appropriate.
Technical Specification 1.20 defines the reactor as "shutdown" when it is subcritical by at least the shutdown margin with the reactivity of all installed experiments included.
The term "Reactor Secured" was defined as part of the answer to RAI question 14.18 to be:
"The reactor is secured when the following conditions are met:
- a. The reactor is shutdown.
- b. The master switch is in the off position and the key is removed from the lock.
- c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods.
- d. No experiments are being moved or serviced."
There is no Technical Specification requirement for safety systems and safety related instrumentation that must be operable when the reactor is subcritical but not secured because it is impossible to do a pre-start checkout to verify the operability of the safety related instrumentation without the reactor being in a non-secured state. Condition b cannot be met because the master switch cannot be in the off position with the key removed in order to perform the pre-start checkout.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Radiation monitors are required for work of the types specified in TS 1.19.1 .c and TS 1.19.1 .d. Technical Specification 1.17 defines the reactor to be in operation whenever it is not secured or shutdown. The answer to RAI question 7.4 provides the list of the radiation monitoring instrumentation that is required to be in operation whenever the reactor is in operation (e.g., movement of fuel in the reactor core). It is possible to verify that these instruments are operable prior to taking the reactor into an unsecured state.
14.82 TS 3.2.2 references the surveillance requirements of TS 4.1.1 and TS 4.1.2.
Explain why TS 3.2.2 references these surveillance requirements.
Technical Specification 3.2 has bee re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87. The new proposed shim safety LCO specifications are covered in Technical Specifications 3.2.1.1 and 3.2.1.2. The corresponding surveillance requirements are covered in Technical Specifications 4.2.1 and 4.2.2, which were submitted as part of the answer to RAI question 14.141.
14.83 TS 3.2.3 references the surveillance requirements of TS 4.2.5 and TS 4.2.6 (the reference to TS 4.2.6 appears to be incorrect). Explain why TS 3.2.3 references these surveillance requirements.
Technical Specification 3.2 has bee re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87. The original TS 3.2.3 referred to the LCO regarding shim safety drop times. This is now covered in Specification 3.2.1.1. The original references to TS 4.2.5 and (incorrectly) 4.2.6 had to do with the surveillance requirement for shim safety drop times.
These reference have been removed, though the surveillances are included as Specifications 4.2.1.1 an 4.2.1.2 in the revised version of Technical Specification 4.2 submitted as part of the answer to RAI question 14.14 1.
14.86 The bases for TS 3.2.1 state, "the period scram limits the rate of rise of the reactor power to periods which are manually controllable." Table 3.1 indicates that the Log N Period trip channel set point is 4 seconds. The SAR does not appear to contain an analysis that shows how a reactor period slightly greater than 4 seconds would be manually controllable. Explain how a reactor period slightly greater than 4 seconds is manually controllable by the reactor operator.
The 4 second period limit serves as an auxiliary protection to assure that the reactor fuel would not be damaged in the event that there was a power transient.
As part of the answer to RAI question 13.7, an analysis was performed for a rapid insertion of 0.6 % dK/K reactivity from very low power. Effectively in this analysis, a step insertion of 0.6 % dK/K reactivity is inserted at low power and the power increases until the true power reaches the limiting safety system setting of 2.3 MW, at which point one of the over power trips cause a scram. It is assumed 3
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 that it takes 100 ms for the control blades to start dropping into the core, and that it takes 1 second for full insertion. The analysis shows that the peak fuel temperature is well below the temperature required to damage the fuel.
An insertion of 0.6 % dK/K corresponds to a period of less than 1 second.
Consequently, the consequences of a power excursion due to a 4 second period is covered by this analysis.
See the answer to RAI question 14.87 for the new basis given for the 4 second period scram.
14.87 The bases for TS 3.2 only discuss the reactor power, reactor period, and coolant flow scrams required by TS 3.2.1. Provide bases for the other safety channels and safety-related instrumentation required by TS 3.2.1, Table 3.1 and Table 3.2.
Technical specification 3.2 has been re-written to conform more closely to ANSI 15.1. Some of the specifications that had been in section 3.2 have been moved.
The following table provides a summary of how things have been changed:
Original Specification New Location Location 3.2.1 Minimum Safety Instrumentation 3.2.1.3 3.2.1.4 3.2.1.5 3.2.2 Operability of Shim Safety Blades 3.2.1.1 3.2.3 Scram Time 3.2.1.1 3.2.4 Reactivity Insertion Rate 3.2.1.2 The radiation monitoring instrumentation described in the new RINSC Technical Specification 3.2.1.3 was taken from the description given as part of the answer to RAI question 7.4. References to specific radiation monitoring instrumentation have been removed in order to allow for more flexibility in using alternative monitoring equipment. References to specific radiation alarm setpoints have been removed. RINSC has a radiation safety program, which has safety committee oversight to ensure that ALARA principles are met. Radiation levels inside the reactor room are contingent on the number, and types of experiments that are in progress. Rather than defining setpoints with the caveat that they can be adjusted higher with the approval of the approval of the facility Director or Assistant Director, setpoints will be set in a manner that ensures that the goals of the Radiation Safety Program are met. Table 3.2 will be replaced with Specifications 3.2.1.3 and 3.2.1.4.
The reactor safety and safety related instrumentation described in the new RINSC Technical Specification 3.2.1.5 was taken from the description given as part of the answer to RAI question 7.1.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 The bases for 3.2.1.1 and 3.2.1.2 refer to transient analyses that were part of the answer to RAI question 13.7.
The basis for Specification 3.2.1.4 is consistent with the answer given for RAI question 14.80 regarding the justification for being able to operate for six hours without the stack gaseous or particulate monitor.
The bases for Specification 3.2.1.5 regarding the safety limits, limiting trip values, and limiting safety system settings are consistent with the answer given for RAI question 14.36, except that the cooling modes for which the pool temperature, and primary coolant flow rate channels are required have been corrected. The basis regarding the inlet temperature channel is consistent with the answer given for RAI question 4.23. The basis regarding the outlet temperature channel is consistent with the answer given for RAI question 14.36. The basis regarding the pool temperature channel refers to the basis for Specification 2.2.2, which was updated as part of the answer to RAI question 14.52.
The new versions of Technical Specification 3.2 is:
3.2 Reactor Safety System Applicability:
This specification applies to the reactor safety system and safety related instrumentation required for critical operation of the reactor.
Objective:
The objective of this specification is to define the minimum set of safety system and safety related channels that must be operable in order for the reactor to be made critical.
