ML20057A051
| ML20057A051 | |
| Person / Time | |
|---|---|
| Issue date: | 08/23/1993 |
| From: | Alexander Adams Office of Nuclear Reactor Regulation |
| To: | Cowan B ECKERT, SEAMANS, CHERIN & MELLOTT |
| References | |
| NUDOCS 9309100383 | |
| Download: ML20057A051 (67) | |
Text
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August 23, 1993 s
Mr. Barton Z. Cowan Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor Pittsburgh, Pennsylvania 15219
Dear Mr. Cowan:
SUBJECT:
REQUEST FOR DOCUMENTS As requested in your letter of August 13, 1993, please find enclosed a copy of draft Chapter 5, " Reactor Coolant and Associated Systems" and draft Chapter 14, " Technical Specifications," of the " Format and Content for Applications for the Licensing of Non-Power Reactors" and " Safety Analysis Report Review Plan and Acceptance Criteria for Non-Power Reactors."
Sincerely, Original signed by:
Alexander Adams, Jr., Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation
Enclosures:
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wassinotow o.c. rosss-oooi August 23, 1993 Mr. Barton Z. Cowan Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor Pittsburgh, Pennsylvania 15219
Dear Mr. Cowan:
l l
SUBJECT:
REQUEST FOR DOCUMENTS As requested in your letter of August 13, 1993, please find enclosed a copy of draft Chapter 5, " Reactor Coolant and Associated Systems" and draft Chapter 14, " Technical Specifications," of the " Format and Content for i
Applications for the Licensing of Non-Power Reactors" and " Safety Analysis Report Review Plan and Acceptance Criteria for Non-Power Reactors."
Sincerely, Alexander Adams, Jr., S n Project Manager Non-Power Reactors and commissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation
Enclosures:
As stated
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i 1
S SAR Standard Format and Content Chapter 5 - Draft July 2, 1993 5 REACTOR COOLANT SYSTEMS 5.1 Introduction In this chapter, the applicant should provide the design bases, descriptions, and the functional analyses of the reactor coolant systems.
The principal purpose of the coolant system is to safely remove the fission and decay heat from the fuel and dissipate it to the environment. The discussions should include all sinnificant heat sources in the reactor and show how the heat is safely removed and transferred to the environment.
The coolant in the primary systems of most non-power reactors serves more functions than just efficient removal of heat. The coolant can act as a radiation shield for the reactor, fuel storage fac',lities, and; in some designs, experimental facilities and experiments.
In open pool reactors, the coolant is the only vertical shielding.
In many designs the reactor coolant also acts as a core moderator and reflector.
Because of these many functions that the reactor coolant serves, the reactor coolant system design is based on selecting among interdependent parameters, including thermal power level, research capability, available fuel type, reactor core physics requirements, and radiation shielding.
Some non-power reactors are licensed to operate at such low power levels that no significant temperature increases will occur during normal operation. Such reactors may not require an engineered coolant system for heat removal.
For those reactors, the analysis in Chapter 4, " Reactor Description," should discuss the disposition of the heat produced, estimate potential temperature increases during operation, and justify why an engineered coolant system for heat removal is not required. This chapter should summarize those considerations and conclusions.
For all other non-power reactors, systems to remove and dispose of the waste heat should be described and discussed in this chapter. The design bases of the reactor cooiing systems for normal and off-normal operation will be based on ensuring acceptable reactor conditions established in Chapter 4.
The design bases of any features of the core cooling system designed to respond to potential accidents or to mitigate the consequences of potential accidents should be derived from the analyses in Chapter 13, " Accidents." These features should be summarized in this chapter and should be discussed in detail in Chapter 6, " Engineered Safety Features." This chapter of the SAR should discuss and reference the technical specifications where analyses are used as the basis for a requirement.
All auxiliary and subsystems that use and contribute to the heat load of either the primary or secondary coolant should be described and discussed in this chapter. Any auxiliary systems using' coolant from other sources should be discussed in Chapter 9, " Auxiliary Systems."
The primary loop of the coolant systems of most licensed non-power reactors are of two basic types, forced convection and natural thermal convection.
Facilities using forced-convection cooling also may be licensed to operate in the natural-convection mode. All non-power reactors with engineered coolant systems that do not have active decay heat removal systems should be capable 1
SAR Standard Format and Content Chapter 5 - Draft '
June 22, 1993 of dissipating decay heat in natural-convection mode. The information required for this chapter should describe the complete coolant systems for the allowed modes of operation, as discussed below.
5.2 Summary Description This section should present a brief description of the reactor coolant systems, summarizing the principal features.
Information should include type of primary coolant:
liquid, gas, solid (conduction to surrounding structures) whether the primary system is open to the atmosphere er closed type of coolant flow in the primary system:
force'd-convection, natural-convection, or both type of secondary system, if one is provide'd, and the method of heat disposal to the environment whether sufficient heat removal capability is provided to support continuous operation at full licensed pow w special or facility-unique features 5.3 Primary Coolant System This section should provide information on the primary coolant system. The basic requirements and the design bases of the primary coolant system are to maintain reactor facility conditions within the range of design conditions and accident analyses assumptions as derived from other chapters of the SAR, especially Chapters 4 and 13. The applicant should show the interrelationships among all SAR chapters and how the designed primary coolant system provides all necessary functions. The following information should be included:
(1)
Description of the design bases and functional requirements of the primary coolant system.
(2)
Schematic and flow diagrams of the system, showing such essential components as the heat source (reactor core), heat sink (heat exchanger), pumps, piping, valves, control and safety instrumentation, interlocks, and other related subsystems.
(3)
Tables of allowable ranges of important design and operating parameters and specifications of the primary system and its components, including coolant material coolant flow rates inlet and outlet temperatures and pressures throughout the system 2
i SAR Standard Format and Content Chapter 5 Draft June 22, 1993 i
elevati~on of components and water levels relative to the reactor core construction materials of components fabrication specifications of safety-related components a
coolant quality requirements for operation and shutdown conditions, including pH and conductivity as a minimum minimum coolant level (4)
Discussions and analyses keyed to drawings showing how the system provides the necessary cooling for all heat loads and all potential reactor conditions analyzed in the thermal-hydraulics section of Chapters 4 and 13 including Removal of heat from the fuel by forced-convection or natural-convection cooling, or both for those reactors licensed to operate in both modes. Discussion and analyses of how the size, shape, and structural features of the primary vessel or pool affect cooling characteristics, the function of the pool as a heat reservoir, and the effect of water depth on natural thermal convection cooling.
Transfer of heat from the primary coolant to a secondary system for all reactor conditions.
This discussion should include any heat exchanger design and operating conditions.
Some non-power reactors may contain only a primary system that functions as a heat reservoir.
For such systems, the analyses should include any factors that limit continuous operation, such as pool water temperature, and the proposed technical specifications that ensure operation within the analyzed limits. Some high power non-power reactors may require shutdown pumps to circulate coolant through the core to remove decay heat.
Safe reactor shutdown, including passive or fail safe transition from forced-to natural-convection cooling and removal of decay heat from the fuel. This discussion should include the loss of offsite electrical power.
Locations, designs, and functions of such essential components as drains, syphon-breaks, pumps, isolation valves, and check valves.
These components ensure that the primary system is operable and that uncontrolled loss or discharge of contaminated coolant from the primary system does not occur.
Radiological effects of potential coolant releases should primarily be analyzed in Chapter 11, " Radiological Waste Management and Radiation Protection."
(5)
Discussion of the control and safety instrumentation, including location and functions of sensors and readout devices. The scram or interlock functions that prevent exceeding safety limits should be shown and discussed, including the related technical specifications.
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SAR Standard Format and Content Chapter 5 - Draft.
June 22, 1993 3
(6)
Description and function of any special features of the primary coolant system, such as removal of the neutron moderator for backup reactor shutdown.
(7)
Brief description and functions of special features or components of the primary system that affect or limit personnel radiation exposures from such radionuclides as nitrogen-16 and argon-41 and from radioactive contaminants and fission products.
(8)
Description of radiation monitors or detectors incorporated into the primary system and discussion of their functions.
(9)
Brief discussion and reference to detailed discussions in later sections for auxiliary systems using primary coolant, such as water cleanup, water makeup, nitrogen-16 control, emer.gency core cooling, experiment cooling, experimental facility cooling, and biological or thermal shield cooling. The direct effect on the design and functioning of the primary system by these auxiliary systems should be discussed.
(10) Brief discussion of radiation shielding provided by the primary coolant.
Most non-power reactors are submerged in a pool or tank so that the primary coolant shields personnel above the pool and at the reactor room floor. The design bases for these shielding functions should be analyzed for anticipated reactor conditions in Chapters 4 and 11 and for postulated reactor accidents in Chapter 13. The effect of any special shielding features, such as fuel storage facility shielding and experimental facility shielding (e.g., beam tubes), on the functioning of the primary coolant system should be discussed.
(11) Discussion of leak detection and allowable leakage limits, if any.
Prompt detection of leakage is very important in reactors with heavy water.
(12) Discussion of normal primary coolant radiation concentration limits including sampling frequency, isotopes of interest, and actions to be taken if limits are exceeded.
(13)
For reactors that have closed systems, a discussion of allowable hydrogen limits in air spaces that are in contact with the primary coolant.
(14) Discussion of technical specification requirements for parameters of the primary system, including the bases, and surveillance requirements.
5.4 Secondary Coolant System Many licensed non-power reactors include fuel cooling systems composed of both primary and secondary coolant systems. Some very low-powered reactors contain no engineered coolant system, and some may contain a single compone-t (primary) cooling system.
Still other low-powered reactors may include a coolant system, with both primary and secondary subsystems, that would not support continuous reactor operation at full licensed power.
4
SAR Standard Format and Content Chapter 5 - Draft June 22, 1993 This section should provide information about those non-power reactors that include a secondury coolant system.
For the others, this section should state that a secondary coolant system is not needed and justify that conclusion.
The following information should be provided:
(1)
Description of the design bases and functional requirements of the system, including whether the system is designed for continuous full-power reactor operation and whether the secondary system is shared with other facilities.
i i
(2)
Schematic and flow diagrams of the secondary system, showing such essentials as how the heat exchanger connects the primary system (the i
heat source) to the secondary system, pumps, piping, valves, control and safety instrumentation, interlocks, and the interface with the environment for ultimate release of the heat.
(3)
Tables of the range of important design and operating parameters and specifications of the secondar' system, including coolant material and its source a
coolant flow rates type of heat dissipation system, such as cooling tower, refrigerator, radiator, or body of water location of heat dissipation system in relation to the reactor and the heat exchanger construction materials and fabrication specifications of components heat dissipation specifications related to environmental factors specific i.ivna and limitations on coolant quality and its effect on environmental chemical contamination factors and corrosion of the secondary system components (4)
Discussion and functional analyses keyed to the drawings showing how the system provides the necessary cooling for all potential reactor l
conditions. These discussions should address inlet and outlet temperatures and pressures throughout the system, e
including the pressure differential between primary and secondary 4
systems in the heat exchanger (Discuss how the pressure in the secondary system is maintained above that in the primary for all operating conditions, or analyze the radiological effect of leakage of contaminated primary coolant into the secondary system.
If the transfer of primary coolant into the secondary system is caused by an abrupt event, such as a tube rupture in the heat exchanger, the analysis should be located in Chapter 13 and summarized here.)
5
SAR Standard Format and Content Chapter 5 - Draft June 22, 1993 control of heat removal from the secondary system necessary to maintain fuel temperatures in the primary system within the limits derived in the thermal-hydraulics analyses in Chapters 4 and 13 removal of heat from the heat exchanger and release to the environment when the primary, system operates in all anticipated and licensed modes, including forced-convection flow and natural-convection flow, as applicable safe reactor shutdown, and removal and dissipation of decay heat.
Includes evaluation of the primary system change from forced-convection flow to natural-convection flow if forced-convection flow is an allowed mode of operation i
response of the seconda.ry system to the loss of primary coolant with or without an eme'rgency core cooling system locations, designs, and functions of such essential components as drains, sumps, pumps, makeup water, and check valves that ensure contaminated primary coolant is not inadvertently transferred to the secondary system and released to the environment.
4 (5)
Discussion of control and safety instrumentation, including locations and functions of sensors and readout devices and interlocks or safety cap >hilities.
(6)
Desci iptions of functions of any radiation monitors or detectors incorporated into the secondary system. Discussion of surveillance to measure secondary coolant activity including frequency, action levels, and action to be taken.
(7)
Brief comments and reference to detailed discussion in other sections for auxiliary cooling systems that transfer heat to the secondary system, such as emergency core culing system, experiment cooling, or biological or thermal shield cooling.
(8)
Discussion of technical' specification requirements for the secondary system, including the bases, and surveillance requirements.
5.5 Primary Coolant Cleanuo System Most licensed non-power reactors contain solid fuel elements immersed in the primary coolant water.
Experience has shown that the metal cladding is susceptible to corrosion if the chemical purity of the water is not high. The water purity must be above that usually available in the potable water supply.
Experience has shown that oxide buildup on aluminum-clad fuel operating at high power densities can reduce heat transfer. The rate of buildup depends, j
among other factors, on the water quality (J. C. Griess et al.,1964). This process should be evaluated in Chapter 4 and summarizad in this chapter if it contributes to determining the requirements for the primary coolant purity.
To delay or prevent component failure by corrosion, non-power reactors should provide a primary coolant cleanup system. The purity of the primary coolant should be maintained as high as reasonably possible to 6
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SAR Standard Format and Content Chapter 5 - Draft June 22, 1993 4
limit the chemical corrosion of fuel cladding, control and safety rod cladding, the reactor vessel or pool, and other essential components in the primary coolant system limit the concentrations of particulate and dissolved contaminants that could be made radioactive by neutron irradiation maintain high transparency of the water for observation of submerged components The following information should be provided:
(1)
Description of the design bases and functional requirements of the primary coolant cleanup system.
Experience at non-power reactors has shown that with a well-planned water cleanup system and good housekeeping practices, primary coolant quality can be maintained within the following ranges:
electrical conductivity $5 pmho/cm pH between 5.5 and 7.5 The design bases should be derived from discussions in Chapter 4 and any recommendations of the fuel vendor also should be addressed.
(2)
Schematic drawings and flow diagrams of the primary coolant cleanup loop.
(3)
Table of specifications for the cleanup system demonstrating that it is designed for the volume and throughput of the primary coolant system.
(4) locations and functions of control and monitoring instrumentation, including sensors, recorders, and meters.
The discussion of monitors should include methods for continuously assessing coolant quality and effectiveness of the cleanup system.
(5) locations and functional designs of cleanup system components such as branch points, pumps, valves, filters, and demineralizers.
(6)
Discussion of schedules and methods for replacing or regenerating resins and filters and disposal of resultant radioactivity to ensure that radiation exposures do not exceed the limits discussed in Chapter 11.
(7)
Summary of methods for predicting, monitoring, and shielding radioactivity deposited in filters and demineralizers from routine operations. The detailed discussion should be in Chapter 11.
(8)
Summary of methods for predicting and limiting exposures of personnel in the event of inadvertent release of excess radioactivity in the primiry system and deposition in filters and demineralizers. The detailed discussion should be in Chapter 13.
(9)
Provisions in the design and operation of the cleanup system to avoid malfunctions that could lead to significant loss of primary coolant or 7
SAR Standard Format and Content Chapter 5 - Draft-1 June 22, 1993 release of contaminated coolant, which could cause radiological exposure of personnel or the unrestricted environment to exceed 10 CFR 20 and the facility ALARA program guidelines.
l (10) Discussion of technical specification requirements for the primary coolant cleanup system; including the bases, and surveillance requirements.
5.6 Primary Coolant Makeup Water System i
During operations at non-power reactors with a water cooled primary system, primary coolant must be replaced or replenished (i.e., makeup). Coolant may i
i be lost as a result of evaporation from open-pool systems, radiolysis, designed leakage as from pump seals, and other operational activities. -The non-power reactor design should include a system or a procedure that meets the projected coolant needs. The makeup system need not provide a total rapid replacement of the primary coolant inventory, but should be able to maintain the minimum acceptable water quantity and quality for reactor operation.
l The following information should be provided:
(1)
Description of the design bases for primary water makeup that account for all activities that could cause a decrease in the primary coolant.
l A large loss of coolant event should be analyzed in Chapter 13.
Although a required emergency core cooling system need not be a part of i
the makeup system, it should be discussed in Chapter 6 if it exists.
(2)
Schematic diagrams and functional discussions that show the source of water, the methods of addition to the primary system, and the requirements for pretreatment before addition. Not all non-power j
reactors need a makeup system attached to the reactor primary system.
l (3)
Locations and functions of control instrumentation, including sensors, readout displays, and interlocks. Methods should be discussed for tracking makeup water additions to detect significant changes that might indicate leaks or other malfunction of the primary coolant system.
