ML053460347
| ML053460347 | |
| Person / Time | |
|---|---|
| Issue date: | 03/07/2006 |
| From: | Charemagne Grimes NRC/NRR/ADRA/DPR |
| To: | |
| Kotzalas, M, NRR/ADRA/DPR, 301-415-0566 | |
| References | |
| RIS-06-004 | |
| Download: ML053460347 (13) | |
See also: RIS 2006-04
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
March 7, 2006
NRC REGULATORY ISSUE SUMMARY 2006-04
EXPERIENCE WITH IMPLEMENTATION OF ALTERNATIVE
SOURCE TERMS
ADDRESSEES
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.
INTENT
The U.S. Nuclear Regulatory Commission (NRC) is issuing this regulatory issue summary (RIS)
to update addressees on experience with implementation of alternative source terms (ASTs) in
design basis accident (DBA) radiological analyses of currently licensed light water reactors. It is
expected that recipients will review the information for applicability to their facilities and consider
actions, as appropriate. However, any suggestions this RIS may contain are not NRC
requirements; therefore, no specific action or written response is required on the part of an
addressee.
In particular, the information in this RIS should be useful to licensing and engineering staffs at
currently operating reactors and to contractors supporting implementation of an AST through a
license amendment request (LAR). This RIS should aid in the reduction of requests for
additional information (RAIs) and help improve the planning for and implementation of an AST.
BACKGROUND INFORMATION
Many of the LARs for implementation of an AST have lacked the necessary information for the
NRC staff to make its safety determination. The purpose of this RIS is to discuss the more
frequent and significant issues encountered by the NRC staff during its review of AST
submittals and to provide information for licensees to consider when developing submittals for
implementation of an AST. Frequently, licensees submit AST analyses using methods that are
not simple variations or technically justifiable extensions of those addressed in the NRC
guidance documents. These methods require additional NRC staff research and review and
frequently result in RAIs. Licensees should consider early interaction with NRC staff during
development of submittals that incorporate assumptions, methods, or analyses not specifically
discussed in the regulatory documents or guidance, since they may require additional NRC staff
review.
Page 2 of 8
SUMMARY OF ISSUE
1.
Level of Detail Contained in LARs
An AST amendment request should describe the licensees analyses of the radiological and
non-radiological impacts and provide a justification for the proposed modification in sufficient
detail to support review by the NRC staff. For example, the AST amendment request should (1)
provide justification for each individual proposed change to the technical specifications (TS), (2)
identify and justify each change to the licensing basis accident analyses, and (3) contain
enough details (e.g., assumptions, computer analyses input and output) to allow the NRC staff
to confirm the dose analyses results in independent calculations. The provision of sufficient
detail is necessary for the NRC staff to be able to conclude, with reasonable assurance,
whether the licensees analyses and changes are acceptable. For a previous NRC staff
discussion on the level of detail necessary for review, see RIS 2001-19, Deficiencies in the
Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License
Amendment Requests (Ref 1).
In response to RAIs, some licensees have made changes to originally proposed LARs and their
supporting analyses. In some cases, these changes were extensive or involved multiple
re-analyses and supplements. Because of the depth and scope of many AST submittals,
multiple changes to the original submittal (particularly those with multiple supplements that
revise portions of previous supplements) can increase the chance of NRC staff using
information that has been superceded during the review. For these cases, NRC staff
recommends that licensees identify the most current analyses, assumptions, and TS changes in
their submittal and supplements to the submittal.
2.
Main Steam Isolation Valve (MSIV) Leakage and Fission Product Deposition in
Piping
For calculation of aerosol settling velocity in the main steamline (MSL) piping of boiling water
reactors, some LARs reference Accident Evaluation Report (AEB) 98-03, Assessment of
Radiological Consequences for the Perry Pilot Plant Application Using the Revised
(NUREG-1465) Source Term (Ref. 2). This is acceptable. However, it is important to note that
the report was written based on the parameters of a particular plant and, therefore, the removal
rate constant is specific to that plant. Any licensee who chooses to reference these AEB 98-03
assumptions should provide appropriate justification that the assumptions are applicable to their
particular design.
