United States Nuclear Regulatory Commission - Protecting People and the Environment

NUREG-0737 Technical Specifications (Generic Letter No. 83-36)



                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                          WASHINGTON, D. C. 20555  

                              November 1, 1983 


TO ALL BOILING WATER REACTOR LICENSEES 

Gentlemen: 

Subject: NUREG-0737 TECHNICAL SPECIFICATIONS (Generic Letter No.  83-36) 

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 83-02 
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 boiling 
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 respond within 90 days of receipt of this 
letter.  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. 


8311010180 
.

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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.l) 

     The staff has determined that no changes in Technical Specifications 
     are required by this Action Plan item for Boiling Water Reactors (BWRs) 
     which do not have isolation condenser.  The staff has also concluded 
     that no changes in Technical Specifications are required for those 
     plants which have isolation condenser, and either a turbine driven high 
     pressure injection system or a feedwater coolant injection system with 
     an auxiliary power source such as a gas turbine. 

     Those BWRs with isolation condenser, and no high pressure injection 
     other than normal feedwater or the control rod drive system must have 
     isolation condenser vents which satisfy the requirements of Item II.B.1
     of NUREG-0737.  These plants should have at least one reactor coolant 
     system vent path (consisting of at least two valves which are powered 
     from emergency buses) operable and closed at all times (except for cold
     shutdown and refueling) at isolation condenser high points.  A typical 
     Technical Specification for reactor coolant system vents is provided in
     Enclosure 3. 

(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 readily 
     available to the operating staff during accident and transient 
     conditions. 

(3)  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 
.

                                    - 2 -

     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 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 system to operable 
     status. 

(4)  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 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. 

(5)  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. 

     Typical surveillance requirements are presented 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. 

(6)  Containment Pressure Monitor (II.F.1.4) 

     Containment pressure should be continuously indicated in the control 
     room of each operating reactor during Power Operation and Startup 
     Modes. 

     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 (LCO) 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. 
.

                                    - 3 -

(7)  Containment Water Level Monitor (II.F.1.5) 

     A continuous indication of suppression pool water level should be 
     provided in the control room of each reactor during Power Operation and
     Startup Modes.  Two channels should be operable at all times when the 
     reactor is operating in any of the above mentioned modes.  Technical 
     Specifications for suppression pool water level monitors should be 
     included with other accident monitoring instrumentation in the present 
     Technical Specifications.  Limiting conditions for operation (LCO) for 
     these monitors 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. 

     The BWRs with dry containment should have at least two channels for 
     wide range instruments and one channel of narrow range instrument 
     operable at all times during above mentioned modes.  LCOs for wide 
     range monitors should be similar to that discussed above.  LCOs for 
     narrow range monitor should include the requirement that the inoperable 
     channel will be restored to operable status within 30 days or the 
     reactor will be brought to hot shutdown condition as required by other 
     accident monitoring instrumentation. 

(8)  Containment Hydrogen Monitor (II.F.1.6) 

     Two independent containment hydrogen monitors should be operable 
     (should be capable of performing the required function) at all times 
     when the reactor is operating in Power Operation and Startup Modes.  
     Technical Specifications for hydrogen monitors should be included with 
     other accident monitoring instrumentation in the present Technical 
     Specification.  Typical acceptable LCO and surveillance requirements 
     are included in Enclosure 3. 

(9)  Control Room Habitability Requirements (II.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 shut down under design basis accident conditions.  If the 
     results of the analyses of postulated accidental release of toxic gases
     (at or near the plant) indicated a 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. 
.

                                    - 4 -

In addition to the above requirements, other aspects of the control room 
habitability requirements should be included in the Technical Specifications
for control room emergency air filtration system.  Two independent control 
room emergency air filtration system 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. 

(2)  Instrumentation for Detection of Inadequate Core Cool (II.F.2) 

     The BWR Owners' group has proposed some modifications in existing 
     instrumentation to satisfy the requirements of this Action Plan Item. 
     The staff is currently evaluating various options to modify existing 
     instrumentation in boiling water reactors.  Changes in Technical 
     Specifications will be determined after the evaluation is completed. 
     No response is required at this time. 