Specification:
3.2.1 The reactor shall not be made critical unless:
3.2.1.1 All shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated.
3.2.1.2 The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02%
dK/K per second.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 3.2.1.3 The following area radiation monitoring instrumentation is operable:
3.2.1.3.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels shall be at the experimental level.
3.2.1.3.2 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.
3.2.1.3.3 If either of these detectors fail during operation, the staff shall have one hour to either repair the detector, or find an acceptable replacement without having to shut the reactor down.
3.2.1.4 The following air radiation monitoring instrumentation is operable:
3.2.1.4.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous effluent shall be operating.
3.2.1.4.2 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement particulate effluent shall be operating.
3.2.1.4.3 If either of these detectors fail during operation, the staff shall have six hours to either repair the detector, or find an acceptable replacement without having to shut the reactor down.
3.2.1.5 The following reactor safety and safety related instrumentation is operable and capable of performing its intended function:
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Protection Cooling Channels Function Set Point Mode Required Over Power Both 2 Scram by Power Level Less than or 105% of Equal to Licensed Power Low Pool Level Both 1 Scram by Pool Level Less than or 23 ft 9.6 Drop Equal to in Primary Coolant Inlet Forced 1 Alarm by Inlet Temp Less than or 111 F Temperature Equal to Primary Coolant Outlet Forced 1 Alarm by Outlet Temp Less than or 117 F Temperature Equal to Forced 1 Scram by Outlet Temp Less than or 120 F Equal to Pool Temperature Natural 1 Scram by Pool Temp Less than or 125 F Equal to Primary Coolant Flow Primary Flow Less than or 1800 Rate Forced 1 Scram by Rate Equal to gpm Rate of Change of Less than or 4 Power Both 1 Scram by Period Equal to seconds Seismic Disturbance Both 1 Scram if Seismic Disturbance Detected Bridge Low Power Position Forced 1 Scram if Bridge Not Seated at HP End Bridge Movement Both 1 Scram if Bridge Movement Detected Coolant Gates Open Forced 1 Scram if Inlet Gate Open Forced 1 Scram if Outlet Gate Open Detector HV Less than or Detector HV Failure Both 1 Scram if Decrease Equal to 50 V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V Detector HV Less than or Both 1 Scram if Decrease Equal to 50 V No Flow Thermal Column Forced 1 Scram by No Flow Detected Manual Scram Both 1 Scram by Button Depressed Both 1 Scram by Button Depressed No Automatic Servo Control Interlock Both 1 Servo if Regulating Blade not Full Out No Automatic 30 Both 1 Servo if Period Less than seconds Shim Safety No SS Withdrawal Both 1 Withdrawal if Count Rate Less than 3 cps No SS Both 1 Withdrawal if Test / Select SW not Off Rod Control Loss of Less than or 10 Communication Both 1 Scram if Communication Equal to jseconds 7
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Basis:
Specification 3.2.1.1 requires that all shim safety blades be capable of being fully inserted into the reactor core within 1 second from the time "that a scram condition is initiated. As part of the Safety Analysis, Argonne National Laboratory analyzed a variety of power transients in which it was assumed that the time between the initiation of a scram signal, and full insertion of all of the shim safety rods was one second. The analysis showed that if the reactor is operated within the safety limits, this time delay will not cause an over power excursion to damage the fuel.
Specification 3.2.1.2 requires that the reactivity insertion rates of individual shim safety and regulating rods do not exceed 0.02%
dK/K per second. As part of the Safety Analysis, Argonne National Laboratory analyzed ramp insertions of 0.02% dK/K reactivity from a variety of initial power levels. The reactivity insertions are stopped by the over power trip. In all cases, peak fuel and cladding temperatures due to the power overshoot are well below the temperatures required to damage the fuel or cladding.
Consequently, this limit ensures that an over power condition due to a reactivity insertion from raising a control rod will not damage the fuel or cladding.
Specification 3.2.1.3 identifies the area radiation monitoring instrumentation that is required to be operable when the reactor is operated. Radiation monitors that are capable of warning personnel of high radiation levels at the experimental elevation, and over the pool serve to ensure that personnel inside the reactor room are made aware when dose rates are higher than anticipated.
Additionally, these monitor alarms provide an indication of a potential fuel failure. In the event of a failure of either of these monitors, the operations staff is afforded the opportunity to rely on alternative monitoring instrumentation without having to shut the reactor down. This configuration has been in use for the life of the facility, without any indication that it is insufficient.
Specification 3.2.1.4 identifies the air radiation monitoring instrumentation that is required to be operable when the reactor is operated. Radiation monitors that are capable of warning personnel of high gaseous and particulate airborne radioactive material levels ensure that personnel are made aware of potential radiological releases from the stack. In the event of a failure of either of these monitors, the operations staff is afforded the opportunity to rely on alternative monitoring instrumentation 8
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 without having to shut the reactor down. This configuration has been in use for the life of the facility, without any indication that it is insufficient.
Specification 3.2.1.5 identifies the safety and safety related instrumentation that is required to be operable when the reactor is operated.
Two independent power level channels are required for both forced and natural convection cooling modes of operation, each of which must be capable of scramming the reactor by 105%
licensed power. The basis section of Specification 2.2.1 shows that this ensures that the power level safety limit of 2.4 MW will not be exceeded. Having two independent power level channels ensures that at least one over power protection will be available in the event of an over power excursion.
One low pool level channel is required for both forced and natural convection cooling modes of operation. This channel ensures that the reactor will not be in operation if the pool level is below the safety limit of 23 ft 6.5 inches above the top of the core.
One primary inlet coolant temperature channel is required for forced convection cooling mode operation. This channel alerts the operator in the event that the inlet temperature reaches 111 F. The steady state thermal hydraulic analysis that was done by Argonne National Laboratory for forced convection flow predicts that the inlet temperature would be 115 F for operation at 2.4 MW, with a primary flow of 1580 gpm and an outlet temperature of 125 F.
One primary outlet temperature channel is required for forced convection cooling mode operation. This channel is capable of scramming the reactor When the temperature reaches 120 F.
The basis section of Specification 2.2.1 shows that this ensures that the coolant outlet temperature safety limit of 125 F will not be exceeded.
One pool temperature channel is required for natural convection cooling mode of operation. This channel is capable of scramming the reactor when the temperature reaches 125 F.
The basis section of Specification 2.2.2 shows that this ensures that the pool temperature safety limit of 130 F will not be exceeded. This channel provides the over temperature 9
. Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 protection when the reactor is operated in the natural convection cooling mode.
