[Federal Register Volume 62, Number 121 (Tuesday, June 24, 1997)]
[Notices]
[Pages 34091-34095]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-16484]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-255, 50-266, 50-301, 50-313, 50-368, 72-5, 72-7, 72-13, 
and 72-1007]


Consumers Power Company (Palisades Nuclear Plant), Wisconsin 
Electric Power Company (Point Beach Nuclear Plant, Units 1 and 2), 
Entergy Operations, Inc. (Arkansas Nuclear One, Units 1, and 2); 
Issuance of Director's Decision Under 10 CFR 2.206

    Notice is hereby given that the Director, Office of Nuclear 
Material Safety and Safeguards, has issued a Director's Decision 
concerning a Petition dated October 18, 1996, filed by Don't Waste 
Michigan and the Lake Michigan Federation (Petitioners) under Section 
2.206 of Title 10 of the Code of Federal Regulations (10 CFR 2.206). 
The Petition requested that the U.S. Nuclear Regulatory Commission 
order all users of Ventilated Storage Casks (VSC-24) systems to refrain 
from loading any casks until the certificate of compliance (COC), 
safety analysis report (SAR), and safety evaluation report (SER) are 
amended to include operating controls and limits to prevent hazardous 
conditions. Such conditions include the generation of explosive gases, 
caused by the interaction between the VSC materials and the 
environments, encountered during loading, storage, and unloading.
    Further, Petitioners claim the VSC-24 should not be used until: (i) 
An independent third-party review team has examined the safety issues 
they raise; (ii) the potential impacts of all material aspects of the 
casks have been fully assessed; (iii) there is experimental 
verification of temperature calculations and heat transfer assessments 
and other design assumptions; (iv) the safety of the material coatings 
on components and structures has been justified; and (v) the SAR, SER, 
and COC are amended to include the necessary operating control and 
limits to direct safe use of the VSC-24.
    The Director of the Office of Nuclear Material Safety and 
Safeguards has determined that the Petition should be denied for the 
reasons stated in the ``Director's Decision Under 10 CFR 2.206'' (DD-
97-15), the complete text of which follows this notice. The decision 
and documents cited in the decision are available for public inspection 
and copying in the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC.
    A copy of this decision has been filed with the Secretary of the 
Commission for the Commission's review in accordance with 10 CFR 
2.206(c). As provided therein, this decision will become the final 
action of the Commission 25 days after issuance unless the Commission, 
on its own motion, institutes review of the decision within that time.

    Dated at Rockville, Maryland, this 18th day of June, 1997.

    For the Nuclear Regulatory Commission.
Malcolm R. Knapp,
Acting Director, Office of Nuclear Material Safety and Safeguards.

Director's Decision Under 10 CFR 2.206

[DD-97-15]

I. Introduction

    On October 18, 1996, Don't Waste Michigan and the Lake Michigan 
Federation (Petitioners) filed a Petition pursuant to Section 2.206 of 
Title 10 of the Code of Federal Regulations (10 CFR 2.206) requesting 
that the U.S. Nuclear Regulatory Commission take the following action:

    Prohibit loading of Ventilated Storage Casks (VSC-24s) until the 
certificate of compliance (COC), the Safety analysis report (SAR), 
and the safety evaluation report (SER) are amended following an 
independent, third-party review of the VSC-24 design, to address 
concerns raised by the Petitioners' engineering consultant, Dr. 
Rudolf Hausler.

    The Petition has been referred to me pursuant to 10 CFR 2.206. By 
letter dated December 10, 1996, to Dr. Mary Sinclair and Ms. Eleanor 
Roemer, on behalf of the Petitioners, NRC acknowledged receipt of the 
Petition and provided the NRC staff's determination that the Petition 
did not require immediate action by the NRC. Notice of receipt was 
published in the Federal Register on January 13, 1997 (62 FR 1783).
    On the basis of the NRC staff's evaluation of the issues and for 
the reasons given below, I have determined that the Petitioners' 
request should be denied.

