ML011860407
| ML011860407 | |
| Person / Time | |
|---|---|
| Issue date: | 10/18/2001 |
| From: | Matthews D Division of Regulatory Improvement Programs |
| To: | |
| Fields E N 301-415-1173 | |
| References | |
| -nr RIS-01-019 | |
| Download: ML011860407 (13) | |
See also: RIS 2001-19
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
October 18, 2001
NRC REGULATORY ISSUE SUMMARY 2001-19: DEFICIENCIES IN
THE DOCUMENTATION OF DESIGN BASIS RADIOLOGICAL
ANALYSES SUBMITTED IN CONJUNCTION WITH LICENSE
AMENDMENT REQUESTS
ADDRESSEES
All holders of operating licenses for power reactors.
INTENT
The U.S. Nuclear Regulatory Commission (NRC) is issuing this regulatory issue summary (RIS)
to inform addressees of inadequacies in licensees documentation of design basis accident
(DBA) radiological analyses in license amendment submittals. It is expected that recipients will
review the information for applicability to their facilities and consider actions, as appropriate.
However, suggestions contained in this RIS are not NRC requirements; therefore, no specific
action or written response is required.
BACKGROUND INFORMATION
Under Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR 50.59), a licensee
may make changes to a nuclear facility without prior NRC approval. Changes made under
10 CFR 50.59 must meet certain criteria and must not involve a revision to a technical
specification. Revisions to technical specifications and proposed changes that do not meet the
criteria of 10 CFR 50.59 are submitted under 10 CFR 50.90 for NRC approval. Under 10 CFR 50.90, a licensee is required to fully describe the changes desired and to follow, as far as
applicable, the format prescribed for original applications. The evaluation of postulated
radiological consequences often constitutes a significant portion of the safety analyses
performed in support of the proposed license amendment. In reviewing these submittals, the
NRC staff considers the licensees description of the analyses performed, the assumptions and
inputs, the methodology, and the results obtained. The NRC staff often finds that licensees
submit insufficient information for an adequate review. Also, in some cases, the NRC staff has
identified deficiencies in analysis assumptions, inputs, and methods that had to be resolved
before the amendment was approved. The purpose of this RIS is to discuss the more frequent
and more significant deficiencies observed by the NRC staff.
Page 2 of 7
SUMMARY OF ISSUE
The DBAs were structured to provide a conservative set of assumptions to test the performance
of one or more aspects of the facility design. Many physical processes and phenomena are
represented by conservative, bounding assumptions rather than by being directly modeled.
The staff has selected assumptions and models that, when used in combination, form a basis
for evaluating the facility design to ensure an appropriate and prudent safety margin against
unpredicted events in the course of an accident and to compensate for uncertainties in plant
parameters, accident progression, human performance, radioactive material transport, and
atmospheric dispersion.
1.
Facility Design Basis
Radiological consequence analyses performed in support of license amendment
requests should use analysis assumptions, inputs, and methods that are consistent with
the current facility design basis and with current facility normal and emergency operating
procedures. Licensees may take analysis credit for plant features that were included in
design basis radiological consequence calculations previously approved by the NRC
staff. Such credit should be taken only if assumptions related to equipment operability
and performance are consistent with the facilitys current design basis and current
normal and emergency operating procedures. The NRC staff generally does not accept
analyses that credit plant features that (a) are not safety-related, (b) are not covered by
technical specifications, (c) do not meet single-failure criteria, or (d) rely on the
availability of offsite power unless the assumptions were previously accepted by the
NRC in a site-specific licensing action and are therefore part of the facility design basis.
Design basis delays in actuation of these features should be considered, especially for
those features that rely on manual operator intervention.
Generally, the NRC staff will consider an assumption made in a licensee analysis,
supporting a docketed amendment request, to be part of the current design basis if the
staff relied upon that assumption when evaluating whether NRC requirements were met
in granting the license amendment.
2.