Both NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (Ref. 3) and
Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design
Basis Accidents at Nuclear Power Plants (Ref. 4) define AST as a fission product release from
the reactor core into the containment. Neither provides sufficient information regarding the
amount and composition of fission products in the reactor vessel or the attached piping. As
indicated in Appendix A to RG 1.183, Regulatory Position 6.0, the NRC staff accepts the
practice of treating fission product concentration in containment (more specifically in the drywell)
as representative of that in the vicinity of the MSIV. Some AST amendment requests have
reduced the drywell activity levels by assuming mixing with the free air volume of the wetwell. If
appropriate justification is provided, the suppression pool free air volume may be included,
provided there is a mechanism to ensure mixing between the drywell and wetwell. For example,
the NRC staff would expect to see thermal and hydraulic analyses in support of crediting mass
exchange between the wetwell and drywell airspace for time periods associated with fission
product releases.
Page 3 of 8
The size distribution of airborne particles in the vicinity of the MSIV is, in general, different from
that in the containment. Since the piping is attached to the source of fission product releases,
the agglomeration process of highly concentrated but small aerosols may substantially differ
from that in containment. Modeling of MSL piping may include volumes between the reactor
pressure vessel and the inboard MSIV (inboard volume), between the inboard and outboard
valves (in-between volume), and outside of the outboard valve (outboard volume). Since a
majority of large (i.e., heavier) particles deposit in the inboard volume, the distribution of the
aerosol that leaks to the subsequent volume is smaller (i.e., lighter) particles. This particle
behavior leads to the conclusion that the choice of an effective settling velocity in any volume
should account for the distribution of particle sizes in that volume. The steam flow rate during
the accident also affects the removal of particles and should be accounted for in the analysis.
For aerosol settling, only horizontal sections of piping should be credited. The effective settling
area should be calculated as the length of horizontal piping multiplied by the pipe diameter.
Deposition of gaseous iodine (elemental and organic) in the piping is a frequent point of
contention between licensees and NRC staff. Some licensees claim that because of chemical
adsorption, a large portion of iodine is deposited on the piping surface. However, this
deposition is strongly dependent upon the thermal and hydraulic conditions in the piping. Given
the large uncertainty associated with iodine behavior in piping, deposition of gaseous iodine in
piping should be omitted unless appropriate justification is provided (including providing
estimates of the thermal and hydraulic conditions in the piping).
3.
When implementing an AST, some licensees have proposed that certain engineered safety
features (ESF) ventilation systems not be credited as a mitigation feature in response to an
accident. In some cases, the licensees revised design basis analysis introduced the
assumption that normal (non-ESF) ventilation systems are operating during all or part of an
accident scenario. Such an assumption is inappropriate unless the non-ESF system meets
certain qualities, attributes, and performance criteria as described in RG 1.183, Regulatory
Positions 4.2.4 and 5.1.2. For example, credit for the operation of non-ESF ventilation systems
should not be assumed unless they have a source of emergency power. In addition, the
operation of ventilation systems establishes certain building or area pressures based upon their
flowrates. These pressures affect leakage and infiltration rates which ultimately affect operator
dose. Therefore, to credit the use of these systems, licensees should incorporate the systems
into the ventilation filter testing program in Section 5 of the TS. In summary, use of non-ESF
ventilation systems during a DBA should not be assumed unless the systems have emergency
power and are part of the ventilation filter testing program in Section 5 of the TS.
Generic Letter (GL) 2003-01, Control Room Habitability (Ref. 5) requested licensees to
confirm the ability of their facilitys control room to meet applicable habitability regulatory
requirements. In addition, licensees were requested to confirm that control room habitability
systems were designed, constructed, configured, operated and maintained in accordance with
the facilitys design and licensing bases. The GL placed emphasis on licensees confirming that
the most limiting unfiltered inleakage into the control room envelope (CRE) was not greater than
the value assumed in the DBA analyses. The tests, measurements, and analyses which were
performed for this confirmation were to be described in the response to the GL. Some AST
amendment requests proposed operating schemes for the control room and other ventilation
systems which affect areas adjacent to the CRE and are different from the manner of operation
and performance described in the response to the GL without providing sufficient justification for
the proposed changes in the operating scheme. In some cases, licensees proposed new
modes of operation that lacked confirmation of the CRE inleakage characteristics.