II. Background

    On May 28, 1996, a hydrogen gas ignition occurred during the 
welding of the shield lid after spent fuel had been loaded into a VSC-
24 at the Point Beach Nuclear Plant. The hydrogen was formed by a 
chemical reaction between a zinc-based coating (Carbo Zinc 11) and the 
borated water in the spent fuel pool. On June 3, 1996, the NRC issued 
confirmatory action letters (CALs) to those licensees using or planning 
to use VSC-24s for dry storage of spent nuclear fuel, i.e., licensees 
for Point Beach Nuclear Plant, Palisades Nuclear Generating Plant, and 
Arkansas Nuclear One (ANO). The CAL issued to the licensee for ANO was 
supplemented on June 21, 1996, and the CALs issued to the licensees for 
Point Beach and Palisades were supplemented on June 27, 1996. The CALs, 
as supplemented, documented the licensees' commitments not to load or 
unload a VSC-24 without resolution of material compatibility issues 
identified in a forthcoming general communication and subsequent NRC 
confirmation of corrective actions taken by the licensees. The generic 
communication was issued on July 5, 1996, in the form of NRC Bulletin 
96-04, ``Chemical, Galvanic, or Other Reactions in Spent Fuel Storage 
and Transportation Casks.'' NRC Bulletin 96-04 notified addressees 
about the potential for adverse 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. The 
actions requested in Bulletin 96-04 included reviewing the cask 
materials for potential adverse reactions, evaluating the short-term 
and long-term effects of any identified reactions, and

[[Page 34092]]

determining the adequacy of cask operating procedures to minimize the 
consequences of any identified reactions. The NRC staff has 
acknowledged that the event demonstrated that the cask vendor's (Sierra 
Nuclear Corporation) SAR for the VSC-24 and related NRC review, as 
documented in the NRC staff's SER, did not adequately address the use 
of a zinc-based coating and its reaction with the acidic water in spent 
fuel pools.
    In response to Bulletin 96-04 and to subsequent NRC staff 
inquiries, the licensees for ANO, Point Beach, and Palisades submitted 
to the NRC evaluations of possible material interactions and the 
effects of such interactions on cask performance and operation. The 
licensees also submitted information on the operating controls and 
limits that were implemented to prevent hazardous conditions which may 
result from adverse material interactions. The operating controls and 
limits included controls for the environments that the casks encounter 
during use, requirements for inspections and environmental sampling, 
and additional precautions for various cask operations.
    The NRC staff evaluated the responses submitted by the licensee for 
ANO. As documented in the staff's safety evaluation dated December 3, 
1996, the staff determined that the licensee's submittals provided the 
necessary level of confidence that the VSC-24 can be used to safely 
store spent fuel over the 20-year period of the certificate. The staff 
also determined that the operating controls and limits proposed by the 
licensee are acceptable and satisfy regulatory requirements. By a 
separate letter, also dated December 3, 1996, the staff informed the 
licensee for ANC that its corrective actions had been verified by 
inspections performed by the NRC staff. Shortly thereafter, the 
licensee initiated cask loading activities.
    The NRC staff also evaluated the responses submitted by the 
licensees for Point Beach and Palisades. As documented in the staff's 
safety evaluations dated respectively April 8, 1997, and June 12, 1997, 
the staff determined that the licensees evaluations and proposed 
operating controls and limits are acceptable and satisfy regulatory 
requirements. However, the CALs placed on Point Beach and Palisades 
still remain in place until an NRC inspection is performed to verify 
that the licensees' corrective actions are properly implemented.

III. Discussion

    The Petition requests an NRC order to users of VSC-24s not to load 
additional casks until: (1) The COC, SAR, and SER are amended to 
contain operating controls and limits to prevent hazardous conditions; 
(2) an independent third-party review team has examined the safety 
issues raised by the Petitioners; (3) the potential impacts of all 
material aspects of the casks have been fully assessed; (4) there is 
experimental verification of temperature calculations and heat transfer 
assessments and other design assumptions; and (5) the safety of the 
material coatings on components and structures has been justified.