Level of Detail in Submittals
The NRC staff reviews licensee amendment requests to ensure that the proposed
change will maintain an adequate level of protection of public health and safety. The
NRC staff accomplishes these reviews by evaluating the information submitted in the
amendment request against the current plant design basis as documented in the Final
Safety Analysis Report (FSAR), previously issued staff safety evaluation reports,
regulatory guidance, other licensee commitments, and staff experience gained in
considering similar requests for other plants. The NRC staff bases its finding on the
acceptability of an amendment on its assessment of the licensees analysis, since it is
the licensees analysis that becomes part of the facilitys design basis. Licensees should
ensure that adequate information, including analysis assumptions, inputs, and methods
is presented in the submittal to support a staff assessment. The NRC staffs
assessment may include performance of independent analyses to confirm the licensees
conclusions. Licensees should expect an NRC staff effort aimed at resolving critical
differences between analysis assumptions, inputs, and methods used by the licensee
and those deemed acceptable to the NRC staff.
Page 3 of 7
Regulatory Guide (RG) 1.70 (Ref. 1) offers guidance on information to be included in
accident analysis descriptions in FSARs, and may be useful in determining the minimum
information that should be submitted in support of a license amendment. Additional
information may be needed, depending on the particular analysis. Licensees may want
to consider submitting the affected FSAR pages annotated to reflect the revised
analyses and or the actual calculation documentation, in addition to the analysis
summary. Licensees who submit electronic FSARS, i.e. CDs may wish to consider
submitting any updates electronically and also provide a list of affected FSAR pages.
3.
Analysis Inputs
Analysis inputs should be the most restrictive values of plant parameters selected from
the range of design values possible during the specific event so that the postulated
consequences of the event are maximized. It is generally inappropriate to use values
characterized as best estimates. Licensee commitments to particular regulatory
guides and standard review plan sections may establish the value of certain parameters
and should continue to be used where applicable. Other considerations follow:
a.
The range of values applicable during an accident may vary from accident to
accident, and will likely differ from the range that applies during normal
operations. For example, a loss-of-offsite-power assumption may affect
ventilation system flow rates.
b.
It may be necessary to use different parameter values in different portions of the
analyses or to perform a sensitivity analysis to determine the limiting value. In
some situations the minimum and maximum value of the range may be
applicable in a single analysis. For example, the minimum containment spray
flow rate is used in determining the spray removal coefficients, but the maximum
flow rate may be appropriate in determining the minimum sump pH.
c.
If a plant parameter is associated with a technical specification limiting condition
for operation (LCO), the value specified in the technical specification should be
used. If the LCO specifies a range, or a value with a tolerance band, the most
restrictive value should be used. The technical specifications may also specify
numeric values in surveillance requirements or action statements; for example,
acceptable emergency core cooling system leakage or transient reactor coolant
system (RCS) iodine concentration. These should be used where appropriate.
d.
Some parameters may change value during the accident; for example, RCS
temperature and pressure decrease during plant cooldown. In these cases, the
calculation should either assume the most restrictive value for the entire duration
or the calculation should be performed in time steps, with the appropriate
parameter values used for each time step. Containment leakage should be
modeled as described in RG 1.3 and 1.4.
e.
For parameters based on the results of less frequent surveillance testing, for
example, nondestructive testing (NDT) of steam generator tubes or efficiency
testing of charcoal filters, the degradation that may occur between periodic tests
should be considered in establishing the analysis value.
Page 4 of 7
f.
Some analysis parameters can be affected by density changes that occur in the
process stream. The NRC staff has noted errors made in converting volumes
and volumetric flow rates ( for example, gpm) to mass units, or vice versa,
particularly in analyses involving primary-to-secondary leakage (Ref. 2).
