**NRC INSPECTION MANUAL** NMSS/SFST

INSPECTION PROCEDURE 60856

REVIEW OF 10 CFR 72.212(b) EVALUATIONS

PROGRAM APPLICABILITY: 2690

FUNCTIONAL AREA: ENGINEERING (ENG)

60856-01 INSPECTION OBJECTIVE

Section 72.210 of Title 10 of the Code of Federal Regulations (10 CFR
72.210) grants a general license for the storage of spent fuel in an
independent spent fuel storage installation (ISFSI) at power reactor
sites to any person authorized to possess or operate nuclear power
reactors under 10 CFR Part 50. Section 72.210 gives the conditions for
this general license and 72.212(b) delineates requirements that the
general licensee shall meet. This inspection procedure (IP) provides
guidance for determining whether a general licensee has met these
requirements before operating its ISFSI.

60856-02 INSPECTION REQUIREMENTS

02.01 [Review of Licensee Evaluations]{.underline}. \[72.212(b)(2)\][^1]

Determine whether:

a\. The licensee performed written evaluations which established that
the conditions set forth in the Certificate of Compliance (CoC) have
been met.

b\. The licensee performed written evaluations which established that
the cask storage pads and areas have been designed to support the stored
static load of the storage casks.

c\. The licensee performed written evaluations which established that
the requirements of 10 CFR 72.104 regarding effluents and direct
radiation from an ISFSI have been met.

d\. Provisions are in place for retaining the evaluations (records)
referenced above until spent fuel is no longer stored under the general
license.

e\. The licensee performed any revisions to the written evaluations
required by 10 CFR 72.212 in accordance with 10 CFR 72.48(c).

02.02 [Review of Site Characteristics Against Safety Analysis Report
(SAR) and Safety Evaluation Report (SER)]{.underline} \[72.212(b)(3)\]

a\. Verify that the licensee reviewed the SAR referenced in the cask CoC
and associated U.S. Nuclear Regulatory Commission (NRC) SER and
determined that the cask design bases used in these reports are
enveloped by the reactor site parameters.

b\. Verify that the results of the above reviews are documented and
included with the evaluations retained in accordance with 02.01.d.

02.03 [Review of ISFSI Activities for Determination of No Adverse Impact
on Site Operations or Technical Specifications]{.underline}
[(TS)]{.underline}. \[72.212(b)(4)\]

a\. Determine if the licensee evaluated whether activities related to
storage of spent fuel under the general license involved a change in the
facility TS or required a license amendment for the facility in
accordance with 10 CFR 50.59(c).

b\. Verify that the results of the above reviews are documented and
included with the evaluations retained in accordance with 02.01.d.

02.04 [Safeguards Program Review]{.underline}. Review applicable parts
of the licensee=s safeguards program and security plan to verify that
spent fuel is protected against the design basis threat of radiological
sabotage in accordance with the requirements of 10 CFR 73.55, with the
conditions given in 10 CFR 72.212(b)(5)(i) through (v).

02.05 [Review of Programs Impacted by ISFSI Operation]{.underline}.
\[72.212(b)(6)\]

Verify that the licensee reviewed the following programs to determine if
their effectiveness is decreased. If so, determine whether the necessary
changes were made and if necessary approvals (internal or external) were
sought and obtained.

a\. Emergency plan;

b\. Quality assurance program (QAP);

c\. Radiation protection program; and

d\. Training program.

02.06 [ISFSI Procedures]{.underline}. \[72.212(b)(9)\]

Verify that activities related to storage of spent fuel under the
general license will be performed only in accordance with licensee
written procedures.

02.07 [Storage of ISFSI Records]{.underline}. \[72.212(b)(7) and (8)\]

Verify that:

a\. The licensee maintains current copies of the CoC and any documents
referenced in the CoC for each cask model used at the ISFSI.

b\. The licensee maintains the records provided by the cask supplier for
each cask design used, and provisions are in place for transferring
these records if a cask is sold, leased, loaned, or otherwise
transferred to another user.

