ML21257A262
| ML21257A262 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/07/2021 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21278A102 | List:
|
| References | |
| NEI 99-04 | |
| Download: ML21257A262 (14) | |
Text
CALVERT CLIFFS UFSAR 1C.1-1 Rev. 50 APPENDIX 1C 1C.0 AEC PROPOSED GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS 1C.1 INTRODUCTION On July 10, 1967, the Atomic Energy Commission published the proposed General Design Criteria (GDC) for Nuclear Power Plants. These 70 criteria were issued for public comment. In 1971, the final version of the GDC was incorporated into regulation, as Appendix A to 10 CFR Part 50. This version contained 64 criteria.
Baltimore Gas and Electric Company obtained a construction permit on July 7, 1969, to construct both Calvert Cliffs Units 1 and 2. The construction permit was obtained from the Atomic Energy Commission after their review of our Preliminary Safety Analysis Report which contained an assessment of our compliance with the draft GDC.
In 1971, Baltimore Gas and Electric Company submitted the Final Safety Analysis Report for Calvert Cliffs Units 1 and 2. Included in the Final Safety Analysis Report was an assessment of the plant design against the draft GDC. That assessment has not been updated since 1974.
For convenience, the draft GDC are reproduced in this appendix as they appeared in 1967. The 1974 assessment of plant design against these draft GDC is not included since it is significantly out-of-date. Calvert Cliffs was designed and constructed to meet the intent of the draft GDC.
As described in SECY-92-223 (Reference 1), the Nuclear Regulatory Commission policy requires no backfit of the published GDC to plants whose construction permit was issued prior to May 21, 1971. Modifications to the facility are evaluated in accordance with 10 CFR 50.59 to assess consistency with the current licensing basis (including the draft GDC, as applicable).
In 1996, Baltimore Gas and Electric Company completed the addition of a new safety-related Societie Alsacienne De Construction Mechaniques Del Melhouse (SACM) diesel generator.
This emergency diesel generator was installed to comply with the Station Blackout Rule (10 CFR 50.63). Calvert Cliffs Nuclear Power Plant compliance with the Station Blackout Rule is described in Section 1.8.2. The new safety-related diesel generator (Emergency Diesel Generator 1A), its support systems, and the building which houses it, were designed to the GDC that appear in Appendix A to 10 CFR Part 50. The specific design criteria that applies to Emergency Diesel Generator 1A and its support systems are:
Criterion 1 Criterion 17 Criterion 2 Criterion 18 Criterion 3 Criterion 44 Criterion 4 Criterion 45 Criterion 5 Criterion 46 Criterion 13
CALVERT CLIFFS UFSAR 1C.2-1 Rev. 47 1C.2 OVERALL PLANT REQUIREMENTS CRITERION 1 - QUALITY STANDARDS (Category A)
Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified.
Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures, and inspection acceptance levels used is required.
CRITERION 2 - PERFORMANCE STANDARDS (Category A)
Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.
CRITERION 3 - FIRE PROTECTION (Category A)
The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features.
CRITERION 4 - SHARING OF SYSTEMS (Category 4)
Reactor facilities shall not share systems or components unless it is shown that safety is not impaired by the sharing.
CRITERION 5 - RECORDS REQUIREMENTS (Category A)
Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor.
CALVERT CLIFFS UFSAR 1C.3-1 Rev. 47 1C.3 PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS CRITERION 6 - REACTOR CORE DESIGN (Category A)
The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.
CRITERION 7 - SUPPRESSION OF POWER OSCILLATIONS (Category B)
The core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.
CRITERION 8 - OVERALL POWER COEFFICIENT (Category B)
The reactor shall be designed so that the overall power coefficient in the power operating range shall not be positive.
CRITERION 9 - REACTOR COOLANT PRESSURE BOUNDARY (Category A)
The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.
CRITERION 10 - CONTAINMENT (Category A)
Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.
CALVERT CLIFFS UFSAR 1C.4-1 Rev. 47 1C.4 NUCLEAR AND RADIATION CONTROLS CRITERION 11 - CONTROL ROOM (Category B)
The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10 CFR Part 20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause.