Licensees may wish to avoid using volumetric units to the extent possible in
these calculations. With regard to the volumetric flow rates specified as LCOs,
the density used should be consistent with the density that is assumed in the
surveillance procedure that demonstrates compliance with the LCO. These
procedures are typically based on cooled water, not on water at RCS operating
temperature and pressure. Similarly, for those pressurized-water reactors
(PWRs) using alternate repair criteria (ARC), the tube burst flow rate correlations
are typically based on measurements of cooled water.
4.
Use of Incompatible Assumptions
Licensees should ensure that their analyses do not use assumptions that are
incompatible with the accident conditions or with other assumptions. For example:
a.
RG 1.3 (Ref. 3) and RG 1.4 (Ref. 4) state that 50 percent of the iodine activity
released from the core during a loss-of-coolant accident (LOCA) can be
assumed to instantaneously plate out on containment surfaces, leaving 25
percent of the core inventory in the containment atmosphere available for
release. Later revisions of the Standard Review Plan (SRP) (Ref. 5) Section
6.5.2 identify a mechanistic treatment of plateout that can be included in the
determination of the containment spray coefficients. It would not be appropriate
to assume 50 percent instantaneous plateout and to incorporate mechanistic
treatment plateout in the same calculation, because this would constitute double
credit of iodine plateout.
b.
RG 1.25 (Ref. 6) contains a footnote that the assumptions in the guide are
acceptable for use if certain fuel parameters, including the amount of burnup, are
not exceeded. However, some extended burnup fuel designs may exceed these
parameters. NUREG/CR-5009 (Ref. 7) considers the impact of extended burnup
fuel and suggests revised isotopic gap fractions for use in fuel handling
accidents. Licensees should justify the use of RG 1.25 or propose alternatives if
the fuel parameters specified in RG 1.25 are exceeded.
5.
Analysis Source Terms
The source terms used in accident analyses should be consistent with the guidance in
applicable RGs and SRPs. Several source terms are tabulated in typical FSARs, each
intended for specific purposes. Licensees should ensure the proper source terms are
used. For analyses performed in support of license amendment requests, the assumed
core inventory data should be appropriate for the currently licensed reactor power, fuel
enrichment, and fuel burnup. Reactor coolant activity should be based on the technical
specification specific activity LCO, including the specified transient specific activity.
Page 5 of 7
6.
Atmospheric Dispersion Values
The NRC guidance on short-term atmospheric dispersion values (/Q) has changed over
time. Many of the early plants were licensed on the basis of analyses that incorporated
the conservative and simplistic dispersion methods described in RG 1.3 and RG 1.4.
Most control room /Qs were based on the guidance of Murphy and Campe (Ref. 8), but
other methods have been used. Later plants may have used the guidance in RG 1.145
(Ref. 9) for determining offsite /Qs. The NRC staff is currently evaluating whether the
ARCON96 (Ref. 10) methodology may be used to determine control room /Q.
Licensees should use /Q values previously approved by the NRC staff and documented
in the FSAR. If the licensee chooses to revise the /Q values using a methodology
different from that accepted by the NRC staff and documented in the FSAR, the
amendment submittal should identify this change in methodology and present sufficient
information for the staff to make a determination regarding the acceptability of the
revised values. Meteorological data used in the offsite and control room assessments
should meet the guidance of Regulatory Position C.1.1 of RG 1.145.
7.
Many amendments submitted for NRC staff review address changes in the offsite dose
consequences, but fail to address the impact of the increased releases on control room
habitability. In approving the amendment, the NRC staff is required, under 10 CFR 50.92, to make a finding that the radiological consequences of the proposed
amendment, if implemented, would comply with 10 CFR Part 100 and with 10 CFR Part 50 (Appendix A, General Design Criterion 19 (GDC 19)). Some believe that the
LOCA dose consequences will be limiting for the control room because of the magnitude
of the source term relative to the source term for other accidents. The NRC staff has
identified several cases in which the LOCA was not the limiting accident for control room
habitability. The following considerations should be evaluated in performing control
room habitability analyses:
a.