One primary coolant flow rate channel is required for forced convection cooling mode operation. This channel assures that the reactor will not be operated at power levels above 100 kW with a primary coolant flow rate that is less than the safety limit of 1580 gpm. The basis section of Specification 2.2.1 shows that if this channel is set to scram at a limiting safety system setting of 1800 gpm, the safety limit will not be exceeded.
One rate of change of power channel is required for both cooling modes of operation. The 4 second period limit serves as an auxiliary protection to assure that the reactor fuel would not be damaged in the event that there was a power transient.
As part of the Safety Analysis, Argonne National Laboratory analyzed a power excursion involving a period of less than 1 second, which was stopped by an over power scram when the true power reached the limiting safety system setting of 2.3 MW. The analysis showed that peak fuel temperatures stayed well below the temperature required to damage the fuel. A 4 second period limit provides an additional layer of protection against this type of transient.
One seismic disturbance scram is required for both modes of operation. In the event of a seismic disturbance, the shim safety blade magnets would be likely to drop the blades due to the vibration caused by the disturbance. However, this scram ensures that the blades will be dropped in the event of a disturbance.
One bridge low power position scram is required for forced convection cooling mode operation. In order for the forced convection cooling system to work, the reactor must be seated against the high power section pool wall. This scram ensures that the reactor is properly positioned in the pool so that the coolant ducts are properly coupled with the cooling system piping.
One bridge movement scram is required for both modes of operation. This scram assures that the reactor will be shut down in the event that the bridge moves during operation.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 One coolant gate open scram on each coolant duct is required during forced convection cooling mode operation. These scrams ensure that coolant flow through the inlet and outlet ducts are not bypassed during forced convection cooling.
One detector HV failure scram is required for each of the power channels, and the period channel. These channels rely on detectors that require high voltage in order to be operable.
These scrams assure that the reactor will not be operated when one of these detectors does not have proper high voltage.
One no flow thermal column scram is required during forced convection cooling mode operation. This scram ensures that there is coolant flow through the thermal column gamma shield during operations above 100 kW.
Two manual scram buttons are required to be operational during both modes of operation. One manual scram button is located in the control room, which provides the operator with a mechanism for manually scramming the reactor. The second scram button is on the reactor bridge, which provides anyone directly over the core with a mechanism for scramming the reactor if there were a reason to do so.
One servo control interlock that prevents the regulating blade from being put into automatic servo mode unless the blade is fully withdrawn is required for both modes of operation. As a result of this interlock, when the regulating blade is transferred to automatic servo control, the blade is unable to insert additional reactivity into the core.
One servo control interlock that prevents the regulating blade from being put into automatic mode if the period is less than 30 seconds is required for both modes of operation. This interlock limits the power overshoot that occurs when the regulating blade is put into automatic mode.
One shim safety interlock that prevents shim safety withdrawal if the start up neutron count rate is less than 3 cps is required for both modes of operation. This interlock ensures that the start up channel, which is the most sensitive indication of subcritical multiplication, is operational during reactor start-ups.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 One shim safety interlock that prevents shim safety withdrawal if the neutron flux monitor test / select switch is not in the off position is required for both modes of operation. This interlock prevents shim safety withdrawal when this instrument is receiving test signals rather than actual signals from the detector that is part of the neutron flux monitor channel.
One rod control communication scram is required for both modes of operation. The control rod drive system has a communication link between the digital display in the control room, and the stepper motor controllers out at the pool top.
There is a watchdog feature that verifies that this communication link is not broken. In the event that the link is broken, a scram will occur within ten seconds of the break. All of the scram signals are sent independently of this link. The transient analysis performed by Argonne National Laboratory shows that if the control rod drive communication were lost while the reactor were on a period, the over power, and period trips would prevent the power from reaching a level that could damage the fuel cladding.
14.88 The bases for TS 3.2 do not provide bases for TS 3.2.2 and TS 3.2.3. Provide bases for these TS.
Technical Specification 3.2 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87 for the bases to these specifications.
14.89 The bases for TS 3.2 reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended.
Technical Specification 3.2 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.87 for the revised bases to these specifications.
14.116 ANSI/ANS-15.1 recommends that the technical specifications specify that experiments will be designed such that they do not contribute to the failure of other experiments or reactor systems and components important to safety. Explain the reason that the proposed TS do not contain any such requirement for experiments, and revise the proposed TS as appropriate.
Technical Specification section 3.8 will be re-written so that it more closely conforms to ANSI 15.1. The reactivity limits on experiments are covered in the re-written section of Technical Specifications 3.1.3 and 3.1.4. See the answer to RAI question 14.137. See the answer to RAI 13.7 for the transient analysis 12
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 associated with a step reactivity insertion of the maximum worth of an experiment.
Technical Specification 3.8 should be revised to say:
3.8 Experiments 3.8.1 Experiment Materials Applicability:
This specification describes the limitations on the types of materials that may be irradiated or installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to prevent damage to the reactor, reactor pool, and reactor experimental facilities.
Specification:
- 1. Corrosives Materials
- 1. Corrosive materials shall be doubly contained in corrosion resistant containers.
- 2. Highly Water Reactive Materials
- 1. Highly water reactive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
- 3. Explosive Materials
- 1. Explosive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
- 4. Fissionable Materials
- 1. The quantity of fissionable materials used in experiments shall not cause the experiment reactivity worth limits to be exceeded.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Basis:
ANSI 15.1 recommends that the kinds of materials used in experiments be taken into consideration in order to limit the possibility of damage to the reactor, reactor pool, or reactor experimental facilities. Specifically, ANSI suggests that:
Damage could arise as a result of corrosive materials reacting with core, or experimental facility materials.
Specification 3.8.1.1 reduces the possibility of this by requiring that corrosive materials be doubly contained so that the likelihood of container breach is minimized.
Damage could arise as a result of highly water reactive materials reacting with the pool water. Specification 3.8.1.2 makes this scenario impossible by prohibiting the use of highly water reactive materials in experiments.
Damage could arise as a result of explosive materials reacting inside and experimental facility. Specification 3.8.1.3 makes this scenario impossible by prohibiting the use of explosive materials in experiments.
Failure of experiments that contain fissionable materials have the potential to have an impact on reactor criticality, or on radioactive material release. The consequence of experiment failure on criticality is bounded by limiting the reactivity worths of experiments. The analysis for this is in SAR Chapter 13 as part of the transient analysis. The radioactive material release is bounded by the analysis in SAR Chapter 13 for the Maximum Hypothetical Accident involving a fuel element failure.
