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
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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.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
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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.
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(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.
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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.
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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.
(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.
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(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.
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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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