The control room design is often optimized for the DBA LOCA, and the protection
afforded for other accident sequences may not be as advantageous. For
example, in most designs, control room isolation is actuated by engineered
safety feature (ESF) signals such as containment high pressure or safety
injection (SI), or radiation monitors, or both. For accidents that rely on radiation
monitor actuation, there may be a time delay in isolation that would not occur for
the immediate SI signal that would result from a LOCA. In such cases,
contaminated air would enter the control room for a longer period preceding
isolation than it would for a LOCA.
b.
The configuration of radiation monitors has an impact on their sensitivity. Ideally,
the radiation monitors would be located outside in air ventilation intake ductwork.
However, there are system designs that place the radiation monitor in
recirculation ductwork or downstream of filters. There are also designs that use
area radiation monitors. In these latter designs, the contaminated air continues
to build up in the control room volume until the concentration is large enough to
actuate the radiation monitor.
1 Although TEDE subsumes both the whole body dose and the thyroid dose, the rule language in GDC-19 and 10 CFR 100.11 specifically identifies whole-body and thyroid doses. The staff is considering changes to GDC-19 to
replace the current dose criterion with one based on TEDE. There are no current plans to revise the §100.11
guidelines due to the synergy that exists between the TID-14844 source terms and the accident dose guidelines.
For further information, see the discussion at 64 Federal Register 12119.
Page 6 of 7
c.
In some cases, control room radiation monitor setpoints may have been based
on external exposure concerns, for example, 2.5 mrem/hour, rather than thyroid
dose from inhalation. The airborne concentration of radioiodines will likely cause
elevated thyroid doses before reaching the concentration of all radionuclides
necessary to alarm the monitor. This condition is typically seen with accidents
that involve a high iodine-to-noble-gas ratio, such as main steam line breaks in
PWRs.
d.
The distance between the control room and the release point, and the associated
wind sectors, may be different for each postulated accident. These differences
are usually not significant with regard to offsite doses, but may be significant for
control room assessments because of the shorter distances typically involved.
The /Q for the DBA LOCA may not be applicable to other DBAs. A ground-level
release associated with a non-LOCA event may be more limiting than the
elevated release associated with LOCAs at plants with secondary containments
or enclosure buildings.
e.
Licensees should ensure that assumptions regarding control room isolation and
infiltration can be supported by appropriate test results or engineering
evaluations. Twenty percent of the licensed power reactors have performed
tracer gas tests of control room integrity. All of the tests performed identified as-
found infiltration rates greater than those assumed in the design basis
calculations.
f.
The use of personal respirators or the use of potassium iodide (KI) as a thyroid
prophylaxis should not be credited as a substitute for process controls or other
engineering controls as discussed in 10 CFR 20.1702.
8.
Dose Conversion Factors
The dose conversion factors (DCFs) used to convert release rate to doses should be
appropriate for use in acute, short-term exposure situations. Whole-body doses have
been traditionally based on semi-infinite cloud models, and thyroid doses have been
based on DCFs presented in Technical Information Document (TID)-14844 (Ref. 11)
(which are based on ICRP-2 (Ref. 12)). The NRC staff considers thyroid dose
conversion factors based on ICRP-30 (Ref. 13), such as those tabulated in Federal
Guidance Report 11 (Ref. 14), to be an acceptable change in methodology that does not
warrant prior review. Licensees using ICRP-30 DCFs in accident calculations should
consider revising the technical specification definition for dose equivalent I-131 to reflect
the DCFs used. However, total effective dose equivalent (TEDE) is not an acceptable
alternative in showing compliance with GDC-19 and Part 100 whole-body and thyroid
dose criteria.1
Page 7 of 7
For control room whole-body dose estimates, it is common to adjust the semi-infinite
cloud DCF to account for the finite size of the control room. This correction is not
applied to beta skin dose estimates since the range of beta particles in air is less than
the typical control room dimensions. It is important to note that the skin dose DCFs
presented in the recent literature (e.g., Federal Guidance Report 12 (Ref. 15)) are based
on both photon and beta emissions. Without the geometry correction, the photon dose
component will be over-estimated. If the geometry correction is included, the beta dose
component will be under-estimated. DOE/EH-0070 (Ref. 16) tabulates the beta and
photon skin dose DCFs separately. Licensees should ensure that the DCFs used are
appropriate for the intended use.
