Bulletin 96-02: Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or over Safety-Related Equipment
OMB No. 3150-0012
NRCB 96-02
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON D.C. 20555-0001
April 11, 1996
NRC BULLETIN 96-02: MOVEMENT OF HEAVY LOADS OVER SPENT FUEL, OVER FUEL
IN THE REACTOR CORE, OR OVER SAFETY-RELATED EQUIPMENT
Addressees
All holders of boiling-water reactor (BWR) and pressurized-water reactor (PWR)
operating licenses for nuclear power reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this bulletin to
accomplish the following:
(1) Alert addressees to the importance of complying with existing regulatory
guidelines associated with the control and handling of heavy loads at
nuclear power plants while the plant is operating (in all modes other
than cold shutdown, refueling, and defueled) and remind addressees of
their responsibilities for ensuring that heavy load activities carried
out under their license are performed safely and within the requirements
specified under Title 10 of the Code of Federal Regulations.
(2) Request that addressees review their plans and capabilities for handling
heavy loads (e.g., spent fuel dry storage casks, reactor cavity
biological shield blocks) in accordance with existing regulatory
guidelines [specifically NUREG-0612 (Phase I) and Generic Letter (GL)
85-11] and within their licensing basis as previously analyzed in the
final safety analysis report (FSAR).
(3) Require addressees to report to the NRC whether and to what extent they
have complied with the requested actions contained in this bulletin.
Although this bulletin is particularly concerned with heavy load movements
while the plant is operating (i.e., in all modes other than cold shutdown,
refueling, and defueled), the staff is considering further generic actions on
the issue of handling all heavy loads both while the plant is operating and
during shutdown.
Background
There are a number of heavy loads being handled in various areas of nuclear
power plants, especially over safety-related equipment, when the plant is
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April 11, 1996
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operating. Some licensees have moved or are planning to move heavy loads such
as spent fuel shipping casks, transfer casks, and reactor cavity biological
shield blocks during plant operations. If these loads experience uncontrolled
movement or are dropped on safety-related equipment, the equipment may be
unable to perform its function.
Guidelines regarding the movement of these and other heavy loads are provided
in a number of documents that in combination make up the framework for the
existing regulatory position on heavy load handling and control. The most
important guidelines are contained in the following three documents:
(1) NUREG-0612, "Control of Heavy Loads at Power Plants," Resolution of
Generic Technical Activity A-36, issued July 1980
(2) Unnumbered generic letter dated December 22, 1980, "Control of Heavy
Loads"
(3) GL 85-11, "Completion of Phase II of Control of Heavy Loads at Nuclear
Power Plants, NUREG-0612," dated June 28, 1985
NUREG-0612 provides guidelines to (1) ensure the safe handling of heavy loads,
(2) reduce the potential for uncontrolled movement of heavy loads or load
drops, and (3) limit the consequences of dropping a heavy load. The
guidelines were supported by historical data and fault tree analyses. Some
portions of the guidelines were generic to all plants, while others were
specific to plant type and location (e.g., the PWR containment building).
The guidelines consider the handling of heavy loads while the reactor is at
power and provide a methodology to do so safely.
The unnumbered generic letter of December 22, 1980, requested that licensees
implement the heavy load control guidelines in NUREG-0612 and identify any
problems that they encountered. The generic letter also requested immediate
implementation of some interim actions (safe load paths, crane design and
inspection, operator training, and procedures), a 6-month followup response on
the status of the implementation of Section 5.1.1 of NUREG-0612 (Phase I), and
a 9-month followup response on the status of the implementation of the
remaining applicable portions of Section 5.1 of NUREG-0612 (Phase II: single-
failure-proof cranes, stops/interlocks, or load-drop analyses).
All affected licensees implemented the interim actions and Phase I of the
generic letter and submitted a response for Phase II. The staff reviewed the
implemented actions and a sample of the Phase II submittals and determined
that the actions taken by the licensees had significantly decreased the
potential for a heavy load drop. The staff performed a limited review of the
remaining Phase II submittals and did not identify any plant-specific safety
concerns associated with the control of heavy loads.
Subsequently, the staff issued GL 85-11, which informed licensees that
implementation of Phase II was not necessary but encouraged licensees to
implement any safety-significant portions they believed were appropriate.
GL 85-11 relieved licensees from performing the actions requested under . NRCB 96-02
April 11, 1996
Page 3 of 9
Phase II of the previous generic letter. However, GL 85-11 did not grant
blanket NRC approval for all load paths identified in the Phase II submittals,
nor did it authorize licensees to exceed their design basis for heavy load
transfer.
Although the generic letter stated that the NRC staff review of the Phase II
submittals did not indicate the need to require further generic action at that
time, it did not preclude the possible future need for the staff to review
additional heavy load handling concerns and to require, as appropriate,
further actions by licensees.