60856-03 INSPECTION GUIDANCE[^2]

03.01 [General Guidance]{.underline}.

a\. Questions on ISFSI activities potentially affecting safety-related
reactor systems, structures or components (SSCs) and all other
ISFSI-related technical questions should be referred to the cognizant
Spent Fuel Storage and Transportation (SFST) project manager (PM). SFST
may request Nuclear Reactor Regulation (NRR) assistance if needed.

b\. In preparation for inspecting the dry cask storage system (DCSS)
used by the licensee, review the commitments and key requirements in the
SAR, the NRC's SER, and the CoC. SARs and SERs have been written for
each type of approved DCSS and describe operation of DCSS components.
Information about commitments for a particular DCSS may also be found in
the associated CoC. In some cases, the CoC may have an attached
Conditions of Use/Technical Specifications document and design bases.
DCSS designs vary and care must be taken to review the correct
documentation. Copies of these documents may be obtained from the region
or the cognizant SFST PM. (Note: Although the SER can document or
clarify commitments made by the licensee or vendor, the SER is not an
independent basis for enforcement actions.)

c\. Because of the large amount of design information to be reviewed in
carrying out this IP, a pre-inspection site visit may be needed.

03.02 [Specific Guidance]{.underline}.

a\. [Inspection Requirement 02.01]{.underline}.

1\. In reviewing CoC conditions, the inspector should:

\(a\) Review the CoC for the DCSS to be used by the licensee and
identify any conditions that should be considered in the licensee's
evaluations. The Conditions section of the CoC may refer to an attached
Condition for Cask Use/TS or similar document. This attachment will
likely contain regulatory requirements such as site-specific parameters
and analyses that the user must verify. It may also contain requirements
for additional procedures, heavy load considerations, training and
training exercises, and for the "first cask in place." The TS Functional
and Operational Limits sections may also contain a number of
surveillance requirements, restrictions on the characteristics of the
spent fuel (or other contents) that may be loaded in the DCSS, and other
physical parameters that must be observed.

\(b\) Review the licensee's evaluations and verify that the conditions
identified in Section 03.02.a.1.(a) above were considered. Perform
independent technical review of selected parts of the evaluations to
help assess whether all the CoC conditions were met. Carefully examine
whether the licensee\'s analysis to ensure that conditions of the CoC
were met used SAR conditions and assumptions. For example, in a
time-to-boil analysis for a loaded cask that is being sealed, if the
licensee used a different initial spent fuel pool water temperature than
that in the SAR, the licensee should have adjusted the time-to-boil
analysis for the actual starting temperature.

\(c\) While performing this review, verify that applicable conditions
have been incorporated into procedures (operational or programmatic, as
applicable). This should be documented by the procedure review of 02.06
of this IP.

2\. In reviewing the cask storage pads and areas designed to support the
static load of the DCSS, the inspector shall use Appendix A of this IP.
As stated in footnote 2, this review shall be performed prior to
construction of the ISFSI.

3\. In reviewing whether the requirements of 10 CFR 72.104 for effluents
and direct radiation from an ISFSI have been met, the inspector should
refer to the radiological protection programs guidance provided in
Section 03.02.e.3 of this IP.

4\. A copy of the licensee=s evaluations from Sections 03.02.a(1), (2),
and (3) of this IP (reference 10 CFR 72.212 (b)(2)) must be maintained
until spent fuel is no longer stored under the general license issued
under 10 CFR 72.210. Additional guidance is in Section 03.02.g of this
IP.

5\. Guidance in this area is provided in IP 60857.

b\. [Inspection Requirement 02.02]{.underline}. Review the licensee's
evaluation that reactor site parameters are enveloped by the cask design
bases used in the DCSS SAR and SER. Verify that the evaluation of the
reactor site parameters includes, at a minimum, earthquake intensity and
tornado-generated missiles. Parameters for earthquake intensity and
tornado missiles are listed in the reactor site's Updated Safety
Analysis Report (UFSAR) or Final Safety Analysis Report (FSAR).

c\. [Inspection Requirement 02.03]{.underline}.

1\. Review the licensee's evaluation of the ISFSI's impact on the
reactor facility and identify any required license amendments or changes
to the reactor TS. Additional guidance for reviewing the reactor
licensee's 10 CFR 50.59 evaluations may be found in IP71111.02. If the
tasks required by Subsection 03.02.c.5 through .8 have been previously
performed under IP 60853, they do not need to be repeated.