BACKFIT DISCUSSION
This RIS does not require any modification to plant structures, systems, components, or design
of facilities, or action or written response; therefore, the staff did not perform a backfit analysis
or require OMB clearance.
FEDERAL REGISTER NOTICE
A notice of opportunity for public comment was not published in the Federal Register
because this RIS is informational and pertains to staff positions that do not represent
departures form current regulatory requirements and practice.
PAPERWORK REDUCTION ACT STATEMENT
This RIS does not request any information collection.
If you have any questions about this matter, please contact one of the persons listed below or
the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
/RA Cynthia A. Carpenter for/
David B. Matthews, Director
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Contacts: Stephen F. LaVie, NRR
F. Mark Reinhart, NRR
301-415-1081
301-415-1185
E-mail: sfl@nrc.gov
E-mail: fmr@nrc.gov
Attachments:
1. List of References
2. List of Recently Issued Regulatory Issue Summaries
Page 7 of 7
For control room whole-body dose estimates, it is common to adjust the semi-infinite
cloud DCF to account for the finite size of the control room. This correction is not
applied to beta skin dose estimates since the range of beta particles in air is less than
the typical control room dimensions. It is important to note that the skin dose DCFs
presented in the recent literature (e.g., Federal Guidance Report 12 (Ref. 15)) are
based on both photon and beta emissions. Without the geometry correction, the photon
dose component will be over-estimated. If the geometry correction is included, the beta
dose component will be under-estimated. DOE/EH-0070 (Ref. 16) tabulates the beta
and photon skin dose DCFs separately. Licensees should ensure that the DCFs used
are appropriate for the intended use.
BACKFIT DISCUSSION
This RIS does not require any modification to plant structures, systems, components, or design
of facilities, or action or written response; therefore, the staff did not perform a backfit analysis
or require OMB clearance.
FEDERAL REGISTER NOTICE
A notice of opportunity for public comment was not published in the Federal Register
because this RIS is informational and pertains to staff positions that do not represent
departures form current regulatory requirements and practice.
PAPERWORK REDUCTION ACT STATEMENT
This RIS does not request any information collection.
If you have any questions about this matter, please contact one of the persons listed below or
the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
/RA Cynthia A. Carpenter for/
David B. Matthews, Director
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Contacts: Stephen F. LaVie, NRR
F. Mark Reinhart, NRR
301-415-1081
301-415-1185
E-mail: sfl@nrc.gov
E-mail: fmr@nrc.gov
Attachments:
1. List of References
2. List of Recently Issued Regulatory Issue Summaries
DISTRIBUTION:
RIS FILE
PUBLIC
SPSB R/F
DOCUMENT NAME: G:SPSB\\LaVIE\\RIS99XXX.WPD
Log No.: 99-258
- See previous concurrence
OFFICIAL RECORD COPY
ADAMS ACCESSION NUMBER # ML011860407
TEMPLATE NUMBER# NRR-052
OFFICE
SPSB
E
SC:SPSB
C:SPSB
Tech Ed
D:DLPM
REXB
C:REXB
D:DRIP
NAME
SLaVie:rmc*
FMReinhart*
RJBarrett*
PKleene*
Jzwolinski*
JTappert*
EImbro*
DMatthews
DATE
06/05/01
06/08/01
06/08/01
06/15/01
07/23/01
08/21/01
08/29/01
10/18/01
Attachment 1
Page 1 of 1
LIST OF REFERENCES
1.
USNRC, Standard Format and Content of Safety Analysis Reports for Nuclear Power
Plants (LWR Edition), RG 1.70, 1978.