I (4)
Discussion of safety systems and administrative controls to ensure that i
the system or procedures for adding makeup water will not lead to j
significant loss of primary coolant and will prevent leakage of contaminated coolant into the potable water supply, i
(5)
Discussion of technical specification requirements for the primary coolant makeup system, including the bases, and surveillance requirements.
5.7 Nitrocen-16 Control Sys_tg When ordinary oxygen is irradiated with neutrons of sufficient energy, Nitrogen-16 (N-16), a high-energy beta and gamma emitter with a 7-second half-life, is formed.
In water-cooled reactors operated above a few hundred kilowatts, the radioactivity of this nuclide may require specific systems or procedures for limiting personnel exposure.
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1 SAR Standard Format and Content Chapter 5 - Draft June 22, 1993 s
In reactors with natural-convection cooling, transport of N-16 to the pool surface may be delayed by a coolant circulator or diffuser.
In reactors with forced-convection cooling, the coolant carrying the N-16 out of the core may j
be passed through a system such as a large shielded and baffled tank.
This delay allows the radioactivity to decay significantly before the coolant i
emerges from the shielding. Another method of radiation control is to shield the entire primary coolant system.
l The potential for personnel exposure to N-16 should be analyzed and control l
systems or procedures should be proposed that include the following:
l (1)
Description of the design bases and functional design of the N-16 l
control system or procedures. The design bases are derived from I
analyses in Chapter 11.
(2)
Schematic drawings and system and component specifications for the N-16 l
control system.
l l
(3)
The method used by the N-16 control system to reduce exposure rates and potential doses in occupied areas.
Potential doses with and without the l
N-16 controls should be analyzed in Chapter 11 and summarized here.
These potential doses include dose from direct radiation and dose from airborne N-16.
(4)
Description of the effect of the N-16 control system on overall reactor safety and operation. For example, diffuser systems in natural convection reactors may affect coolant flow and coolant transparency.
(5)
Description of other reactor design features affected by the N-16 control system.
For example, a large shielded &ay tank may affect coolant flow parameters, pump sizes., access for arveillance or inservice testing, or other factors in the primary coolant system.
(6)
An assessment that the N-16 control systein would not lead to an uncontrolled loss of primary coolant or the release of contaminated primary coolant to exceed 10 CFR 20 and the facility ALARA program guidelines. Methods for analyzing radiation exposures as a result of coolant release should be consistent with analyses in Chapter 11.
(7)
Discussion of any technical specification requirements for the N-16 control system, including the bases, and surveillance requirements.
5.8 Auxiliary Systems Usino Primary Coolant In addition to the systems discussed above that are associated with the primary coolant system, some non-power reactors employ other auxiliary cooling systems or shields that may require the use of primary coolant and may affect the operation or safety of the reactor. Any function of the primary coolant that is principally shielding should be described in Chapter 4 and summarized here and in Chapter 11.
If the reactor design includes an emergency core cooling system, that should be described and discussed in Chapter 6.
The following auxiliary systems that use primary coolant should be discussed in this section:
9
i i
SAR Standard Format and Content Chapter 5 - Draft
- i June 22, 1993 experiment cooling experimental facility cooling experimental facility shielding (e.g., beam tubes) biological shield cooling thermal shield cooling i
fuel storage cooling and shielding l
The following information about these systems should be provided in this section:
i (1)
Description of the design bases and functional requirements of the auxiliary systems based on discussions elsewhere in the SAR, such as Chapters 4, 9, and lo, " Experimental Facilities and Utilization."
(2)
Schematic drawings and flow diagrams that show the source of water, i
locations of sensors and instruments, and locations of the components cooled or shielded.
(3)
Tables of the range of important parameters of the systems and l
specifications of materials and components.
i (4)
Discussion of components to be cooled, the source of heat, the source of the coolant water, heat transfer to the coolant, and coolant heat l
dissipation.
l (5)
Summary of the shielding requirements and the protection factors provided by the coolant.
L i
(6)
Discussion of the provisions in the auxiliary system designs to prevent interference with safe reactor shutdown.
L i
(7)
Discussion of the provisions in the auxiliary system design to prevent the uncontrolled release of primary coolant or radiation exposures that would exceed 10 CFR 20 and the facility ALARA program guidelines.
a (8)
Requirements for minimum water quality.
j (9)
Discussion of any technical specification requirements for the auxiliary cooling system, including the bases, and surveillance requirements.
REFERENCE l
Griess, J. C., et al., Effect of Heat Flux on the Corrosion of Aluminum by I
Water. Part IV, ORNL-3541, February 1964.
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i SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22. 1993
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5.1 Introduction This chapter provides the design bases, descriptions, and functional analyses of the reactor coolant systems.
The principal purpose of the coolant system is to safely remove the fission and decay heat from the fuel and dissipate it to the environment.
H.twever, the coolant in the primary systems of most non-power reactors serves more functions than just efficient removal of heat.
The coolant can act as rr.diation shielding for the reactor, fuel storage facilities, and in some designs, experimental facilities and experiments.
In open pool reactors, the coolant is the only vertical shielding.
In many designs the reactor coolant also acts as a core moderator and reflector.
Because of these many functions that the reactor coolant serves, the reactor coolant system design is based on selecting among interdependent parameters, including thermal power level, research capability, available fuel type, reactor core physics requirements, and radiation shielding.
The principal licensing basis of non-power reactors is the thermal power developed in the core during operation. This basis also applies to the few non-power reactors licensed to operate at such low-power levels that no significant core temperature increases would occur during normal operation.
Such reactors may not require an engineered coolant system.
For those reactors, the analysis in Chapter 4, " Reactor Description," should discuss the dissipation of the heat produced, estimate potential temperature increases during reactor operation, and justify why an engineered coolant system is not required. This chapter should summarize those considerations and conclusions.
j For all other non-power reactors, systems to remove and dispose of the waste heat should be described and discussed in this chspter. The design bases of the core cooling systems for normal and off-normal operation will be derived in Chapter 4.
All auxiliary sys',ms and subsystems that use and contribute to
{
l the heat load of either the prir y or secondary coolant system should also be described and discussed in this chapter. Any auxiliary systems using coolant j
from other sources, such as building service water, should be discussed in Chapter 9, " Auxiliary Systems." The design bases of any features of the core cooling system designed to respond to potential accidents or to mitigate the i
consequences of potential accidents should be derived from the analyses in Chapter 13, " Accidents." These features should be summarized in this chapter and discussed in detail in Chapter 6, " Engineered Safety Features." This chapter of the SAR should discuss and reference the technical specifications that are needed to ensure operability consistent with SAR analyses assumptions.
l The primary loops of the coolant systems of most licensed non-power reactors l
are of two basic types, forced-convection and natural thermal-convection.
l Facilities using forced-convection cooling also may be licensed to operate in the natural-convection mode and should be capable of dissipating decay heat in that mode.
This chapter provides the review and acceptance criteria for information concerning the heat removal systems. The information suggested for this l
section of the SAR is outlined in Chapter 5 of the format and content guide.
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SAR Revicw Plan and Acceptance Criteria Chapter 5 - Draft
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June 22, 1993 e
5.2 Summary Description A summary of the principal features of the reactor cooling systems should be given, including type of primary coolant: liquid, gas, or solid (conduction to surrounding structures) whether the system is open or closed to the atmosphere type of flow in the primary system:
forced convection, natural e
convection, or both type of secondary system, if one is present, and the method of heat disposal to the environment capability to provide sufficient heat removal to support operation at full licensed power special or facility-unique features 5.3 Primary Coolant System 5.3.1 Areas of Review As noted above, non-power reactor design requires choosing among several interdependent variables. Usually, the design represents a compromise between the neutron flux densities required and the need to dissipate thermal power.
The final design depends on the intended uses of the reactor, the resources available, and the priorities of the facility owner. The objective of the final design is ensured safety. The primary coolant system is a key component in the design and should have the capability to remove the fission heat and decay from the fuel during reactor operation and decay heat during reactor shutdown for most non-power reactors, transfer the heat to a secondary system for controlled dissipation to the environment maintain high water quality to limit corrosion of fuel cladding, control and safety rods, the reactor vessel or pool, and other essential components provide radiation shielding of the core and other components such as e
beam tubes and fuel storage facilities provide neutron moderation and reflection in the core prevent uncontrolled leakage or discharge of contaminated coolant to the unrestricted environment The basic requirements for these functions are generally derived and analyzed in other chapters of the SAR. This chapter should describe how the coolarit 2
SAR Review Plan and Acceptance Criteria i
Chapter 5 - Draft l..
June 22, 1993 l
system provides these functions. Specific areas for review for this section I
are discussed in section 5.3 of the format and content guide.
l 5.3.2 Acceptance Criteria Acceptance criteria for the information concerning the primary coolant system are based on the following:
(1)
Chapter 4 analyzes the reactor core including coolant parameters necessary to ensure fuel integrity.
Safety limits (SLs) and limiting safety system settings (LSSSs) to ensure fuel integrity will be derived from those analyses and included in the technical specifications.
Examples of cooling system variables on which LSSSs may be established I
are maximum thermal power level for operation in natural-convection flow, maximum coolant temperature, minimum coolant flow rate, minimum pressure of coolant at core, and minimum pool depth above the core.
To i
be acceptable, the analyses in Chapter 5 should show that the components and the functional design of the primary system will ensure that no LSSS will be exceeded through the normal range of reactor operation. The analyses should address forced flow or natural-convection flow, or both for reactors licensed for both modes.
The design must show that the I
passive or fail safe transition from forced flow to natural-convection
)
flow is reasonably ensured in all forced-flow non-power reactors and that continued fuel integrity is ensured.
The functional design should show that safe reactor shutdown and decay heat removal is sufficient to ensure fuel integrity for all possible reactor conditions, including potential accident scenarios. Scenarios that postulate loss of flow or loss of coolant should be analyzed in Chapter 13 and the results summarized in Chapter 5.
(2)
The descriptions and discussions should show that sufficient instrumentation, coolant parameter sensors, and control systems are provided to monitor and ensure stable coolant flow, respond to changes in reactor power levels, and provide for a rapid reactor shutdown in the event of loss of cooling. There also should be instrumentation for monitoring the radiation of the primary coolant because this could indicate a loss of fuel clad integrity. There should be routine sampling for gross radioactivity in the coolant and less frequent radioactive spectrum analysis to identify the isotopes and concentrations found in the coolant. This spectrum analysis may also detect cladding failure at its earliest stages.
(3)
The primary coolant must provide a chemical environment that limits corrosion of fuel cladding, control and safety rod surf aces, reactor vessels or pools, and other essential components. Aluminum-clad fuel operated at high power density will develop an oxide coating that could decrease heat conductivity (J. C. Griess et al., 1964). Chapter 4 of the SAR should provide discussion and analyses of the dependence of oxide formation on water quality and other factors. Other requirements for water purity should be analyzed in the SAR, and proposed values of conductivity and pH justified.
Experience at non-power reactors has shown that the primary water conditions, electrical conductivity $5 3
SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 199S '
I gmho/cm and pH between 5.5 and 7.5, can usually be attained with good housekeeping and a good filter and demineralizer system. The proposed system will be acceptable if chemical conditions can be maintained, as discussed in the water cleanup section.
(4)
Most non-power reactors consist of a core submerged in a pool or tank of water. The water helps shield personnel in the reactor room and the unrestricted area from core neutrons and gamma rays. The water also decreases potential neutron activation and radiation damage to such reactor components as the pool liner, beam port gaskets, in-pool lead shields, and the concrete biological shield.
The analyses in Chapter 4 discuss these factors; the design of the primary coolant vessel will be acceptable if exposure limits on materials discussed in Chapter 4 are not exceeded, and exposures to personnel, as discussed in Chapter 11, do not exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(5)
Radioactive species may be produced in the primary coolant, including nitrogen-16 (N-16) and argon-41 (Ar-41). Additional radioactivity may occur as a result of neutron activation of coolant contaminants and fission product leakage from the fuel.
Provisions for limiting personnel radiological hazards will be acceptable if they maintain potential exposures from coolant radioactivity below the requirements of 10 CFR 20 and are consistent with the facility ALARA program. Facilities or components for controlling, shielding, or isolating N-16 are acceptable if potential exposures do not exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program. N-16 is discussed in Section 5.7 below.
Ar-41 is a ubiquitous radionuclide produced at non-power reactors.
Because this radionuclide may be the major release to the environment during normal operation, special analyses and discussion of its production and consequences should be provided in Chapter 11.
If any special design or operational features of the primary coolant system are provided to modify or limit exposures from Ar-41, they should be discussed in Chapter 5.
This discussion will be acceptable if it demonstrates that any facilities or components added to the primary system to modify Ar-41 releases can limit potential personnel exposures to the values found acceptable in Chapter 11.
Closed systems also may experience a buildup of hydrogen in air spaces in contact with the coolant.
The discussion should show that it is not possible to have hydrogen build up to concentrations that are combustible.
This may require gas sweep systems and hydrogen concentration monitoring. These systems should be discussed in Chapter 9.
(6)
Because the primary coolant may provide essential fuel cooling and radiation shielding, the system design should avoid uncontrolled release or loss of coolant. Some design features to limit losses include locating components of the primary system above the core level, avoiding drains or valves below core level in the pool or tank, providing syphon-breaks in piping that enters the primary vessel or poo), and providing 4
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SAR Review Plan and Acceptance Criteria j
Chapter 5 - Draft l
June 22, 1993
)s check valves to preclude backflow.
The designs and locations of such features will be acceptable if there is reasonable assurance that loss of coolant that could uncover the core is very unlikely. A potential accident of rapid loss of coolant should be analyzed in Chapter 13 and summarized in Chapter 5.
Heavy water systems require additional design features because of the l
radiological hazards of tritium production in the coolant. Heavy water systems should be designed with systems to detect minor leakage. The systems also should be designed so that heavy water, if lost from the system, will be contained and not released to the environment.
i If contaminated coolant were lost from the primary system, the design and analyses would be acceptable if potential personnel exposures and uncontrolled releases to the unrestricted environment would not exceed acceptable radiological dose consequence limits from the accident analyses. The radiological consequences from the contaminated coolant should be discussed in Chapter 11 and summarized in Chapter 5.
Necessary surveillance provisions should be included in the technical specifications.
I (7)
The discussions should identify operational limits, design parameters and surveillances to be included in the technical specifications.
5.3.3 Review Procedures 1
The staff will compare the functional design and the operating characteristics t
of the primary coolant system with the bases for design presented in this and other relevant chapters of the SAR. The system design should meet the acceptance criteria presented above.
5.3.4 Evaluation Findings Chapter 5 should contain sufficient information to support the following types of conclusions, which will be included in the staff's safety evaluation report:
(1)
The primary coolant system is designed in accordance with the design bases derived from all relevant analyses in the SAR.
(2)
Design features of the primary coolant system and components give reasonable assurance of fuel integrity under all possible reactor conditions. The system is designed to remove sufficient fission heat from the fuel to allow all licensed operations without exceeding the l
l established LSSS that are included in the technical specifications.
(3)
Designs and locations of primary system componente have been specifically selected to avoid coolant loss that could lead to fuel failure, uncontrolled release of excessive radioactivity, or damage to safety systems or experiments. Heavy water systems are designed to quickly detect leakage and prevent the release of heavy water to the environment.
If an emergency core cooling system is required to prevent 5
l l
SAR R: view Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 r
a loss of fuel integrity, it is evaluated in connection with the review of the engineered safety features.
(4)
For reactors licensed to operate with forced-convection coolant flow, the system is designed to convert in a passive or fail-safe method to i
natural-convection flow sufficient to avoid loss of fuel integrity.
This feature is evaluated in conjunction with the reviews of the reactor description and accidents.
t l
(5)
The chemical quality of the primary coolant will limit corrosion of fuel I
cladding (or other primary barrier to release of fission products),
control and safety rod cladding, the inside of the reactor vessel (or pool), and other essential components in the primary system for the duration of the license and for the projected utilization time of the fuel. For a closed primary system, systems are present that will prevent hydrogen concentrations from reaching combustible limits.
(6)
The size and shape of the primary vessel or pool will provide sufficient l
radiation shielding to maintain personnel exposures beloit limits in 10 CFR Part 20 and will provide a heat reservoir sufficient for anticipated reactor cperations.
l (7)
Primary coolant system instrumentation and controls are designed to provide all necessary functions and to transmit information on the operating status to the control room.
(8)
The technical specifications, including testing and surveillance, provide reasonable assurance of necessary primary system operability for reactor operations as analyzed in the SAR.
(9)
The design bases of the primary coolant system provides reasonable assurance that the environment and the health and safety of the public i
will be protected.
5.4 Secondary Coolant System 5.4.1 Areas of Review Secondary coolant systems of non-power reactors are designed to transfer reactor heat from the primary system to the environment. Non-power reactors may be designed in three ways: with a continuously operating secondary ceolant system, with an on-demand secondary coolant system, and without a secondary coolant system.