2\. Verify that any 10 CFR 50.59 evaluations have been performed in
accordance with the licensee's administrative requirements. The licensee
should have evaluated the impact of the ISFSI or DCSS on the reactor
facility and document these evaluations in accordance with the site
administrative procedures and 10 CFR 50.59.

3\. Verify that dropping a storage or fuel transfer cask inside the fuel
handling or reactor building will not challenge any safety-related
facility SSCs. Verify that storage and transfer cask pathways were
approved in accordance with the licensee's heavy-load administrative
requirements. The licensee's analysis should indicate whether the weight
of the DCSS being used is bounded by the cask weight used in the UFSAR
dropped cask analysis and has been evaluated against the NUREG-0612
requirements for controlling heavy loads. Otherwise, the licensee should
have fully analyzed the impact of a dropped DCSS on the reactor
facility.

Review the assumptions used by the licensee in any cask drop analysis
and ensure that these assumptions are consistent with the NRC cask SER.
For example, suppose a licensee has taken credit for a cask meeting the
hypothetical accident conditions specified in 10 CFR 71.73 for a 30-foot
drop onto a flat surface. For this accident, the NRC SER assumed that
the cask closure bolts were all installed and tensioned, valve box
covers were installed, and any impact limiters were installed. However,
these conditions might not be met for a cask while it is being moved
inside the reactor or fuel handling building. Therefore, the licensee
may not take credit for the radiological consequences analyzed for the
10 CFR 71.73 event.

4\. Verify that the fuel handling or reactor building cranes can safely
move the storage or fuel transfer casks (reference NUREG 0612). Verify
that any changes to the interlocks and mechanical stops of the fuel
handling or reactor building cranes have been evaluated and properly
implemented. Be aware that the interlocks or mechanical stops may be
controlled by the facility TS.

5\. Verify, by reviewing records and interviewing licensee personnel,
that the licensee has considered the maximum expected load and any
interferences in evaluating DCSS movement along the proposed roadway and
any staging areas between the fuel or reactor buildings and the ISFSI
storage site. The licensee should have reviewed the pathway for moving
the DCSS to the ISFSI site and determined the right-of-way needed for
such movements. The licensee should have also considered the effects of
weathering and repeated use on the roadway. The inspection should
examine any interferences from overhead lines or nearby structures.
Contact with overhead power lines may impact the safety of the reactor
plant and endanger the cask transporter.

6\. Verify that an analysis indicates that conduit, piping, and
equipment beneath the proposed roadway, ISFSI storage pad, and staging
areas can withstand the heavy loads of the storage casks or that
modifications have been designed and implemented to reinforce the area
and protect any buried systems. The licensee should have determined
whether reinforcements are needed for places where the roadway passes
over or near buried components (e.g., diesel fuel oil transfer lines,
fire mains, service water lines, and electrical conduits). Inspection of
any construction activities to reinforce buried components should be
performed under IP 60853.

7\. Determine, by observation and interviews with licensee personnel,
that the proposed roadway is suitable for the secure movement of spent
fuel storage casks and that a cask temporarily halted during transport
will not pose a traffic obstruction that adversely impacts facility
activities or security. The licensee should have assessed the potential
impact of a breakdown of the vehicle transporting the cask on reactor
site traffic and security activities.

8\. Verify that the transfer route and the storage pad area do not
contain fire hazards beyond those analyzed in the SAR. Movement of a
cask through or by vehicle parking lots may expose the cask to more
flammable material than assumed in the SAR fire hazard analysis, a
possibility that should be reflected in the licensee's analysis.

d\. [Inspection Requirement 02.04]{.underline}. Verify that the
licensee=s security program reflects the requirements of 10 CFR
72.212(b)(5) for protecting spent fuel against radiological sabotage.
Additional guidance for reviewing the reactor licensee's physical
security plan (PSP) may be found in IP 81001.

1\. Review the licensee's evaluation of the ISFSI's impact on the
reactor facility's PSP. Verify that changes have not decreased the
plan's effectiveness. If any changes have decreased the plan's
effectiveness, verify that NRC approval of the changes was obtained.

