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NUREG-0737 Technical Specifications (Generic Letter No. 83-37)
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 November 1, 1983 TO ALL PRESSURIZED WATER REACTOR LICENSEES Gentlemen: Subject: NUREG-0737 TECHNICAL SPECIFICATIONS (Generic Letter No. 83-37) NUREG-0737, "Clarification of TMI Action Plan Requirements," identifies those items for which Technical Specifications are required. Technical Specifications are required to provide assurance that facility operation is maintained within the limits determined acceptable following implementation at each facility. The scope and type of specification should include appropriate actions if limiting conditions for operation cannot be met. Relevant surveillance requirements for installed equipment should also be included. The guidance on Technical Specifications provided in Generic Letter 82-16 covered NUREG-0737 items which were scheduled for implementation by December 31, 1981. A number of NUREG-0737 items which require Technical Specifications were scheduled for implementation after December 31, 1981. Each of those items is presented in either Enclosure 1 or Enclosure 2. Included in the Enclosure 1 is guidance on the scope of Technical Specifications which the staff would find acceptable. Enclosure 2 presents a discussion on items which do not require a response at this time. Enclosure 3 contains samples in Standard Technical Specification format with blanks or parentheses appearing where the information is plant specific. It includes appropriate pages as background information for facilities that do not have Standard Technical Specifications. These samples are for your information only. We solicited comments on proposed Technical Specifications from pressurized water reactor owners group and the Atomic Industrial Forum. Appropriate comments have been incorporated, We request that you review your facility's Technical Specifications to determine if they are consistent with the guidance provided in Enclosure 1. For those items where you identify deviations or absence of a specification, we request that you submit an application for a license amendment. The Bases Section should be revised, as appropriate, to reflect the changes made in Technical Specifications. If some of the items are not yet implemented at your facility, you should submit an amendment request at the time they are implemented. It is recommended that licensees submit Technical Specifications for reactor coolant system vents within 30 days of receipt of this letter, and within 90 days for the remaining items discussed in Enclosure 1. However, it is recognized that some licensees may find this schedule to be stringent considering other activities planned at their facility as well as availability of the manpower. These licensees are encouraged to establish a realistic schedule for submittal of a response to this letter by negotiating with the individual Project Manager assigned to their facility. 8311010182 . - 2 - This request for information was approved by the Office of Management and Budget under clearance number 3150-0065 which expires September 30, 1985. Sincerely, Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation Enclosures: As Stated Licensee's Service Lists: See next page . ENCLOSURE 1 STAFF GUIDANCE ON TECHNICAL SPECIFICATIONS FOR NUREG 0737 ITEMS SCHEDULED AFTER DECEMBER 31, 1981 (1) Reactor Coolant System Vents (II.B.1) At least one reactor coolant system vent path (consisting of at least two valves in series which are powered from emergency buses) shall be operable and closed at all times (except for cold shutdown and refueling) at each of the following locations: a. Reactor Vessel Head b. Pressurizer steam space c. Reactor coolant system high point A typical Technical Specification for reactor coolant system vents is provided in Enclosure 3. For the plants using a power operated relief valve (PORV) as a reactor coolant system vent, the block valve is not required to be closed if the PORV is operable. (2) Post-accident Sampling (II.B.3) Licensees should ensure that their plant has the capability to obtain and analyze reactor coolant and containment atmosphere samples under accident conditions. An administrative program should be established, implemented and maintained to ensure this capability. The program should include: a) training of personnel b) procedures for sampling and analysis, and c) provisions for maintenance of sampling and analysis equipment. It is acceptable to the Staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifications and include a detailed description of the program in the plant operation manuals. A copy of the program should be easily available to the operating staff during accident and transient conditions. (3) Long Term Auxiliary Feedwater System Evaluation (II.E.1.1) The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. Technical Specifications depend on the results of the licensee's evaluation and staff review of each plant. The limiting conditions of operation (LCO) and surveillance requirements for the AFW system should be similar to safety-related systems. Typical generic Technical Specifications are provided in Enclosure 3. These specifications are for a plant which has three auxiliary feedwater pumps. Plant specific Technical Specifications could be established by using the generic Technical Specifications for the AFW system. . - 2 - (4) Noble Gas Effluent Monitors (II.F.1.1) Noble gas effluent monitors provide information, during and following an