For most of the licensed reactors, the coolant systems are designed for continuous operation at licensed power level.
Therefore, the secondary system in these reactors must be designed to dissipate the heat continuously. Some non-power reactors are designed and licensed to operate at low power levels, or for limited time intervals, so that the primary system can absorb or dissipate the heat without a continuously operating secondary system.
Some non-power reactors require no secondary cooling system. This section should justify how any necessary heat dissipation is accomplished. Specific areas for review for this section are discussed in Section 5.4 of the format and content guide.
6
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SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993
!s 5.4.2 Acceptance Criteria l
Acceptance criteria for the information in Chapter 5 on the secondary coolant system are based on the following:
(1)
The required operating characteristics of the primary coolant system are presented in Section 5.3.
To be acceptable, the analyses and i
discussions of Section 5.4 should demonstrate that the secondary system is designed to allow the primary system to transfer the heat as necessary to ensure fuel integrity. The analyses should address primary systems operating with forced flow, natural-convection flow, and both for reactors licensed for both modes.
The design must show that the secondary system is capable of dissipating all necessary fission and I
decay heat for all potential reactor conditions as analyzed in the SAR.
(2)
Some non-power reactors are designed with secondary systems that will not support continuous reactor operation at full licensed power. This is acceptable, provided the capability and such limiting conditions as maximum pool temperature are analyzed in the SAR and included in the technical specifications.
(3)
The primary coolant will usually contain radioactive contamination. The design of the total coolant system should ensure that release of such radioactivity through the secondary system to the unrestricted environment would not lead to potential exposures of the public in excess of the requirements of 10 CFR Part 20 and are consistent with the ALARA program. Designs would be acceptable in which the primary system pressure is lower than the secondary system pressure across the heat exchanger under all anticipated conditions, the secondary system is closed, or radiation monitoring and effective remedial capability are provided. An acceptable design of the secondary system must prevent or acceptably mitigate uncontrolled release of radioactivity to the unrestricted environment.
Periodic samples of secondary coolant should be analyzed for radiation. Action levels and required actions should be discussed.
l l
(4)
An acceptable secondary system design should accommodate any heat load required of it in the event of a potential engineered safety feature operation or accident conditions as analyzed in Chapters 6 and 13.
(5)
An acceptable secondary system design should provide for any necessary l
chemical control to limit corrosion or other degradation of the heat exchanger and prevent chemical contamination of the environment.
(6)
The discussions should identify operational limits, design parameters, l
and surveillances to be included in the technical specifications.
5.4.3 Review Procedures The staff will verify that all reactor conditions, including postulated accidents, requiring transfer of heat from the primary coolant system to the secondary coolant system have been discussed. The review will verify (1) the secondary coolant system design is capable of removing and dissipating the 7
1 SAR Review Plan and Acceptance Criteria Chapter 5 - Draft I
June 22, 1993 i
/
amount of heat and the thermal power necessary to ensure fuel integrity and (2) analyses of secondary coolant system malfunctions and effects on reactor safety, fuel integrity, and the health and safety of the public.
5.4.4 Evaluation Findings f
Chapter 5 should contain sufficient information to support the following types of conclusions, which will be included in the staff's safety evaluation report:
(1)
Design features of the secondary coolant system and components will allow the transfer from the primary system the necessary reactor heat under all possible reactor conditions.
(2)
Locations and design specifications for secondary coolant system components ensure that malfunctions in the system will not lead to reactor damage, fuel failure, or uncontrolled release of radioactivity to the environment.
(3)
Secondary coolant system instrumentation and controls are designed to I
provide all necessary functions and to transmit information on the i
operating status to the control room.
(4)
The secondary coolant system is designed to respond as necessary to such postulated events as a loss of primary coolant accident and a loss of forced coolant flow in the primary system.
l (5)
The technical specifications, including testing and surveillance, provide reasonable assurance of necessary secondary system operability for normal reactor operations.
5.5 Primary Coolant cleanuo System t
5.5.1 Areas of Review Experience has shown that potable water supplies are usually not acceptably pure for use as a reactor primary coolant without additional cleanup. Most licensed non-power reactors contain solid fuel elements immersed in the primary coolant water.
Experience has also shown that oxide buildup on aluminum-clad fuel operated at high power densities can reduce heat transfer l
(J. C. Griess et al.,1964). The rate of buildup depends upon several operational characteristics, including the pH of the coolant. Therefore, a discussion of this process should be included in Chapter 4 and summarized in Chapter 5 if it contributes to establishing requirements for the primary coolant purity. The water purity should be maintained as high as possible for the following reasons:
to limit the chemical corrosion of fuel cladding, control and safety rod cladding, the reactor vessel or pool, and other essential components in the primary coolant system to limit the concentrations of particulate and dissolved contaminants that might become radioactive by neutron irradiation 8
SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 s
to maintain high transparency of the water for observation of submerged operational and utilization components Specific areas for review for this section are discussed in Section 5.5 of the format and content guide.
5.5.2 Acceptance Criteria The acceptance criteria for the primary coolant cleanup system are based on the following:
(1)
The functional design of the cleanup system is acceptable if the primary coolant quality can be maintained in the ranges established as acceptable in Chapters 4 and 11 of the SAR. These analyses for high power reactors (greater than 2 MW) should include the buildup of an oxide film on aluminum cladding.
Experience has shown that quality water conditions, electrical conductivity $5 pmho/cm and pH between 5.5 and 7.5, can usually be achieved by good housekeeping and a cleanup loop with particulate filters and demineralizers.
Such a system is acceptable unless the SAR analyses establish other purity conditions as acceptable.
(2)
Disposal or regeneration of radioactively contaminated resins and filters is acceptable if accomplished in accordance with radiological waste management plans discussed in Chapter 11 and if potential exposures and rcleases to the unrestricted environment do not exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(3)
Location, shielding, and radiation monitoring of the water cleanup system are acceptable for routine operations and potential accidental events if the occupational staff and the public are protected from radiation exposures exceeding the requirements of 10 CFR Part 20 and acceptable radiological consequence dose limits for accidents.
(4)
Location and functional design of the components of the water cleanup system are acceptable if malfunctions or leaks in the system would not cause uncontrolled loss or release of primary coolant, personnel exposure, and/or release of radioactivity would not exceed the requirement of 10 CFR Part 20, are consistent with the facility ALARA program, and safe reactor shutdown would not be prevented.
(5)
The discussions should identify operational limits, design parameters, and surveillances to be included in the technical specifications.
5.5.3 Review Procedures The staff will compare the design bases for the primary water quality and the design bases by which the primary coolant cleanup system will achieve the requirements. The comparison will include performance specifications, the schematic diagrams, and the discussion of the functional characteristics of the cleanup system. The staff will review (1) design features to ensure that leaks or other malfunctions will not cause inadvertent damage to the reactor 9
l SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 '
/ :
or exposure of persoanel and (2) the plan for control and disposal of radioactive filters and demineralizer resins.
5.5.4 Evaluaticn Findings Chapter 5 should contain sufficient information to support the following types of conclusions, which will be included in the staff's safety evaluation report:
(1)
The design bases and functional descriptions of the primary water cleanup system give reasonable assurance that the required water quality can be achieved. The design will ensure that corrosion and oxide buildup of fuel cladding and other essential components in the primary coolant system will not exceed the acceptable limits or the recommendations of the fuel vendor.
(2)
Experience has shown that the pH of the primary coolant can influence the rate of oxide buildup on aluminum-clad fuel. The pH and the proposed system are consistent with the analysis for the effect of oxide on heat transfer from the fuel.
(3)
The primary coolant cleanup system and its components have been designed and selected so that malfunctions are unlikely. Any malfunctions or leaks will not lead to radiation exposure to personnel or releases to the environment that exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(4)
The plans for controlling and disposing of radioactivity accumulated in components of the primary water cleanup system, which results from normal operations and potential accident scenarios, comply with applicable regulations, including 10 CFR Part 20 and acceptable radiological consequence dose limits for accidents.
(5)
The technical specifications, including testing and surveillance, provide reasonable assurance of necessary primary water cleanup system i
operability for normal reactor operations.
l 5.6 Primary Coolant Makeuo Water System 5.6.1 Areas of Review During operations at non-power reactors, primary coolant must be replaced or replenished (i.e., makeup). Coolant may be lost through evaporation in open-pool systems, radiolysis, leaks from the system, and other operational activities. Although each non-power reactor should have a makeup water system or procedure to meet projected operational needs, the makeup system need not be designed to provide a rapid total replacement of the primary coolant i
inventory. Specific areas for review for this section are discussed in i
Section 5.6 of the format and content guide.
10 j
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5 SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 r
5.6.2 Acceptance Criteria The acceptance criteria for the primary water makeup system are based on the l
i following:
l (1)
The projected loss of primary water inventory for anticipated reactor operations should be discussed. An acceptable design or plan for supplying makeup water will ensure that those operational requirements i
are satisfied.
(2)
If storage of treated makeup water is required by the design bases of 4
the primary coolant system, an acceptable makeup water plan or system will provide such water.
i l
(3)
Not all non-power reactors must provide nakeup water through hardware j
systems directly connecting the reactor to the facility potable water 1
supply. However, for those that do, an acceptable n?akeup system or plan j
must provide components or administrative controls that prevent i
potentially contaminated primary coolant from entering the potable water system.
i (4)
An acceptable water makeup system or plan will. include features to
}
prevent loss or release of coolant from the primary coolant system.
1 (5)
The primary water makeup system need not have a functional relationship with any installed emergency core cooling system (ECCS).
If it does, an acceptable design of the water makeup system will not interfere with the availability and operability of the ECCS.
(6)
An acceptable water makeup system or plan will provide a record of the use of makeup water to detect changes that indicate leakage or other malfunction of the primary coolant system.
(7)
The discussions should identify operational limits, design parameters and surveillances to be included in the technical specifications.
4 5.6.3 Review Procedures The staff will compare the design bases and functional requirements for replenishing primary coolant including the quantity and quality of water, the activities or functions that remove primary coolant, systems or procedures to accomplish the water makeup and the safety precautions to preclude overfilling of the reactor coolant system and subsequent loss through the drain system and release of primary coolant back through the makeup system into potable water supplies.
t 5.6.4 Evaluation Findings Chapter 5 should contain sufficient information to support the following types of conclusions, which will be included in the staff's safety evaluation report:
11
l SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 h
?
(1)
The design bases, functional descriptions, and procedures for the i'
primary water makeup systems give reasonable assurance that the quantity and quality of water required will be provided, j
(2)
The system design or procedures will prevent overfilling of the primary system or malfunction of the makeup system and will prevent loss or release of contaminated primary coolant that would exceed the requirements of 10 CFR Part 20 and are consistent with the facility i
ALARA program.
(3)
The system design or procedures will prevent contaminated primary s
coolant from entering the potable water system through the water makeup system.
(4)
The technical specifications, including testing and surveillance, I
provide reasonable assurance of necessary water makeup system operability'for normal reactor operations.
5.7 Nitrocen-16 Control Systpm 5.7.1 Areas of Review Non-power reactors that use either light or heavy water for neutron moderation i
or cooling will produce N-16 by the fast neutron-proton reaction in oxygen-16.
N-16, a high-energy beta and gamma ray emitter with a half-life of i
approximately 7 seconds, is a potential source of high radiation exposure at water-cooled non-power reactors. The N-16 tends to remain dissolved in the primary coolant water as it exits the core. The quantity and concentration of i
N-16 must be considered and provisions made to control personnel exposure.
Because of the relatively short half-life, potential doses.can be decreased by l
delaying the coolant within shielded regions.
For reactors using natural-convection in a large open pool, stirring or diffusing the convection flow to the surface can produce a delay.
For forced-flow cooling, passing the coolant through a large shielded and baffled tank can produce the delay.
In some non-power reactor designs, the entire primary coolant system may be shielded.
Specific areas for review for this section are discussed in Section 5.7 of the format and content guide.
i 5.7.2 Acceptance Criteria j
The acceptance criteria for the information in Chapter 5 are based on the following:
(1)
The reduction in personnel exposure to N-16 should be consistent with the N-16 analyses in Chapter 11 of the SAR. The reduction in exposure is acceptable if total dose does not exceed the requirements of 10 CFR Part 20 and is consistent with the facility ALARA program.
(2)
System design would not decrease cooling efficiency so that any LSSS would be exceeded.
12
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SAR Review Plan and Acceptance Criteria 1
Chapter 5 - Draft i
June 22, 1993 s
1ead to uncontrolled release or loss of coolant if a malfunction 4
were to occur l
prevent safe reactor shutdown and removal of decay heat sufficient to avoid fuel damage 4
(3)
The discussions should identify operational limits, design parameters,
}
and surveillances to be included in the technical specifications.
i 5.7.3 Review Procedures 1
l The staff will study the design bases and functional requirements of the j
system designed to control personnel exposures to N-16.
a Confirm the amount of N-16 predicted by the SAR analysis at the proposed power level and the potential personnel exposure rates.
This should include exposures from direct radiation and airborne N-16.
1 Review the type of system and the decrease in exposure rates.
Review the effect of the proposed system on normal and off-normal reactor enerations.
Review possible effects on reactor safety, safe reactor shutdown, and j
release of contaminated primary coolant as a result of malfunctions of j
the N-16 system.
5.7.4 Evaluation Findings J
Chapter 5 should contain sufficient information to support the following types of conclusions, which are to be included in the staff's safety evaluation i
report:
(1)
Design bases and design features give reasonable assurance that the N-16 control system can function as proposed and reduce potential doses to personnel so that no doses would exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program, (2)
Design and functional operation of the system give reasonable assurance a
it will not interfere with reactor cooling under anticipated reactor 1
operating conditions and will. not reduce cooling below the acceptable thermal-hydraulic performance discussed in Chapter 4 of the SAR.
(3)
Design features give reasonable assurance that malfunction of the N-16 control system will not cause uncontrolled loss or release of primary coolant and will not prevent safe reactor shutdown.
(4)
The technical specifications, including testing and surveillance, provide reasonable assurance of necessary N-16 control system operability for normal reactor operations.
4 13 i
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June 22, 1993 i
e i e
j 5.8 Auxiliary Systems Usino Primary Coolant 5.8.1 Areas of Review 1
(
i The primary coolant of an non-power reactor may serve functions in addition to i
cooling the reactor fuel.
Some of these auxiliary functions involve cooling i
other heated components, which may affect the heat load of the primary system.
Some of the auxiliary functions involve radiation shielding that may not contribute to the heat load but could require diverting or distributing the primary coolant to subsystems not involving core cooling.
J Auxiliary uses of the primary coolant could affect its availability as a fuel coolant, which is its principal use. Although the principal discussions of these systems are located in other chapters of the SAR, summaries of their effects on the coolant systems should be contained in this section. Auxiliary systems that may use primary coolant include the followina:
experiment cooling i
e l
experimental facility cooling l
e l
experimental facility shielding l
biological shield cooling l
thermal shield cooling fuel storage cooling or shielding l
Specific areas for review for this section are discussed in Section 5.8 of the i
a format and content guide.
5.8.2 Acceptance Criteria i
The acceptance criteria for auxiliary systems using primary coolant are based l
on the following:
i (1)
The system would remove sufficient projected heat to avoid damage to the cooled device.
(2)
The system would not interfere with the required operation of the primary core cooling system.
(3)
Any postulated malfunction of an auxiliary system would not cause uncontrolled loss of primary coolant or prevent a safe reactor shutdown.
(4)
The shield system using primary coolant would provide sufficient protection factors to prevent personnel exposures from exceeding the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(5)
The system would not cause radiation exposures or release of radioactivity to the environment that would exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(6)
The discussions should identify operational limits, design parameters and surveillances to be included in the technical specifications.
14
t SAR Review Plan and Acceptance Criteria Chapter 5 - Draft June 22, 1993 t
5.8.3 Review Procedures t
The staff will verify that auxiliary cooling or shielding using primary coolant are described in this section for any component (other than the core) in which potentially damaging temperature increases or excessive radiation exposures are predicted. Wherever the potential exists for radiation heating of components near the reactor core, the staff should verify that the heat source, temperature increases, heat transfer mechanisms, and heat disposal are i
discussed and analyzed. The staff should verify that the potential personnel l
radiation exposures from sources shielded by primary coolant have been i
analyzed and the protection factors provided by the coolant have been discussed.
5.8.4 Evaluation Findings Chapter 5 should contain sufficient information to support the following types of conclusions, which will be included in the staff's safety evaluation report:
I (1)
Auxiliary systems that use primary reactor coolant for functions other than in core fuel cooling have been acceptably described and analyzed.