CRITERION 12 - INSTRUMENTATION AND CONTROL SYSTEMS (Category B)
Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.
CRITERION 13 - FISSION PROCESS MONITORS AND CONTROLS (Category B)
Means shall be provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of the core, such as indication of position of control rods and concentration of soluble reactivity control poisons.
CRITERION 14 - CORE PROTECTION SYSTEMS (Category B)
Core protection systems together with associated equipment shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.
CRITERION 15 - ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS (Category B)
Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features.
CRITERION 16 MONITORING REACTOR COOLANT PRESSURE BOUNDARY (Category B)
Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage.
CRITERION 17 - MONITORING RADIOACTIVITY RELEASES (Category B)
Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions.
CRITERION 18 - MONITORING FUEL AND WASTE STORAGE (Category B)
Monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures.
CALVERT CLIFFS UFSAR 1C.5-1 Rev. 47 1C.5 RELIABILITY AND TESTABILITY OF PROTECTION SYSTEMS CRITERION 19 - PROTECTION SYSTEMS RELIABILITY (Category B)
Protection systems shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed.
CRITERION 20 - PROTECTION SYSTEMS REDUNDANCY AND INDEPENDENCE (Category B)
Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation components.
CRITERION 21 - SINGLE FAILURE DEFINITION (Category B)
Multiple failure resulting from a single event shall be treated as a single failure.
CRITERION 22 - SEPARATION OF PROTECTION AND CONTROL INSTRUMENTATION SYSTEMS (Category B)
Protection systems shall be separated from control instrumentation systems to the extent that failure or removal from service of any control instrumentation system component or channel, or of those common to control instrumentation, and protection circuitry, leaves intact a system satisfying all requirements for the protection channels.
CRITERION 23 - PROTECTION AGAINST MULTIPLE DISABILITY FOR PROTECTION SYSTEMS (Category B)
The effects of adverse conditions to which redundant channels or protection systems might be exposed in common, either under normal conditions or those of an accident, shall not result in loss of protection function.
CRITERION 24 - EMERGENCY POWER FOR PROTECTION SYSTEMS (Category B)
In the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems.
CRITERION 25 - DEMONSTRATION OF FUNCTIONAL OPERABILITY OF PROTECTION SYSTEMS (Category B)
Means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has occurred.
CRITERION 26 - PROTECTION SYSTEMS FAIL-SAFE DESIGN (Category B)
The protection systems shall be designed to fail into a safe state or into a state established as tolerable on a defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or adverse environments (e.g., extreme heat or cold, fire, steam, or water) are experienced.
CALVERT CLIFFS UFSAR 1C.6-1 Rev. 47 1C.6 REACTIVITY CONTROL CRITERION 27 - REDUNDANCY OF REACTIVITY CONTROL (Category A)
At least two independent reactivity control systems, preferably of different principles, shall be provided.
CRITERION 28 - REACTIVITY HOT SHUTDOWN CAPABILITY (Category A)
At least two of the reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits.
CRITERION 29 - REACTIVITY SHUTDOWN CAPABILITY (Category A)
At least one of the reactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided.
CRITERION 30 - REACTIVITY HOLDDOWN CAPABILITY (Category B)
At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies.
CRITERION 31 - REACTIVITY CONTROL SYSTEMS MALFUNCTION (Category B)
The reactivity control systems shall be capable of sustaining any single malfunction, such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.
CRITERION 32 - MAXIMUM REACTIVITY WORTH OF CONTROL RODS (Category A)
Limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.
CALVERT CLIFFS UFSAR 1C.7-1 Rev. 47 1C.7 REACTOR COOLANT PRESSURE BOUNDARY CRITERION 33 - REACTOR COOLANT PRESSURE BOUNDARY CAPABILITY (Category A)
The reactor coolant pressure boundary shall be capable of accommodating without rupture, and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.
CRITERION 34 - REACTOR COOLANT PRESSURE BOUNDARY RAPID PROPAGATION FAILURE PREVENTION (Category A)
The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagation type failures. Consideration shall be given (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) the provisions for control over service temperature and irradiation effects which may require operational restrictions.