Item 1: Prohibit Loading of VSC-24s Pending Amendment of Documents

    As noted in the NRC letter to the Petitioners on December 10, 1996, 
the Petitioners' request to amend the COC, SAR, and SER is similar to a 
request made by the Citizen's Utility Board (CUB) in a Petition dated 
September 30, 1996. The NRC staff denied the CUB petition on April 17, 
1997, for the reasons that are identical to the reasons stated here in 
denying the first part of the Petitioners' request.
    The circumstances set forth above made clear that, following the 
event at Point Beach, the NRC staff recognized that additional 
evaluation of potential material interactions was warranted for all 
spent fuel transportation and storage casks. In regard to the VSC-24, 
the event and subsequent NRC inspections made it apparent that actual 
changes in the operating procedures or the design of the cask would be 
necessary. CALs were issued to confirm licensees' commitments to 
refrain from loading VSC-24s pending completion of the NRC staff's 
review of the responses to Bulletin 96-04 and verification of the 
associated corrective actions. As discussed, the CALs established a 
process by which the NRC staff could obtain confidence that operating 
controls and limits to address potential hazardous conditions are 
developed and implemented by each licensee using VSC-24s.
    In particular, the CAL process ensures that licensees will 
incorporate the necessary operating controls and limits into revised 
plant procedures. Moreover, under existing NRC requirements, the 
licensee must adequately implement those revised procedures. For this 
reason, no changes to the COC or SAR are needed to ensure that 
enforceable operating controls and limits are in place to address 
potential hazardous conditions during the loading or unloading of a 
cask. Further, as previously indicated, the NRC staff has documented 
the process, information, and results of its review of the licensees' 
responses to Bulletin 96-04 for use of the VSC-24 at ANO, Point Beach, 
and Palisades in safety evaluations available for public review.
    Although the actions taken as part of the CAL process provide 
adequate assurance that technical and regulatory compliance issues 
raised by the event at Point Beach will be resolved before a licensee 
loads or unloads a VSC-24, the NRC staff agrees with the Petitioners 
that it would be beneficial if the SAR and other licensing basis 
documents accurately describe the identified chemical reaction and the 
associated operating controls and limits. The NRC staff is currently 
reviewing a proposed amendment to the SAR and COC for the VSC-24 design 
and will ensure that the information related to the identified chemical 
reaction and associated operating controls is adequately addressed in 
the appropriate licensing-basis documents. In addition, the NRC staff 
is processing a petition for rulemaking, PRM-72-3, that may lead to 
additional updating of independent spent fuel storage installation SARs 
and the inclusion of information on operating controls and limits 
implemented as a result of the event at Point Beach. However, the 
previously discussed controls to be implemented by the licensees and 
verified by the staff as part of the CAL process, and the 
enforceability of those controls under existing NRC requirements, make 
it unnecessary to require revision of the specific licensing documents 
cited by the Petitioners as a precondition for resuming cask operations 
at the facilities using VSC-24s. Therefore, there would be no 
regulatory basis for granting the first part of the Petition to require 
amendment of the COC, SAR, or SER before further loading of VSC-24s.

Item 2: Prohibit Loading of VSC-24s Pending Independent, Third-Party 
Review

    Petitioners request the NRC to prohibit loading of VSC-24s until 
the COC, SAR and SER are amended following an independent, third-party 
review to address concerns raised by the Petitioners. The NRC staff 
performed a review of the VSC-24 design prior to certification in 1993. 
As a result of the review, the staff determined that the design and 
operation of the cask system is in compliance with 10 CFR Part 72. The 
staff also concluded, with a high degree of assurance, that the VSC-24 
will safely store spent fuel over the 20-year period of the 
certificate. Notwithstanding the staff's review and

[[Page 34093]]

