[Federal Register Volume 70, Number 171 (Tuesday, September 6, 2005)]
[Rules and Regulations]
[Pages 52893-52899]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-17589]



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Rules and Regulations
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Federal Register / Vol. 70, No. 171 / Tuesday, September 6, 2005 / 
Rules and Regulations

[[Page 52893]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[Docket No. PRM-50-76]


Robert H. Leyse; Denial of Petition for Rulemaking

AGENCY: Nuclear Regulatory Commission.

ACTION: Petition for rulemaking; denial.

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is denying a petition 
for rulemaking submitted by Mr. Robert H. Leyse (PRM-50-76). The 
petitioner requests that the NRC's regulations concerning the specified 
evaluation models for emergency core cooling systems (ECCS) and 
associated guidance documents be amended. The petitioner asserts that 
amendments are necessary to correct technical deficiencies in the 
correlations and data used for calculation of metal-water oxidation. 
The petitioner states that the correlations and data do not consider 
the complex thermal-hydraulic conditions present during a loss-of-
coolant accident (LOCA), including the potential for very high fluid 
temperature. The Commission is denying Mr. Leyse's petition for 
rulemaking (PRM-50-76). None of the specific technical issues raised by 
the petitioner have shown safety-significant deficiencies in the 
research, calculation methods, or data used to support ECCS performance 
evaluations. NRC's technical safety analysis demonstrates that current 
procedures for evaluating ECCS performance are based on sound science 
and that no amendments to the NRC's regulations and guidance documents 
are necessary.

ADDRESSES: The NRC is making the documents identified in the table 
below available to interested persons through several means. Publicly 
available documents related to this petition, including the petition 
for rulemaking, public comments received, and the NRC's letter of 
denial to the petitioner, may be viewed electronically on public 
computers in the NRC's Public Document Room (PDR), O-1 F21, One White 
Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. The PDR 
reproduction contractor will copy documents for a fee. Selected 
documents, including comments, may be viewed and downloaded 
electronically via the NRC rulemaking Web site at http://ruleforum.llnl.gov.
    Publicly available documents created or received at the NRC after 
November 1, 1999, are also available electronically at the NRC's 
Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. 
From this site, the public can gain access into the NRC's Agencywide 
Documents Access and Management System (ADAMS), which provides text and 
image files of NRC's public documents. If you do not have access to 
ADAMS or if you have problems in accessing the documents in ADAMS, 
contact the PDR reference staff at (800) 387-4209 or (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#19697d6b59776b7a377e766f"><span class="__cf_email__" data-cfemail="14647066547a66773a737b62">[email protected]</span></a>.

------------------------------------------------------------------------
            Document               PDR      Web            ADAMS
------------------------------------------------------------------------
Federal Register Notice--             X        X   ML022800472
 Receipt of Petition for
 Rulemaking (67 FR 51783; Aug.
 9, 2002).
Letter of Denial to the               X        X   ML052220454
 Petitioner.
Penn State/US NRC ``Rod Bundle                     ML023040657
 Test Facility and Reflood Heat
 Transfer Program''.
Petition for Rulemaking (PRM-50-      X        X   ML022240009
 76).
Public Comments for PRM-50-76..       X        X   ML042740105
US NRC Office of Nuclear              X        X   ML041210109
 Research (RES) ``Technical
 Safety Analysis of PRM-50-76,
 A Petition for Rulemaking to
 Amend Appendix K to 10 CFR
 Part 50 and Regulatory Guide
 1.157''.
US NRC, ``Updated Program Plan   .......  .......  ML031810103
 for High-Burnup Light-Water
 Reactor Fuel''.
Studies of Metal Water           .......  .......  ML050550198
 Reactions at High
 Temperatures, III.
 Experimental and Theoretical
 Studies of the Zirconium-Water
 Reaction,'' L. Baker and L.C.
 Just, ANL-6548 (May 1962).
PWR FLECHT (Full Length          .......  .......  ML052230221
 Emergency Cooling Heat
 Transfer) Final Report,''
 April 1971.
Zirconium Metal-Water Oxidation  .......  .......  ML052230079
 Kinetics IV. Reaction Rate
 Studies,'' ORNL/NUREG-17,
 August 1977..
------------------------------------------------------------------------


FOR FURTHER INFORMATION CONTACT: Timothy A. Reed, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone (301) 415-1462, e-mail <a href="/cdn-cgi/l/email-protection#40140112002e32236e272f36"><span class="__cf_email__" data-cfemail="cd998c9f8da3bfaee3aaa2bb">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

Background

    The petition for rulemaking designated PRM-50-76 was received by 
the NRC on May 1, 2002. A notice of receipt of the petition and request 
for public comment was published in the Federal Register (FR) on August 
9, 2002 (67 FR 51783). The notice of receipt requested comment on two 
questions: (1) Are the petitioner's three concerns about ECCS cooling 
valid, and if so, do these concerns constitute a significant safety 
concern? (2) Are there actions available to the Commission other than 
rulemaking that would effectively address the concerns raised by the 
petitioner?

The Petition

    The petition, PRM-50-76, covers three broad issues: (1) Amending 
Appendix K to Part 50 of the Commission's regulations, (2) amending 
Regulatory Guide (RG) 1.157, and (3) the need for further analysis of 
the 10 CFR Part 50, Appendix K, backup data.

