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

NUCLEAR REGULATORY COMMISSION

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

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 Reactor

Regulation, has issued a Director's Decision concerning a Petition

dated September 30, 1996, filed by Citizens' Utility Board (Petitioner)

under Section 2.206 of Title 10 of the Code of Federal Regulations (10

CFR 2.206). The Petition requested that the NRC (1) Require Wisconsin

Electric Power Company to retain 24 empty and available spaces in the

Point Beach Nuclear Plant spent fuel pool to accommodate retrieval of

spent fuel from a VSC-24 cask, and (2) prohibit loading of VSC-24 casks

until the Certificate of Compliance, the Safety Analysis Report, and

the Safety Evaluation Report are amended to contain operating controls

and limits to prevent hazardous conditions.

The Director of the Office of Nuclear Reactor Regulation has

determined that the Petition should be denied for the reasons stated in

the ``Director's Decision Under 10 CFR 2.206'' (DD-97-09), 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 17th day of April 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor Regulation.

Director's Decision Under 10 CFR 2.206

I. Introduction

On September 30, 1996, Citizens' Utility Board 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 (NRC) take the following actions:

1. Order Wisconsin Electric Power Company (WEPCO) to retain 24

empty and available spaces in the Point Beach Nuclear Plant spent

fuel pool to provide the capability to permit retrieval of spent

fuel from a VSC-24 cask in the event of an accident requiring

removal of spent fuel from the cask or in the event that conditions

of the certificate of compliance (COC) for the VSC-24 require

removal of spent fuel from the cask, until such time that WEPCO has

other options available to it to remove spent fuel from a cask in

the event conditions warrant it; and

2. Order users of the VSC-24 cask not to load VSC-24 casks until

the COC, safety analysis report (SAR), and safety evaluation report

(SER) are amended to contain operating controls and limits that

prevent hazardous conditions, including but not limited to the

generation of explosive gases, due to VSC-24 material reactions with

environments encountered during loading, storage, and unloading of

the VSC-24 cask. The SAR and SER must be amended such that each

operating control and limit is clearly documented and justified in

the technical review sections of the SAR and associated SER as

necessary and sufficient for safe cask operation.

The Petition has been referred to me pursuant to 10 CFR 2.206. The

NRC letters dated October 11 and December 10, 1996, to Mr. Dennis Dums,

on behalf of the Petitioner, 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 December 16, 1996 (61 FR 66063).

[[Page 19821]]

On the basis of the NRC staff's evaluation of the issues and for

the reasons given below, the Petitioner's requests are denied.

II. Background

The Petitioner's first request is for the NRC to order WEPCO to

maintain sufficient empty space in the spent fuel pool at Point Beach

to accommodate the unloading of a VSC-24 spent fuel storage cask. NRC

regulations include a requirement that an independent spent fuel

storage installation (ISFSI) be designed to provide for the ready

retrieval of spent fuel or high-level radioactive waste for further

processing or disposal. This requirement is applicable to ISFSIs so

that the stored spent fuel can be retrieved for transport to either a

monitored retrievable storage installation (MRS) or a high-level waste

repository whenever it becomes available. This regulation, 10 CFR

72.122(l), provides as follows:

(1) Retrievability. Storage systems must be designed to allow

ready retrieval of spent fuel or high-level radioactive waste for

further processing or disposal.

In addition to the regulatory requirements in Section 72.122(l)

pertaining to retrieval of the fuel assemblies for further processing

or disposal, there are certain events or conditions that could warrant

removing a VSC-24 cask from an ISFSI and returning the multi-assembly

sealed basket (MSB) to the spent fuel pool and unloading the stored

fuel assemblies. The COC requires a VSC-24 cask to be returned to the

spent fuel pool in response to those design basis events or conditions

that may challenge the integrity of the storage cask or the cladding of

the spent fuel assemblies.1

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\1\ The following sections of the COC include requirements for

returning a VSC-24 cask to the spent fuel pool and/or unloading the

cask:

Section 1.2.3, ``Maximum Permissible Air Outlet Temperature'';

Section 1.2.10, ``Time Limit for Draining the MSB'';

Section 1.2.15, ``Handling Height''; and

Section 1.3.4, ``Thermal Performance.''