determination in 1993, the Petitioners are claiming that a new, 
independent review is needed before further VSC-24s are loaded.
    While the event at Point Beach revealed the need for additional 
evaluation by licensees and NRC of potential material interactions in 
the VSC-24 (and other transportation and storage casks), the actions 
already taken, in the staff's judgment, provide an adequate response. 
In particular, Bulletin 96-04 was issued to request additional 
information from licensees using the VSC-24 on material interactions 
and compatibility in the VSC-24 and on the corrective actions 
implemented. The NRC staff then received and reviewed the responses 
submitted by the licensees for ANO, Point Beach, and Palisades. The 
staff's reviews (as well as the licensees') have been exhaustive and 
were performed by an inter-disciplinary team of engineers knowledgeable 
in materials, corrosion, metallurgy, chemistry, structural engineering, 
heat transfer, nuclear engineering, and other technical fields needed 
to perform the review. The results of the staff's reviews, including 
the necessary corrective actions, are documented and justified in the 
staff's December 3, 1996, April 8, 1997, and June 12, 1997, safety 
evaluations. These corrective actions include: cleanliness checks 
before placing the cask in the spent fuel pool, venting and monitoring 
of the air space beneath the VSC-24 shield lid during welding or 
cutting activities, discontinuing welding or cutting should the 
hydrogen concentration exceed 0.4% by volume (10% of the minimum amount 
necessary for a combustible concentration), and sampling the boron 
concentration in the spent fuel pool and multi-assembly sealed basket 
(MSB) water. While the staff agreed that the corrective actions were 
necessary to prevent hazardous conditions during the loading and 
unloading of VSC-24s, the information submitted by the Petitioners does 
not raise any new issues or provide any reason for the staff to 
question its conclusion that the VSC-24 will safely store spent fuel 
over the 20-year period of the certificate.
    In reaching this conclusion, the NRC staff evaluated the specific 
concerns raised by the Petitioners related to the design of the VSC-24. 
The staff believes that these concerns have already been addressed by 
the recent evaluations submitted in response to Bulletin 96-04, by 
information submitted to NRC to support the certification of the VSC-24 
design in 1993, or by other information submitted in support of NRC 
review and inspection activities. Each of the Petitioners's specific 
concerns is addressed below.
    (i) The Petitioners claim that the cask design allows for fuel 
elements to be in contact with the zinc primer creating a galvanic 
couple which will accelerate the corrosion of the zinc. The NRC staff 
considered galvanic effects between the Zircaloy fuel rods and the 
Carbo Zinc 11 coating. The staff agrees that a galvanic effect would 
increase the corrosion rate of the zinc, with a corresponding increase 
in the hydrogen gas generation rate, as the zinc in the Carbo Zinc 11 
coating is polarized to a more active potential. However, in the VSC-24 
design, several factors reduce the amount of zinc polarization such 
that there would not be a significant increase in hydrogen generation. 
One factor is the contact resistances between the stainless steel fuel 
assembly end-fittings and the Zircaloy fuel rods and between the end-
fittings and the Carbo Zinc 11 paint. Another factor is the geometry of 
the VSC-24 and the fuel assemblies. The fuel assemblies are placed in 
fuel storage sleeves with a clearance of approximately 0.1 inch to 0.5 
inch between the sides of the fuel assembly and the sleeves. This 