3.8.2 Experiment Failures or Malfunctions Applicability:
This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Objective:
The objective of this specification is to ensure that experiments cannot fail in such a way that they contribute to the failure of other experiments, core components, or principle barriers to the release of radioactive material.
Specification:
- 1. Experiment design shall be reviewed to ensure that credible failure of any experiment will not result in releases or exposures in excess of limits established in 10 CFR 20.
- 2. Experiment design shall be reviewed to ensure that no reactor transient can cause the experiment to fail in such a way that it contributes to an accident.
- 3. Experiment design shall be reviewed to ensure that credible failure of any experiment will not contribute to the failure of:
- 1. Other Experiments
- 2. Core Components
- 3. Principle physical barriers to uncontrolled release of radioactivity
- 4. Experiments which could increase reactivity by flooding shall not remain in the core, or adjacent to the core unless the minimum core shutdown margin required would be satisfied with the experiment in the flooded condition.
Basis:
ANSI 15.1 recommends that experiment design be taken into consideration in order to limit the possibility that an experiment failure or malfunction could result in other failures, accidents, or significant releases of radioactive material.
Experiments are reviewed by the RINSC Nuclear and Radiation Safety Committee prior to being authorized to be installed in the reactor pool, or inside the reactor experimental facilities. These specifications ensure that experimental design is considered as part of the review, in 15
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 order to minimize the possibility of these types of problems due to experiment failure or malfunction.
In order to determine the reactivity worth of a new experiment for which there is no data based on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core. In that case, it is possible that an experiment could be found to have enough positive reactivity that if additional positive reactivity were added due to flooding, the shutdown margin would be less than 1.0 % dK/K. In that event, Technical Specification 3.8.2.4 requires that the experiment be removed immediately.
14.117 TS 3.8.3 states, "Fissionable materials shall have total iodine and strontium inventory less than that allowed by the facility by-product license." What facility by-product license does this specification reference? What inventory limits does that by-product license specify? Why are iodine and strontium the only elements of concern for experiments involving fissionable materials? Provide an analysis of the consequences of the failure of an experiment involving fissionable materials that shows the consequences are bounded by the analysis of the MHA presented in Chapter 13 of the SAR. Discuss all assumptions used in the analysis, including justification for the use of the assumptions.
Please change proposed technical specification 3.8.3 to read: "Each experiment containing fissionable materials shall be limited to a maximum reactivity worth of 0.60 % AK/K if secured or 0.08 % AK/K if moveable. The total reactivity of all experiments shall not exceed 0.60 % AK/K." The basis for the proposed technical specification can be found in Chapter 10 of the SAR. SAR Section 10.3, "Experiment Review," states that the reactivity worth of any single independent experiment or combination of connected experiments that can be added to the core simultaneously cannot exceed 0.60% AK/K and the calculated reactivity worth of any single independent experiment not rigidly fixed in place or the combination of connected or related experiments added to the core simultaneously cannot exceed 0.08% AK/K. Positive reactivity is the result of the insertion of either fissile materials or reflector materials into the core.
To address the questions posed in the RAI, when the SAR was written, the facility had an Agreement State broad-scope radioactive materials license and that is the license to which the quoted statement refers. When the facility was governed by both licenses, the broad-scope license allowed inventory was limiting. Among the numerous radionuclides formed when fissile material is fissioned in an experiment, the limitations on strontium and iodine were the most restrictive of the inventory limits in the broad-scope license. During the long delay between when the SAR was submitted and the NRC completed its review and issued this 16
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 RAI, the broad-scope license was dropped. It should be noted that our current technical specifications contain the same restriction and we were asked by the NRC not to submit any requests for amendments to our license while we were awaiting review of our SAR. Thus, since the limiting broad-scope license inventory items no longer exist, the questions posed in the first portion of this RAI are essentially moot at this point.
It is our contention that the limitations on the reactivity worth of an experiment essentially assures that the consequences of failure of that experiment will remain within the dose equivalent consequences of the fuel element failure. It should be noted that the RINSC is currently licensed to increase the core fuel elements from fourteen to seventeen. Each additional fuel element provides approximately 275 grams of uranium-235 far exceeding the reactivity of any single experiment or combination of experiments containing fissionable material. Additionally, there are technical specification limits on core excess reactivity and core shutdown margin that must be met taking the experiment into account. The inventory of radioactivity in the core is dependent on core power level and the RINSC is limited to 2 MWI. The MHA assumes the failure of a fuel element containing the fission products from far more fissionable material than any single experiment or combination of experiments. Thus, the MHA bounds the radiological consequences of the failure of an experiment containing fissionable materials.
14.118 TS 3.8.5 states, "experiments shall be designed against failure from internal and external heating at the true values associated with the LSSS for reactor power level and other process variables." ANSI/ANS-15.1 recommends that experiments also be able to withstand reactor transients. Section 4.6.4 of the SAR states that a rising power transient could result in a maximum reactor power of 2.78 MW, which is greater than the LSSS value of 2.30 MW. Explain how TS 3.8.5 ensures that experiments will be designed to withstand reactor transients, and revise the proposed TS as appropriate.
Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. There is no specific discussion about internal or external heating. Refer to the answer provided for RAI question 14.116.
14.119 The requirements of TS 3.8.10 imply that accidents involving experiments could result in occupantional and public radiation doses up to the regulatory limits.
These doses would be greater than the consequences of a fuel failure accident analyzed in Section 13.2.1 of the SAR. Explain why the SAR considers the fuel failure accident to be the MHA if the failure of an experiment could have greater consequences. Provide an analysis of the occupational and public dose consequences of the worst-case failure of an experiment that is consistent with the requirements of the proposed TS. Discuss all assumptions used in the analysis, including justification for the use of the assumptions.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Once more the RAI shows the inherent confusion in the guidance provided by the NRC in NUREG-1537, Part 1. The NUREG uses ANSI/ANS-15.1-1990 as its recommended (required) guide. In keeping with ANSI/ANS-15.1-1990, Section 3.8.3 (1), credible failure of any experiment shall not result in releases or exposures in excess of established limits nor in excess of annual limits established in 10 CFR 20. Proposed technical specification 6.5.9, "Operating Procedures,"
states, in part: "Experiment review on a case-by-case basis assuring that section 3.8.3(2) of ANSI/ANS 15.1 is satisfied." Experiments are reviewed by the Nuclear and Radiation Safety Committee prior to initiation (see proposed technical specification 6.4.2.b).
14.121 TS 3.8.10 specifies requirements related to failure of an experiment encapsulation. Explain what specific types of encapsulation are covered by TS 3.8.10, and revise the proposed TS as appropriate.