Each section is discussed later in this decision.

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Petitioner's second request is for an NRC order to WEPCO and other

users of VSC-24 casks not to load additional casks until the COC, the

SAR, and the SER are amended to contain operating controls and limits

to prevent hazardous conditions. 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 cask 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.

Following the event, the NRC issued confirmatory action letters (CALs)

to those licensees using or planning to use VSC-24 casks for the

storage of spent nuclear fuel (i.e., licensees for Point Beach,

Palisades, and Arkansas Nuclear One). The CALs documented the

licensees' commitments not to load or unload a VSC-24 cask without

resolution of material compatibility issues identified in NRC Bulletin

96-04, ``Chemical, Galvanic, or Other Reactions in Spent Fuel Storage

and Transportation Casks,'' dated July 5, 1996, and subsequent

confirmation of corrective actions by the NRC. The staff has

acknowledged that the event demonstrated that the SAR and related NRC

review, as documented in the SER, did not adequately address the use of

a zinc-based coating and its reaction with the acidic water in spent

fuel pools.

The licensees using VSC-24 casks submitted to the NRC information

on operating controls and limits to prevent hazardous conditions

implemented in response to NRC Bulletin 96-04 and subsequent staff

inquiries. The submittals from the licensees included evaluations of

possible material interactions and provided descriptions of how

procedures were revised. The revisions include 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 has evaluated these responses

for Arkansas Nuclear One (ANO) and Point Beach and, as documented in

the safety evaluations dated December 3, 1996, and April 8, 1997,

determined that the operating controls and limits proposed by these

licensees are acceptable and satisfy regulatory requirements. By a

separate letter also dated December 3, 1996, the staff informed the

licensee for ANO that its corrective actions had been verified by

inspections performed by the NRC staff. Shortly thereafter, the

licensee initiated cask loading activities.2 The NRC will

perform inspections in the near future in order to verify corrective

actions implemented at Point Beach. The review of responses to the

bulletin related to Palisades is ongoing. Cask operations at Point

Beach and Palisades continue to be limited by the licensees'

commitments described in CALs.

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\2\ The NRC staff is looking into reports from licensees on the

need to perform weld repairs during the welding of the shield lid

into the MSBs of several VSC-24 casks. This potential problem is not

related to the requested actions or supporting information cited in

the Petition. The NRC staff determined that the issuance of this

Director's Decision should not be delayed pending resolution of

potential problems associated with the weld repairs because the weld

repairs are not related to concerns presented in the Petition and

the welding issue is being addressed by ongoing NRC activities. The

Petitioner was informed of the welding issue and the NRC staff's

decision to not include the issue in the staff's evaluation of the

Petition.

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III. Discussion

As noted, the Petition requests two actions be taken by the NRC.

They are addressed below.

Item 1: Order WEPCO To Retain 24 Spaces in the Point Beach Spent Fuel

Pool

The first requested action calls for the NRC to issue an order to

WEPCO to retain 24 empty and available spaces in the Point Beach spent

fuel pool to provide the capability to unload a VSC-24 dry storage

cask. The two basic reasons to return a cask to the spent fuel pool

would be either to (1) Retrieve the fuel assemblies for further

processing or disposal pursuant to 10 CFR 72.122(l) or (2) respond to

an event or condition that has potentially degraded the cask or spent

fuel in regard to the requirements established in the COC.