2\. Verify that any ISFSI-specific requirements have been adequately
addressed in the facility's PSP. Verify that changes have been
appropriately incorporated into the plan's implementing procedures and
that personnel have been trained regarding these changes.

e\. [Inspection Requirement.02.05]{.underline}.

1\. Review the licensee's evaluation of ISFSI's impact on the reactor
emergency plan (EP). Additional guidance for reviewing the EP can be
found in IP 82701.

\(a\) Verify that changes have not decreased the plan's effectiveness.
If any changes decreased the plan's effectiveness, verify that NRC
approval of the changes was obtained.

\(b\) Verify that any ISFSI-specific requirements have been adequately
addressed in the EP. Verify that changes have been incorporated into the
EP implementing procedures and that personnel have been trained
regarding these changes.

2\. Review the licensee's evaluation of ISFSI impact on the reactor QAP.

\(a\) Verify that changes have not decreased the QAP=s effectiveness. If
any changes decreased the QAP's effectiveness, verify that NRC approval
of the changes was obtained.

\(b\) Verify that any ISFSI-specific requirements have been adequately
addressed in the QAP. Verify that changes have been incorporated into
the QAP's implementing procedures and that personnel have been trained
regarding these changes.

3\. Review the licensee=s evaluation of the ISFSI=s impact on the
reactor radiological protection program (RPP). Additional guidance for
reviewing the RPP can be found in IPs 83750 and 83729.

\(a\) Verify that changes have not decreased the program's
effectiveness. If any changes have decreased the program's
effectiveness, verify that the necessary approvals (internal or
external) were obtained. Changes to an RPP generally do not require NRC
approval.

\(b\) Verify that ISFSI-specific requirements have been adequately
addressed in the RPP. Verify that changes have been incorporated into
the program's implementing procedures and personnel trained regarding
these changes.

\(c\) Review the licensee's evaluation of the ISFSI=s impact on the
annual dose-equivalent exposure for normal operations and anticipated
occurrences to a real individual who is beyond the controlled area
defined in 10 CFR 72.106. Verify that the limits specified in 10 CFR
72.104(a) have not been exceeded using the exposure requirements in 10
CFR 72.104(a)(1) through (3) (10 CFR 72.104(a)(3) is interpreted to mean
any radiation from the reactor facility's licensed activities;
consequently, the annual exposures should consider the sum of all these
sources).

\(d\) Assess whether the licensee's operating limits on ISFSI
radioactive effluents and direct radiation will meet annual
dose-equivalent exposure limits.

\(e\) Assess whether the licensee's operating restrictions on ISFSI
radioactive effluents and direct radiation will meet as low as
reasonably achievable (ALARA) objectives.

\(f\) Determine if the licensee has measured the ISFSI pad's background
radiation levels before storing spent fuel in the ISFSI. Although
information on storage pad background radiation levels is not required,
the licensee may want to measure radiation levels to establish the
baseline level for future ISFSI decommissioning activities.

4\. Review the licensee=s evaluation of the ISFSI=s impact on the
reactor training program. Additional guidance for reviewing the training
program is given in IP 41500.

\(a\) Verify that changes have not decreased the program's
effectiveness. If any changes have decreased the program's
effectiveness, verify that the necessary approvals (internal or
external) were obtained. Training program changes generally do not
require NRC approval.

\(b\) Verify that ISFSI-specific requirements have been adequately
addressed in the training program. Verify that changes have been
incorporated into the program's implementing procedures and personnel
trained regarding these changes.

f\. [Inspection Requirement 02.06]{.underline}. No specific guidance.

g\. [Inspection Requirement 02.07]{.underline}. The licensee should have
specified a retention time limit and a quality standard for any storage
facilities used to store the evaluations and records required to be kept
by 10 CFR 72.212(b) and 72.174. Retention time limits may vary because
some documents must be kept as long as the ISFSI exists, while others
are related to an individual loaded DCSS cask and are transferred with
the cask to any new licensee (another ISFSI or a disposal facility).
Review the licensee=s records management program for DCSS activities and
verify that:

1\. The licensee established appropriate retention limits for the
evaluations required under 10 CFR 72.212(b).

2\. The licensee established appropriate retention limits for copies of
the CoC and documents referenced in the CoC, as required by 10 CFR
72.212(b)(7).

