This section of the FEDERAL REGISTER

contains regulatory documents having general

applicability and legal effect, most of which

are keyed to and codified in the Code of

Federal Regulations, which is published under

50 titles pursuant to 44 U.S.C. 1510.

The Code of Federal Regulations is sold by

the Superintendent of Documents. Prices of

new books are listed in the first FEDERAL

REGISTER issue of each week.

Rules and Regulations Federal Register

52893 

Vol. 70, No. 171 

Tuesday, September 6, 2005 

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-ofcoolant 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 pdr@nrc.gov. 

Document PDR Web ADAMS 

Federal Register Notice—Receipt of Petition for Rulemaking (67 FR 51783; Aug. 9, 2002) .............................. X X ML022800472 

Letter of Denial to the Petitioner ............................................................................................................................. X X ML052220454 

Penn State/US NRC ‘‘Rod Bundle Test Facility and Reflood Heat Transfer Program’’ ........................................ ML023040657 

Petition for Rulemaking (PRM–50–76) ................................................................................................................... X X ML022240009 

Public Comments for PRM–50–76 ......................................................................................................................... X X ML042740105 

US NRC Office of Nuclear 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’’.X X ML041210109 

US NRC, ‘‘Updated Program Plan for High-Burnup Light-Water Reactor Fuel’’ ................................................... .......... .......... ML031810103 

Studies of Metal Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium-Water Reaction,’’ L. Baker and L.C. Just, ANL–6548 (May 1962)........... .......... ML050550198 

PWR FLECHT (Full Length Emergency Cooling Heat Transfer) Final Report,’’ April 1971 .................................. .......... .......... ML052230221 

Zirconium Metal-Water Oxidation Kinetics IV. Reaction Rate Studies,’’ ORNL/NUREG–17, August 1977. ......... .......... .......... ML052230079 

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 TAR@nrc.gov. 

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 

52894 Federal Register / Vol. 70, No. 171 / Tuesday, September 6, 2005 / Rules and Regulations 

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: 

• The bulk water temperature was no 

greater than 315 °C (599 °F). 

• 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. 

• Zirconium specimens were exposed 

to water only, while LOCA conditions 

include steam and nonequilibrium 

water-steam mixtures that reached 

higher bulk fluid temperatures. 

• 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 °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: 

• Zircaloy-4 specimens exposed only 

to steam, rather than fluid conditions as 

present in a LOCA. 

• No documented heat transfer from 

the Zircaloy surface to the slow-flowing 

steam. 

• Small-scale laboratory testing 

without conditions typical of the 

complex thermal-hydraulic conditions 

that prevail during a LOCA. 

• An unexplained shift from the 

MaxiZWOK (testing apparatus for 

investigations in the temperature range 

1652 °F to 1832 °F) to the MiniZWOK 

(a different testing apparatus for 

investigations in the temperature range 

1832 °F to 2734 °F). 

The petitioner believes that the 

investigators’ conclusions include a 

statement that ‘‘overlooks the very 

substantially greater mass transfer 

coefficients that accompany the socalled 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 °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 

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

52896 Federal Register / Vol. 70, No. 171 / Tuesday, September 6, 2005 / Rules and Regulations 

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 °F) for one or two seconds are not 

typical of fuel rod cladding at 

temperatures in the range of 1800 °F–

2200 °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 °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 § 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 °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 zirconiumwater reaction kinetics at very high 

temperatures. 

One interesting feature of the BakerJust 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 temperaturesensitive 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 

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

°F and 1832 °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°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 § 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 

52898 Federal Register / Vol. 70, No. 171 / Tuesday, September 6, 2005 / Rules and Regulations 

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 zirconiumwater 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 hightemperature 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 largebreak 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 UO2 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 BakerJust 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, wellcharacterized isothermal tests are more 

important than the complex thermal 

hydraulics suggested by the petitioner. 

Federal Register / Vol. 70, No. 171 / Tuesday, September 6, 2005 / Rules and Regulations 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 ZrO2 thickness 

equations to the four FLECHT Zircaloy 

experiments, confirming the bestestimate 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 

DEPARTMENT OF TRANSPORTATION 

Federal Aviation Administration 

14 CFR Part 39 

[Docket No. FAA–2004–18877; Directorate 

Identifier 2002–NM–340–AD; Amendment 

39–14248; AD 2005–18–08] 

RIN 2120–AA64 

Airworthiness Directives; Boeing 

Model 737–100, –200, –200C, and –300 

Series Airplanes 

AGENCY: Federal Aviation 

Administration (FAA), Department of 

Transportation (DOT). 

ACTION: Final rule. 

SUMMARY: The FAA is adopting a new 

airworthiness directive (AD) for certain 

Boeing Model 737–100, –200, –200C, 

and –300 series airplanes. This AD 

requires repetitive detailed inspections 

to detect discrepancies of the retaining 

pin lugs on the support fitting of the 

main landing gear (MLG) beam, and 

rework of the support fitting, or 

replacement of the fitting if necessary. 

This AD is prompted by reports of 

discrepancies of the lugs. We are issuing 

this AD to prevent separation of the 

support beam of the MLG from the rear 

spar, which could cause cracking of the 

MLG support fitting and a consequent 

leak in the wing fuel tank or collapse of 

the MLG. 

DATES: This AD becomes effective 

October 11, 2005. 

The incorporation by reference of 

certain publications listed in the AD is 

approved by the Director of the Federal 

Register as of October 11, 2005. 

ADDRESSES: For service information 

identified in this AD, contact Boeing 

Commercial Airplanes, P.O. Box 3707, 

Seattle, Washington 98124–2207. 

Docket: The AD docket contains the 

proposed AD, comments, and any final 

disposition. You can examine the AD 

docket on the Internet at http:// 

dms.dot.gov, or in person at the Docket 

Management Facility office between 9 

a.m. and 5 p.m., Monday through 

Friday, except Federal holidays. The 

Docket Management Facility office 

(telephone (800) 647–5227) is located on 

the plaza level of the Nassif Building at 

the U.S. Department of Transportation, 

400 Seventh Street, SW., room PL–401, 

Washington, DC. This docket number is 

FAA–2004–18877; the directorate 

identifier for this docket is 2002–NM–

340–AD. 

FOR FURTHER INFORMATION CONTACT: 

Robert C. Hardwick, Aerospace 

Engineer, Airframe Branch, ANM–120S, 

FAA, Seattle Aircraft Certification 

Office, 1601 Lind Avenue, SW., Renton, 

Washington 98055–4056; telephone 

(425) 917–6457; fax (425) 917–6590. 

SUPPLEMENTARY INFORMATION: The FAA 

proposed to amend 14 CFR part 39 with 

an AD for certain Boeing Model 737–

100, –200, –200C, and –300 series 

airplanes. That action, published in the 

Federal Register on August 17, 2004 (69 

FR 51017), proposed to require 

repetitive detailed inspections to detect 

discrepancies of the retaining pin lugs 

on the support fitting of the main 

landing gear (MLG) beam, and rework of 

the support fitting or replacement of the 

fitting if necessary. 

Comments 

We provided the public the 

opportunity to participate in the 

development of this AD. We have 

considered the comments that have 

been submitted on the proposed AD. 

Agreement With the Proposed AD 

One commenter, the manufacturer, 

agrees with the proposed AD. 

Conditional Agreement With the 

Proposed AD 

One commenter, an operator, agrees 

with the proposed AD provided that 

there are adequate parts available if the 

discrepant condition is found. 

The FAA agrees that adequate 

availability of parts is necessary. We