Issue 1: Amending Appendix K to Part 50

    The petitioner describes at length alleged technical deficiencies 
in Appendix K Section I.A.5, ``Metal-Water Reaction Rate.'' The 
petitioner claims that Section I.A.5 does not accurately describe the 
extent of zirconium-water reactions that may occur during a LOCA. The 
petitioner states that the

[[Page 52894]]

Baker-Just equation, which is used to calculate the metal-water 
reaction in assessing ECCS performance, does not include any allowance 
for the complex thermal-hydraulic conditions during a LOCA, including 
the potential for very high bulk fluid temperatures within the cooling 
channels of the zirconium-clad fuel elements.
    The petitioner cites the abstract of an Argonne National Laboratory 
(ANL) report (ANL-6548 ``Studies of Metal Water Reactions at High 
Temperatures, III. Experimental and Theoretical Studies of the 
Zirconium-Water Reaction,'' L. Baker and L.C. Just, May 1962) and 
disputes the conclusions based on the petitioner's opinion that the 
tests discussed in ANL-6548 do not accurately reflect the conditions 
present during a LOCA. The petitioner makes the following points to 
support his views:
    <bullet> The bulk water temperature was no greater than 315 [deg]C 
(599 [deg]F).
    <bullet> The volume of water within the test apparatus was 
substantially greater than the volume of zirconium specimens, creating 
a vastly greater capacity to cool the heated zirconium particles of the 
Baker and Just experiment than would exist under LOCA conditions.
    <bullet> Zirconium specimens were exposed to water only, while LOCA 
conditions include steam and nonequilibrium water-steam mixtures that 
reached higher bulk fluid temperatures.
    <bullet> A footnote in ANL-6548 states: ``This discussion is of a 
preliminary nature: work in this area is continuing.'' Based on this 
footnote, the petitioner concludes that it is not appropriate to apply 
the Baker-Just equation as prescribed in Appendix K Section I.A.5 for 
the calculation of energy release rates, hydrogen generation, and 
cladding oxidation from the metal-water reaction.

Issue 2: Amending Regulatory Guide 1.157

    The petitioner states that RG 1.157, which allows use of data from 
NUREG-17 (ORNL/NUREG-17, ``Zirconium Metal-Water Oxidation Kinetics IV, 
Reaction Rate Studies,'' by Cathcart et al., August 1977) for 
calculating energy release rates, hydrogen generation, and cladding 
oxidation for cladding temperatures greater than 1900 [deg]F, results 
in flawed ECCS performance evaluations. The petitioner claims the 
NUREG-17 data is based on very limited test conditions and consequently 
the results should not be used for evaluating LOCA conditions.
    In support of this contention, the petitioner describes the 
following test conditions:
    <bullet> Zircaloy-4 specimens exposed only to steam, rather than 
fluid conditions as present in a LOCA.
    <bullet> No documented heat transfer from the Zircaloy surface to 
the slow-flowing steam.
    <bullet> Small-scale laboratory testing without conditions typical 
of the complex thermal-hydraulic conditions that prevail during a LOCA.
    <bullet> An unexplained shift from the MaxiZWOK (testing apparatus 
for investigations in the temperature range 1652 [deg]F to 1832 [deg]F) 
to the MiniZWOK (a different testing apparatus for investigations in 
the temperature range 1832 [deg]F to 2734 [deg]F).
    The petitioner believes that the investigators' conclusions include 
a statement that ``overlooks the very substantially greater mass 
transfer coefficients that accompany the so-called appropriate heat 
transfer coefficients.'' The petitioner concludes that ``it is those 
very substantially greater mass transfer coefficients that led to the 
temperature overshoot of the MaxiZWOK test at 1832 [deg]F, and that 
would have led to very substantially greater temperature overshoots and 
likely destruction of the Zircaloy tubing if MaxiZWOK had been operated 
over the temperature range of the MiniZWOK runs.''
    The petitioner contends that the NUREG-17 investigators do not 
warrant their work, and specifically assume no responsibility for the 
accuracy of their work, and therefore, that NUREG-17 is not applicable 
to the regulation of nuclear power reactors in the United States of 
America. To support this contention, the petitioner cites the following 
statement on the introductory page of NUREG-17: This report was 
prepared as an account of work sponsored by the United States 
Government. Neither the United States nor the Energy Research and 
Development Administration/United States Nuclear Regulatory Commission, 
nor any of their employees, nor any of their contractors, 
subcontractors, or their employees, makes any warranty, express or 
implied, or assumes any legal liability or responsibility for the 
accuracy, completeness or usefulness of any information, apparatus, 
product or process disclosed, or represents that its use would not 
infringe privately owned rights.''