Page 4 of 8
1 Use of parametric studies in which inleakage rates are varied is not a preferable alternative to CRE
inleakage measurements. (see GL 2003-01)
Measurements1 of these characteristics are important to confirm inleakage assumptions used in
the analyses for an AST amendment, even for those situations in which the air in the control
room would appear to be stagnant.
4.
Atmospheric Dispersion
Licensees may continue to use atmospheric relative concentration (P/Q) values and
methodologies from their existing licensing-basis analyses when appropriate. Licensees also
have the option to adopt the generally less conservative (more realistic) updated NRC staff
guidance on determining P/Q values in support of design basis control room radiological
habitability assessments provided in RG 1.194, Atmospheric Relative Concentrations for
Control Room Radiological Habitability Assessments at Nuclear Power Plants (Ref. 6).
Regulatory positions on P/Q values for offsite (i.e., exclusion area boundary and low population
zone) accident radiological consequence assessments are provided in RG 1.145, Atmospheric
Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants
(Ref. 7).
Based on submittal reviews, the NRC staff identified the following areas of improvement for
licensee submittals that propose revision of the design basis atmospheric dispersion analyses
for implementing AST. They should include the following information:
A site plan showing true North and indicating locations of all potential accident release
pathways and control room intake and unfiltered inleakage pathways (whether assumed or
identified during inleakage testing).
Justification for using control room intake P/Q values for modeling the unfiltered inleakage,
if applicable.
A copy of the meteorological data inputs and program outputs along with a discussion of
assumptions and potential deviations from staff guidelines. Meteorological data input files
should be checked to ensure quality (e.g., compared against historical or other data and
against the raw data to ensure that the electronic file has been properly formatted, any unit
conversions are correct, and invalid data are properly identified).
When running the control room atmospheric dispersion model ARCON96, two or more files of
meteorological data representative of each potential release height should be used if P/Q values
are being calculated for both ground-level and elevated releases (see RG 1.23, Onsite
Meteorological Programs, Regulatory Position 2 (Ref. 8) and Table A-2 in Appendix A to
RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability
Assessments at Nuclear Power Plants). In addition, licensees should be aware that (1) two
levels of wind speed and direction data should always be provided as input to each data file,
(2) fields of nines (e.g., 9999) should be used to indicate invalid or missing data, and (3) valid
wind direction data should range from 1° to 360°. Licensees should also provide detailed
engineering information when applying the default plume rise adjustment cited in RG 1.194 to
control room P/Q values to account for buoyancy or mechanical jets of high energy releases.
Page 5 of 8
This information should demonstrate that the minimum effluent velocity during any time of the
release over which the adjustment is being applied is greater than the 95th percentile wind
speed at the height of release.
When running the offsite atmospheric dispersion model PAVAN, two or more files of
meteorological data representative of each potential release height should be used if P/Q values
are being calculated for pathways with significantly different release heights (e.g., ground level
versus elevated stack). The joint frequency distributions of wind speed, wind direction, and
atmospheric stability data used as input to PAVAN should have a large number of wind speed
categories at the lower wind speeds in order to produce the best results (e.g., Section 4.6 of
NUREG/CR-2858, PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis
Accidental Releases of Radioactive Materials from Nuclear Power Stations (Ref. 9), suggests
wind speed categories of calm, 0.5, 0.75, 1.0, 1.25, 1.5, 2.0, 3.0, 4.0 5.0, 6.0, 8.0 and 10.0
meters per second).
5.
Modeling of ESF Leakage
ESF systems that recirculate sump water outside the primary containment may leak during their
intended operation. This release source includes leakage through valve packing glands, pump
shaft seals, flanged connections, and other similar components. This release source may also
include leakage through valves isolating interfacing systems (e.g., refueling water storage tank).
Appendix A to RG 1.183, Regulatory Position 5, states that the radiological consequences from
the postulated [ESF] leakage should be analyzed and combined with consequences postulated
for other fission product release paths to determine the total calculated radiological
consequences from the [loss-of-coolant accident] LOCA.