As originally submitted, specification 3.8.1 indicates that sample materials that are going to be irradiated must be contained in containers or encapsulation materials that do. not react with water, and do not induce corrosion of core and core structural materials. The type of encapsulation is not relevant, as long as it is able to contain the sample material, and as long as the material used for the containers will not react with any of the core, core structure, or coolant materials.
As originally submitted, specification 3.8.10 describes actions to be taken in the event of the failure of a sample container.
Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. There is no discussion about types of encapsulation. Refer to the answer provided for RAI question 14.116.
14.122It appears that the first sentence of the second paragraph of TS 3.8.10 explicitly excludes "fuel materials" from the requirements of TS 3.8.10. Clarify whether "fuel materials" is synonymous with "fissionable materials" as used in TS 3.8.3. If the requirements of TS 3.8.10 exclude fissionable materials, explain the reason for not including similar requirements for experiments that contain "fuel materials," and revise the proposed TS as appropriate.
Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. The radiological release, of all experiments is now covered in Specification 3.8.2.1. Refer to the answer provided for RAI question 14.116.
14.126 The bases for TS 3.8 state that several of the specifications are "self explanatory." In accordance with 10 CFR 50.36, provide bases for all of the specifications in TS 3.8.
Technical Specification 3.8 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.116.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 14.128 The bases for TS 3.9.a reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended.
TS 3.9a references "Part A Section VIII" of the SAR. This section is from a previous SAR. TS 3.9a should reference Section 4.2.3, 'Neutron Moderators and Reflectors'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 14.132 TS 4.1.1 does not require surveillance of the shim safety blades following maintenance or replacement. Explain the reason for not requiring surveillance of the shim safety blades following maintenance or replacement, and revise the proposed TS as appropriate.
Technical Specifications 4.1 and 4.2 have been re-written in order to make them conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138 for the revised version of Technical Specification 4.1, and refer to the answer provided for 14.141 for the revised version of Technical Specification 4.2. Specification 4.1.1.2 provides the surveillance requirements for shim safety blade reactivity worths. Specification 4.2.2 provides the surveillance requirements for shim safety blade reactivity insertion rates. In both cases, a surveillance requirement has been added to cover activities that could have an effect on these parameters.
14.133 TS 4.1.1 .b references the startup core and three other analyzed cores. Explain the reason for referencing the startup core, and revise the proposed TS as appropriate.
There is no reason for specific core configurations to be referenced in this section of the specification. Technical Specification 4.1 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138. The surveillance requirements have been written so that they do not refer to specific core configurations.
14.134 TS 4.1.1.b implies that there are only three allowed core configurations for the RINSC reactor. The proposed TS do not contain an LCO restricting the configuration of the RINSC core to three configurations. Explain the reason for only requiring surveillance of the shim safety blades when switching to one of the three referenced core configurations, and revise the proposed TS as appropriate.
Technical Specification 4.1 has been re-written in order to make it conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138. The surveillance requirements have been written so that they do not refer to specific core configurations.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 14.135 TS 4.1.1.b references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.
Technical Specifications 4.1 and 4.2 have been re-written in order to make them conform more closely to ANSI 15.1. Refer to the answer provided for RAI question 14.138 for the revised version of Technical Specification 4.1, and refer to the answer provided for 14.141 for the revised version of Technical Specification 4.2. Specification 4.1.1.2 provides the surveillance requirements for shim safety blade reactivity worths. Specification 4.2.2 provides the surveillance requirements for shim safety blade reactivity insertion rates.
References to the SAR have been removed.
14.138The bases for TS 4.1.3 state that the specification "provides assurance that experiment reactivity worths do not increase beyond the established limits due to core configuration changes." The specification does not appear to require any surveillance of experiment reactivity worths following core configuration changes. Explain the apparent inconsistency between the specification and the bases, and revise the proposed TS as appropriate.
Technical Specifications 3.1 and 4.1 have been re-written in order to make them conform more closely to ANSI 15.1. Some of the specifications that had been in section 3.1 have been moved. The table that follows provides a summary of how things have been changed. The new Technical Specification 4.1.1.3.2 addresses the issue regarding experiment worth changes due to core configuration changes.
Original Specification New Location Location 3.1.1 Shutdown Margin Reactivity 3.1.1.1.1 3.1.2 Core Excess Reactivity 3.1.1.1.2 3.1.3 Total Experiment Reactivity Worth 3.1.1.3.1 3.1.4 Individual Experiment Reactivity Worth 3.1.1.3.2 3.1.5 Criticality During Fuel Loading 3.1.1.1.4 3.1.6 Regulating Rod Reactivity Worth 3.1.1.2.1 3.1.7 Flooded Experiment 3.8.2.4 3.1.8 Negative Temperature Coefficient 3.1.1.1.3 Temperature Coefficient Surveillance 4.1.1.1.3 3.1.9 FC Mode Operation Core Grid Filled 3.1.2.1 3.1.10 FC Mode Operation Coolant Gate Stored 3.1.2.2 The basis section of Specifications 3.1.1.3.1 and 3.1.1.3.2 refer to analyses performed for reactivity insertions. The determination of a period due to a 0.08 %
dK/K reactivity insertion is part of the answer to RAI question 14.61.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 The basis section of Specification 4.1.1.1.3 refers to an analysis that estimates the temperature coefficient. This analysis is part of the answer to RAI questions 4.12 and 4.13.
The new versions of Technical Specification 3.1 and 4.1 are:
3.1 Core Parameters 3.1.1 Reactivity Limits Applicability:
This specification applies to all core configurations, including configurations that have experiments installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to make certain that core reactivity parameters will not exceed the limits used in the safety analysis to ensure that a reactor transient will not result in damage to the fuel.
Specification:
3.1.1.1 Core 3.1.1.1.1 The core shutdown margin shall be at least 1.0 % dK/K.
3.1.1.1.2 The core excess reactivity shall not exceed 4.7 % dK/K.
3.1.1.1.3 The temperature coefficient shall be negative.
3.1.1.1.4 The reactor shall be subcritical by at least 3.0 %dK/K during fuel loading changes.
3.1.1.2 Control Rods 3.1.1.2.1 The reactivity worth of the regulating rod shall not exceed 0.6 % dK/K.
3.1.1.3 Experiments 21
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 3.1.1.3.1 The total reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth:
Total Moveable and Fixed 0.6 %dKiK Total Moveable 0.08 %dK/K 3.1.1.3.2 The maximum reactivity worth of any individual experiment shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth:
Fixed 0.6 % dK.K Moveable 0.08 % dK/K Basis:
Specification 3.1.1.1.1 provides a limit for the minimum shutdown reactivity margin that must be available for all core configurations. The shutdown margin is necessary to ensure that the reactor can be made subcritical from any operating condition, and to ensure that it will remain subcritical after cool down and xenon decay, even if the most reactive control rod failed in the fully withdrawn position. No credit is taken for the negative reactivity worth of the regulating rod because it would not be available as part of the negative reactivity insertion in the event of a scram.
