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United States Nuclear Regulatory Commission - Protecting People and the Environment

                                                            OMB No. 3150-0011
                                                            NRCB 96-04


                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                                 July 5, 1996


NRC BULLETIN 96-04:  CHEMICAL, GALVANIC, OR OTHER REACTIONS IN SPENT FUEL
                     STORAGE AND TRANSPORTATION CASKS


Addressees 

This bulletin is being sent to:

      All holders of operating licenses or construction permits for
      nuclear power reactors.

      All holders of, and applicants for, certificates of compliance for
      transportation casks for commercial spent fuel.

      All holders of, and applicants for, certificates of compliance for
      storage casks for commercial spent fuel.

      All vendors of storage and transportation casks for commercial
      spent fuel.

      All registered users of transportation casks for commercial spent
      fuel.

It is expected that all recipients will review the information for
applicability to their facilities and consider actions as appropriate to avoid 
problems similar to those discussed here.  However, action is only requested
from those addressees who are licensees with independent spent fuel storage
installations, vendors of spent fuel storage or transportation casks, and
holders of certificates of compliance for spent fuel storage or transportation
casks (�action addressees�).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this bulletin to: 
(1) notify addressees about the potential for chemical, galvanic, or other
reactions among the materials of a spent fuel storage or transportation cask,
its contents, and the environments the cask may encounter during use, that may
produce adverse conditions in cask loading/unloading/handling operations or
degrade the integrity and performance of the cask; (2) request that all action
addressees implement the actions described herein; and (3) require that all
action addressees provide, to NRC, written responses to this bulletin,
relating to implementation of the requested actions.  This bulletin also 


9607020241.                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 2 of 9


requires that certain information on the subject matter of this bulletin be
submitted to NRC.

Background

On May 31, 1996, NRC Information Notice 96-34 was issued as notification of a
hydrogen gas ignition event that occurred during the welding of the shield lid
on a spent fuel storage cask at the Point Beach Nuclear Plant.

Description of Circumstances

On May 28, 1996, a hydrogen gas ignition occurred during the welding of the
shield lid on a ventilated storage cask (VSC-24) multi-assembly sealed basket
(MSB).  The gas ignition displaced the shield lid (weighing about 2898
kilograms [6390 pounds]), leaving it in place but tipped at a slight angle,
with one edge about 7.6 centimeters [3 inches] higher than normal.

The loaded VSC-24 multi-assembly transfer cask (MTC), a shielded lifting
device used to transfer the MSB loaded with spent fuel to the ventilated
concrete cask, had been placed in the cask decontamination work area in the
auxiliary building.  Approximately 114 liters [30 gallons] of borated spent
fuel pool water had been drained from the MSB to facilitate welding of the
shield lid, creating an air space below the lid.  The hydrogen gas ignition
occurred during the initiation of the shield lid welding, approximately  
11 hours after the loaded MTC had been removed from the spent fuel storage
pool.

After the event, gas and water samples collected from the MSB internals showed
detectable levels of hydrogen both in the air space beneath the shield lid and
dissolved in the MSB water.  The licensee then continuously purged the air
space beneath the lid with nitrogen to prevent the accumulation of combustible
gases and returned the shield lid to its original position.

The MSB was then fully flooded to eliminate the air space under the shield lid
and returned to the spent fuel storage pool.  On removal of the shield lid,
the licensee observed a white, foam-like substance under the shield lid.  Most
of this substance floated to the top of the pool; however, some remained below
the surface of the water.  Samples of this substance were taken to Argonne
National Laboratory for analysis.

The licensee unloaded the spent fuel assemblies and placed them in the spent
fuel pool storage racks.  A visual examination of the MSB, the MTC, and the
spent fuel assemblies showed no evidence of damage as a result of the
combustible gas ignition.

The unloaded MTC/MSB was subsequently moved back to the decontamination work
area for further inspection as part of the licensee's investigation of the
combustible gas ignition.  At the decontamination work area, the MSB remained
filled with borated spent fuel pool water for several days.  Hydrogen
continued to be generated during that time, as evidenced by the effervescence
off the MSB shell and fuel grid.
.                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 3 of 9


The licensee investigation determined that the source of hydrogen was 
oxidation of zinc in the Carbo Zinc 11 coating (used as a coating to prevent
corrosion) of the MSB when in contact with the borated water in the spent fuel
pool.  The zinc reacted chemically with the acidic borated water from the
spent fuel storage pool to produce hydrogen, and zinc oxide and hydroxide. 
The foam-like substance was determined to be a mixture of a precipitate formed
by the reaction, residual soap used during decontamination of the MSB, and
air.