The design bases have been derived from other chapters of the SAR. Any reactor components located in high radiation areas near the core have i
been analyzed for potential heating that could cause damage to the i
reactor core or failure of the component. Acceptable methods have been t
planned to remove sufficient heat to ensure the integrity'of the components. The coolant for these systems is obtained from the purified primary coolant system without decreasing the capability of the primary i
system below its acceptable performance criteria for core cooling.
l (2)
Any reactor components or auxiliary systems for which primary coolant helps shield personnel from excessive radiation exposures have been analyzed. The use of the coolant for these purposes is acceptable, and j
the estimated protection factors limit the exposures to the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
There is reasonable assurance that credible and postulated malfunctions of the auxiliary cooling systems will not lead to uncontrolled loss of primary coolant, radiation exposures, or release of radioactivity to the unrestricted environment that exceed the requirements of 10 CFR Part 20 and are consistent with the facility ALARA program.
(3)
The technical specifications, including testing and surveillance, provide reasonable assurance of necessary auxiliary cooling system operability for normal reactor operations.
REFERENCE Griess, J. C., et al., Effftgt of Heat Flux on the Corrosion of Aluminum by Water. Part IV, ORNL-3541, February 1964.
15
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SAR Standard Fermat cnd Centent
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i June 23, 1993
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l 14 TECHNICAL SPECIFICATIONS l
l j
14.1 Introduction j
l This section contains the guidelines to develop the technical specifications l
for NRC-licensed non-power reactors.
4 The NRC requires each operating non-power reactor applicant to develop technical specifications that set forth the limits, operating conditions, and j
other requirements imposed on facility operation to protect the health and safety of the public in accordance with 10 CFR 50.36. The technical specifi-1 j
cations are typically derived from the facility descriptions and safety con-i l
siderations contained in the SAR and represent a comprehensive envelope of safe operation.
j 3;
j Applications for construction permits / operating licenses and renewals of j
operating license's must contain proposed technical specifications that are
]
incorporated in the operating license. During its review of the application, l
l the NRC staff will review the SAR and proposed technical specifications to ensure they are complete, comprehensive, and that the public health and safety I
j will be protected. After final acceptance by the NRC staff, the technical-j specifications will be included as Appendix A to the operating license.
j i
j The format and content of the technical specifications discussed in this document follow that of the 1990 revision to ANSI /ANS 15.1, " Development of
}
Technical Specifications for Research Reactors." Additional guidance on the format and content of technical specifications can be found in previously i
j accepted and approved technical specifications for non-power reactors of l
similar design, operating characteristics, site and environmental conditions, I
and use.
lL This chapter of the SAR normally is very short. The applicant should be able i
to state conclusively in this chapter that the technical specifications were prepared following an accepted format, that normal operation of the reactor l
within the limits of the technical. specifications will not result in offsite i
radiation exposure in excess of 10 CFR Part 20 guidelines, and that the tech-nical specifications limit the likelihood and consequences of malfunctions.
The reader is referred to the technical specifications, which are contained in a separate document from the SAR. The technical specifications are neither-derived or justified in this chapter of the SAR. The actual technical speci-1 fications are determined by the analysis that appears in the other chapters of i
the SAR. Each of the technical specifications should be supported by the SAR and it is useful to refer to the supporting SAR analysis in the basis of each technical specification.
In the text that follows, all sections of ANSI /ANS 15.1 are addressed.
If modifications or clarifications for ANSI /ANS 15.1 are required to provide acceptable technical specifications, the additional guidance is provided.
Sections that provide acceptable guidance as written also are noted.
2
)
l 1
i' L
SAR Standard Fermat and Ccntent Chapter 14 - Draft June 23, 1993 14.2 Format and Content of Technical Specifications The numbering in this part (Sections 1 through 6.8) corresponds to the section numbering in ANSI /ANS 15.1-1990.
1 Introduction 1.1 Scope The NRC accepts the guidance provided in this section of ANSI /ANS 15.1. This section confirms that the technical specifications for non-power reactors should include all the categories in 10 CFR 50.36 for production and utilization facilities.
1.2 Application 1.2.1 Purpose The NRC accepts the guidance provided in this section of ANSI /ANS 15.1.
Technical specifications represent a set of operating requirements for a reactor that the licensee and the NRC have agreed on. The specifications become part of the operating license.
1.2.2 Format Sections of the technical specifications must be numbered as indicated in Section 1.2.2 of ANSI /ANS 15.1.
Subsections may be left out if not applicable for a particular reactor or altered if necessary, but the subsections included must be numbered in consecutive order.
For individual specifications in Sections 2, 3, and 4, applicability, objective, specification, and basis information must be included in the specified format. For Sections 5 and 6 of the technical specifications, ANSI /ANS 15.1 suggests the specifications be stated without providing applicability, objective, or basis. Although this format is preferred, it is acceptable to NRC if these sections include applicability, objective, or basis statements.
Technical specifications that use the SAR as a basis should explicitly reference the SAR section number.
In addition, any other sources used to support the technical specification should be explicitly referenced.
1.3 Definitions The NRC and the non-power reactor community have agreed on most of the definitions given in this section of ANSI /ANS 15.1. Those applicable to a particular facility should be included verbatim. Facility-specific defini-tions may be added to clarify terms referred to in the technical specifica-tions. Modifications and additional definitions presented below help clarify the meaning of terms used in ANSI /ANS 15.1.
2
r SAR Standard Format and Cent nt Chapter 14 - Draft j
June 23, 1993 The following definitions should be modified as indicated:
Class A reactor operator. The term acceptable to the NRC is senior reactor operator.
Class B reactor operator. The term acceptable to the NRC is reactor operator.
1 reactor shutdown. The reactor is shut down if it is subcritical by at least I dollar both in the reference core condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included.
reference core condition. The reactivity condition of the core when it is at 20 *C and the reactivity worth of xenon is zero (i.e. cold, clean, and critical).
shutdown margin. Shutdown margin is the minimum shutdown reactivity l
necessary to provide confidence that the reactor can be made subcritical i
l by means of the control and safety systems starting from any permissible l
cperating condition.
It should be assumed that the most reactive scrammable rod and all non-scrammable rods are in their most reactive position and that the reactor will remain subcritical without further operator action.
Note. Shutdown margin has a single value mutually acceptable to the NRC and the licensee, to be determined on a case-by-case basis.
The following definitions should be added secured shutdown.
Secured shutdown is achieved when the reactor meets the requirements of the definition of " reactor secured" and the facility administrative requirements for leaving the facility with no licensed reactor operators present.
shutdown reactivity.
Shutdown reactivity is the value of the reactivity of the reactor with all control rods in their least reactive positions (e.g., inserted). The value of shutdown reactivity includes the reactiv-ity value of all installed experiments and is determined with the reactor at ambient conditions.
2 safety Limits and Limiting safety system settings 2.1 Safety Limits All reactor licensees are required by 10 CFR 50.36(c) to specify safety limits in the technical specifications. These safety limits will be placed on impor-tant process variables identified in the SAR as necessary to reasonably pro-tect the integrity of the primary barrier against the uncontrolled release of radioactivity.
For non-power reactors, the radioactivity of concern is gener511y the fission products in the fuel.
For heterogeneous-core non-power reactors, the primary barrier is the cladding of fuel plates, rods, or pins.
Cladding integrity could be lost by softening, melting, blistering, or 3
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i SAR Standard Fermat and Ccntcnt l
Chapter 14 - Draft June 23, 1993 A
j yielding to excess internal pressure, all of which are dependent on temperature and operating history. For homogeneous-core reactors, this primary barrier may be the fuel matrix, the primary vessel or some other component that contains the fuel and the fission products.
l Reactor conditions and safety limits should be developed to avoid failure of I
the fuel and be supported by SAR analyses. Manufacturers have studied failure l
modes and failure parameters during fuel development programs. The NRC has i
issued NUREG reports approving the u::e of some low-enriched uranium fuel types in non-power reactors. The applicant should make maximum use of the appropriate references; some of which are listed at the end of this document.
The applicant should consult NUREG-1313, " Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Nonpower Reactors," for evaluating aluminum-clad, aluminum matrix l
plate-type fuels, using both high-enriched uranium (HEU) and new uranium-silicide low-enriched uranium (LEU).
It discusses temperatures from experimental irradiation tests at which plate blistering has been observed, a t
possible forerunner of failure. The NRC finds 530 *C an acceptable fuel and cladding temperature limit not to be exceeded under any conditions of i
operation. NUREG-1313 also reTerences tests with HEU plate fuel that have led I
to similar conclusions (Beeston et al., 1980; Gibson, 1967; Nazare et al.,
1975; Stahl, 1982).
l There are several reports on training reactor and isotope production, General l
Atomics (TRIGA)-type fuels (NUREG-1282; Simnad et al., 1976 and 1981; Simnad l
and West,1986; West et al.,1986). For stainless-steel-clad UZrH uranium weight percent (w/o) TRIGA fuel, stainless-steel-clad UZrH.,, s LEU 8.5 HEU (70%
U-235 enriched) 8.5 w/o fuel lifetime improvement program (FLIP) TRNA fuel, i
and stainless-steel-clad UZrH LEU 20 w/o and 30 w/o TRIGA fuel, General 65 Atomics has shown and NRC has,, accepted that integrity is not compromised under the following cases and conditions:
I for cladding temperature at or less than 500 *C, peak fuel temperature at or less than 1150 *C r
for cladding temperature greater than 500 *C, peak fuel temperature at or less than 950 *C For aluminum-clad UZrH, LEU 8 w/o TRIGA fuel, NRC accepts that the peak fuel temperature should not,, exceed 500 *C.
i For pulsed training assembled reactor (PULSTAR) types, NRC accepts that the U0, fuel temperature should not exceed 2400 *C and the Zircaloy-2 cladding temperature should not exceed 1500 *C.
For Aerojet-General Nucleonics (AGN)-201 reactor types, NRC accepts that the fuel temperature should not exceed 200 *C.
The applicant should base SAR analyses on the applicable fuel developer's reported test results to ensure fuel integrity under all operating conditions.
4
F SAR Standard Fermat and C:ntcnt Chapter 14 - Draft June 23, 1993 2.1.1 Important Process Variables ANSI /ANS 15.1 proposes a list of parameters that may be acceptable as process variables for non-power reactors and states that safety limits will be measur-able parameters. However, as discussed in the Sections 2.1.2 and 2.1.3 below, not all safety limits for non-power reactors must be monitored and actually measurable. Safety limits could be inferred from limitations on other process variables.
It is convenient in discussing fuel integrity to divide the non-power reactors into two groups:
those with engineered cooling systems (forced-convection cooling) and those without engineered cooling systems (natural-convection cooling or no active cooling system). Safety limits for these reactors are discussed in the following two sections.
2.1.2 Criteria-Reactors With Engineered Cooling Systems The NRC :.;odifies this section of ANSI /ANS 15.1 as follows. Operation of the cooling system for reactors with forced-convection cooling maintains fuel temperature within acceptable limits to ensure cladding integrity.
Important parameters include fuel temperature, coolant flow rate, coolant inlet temperature, height of coolant above core, and reactor power level. These parameters should be controlled and measured. When all values are jointly maintained within the limits determined by the safety analyses, fuel cladding integrity will not be lost. These parameters are important process variables on which safety limits should be established and specified in the technical specifications. Safety limits should preclude flow instabilities in the hottest channel and ensure that the minimum departure from nucleate boiling ratio (DNBR) is at least 2.0 (which has been an acceptable margin to the onset of nucleate boiling). The analyses should be over the range of all physical and engineering parameters of the fuel components, the core configurations, and the coolant systems, as well as include consideration for uncertainties.
For reactors that will operate with both natural-convection cooling and forced-convection cooling, safety limits should be specified for appropriate process variables in both modes of operation (see Section 2.1.3 below). All non-power reactors should be designed so that both fission heat and decay heat can be dissipated without fuel damage. The analyses also should show that the safety limits are not exceeded during all anticipated modes of operation.
2.1.3 Criteria-Reactors Without Engineered Cooling Systems For reactors that will be licensed to operate without forced-convection cooling, only criterion 2.1.3(2) of ANSI /ANS 15.1 is acceptable according to 10 CFR 50.36(c). The NRC modifies Section 2.1.3(2) of ANSI /ANS 15.1 as follows. For reactors that either circulate coolant by natural thermal convection or have no specific coolant or cooling systems, thermal-hydraulic coolant parameters are not separately controllable. The SAR should ensure fuel cladding integrity and identify appropriate parameters chosen for safety limits in the technical specifications.
High fuel temperature is the likely precursor of fuel failure. Therefore, a maximum allowable fuel temperature safety limit should be established below 5
SAR Standard Format and Centent Chapter 14 - Draft June 23, 1993 1
which fuel integrity is ensured. On the basis of this fuel temperature, a power level should be calculated using an appropriate n;argin that ensures the fuel remains below the fuel temperature safety limit.
If the license will contain a provision to measure fuel temperature, the maximum fuel temperature in the core would be the parameter on which a safety i
limit is established. The SAR should show the relationship between the mea-sured fuel temperature and the maximum fuel temperature for the proposed reactor conditions.
If there is no provision for measuring fuel temperature directly, the calculated power level should be selected as the safety limit, t
on the basis of the maximum allowable fuel temperature and appropriate margin.
However, the basis for the safety limit still should be the maximum allowable fuel temperature.
Because most TRIGA-fueled cores have at least one instrumented fuel rod, the NRC has accepted fuel temperature alone as the safety limit for these reac-tors. Because the point of measured fuel temperature is normally not the point of maximum fuel temperature, the SAR must show the relationship between the measured fuel temperature and the maximum fuel temperature.
i For plate-type fuel, fuel temperature measuring capability is generally not available. Therefore, the licensee should determine a fuel cladding temperature below which cladding damage (softening or blistering) can be precluded. The licensee should then establish a corresponding power level, i
reactor conditions, and uncertainties that limits cladding temperature below the damage limit.
1 For reactors without fuel elements, such as the homogeneous AGN-201 reactors, safety limits should be based on considerations similar to those for plate-type fuel. The power level established in the SAR as a safety limit must provide reasonable assurance that fission products will not be released from their confining barrier, which could be either the fuel matrix or the fuel i
canister.
Safety limits should be based on the SAR. The technical specifications should discuss the mechanism and magnitude of the fuel limitation, including a primary reference for the fuel development studies that support the safety I
limit presented (NUREG-1282 and -1313; Simnad et al.,1976).
The analysis should address authorized core configurations and limiting thermal power levels and conditions for the reactor.
Safety limits acceptable to NRC for various reactor fuels are discussed in Section 2.1 above.
2.2 Limiting Safety System Settings The NRC accepts the guidance of this section of ANSI /ANS 15.1. The SAR should address normal operating conditions, off-normal operations, and all pertinent postulated accident scenarios. For each parameter on which a safety limit is established by the SAR, a protective channel should be identified that prevents the value of that parameter from exceeding the safety limit. The calculated set point for this protective action, providing the minimum i
acceptable safety margin considering process uncertainty, overall measurement 6
~~
SAR Standard Fermat and Cent;nt Chapter 14 - Draft June 23, 1993 r
uncertainty, and the transient phenomena of the process instrumentation, is defined as the " limiting safety system setting (LSSS)." Because the LSSSs are analytical limits, the protective channels may be set to actuate at more conservative values. The more conservative values may be established as limiting conditions of operation (LCOs).
LCOs also may be determined on the basis of experience, which has shown that safety system channels can be set readily within 20 percent of the normal operating value for a measured parameter, if the LSSS is not exceeded, without undue interference on operations.
In many cases the LCO can even be within 10 j
percent of the operating value.
If LCOs are set more than 20 percent from the operating value of the parameter, the SAR justification should be referenced.
2.2.1 Criteria-Reactors With Engineered Cooling Systems j
The NRC accepts this section of ANSI /ANS 15.1 for the forced-convection cooling mode of operation. For reactors licensed to operate with forced-convection cooling, this specification should list the LSSS derived in the SAR for each reactor parameter for which a safety limit was established. The Lases part of this specification should indicate the SAR assumptions and limits of uncertainty for each analyzed LSSS.
For reactors licensed to operate in forced-and natural-convection cooling modes, there should be appropriate LSSSs for both modes.
i i
2.2.2 Criteria-Reactors Without Engineered Cooling Systems The NRC substitutes the following guidance for Section 2.2.2 of ANSI /ANS 15.1.
Section 2.1.3 above requires safety limits be established by SAR analysis for all licensed reactors; therefore, channels should be established on the basis of SAR analysis to not violate each of these safety limits. Calculated LSSSs defined in Section 2.2 c; ANSI /ANS 15.1 and this document should be provided as technical specifications.
3 Liraiting Conditions for Operations LCOs are derived from the safety analyses in the SAR, which provide the bases for the LCOs.
LCOs are implemented administrative 1y or by control and monitoring circuitry to ensure that the reactor is not damaged, that the reactor is capable of performing its intended function, and that no one suffers undue radiological exposures because of reactor operations.