Specification 3.1.1.1.2 provides a maximum limit for excess reactivity available for all core configurations.
Excess reactivity is necessary to overcome the negative reactivity effects of coolant temperature increase, coolant void increase, fuel temperature increase, and xenon build-up that occur during sustained operations. Excess reactivity is also required to be available in order to overcome any negative reactivity effects of experiments that are installed in the core.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Specification 3.1.1.1.3 requires that the temperature coefficient be negative. This requirement ensures that a temperature rise due to a reactor transient will not cause a further increase in reactivity.
Specification 3.1.1.1.4 provides a limit for the minimum core shutdown reactivity during fuel loading changes. This limit takes advantage of the negative reactivity that can be added to the core above and beyond the shutdown margin by the insertion of the highest reactivity worth, and regulating control rods. This limit assures that the core will remain subcritical during these operations.
Specification 3.1.1.2.1 provides a limit for the reactivity worth of the regulating rod. The reactivity limit is set to a value less than the delayed neutron fraction so that a failure of the automatic servo system could not result in a prompt critical condition.
Specification 3.1.1.3.1 provides total reactivity limits for all experiments installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The limit on total experiment worth is set to a value less than the delayed neutron fraction so that an experiment failure could not result in a prompt critical condition. The limit on total moveable experiment worth is set to a value that will not produce a stable period of less than 30 seconds, so that the reactivity insertion can be easily compensated for by the action of the control and safety systems. As part of the Safety Analysis, Argonne National Laboratory modeled a reactivity insertion of + 0.08 % dK/K over a 0.1 second interval, and determined that this reactivity insertion resulted in a stable period of approximately 75 seconds.
Specification 3.1.1.3.2 provides total reactivity limits for any individual experiment installed in the reactor, the reactor pool, or inside the reactor experimental facilities.
The reactivity limits for both, fixed and moveable experiments are the same as the limits for total fixed and moveable experiments. Consequently, the safety analysis done for Specification 3.1.1.3.1 applies to this specification as well.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 3.1.2 Core Configuration Limits Applicability:
This specification applies to core configurations during operations above 0.1 MW when the reactor is in the forced convection cooling mode.
Objective:
The objective of this specification is to ensure that there is sufficient coolant to remove heat from the fuel elements when the reactor is in operation at power levels greater than 0.1 MW.
Specification:
3.1.2.1 All core grid positions shall contain fuel elements, baskets, reflector elements, or experimental facilities during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.
3.1.2.2 The pool gate shall be in its storage location during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.
Basis:
Specification 3.1.2.1 requires that all of the core grid spaces be filled when the reactor is operated at higher power levels that require forced convection cooling. This requirement prevents the degradation of coolant flow through the fuel channels due to flow bypassing the actively fueled region of the core through unoccupied grid plate positions.
Specification 3.1.2.2 requires that the pool gate that is used for separating the sections of the pool, be in its storage location when the reactor is in operation at higher power levels that require forced convection cooling. This requirement ensures that there will be a sufficient heat sink for high power operations, and ensures that the full volume of the pool water will be available in the event of a loss of coolant accident.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 4.1 Core Parameter Surveillance 4.1.1 Reactivity Limit Surveillance I
Applicability:
This specification applies to the surveillance requirements for reactivity limits.
Objective:
The objective of this specification -is to ensure that reactivity limits are not exceeded.
Specification:
4.1.1.1 Core Reactivity Limit Surveillance 4.1.1.1.1 The core shutdown margin shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed 4.1.1.1.2 The core excess reactivity shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed 4.1.1.1.3 The temperature coefficient shall be shown to be negative at the initial start-up after a fuel type change.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 4.1.1.2 Control Rod Reactivity Limit Surveillance 4.1.1.2.1 The reactivity worth of the regulating rod shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.2.2 The reactivity worth of the shim safety rods shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.3 Experiment Reactivity Limit Surveillance 4.1.1.3.1 The reactivity worth of new experiments shall be determined prior to the experiments initial use.
4.1.1.3.2 The reactivity worth of any on going experiments shall be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously.
Basis:
Specification 4.1.1.1.1 requires that the core shutdown margin be determined annually, and whenever there is a change in core loading or core reflection. The annual measurement of the shutdown margin provides a snapshot 26
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 of how the shutdown margin is increasing due to fuel burn-up. Measurements made whenever the core loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.
Specification 4.1.1.1.2 requires that the core excess reactivity be determined annually, and whenever there is a change in core loading or core reflection. The annual' measurement of the excess reactivity provides a snapshot of how it is decreasing due to fuel burn-up. Measurements made whenever the core loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.
Specification 4.1.1.1.3 requires that the temperature coefficient be shown to be negative at the initial start-up after a fuel type change. A negative temperature coefficient makes power increases self limiting by inserting a negative reactivity effect as fuel and coolant temperatures rise. As part of the Safety Analysis, Argonne National Laboratory determined that for the equilibrium core, the temperature and void coefficients are negative over a temperature range of 20 C to 100 C. The fuel temperature coefficient was determined to be negative over a temperature range of 20 C to 600 C.
Specification 4.1.1.2.1 requires that the regulating rod reactivity be determined annually, and whenever there is a change in core loading or core reflection. These determinations provide assurance that the rod worth does not exceed its reactivity limit due to fuel burn-up, changes in core configuration, or control rod degradation.
Specification 4.1.1.2.2 requires that the shim safety rod reactivities be determined annually, and whenever there is a change in core loading or core reflection. These determinations provide assurance that the rod worths do not degrade due to rod changes, or changes in core configuration.
Specification 4.1.1.3.1 requires that the reactivity worth of new experiments be determined prior to initial use. This ensures that reactivity worth limits are not exceeded.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Specification 4.1.1.3.2 requires that the reactivity worth of on going experiments be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously. This provides assurance that core configuration changes do not cause experiment reactivity worth limits to be exceeded, without requiring that experiment worths be re-determined every time that a recurring core configuration change, such as equilibrium core re-fuelling, occurs.