On June 3, 1996, Confirmatory Action Letters (CALs) were issued to licensees
using the VSC-24 cask (i.e., Arkansas Nuclear One, Palisades Nuclear
Generating Plant, and Point Beach Nuclear Plant).  The CALs confirmed a
commitment, made by each of the three licensees to NRC, that prior to loading
or unloading a VSC-24 cask with spent fuel, or placing a VSC-24 cask into the
spent fuel pool, the licensees will:  (1) assess the potential for the
generation and ignition of explosive gases during all phases of operation of
the VSC-24 system; (2) have compensatory actions in place to minimize the
potential for the generation and ignition of explosive gases; and (3) have
procedures in place to respond in the event of a gas ignition and the
applicable personnel trained and briefed accordingly.

Supplements to each of the June 3, 1996, CALs were subsequently issued on  
June 21 and June 27, 1996, to confirm a further commitment, made by each of
the three licensees to NRC, that the licensees will:  (1) provide the required
response specified in the staff forthcoming generic communication on the Point
Beach VSC-24 event (i.e., this bulletin); and (2) refrain from loading or
unloading a VSC-24 cask, or placing a VSC-24 cask into the spent fuel pool,
pending staff review and acceptance of the response and verification of any
subsequent actions taken in response to this bulletin.

Discussion

An NRC Augmented Inspection Team (AIT) was formed and sent to the site on
May 29, 1996, to examine the circumstances surrounding the event.  The AIT
Charter consisted of evaluations of the licensee response to the event,
including the radiation protection consequences of the event to both the plant
staff and the general public, the effectiveness of the licensee root-cause
investigation, and determination of any potential generic implications of the
event.  The findings of the AIT are summarized below.  Details of the AIT
findings are documented in Report Nos. 50-266/96005 and 50-301/96005.

     �      The AIT agrees with the licensee that the hydrogen was formed by a
            chemical reaction among the Carbo Zinc 11 coating and the borated
            pool water..                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 4 of 9

     
      �     The licensee believes that its internal processes associated with
            the design, design review, design specifications, and the
            independent review of the VSC-24 cask were deficient in that they
            did not recognize that the chemical reaction of the Carbo Zinc 11
            coating with borated water would result in the production of
            hydrogen.

     �      The licensee had several opportunities to identify the generation
            of gas inside the MSB during previous cask loading operations,
            because of several noted abnormalities.  However, the
            abnormalities were not documented and were not viewed
            collectively.  This is of particular concern because in at least
            one case the licensee had indications that the gas being produced
            may have been combustible.

     �      The generic implications of the event extend beyond the use of the
            VSC-24 system.  The AIT recommends that the adequacy of the
            chemical compatibility studies conducted during design reviews for
            all dry cask storage designs and facility environments should be
            reviewed.  In addition, the suitability of Carbo Zinc 11 and other
            similar coatings used in nuclear applications, where there is the
            potential for exposure to boric acid, should be reviewed.

The gas ignition caused no injuries, no radiological releases, and no apparent
damage to the spent fuel, storage cask, or the reactor facility itself. 
However, the event has raised some generic issues pertaining to both spent
fuel storage and transportation casks.  These generic issues extend to spent
fuel transportation casks because transportation casks have designs similar to
storage casks.  Also, the current trend in the industry is to pursue licensing
of dual-purpose, storage/transport cask designs.

Provisions for material suitability exist in both 10 CFR Part 71 and Part 72. 
Notwithstanding, the event at Point Beach suggests that the VSC-24 vendor and
licensees did not adequately consider material reactions, and material
compatibility with possible environments, in the design and design review of
the VSC-24 cask.  Also, NRC did not fully consider material reactions and
material compatibility in its licensing review of the VSC-24 cask and other
storage and transportation casks.