As previously discussed, 10 CFR 72.122(l) sets forth requirements

pertaining to retrieval of the fuel for further processing or disposal;

however, it provides no basis for the NRC to require a licensee to

maintain a specified reserve capacity in the spent fuel pool. Licensees

will have considerable opportunity to plan and schedule the activities

associated with retrieving fuel assemblies from existing storage casks

for transfer to other casks for further processing or disposal. This

ability to control the activity includes either ensuring that existing

spent fuel pool facilities will support the transfer or developing

alternate approaches. Alternate approaches could involve, for example,

making room in spent fuel pools by use of other storage or

transportation casks, expanding the wet storage capacity by making

changes to the spent fuel pool or other parts of the reactor facility,

or development of a system for direct cask-to-cask transfer under dry

conditions. Therefore, the design requirement for ready retrieval in 10

CFR 72.122(l) does not provide a basis for issuing an order as

requested by the Petitioner.

Similarly, requiring the licensee to maintain space in the spent

fuel pool is not necessary as a contingency for certain events or

conditions for which a cask must be returned to the spent fuel pool to

facilitate inspections or ensure adequate cooling of the fuel

assemblies. During its reviews performed during certification of the

VSC-24 design, the NRC staff confirmed that the design features of the

cask provide reasonable

[[Page 19822]]

assurance that the cask and fuel assemblies will confine the

radioactive materials following the design basis events established for

dry storage casks. These design features include the confinement

function provided by the welded MSB, the cooling and shielding

functions provided by the ventilated concrete cask (VCC), the

limitations on the fuel to be stored, and other cask characteristics

and limitations placed on its use that were relied upon during the

NRC's certification of the cask. Although the NRC staff considered it

prudent to require a cask to be returned to the spent fuel pool to

ensure cooling of the spent fuel and support inspections to confirm

that the cask could remain in service following certain design basis

events, the ability of the VSC-24 casks to withstand such events made

it unnecessary for the NRC to include specific time constraints in

which the operation needed to be completed.3

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\3\ The position that a time-urgent unloading of a cask need not

be considered is also supported by the analysis of a hypothetical

event involving the failure of the stored fuel pins with subsequent

ground level breach of an MSB that was presented in the SAR for the

VSC-24 design. Although no identified accident results in such

failures, the event was analyzed to demonstrate the limited

radiological consequences from accidents involving VSC-24 casks.

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In the event that a condition would arise requiring a cask to be

returned to the spent fuel pool, the continued confinement of the

radioactive materials within the MSB would afford the licensee ample

time to develop corrective actions that would maintain safe storage

conditions and minimize occupational exposures. The design features of

the cask, the unlikely nature of events that may require unloading a

cask, and the NRC staff's judgment that licensees could develop an

alternate approach if a spent fuel pool could not support an immediate

unloading of a cask have previously been cited as reasonable

justification for not requiring licensees to maintain a fixed reserve

capacity in spent fuel pools.4

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\4\ See resolution of public comments published with rulemakings

to add the VSC-24 cask (58 FR 17948) and TN-24 cask (58 FR 51762) to

the list of NRC-certified casks.

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Requirements defining conditions for returning a cask to the spent

fuel pool were included in the COC for the VSC-24 cask in order to

maintain the cask components and stored spent fuel assemblies within

the boundaries evaluated and accepted by the NRC staff during the

certification process. The COC addresses those events or conditions

which might lead to degradation of the cask or fuel assemblies. The

required actions normally include restoring operations to within the

acceptable limits or otherwise ensuring the spent fuel is placed in a

safe storage condition. The COC requirements for some events or

conditions include returning the MSB to the spent fuel pool to provide

a safe storage condition and unloading of the spent fuel assemblies in

order to support inspections of the cask.