2.
USNRC, Steam Generator Tube Rupture Analysis Deficiency, IN 88-13, May 25,1988.
3.
USNRC, Assumptions Used for Evaluation the Potential Radiological Consequences of
a Loss of Coolant Accident for Boiling Water Reactors, RG 1.3, 1974.
4.
USNRC, Assumptions Used for Evaluating the Potential Radiological Consequence of a
Loss of Coolant Accident for Pressurized Water Reactors, RG 1.4, 1974.
5.
USNRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear
Power Plants, NUREG-0800, 1987.
6.
USNRC, Assumptions Used for evaluation the Potential Radiological Consequences of
a fuel Handling Accident in the Fuel Handling and Storage facility for Boiling and
Pressurized Water Reactors, RG 1.25, 1972.
7.
USNRC, Assessment of the Use of Extended Burnup Fuel in Light Water Power
Reactors, NUREG/CR-5009, 1988.
8.
K.G. Murphy and K.W. Campe, Nuclear Power Plant Control Room Ventilation System
Design for Meeting General Criterion 19, published in proceedings of the 13th AEC Air
Cleaning Conference.
9.
USNRC, Atmospheric Dispersion Models for Potential Accident Consequence
Assessments at Nuclear Power Plants, RG 1.145, 1982.
10.
J.V. Ramsdell and C.A. Simonen, Atmospheric Relative Concentrations in Building
Wakes, NUREG/CR-631, Revision 1, 1997.
11.
J.J. DiNunno, et al., Calculation of distance factors for Power and Test Reactors Sites,
USAEC TID-14844, 1962.
12.
ICRP, Report of Committee II on Permissible Dose for Internal Radiation, ICRP
Publication 2, 1959.
13.
ICRP, Limits for Intakes of Radionuclides by workers, ICRP Publication 30, 1978.
14.
Eckerman, K.F., et al., Limiting values of Radionuclide Intake and Air Concentration
and dose Conversion factors for Inhalation, submersion, and Ingestion; Federal
Guidance Report 11, EPA-520/1-88-020, 1988.
15.
K.F. Eckerman and J.C. Ryman, External Exposure to Radionuclides in Air, Water, and
Soil, Federal Guidance Report 12, EPA-402-R-93-081.
16.
USDOE, External Dose-Rate Conversion Factors for Calculation of Dose to the Public,
DOE/EH-0070.
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment 2
Page 1 of 1
LIST OF RECENTLY ISSUED
NRC REGULATORY ISSUE SUMMARIES
_____________________________________________________________________________________
Regulatory Issue
Date of
Summary No.
Subject
Issuance
Issued to
_____ _______________________________________________________________________________
2001-18
Requirements for Oath or
Affirmation
08/22/2001
All holders of construction permits
or operating licenses for nuclear
power reactors and non-power
reactors under Part 50 of Title 10
of the Cod of Federal Regulations
(10 CFR Part 50), including those
who have permanently ceased
operations and have certified that
fuel has been permanently
removed from the reactor vessel,
holders of licenses issued under
10 CFR Part 72, and holders of
certificates issued under 10 CFR Part 76
2001-17
Preparation and Scheduling of
Operator Licensing Examinations
08/22/2001
All holders of operating licenses
for nuclear power reactors, except
those who have permanently
ceased operations and have
certified that fuel has been
permanently removed from the
reactor vessel
2001-16
Update of Evacuation Time
Estimates
08/01/2001
All holders of operating licenses
for nuclear power plants
2001-15
Performance of DC-Powered
Motor-Operated Valve Actuators
08/01/2001
All holders of operating licenses
for nuclear power reactors, except
those who have permanently
ceased operations and have
certified that fuel has been
permanently removed from the
reactor vessel
2001-14
Position on Reportability
Requirements for Reactor Core
Isolation Cooling System Failure
07/19/2001
All holders of boiling-water reactor
(BWR) operating licenses