14.139 ANSI/ANS-15.1 recommends annual thermal power verification. Explain the reason that the proposed TS do not contain any such requirement, and revise the proposed TS as appropriate.
RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. Specification 4.2.7.5 requires that the power level channels be calibrated annually. This calibration is done by thermal power verification.
14.140 ANSI/ANS-15.1 recommends annual surveillance of required interlocks.
Explain the reason that the proposed TS do not contain any such requirements, and revise the proposed TS as appropriate.
RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. Interlock surveillance is covered in Specification 4.2.6.
14.141 TS 4.2 specifies, surveillance requirements for the safety system and safety-related instrumentation required by TS 3.2.1. However, the proposed TS do not specify surveillance requirements for many of the items required by TS 3.2.1, Table 3.1 and Table 3.2. In accordance with 10 CFR 50.36(c)(3), propose surveillance requirements for the safety system and safety related instrumentation required by TS 3.2.1.
Specification 4.2 has been revised to more closely reflect what ANSI 15.1 suggests should be covered by this specification. The following table shows how the locations of the original specifications have changed.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Specification Original ANSI New Location Standard Location Channel Test of Neutron Flux Level Safeties 4.2.1 4.2.5 4.2.4.2 Channel Test of Period Safety 4.2.1 4.2.5 4.2.4.3 Channel Calibration of Channels in Table 3.1 4.2.2 4.2.5 4.2.7 Radiation Monitors in Table 3.2 Operable 4.2.3 4.2.3 Rod Drop ime4.2.4 4.2.4 4.2.1.1 Rod Drop Time 4.2.5 4.2.4 4.2.1.2 Shutdown Margin 4.2.6 4.1.2 4.1.1.1.1 Excess Reactivity 4.2.7 4.1.1 4.1.1.1.2 Reactivity Insertion Rate 4.2.8 4.2.2 4.2.2 Control Rod Reactivity Worth 4.1.1 4.2.1 4.1.1.2 Power Calibration 4.2.8 4.2.7.5 The following is the new proposed Specification:
4.2 Reactor Controland Safety System Applicability:
This specification applies to the safety and safety related instrumentation.
Objective:
The objective of this specification is to ensure that the safety and safety related instrumentation is operable, and calibrated when in use.
Specification:
4.2.1 Shim safety drop times shall be measured:
4.2.1.1 Annually 4.2.1.2 Whenever maintenance is performed which could affect the drop time of the blade 4.2.2 All shim safety reactivity insertion rates shall be measured:
4.2.2.1 Annually 4.2.2.2 Whenever maintenance is performed which could affect the reactivity insertion rate of the blade 4.2.3 The following radiation monitors shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized:
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 4.2.3.1 The experimental level area radiation monitor 4.2.3.2 The pool top area radiation monitor 4.2.3.3 The gaseous effluent air monitor 4.2.3.4 The particulate air monitor 4.2.4 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized:
4.2.4.1 Control room manual scram button 4.2.4.2 Power level channels 4.2.4.3 Period channel 4.2.4.4. Rod control communication watchdog scram 4.2.5 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized for which reactor power level will be greater than 100 kW:
4.2.5.1 All. of the reactor safety and safety related instrumentation listed in 4.2.4 4.2.5.2 Primary coolant flow rate scram 4.2.6 The following reactor safety and safety related instrumentation alarms, scrams, and interlocks shall be tested annually:
4.2.6.1 The following detector HV failure scrams:
4.6.2.1.1 Power level channels 4.6.2.1.2 Period channel 4.2.6.2 The following shim safety withdrawal interlocks:
4.2.6.2.1 Start-up count rate 4.2.6.2.2 Test / Select switch position 4.2.6.3 The following servo control interlocks:
4.2.6.3.1 Regulating blade not full out 4.2.6.3.2 Period less than 30 seconds 30
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, -2010 4.2.6.4 The following coolant system channel temperature alarms and scrams:
4.2.6.4.1 Primary inlet temperature alarm 4.2.6.4.2 Primary outlet temperature alarm 4.2.6.4.3 Primary outlet temperature scram 4.2.6.4.4 Pool temperature alarm 4.2.6.4.5 Pool temperature scram 4.2.6.5 The following coolant system channel flow scrams:
4.2.6.5.1 Primary flow scram 4.2.6.5.2 Inlet and outlet coolant gates open scrams 4.2.6.5.3 No flow thermal column scram 4.2.6.6 Low pool level scram 4.2.6.7 The following bridge scrams:
4.2.6.7.1 Bridge manual scram 4.2.6.7.2 Bridge movement scram 4.2.6.7.3 Bridge low power position scram 4.2.6.8 Seismic scram 4.2.7 The following reactor safety and safety related instrumentation shall be calibrated annually:
4.2.7.1 The experimental level area radiation monitor 4.2.7.2 The pool top area radiation monitor 4.2.7.3 The gaseous effluent air monitor 4.2.7.4 The particulate air monitor 4.2.7.5 Power level channels 4.2.7.6 Primary flow channel 4.2.7.7 Primary inlet temperature channel 4.2.7.8 Primary outlet temperature channel 4.2.7.9 Pool temperature channel Basis:
Specification 4.2.1 defines the surveillance interval for measuring the shim safety drop times. The annual, requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1. The requirement that this parameter be measured after maintenance is performed which could affect the drop time of the 31
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 blade assures that the reactor will not be operated with a shim safety blade that does not meet the LCO requirements due to maintenance activities.
Specification 4.2.2 requires that all shim safety reactivity insertion rates shall be measured annually. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.
Specification 4.2.3 indicates the radiation monitors that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.
Specification 4.2.4 indicates the reactor safety and safety related instrumentation that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.
Specification 4.2.5 provides for the fact that if the reactor is operated at power levels less than or equal to 100 kW, the forced cooling system is not required to be operational. However, for operations above 100 kW, this specification requires that the primary coolant flow rate scram be verified to be operable prior to the initial start-up of the reactor. This requirement is consistent with the historical facility requirements.
Specification 4.2.6 defines the surveillance interval for testing the reactor safety and safety related instrumentation alarms, scrams, and interlocks that are not tested as part of the requirements of Specifications 4.2.4 and 4.2.5. The annual requirement is consistent with the historical facility frequency.
Specification 4.2.7 defines the surveillance interval for calibrating the safety and safety related instrumentation. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.
14.142 TS 4.2.1.a requires channel tests of nuclear instrumentation "prior to each reactor startup following a period when the reactor was secured." Given that the TS do not require the reactor to be secured on a periodic basis, explain the reason for not requiring periodic (e.g., quarterly) surveillance of the nuclear instrumentation, and revise the proposed TS as appropriate.