CRITERION 35 - REACTOR COOLANT PRESSURE BOUNDARY BRITTLE FRACTURE PREVENTION (Category A)
Under conditions where reactor coolant pressure boundary system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120F above the nil-ductility transition temperature (NDTT) of the component material if the resulting energy release is expected to be absorbed by plastic deformation or 60F above the NDTT of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.
CRITERION 36 - REACTOR COOLANT PRESSURE BOUNDARY SURVEILLANCE (Category A)
Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming with American Society for Testing and Materials (ASTM)
E-185-66 shall be provided.
CALVERT CLIFFS UFSAR 1C.8-1 Rev. 47 1C.8 ENGINEERED SAFETY FEATURES A.
GENERAL REQUIREMENTS CRITERION 37 - ENGINEERED SAFETY FEATURES BASIS FOR DESIGN (Category A)
Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary, assuming unobstructed discharge from both ends.
CRITERION 38 - RELIABILITY AND TESTABILITY OF ENGINEERED SAFETY FEATURES (Category A)
All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the system, including engineered safety features, will be influenced by the known and the demonstrated performance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of the plant.
CRITERION 39 - EMERGENCY POWER FOR ENGINEERED SAFETY FEATURES (Category A)
Alternate power system shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning required of the engineered safety features. As a minimum, the onsite power system and the offsite power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system.
CRITERION 40 - MISSILE PROTECTION (Category A)
Protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures.
CRITERION 41 - ENGINEERED SAFETY FEATURES PERFORMANCE CAPABILITY (Category A)
Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
CRITERION 42 - ENGINEERED SAFETY FEATURES COMPONENTS CAPABILITY (Category A)
Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident (LOCA).
CRITERION 43 - ACCIDENT AGGRAVATION PREVENTION (Category A)
Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse after-effects of the loss of normal cooling is avoided.
CALVERT CLIFFS UFSAR 1C.8-2 Rev. 47 B.
EMERGENCY CORE COOLING CRITERION 44 - EMERGENCY CORE COOLING SYSTEMS CAPABILITY (Category A)
At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features or components unless it can be demonstrated that (a) the capability of the shared feature or component to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a LOCA, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a LOCA and is not lost during the entire period this function is required following the accident.
CRITERION 45 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMS (Category A)
Design provisions shall be made to facilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internals and water injection nozzles.
CRITERION 46 - TESTING OF EMERGENCY CORE COOLING SYSTEMS COMPONENTS (Category A)
Design provisions shall be made so that active components of the emergency core cooling systems, such as pumps and valves, can be tested periodically for operability and required functional performance.
CRITERION 47 - TESTS OF EMERGENCY CORE COOLING SYSTEMS (Category A)
A capability shall be provided to test periodically the delivery capability of the emergency core cooling systems at a location as close to the core as is practical.
CRITERION 48 - TESTING OF OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS (Category A)
A capability shall be provided to test under conditions as close to design as practical the full operational sequence that would bring the emergency core cooling systems into action, including the transfer to alternate power sources.
C.
CONTAINMENT CRITERION 49 - CONTAINMENT DESIGN BASIS (Category A)
The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.
CRITERION 50 - NIL-DUCTILITY TEMPERATURE REQUIREMENT FOR CONTAINMENT MATERIAL (Category A)
Principal load-carrying components of ferritic materials exposed to the external environment shall be selected so that their temperatures under normal operating and testing conditions are not less than 30F above NDTT.
CALVERT CLIFFS UFSAR 1C.8-3 Rev. 47 CRITERION 51 - REACTOR COOLANT PRESSURE BOUNDARY OUTSIDE CONTAINMENT (Category A)
If part of the reactor coolant pressure boundary is outside the containment, appropriate features as necessary shall be provided to protect the health and safety of the public in case of an accidental rupture in that part. Determination of the appropriateness of features such as isolation valves and additional containment shall include consideration of the environmental and population conditions surrounding the site.
CRITERION 52 - CONTAINMENT HEAT REMOVAL SYSTEMS (Category A)
Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided.
CRITERION 53 - CONTAINMENT ISOLATION VALVES (Category A)
Penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus.