Description of Circumstances
In 1996, GPU Nuclear (GPUN) Corporation, the licensee for the Oyster Creek
Nuclear Power Plant, is scheduled to begin moving heavy loads involving dry
storage casks within the Oyster Creek facility. GPUN is planning to load
spent fuel from the Oyster Creek plant into dry storage casks that will be
placed in an independent spent fuel storage installation. The loaded casks,
each weighing 100 tons, must be moved over safety-related equipment during
this process. The licensee's plans involve loading and moving the casks
during power operation because performing these activities during a refueling
outage would significantly increase the outage time.
The licensee prepared an initial evaluation pursuant to 10 CFR 50.59 regarding
the planned activities for handling the dry storage casks, including the use
of the non-single-failure-proof reactor building crane to transfer spent fuel
to the dry cask storage facility during plant operation. To reduce the
probability of a load drop, GPUN modified its crane; proposed to use a crush
pad along part of the load path; and proposed to institute an "Error Free
Plan," which includes upgrading its training, management and oversight, and
cask-handling procedures specific to this evolution and development. However,
during two portions of the proposed cask movement inside the reactor building,
a cask drop could damage both isolation condensers and the torus, possibly
creating an unisolable loss-of-coolant accident outside containment. This
drop could occur in those areas near the spent fuel pool or near the equipment
hatch where the crush pad proposed by the licensee to protect against drops on
the 119-foot level is not installed. A cask dropped from either of these
locations on the 119-foot level could fall through all of the lower floors and
into the torus, damaging all equipment in its path. The licensee stated that
core cooling could be maintained by steaming to the condenser using the normal
feedwater system and providing makeup from the condensate storage tank and
fire water systems by way of the core spray system. While GPUN had reduced
the probability of dropping the cask, the staff was concerned that because the
casks are heavier than previously considered in the FSAR, a cask drop could
result in higher consequences than those previously analyzed.
As a result of concerns raised by the staff and GPUN's efforts to improve the
efficiency of handling the spent fuel storage casks and to minimize the
probability of a cask drop, GPUN updated its 10 CFR 50.59 evaluation to
include a number of improvements applicable to the criteria of NUREG-0612,
Phase I. GPUN adjusted the load path, eliminated the crush pad, and upgraded . NRCB 96-02
April 11, 1996
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the reactor building crane (but not to the level of a single-failure-proof
crane as defined in NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power
Plants") by installing a fixed link support system. The fixed links provide
redundant rigging for the cask while it is transported on the 119-foot level,
especially in the area over the isolation condensers. It uses horizontal
support beams attached to the cask-lifting yoke and vertical tie-rods
connected to the crane trolley to support the cask in the event of a failure
of a crane hoist component.
GPUN evaluated postulated load drops while the cask is in the reactor building
equipment hatchway (from the 119-foot elevation to the 23-foot elevation) and
at the laydown area on the 119-foot elevation where the fixed links are not
engaged and concluded that if a cask is dropped in either of these areas, the
cask could damage the torus, causing it to drain. Consequently, the pressure
suppression function of the primary containment could be disabled. The
reactor is expected to scram successfully, reducing power so that only post-
scram decay heat would have to be removed. The primary coolant system piping
would not be affected by the drop; therefore, the need for vessel inventory
makeup would not be required immediately. Some safety-related equipment would
be damaged, for example, one set of containment spray pumps and one contain-
ment spray heat exchanger. However, containment spray would be unavailable in
any event since GPUN has assumed no water would be present in the torus. The
isolation condenser system would be available to provide long-term heat
removal from the reactor vessel. Makeup to the isolation condenser shell
could be accomplished remotely by using condensate transfer. If needed, a
reactor building entry to establish shell-side makeup could be performed after
approximately 1 hour. The load-drop analysis concluded that the reactor could
be safely shut down following a drop of the cask and that the offsite
consequences of a load drop are bounded by high-energy line break evaluations.
The licensee determined that releases resulting from damage to the 52 fuel
assemblies in the cask would not exceed 25 percent of the limits set out in
10 CFR Part 100 because the fuel assemblies will be more than 10 years old.
GPUN's 10 CFR 50.59 evaluation concludes that no unreviewed safety questions
are involved, that movement of the casks can be accomplished in a safe manner
because of GPUN's reduction of the probability of dropping the load, and that
all license requirements would be satisfied. GPUN based this conclusion on
its completion of the Phase I guidelines (Section 5.1.1 of NUREG-0612) for the
control of heavy loads at nuclear power plants. The staff states in GL 85-11
that "our review has indicated that satisfaction of the Phase I guidelines
assures that the potential for a load drop is extremely small." This
conclusion is further based on GPUN's evaluation that (1) the fixed links
provide redundant load support for the transfer cask, equivalent to a single-
failure-proof crane for nearly the entire travel path; (2) safe shutdown can
be achieved where the fixed link support system does not provide protection;
and (3) although a postulated load drop could damage safety-related equipment,
the probability of a drop is extremely low. The licensee also noted that the
only load drop previously evaluated in the plant safety analysis report (SAR)
is the drop of a 100-ton fuel shipping cask in the vicinity of the fuel pool.