Issue 3: Need for Further Analysis of Appendix K Backup Data

    The petitioner states that the results of Zircaloy bundle test no. 
9573, which was a test done for the Full Length Emergency Cooling Heat 
Transfer (FLECHT) tests and documented in WCAP-7665 (``PWR FLECHT (Full 
Length Emergency Cooling Heat Transfer) Final Report, Westinghouse 
Report WCAP-7665, April 1971''), are applicable to the calculation of 
the metal-water reaction and shows that the Baker-Just equation 
(referenced in Section I.A.5 of Appendix K for calculating the metal-
water reaction) is not conservative. The petitioner states that the 
data in WCAP-7665, which includes test run 9573, includes the complex 
thermal-hydraulic conditions and Zircaloy-water reactions that 
characterize the reflood portion of the LOCA transient. The petitioner 
states that these conditions are not found in the narrow test 
procedures of ANL-6548 or NUREG-17.
    The petitioner states that a pertinent description of the 
complexities of thermal-hydraulic conditions during reflood, including 
negative heat transfer coefficients, is included in Section 3.2.3 of 
WCAP-7665 and that this description applies to data collected with 
FLECHT bundles with stainless steel cladding. The petitioner feels that 
another FLECHT Zircaloy bundle test, run 8874, is also pertinent to 
issues raised in this petition.
    The petitioner cites Section 5.6 of WCAP-7665 and finds statements 
comparing Zircaloy to stainless steel to be misleading because they 
imply that stainless steel heat transfer coefficients may be used as a 
conservative representation of Zircaloy behavior. The petitioner 
believes that the differences in behavior for various test runs are 
explained by the differences in the thermal-hydraulic conditions 
leading to a different combination of heat transfer and mass transfer 
factors, and are not due to inconsistency of the data, as implied by 
the report.
    The petitioner also finds WCAP-7665, Section 5.11, ``Materials 
Evaluation,'' to be misleading in view of the total experience with 
FLECHT run 9573. Finally, the petitioner notes that the same warning 
language used in NUREG-17 is on the cover page of WCAP-7665.
    The petitioner further identifies several aspects of the data 
supporting the document entitled ``Acceptance Criteria for Emergency 
Core Cooling Systems for Light-Water Cooled Nuclear Reactors-Opinion of 
the Commission,'' (Docket No. RM50-1, December 28, 1973) and notes the 
Commission concluded: ``It is apparent, however, that more experiments 
with Zircaloy cladding are needed to overcome the impression left from 
run 9573.'' The petitioner finds that there has been a lack of 
appropriate response to the Commission's expressed wish for more

[[Page 52895]]

experiments, and believes that at the very least, run 9573 should have 
been repeated. The petitioner emphasizes that although at least $1 
billion had been expended on other analytical efforts, there has been 
no reported analysis of FLECHT run 9573.
    The petitioner states that the test programs discussed in the 
petition were funded by Government agencies. He believes that most of 
the programs were firmly controlled by those ``who were indoctrinated 
in the methods of the tightly regimented Naval Reactors Program.'' The 
petitioner finds that the ``biased reporting of WCAP-7665 may be traced 
to these controls'' and believes that ``the lack of application of the 
MaxiZWOK apparatus beyond 1832 [deg]F in NUREG-17 may likely be traced 
to rigid restrictions by management at the NRC.'' The petitioner 
further contends that while the Argonne work in ANL-6548 was likely 
less impacted by these controls, the controls likely did inhibit 
further analysis or reporting of FLECHT run 9573.
    The petitioner notes that he has made several requests to the 
Knolls Atomic Power Laboratory for report KAPL-1534 and that his 
requests have been ignored.

Public Comments on the Petition

    Six letters of public comment were received on the petition in 
response to the request for public comment. Three of these letters were 
from the petitioner. These letters are summarized below.
    By letter dated September 11, 2002, the petitioner provided 
comments that did not raise new issues. The petitioner stated that the 
Baker-Just equation and the Cathcart-Pawel equation in NUREG-17 have 
been grossly misapplied by the NRC. According to him, it is 
fundamentally important that the determinations of LOCA transient 
chemical kinetics include the geometry of the stationary Zircaloy 
reactant in combination with the thermal-hydraulic conditions of the 
flowing water/steam reactant. In addition, he repeated in his letter 
that there are deficiencies in RG 1.157, since it references documents 
such as NUREG-17 that do not consider the complex thermal-hydraulic 
conditions during LOCAs, including the potential for very high fluid 
temperatures. The petitioner also stated that the Commission should 
provide a rational basis for regulation of ECCS performance and perform 
additional experiments with Zircaloy cladding due to the cladding 
failure reported in Westinghouse report WCAP-7665.
    By letter dated October 23, 2002, Westinghouse Electric Company 
submitted comments that opposed the proposed changes. Westinghouse 
commented that runaway oxidation is prevented by the 2200 [deg]F peak 
cladding temperature limit. Additionally, Westinghouse commented that 
the Baker-Just correlation is known to be conservative, over-predicting 
the zirconium-water reaction by as much as 30 percent at the limiting 
temperature (2200 [deg]F). Westinghouse stated that the conditions of 
FLECHT run 9573 (high power and high initial temperatures) were 
extremely severe, intentionally beyond design basis for ECCS 
performance. Westinghouse stated that the Cathcart-Pawel tests had 
adequate steam flow so that the zirconium-water reaction rate was not 
limited by the availability of steam, and as a result, the tests were 
valid. Westinghouse commented that differences between ECCS test 
conditions and reactor core fluid conditions during postulated LOCAs do 
not prevent the current zirconium-water reaction database from being 
applicable to ECCS analysis.
    By letter dated October 25, 2002, the Nuclear Energy Institute 
(NEI) submitted comments supporting the Westinghouse comments, stating 
that extensive testing and analysis by the nuclear industry and 
national laboratories indicate that the Cathcart-Pawel correlation test 
is conservative. The NRC notes that the Cathcart-Pawel correlation is 
intended to be a best estimate, and is not intended to conservatively 
bound metal-water reaction rates. NEI commented that the test run, 
FLECHT 9573, was intentionally performed under very severe, beyond 
design-basis conditions, that post-test evaluations showed oxidation 
was within the expected range, and that runaway oxidation did not occur 
until the cladding temperature was well beyond 2300 [deg]F. NEI further 
commented that the petitioner's concerns do not constitute a 
significant safety concern and thus, there is no need to revise 
Appendix K to Part 50 or RG 1.157.
    By letter dated November 6, 2002, Strategic Teaming and Resource 
Sharing (STARS), a group of six utilities, submitted comments opposing 
the petition. These comments stated that within the range of test 
parameters applicable to ECCS evaluation models, as specified in 
Appendix K and RG 1.157, the regulations and guidance are valid and 
conservative. STARS notes that all of the data referenced in the 
petition was either available to the Commission and industry when the 
regulations and guidance were created or was assessed later when the 
test information became available.
    On November 22, 2002, the petitioner submitted a reply to STARS but 
raised no new issues. On December 14, 2002, the petitioner responded to 
Westinghouse and NEI comments by discussing runaway oxidation in the 
WCAP-12610 report and severe fouling of fuel cladding during a LOCA. 
The petitioner stated that no allowance for higher temperatures due to 
fouling was made in run 9573, and repeated his request for more 
experiments with Zircaloy cladding.