CRITERION 54 - CONTAINMENT LEAKAGE RATE TESTING (Category A)
Containment shall be designed so that an integrated leakage rate testing can be conducted at design pressure after completion and installation of all penetrations and the leakage rate measured over a sufficient period of time to verify its conformance with required performance.
CRITERION 55 - CONTAINMENT PERIODIC LEAKAGE RATE TESTING (Category A)
The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime.
CRITERION 56 - PROVISIONS FOR TESTING OF PENETRATIONS (Category A)
Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leaktightness to be demonstrated at design pressure at any time.
CRITERION 57 - PROVISIONS FOR TESTING OF ISOLATION VALVES (Category A)
Capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limits.
D.
CONTAINMENT PRESSURE REDUCING CRITERION 58 - INSPECTION OF CONTAINMENT PRESSURE-REDUCING SYSTEMS (Category A)
Design provisions shall be made to facilitate the periodic physical inspection of all important components of the containment pressure-reducing systems, such as, pumps, valves, spray nozzles, torus, and sumps.
CRITERION 59 - TESTING OF CONTAINMENT PRESSURE-REDUCING SYSTEMS COMPONENTS (Category A)
The containment pressure-reducing systems shall be designed so that active components, such as pumps and valves, can be tested periodically for operability and required functional performance.
CALVERT CLIFFS UFSAR 1C.8-4 Rev. 47 CRITERION 60 - TESTING OF CONTAINMENT SPRAY SYSTEMS (Category A)
A capability shall be provided to test periodically the delivery capability of the containment spray system at a position as close to the spray nozzles as is practical.
CRITERION 61 - TESTING OF OPERATIONAL SEQUENCE OF CONTAINMENT PRESSURE-REDUCING SYSTEM (Category A)
A capability shall be provided to test, under conditions as close to the design as practical, the full operational sequence that would bring the containment pressure-reducing systems into action, including the transfer to alternate power sources.
E.
AIR CLEANUP CRITERION 62 - INSPECTION OF AIR CLEANUP SYSTEMS (Category A)
Design provisions shall be made to facilitate physical inspection of all critical parts of containment air cleanup systems, such as, ducts, filters, fans, and dampers.
CRITERION 63 - TESTING OF AIR CLEANUP SYSTEMS COMPONENTS (Category A)
Design provisions shall be made so that active components of the air cleanup systems, such as fans and dampers, can be tested periodically for operability and required functional performance.
CRITERION 64 - TESTING OF AIR CLEANUP SYSTEMS (Category A)
A capability shall be provided for in-situ periodic testing and surveillance of the air cleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits.
CRITERION 65 - TESTING OF OPERATION SEQUENCE OF AIR CLEANUP SYSTEMS (Category A)
A capability shall be provided to test, under conditions as close to design as practical, the full operational sequence that would bring the air cleanup systems into action, including the transfer to alternate power sources and the design air flow delivery capability.
CALVERT CLIFFS UFSAR 1C.9-1 Rev. 47 1C.9 FUEL AND WASTE STORAGE SYSTEMS CRITERION 66 - PREVENTION OF FUEL STORAGE CRITICALITY (Category B)
Criticality in new and spent fuel storage shall be prevented by physical systems or processes.
Such means as geometrically safe configurations shall be emphasized over procedural controls.
CRITERION 67 - FUEL AND WASTE STORAGE DECAY HEAT (Category B)
Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs.
CRITERION 68 - FUEL AND WASTE STORAGE RADIATION SHIELDING (Category B)
Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10 CFR Part 20.
CRITERION 69 - PROTECTION AGAINST RADIOACTIVITY RELEASE FROM SPENT FUEL AND WASTE STORAGE (Category B)
Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs.
CALVERT CLIFFS UFSAR 1C.10-1 Rev. 47 1C.10 PLANT EFFLUENTS CRITERION 70 - CONTROL OF RELEASES OF RADIOACTIVITY TO THE ENVIRONMENT (Category B)
The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR Part 20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10 CFR Part 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents.
CALVERT CLIFFS UFSAR 1C.11-1 Rev. 50 1C.11 REFERENCES Memo from S. J. Chilk (NRC) to J. M. Taylor (NRC), dated September 18, 1992, SECY-92-223, Resolution of Deviations Identified During the Systematic Evaluation Program