. NRCB 96-02
April 11, 1996
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Discussion
In 10 CFR 50.59(a)(1), it is stated that "the holder of a license authorizing
operation of a production or utilization facility may (i) make changes in the
facility as described in the safety analysis report, (ii) make changes in the
procedures as described in the safety analysis report, and (iii) conduct tests
or experiments not described in the safety analysis report, without prior
Commission approval, unless the proposed change, test or experiment involves a
change in the technical specifications incorporated in the license or an
unreviewed safety question." Section 50.59(a)(2) states that "a proposed
change, test, or experiment shall be deemed to involve an unreviewed safety
question (i) if the probability of occurrence or the consequences of an
accident or malfunction of equipment important to safety previously evaluated
in the safety analysis report may be increased; or (ii) if a possibility for
an accident or malfunction of a different type than any evaluated previously
in the safety analysis report may be created; or (iii) if the margin of safety
as defined in the basis for any technical specification is reduced."
The NRC staff audited both the initial and updated 10 CFR 50.59 evaluations
performed by the licensee and determined that the proposed cask movement
activities represent an unreviewed safety question that should be submitted to
the NRC for review and approval pursuant to the requirements of 10 CFR 50.59
and 50.90. The staff based its determination on the fact that, as noted by
the licensee, the activity involves movement of loads heavier than those
previously analyzed in the FSAR (except over the cask drop protection system
in the fuel pool, where a 100-ton cask drop had been previously analyzed).
This determination is also based on the fact that the load drop had not been
previously evaluated along the remainder of the load path, and on the
possibility that a load drop in the reactor building while the reactor is at
power could result in consequences that are greater than those previously
postulated in the FSAR. Therefore, although the licensee had reduced the
probability of dropping the cask, the staff was concerned that a load drop
could result in an increase in the potential consequences. Accordingly, as
defined in 10 CFR 50.59(c), if an activity is found to involve an unreviewed
safety question, an application for a license amendment must be filed with the
Commission pursuant to 10 CFR 50.90.
Based on the NRC staff's audit of GPUN's 10 CFR 50.59 evaluation, the staff is
concerned that other licensees may believe that their heavy load operations
are in compliance with the regulations because they have completed Phase I of
the generic letter of December 22, 1980, and the closeout of Phase II by
GL 85-11. GL 85-11 did not relieve licensees of their responsibility under
10 CFR 50.59 to evaluate new activities with respect to the SAR and the
Technical Specifications to determine whether the activity involves an
unreviewed safety question or a change in the Technical Specifications. In
addition, GL 85-11 concluded that the risks associated with damage to safety-
related systems are relatively small because (1) nearly all load paths avoid
this equipment, (2) most equipment is protected by an intervening floor,
(3) there is redundancy of components, and (4) crane failure probability is
generally independent of safety-related systems. As is demonstrated by Oyster
Creek's proposed activities, this conclusion may not always be valid.
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April 11, 1996
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Therefore, the staff has concluded that although some licensees have under-
taken efforts to further reduce the probability of an accident involving heavy
loads beyond that previously accepted for NUREG-0612, Phase I, if the loads
are heavier and the load paths and potential consequences of a load drop are
different than those previously considered in the FSAR, the probability of an
occurrence or the consequences of an accident may be increased.
Requested Actions
To ensure that the handling of heavy loads is performed safely and within the
conditions and requirements specified under Title 10 of the Code of Federal
Regulations, all addressees are requested to take the following actions:
. Review plans and capabilities for handling heavy loads while the reactor
is at power (in all modes other than cold shutdown, refueling, and
defueled) in accordance with existing regulatory guidelines. Determine
whether the activities are within the licensing basis and, if necessary,
submit a license amendment request. Determine whether changes to
Technical Specifications will be required in order to allow the handling
of heavy loads (e.g., the dry storage canister shield plug and associated
lifting devices) over fuel assemblies in the spent fuel pool.