Two specific concerns or issues, one short-term and one long-term, arise from
the event at Point Beach.  The short-term concern is that combustible gases
generated by material reactions may create hazardous conditions while loading
or unloading a cask.  The reactions, which form the combustible gases and
precipitate, would not continue after the loaded casks are drained, evacuated,
and back-filled with inert helium gas.  Thus, the reactions would not continue
while the cask and spent fuel are in the dry storage or transportation mode. 
However, the reactions could resume when the cask is reflooded prior to
unloading, and therefore, need to be considered in cask unloading operations.
.                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 5 of 9


The long-term concern is that any remaining product of these reactions, e.g.,
the precipitate, may degrade the structural integrity and adversely impact the
retrievability of the stored spent fuel.  Degradation of the fuel structural
integrity is not expected to present an immediate health or safety concern. 
However, the overall requirements specified in 10 CFR 72.122 for spent fuel
storage casks include protection of the fuel cladding from degradation and
ready retrieval of the spent fuel.  The staff, based on available information,
does not anticipate any impact on the continued safe confinement of spent
nuclear fuel in existing dry storage installations.

Based on available information and operational experience, it appears that the
VSC-24 design is the system most susceptible to hydrogen gas generation.  This
is due to the VSC-24 relatively large use of carbon steel and anti-corrosion
coatings (i.e., Carbo Zinc 11).  Other licensed or certified cask designs are
primarily constructed of stainless steel, and either use no or limited
quantities of anti-corrosion coatings.  Over 75 non-VSC-24 storage casks have
been loaded, and spent fuel transportation casks have been in use for over 
20 years.  Operational experience with these casks has not evidenced any
problems with material reactions, material incompatibility, and combustible
gas generation during loading or unloading.  Nevertheless, since the licensing
review did not evaluate this issue, this bulletin is intended to request
information to confirm that chemical, galvanic, or other reactions are not a
concern for other storage and transportation cask designs.

Requested Actions

Action addressees are requested to take the following actions:

1.    Address the following items relating to the susceptibility of the spent
      fuel storage or transportation cask design to chemical, galvanic, or
      other reactions:

      (a)   Review the cask materials, including coatings, lubricants, and
            cleaning agents, to determine whether chemical, galvanic, or other
            reactions among the materials, contents, and environment can occur
            during any phase of loading, unloading, handling, storage, and
            transportation.  Consideration should be given to all environments
            that may be encountered under normal, off-normal, or accident
            conditions.

      (b)   Evaluate the effects of any identified reactions to determine if
            any adverse conditions could result during cask operations,
            including loading and unloading.  Consideration should be given,
            but not limited, to:

            (i)   generation of flammable or explosive quantities of hydrogen
                  or other combustible gases; and

            (ii)  increased neutron multiplication in the fuel in a cask 
                  because of boron precipitation from a chemical reaction
                  among the borated water and cask materials..                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 6 of 9

      
      (c)   Review current cask operating procedures to determine if adequate
            controls and procedures are in place to minimize hazardous
            conditions that may be created by any identified reactions.

      (d)   Evaluate the effects of any identified reactions to determine if
            their reaction products could reduce the overall integrity of the
            cask or its contents during storage or transportation.  Determine
            if the reaction products could adversely affect the cask ability
            to maintain the structural integrity and retrievability of the
            spent fuel throughout the term of the license or to transport fuel
            safely.  Consideration should be given, but not limited, to:

            (i)   changes in cask and fuel cladding thermal properties, such
                  as emissivity;

            (ii)  binding of mechanical surfaces, especially fuel-to-basket
                  clearances; and

            (iii) degradation of any safety components, either caused directly
                  by the effects of the reactions, or by the effects of the
                  reactions combined with the effects of long-term exposure of
                  the materials to neutron and gamma radiation, high
                  temperatures, or other possible conditions.

2.    For storage casks currently loaded with spent fuel, determine the
      extent, if any, of the chemical, galvanic, or other reactions that have
      occurred, and the effect of these reactions on the cask ability to
      maintain the structural integrity and retrievability of the spent fuel
      throughout the term of the license. 

In addition to the items above, the vendor and users of the VSC-24 spent fuel
storage cask (Sierra Nuclear Corporation, Arkansas Nuclear One, Palisades
Nuclear Generating Plant, and Point Beach Nuclear Plant) are requested to take
the following actions:

3.    Evaluate the effects of the reaction among Carbo Zinc 11 (or other
      equivalent coating used) and the water environments the cask may
      encounter.  Show that the ability of the cask to maintain the structural
      integrity and retrievability of the spent fuel over a 20-year period has
      not been adversely affected by the formation of precipitate, or by any
      other effects of the reaction.  Justify the continued use of VSC-24
      storage casks already loaded with spent fuel.  In this evaluation,
      consideration should be given, but not limited, to:

      (a)   the effect of the precipitate on fuel cladding integrity;

      (b)   the effect of the precipitate on the heat transfer characteristics
            of the cask; and.                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 7 of 9

      
      (c)   behavior of the precipitate under long-term exposure to neutron
            and gamma radiation, high temperatures, and other possible
            conditions.