3\. The licensee established appropriate retention limits for
information required by 10 CFR 72.212(b)(8)(i) and (ii).

4\. The licensee has provided for transferring specific records to any
new registered user, as required by 10 CFR 72.212(b)(8)(iii).

60856-04 INSPECTION RESOURCES

To prepare for these inspections each inspector should spend
approximately forty hours on in-office review. Inspection activities
will require approximately forty hours, by each of three inspectors.
Documentation is estimated to require sixteen hours per inspector.
Regional inspection staff are expected to have the lead in performing
this inspection; however, they may be augmented by SFST staff. SFST may
request NRR assistance if needed in certain technical areas such as 10
CFR 50.59 evaluations.

60856-05 REFERENCES

ANSI/N14.6-1993, "Special Lifting Devices for Shipping Containers
Weighing 10,000 Pounds (4500 kg) or More for Radioactive Materials."

NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," July 1980.
Licensee implementation of this NUREG may vary. Specific commitments to
this guidance are covered by licensee responses to NRC Generic Letters
80-113 and 85‑11.

NRC Bulletin 96-02, "Movement of Heavy Loads over Spent Fuel, Over Fuel
in the Reactor Core, or Over Safety-related Equipment," April 11, 1996.

END

Appendix

A. Review of ISFSI Storage Pad Design

APPENDIX A

REVIEW OF ISFSI STORAGE PAD DESIGN

A. PURPOSE AND OBJECTIVE

This appendix provides guidance on reviewing a general licensee=s
evaluation of its Independent Spent Fuel Storage Installation (ISFSI)
storage pad design before constructing the pad.

B. INSPECTION REQUIREMENTS AND GUIDANCE

a\. [General Guidance]{.underline}. A general licensee\'s ISFSI storage
pad is typically classified as not-important-to-safety, so generally
accepted commercial standards and practices are considered acceptable in
designing the support pad. NUREG-1567, Section 5.3.3 contains additional
guidance on construction standards for reinforced concrete structures.
If the ISFSI site characteristics require a modification of the storage
pad\'s design as described in the Safety Analysis Report (SAR), verify
the safety classification of the storage pad before performing any
inspections. Division of Spent Fuel Storage and Transportation (SFST)
may request Nuclear Reactor Regulation (NRR) assistance if needed.

b\. [Requirements and Specific Guidance]{.underline}.

1\. Find out the following from records: whether the reactor site soil
structures differ from the soil structures under the ISFSI storage pad,
the ground acceleration associated with the safe shutdown earthquake
(SSE), and the elevation at which this ground acceleration was applied
to the reactor structures.

Review this information to determine whether the licensee used
appropriate assumptions in the seismic and liquefaction analyses for the
storage pad.

2\. Assess the licensee's conclusions about the acceptability of the
storage pad's design with respect to the site's hydrology, geology and
seismology.

Hydrological data should include the effects of potential flooding and
soil erosion on the ISFSI site. Geotechnical data from soil borings and
standard penetration tests (SPTs) should include soil descriptions,
groundwater table level, effects of long-term consolidation of soil
beneath the pad, and the potential for swelling on excavation or removal
of overburden.

Geological information should include evidence that the soil borings
were to the proper depth (i.e., bedrock or very dense material, as
evidenced by high SPT blow counts), the thicknesses of soil layers below
the storage pad, and other significant properties of the soil layers
(i.e., unit weight, Poisson's ratio, low-strain shear modulus, moisture
content, fines content, relative density, and shear-wave velocity).

3\. Verify that the ISFSI site's seismic ground motion acceleration
analysis envelops the seismic ground motion acceleration specified in
the Dry Cask Storage System (DCSS) SAR design bases. Verify that the
ISFSI site's seismic ground motion acceleration has been used in the
pad's soil liquefaction analysis.

If the soil structures under the ISFSI storage pad are different from
the soil structures under the reactor site, the licensee's analysis
should consider these differences in providing an accurate analysis of
the seismic conditions under the ISFSI storage pad. Similarly, if the
seismic acceleration used in the reactor site Updated Final Safety
Analysis Report (UFSAR) was applied at the base of the seismic Category
I reactor containment building foundation, the licensee's analysis
should translate this acceleration to the free field ground surface for
the ISFSI storage pad and its foundations.