The allowable ESF leakage is typically contained in the plants TS or procedures. The ESF
leakage at accident conditions may differ from the ESF leakage at normal operating conditions.
Licensees should account for ESF leakage at accident conditions in their dose analyses so as
not to underestimate the release rate.
In Appendix A to RG 1.183, Regulatory Position 5.5, the NRC staff provided a conservative
value of 10 percent as the assumed amount of iodine that may become airborne from ESF
leakage that is less than 212 °F. The NRC staff structured this regulatory position to be
deterministic and conservative. The 10 percent value also compensates for the lack of research
concerning iodine speciation beyond the containment and the uncertainties of applying
laboratory data to the post-accident environment of the plant. Regulatory Position 5.5 states
that a smaller flash fraction could be justified. Some licensees have referenced
NUREG/CR-5950, Iodine Evolution and pH Control (Ref. 10) to justify a smaller flash fraction.
However, NUREG/CR-5950 was developed for very specific laboratory conditions and the
results have a degree of uncertainty. The mechanism for release of the fluid is also uncertain.
Leaked fluid may spray onto surfaces and evaporate, or be sprayed in fine droplets into the air.
A value of less than 10 percent can be justified by including considerations for plant-specific
variables, including the post-accident environment (e.g., impurities in the water or the presence
of organic substances) and the uncertainties in the application of research situations to plant
environments.
Figure 3.1 in NUREG/CR-5950 can be used to quantify the amount of elemental iodine as a
function of the sump water pH and the concentration of iodine in the solution. In some cases,
however, licensees have misapplied this figure. Rather than using the total concentration of
Page 6 of 8
iodine (i.e., stable and radioactive), licensees based their assessment on only the radioactive
iodine in the sump water. By using only the radioactive iodine, licensees have underestimated
how much iodine evolves during post-accident conditions.
6.
Release Pathways
Changes to the plant configuration associated with an LAR (e.g., an open containment during
refueling) may require a re-analysis of the design basis dose calculations. A request for TS
modifications allowing containment penetrations (i.e., personnel air lock, equipment hatch) to be
open during refueling cannot rely on the current dose analysis if this analysis has not already
considered these release pathways. RG 1.194, Regulatory Position 3.2.4.2 supports review of
penetration pathways, by stating that leakage is more likely to occur at a penetration, [and that
the] analysts must consider the potential impact of leakage from building penetrations exposed
to the environment. Therefore, releases from personnel air locks and equipment hatches
exposed to the environment and containment purge releases prior to containment isolation need
to be addressed.
Some licensees have identified unique release pathways that had not been previously
considered. For example, a recent submittal noted that containment hatches and containment
plugs may be removed during refueling. The removal of these barriers creates new release
pathways. Licensees are responsible for identifying all release pathways and for considering
these pathways in their AST analyses, consistent with any proposed modification.
7.
Primary to Secondary Leakage
Some analysis parameters can be affected by density changes that occur in the process steam.
The NRC staff continues to find errors in LAR submittals concerning the modeling of primary to
secondary leakage during a postulated accident. This issue is discussed in Information Notice (IN) 88-31, "Steam Generator Tube Rupture Analysis Deficiency," (Ref. 11) and Item 3.f in
RIS 2001-19. An acceptable methodology for modeling this leakage is provided in Appendix F
to RG 1.183, Regulatory Position 5.2.
8.
Elemental Iodine Decontamination Factor (DF)
Appendix B to RG 1.183, provides assumptions for evaluating the radiological consequences of
a fuel handling accident. If the water depth above the damaged fuel is 23 feet or greater,
Regulatory Position 2 states that the decontamination factors for the elemental and organic
[iodine] species are 500 and 1, respectively, giving an overall effective decontamination factor of
200. However, an overall DF of 200 is achieved when the DF for elemental iodine is 285, not
500.
9.
Isotopes Used in Dose Assessments
For some accidents (e.g., main steamline break and rod drop), licensees have excluded noble
gas and cesium isotopes from the dose assessment. The inclusion of these isotopes should be
addressed in the dose assessments for AST implementation.
10.