NRC Requirements for ECCS Evaluations

    Section 50.46 specifies the performance criteria against which the 
ECCS must be evaluated. The criteria include the maximum peak cladding 
temperature, the maximum cladding oxidation thickness, the maximum 
total hydrogen generation, and requirements to assure a coolable core 
geometry and abundant long-term cooling. This regulation also states 
that the ECCS cooling performance following postulated LOCAs must be 
calculated in accordance with either a realistic (also called a best-
estimate) evaluation model that accounts for uncertainty or a 
conservative evaluation model that conforms with the required features 
of appendix K to 10 CFR part 50. If a licensee elects to calculate ECCS 
performance using an Appendix K evaluation model, then one important 
feature of that model is the way the metal-water reaction is 
calculated. For this calculation, Appendix K prescribes the use of the 
Baker-Just equation from ANL report ANL-6548 (L. Baker, L.C. Just, 
``Studies of Metal Water Reactions at High Temperatures, III. 
Experimental and Theoretical Studies of the Zirconium-Water Reaction'' 
May 1962). The metal-water reaction, which is predicted to occur during 
the LOCA and which is calculated using the Baker-Just equation, is the 
subject of much of this petition. The Baker-Just equation calculates a 
conservative rate of hydrogen generation and fuel cladding oxidation 
during the LOCA transient. Additionally, for licensees electing to use 
best-estimate calculations to evaluate ECCS performance, NRC RG 1.157 
provides guidance for such evaluations. RG 1.157 allows the use of data 
from NUREG-17 for the calculation of the metal-water reaction.

NRC Technical Evaluation

    The NRC reviewed the petitioner's request and concluded that none 
of the issues raised by the petitioner justified the initiation of 
rulemaking. The NRC's response to the technical issues raised in PRM-
50-76 is based largely on a technical study by the Office of Nuclear 
Regulatory Research (RES) ``Technical

[[Page 52896]]

Safety Analysis of PRM-50-76, A Petition for Rulemaking To Amend 
appendix K to 10 CFR part 50 and Regulatory Guide 1.157.'' The NRC's 
responses to the petitioner's issues are as follows:

Issue 1: Amending Appendix K to Part 50

    The petitioner claims that the requirement to use the Baker-Just 
equation in Section I.A.5 of Appendix K to 10 CFR Part 50, does not 
accurately describe the extent of zirconium-water reaction that may 
occur during a LOCA. He states that the Baker-Just equation does not 
include any allowance for the complex thermal-hydraulic conditions 
during a LOCA. The NRC disagrees with the petitioner's assertions.
    In Section 3.1 of the petition, the petitioner discusses the 
inapplicability of the Baker-Just equation for calculating zirconium-
water reaction rates during a LOCA. The NRC notes that it is important 
to distinguish between the experiments performed by Baker and Just, and 
the equation developed by them and adopted in Appendix K to Part 50. 
Experiments run with 40-60 mil wires at temperatures at, or near, the 
zirconium melting point (3400 [deg]F) for one or two seconds are not 
typical of fuel rod cladding at temperatures in the range of 1800 
[deg]F-2200 [deg]F for 50 to 400 seconds that are postulated to occur 
in a design basis LOCA. In the Baker-Just report, only one data point 
from their experiments (at 3366 [deg]F) is used in developing the 
Baker-Just equation. This one data point was used to anchor the Baker-
Just equation at the melting point of zirconium. The remaining data 
from Bostrum (``The High Temperature Oxidation of Zircaloy in Water,'' 
W. A. Bostrum, WAPD-104 March 1954) and Lemmon (``Studies Relating to 
the Reaction Between Zirconium and Water at High Temperatures,'' A. W. 
Lemmon, Jr., BMI-1154, January 1957), at more relevant zirconium 
cladding conditions, were used by Baker and Just in the derivation of 
their equation. The use of the single data point at the melting 
temperature makes the Baker-Just equation very conservative. At the 
time of the promulgation of Sec.  50.46, the Commission expected the 
NRC staff to obtain new and better zirconium-water reaction data. The 
petitioner also expressed concerns about the need for additional data. 
The substantial work of Cathcart and Pawel was performed for the NRC in 
response to the Commission's expectation.
    The NRC compares the Baker-Just correlation to other correlations 
in a technical study (ADAMS accession ML041210109). The comparisons 
show the conservatism of the Baker-Just correlation in the temperature 
range important for clad oxidation calculations for LOCAs. In the 
discussion of Issue 3, comparisons of the Baker-Just correlation to 
relevant data demonstrate the substantial conservatism of the Baker-
Just correlation. The petitioner expresses concern about the low water 
temperature (no greater than 599 [deg]F) in the Baker-Just experiments. 
This temperature corresponds to the saturation temperature at 1530 
psia, which was the pressure for that particular experiment. While a 
few degrees of liquid superheat may be possible under LOCA/ECCS 
conditions, the degree of nonequilibrium required for higher liquid or 
``bulk'' temperatures postulated by the petitioner is not possible.
    The petitioner is also concerned about the large water volume 
compared to the zirconium sample size with respect to the quench 
capability of zirconium-clad fuel rods. As noted, these experiments 
were atypical in that respect, but barely used in the formulation of 
the Baker-Just correlation. Further, it should be noted that the Baker-
Just report was not intended to be a heat transfer study, but rather an 
investigation of zirconium-water reaction kinetics at very high 
temperatures.
    One interesting feature of the Baker-Just report is the heat and 
mass transfer analysis of an example case analyzed to examine the 
processes limiting the reaction rate. In this severe case, a 0.21 cm 
zirconium sphere at its melting point was dropped into water. Baker 
andJust were concerned that the reaction could be limited by gas phase 
diffusion of steam through a film of steam and hydrogen. This appears 
to be similar to the petitioner's concern. As explained in the Baker-
Just report, water cannot stay in contact with the hot metal and a 
vapor film immediately forms around the sphere. Figure 15 in that 
report shows that vapor phase diffusion is the limiting steam transport 
process for less than 0.2 seconds, during which a slight film of oxide 
is forming on the surface of the sphere. After that, the parabolic rate 
equation, (e.g., the Baker-Just equation) becomes limiting. The figure 
also shows that the gas phase diffusion is far less temperature-
sensitive than the parabolic rate law. Certainly at lower temperatures 
more typical of a LOCA, the parabolic law is even more limiting than 
gas phase diffusion as long as the reaction is not steam starved.
    Comparison of the Baker-Just equation to numerous data sets has 
shown the equation to be conservative. A significant example of this 
conservatism is discussed under Issue 3.
    In summary, the NRC found no technical basis in the petition or in 
NRC records for the assertion that the NRC requirement to use the 
Baker-Just equation, along with other requirements of Appendix K, is 
flawed and is a significant safety concern.

Issue 2: Amending Regulatory Guide 1.157

    The petitioner stated that RG 1.157, which allows use of the data 
and the Cathcart-Pawel equation presented in NUREG-17, results in 
flawed evaluations of ECCS performance. The NRC disagrees with the 
petitioner's assertions on this issue. In Section 3.2 of the petition, 
the petitioner states that the limited test conditions described in 
NUREG-17 preclude the use of the results for LOCA calculations. He 
further states that Zircaloy-4 specimens were not exposed to LOCA fluid 
conditions and that only steam was applied at very low velocities for 
the main test series. The petitioner states that there was no 
documented heat transfer from the Zircaloy surface to the slow-flowing 
steam and that as a result the conditions of the small-scale laboratory 
tests were not typical of the complex thermal-hydraulic conditions that 
prevail during a LOCA.
    The petitioner suggests that without liquid water, the tests are 
invalid. The NRC disagrees. The presence of liquid water would 
invalidate the tests. Accurate steady-flow measurement would be 
extremely difficult. The droplets or liquid film would make it 
difficult to achieve the relatively constant sample temperatures that 
are necessary in these reaction kinetics tests. However, adequate steam 
flow is a concern. If the flow is too low, the reaction becomes steam 
starved. Otherwise, it is unnecessary to have steam flow typical of 
LOCA/ECCS conditions. These are not heat transfer tests. Once a 
reaction rate model is developed using data from experiments like 
these, the model should be validated against transient tests under LOCA 
conditions, as in the four Zircaloy tests described in WCAP-7665 and 
the transient tests described in the Cathcart-Pawel report.
    Calculations were performed to assure that there was adequate steam 
flow for the MiniZWOK experiments used to derive the Cathcart-Pawel 
correlation in NUREG-17. These calculations are described in the RES 
technical study.
    An important argument for the absence of steam starvation is how 
the isothermal Cathcart-Pawel experiments

