02-2075. Robert H. Leyse; Supplement to a Petition for Rulemaking

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Supplemental petition for rulemaking; notice of receipt.

SUMMARY:

The Nuclear Regulatory Commission (NRC) has received and requests public comment on a supplement to his original petition for rulemaking (PRM-50-73) filed with the Commission by Robert H. Leyse. The supplemental petition was docketed by the Commission and has been assigned Docket No. PRM-50-73A. The petitioner requests, in this supplement to his earlier petition, that the NRC amend its regulations on the acceptance criteria for emergency core cooling systems for light-water nuclear power reactors to address the impact of severe crud deposits on fuel bundle coolability during normal operation of a light-water-reactor (LWR).

DATES:

Submit comments by April 15, 2002. Comments received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date.

ADDRESSES:

Submit written comments to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff. Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4:15 p.m. Federal workdays.

For a copy of the petition, write to Michael T. Lesar, Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

You may also provide comments via the NRC's interactive rulemaking Web site at http://ruleforum.llnl.gov. This site provides the capability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking Web site, contact Ms. Carol Gallagher, 301-415-5905 (e-mail: [e-mail]).

The petition and copies of comments received may be inspected and copied for a fee at the NRC Public Document Room, 11555 Rockville Pike, Public File Area O1F21, Rockville, Maryland. Copies of comments received are also available through the NRC's Agencywide Documents Access and Management System (ADAMS), which provides text and image files of NRC's public documents. These documents may be accessed through the NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/​NRC/​ADAMS/​index.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737or by e-mail to [e-mail].

FOR FURTHER INFORMATION CONTACT:

Michael T. Lesar, Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: [phone] or Toll Free: 800-368-5642.

SUPPLEMENTARY INFORMATION:

Background

The NRC received a petition for rulemaking dated September 4, 2001, submitted by Mr. Robert H. Leyse, on his own behalf. The petition was docketed as PRM-50-73 on September 6, 2001. The notice of receipt of this petition was published on October 12, 2001, (66 FR 52065). On November 5, 2001, the NRC received a supplement to PRM-50-73 submitted by Mr. Leyse. The supplement to the petition was assigned docket number PRM-50-73A.

In his original petition, the petitioner requested that the NRC amend its regulations on the acceptance criteria for emergency core cooling systems for light-water nuclear power reactors to address the impact of crud on cooling capability during a fast-moving, large-break, loss-of-coolant accident (LOCA).

The petitioner requested that elements in § 50.46 concerning comparisons to applicable experimental data, and the following paragraphs in Appendix K to part 50, be revised to include the impact of crud deposits on fuel pins:

I.B. Swelling and Rupture of the Cladding and Fuel Rod Thermal Parameters;

I.C.2 Frictional Pressure Drops;

I.C.4 Critical Heat Flux;

I.C.5 Post-CHF Heat Transfer Correlations;

I.C.7 Core Flow Distribution During Blowdown;

I.D.3 Calculation of Reflood Rate for Pressurized Water Reactors;

I.D.6 Convective Heat Transfer Coefficients for Boiling Water Reactor Fuel Rods Under Spray Cooling; and

I.D.7 The Boiling-Water Reactor Channel Box Under Spray Cooling.

II.1.a The documentation requirements in this paragraph should include a description of each evaluation model used for estimation of the effects of crud deposits on fuel pins.

The Petitioner's Request

In his supplemental petition (PRM-50-73A), the petitioner requests that the NRC revise its regulations on the acceptance criteria for emergency core cooling systems for light-water nuclear power reactors to address the impact severe crud buildup will have on core coolability during normal reactor operations.

The petitioner states that a certain licensed power reactor has operated with unusually heavy crud deposits within several fuel bundles. The petitioner states that these deposits were found and at least partially classified during a refueling outage. The petitioner believes that if these deposits had continued to build during normal reactor operation at power, the unusually heavy crud deposits would have become severe crud deposits. Blockage of the flow channels within the fuel bundles would likely have developed. The petitioner believes that severe crud deposits within the fuel bundles can lead to a loss of coolability with consequent overheating of zirconium cladding within the bundles, autocatalytic zirconium-water reactors of the fuel cladding, chemical reactions between the fuel cladding and uranium oxide fuel pellets, initiation of zirconium water reactions involving zirconium core structures such as fuel bundle spacer grids and channel boxes, melting of certain control element materials, melting of braze materials in certain fuel bundle spacer grids, metallurgical reactions between certain fuel bundle spacer grid springs and the zirconium cladding on the fuel pins, and additional sources of structural degradation. The petitioner states that these factors can initiate substantial and rapid localized core melting while the LWR is at power. The petitioner states that if the LWR is then shut down, the core meltdown may rapidly propagate among the fuel bundles and core structures with sequential and parallel destruction of the barriers that constitute defense in depth. Thus, the single entity, unusually heavy crud deposits on the fuel pins, might be only one step before unusually heavy crud deposits thicken and become severe crud deposits. The petitioner states that severe crud deposits then threaten the integrity of all of the barriers that in total constitute the defense in depth.

The petitioner states that performance-based experience reveals that when unusually heavy crud deposition on fuel bundles occurs during normal operation of an LWR, there are likely to be indications of fuel element cladding defects by increases in the offgas activity. However, the petitioner states that this increase in the offgas activity is not regarded as an indicator of a possible heavy crud deposition. The petitioner believes that an LWR may be operated within its Licensing Basis and the Technical Specifications until the transition from unusually heavy crud deposition to severe crud deposition is effected. The petitioner believes that at this point it is likely that rapid localized core melting will be initiated while the LWR is at power. The petitioner also believes that there will likely be delays (several seconds) before the LWR is shut down. The petitioner believes that by then the rapid propagation of the meltdown will likely be well underway and it will likely continue even though the LWR is shut down.

The petitioner requests that elements in § 50.46 and the following paragraphs in Appendix K to part 50, and perhaps other regulations, be revised to include the impact of crud deposits on the fuel bundles during normal operation:

I.B. Swelling and Rupture of the Cladding and Fuel Rod Thermal Parameters;

I.C.2 Frictional Pressure Drops;

I.C.4 Critical Heat Flux;

I.C.5 Post-CHF Heat Transfer Correlations;

I.C.7 Core Flow Distribution During Blowdown;

I.D.3 Calculation of Reflood Rate for Pressurized Water Reactors;

I.D.6 Convective Heat Transfer Coefficients for Boiling Water Reactor Fuel Rods Under Spray Cooling; and

I.D.7 The Boiling-Water Reactor Channel Box Under Spray Cooling.

II.1.a The documentation requirements in this paragraph should include a description of each evaluation model used for estimation of the effects of crud deposits on fuel pins.

Dated at Rockville, Maryland, this 22nd day of January 2002.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook,

Secretary of the Commission.