Required Response
Pursuant to Section 182a, the Atomic Energy Act of 1954, as amended, and
10 CFR 50.54(f), all addressees must submit the following written information:
(1) For licensees planning to implement activities involving the handling of
heavy loads over spent fuel, fuel in the reactor core, or safety-related
equipment within the next 2 years from the date of this bulletin, provide
the following:
. A report, within 30 days of the date of this bulletin, that
addresses the licensee's review of its plans and capabilities to
handle heavy loads while the reactor is at power (in all modes other
than cold shutdown, refueling, and defueled) in accordance with
existing regulatory guidelines. The report should also indicate
whether the activities are within the licensing basis and should
include, if necessary, a schedule for submission of a license
amendment request. Additionally, the report should indicate whether
changes to Technical Specifications will be required.
(2) For licensees planning to perform activities involving the handling of
heavy loads over spent fuel, fuel in the reactor core, or safety-related
equipment while the reactor is at power (in all modes other than cold
shutdown, refueling, and defueled) and that involve a potential load drop
accident that has not previously been evaluated in the FSAR, submit a
license amendment request in advance (6-9 months) of the planned movement
of the loads so as to afford the staff sufficient time to perform an
appropriate review.
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April 11, 1996
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(3) For licensees planning to move dry storage casks over spent fuel, fuel in
the reactor core, or safety-related equipment while the reactor is at
power (in all modes other than cold shutdown, refueling, and defueled)
include in item 2 above, a statement of the capability of performing the
actions necessary for safe shutdown in the presence of radiological
source term that may result from a breach of the dry storage cask, damage
to the fuel, and damage to safety-related equipment as a result of a load
drop inside the facility.
(4) For licensees planning to perform activities involving the handling of
heavy loads over spent fuel, fuel in the reactor core, or safety-related
equipment while the reactor is at power (in all modes other than cold
shutdown, refueling, and defueled), determine whether changes to
Technical Specifications will be required in order to allow the handling
of heavy loads (e.g., the dry storage canister shield plug) over fuel
assemblies in the spent fuel pool and submit the appropriate information
in advance (6-9 months) of the planned movement of the loads for NRC
review and approval.
Address the required written report(s) to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, under
oath or affirmation under the provisions of Section 182a, Atomic Energy Act of
1954, as amended, and 10 CFR 50.54(f). In addition, submit a copy of the
report to the appropriate regional administrator.
Related Generic Communications
. NUREG-0612, "Control of Heavy Loads at Power Plants," Resolution of
Generic Technical Activity A-36, issued in July 1980
. Unnumbered generic letter dated December 22, 1980, "Control of Heavy
Loads"
. GL 85-11: "Completion of Phase II of Control of Heavy Loads at Nuclear
Power Plants, NUREG-0612," June 28, 1985
Backfit Discussion
This bulletin is an information request made pursuant to 10 CFR 50.54(f).
The objective of the actions requested in this bulletin is to verify that
licensees are complying with the current licensing basis for their facility
with respect to the proper handling and control of heavy loads at nuclear
power plants when the plant is operating (in all modes other than cold shut-
down, refueling, and defueled). The issuance of the bulletin is justified on
the basis of the need to ensure compliance with the current licensing basis
with respect to the weight of the heavy loads being moved over spent fuel,
over fuel in the reactor core, or over safety-related equipment, and the
potentially severe consequences that can result if a load is dropped.
. NRCB 96-02
April 11, 1996
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Paperwork Reduction Act Statement
This bulletin contains information collections that are subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information
collections were approved by the Office of Management and Budget (OMB),
approval number 3150-0012, which expires June 30, 1997.
The public reporting burden for this collection of information is estimated to
average 600 hours per response, including the time for reviewing instructions,
searching existing data sources, gathering and maintaining the data needed,
and completing and reviewing the collection of information. The NRC is
seeking public comment on the potential impact of the collection of infor-
mation contained in the generic bulletin and on the following issues:
(1) Is the proposed collection of information necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
(2) Is the estimate of burden accurate?
(3) Is there a way to enhance the quality, utility, and clarity of the
information to be collected?
(4) How can the burden of the collection of information be minimized,
including the use of automated collection techniques?
Send comments on any aspect of this collection of information, including
suggestions for reducing the burden, to the Information and Records Manage-
ment Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, or by Internet electronic mail at bjs1@nrc.gov; and to the Desk
Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0012),
Office of Management and Budget, Washington, DC 20503.
The NRC may not conduct or sponsor, and a person is not required to respond
to, a collection of information unless it displays a currently valid OMB
control number.
. NRCB 96-02
April 11, 1996
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If you have any questions about this matter, please contact the technical
contact listed below or the appropriate Office of Nuclear Reactor Regulation
(NRR) project manager.
signed by
Dennis M. Crutchfield, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contact: Brian E. Thomas, NRR
(301) 415-1210
Internet: bet@nrc.gov
Lead Project Manager: Kevin A. Connaughton, NRR
(301) 415-3018
Internet: kac@nrc.gov
Attachment: List of Recently Issued NRC Bulletins
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