4.    Evaluate the procedures for unloading the cask to consider the likely
      presence of hydrogen gas or precipitate inside the MSB and the possible
      adverse effects of the hydrogen gas or precipitate on cask handling and
      performance.  Inform the NRC of any changes made to the unloading
      procedures.

Required Response

Pursuant to 10 CFR 2.204, 10 CFR 71.39, and 10 CFR 72.44(b)(3), in order to
determine whether any license or certificate should be modified, suspended, or
revoked, or other action taken, within 45 days of the date of this bulletin,
action addressees are required to submit a written response indicating whether
the addressee will implement the requested actions.

1.    If the addressee intends to implement the requested actions:

      (i)   The 45-day response should include a report confirming completion
            of the actions requested in Items 1(a), 1(b), and 1(c), above, and
            a schedule for completing implementation of the actions requested
            in Items 1(d) and 2, and if applicable, Items 3 and 4, above.  

      (ii)  Within 30 days of completion of the actions requested in 
            Items 1(d) and 2, and if applicable, Items 3 and 4, above, provide
            a report confirming completion.

      The reports indicated in (i) and (ii) above should provide detailed
      descriptions of the reviews and evaluations performed in response to the
      actions requested, including any compensatory measures implemented as a
      result of the reviews and evaluations performed.

2.    If an addressee chooses not to take the requested actions, the 45-day
      response should provide a description of any proposed alternative course
      of action, the schedule for completing the alternative course of action
      (if applicable), and the safety basis for determining the acceptability
      of the planned alternative course of action.

Address the written response(s) to the U.S. Nuclear Regulatory Commission,
ATTN.:  Document Control Desk, Washington, D.C. 20555, under oath or
affirmation under the provisions of 10 CFR 2.204, 10 CFR 71.39, and
10 CFR 72.44(b)(3).  In addition, submit a copy to the appropriate regional
administrator..                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 8 of 9


Related Generic Communications

IN 95-29,   �Oversight of Design and Fabrication Activities for Metal
            Components Used in Spent Fuel Dry Storage Systems.�

IN 96-34,   �Hydrogen Gas Ignition during Closure Welding of a VSC-24 Multi-
            Assembly Sealed Basket�

Backfit Discussion

This bulletin is an information request made pursuant to 10 CFR 2.204, 10 CFR
71.39, and 10 CFR 72.44(b)(3).  The objective of the actions requested in this
bulletin is to verify that licensees are in compliance with existing NRC rules
and regulations pertaining to the appropriateness and adequacy of the design
of spent fuel storage and transportation casks including, and without
limitation, 10 CFR 71.43(d), 72.122(h), 72.122(l), 72.236(c), 72.236(f),
72.236(g), 72.40(a)(5), 72.212(b)(9), 72.236(h), 72.234(b), and 72.146(b).

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 3150-0011, which expires July 31, 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 U.S. Nuclear
Regulatory Commission is seeking public comment on the potential impact of the
collection of information contained in the bulletin and on the following
issues:

      1.    Is the proposed collection of information necessary for the proper
            performance of the functions of 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 this burden, to the Information and Records
Management Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and to the Desk Officer, Office of Information and Regulatory 
Affairs, NEOB-10202 (3150-0011), Office of Management and Budget,             
Washington, DC 20503..                                                            NRCB 96-04
                                                            July 5, 1996
                                                            Page 9 of 9


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.

Nuclear power reactor licensees with questions regarding this matter should
contact the appropriate Office of Nuclear Reactor Regulation (NRR) project
manager.  All other addressees with questions regarding this matter should
contact the technical contact listed below or the appropriate Office of
Nuclear Material Safety and Safeguards (NMSS) project manager.


ORIGINAL SIGNED BY                              ORIGINAL SIGNED BY

Brian K. Grimes, Acting Director                William D. Travers, Director
Division of Reactor Program Management          Spent Fuel Project Office
Office of Nuclear Reactor Regulation            Office of Nuclear Material
                                                  Safety and Safeguards

Technical Contact:  Marissa Bailey, NMSS
                    (301) 415-8531
                    E-mail:  mgb@nrc.gov

Lead Project Manager:  William Reckley, NRR
                       (301) 415-1314
                       E-mail:  wdr@nrc.gov


 

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