The NRC accepts this section of ANSI /ANS 15.1 as amplified in the following sections. Many of the LCOs have evolved from experience. Many are facility specific, depending on reactor type, operatireg characteristics, and site loca-tion. The NRC accepts the LCOs discussed in this section provided the appli-cant justifies the LCO and shows the applicability to the specific facility.
Additional specifications may be appropriate for unique facility designs or experimental features or for additional conservatism in operations required by the applicant or NRC. As noted above, LCOs can in many cases be set within 10 percent of the normal operating level of a parameter. Specifications on surveillance intervals for LCOs and other parameters and facility design features are provided in Sections 4 and 5, respectively, of ANS!/ANS 15.1.
LCOs should be provided as outlined in the remainder of this section.
7
SAR Standard Farmat and Ccntcnt Chapter 14 - Draft June 23, 1993 3.1 Reactor Core Parameters (1) Excess Reactivity The upper limit for allowed excess reactivity should be specified. The referenced SAR analyses should discuss all operations that require excess i
reactivity and the safety implications for the excess reactivity pro-posed. The discussions should include operational flexibility, potential accidents, and the relationship to shutdown margin. The SAR should contain a discussion of the safety implications of the excess reactivity, j
including resultant shutdown reactivity with all control rods inserted effects on the reactor of any credible rapid removal of a control or safety rod potential effects of other maximum credible rapid additions of excess reactivity
)
possible reactivity changes caused by experiment failure or displacement interrelationship between shutdown margin and excess reactivity i
If none of the postulated events would lead to loss of fuel integrity or to uncontrolled release of radioactivity, the proposed excess reactivity i
would be acceptable.
(2) Shutdown Margin A single value for the shutdown margin, as defined in Section 1.3 above, should be specified. The specification should state that compliance with the shutdown margin takes precedence over the excess reactivity specifi-i cation.
In addition, other reactor parameters that apply to the shutdown margin should be stated. These should include cold, clean core reactivity conditions (e.g., temperature and poi. cans), core configuration (e.g., fuel and control rods), and the status of experiments (e.g.,
movable experiments in their most reactive state). The value of the shutdown margin should be large enough to be readily determined i
experimentally, for example, 20.5% Ak/k or 20.50 dollars.
(3) Pulse limits Because the TRIGA design is the only reactor design the NRC currently licenses to pulse at this time, the specific values given in the discussion below apply only to TRIGA design pulsing reactors. However, the general design criteria discussed below may be applicable to other potential pulsing non-power reactor designs.
The maximum reactivity addition for a pulse is a license condition similar to maximum thermal power and is determined case by case. The value should be based on the SAR analysis for maintaining fuel integrity, 8
e t
SAR Standard Fcrmat and Centant l
l Chapter 14 - Draft June 23, 1993 t
which considers fuel type, limiting core cenfigurations, reactivity l
l feedback coefficients, operating history, heat capacity, and peak fuel i
temperature limitations. This LCO on the maximum reactivity addition administratively gives essurance that the maximum pulse reactivity addition license condition and the safety limit on maximum fuel
}
temperature will not be exceeded.
l i
The SAR should show that the maximum reactor pulse for a TRIGA reactor j
with stainless-steel-clad UZrH fuel would not raise the peak fuel t
us temperature of any element above 1000 "C (Simnad et_al., 1976).
(This is l
i l
a conservative limit, proposed by General Atomics and accepted by the j
NRC, that is not to be confused wi% the safety limit temperature value.)
j For a TRIGA reactor with aluminum-clad UZrH fuel, the analysis should showthatthepeakfueltemperaturewill'nd,, exceed 500*Cforthe l
l maximum reactor puise. The analysis should be applicable to the specific j
reactor considering its core size, operating history, fuel types, feedback coefficients, temperature gradients and the power peaking of all authorized core configurai. ions. The potential effects of pulsing on in-l core experiments or detectors should be' included in the. analysis. Any 1
required limitations-on experiments should be included in Section 3.8 of j
the technical specifications document.
i The report by Simnad et al., (1991) discussing fuel damage at Texas A&M University TRIGA Reactor should be reviewed to determine if reactor operating history and power level would require a lower peak pulse fuel i
j temperature because of damage to the fuel during pulsing operation.
j If the analysis of the core shows that the worth of the pulte rod could i
j exceed the mar.imum reactivity insertion limit and allow an amount of j
reactivity to-be inserted into the core that could damage the fuel, there shoutJ be a limit on the total worth of the pulse rod. There should be a steady-state power level alove which pulses shall not be initiated.
For q
j_
TRIGA reactors, the NRC has accepted a power level of 1 kW in conjunction with an interlock that prevents movement of the steady-state control rods j
when the reactor-is in the pulse mode.
a For other pulsing reactors, the proposed limiting fue1~ temperature and
(
reae.tivity insertion should be justified by reference to appropriate tests and analyses.
~
I
}
(4) Core Configurations The applicant should specify special core configurations, experirental facilities internal to the core, special neutron reflectors,-burnable poisons, or mixed fuel types assumed in the SAR. The following i
specificatiens should be included in the LC0 for core configurations:
If analysis shows the reactivity effects of waterholes in the core
.i j
needs to be limited due to reactivity insertion _ accidents, the
'l reactor should have a closely packed core tiin acceptable vacancies i
in the core' center and periphery describec rhis does not prevent t
the use of in-core experimental facilitier Reactors with thermal power levels in excess of I megawatt and cross-section area of core-9
SAR Standard Fcrmat cnd Centerd Chapter 14 - Draft June 23, 1993 experimental facilities greater than 16 square inches will be licensed as testing facilities.
No fuel should be inserted or removed from the core unless the reactor is subcritical by more than the worth of the most reactive fuel element.
If control rods need to be removed from the reactor core for inspection, an LCO should state the negative reactivity necessary in the core before a control rod can be removed.
Non-power reactors should be designed with reactivity and void coefficients and a power defect sufficiently negative that many reactor transients are inherently counteracted to avoid loss of fuel integrity.
Although the individual reactivity coefficients and power defect are addressed in the new soecification below, this LCO should be used to develop specifications on allowed core configurations to ensure the assumptions used in the development of limits on those parameters are met.
The specified conditions of core configuration are acceptable to NRC if the SAR shews that none of the conditions analyzed could lead to loss of fuel integrity, uncontrolled release of radioactivity, or potential exposures exceeding 10 CFR Part 20.
l (5) Reactivity Coefficients (Added by NRC) f Non-power reactors should specify reactivity coefficients for fuel temperature, moderator temperature, and VGid volume and a power defect.
The net effect of the coefficients and the power defect should be negative over most of the range of reactor operations. The SAR analyses i
of both routine operation and potential accident scenarios should show that the net negative effect of these core characteristics is sufficient to mitigate any anticipated event or postulated accident scenario.
Reasonable values should be designed into the reactor (e.g., by under-i moderation of the neutron spectrum). Values for surveillance should be specified for those negative reactivity coefficients and the negative power defect that can 'oe measured. The values of the coefficients and the power defect are acceptable if they ensure that the assumptions and j
initial conditions of the analyses are enveloped to prevent compromise of the fuel integrity during reactor transients and other applicable accident scenarios.
(6) Fuel Para,seters (Added by MRC)
An LCO should be specified for certain fuel parameters or characteristics.
Design features of the approved fuel should be inc;uded in Section 5 of the technical specifications document.
Fuel-related LCOs include the following:
10
SAR Standard Format and Centcnt Chapter 14 - Draft June 23, 1993 1
(a) All Fuel Types I
No operation with damaged fuel except to locate such fuel. The e
definition of damaged fuel should specify limits on longitudinal growth, bowing, or bending and limits on detectable amounts of fission products that could escape through the primary barrier.
Periodic visual inspection of fuel. This specification should be clear and explicit and reference fuel manufacturers' guidance or recommendations for detecting deterioration. The intervals and methods of fuel inspection should be specified in i
l Section 4 of the technical specifications document. The purpose of inspection is to detect cladding deterioration that results from erosir~, corrosion, or other damage.
(b) TRIGA Fuel l
Additional technical specifications limit fuel rod elongttion, l
bowing, and uranium burnup. Limits listed below were proposed by l
General Atomics and accepted by the Atomic Energy. Commission (AEC) during initial licensing of pulsing TRIGA reactors.
If these limits are reached, the fuel element is defined as " damaged fuel.
Acceptable specifications for TRIGA fuel for both steady-state and l
pulsed operation include the following:
Bowina - For stainless steel-clad UZrH TRI sagitta shall not exccad 0.125 in. (0.$N cm)GA fuel, the over the length of the cladding in a hexagonal-grid core arrangem.nt or 0.0625 in. (0.159 cm) elongation over the original length of the cladding in. circular-grid core arrangement.
For aluminum-clad UZrH,o fuel, the limit on sagitta should be 0.125 in, (0.318 cm Eloncation - For stainless steel-clad UZrH TRIGA fuel, the g 63 total length of the fuel element shall not, exceed its original length by more than 0.125 in. (0.318 cm). For aluminum-clad UZrH,,o fuel, the limit on elongation should not exceed 0.5 in.
(1.2T cm).
Barnup - The burnup of uranium-235 in the UZrH fuel matrix shall not exceed 50 percent of the initial concentration (NUREG
-1282 and Simnad and West, 1986).
(c) Materials Testing Reactor (MTR)-Type Fuel l
To prevent fuel swelling there should be burnup limitations on the i
fuel. Aluminum-clad aluminum-matrix MTR-type fuel plate non-power reactors should have technical specifications that limit U-235 burnup or fission density. The specifications are acceptable if they are consistent with the SAR, which accounts for all relevant t
thermal-hydraulic and metallurgical considerations. The NRC is specifically concerned with the maximum burnup limit for plate-type 11
SAR Standard Format and Centcnt
)
Chaptsr 14 - Draft j
1 June 23, 1993 i
fuels because of the build up of oxide on the fuel cladding. This can be a concern when licensees apply to increase the maximum i
acceptable burnup. The increased resistance to heat transfer to the coolant may affect consequences considered in the accident analysis chapter of the SAR. NRC has accepted uranium burnup densities in fuelsbagedonauran{umaluminidematrixuptoafissiondensityof 2.3 x 10 fissions /cm and up to 50 percent of the initial concentration of uranium-235 (Beeston et al.,1980; Gibson,1967; Nazare et al.,1975; Stahl,1982).
(d) PULSTAR Fuel Burnup of pin-type PULSTAR fuel should be limited by a specification based on testing and the SAR analysis. The NRC has accepted burnup limits up to 20,000 MWed/ tonne uranium.
LCOs are acceptable if they are analyzed in the SAR and consistent with I
the values given above. The analyses should verify for these fuel J
parameter conditions that the fuel will not exceed safety limits for normal and off-normal operations.
.3.2 Reactor Control and Safety Systems (1) Operable Control Rods The number and type of operable control and safety rods should be speci-fied. There is no prescribed minimum number of operable control and safety rods for non-power reactors. The specification regarding the number of operable control rods is acceptable if the excess reactivity and shutdown margin specifications required by the SAR analyses can be ensured for all operating conditions. The individual or total reactivity i
worths need not be specifically listed. A rod of lesser worth might be designated the " regulating rod" and is used as a fine power adjustment mechanism.
In some cases the worth of a control rod (s) connected to an automatic control system (which can add reactivity) may be limited to a maximum amount that was assumed in the SAR in this LCO. This regulating rod need not have scram capability but rods without scram capability i
should not be used when showing compliance with shutdown margin requirements. Other rods of greater worth, with automatic protective L
(scram) function, should be capable of achieving the specified shutdown margin.
The maximum scram time should be specified for each scrammable rod. The specification should ensure that the drop times are consistent with the j
SAR analysis of reactivity required as a function of time to terminate a reactivity addition event accounting for measurement and calculational uncertainties.
In most non-power reactors, for rods 2--to 3-feet long, full rod insertion time in the absence of excess mechanical friction or interference is less than I second.
If a specification proposes a longer l
scram time, it requires appropriate SAR analysis. The NRC-finds it acceptable to shut down a non-power reactor by intentionally scramming the control and safety rods.
12 i
SAR Standard Fcrmat and C nt:nt Chapter 14 - Draft June 23, 1993 (2) Reactivity Insertion Rates The maximum rates of adding positive reactivity should be specified for the control and safety rods. The specification should specifically state that gang or multiple rod withdrawal is allowed. Control rod (s) con-nected to an automatic control system may have maximum rates of reactiv-ity addition that differ from the rest of the control rods. The accept-able rates should be based on the SAR, including inadvertent addition of ramp reactivity at the maximum rate for the most conservative power, rod position, and reactor conditions.
(3) Pulsed Operation Limitations on reactivity additions are discussed in Section 3.l(3) above and need not be repeated here.
If any hardware systems require special limitations for pulsing, they should be discussed in this section of the technical specification document. Examples might include (a) special l
core configuration, (b) specific location of pulse rod, (c) number of i
pulse rods, and (d) removal of in-core fueled experiments. These specifications are acceptable if the assumptions of the SAR are ensured and damage to the reactor by authorized pulses is precluded.
If an experiment containing fissionable material could be damaged by reactor pulsing, limiting specifications must be provided to preclude this event in Section 3.8 of the technical specifications document.
(4) Scram Channels i
A table should specify all required scram channels and set points, the minimum number of channels, other functions performed by the channel, and reactor operating mode, such as steady-state power or pulsed, and cooling method, such as forced-or natural-convection coolant flow. The safety limits that the scram protects should be discussed in the basis for the table. Table 1 provides an example of how the information could be displayed.
Reactor scrams should be based on the SAR. There should be at least two completely independent power level scram channels and they should provide diversity and redundancy.
Historically, NRC has accepted power level scrams as high as 1.2 times the licensed power. This is the only non-power reactor scram set point that, if reached, violates the license (maximum power level). Some licensees have incorrectly interpreted this scram set point as allowing limited operation above the license power level. Although this is gen-erally not a safety concern, the NRC staff recognizes it as a regulatory problem. For example, if the reactor power measuring channels are out of calibration, it is possible that the reactor has been operated at several percent above the maximum licensed power level for a period of time. To ensure that licensed power levels are not exceeded for non-power reac-tors, power level scrams should be set below the licensed power level.
The NRC staff, upon submission of a license amendment request, has approved license amendments for non-power reactors that raise the licensed power 10 percent above the power value at which the reactor will be operated. This allows the operating power to remain the same while 13
SAR Standard Fermat and Centcnt Chapter 14 - Draft June 23, 1993 Table 1 Typical required scrams and power reverses
- minism ruuber chwret set point **
required and Function Period safety Scram if period 5 3 see 1
Period reverse Rod rm in if period 510 sec 1
Power tevet safety (linear and safety)
Scram if power > 1001 2
Power level reverse (safety)
Rod re in if power > 971 1
Nish power /no coolant flow scram if flow < 56.8 t/sec (900 spm) 1 and power > 100 W Mish power / flapper open scram if power > 100 W 1
and ftapper open Flapper closed /no coolant flow scram if flow < 56.8 t/sec (900 spa) 1 and fIapper c!osed sof tware (digital) malfmetion Scram upon malfmetton 1
i Loss of high voltage to detectors scram if voltage lost 1
Poot water levet scram if level < 4.88 m (16 ft) 1 r
above core top Bridge not clanq:ed Scram when clamps released 1
Bridge radiation level and Scram if radiation a 50 anram/hr and 1
Building exhaust air radiation level concentration t 2 x 10* Acl/mL 1
Manual scram switch Scram when switch depressed 1
Rod magnet power keyswitch Scram when magnet power turned off 1
Fuel temperature Scram if temperature t $50 *c (1022 *F) 2 j
Reactor coolant exit tenperature scram if temperature t 55 *c (131 *F) 1 Automatic control system out of timit Rod rm in if out of specification 1
Experiment scram if setpoint is vlotated 1
Loss of site power Scram if power lost 1
1
- As Ittustrative values, the set points and channels listed do not apply to any one reactor.
4
- Values listed are limiting set points. For operational convenience, set points may be changed to more conservative val es.
l t
t 14 i
9 SAR Standard Fermat and Cent nt Chapter 14 - Draft June 23, 1993 retaining the scram set points within the license and near the previous values.
In this example, the scram set points will be less than the licensed power level and will scram the reactor before the licensed power level is exceeded.
Power level scram set points above the licensed power level will 1
only be acceptable on a case-by-case basis when justified by analysis in the SAR which includes conservative assumptions with regard to fission product inventory, decay heat, fuel temperature, and other pertinent reactor conditions.
(5) Interlocks Required interlocks that inhibit or prevent control rod withdrawal or 1
reactor startup should be specified by a table (see Table 2 as an i
example).