Definition of Dose Equivalent 131I
In the conversion to an AST, licensees have proposed a modification to the TS definition of
dose equivalent 131I. Some have modified the definition to base it upon the thyroid dose
conversion factors of International Commission on Radiation Protection (ICRP) Publication 2,
Page 7 of 8
2 An off-gas or waste gas system release does not need to be addressed for a full AST implementation
unless a design change is being proposed for the waste gas tank or systems at the same time.
3 Branch Technical Position ETSB 11-5, Standard Review Plan for the Review of Safety Analysis Reports
for Nuclear Power Plants (Ref. 16).
Report of Committee II on Permissible Dose for Internal Radiation (Ref. 12) or ICRP
Publication 30, Limits for Intakes of Radionuclides by Workers (Ref. 13). Others have
proposed a definition which is a combination of different iodine dose conversion factors, (e.g.,
RG 1.109, Revision 1, Calculation of Annual Doses to Man from Routine Releases of Reactor
Effluents for the Purpose of Evaluating Compliance with 10 CFR [Part] 50, Appendix I (Ref. 14),
ICRP Publication 2, Federal Guidance Report 11, Limiting Values of Radionuclide Intake and
Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion (Ref.
15). Although different references are available for dose conversion factors, the TS definition
should be based on the same dose conversion factors that are used in the determination of the
reactor coolant dose equivalent iodine curie content for the main steamline break and steam
generator tube rupture accident analyses.
11.
Acceptance Criteria for Off-Gas or Waste Gas System Release
As part of full AST implementation,2 some licensees have included an accident involving a
release from their off-gas or waste gas system. For this accident, they have proposed
acceptance criteria of 500 millirem (mrem) total effective dose equivalent (TEDE).
The acceptance criteria for this event is that associated with the dose to an individual member
of the public as described in 10 CFR Part 20, Standards for Protection Against Radiation.3
When the NRC revised 10 CFR Part 20 to incorporate a TEDE dose, the offsite dose to an
individual member of the public was changed from 500 mrem whole body to 100 mrem TEDE.
Therefore, any licensee who chooses to implement AST for an off-gas or waste gas system
release should base its acceptance criteria on 100 mrem TEDE. Licensees may also choose
not to implement AST for this accident and continue with their existing analysis and acceptance
criteria of 500 mrem whole body.
12.
Containment Spray Mixing
Some plants with mechanical means for mixing containment air have assumed that the
containment fans intake air solely from a sprayed area and discharge it solely to an unsprayed
region or vice versa. Without additional analysis, test measurements or further justification, it
should be assumed that the intake of air by containment ventilation systems is supplied
proportionally to the sprayed and unsprayed volumes in containment.
BACKFIT DISCUSSION
This RIS provides guidance for licensees who voluntarily request to amend their licenses to
implement an AST. Accordingly, this RIS is not a backfit under 10 CFR 50.109, and the NRC
staff did not prepare a backfit analysis.
FEDERAL REGISTER NOTIFICATION
A notice of opportunity for public comment was not published in the Federal Register
because this RIS is informational and pertains to NRC staff positions that do not represent
departures from current regulatory requirements and practice.
Page 8 of 8
SMALL BUSINESS REGULATORY ENFORCEMENT FAIRNESS ACT OF 1996
In accordance with the Small Business Regulatory Enforcement Fairness act of 1996, the NRC
has determined that this action is not a major rule and has verified this determination with the
Office of Information and Regulatory Affairs of the Office of Management and Budget (OMB).
PAPERWORK REDUCTION ACT STATEMENT
This Generic Letter (Bulletin, etc.) contains information collection requirements that are subject
to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections
were approved by the OMB, approval number 3150-0011.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to respond to, a request for
information or an information collection requirement unless the requesting document displays a
currently valid OMB control number.
CONTACT
Please direct any questions about this matter to the technical contacts listed below, or to the
appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
/RA/
Christopher I. Grimes, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contacts: Margie Kotzalas, NRR
R. Brad Harvey, NRR
301-415-0566
301-415-4118
E-mail: MXK5@nrc.gov
E-mail: RBH@nrc.gov
Enclosures: 1. References
2. Recently Issued Regulatory Issue Summaries
Note: NRC generic communications may be found on the NRC public website,
http://www.nrc.gov, under Electronic Reading Room/Document Collections.