clearance and the physical design of the fuel assemblies create 
shielding between the fuel rod surfaces and the Carbo Zinc 11 coating. 
This shielding effectively reduces the galvanic action between the 
Zircaloy fuel rods and the Carbo Zinc 11 coating. The Zircaloy fuel 
rods could contact the Carbo Zinc coated sleeves if the fuel assembly 
is not centered in the storage sleeves or if the fuel rods are bowed. 
However, the shielding effect and small Carbo Zinc/Zircaloy contact 
area would still prevent significant galvanic action. Hydrogen 
concentration measurements made at Point Beach and the hydrogen 
monitoring performed at ANO during loading of a VSC-24 in December 1996 
(NRC Inspection Report Nos. 50-313/96-25 and 72-13/96-02) support the 
conclusion that significant galvanic action between the Zircaloy and 
zinc coating, and hence, increased hydrogen generation, is not 
occurring in the VSC-24. In addition, even if there was an increase in 
hydrogen generation because of the galvanic action, the staff has 
determined that the controls implemented by the licensees for ANO and 
Point Beach would prevent accumulation of a combustible concentration 
of hydrogen and its ignition. The staff will also review and verify the 
adequacy of the controls implemented by the licensee for Palisades.
    (ii) The Petitioners claim that there were numerous discrepancies 
in the responses to Bulletin 96-04. As noted, the NRC staff completed 
its review of responses for ANO, Point Beach, and Palisades. The staff 
found these responses to be acceptable and found no discrepancies of 
concern. There were minor differences in the operating controls 
implemented at the three facilities. However, the staff reviewed these 
controls and concluded that all three sets of controls are adequate to 
preclude hazardous conditions during cask operation.
    (iii) The Petitioners claim that the epoxy-coating applied to the 
exterior of the Multi-Assembly Sealed Basket (MSB) could not withstand 
the temperatures developed during long-term storage. Technical data on 
the type of epoxy coating used on the MSB were provided by the 
licensees in their responses to Bulletin 96-04. The data show that the 
epoxy is temperature-resistant up to 350 deg.F. The SAR for the VSC-24 
(which the staff reviewed and accepted prior to certification in 1993) 
shows that under normal or off-normal storage conditions, the 
temperature of the MSB exterior will not exceed 300 deg.F. for the 
maximum allowable heat load of 24 kW and, therefore, will not degrade 
the epoxy.
    (iv) The Petitioners claim that the low-temperature specification 
in the COC for moving the VSC-24 MSB was not properly translated to the 
MSB shell material compositions. Low-temperature embrittlement of the 
MSB shell material was evaluated by the NRC staff during its safety 
review before certification of the VSC-24. The composition of the MSB 
shell material (SA516, Grade 70 carbon steel) is specified in the 
American Society for Mechanical Engineers, Boiler & Pressure Vessel 
Code, Section II, SA-516, ``Specification for Pressure Vessel Plates, 
Carbon Steel, for Moderate- and Lower-Temperature Service.'' The impact 
testing requirements for the MSB material are found in American Society 
for Testing and Materials Specification A370 (ASTM A370). ``Methods and 
Definitions for Mechanical Testing of Steel Products.'' As specified in 
the COC, SER, and SAR, each MSB shell material must be shown, during 
fabrication, by Charpy test per ASTM A370, to have 15 ft-lbs of 
absorbed energy at -50  deg.F. Further, movement of the MSB must occur 
only at ambient temperatures of 0  deg.F or above to avoid potential 
brittle fracture of the MSB material.\1\ The NRC staff considers the