Interlocks should be specific to the facility and based on the SAR. These interlocks include operability of area or other radiation monitors experiment facilities confinement and ventilation systems i
e initial conditions for pulsing e
detected neutrons for startup Operability of measuring channel components, such as ion chamber power supplies and recorders as discussed in the SAR Table 2 Typical required interlocks
- Minimun N&r channet Reg. sired Fmetion Recorders not operating 3
Prevent rod withdrawal (start y irsibit)
Neutron comt rate (starte) 1 Prevent rod withdrawal (start @ frhibit)
If comtrate s 2 eps simultaneous rod withdrawat 5
Prevent withdrawet of 2 or more rods Norpulse condition 1
Prevent movement of pulse rod in steady state mode
- Pulse withdrawet Prevent movement of standard control rods in pulse mode Transient withdrawet 1
Prevent movement of pulse rod sith reactor power above 1 kW Watues listed are timiting set points. For operationet convenience, set points may be changed to more conservative values.
If the reactor will be licensed to operate in more than one mode, the specification should include the mode for which the interlock is required.
If permanent interlocks are established for special experi-15
SAR Standard Fermat and Centcnt a
Chapter 14 - Draft June 23, 1993 i
ments, shields, or access control, they should be included in the technical specifications, as described in the SAR.
(6) Backup Shutdown Mechanisms Most non-power reactors are required to use only control and safety rods i
for shutdown.
If the SAR identifies a need for backup mechanisms (for example, moderator dump in a critical facility), they should be specified with appropriate requirements placed on their operability (LCOs).
(7) Bypassing Channels i
Any individual channels identified in items 4, 5, or 6 above for which bypassing is allowed during reactor operation should be justified in the SAR and specified under this item. Only minimal bysassing should be permitted in safety systems and never in a system t1at could compromise scram capability of the other channels.
Bypassing temporary scrams or interlacks associated with experiments need not be included in the tech-nical specifications but should be addressed in specific experiment protocol.
(8) Control Systens and Instrumentation Requirements for Operation (Added by l
NRC)
Non-power reactor technical specifications should have redundant and 1
accurate power level monitors that cover the range from subcritical j
source multiplication to above the full licensed power level. Not all monitors are required to include scram capability (see Table 3 for a typical minimum set). These include a startup channel, linear power monitor, logarithmic power monitor, and safety channel (s).
In addition,
(
most non-power reactors have a period channel (meter), including a period scram. One should be specified as analyzed in reactor transient response section of the SAR.
[
t i
Table 3 Typical required minimum measuring channels
- Minlaun l
rumber i
Channel regstred Fisiction starte 1
Nonitor otscritical multiplication for start,
Power level 2
Irput for safety power levet scran Putse power 1
frput for putse power teveI seram Fuel temperature 2
frput for fuel temperature scram Los N/ period 1
wide rance power tevet and input for period i
meter and period scram j
Linear power levet 1
Display power for control u 16 1
Dispter power level
- As (Llustrative vetues,.thess channels do not apply to any one reactor. Minisua channels for e particular I
l facility are determined from the SAR anetysis.
t 16
[
t
SAR Standard Format and Cent;nt Chaptcr 14 - Draft June 23, 1993 s
1 Some non-power reactors with forced-convection cooling have a channel that displays the radiation level of N-16 in the primary flow. Although the N-16 channel is not required in the SAR for mitigating transients it can be an important channel because (1) it has greater stability than delta temperature across the core during power changes and (2) it is not affected by changes in core flux distribution caused by fission product buildup in the core that can affect the ionization chambers.
If it is necessary for the operators to use the N-16 channel in reactor operations, it should be on the list of specified channels.
If digital control and safety instrumentation is used, an analog system should be specified to provide diversity and redundancy.
Specifications in this section should include the entire channel, including readout meters and recorders and the protective fun utons they l
perform, such as to prevent an LSSS from being exceeded.
Each non-power reactor should have more than one power level channel indication in the control room when operating at full licensed power.
However, because sensors and channel electronics might not be identical, I
the channels may indicate slightly different power levels.
Power level is a principal license condition, and each applicant may consider designating a primary channel for power level monitoring. That channel l
should be calibrated for thermal power in the region of maximun licensed power and should be recorded in a way that allows auditing for later t
proof of authorized operation within the license condition. Facility procedures should identify this designated channel and allow for l
alternate designations using analytic comparisons to achieve operational flexibility, if necessary. However, technical specifications or facility l
procedures that do not include this concept are acceptable to NRC because a designated channel is not a requirement upon licensees.
l 3.3 Coolant Systems i
l The basic systems required for cooling the fue' nd other components, for limiting corrosion, and for monitoring coolant radioactivity in non-power reactors should be specified in this section. All non-power reactors should have the capability to remove both fission and decay heat to ensure fuel integrity under all potential conditions. All reactors requiring forced-convection cooling systems shall have specifications assuring operability of syttems and reactor configurations for fail safe changeover from forced-convection to natural-convection cooling. An adequate heat sink, as described l
l in the SAR, is a necessary component of such a system.
For reactors licensed to operate in both forced-and natural-convection cool-ing modes, the appropriate coolant system configurations and the relevant power levels for both modes should be specified as analyzed in the SAR.
Not all of the following items apply to all types of non-power reactors.
However, when applicable, they should be limited by technical specifications, l
on the basis of the analyses and justifications in the SAR.
17
SAR Standard Format and Centcnt Chapter 14 - Draft June 23, 1993 (1) Shutdown Cooling or Pump Requirements As a minimum, the requirements for natural-convection cooling and the operability and status of related systems required for shutdown should be specified as LCOs.
If additional requirements are necessary for tem-porary forced-convection cooling following reactor shutdown from extended high-power operation, the technical specifications should state them, using the SAR as the basis.
(2) Isolation Valves The existence, location, operability, and status of any valves required to isolate subsystems or components for operational needs, including removal of decay heat, should be specified as LCOs.
l (3) Coolant level limits Both the coolant pressure (boiling temperature at the fuel) and adequate natural-convection flow depend on the level of water above the core.
In addition, vertical and horizontal radiation shielding by the coolant might be required.
Pool water level should be an LCO for both reasons, using the SAR as the basis.
(4) Detection of Leakage or loss of Coolant If leakage or other loss of coolant could lead to an uncontrolled release t
of radioactivity (see items 5 and 8 below) to the environment, an LCO should state the need for operability of detection systems.
If credit is taken in the SAR, this includes systems that monitor the pressure differ-ence between the primary and secondary cooling systems at the heat l
exchanger to detect conditions that wou'.d allow loss of primary coolant in the event of a heat exchanger leak. Heavy water systems should always have leak detection capability.
(5) Detection of Fission Product Activity The technical specifications should provide for prompt detection of fission products escaping from the fuel barrier. The method could be a radiation detector placed in the primary coolant flow loop or a strategi-cally located continuous air monitor in the reactor room or a ventilation duct. Temporary substitutions, in case the fission product monitor is inoperable, should follow guidance in Section 3.7.1 of ANSI /ANS 15.1.
This specification may be combined with the specification discussed in i
Section 3.7.1 (2) on fission product monitors.
The specified fission product monitor should be able to initiate action, such as a reactor scram or reactor room isolation. The SAR should pro-vide the bases and describe how fission products are distinguished from other waterborne or airborne radioactivity.
18
IiO SAR Standard Format and Centent Chapter 14 - Draft June 23, 1993 i
(6) Hydrogen Concentration (Off-Gas) Limits l
If the SAR has shown any of the isotopes of hydrogen (hydrogen, deuterium, and tritium (H,D,T)) to be a significant risk to personnel 4
or the facility, an LCO should provide for detection or adequate control, as discussed in the SAR.
1 (7) Energency Core Cooling Systens 1
4 l
If the SAR indicates a need for supplemental core cooling to mitigate a 1
loss-of-primary-coolant event, the technical specifications should con-tain an LC0 requiring an operable and adequate system. The system should' j
satisfy the cooling requirements for the SAR scenario and should not 1
i depend on continued availability of normal electrical service.
.l (8) Secondary and Primary Coolant Radioactivity Limits In addition to the prompt detection of fission products from failed fuel or experiment malfunctions, LCOs should limit radioactivity in' the 1
coolant. The technical specifications should require periodic sampling i
j and appropriate analyses to detect'and quantify radioactivity in both the primarv and secondary coolant. The coolant should be sampled for gross activity on a short interval, for example, weekly, and sampled for. iso-t i
tope identification on a longer interval, for example, quarterly. The l
purpose of this LCO is to detect deterioration of components in the pri-l i
mary coolant loop, such as a control element, and leakage in a heat _
jl exchanger into the secondary coolant loop. These specifications should be stated in such a way that significant changes in radioactivity, as defined in the SAR, trigger remedial action.
1 (9) Water Chemistry Requirements
(
i To control corrosion of such components as the reactor fuel, structure, i
and pool, control activation of impurities in the reactor coolant, and-
)
maintain visual clarity of the reactor coolant, there should be LCOs on 4
both electrical conductivity and pH of the primary coolant. These l
s specifications also should apply to any water that comes into contact j
with the fuel, such as in fuel storage tanks and pits. The limits and J
ranges of values should be given explicitly and be consistent with recommended values given by both fuel vendors and water chemistry guidelines. The conductivity should be monitored continuously. There l
should be a definite schedule for measuring pH, during both operating and a
shutdown periods. The bases should clearly address the appropriate -
ranges and give meaningful references.= The SAR should justify the values for conductivity and pH for the particular reactor. Acceptable ranges i
for these process variables have traditionally been s5 pahos/cm for conductivity and between 5.0 and 7.5 for pH. These values can usually be achieved by demineralization, filtration, and good housekeeping practices, but chemical methods should be described and specified, if applicable.
W Ig 4
__m-o.
...r.
,._.-,_4-.
iI SAR Standard Format and Centent l
q Chapter 14 - Draft -
l l
June 23, 1993 l
l j
)
3.4 Containment or Confinement l
Because accidents that result in release of steam and building over pressure are uncommon, most non-power reactors are housed in a confinement system, not a containment. There should be an LCO requiring that the system specifically.
i described in the SAR exists as stated. The system should be operable during l
operation and for other applicable times such as before operation and follow-l ing shutdown, as noted in Sections 3.4.1 and 3.4.2 of ANSI /ANS 15.1.
If i
t interlocks or administrative controls to ensure operability are required,-
there should be appropriate specifications. Whether the facility has confine-ment or containment depends on the reactor design, operating characteristics, j
and facility location. Specifications should require nominal exhaust rates for air under the operating and accident conditions analyzed in the SAR..
i Specifications should limit building leak rates to those described in the SAR.
3.5 Ventilation Systems Ventilation and exhaust flow rates and the systeps to achieve the controlled release of effluents, as analyzed in the SAR, should be specified as LCOs.
3 These LCOs should be estabihhed to achieve controlled release of effluents.
Automatic fail-safe closure of vents should be specified for confinement sys-tems. Provisions to initiate controlled, filtered, and monitored exhaust and ventilation for radiological accidents should be included.
In some cases, depending on the results of the analysis, minimum airflow rates may be LCOs.
The specified ventilation system should maintain a lower air oressure in the reactor room than in adjacent spaces. Reactor room air should not be distri-buted to other occupied spaces within buildings. The location and height of the air exhaust system stack or release point should be specified as an LCO here or as a design feature in Section 5 of the technical specifications document. The dimensions of the stack should be consistent with the assumptions used in the SAR to predict potential radiation doses in the
-l unrestricted environment.
It is acceptable that the concentration of airborne radioactivity at the point of exhaust for normal operation be higher than the regulatory limit for restricted areas, provided that this point is not readily accessible to the public, the analyzed doses to the public are well below regulatory limits for unrestricted areas, and the potential doses to the facility staff are within regulatory limits. Requirements for the as-low-as-reasonably-achievable (ALARA) program should be applied in all analyses (see Section 3.7 below).
3.6 Emergency Power l
t Any requirement for emergency electrical power for non-power reactor. facili-ties should be analyzed in the SAR on a case-by-case basis. Any necessary-facility functions, such as radiation monitoring, emergency core cooling, or
-isolating containment or confinement, that need to be maintained if normal.
electrical power is lost should be described in the SAR.
If emergency power i
is required, an LCO should ensure operability of the system.. The technical specification thould specify automatic startup of emergency electrical. power if indicated in the SAR.
20
1
'I SAR Standard Fcrmat and Ccntcnt Chapter 14 - Draft June 23, 1993 s
3.7 Radiation Monitoring Systems and Effluents Monitoring systems and effluents may be addressed in the technical specifi-cations under separate principal headings.
The following discussion is consistent with the corresponding sections of ANSI /ANS 15.1.
3.7.1 Monitoring Systems A separate table in the technical specifications document (see Table 4) should list the required radiation monitors, the function each performs (e.g., scram I
or containment isolation), the approximate location, the type of radiation l
detected, and the alarm and/or automatic action setting, as analyzed in the l
SAR. The set points and calibrations should be listed in terms of radiation exposure rates and concentrations rather than count rates that can change with calibration. Specific count r'ates for alarms and action settings can be presented in a facility procedure that can be amended in accord with the technical specifications document procedures section. For specified monitors that b(come inoperable, the specification should require that reactor operations may continue only if the monitor is replaced by a substitute or portable monitor. The replacement monitor must perform essentially the same function until the original monitor is repaired or replaced (generally not to exceed I work week unless justified in the SAR). The specification also should require that if the specified monitor was displayed in the control l
room, the temporary monitor also should be observable by the operator on duty.
l The applicant should provide a table applicable to the specific facility on the basis of the SAR.
j (1) Air Monitors (Gas and Particulate)
Monitors should be specified for both radioactive gas and those radio-active particulates that might be airborne in the reactor room. There should be at least one continuous air monitor (CAM) with an audible alarm and data recorders. These monitors should be capable of alerting facil-ity personne' to the presence of radioactivity. They should be cali-brated for anticipated radioactive species.
Potential sources of airborne radioactivity should be analyzed in the SAR.
l There should be specifications requiring operability of properly
~
calibrated effluent monitors, preferably with recorded outputs for long-l term records that provide documentation of the concentration and total l
quantity of radioactive effluents, as required in Section 3.7.2 below.
For reactors operated at power levels below a few hundred kilowatts, the concentrations of airborne radionuclides may be too low to measure during normal operation.
For these, calculated concentrations of released quant < ties are acceptable as specifications, using the SAR as the basis.
(2) Fission Product Monitors The specified fission product monitor could be the CAM or the primary coolant monitor, depending on the release scenarios analyzed in the SAR.
Release of fission products both from fuel and fueled experiments should 21 1
SAR Standard Format and Centcnt Chapter 14 - Draft June 23, 1993 j
Table 4 Typical Required Radiation Measuring Channels
- i Minlaun rumber Set point egaal Channet regaired F metion or less than Area radiation monitors 4
Alarm 0.15 esv/hr 15 nren/hr hot cett monitor 1
Alarm and door Interlock 1 mSv/hr i
100 mram/hr Reactor bridge 1
Alarm (Isolates contatraent 0.5 esv/hr l
with building particulate) 50 mram/hr l
Primary coolant 1
Alarm 0.5 utv/hr f
50 aren/hr i
buildine particulate 1
Atars (Isolates containment 2 x 10* pct / car I
with reactor bridge) 2 hr particulate Building gas (Ar-41) 1 Atars 2 x 10* SCf/ca' Daily reteese i
stack particulate 1
Alarm 2 x 1D* ACI/ca' 2-hr particulate stack pas (Ar 41) 1 Alarm 2 x 10* pcl/cef Daily release 4 x 1D* pCi/ce' i
Annual everage
- As litustrative values, these channels and set points do not apply to any one reactor. Set points for a particular facility must be determined in the SAR analysis, s
be included. This specification may be combined with the specification discussed in Section 3.3 (5) above.
i i
(3) Area Monitors There should be a specification requiring operable area monitors in and near the reactor room. The type of radiation detected, such as gamma i
rays or neutrons, should be specified. Brand name, efficiency, and specific design should be avoided as specifications, but the range of f
exposure rates monitored may be specified. These area monitors should r
provide information on the potential exposure rates from reactor-related radiation. Alarm and automatic action set points should be specified to ensure that personnel exposures arJ ootential doses remain well below limits of 10 CFR Part 20 and are c h istent with the facility ALARA program.
(4) Environmental Monitors There should be at least one environmental monitoring station in the near environment to the facility, preferably at the site boundary or other areas of concern, such as at population centers or student dormitories.
These monitors should be specified to match the types of radiation anti-cipated and should be either in line-of-sight from the air exhaust point J
l l
22 t
i
SAR Standard Fcrmat and Ccnt:nt Chapter 14 - Draft June 23, 1993 s
or down-wind in the prevailing wind, as appropriate. The types of moni-tors should be specified (see Section 3.7.1 (4) of ANSI /ANS 15.1). The location and method of determining background readings should be dis-cussed in the SAR and the basis of the specification. The specification should state that environmental monitors are used to verify that the potential maximum dose, annual or other, in the unrestricted environment is within the values analyzed in the SAR. The specification should address both potential accident scenarios and normal operations.