4\. Verify that the licensee's evaluation addresses the following
criteria:

a\. The pad's design supports the static load of the DCSS.

b\. The pad's design uses the total and differential settlements for
both static and seismic loadings.

c\. The pad's design uses ultimate and allowable soil-bearing
capacities, including appropriate safety factors, that are less than the
actual soil-bearing capacity.

d\. The pad's design uses the sequential, partial, and total loads that
the pad may be subjected to.

e\. The pad's design uses the soil liquefaction analysis to evaluate the
stability of any slopes adjacent to the ISFSI. Adjacent slopes are
stable or any potential soil movement during a seismic event will not
adversely affect the safety of the ISFSI.

The licensee=s analysis should show that the ISFSI storage pad will
adequately support both static and dynamic loads, as required by 10 CFR
72.212(b)(2)(ii) and 72.212(b)(3), respectively. This analysis should
include the methods used to calculate the total and differential
settlements of the pad, the ultimate and allowable soil bearing
capabilities, and the stresses and strains resulting from the
sequential, partial, and total loads the pad could experience.

These calculational methods should be able to realistically model both
static and dynamic soil-structure interaction (SSI) phenomena. The
analysis should recognize the uncertainties in SSI phenomena, including:
(1) the soil and rock configuration, material characteristics, and lack
of symmetry in the soil deposits; (2) modeling of soil constituents and
soil properties; (3) the effects of a varying water table over time; (4)
the effects of soil swelling and corresponding reversed tensile and
compressive stresses on the storage pad; and (5) the effects of partial
separation of, or loss of contact between, the storage pad and the soil
during sequential loading of DCSSs or seismic events.

If the modeling method used a finite element analysis, the criteria for
determining the location of the side and bottom boundaries should be
assessed. Additional guidance on soil-structure interaction analysis may
be found in the references. Questions on the adequacy of the licensee's
ISFSI storage pad analysis should be referred to the cognizant SFST
project manager (PM). SFST may request NRR assistance if needed.

C. REFERENCES

NRC Regulatory Guide 1.132, "Site Investigations for Foundations of
Nuclear Power Plants," March 1979.

NRC Regulatory Guide 1.138, "Laboratory Investigations of Soils for
Engineering Analysis and Design of Nuclear Power Plants," April 1978.

NRC Information Notice 95-28, "Emplacement of Support Pads for Spent
Fuel Storage Installations at Reactor Sites," June 5, 1995.

END

ATTACHMENT 1

[Revision History for IP 60856]{.underline}

+--------+--------+------------------------+-----+-------+-------------+
| Comm   | Issue  | Description of Change  | Tr  | Tra   | Comment     |
| itment | Date   |                        | ain | ining | Resolution  |
| Tr     |        |                        | ing | Compl | Accession   |
| acking |        |                        | Nee | etion | Number      |
| Number |        |                        | ded | Date  |             |
+--------+--------+------------------------+-----+-------+-------------+
| N/A    | 03     | This document has been | N/A | N/A   | N/A         |
|        | /05/08 | revised to change SFPO |     |       |             |
|        |        | to SFST and some other |     |       |             |
|        | CN     | minor editorial        |     |       |             |
|        | 08-010 | changes. No other      |     |       |             |
|        |        | changes are proposed   |     |       |             |
|        |        | by 10/24/2007.         |     |       |             |
+--------+--------+------------------------+-----+-------+-------------+
|        |        |                        |     |       |             |
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[^1]: Numbers in brackets refer to the applicable Section of 10 CFR Part
    72, Subpart K.

[^2]: Inspection Manual Chapter 2690, Appendix A, requires that review
    of the licensee=s 10 CFR 72.212(b) evaluations of the ISFSI support
    pad be completed before the licensee begins constructing the pad.
    Appendix A of this IP should be used in meeting inspection objective
    02.01.b. The remainder of this IP should be performed before the
    ISFSI becomes operational. IP 60853 will be used for inspection of
    the ISFSI during construction. If after completion of this IP the
    licensee decides to use a different model or type of DCSS,
    applicable parts of this IP should be revisited.