[[Page 34094]]

50  deg.F temperature difference to provide sufficient margin because 
it places the MSB material at a temperature that is significantly above 
the temperature where brittle fracture could occur. It should also be 
noted that the temperature of the MSB shell itself would actually be 
substantially higher than the ambient temperature (e.g., 20 deg.F for 
25-year-old fuel), thus providing an even higher margin. In addition, 
it is highly unlikely that any MSB movement activity would take place 
at temperature below 0 deg.F.
---------------------------------------------------------------------------

    \1\ At Palisades, the licensee has administratively set a 
minimum ambient temperature of 10  deg.F for moving the first four 
MSBs (CMSB-01 through -04) to be loaded because the shell material 
for these MSBs does not have 15 ft-lbs of absorbed energy at -50  
deg.F. Rather, these MSBs have 15 ft-lbs of absorbed energy at -40  
deg.F. Thus, to retain the 50  deg.F temperature margin, the 
licensee has restricted movement of these four MSBs to an ambient 
temperature of 10  deg.F or above. The NRC staff has reviewed and 
approved the licensee's administrative limit, as documented in NRC 
safety evaluation dated September 26, 1995.
---------------------------------------------------------------------------

    (v) The Petitioners claim that zinc-steel interaction at 800  deg.F 
to 1000  deg.F and possible steel embrittlement over a 20-year period 
were not considered. Zinc-steel interaction at the 800  deg.F to 1000  
deg.F temperature range was not considered and is not a concern 
because, as documented in the VSC-24 SAR, temperatures in the MSB will 
not reach 800  deg.F during storage. Maximum temperatures would be 688  
deg.F under normal conditions and 708  deg.F under off-normal 
conditions, for the maximum allowable heat load of 24 kW. Furthermore, 
over the storage period, the temperatures within the MSB will continue 
to decrease as the heat load decreases due to the decay of the spent 
fuel.
    (vi) The Petitioners claim that the effect of molten zinc on 
Zircaloy has not been verified experimentally. The NRC staff evaluated 
the durability and behavior of the zinc coating under the range of 
storage temperatures. The presence of molten zinc is not expected under 
the storage temperatures and conditions, thus the effect of molten zinc 
on Zircaloy is not a concern. However, as documented in the staff's 
safety evaluations for ANO (dated December 3, 1996), Point Beach (dated 
April 8, 1997), and Palisades (dated June 12, 1997), the staff did 
evaluate the potential interaction between zinc vapor and Zircaloy and 
the effect of this interaction. Based on the information provided in 
the responses to Bulletin 96-04, the staff concluded that the potential 
interaction between zinc vapor and Zircaloy presented no immediate or 
long-term safety concern for the spent fuel stored in the VSC-24.
    (vii) The Petitioners claim that the vacuum-drying process does not 
seem to have been experimentally verified. Vacuum drying is a well-
established, widely used method for removing moisture from spent fuel 
storage and transportation casks. The process used for the VSC-24 is a 
common process, which the NRC staff evaluated and determined to be 
acceptable during the safety review before certification in 1993. In 
the staff's judgment, experimental testing to verify a well-established 
process is unnecessary.
    (viii) The Petitioners claim that the thermal analyses for the VSC-
24 have not been experimentally verified. The thermal analyses for the 
VSC-24 contained conservative key assumptions, including a total heat 
generation of 1 kW per assembly (a total of 24 kW per cask). This 
assumption is conservative because it is highly unlikely that each 
assembly loaded in the cask will generate 1 kW of heat. In addition, 
the assembly and total cask heat loads will continually decrease over 
time as the spent fuel decays. In light of the conservatisms in the 
thermal analyses, the staff does not see the need for requiring 
experimental verification of the VSC-24 thermal analyses. Nevertheless, 
the COC requires that a thermal test be performed on the first VSC-24 
to be loaded. The purpose of the test is to measure the heat removal 
performance of the VSC-24 system. The licensee for Palisades performed 
such a test and summarized its results in a letter to NRC dated June 
10, 1993. The temperatures measured during the test were lower than the 
predicted temperatures. The results thus indicate that the VSC-24 
performs its intended heat removal function. The thermal test at 
Palisades was performed with a 12 kW heat load. To date, no VSC-24s 
have been loaded with greater than 12 kW heat load. As required by the 
COC, the thermal test must be performed for the first cask to use any 
higher heat loads, up to 24kW.
    The NRC staff believes, based on the foregoing, that an 
independent, third-party review is not warranted by the Petitioners' 
specific concerns. However, NRC review activities relating to the VSC-
24 will nonetheless continue. In particular, NRC inspection activities 
at the facilities operated by the licensees, the VSC-24 vendor, and the 
VSC-24 fabricators may lead to additional reviews of the VSC-24. In 
addition, the staff is currently reviewing a proposed amendment, 
submitted by the VSC-24 vendor, to the SAR and COC for the VSC-24 
design. This review will be performed in accordance with the staff's 
``Standard Review Plan for Dry Cask Storage Systems'' (NUREG-1536) to 
ensure the thoroughness, quality, and consistency of the review. Where 
relevant, recent operational, technical, and safety issues related to 
the VSC-24 design will be considered by the staff in this review.\2\
---------------------------------------------------------------------------

    \2\ Recent concerns relating to the MSB closure welds, as 
documented in NRC Inspection Report No. 72-1007/97-204, dated April 
15, 1997, may result in further evaluations of the VSC-24 design 
and, if necessary, appropriate regulatory action to ensure continued 
safe use of the VSC-24.
---------------------------------------------------------------------------

    In addition, it is my judgment that the NRC staff is fully capable 
of fulfilling the responsibility for reviewing, approving, and 
certifying dry cask storage systems to be used under 10 CFR Part 72 
which, by law, belongs to the NRC. In conducting its review, the NRC 
staff must have reasonable assurance that the cask system will safely 
store spent fuel over the period of the certificate. Further, the staff 
will assign the necessary resources and expertise to perform such 
reviews. When the NRC staff lacks either the resources or expertise to 
perform all or portions of the review in-house, the NRC may, and does, 
supplement its own ranks by using outside specialists.