The COC-required action in Section 1.2.10, ``Time Limit for

Draining the MSB,'' states that a cask should be returned to the spent

fuel pool for cooling if the water cannot be drained within the

specified time after the MSB is removed from the spent fuel pool with

24 spent fuel assemblies. The referenced draining operation is part of

the cask-loading sequence and it is reasonable to assume, therefore,

that the cask-loading area within or adjacent to the spent fuel pool

would be available for the cask should this contingency need to be

implemented. Further, the COC-required action is meant to restore

cooling to maintain safety margins pertaining to fuel assembly

subcriticality and can be accomplished without unloading the fuel

assemblies from the MSB. It is likely, however, that the locations in

the spent fuel pool that had contained the fuel assemblies loaded into

the storage cask would remain available during the loading and draining

of the cask.

Section 1.2.15, ``Handling Height,'' requires fuel assemblies to be

returned to the spent fuel pool, and inspections and evaluations

performed for cask components in the event a loaded cask is dropped

from a height greater than 18 inches. The COC prohibits handling of a

loaded VCC at a height greater than 80 inches. The NRC evaluation of

the MSB drop analysis concurred that drops up to 80 inches of the MSB

inside the VCC can be sustained without breaching the confinement

boundary, preventing removal of the spent fuel assemblies, or causing a

criticality accident. However, it is deemed prudent to return the cask

to the spent fuel pool to perform inspections and evaluations in the

event a cask experiences a significant drop, which is considered to be

a drop from a height greater than 18 inches. The requirement to perform

such inspections and evaluations was, therefore, included in the COC in

the event that a cask were to be dropped during movement. However,

since the most likely time for a cask drop event to occur would be

during movement of a newly loaded cask to the ISFSI, it is reasonable

to assume that the spaces in the spent fuel pool that had contained the

fuel assemblies loaded into the cask would remain available. Moreover,

even assuming for the sake of this analysis that the drop occurs when

spaces might not be available in the spent fuel pool, reviews of the

cask have shown that the cask and fuel will remain intact following a

drop from the maximum allowable height. Because a drop from the maximum

allowable height would not pose an immediate threat to the safety of

the public or plant personnel, adequate time would be available for the

licensee to develop and implement approaches to perform the required

inspections and evaluations if spaces were not available in the spent

fuel pool to support an immediate unloading of the cask. Temporary

shielding, loading the affected MSB into a spare VCC, placing the

affected MSB into the cask loading area within or adjacent to the spent

fuel pool, or other contingency actions could ensure safe storage

conditions while the licensee developed and implemented an approach to

allow for the actual unloading of the cask that had been dropped.

The requirements contained in Sections 1.2.3, ``Maximum Permissible

Air Outlet Temperature,'' and 1.3.4, ``Thermal Performance,'' were

included in the COC to provide reasonable assurance that the

temperatures of the fuel cladding and the VSC-24 concrete do not exceed

design limits. Concrete temperature limits are intended to prevent

gradual degradation of the VCC and the shielding it provides for the

MSB, which is the containment vessel for the spent fuel. Other

temperature limits pertain to the fuel cladding and are intended to

maintain the stored fuel assemblies below the temperatures at which

damage might occur. However, in the event that excessive temperatures

are detected, cooling of the cask and subsequent placement of the MSB

into the spent fuel pool, if necessary, are sufficient to avoid

immediate safety concerns. Because safe storage of the fuel assemblies

is achieved by placing the affected MSB into the cask loading area

adjacent to or within the spent fuel pool, the actual unloading of the

assemblies from the MSB to the storage racks within the spent fuel pool

can await the licensee's development of alternative approaches if that

were necessary due to a lack of storage space in the spent fuel pool.

Such approaches may require the licensee to make modifications to the

spent fuel pool or other parts of the reactor facility.

In addition to the specific COC requirements previously discussed,

a cask might need to be returned to the spent fuel pool if the cask

fails to meet some criteria provided in NRC regulations or the COC and

should, therefore, be removed from service.