3.7.2 Effluents All radioactive species listed in Section 3.7.2 of ANSI /ANS 15.1 should be addressed for normal operations, and the releases should be limited by techni-cal specifications. The NRC accepts the proposed concentration limits, pro-vided the SAR shows that potential doses from these concentrations comply with 10 CFR Part 20 for the maximum exposed member of the public on a facility-specific basis, j
1 Ar-41 is the principal radionuclide released by most non-power reactors.
Even though the doses related to Ar-41 are generally small, a specification should address the average and maximum concentrations in both the restricted and I
unrestricted areas and the total curies (becquerels) released during a calen-l dar year. The calculated potential doses to the most exposed persons in restricted and unrestricted areas must conform with 10 CFR Part 20 and the facility ALARA program.
Because of diffusion and dispersion of the release, if the point of release is inaccessible to the public and generally not accessed by restricted area per-sonnel, the maximum normal concentration at that location may be higher than the concentration allowed in 10 CFR Part 20 for restricted areas. The SAR must show that the diffused and dispersed release at the point of contact with members of the public is within 10 CFR Part 20 limits. The calculations in the SAR for diffusion and dispersion should be realistic but conservative, and be based on logical models and specified effluent levels.
Because an infinite cloud assumption is extremely conservative for Ar-41 releases, a finite cloud should be considered, as discussed in NUREG-0851 and accepted by the NRC.
3.8 Experiments Experimental facilities should be described in the SAR, and their basic features should be included in Section 5 (Design Features) of the technical specifications document. The experiments to be performed in the experimental facilities need only be noted briefly, if at all, in the SAR unless they could present a hazard to the reactor facility, the public, or facility staff. Any LCOs for experiments should be performance-based to ensure that no regulations are violated, that experiment safety analysis limits are not exceeded, and that the reactor is not damaged by experiment failure or malfunction.
Regulatory Guide 2.2, " Development of Technical Specifications for Experiments in Research Reactors," provides detailed guidance to applicants on the scope of the discussions for experiments to be included in the SAR. The regulatory guide also provides guidance on the technical specifications needed to govern the experiments performed. The technical specifications should follow the 23
SAR Standard Fermat and Ccntcnt Chapter 14 - Draft June 23, 1993 guidance of Section 3.8 of ANSI /ANS 15.1 and Section C of Regulatory Guide 2.2, as supplemented by the guidance below.
3.8.1 Reactivity Limits Limits should be specified on absolute values of reactivity associated with each type of experiment:
secured, unsecured, and movable (see ANSI /ANS 15.1 and Regulatory Guide 2.2 for definitions). Generally, the limits on secured experiments should be approximately twice the limits on unsecured and movable experiments, where the latter should be no more than I dollar. The 1-dollar limit is such that inadvertent prompt criticality is avoided even if failure of the experiment were to occur. Movable experiments must be clearly defined to include those to be inserted or removed while the reactor is operating.
Unsecured experiments include those installed before reactor startup that change position or other conditions while the re:4ctor continues to operate.
Reactivity limits of experiments -that change position while the reactor is operating must n't exceed the ability of the reactor operator or automatic o
servo system to maintain control of the raactor. The specified reactivity limits on movable experiments should not permit the violation of the shutdown margin specification.
The specified sum of the absolute values of the reactivity worths of all experiments should not be more than twice the limit on individual secured experiments. The value should be consistent with the SAR analysis of inad-vertent reactivity insertions, as explained in Section C.I.a. of Regulatory Guide 2.2.
There should be a specification requiring that the reactor be shut down during the changing or moving of any secured experiment.
3.8.2 Materials For fissile materials in experiments, limits should be specified on the allowed thermal power and on the equilibrium or maximum inventory of specific fission products, such as iodines and strontium. Specifications such as those indicated in Section C.2.a of Regulatory Guide 2.2 are acceptable.
A specification should require double encapsulation of liquid, gas, and potentially corrosive materials. The failure of an encapsulation of material that could damage the reactor should require removal and physical inspection of potentially damaged components.
Specifications should limit the quantity of explosive material permitted in the reactor experimental facilities and in the reactor facility. For experi-mental facilities, the upper limit should be 25 mg TNT or its equivalent, as indicated in Section C.2.d of Regulatory Guide 2.2.
For the overall reactor facility, the upper limit will be no higher than 100 mg TNT or its equivalent unless analyzed in the SAR and approved by NRC. An additional specification should require prior testing or analyses of explosive material encapsulations to ensure no reactor damage in the event of detonation, regardless of the l
limit.
24
)
I l '/
SAR Standard Format and Centent l
Chapter 14 - Draft l
June 23, 1993 s
i:
j y-A specification should limit the quantities of unknown materials that could be placed in certain experimental facilities for exploratory studies. Confor-4 mance with Section C.2.1 of Regulatory Guide 2.2 would be acceptable.
3.8.3 Failure and Malfunctions l
-Specifications that address the failure and malfunction of an experiment and limit the experiment parameters should be included on a case-by-case basis, as discussed in the SAR. The guidance of Section 3.8.3(2) of ANSI /ANS 15.1 1
should be followed, but specifications that require compliance with regu-l lations are redundant and are not necessary (see Section 3.8.3(1) of ANSI /ANS
(
15.1).
i i
For experiments that may off-gas, sublime, volatilize, or produce aerosols, i
standard assumptions are often specified for calculating the activity that could be released under normal operating conditions, accident conditions in
]
the reactor, and accident conditions in the experiment. Such specifications ensure conservatism in the safety analysis of the experiment. These specifi-J cations have included such assumptions as (1) if an experiment falls and i
j releases radioactive gases or aerosols to the reactor bay or atmosphere, 100 l
percent of the radioactive gases or aerosols escape; (2) if an effluent holdup 1
' tank isolates on a high radiation signal, at least 10 percent of the radioac-
)
tive gases or aerosols escape; (3) if the effluent exhausts through a filter j
i with 99 percent efficiency for 0.3 micron particles, at least 10 percent of j
the vapors escape; and (4) if an experiment fails that contains materials with j
a boiling point above 130 *F (54 *C), the vapors of at least 10 percent of the 1
l materials escape through an undisturbed column of water above the core. The
]
particular assumptions used, if any, must be derived from the SAR.
1 j
Applicable limits for specific experiments are normally not part of the technical specifications and should be derived from the experiment safety review discussed in Section 6.5 below.
j 3.9 Facility-Specific LCOs The LCOs discussed above apply to most non-power reactors..Each reactor may also have technical specifications containing facility-unique LCOs. These j
should be based in the SAR and facility design.
}
4 Surveillance Requirements Certain LCOs established in Section 3 of the technical specification document should be accompanied by a surveillance requirement in Section 4.
These surveillance-related specifications should clearly identify the parameter or j
function to be measured or tested, the method, the frequency, and the accept-able deviation or error. Acceptable deviations might be limited by license-conditions (such as thermal power level) or by regulations.(such as 10 CFR '
l Part 20).
The NRC accepts the surveillance frequencies stated in this section of ANSI /
ANS 15.1 as amplified in the following sections. The actuk1 wording of the specifications should not be ambiguous. Wording in ANSI /ANS 15.1 has been interpreted incorrectly by some licensees to allow the extended interval
!j 25 4
SAR Standard Format and Content l
Chapter 14 - Draft June 23, 1993 j
j t
(interval not to exceed statement) as the average.
If the extended interval j
is used for a particular surveillance test, a shorter interval should be used j
as soon afterwards as possible to adhere to the average. The intervals of ANSI /ANS 15.1 should be explicitly listed in the applicant's technical l
specifications.
i In addition to surveillance verification of LCOs, other surveillance activi-ties should be specified. These include specific specifications, such as l
periodic pulse rod maintenance and cleaning, thermal power level calibration, preventive maintenance and inspection of control / safety rod drive systems, j
fuel element inspections, preventive maintenance on omer important components to provide assurance of operability, and calibration of effluent monitoring j
systems.
l If a surveillance specification is not required for safety while the reactor l
is shut down, it may he deferred, but must be perfomed before reactor startup.
If the reactor is not to be operated in a particular mode (e.g...
pulse mode) for an interval of time that exceeds the surveillance intervals for that particular mode, surveillance specifications not required for safety (an example is the requirement for a standard pulse to be performed every year) while the reactor is operated in other modes may be deferred, but must be performed before the reactor is considered operational in the mode in which surveillances were deferred. Scheduled surveillance that cannot be performed while the reactor is operating may be deferred until the next planned reactor shutdown. Surveillances that may be deferred and the reasons for deferment l
must be clearly stated in the technical specification, justified in the SAR, and noted in the basis of the specification.
In general, any time that a reactor system or component is modified or repaired, the surveillance requirements for that system should be performed as part of the operability check of the system or component. This should be done regardless of when the surveillance was last performed or when it is next due.
l This special surveillance may change the due date of the next regularly scheduled surveillance of that type.
l 4.1 Reactor Core Parameters The excess reactivity and shutdown margin LCOs specified in Section 3 of the technical specification document are applicable for all authorized operating conditions. As an example, for a movable experiment, the specifications for excess reactivity and shutdown margin surveillance mea.surements should be based on that experiment being in its most reactive location.
In addition, other reactor parameters that affect reactivity during operation should be explicitly specified.
For the following specified surveillance requirements, j
the parameters may be determined by an appropriate' combination of measurements j
and calculations.
g 1
(1) Excess React M ty l
l Excess reactivity should be determined at'least annually and after changes in either the core, in-core experiments, or control rods for
)
which the predicted change in reactivity _ exceeds the absolute value of the specified shutdown margin.
26
-l i
i
^
SAR Standard Fermat cnd Centent Chapter 14 - Draft June 23, 1993 (2) Shutdown Margin The shutdown margin should be determined at least annually and after changes in either the core, in-core experiments, or control rods.
(3) Pulse limits (Added by NRC)
The relationship between peak fuel temperature and inserted reactivity for pulses should be determined when changes in the core (see item 1 above) are made.
l l
(4) Core Configuration (Added by NRC)
Limitations on core configurations are intended to ensure that reactor physics and thermal-hydraulic parameters specific to the core are within the limits analyzed in the SAR. Core configuration parameters specified in item (4) of Section 3.1 or in Section 5 of the technical specifica-tions document must be met during reactor operations. Therefore, an l
acceptable surveillance specification is to verify compliance with all applicable specifications in those sections when any change occurs in the reactor core configuration.
(5) Reactivity Coefficients (Added by NRC)
Item 5 of Section 3.1 of the technical specifications document will limit reactivity coefficients, which are largely determined by reactor design and fuel type. Measuring and verifying reactivity coefficients can be a difficult task. An acceptable schedule for surveillance of reactivity coefficients is at initial reactor startup and when any change in the reactor core configuration or fuel type requires changes in the specifications of Section 5.
(6) Fuel Parameters (Added by HRC)
All TRIGA fuel should be inspected for damage and all TRIGA non-instrumented fuel measured for length and bend at the following frequencies:
For non-pulsing TRIGA reactors, the fuel should be inspected and i
measured on at least a 5-year cycle. Approximately 20 percent of l
the fuel can be inspected and measured annually.
If an annual inspection identifies damaged fuel, then the entire core should be l
inspected and measured.
For pulsing TRIGA reactors, the fuel should be inspected and measured annually.
If the reactor is pulsed infrequently (less than 10 pulses annually), the annual inspection requirement may be
{
l relaxed if analyzed and justified in the SAR.
If the reactor is pulsed to reactivity insertions over 4 dollars, additional inspec-i tion requirements based on the number of pulses may be necessary.
Facilities in this situation should present and justify inspection frequency requirements determined by the fuel vendor.
27 i
l
SAR Standard Format and Ccnt::nt Chapter 14 - Draft June 23, 1993 Routine inspections of fuel used in AC*i-201 and PULSTAR reactors have not been required by technical specification.
If the opportunity is pre-sented to conduct an inspection of fuel, such as core disassembly of an AGN-201 or disassembly of a PULSTAR fuel element to replace fuel pins, the licensee should consider taking advantage of this opportunity. This type of insp,ection need not be a technical specification requirement.
NRC may require fuel inspection as a license condition to increase burnup limits on such fuels. This would be determined on a case-by-case basis.
These inspections are qualitative, to detect evidence of excessive corrosion / erosion and mechanical wear or damage.
Inspections for reactors with plate fuels have not been required by technical specification except for higher power reactors that frequently refuel. However, for reactors that remove plate fuel from service because the fuel has reached its burnup limit, there should be & require-ment tt inspect representative fuel elements (e.g.,1 in every 10) for excessive corrosion / erosion, mechanical wear or damage, or plate swel-ling. The surveillance procedures should follow guidance provided by he fuel supplier, if available.
In all cases, the specification should describe briefly how the inspection will be performed.
For reactors with technical specification limits or SAR analyses imposing limits on uranium burnup or fission density, confirmatory estimates should be made at intervals during the life of the fuel, such as at 50, 60, and 70 percent of the fuel life or semiannually when the NRC/ DOE Form 742 is submitted. The SAR should justify the surveillance method and intervals that ensure that the limit is not exceeded.
For reactors with HEU fuel and subject to 10 CFR 73.6(b) exemption (self-protection), confirmatory radiation measurements or analyses should be made at intervals justified in the SAR.
4.2 Reactor Control and Safety Systems (1) Reactivity Worth of Control Rods The integral and differential worths of all control and safety rods will be determined at initial fuel loading.
Integral and differential worths will be determined at least annually and after changes of the core or control rods, as noted in item 1 of Section 4.1 above.
(2) Rod Withdrawal and Insertion Speeds ANSI /ANS 15.1 acceptable.
(3) Transient Rod and Associated Mechanism This system will be inspected, disassembled, cleaned, and if applicable, t
lubricated annually. The reactor should be pulsed at least annually and after changes to core or control rods, as indicated in item 1 of Sec-tion 4.1 above, with a well-documented reactivity insertion.
If the reactor is not routinely pulsed, this standard pulse (a reactivity 28 i
SAR Standard Format and C';ntent Chapter 14 - Draft June 23, 1993 addition whose results are well know) may be deferred for more than a year, but should be performed before resumption of normal pulsing.
(4) Scram Times of Control and Safety Rods A specific interval should be stated for the surveillance intervals given in item 4 of Section 4.2 of ANSI /ANS 15.1.
(5) Scram and Measuring Channels Channel tests of all scram and power measuring channels required by technical specifications, including scram actions with safety rod release and interlocks, are required before each reactor startup following a shutdown of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or each secured shutdown.
If the reac-tor operating schedule calls for no secured shutdowns, the channel tests should be performed at least quarterly. Many facilities perform these tests before each reactor startup and NRC recommends this practice.
(6) Operability Tests ANSI /ANS 15.1 acceptable.
(7) Thermal Power Calibration for Forced Convection Cooled Reactors Thermal power calibration should be performed at least annually, with a heat balance verification at least monthly.
(8) Thermal Power Calibration for Reactors not Cooled by Forced Convection Thermal power calibration should be performed at least annually. The basis should indicate the method to be used.
i (9) Rod Inspection (Added by NRC) i The rod-drive and scram mechanisms of each control and safety rod should be inspected annually. The poison sections of control and safety rods should be inspected biennially for indications of deterioration or damage. This can be a visual inspection or an inspection that requires the rod to pass through a measuring device that detects swelling.
4.3 Coolant Systems Only a small fraction of the licensed non-power reactors will need to consider all of the following surveillance items. The SAR should discuss the applicability and identify the parameters that should be tested. The applicable parameters should be specified, and the functions should be explicitly stated in the specification.
(1) Starting Function of Energency Shutdown and Susp Pumps ANSI /ANS 15.1 acceptable.
29
l SAR Standard Format and Ccntcnt Chapter 14 - Draft June 23, 1993 (2) Test of Energency Coolant Sources and Systens ANSI /ANS 15.1 acceptable.
(3) Inservice Inspections If any inservice inspections of cooling system components are identified and required in the SAR, they shall be performed according to manufac-turer's recommendations.
If the manufacturer's recommendation is not available, the frequency should be as established in the SAR from engineering judgement and similar component inservice inspection requirements and experience.
(4) Analysis of Coolants for Radioactivity Analyses for isotope identification of primary and, if applicable, secondary coolant, should be performed by sampling quarterly. Sampling weekly for gross analysis should be considered to establish trends to quickly identify fuel or heat exchanger failure.
(5) Hydrogen Concentration in Off-Gas If applicable, this test should be performed at least annually and after i
maintenance or repair that could affect the system or measurement instrumentation.
(6) Conductivity and pH When the reactor is operating on a routine schedule, conductivity and pH should be measured at least weekly. This requirement could be met by a system that monitors conductivity and pH continuously while the reactor is operating.
If the reactor is not operated for long periods, the interval between conductivity and pH measurements may be increased to monthly if reasonable justification is provided in the SAR.