Item 3: Prohibit Loading of VSC-24s Pending Assessment of Cask 
Materials

    Petitioners request the NRC to prohibit loading of VSC-24s until 
the potential impacts of all material aspects of the casks have been 
fully assessed. As previously stated, Bulletin 96-04 was issued to 
request information on material interactions and compatibility in spent 
fuel storage and transportation casks. In response to this request, the 
licensees for ANO, Point Beach, and Palisades submitted evaluations on 
possible material interactions in the VSC-24 and the effects of such 
interactions on cask performance and operation. The only significant 
material interaction identified was between the zinc-based coating and 
the borated spent fuel pool water. As previously discussed, the 
operating controls and limits put in place by the licensees provide an 
adequate level of confidence to prevent the adverse effects of this 
interaction (generation and possible ignition of hydrogen gas and 
possible depletion of boron in the water). The staff reviewed these 
evaluations and, based on the information provided, concluded that none 
of the identified material interactions would adversely affect the VSC-
24's ability to safely store spent fuel over the 20-year period of the 
certificate. The results of the staff's reviews are documented in the 
staff's December 3, 1996, April 8, 1997, and June 12, 1997, safety 
evaluations for ANO, Point Beach, and Palisades, respectively.

[[Page 34095]]

Item 4: Prohibit Loading of VSC-24s Pending Experimental Verification 
of Thermal and Other Design Assumptions

    Petitioners request the NRC to prohibit loading of VSC-24s until 
there is experimental verification of temperature calculations and heat 
transfer assessments and other design assumptions. The thermal and 
other engineering and design analyses for the VSC-24 contained 
conservative key assumptions which are discussed in the SAR and SER. In 
addition, the acceptance criteria for these analyses have margins of 
safety that the staff considers to be sufficient. In light of the 
conservatisms and safety margins in the thermal and other analyses, the 
staff does not see the need for requiring experimental verification of 
the thermal and other design assumptions used in evaluating the VSC-24.

Item 5: Prohibit Loading of VSC-24s Pending Assessment of Material 
Coatings

    Petitioners request the NRC to prohibit loading of VSC-24s until 
the safety of the material coatings on components and structures has 
been justified. As discussed above, material interactions within the 
VSC-24 and their effect on cask operations and performance were 
evaluated by the licensees in response to Bulletin 96-04 and reviewed 
by the staff. Specifically, the licensees evaluated, and the staff 
reviewed, the use of the zinc-based coating, its reaction with borated 
water and other cask environments, and the effect of the reaction or 
reaction products on cask operations and on the performance of the 
various cask components and structures. The staff concluded that use of 
existing VSC-24s with the zinc-based coating is acceptable in light of 
the operating controls and limits for preventing hazardous conditions 
that must be properly implemented by licensees during cask loading and 
unloading. Based on the information provided, the staff also concluded 
that neither the coating itself, nor its reaction with borated water or 
other cask environments, would have an adverse effect on the 
performance of the cask components or structures during the period of 
spent fuel storage.

IV. Conclusion

    The Petitioners requested that the NRC prohibit loading of VSC-24s 
until the COC, SAR, and SER are amended to contain operating controls 
and limits to prevent hazardous conditions. After reviewing each of the 
Petitioners' claims, I conclude that, for the reasons discussed above, 
no adequate basis exists for granting the Petitioners' request to 
prohibit licensees' use of the VSC-24 for dry cask storage of spent 
nuclear fuel at Palisades, Point Beach, or ANO pending: (1) Revision of 
the SAR, SER, and COC for the VSC-24 to contain operating controls and 
limits to prevent hazardous conditions: (2) an independent third-party 
review to examine the safety issues raised by the Petitioners; and (3) 
experimental verification of temperature calculations and heat transfer 
assessments and other design assumptions. Furthermore, I conclude that 
the Petitioners' other two requests, an assessment of potential impacts 
of VSC-24 material aspects and a safety justification of material 
coatings on components and structures, have already been fulfilled 
through the staff's review of the licensees' responses to Bulletin 96-
04.
    A copy of this decision will be filed with the Secretary of the 
Commission for the Commission to review in accordance with 10 CFR 
2.206(c).
    As provided by this regulation, this decision will constitute the 
final action of the Commission 25 days after issuance, unless the 
Commission, on its own motion, institutes a review of the decision 
within that time.

    Dated at Rockville, Maryland, this 18th day of June, 1997.

    For the Nuclear Regulatory Commission.
Malcolm R. Knapp,
Acting Director, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 97-16484 Filed 6-23-97; 8:45 am]
BILLING CODE 7590-01-M