[[Page 52897]]

described in NUREG-17 give consistent results that support the 
parabolic/Arrhenius behavior. This is also discussed in the RES 
technical study.
    Much of the petitioner's criticism of the Cathcart-Pawel work is 
related to a comparison of MiniZWOK and MaxiZWOK experimental 
conditions. MiniZWOK was used to develop a consistent set of data for 
correlation development. Controlling sample temperature by adjusting 
heater power (MiniZWOK) was much more successful than adjusting steam 
flow (MaxiZWOK). As the petitioner notes, temperature overshoot was a 
problem with MaxiZWOK and at high temperatures could have led to 
temperature runaway. As noted previously, temperature control is 
absolutely necessary in reaction kinetics experiments such as these. 
The petitioner implies that the experimenters abandoned MaxiZWOK in 
favor of MiniZWOK. Actually, the isothermal MiniZWOK experiments were 
essentially complete before the MaxiZWOK experiments were begun. 
Results from MaxiZWOK between 1652 [deg]F and 1832 [deg]F agreed well 
with MiniZWOK data at the same temperatures. Cathcart and Pawel state 
that:

    The very good agreement between these two data sets is regarded 
as evidence that steam flow rate and steam insertion temperature do 
not affect significantly the kinetics of the steam oxidation of 
Zircaloy, at least in this temperature range.

    Certainly, with steam velocities at least an order of magnitude 
greater in MaxiZWOK than MiniZWOK, the potential for more rapid gas 
phase diffusion of steam to the sample surface ``mass transfer'' is 
greater for MaxiZWOK. But clearly this is not the limiting phenomenon. 
This was demonstrated by the good agreement between MiniZWOK and 
MaxiZWOK data and the good agreement of MiniZWOK data to parabolic/
Arrhenius behavior. There is no evidence to suggest that high ``mass 
transfer coefficients'' in MaxiZWOK caused temperature overshoot in 
MaxiZWOK at 1832[deg]F, as the petitioner proposes. It is true, as the 
petitioner suggests, that ``[i]t is not possible to achieve an 
isothermal rate of oxidation of Zircaloy-4 if the Zircaloy-4 is exposed 
to LOCA fluid conditions at elevated conditions,'' but not for the 
reasons postulated by the petitioner. Rather, large-break LOCA reflood 
conditions are characterized by constantly decreasing power (decay 
heat) and increasing heat transfer coefficients after a few seconds. 
Under these conditions, isothermal conditions are impossible. WCAP-7665 
showed that this kind of heat transfer and power behavior was universal 
for all tests done under design basis conditions, and as a result, 
these heat transfer tests did not exhibit isothermal cladding 
temperature behavior.
    The petitioner implies that Cathcart and Pawel's statement, that 
scoping tests on the effect of steam pressure were in progress, is an 
admission of inapplicability of their work. On the contrary, the 
scoping work was completed and subsequent work by others has been 
undertaken to examine pressure effects. The petitioner's notion that 
the authors' statement about ongoing work applies to very low steam 
velocities is also unsupported.
    Work in this area did not end in 1977. The NRC, foreign partners, 
and the industry have continued to conduct and evaluate experimental 
and analytical programs on fuel cladding behavior. As in the case with 
many other research activities and their link to the agency's 
regulatory framework, an important objective of this work is the 
confirmation of current Sec.  50.46 criteria and models and the 
development of more realistic, performance-based, and contemporary 
criteria and models. An important link to the current work is the 
extensive research reported by Cathcart and Pawel.
    The NRC disagrees with the petitioner's assertion that the 
disclaimer in the introduction to NUREG-17 causes the technical work to 
be inapplicable to reactor regulation. The disclaimer protects the 
United States Government from potential litigation. It is not intended 
to discredit the technical validity of the work documented in NUREG-17. 
As such, the disclaimer is irrelevant to whether the NUREG-17 work is 
an adequate basis for reactor regulation. That is a question that 
should be decided solely on the technical merits of the work.
    The NRC found no technical basis in the petition nor in NRC records 
to support the assertion that the Regulatory Guide 1.157 conditions for 
acceptance of the use of ORNL/NUREG-17 information result in flawed 
evaluation of ECCS performance.

Issue 3: Need for Further Analysis of Appendix K Backup Data

    In Section 3.4 of his petition, the petitioner quotes from the AEC 
decision on the ECCS rulemaking [See Rulemaking Hearing, Acceptance 
Criteria for Emergency Core Cooling Systems for Light-Water Cooled 
Nuclear Power Reactors, RM-50-1, CLI-73-39, 6AEC1085, at 1124]: ``It is 
apparent, however, that more experiments with Zircaloy cladding are 
needed to overcome the impression left from run 9573.'' The petitioner 
claims that such experiments have not been performed and are necessary. 
The NRC disagrees.
    Run 9573 refers to one of four Zircaloy clad FLECHT experiments 
performed in 1969 and reported in WCAP-7665. The ``impression'' 
referred to by the AEC Commissioners in 1973 appears to be the fact 
that run 9573 indicates lower ``measured'' heat transfer coefficients 
than the other three Zircaloy clad tests reported in WCAP-7665 when 
compared to the equivalent stainless steel tests. This is not a concern 
about the zirconium-water reaction models. The AEC Commissioners 
believed that this anomaly could be cleared up with more experiments on 
Zircaloy cladding. Some of the anomaly can probably be explained by a 
deficiency in the data reduction process. As will be discussed later, 
additional Zircaloy clad tests were performed in the 1980s.
    Regarding the data reduction process, heat transfer coefficients 
are not directly measurable quantities. They must be calculated from 
measured temperatures, known heat sources, and known thermal 
properties. WCAP-7665 describes the heat transfer data reduction 
process using the DATAR code. For these experiments, the decay heat 
simulation was well known, as was the time of heater failure. However, 
the heat source, due to the zirconium-water reaction, had to be 
estimated in some way. The Baker-Just correlation was used for that 
purpose. Because of its conservatism, the Baker-Just correlation 
overestimates the amount of reaction and the associated heat generation 
rate. At 21 locations on 19 rods among the four Zircaloy tests, post-
test oxide thickness measurements were made. Westinghouse applied the 
Baker-Just correlation to each temperature transient measured at or 
very near to each oxide thickness measurement. The comparison between 
predicted and measured oxide thickness was presented in Figure B-12 of 
WCAP-7665. The Baker-Just calculated oxide thickness is about 1.6 times 
the measured value. Thus for this data set, the Baker-Just correlation 
overpredicts the data by about 60 percent, which is quite conservative.
    The NRC obtained tabular time/temperature data from Westinghouse 
for 19 of the 21 locations analyzed by Westinghouse for the four 
Zircaloy FLECHT tests. The Baker-Just correlation was applied to these 
19 data sets as a check on the analysis in WCAP-7665. The RES technical 
study clearly demonstrates that the analysis in WCAP-7665 is correct 
and that the