accident, which are considered helpful to the operator in accessing the plant condition. It is desired that these monitors be operable at all times during plant operation, but they are not required for safe shutdown of the plant. In case of failure of the monitor, appropriate actions should be taken to restore its operational capability in a reasonable period of time. Considering the importance of the availability of the equipment and possible delays involved in administrative controls, 7 days is considered to be the appropriate time period to restore the operability of the monitor. An alternate method for monitoring the effluent should be initiated as soon as practical, but no later than 72 hours after the identification of the failure of the monitor. If the monitor is not restored to operable conditions within 7 days after the failure a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the system to operable status. (5) Sampling and Analysis of Plant Effluents (II.F.1.2) Each operating nuclear power reactor should have the capability to collect and analyze or measure representative samples of radioactive iodides and particulates in plant gaseous effluents during and following an accident. An administrative program should be established, implemented and maintained to ensure this capability. The program should include: a) training of personnel b) procedures for sampling and analysis, and c) provisions for maintenance of sampling and analysis equipment It is acceptable to the staff, if the licensee elects to reference this program in the administrative controls section of the Technical Specifications and include a detailed description of the program in the plant operation manuals. A copy of the program should be readily available to the operating staff during accident and transient conditions. (6) Containment High-Range Radiation Monitor (II.F.1.3) A minimum of two in containment radiation-level monitors with a maximum range of 108 rad/hr (107 R/hr for photon only) should be operable at all times except for cold shutdown and refueling outages. In case of failure of the monitor, appropriate actions should be taken to restore its operational capability as soon as possible. If the monitor is not restored to operable condition within 7 days after the failure, a special report should be submitted to the NRC within 14 days following the event, outlining the cause of inoperability, actions taken and the planned schedule for restoring the equipment to operable status. . - 3 - Typical surveillance requirements are shown in Enclosure 3. The setpoint for the high radiation level alarm should be determined such that spurious alarms will be precluded. Note that the acceptable calibration techniques for these monitors are discussed in NUREG-0737. (7) Containment Pressure Monitor (II.F.1.4) Containment pressure should be continuously indicated in the control room of each operating reactor during Power Operation, Startup and Hot Standby modes of operation. Two channels should be operable at all times when the reactor is operating in any of the above mentioned modes. Technical Specifications for these monitors should be included with other accident monitoring instrumentation in the present Technical Specifications. Limiting conditions for operation (including the required Actions) for the containment pressure monitor should be similar to other accident monitoring instrumentation included in the present Technical Specifications. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3. (8) Containment Water Level Monitor (II.F.1.5) A continuous indication of containment water level should be provided in the control room of each reactor during Power Operation, Startup and Hot Standby modes of operation. At least one channel for narrow range and two channels for wide range instruments should be operable at all times when the reactor is operating in any of the above modes. Narrow range instruments should covert the range from the bottom to the top of the containment sump. Wide range instruments should cover the range.from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity. Technical Specifications for containment water level monitors should be included with other accident monitoring instrumentation in the present Technical Specifications. LCOs (including the required Actions) for wide range monitors should be similar to other accident monitoring instrumentation included in the present Technical Specifications. LCOs for narrow range monitor should include the requirement that the inoperable channel will be restored to operable status within 30 days or the plant will be brought to Hot Shutdown condition as required for other accident monitoring instrumentation. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3. (9) Containment Hydrogen Monitor (II.F.1.6) Two independent containment hydrogen monitors should be operable at all times when the reactor is operating in Power Operation or Startup modes. LCO for these monitors should include the requirement that with one hydrogen monitor inoperable, the monitor should be restored to operable status within 30 days or the plant should be brought to . - 4 - at least a hot standby condition within the next 6 hours. If both monitors are inoperable, at least one monitor should be restored to operable status within 72 hours or the plant should be brought to at least hot standby condition within the next 6 hours. Typical surveillance requirements are provided in Enclosure 3. (10) Instrumentation for Detection of Inadequate Core Cooling (II.F.2) Subcooling margin