United States Nuclear Regulatory Commission - Protecting People and the Environment

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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