[[Page 52898]]

Baker-Just correlation is conservative even under the severe conditions 
of run 9573.
    The petitioner asserts that a detailed thermal-hydraulic analysis 
of run 9573, including evaluation of the heating from Zircaloy-water 
reactions, was never performed. Contrary to that assertion, not only 
was an evaluation of the heating from Zircaloy-water reaction performed 
for run 9573, it was done for all four Zircaloy tests. Unfortunately, 
using the conservative Baker-Just correlation to estimate the 
zirconium-water heat release results is an overestimation of the 
derived heat transfer coefficients. Thirty-five years later, it would 
be difficult to replicate the DATAR code, substitute a better metal-
water model, and re-derive the heat transfer coefficients. The 
difficulty would be in addition to the significant monetary expense of 
conducting high-temperature Zircaloy tests and would have marginal 
benefit in terms of increased understanding of large-break LOCA heat 
transfer and metal-water reaction kinetics. The current programs being 
conducted at Pennsylvania State University and Argonne National 
Laboratory are far more cost-effective.
    High-temperature tests similar to run 9573 would require rod bundle 
powers well outside the range of operation of any current or proposed 
pressurized water reactors (PWRs) and would produce very little useful 
heat transfer information. Therefore, the NRC does not believe that 
such tests are necessary.
    The petitioner states that more experiments with Zircaloy cladding 
have not been conducted on the scale necessary to overcome the 
impression left from run 9573. The NRC disagrees. In fact additional 
Zircaloy tests have been performed. In the early 1980s, the NRC 
contracted with National Research Universal (NRU) at Chalk River, 
Ontario, Canada to run a series of LOCA tests in the NRU reactor. More 
than 50 tests were conducted to evaluate the thermal-hydraulic and 
mechanical deformation behavior of a full-length 32-rod nuclear bundle 
during the heatup, reflood, and quench phases of a large-break LOCA. 
The NRC is reviewing the data from this program to determine its value 
for assessing the current generation of codes such as TRAC-M (now 
renamed TRACE).
    In assessing the need for further experiments like the Zircaloy-
clad FLECHT tests, it is important to understand the past and current 
role of rod bundle reflood heat transfer tests. In the late 1960s, a 
mechanistic understanding of reflood heat transfer did not exist. To 
develop heat transfer models as expeditiously as possible, the Atomic 
Energy Commission (AEC), Westinghouse, and Electric Power Research 
Institute (EPRI), cooperatively developed the PWR FLECHT program. The 
principal objective was to determine reflood heat transfer coefficients 
as a function of key initial and boundary conditions, rod elevation, 
and time after the beginning of reflood and to develop empirical 
correlations based on that dependency. As long as a sufficiently large 
matrix of tests was performed with full-scale rod bundles, there was no 
great need for a comprehensive mechanistic understanding. The key 
parameters were:

A. Pressure
B. Peak power
C. Decay power
D. Flooding rate
E. Inlet subcooling
F. Initial temperature
G. Bundle size
H. Cladding material
I. Housing temperature

    When nuclear plant behavior and design conditions are outside the 
envelope defined by these test parameters or the design of the 
experimental system, there is no basis for extrapolation, since the 
derived heat transfer models are not necessarily based on the physical 
models governing the reflood heat transfer processes. For the very 
empirical process used in the early FLECHT experiments, limited effort 
was expended obtaining data needed for development of mechanistic 
physical models. It would have been impractical to obtain sufficient 
Zircaloy heat transfer coefficient data for the empirical process used 
with the early FLECHT experiments.
    As the FLECHT program and other rod bundle reflood heat transfer 
programs have progressed over the last 30 years, more information 
appropriate for mechanistic model development has been obtained. As 
better mechanistic models are developed, careful extrapolation has a 
better chance of success, and the role of experiments like FLECHT has 
shifted from model development to developmental assessment. In fact, 
many of the FLECHT-SEASET experiments are used to assess the new code 
models. As mentioned previously, the NRC is reviewing the NRU Zircaloy-
clad nuclear fuel bundle test results to establish their value for 
further code assessment.