(3)  Isolation of Isolation Condensers on High Radiation (II.K.3.14) 

     This Action Plan item is applicable to only seven boiling water 
     reactors.  Licensees of all seven plants have submitted the responses 
     on this item and the staff has determined that no changes are required 
     in the present design.  No changes in Technical Specifications are 
     needed. 

(4)  Reduction of Challenges and Failures of Relief Valves (II.K.3.16) 

     The staff has reviewed the information submitted by the BWR Owners' 
     group in response to Item II.K.3.16, and identified acceptable 
     modifications which will reduce safety/relief valve challenges and 
     failures.  One of these modifications involves the design of Low-Low 
     Set (LLS) Relief Logic System.  This system may require changes in the 
     Technical Specifications.  However, for the BWRs with Mark I 
     containment, the Technical Specifications changes will be reviewed as 
     part of the Mark I containment modifications review.  For BWRs, with 
     mark II containment the need for changing the Technical Specifications 
     will be determined on a case by case basis.  Some licensees may decide 
     to change the water level setpoint for the closures of main steam 
     isolation valves (MSIVs) as a part of the implementation of this item. 
     This will require the changes in the Technical Specifications.  These 
     changes will be reviewed on a case by case basis.  No other changes are
     required. 

(5)  Emergency Core-Cool ing Systems (ECCS) Outage (II.K.3.17) 

     The staff has completed the review of ECCS outage data provided by the 
     licensees, and determined that no changes in Technical Specifications 
     are required at this time.  No response is required. 
.

                                    - 2 -

(6)  Automatic Depressurization System Logic Modification (II.K.3.18) 

     Licensees are required to perform a feasibility and risk assessment 
     study to determine the optimum approach for modifying automatic 
     depressurization system (ADS) actuation logic to eliminate the need for
     manual actuation to assure adequate core cooling.  The BWR Owners' 
     group has submitted an evaluation to the staff.  The staff has 
     identified the acceptable options for modifications of ADS logic.  Each 
     licensee was requested to select appropriate modifications approved by 
     the staff.  Technical Specifications changes resulting from the 
     modifications will be reviewed on a case by case basis. 

(7)  Adequacy of Space Cooling for High-Pressure Coolant Injection and 
     Reactor Core Isolation Cooling Systems (II.K.3.24) 

     The staff has reviewed the responses from all licensees for this Action
     Plan item and concluded that space cooling system for high pressure 
     coolant injection (HPCI) and Reactor Core Cooling Isolation (RCIC) 
     systems is powered from diesel generators in case of loss of offsite 
     power.  As the space.cooling system is considered to be a supporting 
     system for HPCI and RCIC systems, the operability requirements of this 
     system should be already included in the Technical Specifications for 
     HPCI and RCIC.  No further changes are required. 

(8)  Qualification of Accumulators on Automatic Depressurization System 
     Valves (II.K.3.28) 

     The staff is currently reviewing information provided by the licensees. 
     Changes in the Technical Specifications will be determined after our 
     review is completed.  No response is required at this time. 

(9)  Compliance with 10 CFR Part 50.46 (II.K.3.31) 

     This Action Plan item requires 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 has reviewed the generic response submitted by 
     General Electric in response to Item II.K.3.30.  Pending formal 
     documentation of the staff review, it is anticipated that no changes in
     the Technical Specifications will be required by this Action Plan item.

(10) Evaluation of Anticipated Transients with Single Failure (II.K.3.44) 

     The staff has completed the review of the evaluation submitted by the 
     BWR Owners' group and determined that no changes are required in the 
     design.  No changes in Technical Specifications are required. 
.

                                    - 3 -

11)  The Upgrade of Emergency Support Facility (III.A.1.2) 

     Meteorological Data (III.A.2.2) 

     These two items are covered under  supplement 1 to NUREG-0737. 
     "Requirements for Emergency Response Capability" (Generic Letter 
     82-33). 

     No response is required at this time. 
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