[[Page 19823]]

Tests and surveillances performed before and after loading spent fuel

into a storage cask are designed to detect failures to conform to

design or regulatory requirements before a problem presents an imminent

threat to the cask or stored fuel. Therefore, while discovery of a

nonconformance or previously unidentified vulnerability may require

removing a cask from service as part of a licensee's corrective

actions, it is highly improbable that the discovery of such a condition

would pose an immediate safety concern. As in the previous examples,

safe storage of the spent fuel could be accomplished by returning the

affected MSB to the cask loading area within or adjacent to the spent

fuel pool and the MSB and spent fuel could remain there while the

licensee determined an appropriate course of action, including

provisions for unloading the cask, if necessary.

In sum, no credible accident has been identified that would require

the immediate unloading of a storage cask as a necessary protective

measure to avoid significant radiological consequences to members of

the public. In addition, there is no event or condition that was

identified during the certification of the VSC-24 cask that would

require a time-urgent unloading of a cask. Therefore, there is no need

for NRC to require continuous availability of space in the spent fuel

pool to accommodate the potential need to unload a cask. Further, the

NRC staff has reasonable assurance that licensees could, if necessary,

develop and implement an approach to unload a cask if required to do so

by unplanned events or conditions, such as those identified in the COC.

If space is not immediately available in the spent fuel pool, there

would be time to make it available by relocating other spent fuel

assemblies or removing them for temporary storage in a cask or by

making modifications to the spent fuel pool or other parts of the

reactor facility. Therefore, the NRC does not see a need to require the

licensee to reserve a fixed number of vacant spaces in the spent fuel

pool or to maintain the capability to retrieve the spent fuel from a

cask within a specified period of time, particularly when there is no

such prescriptive requirement stated in NRC rules.

Item 2: Order VSC-24 Users Not To Load Casks Pending Amendment of

Documents

The Petitioner's second request was for the NRC to order all users

of the VSC-24 cask not to load VSC-24 casks until the COC, the SAR, and

the SER are amended to contain operating controls and limits that

prevent hazardous conditions. As noted previously, following the event

at Point Beach, the NRC staff recognized that additional evaluation of

potential material interactions was warranted for all transportation

and storage casks. In regard to the VSC-24 cask, 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-24 casks pending completion of the staff's review of the responses

to NRC 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-24 casks.

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 the 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 staff has documented the

process, information, and results of its review of the licensee's

response to Bulletin 96-04 for use of the VSC-24 at ANO and Point Beach

in safety evaluations available for public review. The NRC staff is

currently reviewing the responses to the bulletin submitted by the

licensee for Palisades.

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 cask, the NRC staff agrees with the

Petitioner that it would be beneficial if the SAR and other licensing

basis documents accurately described the identified chemical reaction

and the associated operating controls and limits. The NRC staff is

currently reviewing a proposed amendment to the SAR and the COC for the

VSC-24 cask 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 document. In

addition, the NRC staff is processing a petition for rulemaking, PRM-

72-3, that may lead to additional updating of ISFSI 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 NRC 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

Petitioner as a precondition for resuming cask operations at the

facilities using VSC-24 casks.

IV. Conclusion

The Petitioner requested that the NRC (1) Require WEPCO to retain

24 empty and available spaces in the Point Beach Nuclear Plant spent

fuel pool to accommodate retrieval of spent fuel from a VSC-24 cask,

and (2) prohibit loading of VSC-24 casks until the COC, the SAR, and

the SER are amended to contain operating controls and limits to prevent

hazardous conditions. Each of the claims by the Petitioner has been

reviewed. I conclude that for the reasons discussed above, no adequate

basis exists for granting the Petitioner's request for either (1)

Requiring the licensee for Point Beach to reserve a fixed number of

vacant spaces in the spent fuel pool or (2) suspension of the

licensees' use of the general license for dry cask storage of spent

nuclear fuel at Palisades, Point Beach, or Arkansas Nuclear One pending

revision of the SAR, the SER, and the COC for the VSC-24 cask.

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 17th day of April 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor Regulation.

[FR Doc. 97-10522 Filed 4-22-97; 8:45 am]

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