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
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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
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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.
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(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.
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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
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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.
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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.
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(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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