monitors, core exit thermocouples, and a reactor coolant inventory tracking system (e.g., differential pressure measurement system designed by Westinghouse, Heated Junction Thermocouple System designed by Combustion Engineering, etc.) may be used to provide indication of the approach to, existence of, and recovery from inadequate core cooling (ICC). These instrumentation should be operable during Power Operation, Startup, and Hot Shutdown modes of operation for each reactor. Subcooling margin monitors should have already been included in the present Technical Specifications. Technical Specifications for core exit thermocouples and the reactor coolant inventory tracking system should be included with other accident monitoring instrumentation in the present Technical Specifications. Four core-exit thermocouples in each core quadrant and two channels in the reactor coolant tracking system are required to be operable when the reactor is operating in any of the above mentioned modes. Minimum of two core-exit thermocouples in each quadrant and one channel in the reactor coolant tracking system should be operable at all times when the reactor is operating in any of the above mentioned modes. Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are provided in Enclosure 3. (11) Control Room Habitability Requirements (III.D.3.4) Licensees should assure that control room operators will be adequately protected against the effects of the accidental release of toxic and/or radioactive gases and that the nuclear power plant can be safely operated or shutdown under design basis accident conditions. If the results of the analyses of postulated accidental release of toxic gases (at or near the plant) indicate any need for installing the toxic gas detection system, it should be included in the Technical Specifications. Typical acceptable LCO and surveillance requirements for such a detection system (e.g. chlorine detection system) are provided in Enclosure 3. All detection systems should be included in the Technical Specifications. In addition to the above requirements, other aspects of the control room habitability requirements should be included in the Technical Specifications for the control room emergency air cleanup system. Two independent control room emergency air cleanup systems should be operable continuously during all modes of plant operation and capable of meeting design requirements. Sample Technical Specifications are provided in Enclosure 3. . ENCLOSURE 2 DISCUSSION OF NUREG-0737 ITEMS SCHEDULED AFTER DECEMBER 31, 1981, WHICH DO NOT REQUIRE THE RESPONSE (1) Minimum Shift Crew (I.A.1.3.2) The requirements of this Action Plan item are superceded by a recent rule concerning staffing of licensed operators at Nuclear Power Plants. The effective date of this rule is January 1, 1984. The rule was promulgated on July 11, 1983. No response is required at this time. (2) Thermal Mechanical Report (II.K.2.13) Licensees of Westinghouse and Combustion Engineering operating reactors were required to submit by January 1, 1982 an analysis of the thermal mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. The staff has received the above mentioned reports for all PWR vendor designs. Changes to Technical Specifications will be determined after the staff has completed the review of these reports. No response is required at this time. (3) Auto PORV Isolation (II.K.3.1) Implementation of this Action Plan item is to be required only if the studies specified in TMI Action Plan Item II.K.3.2 confirmed the need, for automatic isolation system for the power operated relief valves (PORV). The staff has completed the review of the information provided by the licensees as part of the implementation of Item II.K.3.2. The staff has concluded that Automatic PORV Isolation System will not be required on a generic basis. Each licensee will be informed separately about our conclusion. No changes in Technical Specifications are required where II.K.3.1 implementation is not required. (4) Auto Trip of Reactor Coolant Pumps (II.K.3.5) The staff has informed all licensees by a separate letter to evaluate the need for tripping reactor coolant pumps in each plant. The need for changing Technical Specifications will be determined by reviewing each plant on a case by case basis. No response is required at this time. (5) Emergency Core-Cooling Systems (ECCS) Outage (II.K.3.17) The staff has completed the review of ECCS data provided by the licensees, and determined that no changes in the Technical Specifications are required at this time. No response is required. . - 2 - (6) Compliance with 10 CFR Part 50.46 (II.K.3.31) This Action Plan item requires the licensees to submit plant specific calculations to show compliance with 10 CFR,Part 50.46, if changes have been made in the small break loss of coolant accident (LOCA) evaluation model to show compliance with 10 CFR Part 50, Appendix K (Item II.K.3.30). The staff is currently reviewing the information provided by the licensees in response to Item II.K.3.30. Changes to Technical Specifications, if found necessary, will be determined after the staff has approved the revised evaluation model and plant specific calculations submitted by the licensees to show compliance with 10 CFR Part 50.46. No response is required at this time. (7) The Upgrade of Emergency Support Facility (III.A.1.2) Meteorological Data (III.A.2.2) These two items are covered under Supplement No. 1 to NUREG-0737. No response is required at this time.
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