If fuel is stored in water in separate fuel storage from the reactor pool, the pH and conductivity of this water should be measured at regular intervals as determined and justified in the SAR.
l (7) Primary Coolant level (Added by HRC)
If the primary coolant level above the core is not continuously displayed i
during operation, the primary coolant level in the pool or tank should be verified daily if the reactor is operating or before reactor startup.
1 (8) Primary Coolant Sensors and Channels (Added by NRC)
Channel tests of sensor operability and channels not included elsewhere in the technical specifications document that are identified in the SAR should be performed quarterly and before startup after maintenance.
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SAR Standard Format and Cont;nt Chapter 14 - Draft June 23, 1993 All channels should be calibrated annually and before startup after major modification or component replacement.
4.4 Containment or Confinement 4.4.1 Containment Few licensed non-power reactors are required to provide a containment system.
For those required by the SAR, the surveillance intervals given in ANSI /ANS 15.1 are acceptable.
4.4.2 Confinement Confinement is a system that provides a temporary holdup or controlled release of radioactive effluents to the environment. Most non-power reactors are equipped with confinements and should have a functional test of the overall system oescribed in the SAR quarterly.
In addition, an. efficiency test of the filters should be pe: formed annually or in accordance with manufacturer recomendations and acceptance criteria.
4.5 Ventilation Systems Ventilation systems at most licensed non-power reactors are an integral part of the containment or confinement system and surveillance activities may be interrelated. An operability check, including dampers and blowers, should be performed quarterly and following repair or maintenance to declare the system operable. A functional and efficiency test of filters should be performed annually or in accordance with manufacturer recomendations and acceptance criteria and following repair or maintenance to declare the system operable.
4.6 Emergency Electrical Power For all emergency electrical power systems, channel checks or other operabil-ity checks should be performed before reactor startup and after maintenance.
Maintenance should be performed according to the manufacturer's recommendations.
If the manufacturer's recommendation is not available, the frequency should be as determined in the SAR.
4.6.1 Diesels and Other Devices The shorter of the surveillance intervals given in ANSI /ANS 15.1 is acceptable.
4.6.2 Emergency Batteries 4
The shorter of the surveillance intervals given in ANSI /ANS 15.1 is acceptable.
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SAR Standard Fcrmat and Ccntent Chapter 14 - Draft June 23, 1993
- 4.7 Radiation Monitoring Systems and Effluents 4.7.1 Monitoring Systems A channel check should be performed daily before reactor startup. Where physically possible, a channel test using a r diation source should be performed at least monthly. The SAR should describe such capability.
All required radiation monitoring systems, including effluent monitors, should be calibrated at least annually according to the manufacturer's recommenda-tions.
Individual systems should have separate specifications.
4.7.2 Effluents Quantities of radioactive effluents released to the environment are LCOs.
If the SAR indicates that it is not feasible to monitor such effluents from low power reactors in real time at the point of release, calculated releases may be substituted. The SAR should specify surveillance methods and intervals for confirming these releases or for verifying upper limits.
For gaseous airborne radioactive effluent, confirmation of annual upper limits by integrating dosimeters such as thermoluminescent dosimeters (TLDs) or film is acceptable.
For particulate airborne or waterborne radioactive effluent, confirmation of annual upper limits by surveillance of environmental factors given in Section 4.7.2 of ANSI /ANS 15.1 is acceptable.
4.8 Experiments If any experiment discussed in the SAR is designed to operate with emergency systems or with connections to the reactor protective systems, a channel check should be specified both daily and before reactor startup when the particular experiment is being performed.
Surveillance activities for experiments that are included in the experiment protocol and the review and approval process need not be included explicitly in the technical specifications.
4.9 Facility-Specific Surveillance There should be applicable surveillance specifications for any facility-specific LCOs in Section 3.9 of the technical specifications document not explicitly included in Section 4.
These surveillances should be to verify significant safety features from the SAR.
5 Design Features
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The SAR forms the basis for NRC to issue an operating license for a non-power reactor. Essential information includes the type and enrichment of fuel, core and fuel configurations, fuel storage facilities, thermal power level, potential accident scenarios and mitigating features, environmental conditions at the site, and other factors. To ensure that the issued license remains valid, design features should not be changed without prior NRC review and approval. These major design features are provided in Section 5 of the 32
SAR Standard Format and Ccntcnt Chaptcr 14 - Draft June 23, 1993 technical specifications document, if they have not already been specified in Sections 2 or 3.
The NRC accepts the guidance in this section of ANSI /ANS 15.1. The applicant should provide concise but explicit information on all noted features.
6 Administrative controls The specified information and controls on staffing and operations of the reactor facility will ensure that the management and staff of the facility are acceptably knowledgeable and aware of the technical requirements to operate a safe facility, to comply with regulations cnd the license conditions, and to practice a meaningful ALARA program, which will protect the health and safety of the public, the facility users, and the staff.
Not all owners and operators of non-power reactors will have the same management organi'zation or office titles.
Regardless of the details of the l
management organization, or of ihe complexity of the facility, the administra-tive functions presented in this section of ANSI /ANS 15.1 must be established and specified. The NRC accepts the ANSI /ANS 15.1 position as modified and amplified in the following sections.
6.1 Organization 6.1.1 Structure The information requested by ANSI /ANS 15.1 should be clearly stated, including how and when the radiation safety staff communicates with the facility manager and level 1 management to resolve safety issues.
6.1.2 Responsibility Follow ANSI /ANS 15.1.
6.1.3 Staffing Applicants should use the terms reactor operator (RO) and senior reactor operator (SRO) instead of Class B and Class A, respectively (see Figure 1 in ANSI /ANS 15.1).
6.1.4 Selection and Training of Personnel Compliance with 10 CFR Part 55 is required of the licensee and licensed operators, unless NRC has issued an exemption. ANSI /ANS 15.4-1988, " Selection and Training of Personnel for Research Reactors," provides additional guidance for non-power reactors.
6.2 Review and Audit l
The committee established for the review function may be assigned approval authority by the facility manager or the facility manager may retain it. Sec-tion 6.2 of the technical specifications document should explicitly state who holds the approval authority and should specify the committee's authority and 33
SAR Standard Fcrmat and C nt:nt Chapter 14 - Draft June 23, 1993 -
how it comunicates and interacts with management levels I and 2.
There must be a qualified independent review comittee.
6.2.1 Composition and Qualifications One or more voting members of the comittee should be from organizations other than that operating the reactor.
6.2.2 Charter and Rules Follow ANSI /ANS 15.1.
6.2.3 Review Function The fact that this section of ANSI /ANS 15.1 addresses the review function required by 10 CFR 50.59 should be explicitly stated in the specifications.
6.2.4 Audit Function In addition to the emergency plan, all required plans, such as physical security and operator requalification, should be specified for auditing. The requirement to audit these plans may be part of the plan itself.
If that is the case, the requirement to audit does not need to be repeated in the technical specifications.
6.3 Radiation Safety The technical specifications should clearly state that 10 CFR Part 20 establishes requirements that the radiation safety program must achieve.
Additional guidance for radiation safety programs at non-power reactors may be found in ANSI /ANS 15.11, " Radiological Protection of Research Reactor Facili ti es.
The authority of the safety staff to interdict or terminate safety-related activities should be specifically stated. An explicit statement in the tech-nical specifications should state management's comitment to practice an effective ALARA program. This program should apply to facility staff, facility users, general public, and the environment.
6.4 Procedures Procedures in addition to those identified in ANSI /ANS 15.1 should be required at facilities to address operational situations recognized in the SAR.
For example, if byproduct material whose possession is authorized under the reactor license is used in facility laboratories that are part of the reactor license and/or transferred to other licensees, procedures for control and transfer of this byproduct material should be part of the set of minimum procedures required by the technical specifications. The specifications should be written to ensure a minimum necessary set of procedures, but allowing for future additions as necessary.
i The minor modifications and temporary deviations allowed by ANSI /ANS 15.1 should not be spelled out in the technical specifications. However, the 34 1
SAR Standard Fermat and Centcnt Chapter 14 - Draft
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June 23, 1993 methodology for establishing and changing procedures should be stated in the specifications.
6.5 Experiments Review and Approval In addition to guidance of ANSI /ANS 15.1, the ieview and approval of experi-ments should be consistent with the guidance provided in Section C.3 of Regu-latory Guide 2.2 and Regulatory Guide 2.4, " Review of Experiments for Research Reactors." The specifications should make clear that " established and approved procedures" means written procedures, properly reviewed and approved.
Changes to these procedures should follow Section 6.4 of ANSI /ANS 15.1.
6.6 Required Actions 6.6.1 Action to Be Taken in Case c.' Safety Limit Violation Follow ANSI /ANS 15.1.
6.6.2 Action To Be Taken in the Event of an Occurrence of the Type Identified in Sections 6.7.2(1)(b) and 6.7.2(1)(c) t The first sentence of this section of ANSI /ANS 15.1 says, " reactor conditions shall be returned to normal or the reactor shall be shut down." The specification should be written to provide that the licensee establish, in advance, specific criteria for the two alternate actions:
shutdown or return to normal. For example, a return-to-normal event is a reactor scram resulting from a known cause, such as an electric transient.
6.7 Reports 6.7.1 Operating Reports The technical specifications should state explicitly that operating reports should be sent to the NRC Document Control Desk and a copy to the appropriate regional administrator annually.
Section 6.7.l(4) of ANSI /ANS 15.1 refers to the reporting required by 10 CFR 50.59. The specification should make reference to the rule.
6.7.2 Special Reports The technical specifications should state that special written reports of events should be sent to the NRC Document Control Desk and a copy to the appropriate regional administrator and special telephone reports should be of events made to the NRC Operations Center and the regional staff.
6.8 Records Follow ANSI /ANS 15.1.
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REFERENCES American National Standards Institute and American Nuclear Society, ANSI /ANS 15.1-1990, "The Development of Technical Specifications for Research Reactors," LaGrange, Illinois.
--, ANSI /ANS 15.4-1988, " Selection and Training of Personnel for Research Reactors."
--, ANSI /ANS 15.11-1987, " Radiological Protection at Research Reactor Facilities."
Beeston, J. M., R. R. Hobbins, G. W. Gibson, and W. C. Francis, " Development and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors,"
Nuclear Technoloov 49, pp. 136-149, June 1980.
Gibson, G. W., The Development of Powdered Uranium-Aluminide Compounds for Use as Nuclear Reactor Fuels, IN-ll33, Idaho Nuclear Corporation, Idaho Falls, Idaho, December 1967.
l Nazare, S., G. Ondracek, and F. Thummler, " Investigations on UAlx-Al Dispersion Fuels for High-Flux Reactors," Journal of Nuclear Materials 56, pp. 251-259, 1975.
l Simnad, M.
T., et al., " Fuel Elements for Pulsed TRIGA Research Reactors,"
Nuclehr Technoloav. 28, p. 31,1976.
--, " Interpretation of Damage to the FLIP Fuel During Operation of the Nuclear Science Center Reactor at Texas A&M University," GA-A16613, December 1981.
Simnad, M.
T., and G. B. West, "Postirradiation Examination and Evaluation of TRIGA LEU Fuel Irradiated in the Oak Ridge Research Reactor," GA-A18599, GA Technologies, Inc., San Diego, California, May 1986.
Stahl, D., Fuels for Research and Test Reactors. Status Review, ANL-85-5, Argonne National Laboratory, Argonne, Illinois, December 1982.
U.S. Nuclear Regulatory Commission, Regulatory Guide 2.2, " Development of Technical Specifications for Experiments in Research Reactors," November 1973.
--, Regul? tory Guide 2.4, " Review of Experiments for Research Reactors,"
Revision 0-R, July 1976.
--, NUREG-0851, "Nomoarams for Evaluation of Doses From Finite Noble Gu Clouds," January 1983.
--, NUREG-1282, Safety Evaluation Reoort on Hiah-Uranium Content. Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors, August 1987.
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--, NUREG-1313, Egfety Etpjuation Renart Related to the Evaluation of Low-i Epriched Uranium Sf,ticide-Aluminum Dispersion Fuel for Use in Nonoower-j 4
Reactors, July 1988.
I i
1 West, G. G., M. T. Simnad, and G. L. Copeland, " Final Rs.sults From TRIGA LEU Fuel Postirradiation Examination and Evaluation Following Long-Tern l
Irradiation Testing in the ORR,' presented at the International Meetina on Reduced Enrichment for Research and Test Reactors. Gat 11nbura. Tennessee.
l November 3-6. 1985.
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June 22, 1993 14 TECHNICAL SPECIFICATIONS 14.1 Introduction This section provides guidance for NRC staff review and acceptance of the technical specifications for non-power reactors.
The format for the acceptance criteria (Section 14.6) follows the format of ANSI /ANS 15.1-1990, " Development of Technical Specifications for Research Reactors."
In addition to providing the information specified in ANSI /ANS 15.1, the technical specifications must be consistent with 10 CFR 50.34 and must address all applicable paragraphs of 10 CFR 50.36. The content of technical specifications for non-power reactors are considerably simpler than that for power -eactors consistent with the difference in size and complexity between non-power reactors and power reactors. Maintaining system performance should provide the basis for the technical specifications of non-power reactors.
By addressing limiting or enveloping conditions of design and operation, emphi. sis is placed on ensuring the safety of the public, the facility staff, and the environment.
Because the performance-based concept is used for non-power reactors, standardization is possible across the entire set of technical spec.ification parameters, even for the diverse types of non-power reactors.
ANSI /ANS 15.1 provides the parameters and operating characteristics of a non-power reactor that should be included in the technical specifications.
Because of the wide diversity of non-power reactor designs and operating characteristics, some items may not be applicable to all facilities.
In addition, experience has shown that some of the factors included in ANSI /ANS 15.1 must be explained and supplemented. The NRC staff discusses these factors in the " Format and Content for Applications for the Licensing of Non-power Reactors" (format and content guide), NUREG-XXXX. Guidance is provided herein for the NRC staff who review non-power reactor technical specifications against the requirements of ANSI /ANS 15.1 and the format and content guide.
14.2 Areas of Revigw Under 10 CFR 50.36, every operating license for a nuclear reactor must include technical specifications that state the limits, operating conditions, and
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other requirements for facility operation. These specifications are designed to protect the health and safety of the public.
For non-power reactors, the NRC recommends conformance to ANSI /ANS 15.1 and the format and content guide.
j 14.3 Acceptance Criterja Technical specifications will satisfy 10 CFR 50.34 and 10 CFR 50.36 if they adequately address the issues and parameters of ANSI /ANS 15.1 as supplemented in the format and content guide for non-power reactors.
In addition, an efficient review will be promoted if the technical specifications are standardized and consistent with this guidance. Additional guidance may be found in previously approved technical specifications for non-power reactors 1
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SAR Revicw Plan and Acceptance Criteria Chapter 14 - Draft June 22, 1993 d
of similar design, operating characteristics, site and environmental conditions, and use.
14.4 Review Procedures The reviewer will compare the proposed technical specifications with ANSI /ANS 15.1, as supplemented in the format and content guide; with previously accepted technical specifications for similar facilities; and with the facility SAR.
The technical specifications and basis are determined from the analysis in the SAR. The reviewer will confirm that the technical specifications are complete and follow the correct format. This review is part of the process to review and approve the technical specifications. The reviewer will confirm that each technical specification is supported by appropriate reference to SAR analysis and statements. The NRC review of those SAR chapters should sup;;rt the finding that each of the technical specifications are acceptable to NRC.
14.5 Evaluation Findinos The reviewer will determine that all conditions that provide reasonable assurance that the facility will function as analyzed in the SAR are in the technical specifications and that the technical specifications are in an acceptable format.
Further, in conjunction with the findings made in other chapters of the SER, the reviewer will determine that limits for the facility design, construction, and operation are acceptable because they provide reasonable assurance that the facility can be operated without endangering the health and safety of the public, the environment, or the facility staff.
The following statement is the type that the NRC staff evaluation should support in the SER:
The staff has evaluated the applicant's (licensee's) technical specifications in this licensing action. These technical specifications define certain features, characteristics, and conditions governing the operation of the (name) facility and are explicitly included in the (renewal) license as Appendix A.
The staff has reviewed the format and contents of the technical specifications using the guidance of ANSI /ANS 15.I-1990,
" Development of Technical Specifications for Research Reactors," and NUREG-XXXX, " Format and Content for Applications for the Licensing of Non-Power Reactors."
On the basis of its review, the staff finds the technical specifications acceptable and concludes that normal plant operation within the limits of the technical specifications will not result in offsite radiation exposures in excess of 10 CFR Part 20 guidelines and provide reasonable assurance that the facility will function as analyzed in the safety analysis report.
Furthermore, the technical specifications will limit the likelihood of malfunctions and mitigate the consequences to the public of off-normal or accident events.
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- .i 14.6 Acceptance Criteria - Technical Soecifications for Non-nower Reactors The technical specifications acceptance criteria'are presented in.Section 14.2 i
of the format and content guide.
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