- previous concurrence
OFFICE
AADB:DRA
TechEditor
BC:AADB:DRA
D:DRA
BC:ITSB:DIRS
NAME
RBHarvey*
HChang*
MKotzalas*
JLyons*
TBoyce*
DATE
2 /6 /2006
12 /16 /2005
2 /3 /2006
2/17 /2006
2 /24 /2006
OFFICE
D:DIRS
D:DORL
D:DSS
OGC (SBREFA)
NAME
MCase*
CHaney*
TMartin*
CNolan*
SUttal*
SCrockett*
DATE
2/27/2006
2 /27/2006
2 /22/2006
2 /23 /2006
3 / 2 /2006
3 /1 /2006
OFFICE
PMAS:NRR
OIS
LA:PGCB:DPR
PGCB:DPR
BC:PGCB:DPR
D:DPR
NAME
DMcCain*
BShelton*
CHawes*
AMarkley*
CJackson*
CGrimes*
DATE
02/20/2006
2/23/2006
03/06/2006
3/3/2006
03/06/2006
03/07/2006
Enclosure 1
Page 1 of 2
REFERENCES
1.
U.S. Nuclear Regulatory Commission, "Deficiencies in the Documentation of Design Basis
Radiological Analyses Submitted in Conjunction with License Amendment Requests," RIS 2001-19, October 18, 2001.
2.
J. Schaperow et al., Assessment of Radiological Consequences for the Perry Pilot Plant
Application using the Revised (NUREG-1465) Source Term, U.S. Nuclear Regulatory
Commission, AEB 98-03, December 9, 1998
3.
L. Soffer et al., Accident Source Terms for Light-Water Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, NUREG-1465, February 1995.
4.
U.S. Nuclear Regulatory Commission, Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Plants, Regulatory Guide 1.183,
July 2000.
5.
U.S. Nuclear Regulatory Commission, Control Room Habitability, NRC Generic Letter 2003-01, June 12, 2003.
6.
U.S. Nuclear Regulatory Commission, Atmospheric Relative Concentrations for Control
Room Radiological Habitability Assessments at Nuclear Power Plants, Regulatory
Guide 1.194, June 2003.
7.
U.S. Nuclear Regulatory Commission, Atmospheric Dispersion Models for Potential
Accident Consequence Assessments at Nuclear Power Plants, Regulatory Guide 1.145,
Revision 1, November 1982.
8.
U.S. Nuclear Regulatory Commission, Onsite Meteorological Programs, Regulatory
Guide 1.23, February 1972.
9.
T.J. Bander, PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis
Accidental Releases of Radioactive Materials from Nuclear Power Stations, U.S. Nuclear
Regulatory Commission, NUREG/CR-2858, November 1982.
10.
E.C. Beahm, et al., Iodine Evolution and pH Control, U.S. Nuclear Regulatory
Commission, NUREG/CR-5950, December 1992.
11.
U. S. Nuclear Regulatory Commission, "Steam Generator Tube Rupture Analysis
Deficiency, IN 1988-31, May 25, 1988.
12.
International Commission on Radiation Protection, Report of Committee II on Permissible
Dose for Internal Radiation, ICRP Publication 2, 1959.
13.
International Commission on Radiation Protection, Limits for Intakes of Radionuclides by
Workers, ICRP Publication 30, 1978.
14.
U.S. Nuclear Regulatory Commission, Calculation of Annual Doses to Man from Routine
Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR
[Part] 50, Appendix I, Regulatory Guide 1.190, Revision 1, October 1977.
Enclosure 1
Page 2 of 2
15.
K.F. Eckerman et al., Limiting Values of Radionuclide Intake and Air Concentration and
Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance
Report 11, Environmental Protection Agency, EPA-520/1-88-020,1988.
16.
U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety
Analysis Reports for Nuclear Power Plants, NUREG-0800, Branch Technical Position
ETSB 11-5, July 1981.
Enclosure 2
RECENTLY ISSUED REGULATION ISSUE SUMMARIES
1.
USNRC, Deficiencies in the Documentation of Design Basis Radiological Analyses
Submitted in Conjunction with License Amendment Requests, RIS 01-019,
October 18, 2001