Conclusions

    The NRC investigated each of the petitioner's key concerns. The NRC 
concludes that Appendix K of 10 CFR Part 50 and the existing guidance 
on best-estimate ECCS evaluation models are adequate to assess ECCS 
performance for U.S. light water reactors (LWRs) using Zircaloy-clad 
UO<INF>2</INF> at burnup levels currently permitted by regulations. 
This general conclusion is based on the following considerations:
    The Baker-Just correlation using the current range of parameter 
inputs is conservative and adequate to assess Appendix K ECCS 
performance. Virtually every data set published since the Baker-Just 
correlation was developed has clearly demonstrated the conservatism of 
the correlation for the temperature range important to clad oxidation 
calculations for LOCAs.
    The parabolic/Arrhenius behavior of the Cathcart-Pawel isothermal 
experiments confirmed that there was adequate availability of steam. An 
NRC analysis confirms the ORNL/ANL assessment that the Cathcart-Pawel 
isothermal experiments were not steam starved by at least two orders of 
magnitude. Therefore, the experimental data is valid.
    NRC has continued to study complex thermal hydraulic effects on 
ECCS heat transfer processes during LOCA accident conditions consistent 
with Commission direction. As part of that initiative, the NRC funded 
more than 50 Zircaloy-clad nuclear fueled bundle reflood experiments at 
the NRU reactor. These experiments evaluated fuel rod and heat transfer 
behavior but did not include metallurgical examination to evaluate 
oxidation behavior. The NRC is continuing to conduct and evaluate 
experimental and analytical programs on fuel cladding behavior.
    The petitioner did not take into account Westinghouse's 
metallurgical analyses performed on the cladding for all four FLECHT 
Zircaloy-clad experiments reported in WCAP-7665. The petitioner also 
ignored the Westinghouse application of the Baker-Just correlation to 
these experiments, which had the ``complex thermal hydraulic 
phenomena'' deemed important by the petitioner. This application of the 
correlation to the metallurgical data clearly demonstrates the 
conservatism of the Baker-Just correlation for 21 typical temperature 
transients. The NRC also applied the Baker-Just correlation to the 
FLECHT Zircaloy experiments with nearly identical results, confirming 
the WCAP-7665 results.
    For the development of oxidation correlations, limited by oxygen 
diffusion into the metal, well-characterized isothermal tests are more 
important than the complex thermal hydraulics suggested by the 
petitioner.

[[Page 52899]]

The petitioner's suggested use of complex thermal-hydraulic conditions 
would be counter-productive in reaction kinetics tests because 
temperature control is required to develop a consistent set of data for 
correlation development. Isothermal tests allow this needed temperature 
control. It is more appropriate to apply the developed correlations to 
more prototypic transients (including complex thermal hydraulic 
conditions) to verify that the proposed phenomena embodied in the 
correlations are indeed limiting. This is what was done by Westinghouse 
in WCAP-7665, by Cathcart and Pawel in NUREG-17 and by the NRC in its 
technical safety analysis of PRM-50-76.
    The NRC applied the Cathcart-Pawel oxygen uptake and 
ZrO<INF>2</INF> thickness equations to the four FLECHT Zircaloy 
experiments, confirming the best-estimate behavior of the Cathcart-
Pawel equations for large-break LOCA reflood transients.
    Cathcart and Pawel applied their oxide thickness equation, using 
the BILD5 program, to 15 of their transient temperature experiments as 
described in ORNL/NUREG-17. The results showed that the correlation, 
based on numerous isothermal experiments, was conservative or best-
estimate when applied to this transient data set.

Petitioner's Public Comments

    The petitioner submitted two public comment letters in which he 
again asserted that the Baker-Just and Cathcart-Pawel equations are 
grossly misapplied by the NRC. The first comment letter basically 
repeated the arguments in the petition. No new technical information 
was supplied. The second comment letter introduced the issue of severe 
fouling, which was the subject of PRM-50-78 and addressed by the 
staff's evaluation of that petition for rulemaking. Other issues 
addressed in the second letter are related to the issues already 
discussed in this document, and therefore, no further response is 
necessary.

Reasons for Denial

    For the reasons cited in this document, the Commission is denying 
the petition for rulemaking (PRM-50-76) submitted by Mr. Robert Leyse. 
The NRC believes that the requested rulemaking would not make a 
significant contribution to maintaining safety because current 
regulations and regulatory guidance already adequately address the 
evaluation of performance of the ECCS. No data or evidence was provided 
by the petitioner or found in NRC records to suggest that the research, 
calculation methods, or data used to support ECCS performance 
evaluations were sufficiently flawed so as to create significant safety 
problems. NRC's technical safety analysis demonstrates that current 
procedures for evaluating performance of ECCS are based on sound 
science and that no amendments to the NRC's regulations and guidance 
documents are necessary. Additionally, the petitioner has not shown, 
nor has the NRC found, the existence of any safety issues regarding 
calculation methods or data used to support ECCS performance 
evaluations that would compromise the secure use of licensed 
radioactive material. The proposed revisions would not improve 
efficiency, effectiveness, and realism because licensees and the NRC 
would be required to generate additional information (as part of the 
evaluation of ECCS performance) that has no safety value and does not 
significantly improve realism.

    Dated at Rockville, Maryland, this 26th day of August, 2005.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 05-17589 Filed 9-2-05; 8:45 am]
BILLING CODE 7590-01-P


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