While we do not want to limit our ability to operate the reactor for extended runs over multiple days, the current typical operating schedule at RINSC is one shift per day. Our desire is to set this surveillance such that these channel checks are performed once prior to the initial start-up of the day, so 32
Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 that if there are multiple start-ups for the day, additional channel checks are not required.
In the event that there were a multi-day operation, it is not considered likely that RINSC could operate for a quarter of a year without re-fuelling. RINSC reached its equilibrium core in October of 2008. Based on operating data, we expect to have to refuel after 1550 MWH of operation. Therefore, if we were start with a fresh core, and operate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week, we would reach our limit of 1550 MWH of operation in 2.1 months.
Consequently, quarterly surveillance in lieu of prior to initial start-up of the day is redundant.
If the wording is changed to make sure that pre-start checkouts are performed prior to the initial start-up each day, that the reactor is started up from the shutdown condition, rather than after it has been secured, these conditions can be met.
RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. The new proposed specification regarding the operability of the Neutron Flux Level Safety and Period Safety Channels is:
4.2.4 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired or de-energized:
4.2.4.2 Power level channels 4.2.4.3 Period channel 14.143 TS 4.2.2 states, "A channel calibration of the safety channels listed in Table 3.1, which can be calibrated, shall be performed annually." Revise the proposed TS to explicitly state which channels listed in Table 3.1 will be calibrated annually.
Table 3.1 was updated as part of the answer to RAI question 7.1. RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141. The new proposed specification regarding channel calibrations is:
4.2.7 The following reactor safety and safety related instrumentation shall be calibrated annually:
4.2.7.1 The experimental level area radiation monitor 4.2.7.2 The pool top area radiation monitor 4.2.7.3 The gaseous effluent air monitor 4.2.7.4 The particulate air monitor 4.2.7.5 Power level channels 4.2.7.6 Primary flow channel
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 4.2.7.7 Primary inlet temperature channel 4.2.7.8 Primary outlet temperature channel 4.2.7.9 Pool temperature channel 14.144 TS 4.2.3 appears to be an LCO and not a surveillance requirement. Explain the reason for including TS 4.2.3 in the surveillance requirements, and revise the proposed TS as appropriate.
The LCO regarding the required radiation monitoring instrumentation is covered in the new proposed Specifications 3.2.1.3 and 3.2.1.4 which were submitted as part of the answer to RAI question 14.87. The corresponding surveillance requirements are part of the revised Specification 4.2 which is part of the answer to RAI question 14.141.
14.146 TS 4.2.6 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.
Technical Specification 4.2 has been re-written as part of the answer to RAI 14.141. The reference to the SAR has been removed.
14.148 TS 4.2.7 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed IS will become part of the IS and license.
Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.
Technical Specification 4.2 has been re-written as part of the answer to RAI 14.141. The reference to the SAR has been removed.
14.150 TS 4.2.8 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed IS will become part of the IS and license.
Explain why it is necessary to make the referenced portion of the SAR a requirement in the proposed TS.
Technical Specification 4.2 has been re-written as part of the answer to RAI 14.141. The reference to the SAR has been removed.
14.151 The "Bases" section of TS 4.2 does not contain bases for TS 4.2.6, 4.2.7, or 4.2.8. Provide bases for these specifications.
RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 14.152 The bases for TS 4.2.3 states, "Radiation monitors are checked for proper operation in Specification 4.2.3. Calibration and setpoint verification involve..."
However, TS 4.2.3 appears to be an LCO and does not specify surveillance requirements (e.g., channel tests, channel checks, or channel calibrations).
Explain this apparent inconsistency between the specification and the bases for IS 4.2.3, and revise the proposed IS as appropriate.
RINSC Technical Specification 4.2 has been revised. See the answer to RAI question 14.141.
14.163 The bases for TS 4.8 state, "Review of the experiments... assures that the insertion of experiments will not negate the consideration implicit in the Safety Limits."
Explain what "consideration implicit in the Safety Limits" means in terms of experiments.
This statement was intended to mean that the safety review of experiments will ensure that the installation of experiments will not put the reactor in a condition that makes reaching a safety limit credible.
Technical Specification 4.8 will be re-written to say:
4.8 Experiments Applicability:
This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to ensure that experiments have been reviewed to verify that the design is within the limitations of the RINSC Technical Specifications and 10 CFR 50.59.
Specification:
4.8.1 Experimehts shall be reviewed to ensure that the design is within the limitations of the R1NSC Technical Specifications and 10 CFR 50.59 prior to the experiments initial use.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Basis:
This specification ensures that all experiments will be reviewed to verify that the experiment designs are within the limitations of the RINSC Technical Specifications and 10 CFR 50.59 prior to its initial use.
14.165 The bases for TS 4.9.a reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application. Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended.
TS 4.9a references "Part A Section VIII" of the SAR. This section is from a previous SAR. TS 4.9a should reference Section 4.2.3, 'Neutron Moderators and Reflectors'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 14.168 The bases for TS 4.9.b reference a version of the SAR that is different than the version of the SAR submitted with the license renewal application.
Revise the proposed TS to refer to the SAR submitted with the license renewal application, as amended.
TS 4.9b references "Part A Section VI" of the SAR. This section is from a previous SAR. TS 4.9b should reference Section 4.5, 'Nuclear Design'. The TS shall be changed to reference the current SAR. (See response for RAI 14.1) 14.169 In accordance with 10 CFR 50.36(a)(1), provide bases for proposed technical specifications in Section 5, "Design Features."
Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172.
14.170 ANSI/ANS-15.1 recommends that the number and type of control blades be included in the technical specifications. Explain the reason that the regulating blade is not specified in TS 5.3. Explain the reason that the control blade materials are not specified in the proposed TS, and revise the proposed TS as appropriate.
Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. The number and type of control blades have been included in section 5.3.
14.171 Proposed TS 5.3 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced section of the SAR a requirement in the proposed TS.
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Rhode Island Nuclear Science Center Eighth Response to NRC Request for Additional Information Dated April 13, 2010 Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. References to the SAR have been removed.
14.176 Proposed TS 5.5 references the SAR. Any portion of the SAR referenced in the "Specification" section of the proposed TS will become part of the TS and license. Explain why it is necessary to make the referenced section of the SAR a requirement in the proposed TS.
Chapter 14, section 5 has been re-written in order to make it conform more closely to ANSI 15.1. See the answer to RAI question 14.172. References to the SAR have been removed.
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