United States Nuclear Regulatory Commission

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No. 96-97 FOR IMMEDIATE RELEASE

(Tuesday, July 9, 1996)

NOTE TO EDITORS:

The Nuclear Regulatory Commission has received the attached

report from its Advisory Committee on Reactor Safeguards. The

report, in the form of a letter, provides comments on the NRC's

severe accident research program.

In addition, the NRC's executive director for operations has

received a letter report on a draft Regulatory Guide, DG-1047,

"Standard Format and Content for Applications to Renew Nuclear

Power Plant Operating Licenses."

#

Attachments:

As stated
June 28, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: SEVERE ACCIDENT RESEARCH

During the 432nd meeting of the Advisory Committee on Reactor

Safeguards, April 11-13, 1996, we completed our review of the

status of the NRC severe accident research program and severe

accident codes. Our Subcommittee on Severe Accidents held

meetings on these matters on March 1 and April 8, 1996. During

this review, we had the benefit of discussions with

representatives of the NRC staff and of the documents referenced.

Conclusions and Recommendations

1. Severe accident research provides information essential to

the development of risk-informed regulation.

2. Severe accident research provides the basis for evaluating

severe accident management strategies.

3. The NRC nuclear safety research program budget continues to

decline, and various research efforts are being reduced or

eliminated. Periodic analysis should be performed to assure

that the remaining severe accident research efforts are

focused on topics that have the greatest impact on risk and

the associated uncertainties. Criteria should be developed

for determining when programs have met their objectives.

4. Results of the severe accident research have shown that

there is no threat of prompt containment failure posed by

direct containment heating (DCH) in Westinghouse large dry

containments, alpha-mode steam explosions, and Mark I liner

melt-through. Research should continue to:

ˇ determine the impact of DCH on other containment types,

ˇ develop codes to better model the hydrogen

stratification and detonation,

ˇ determine the impact of ex-vessel steam explosions on

the BWR containments,

ˇ understand the phenomenological aspects associated with

molten debris coolability,
ˇ determine the impact of fuel coolant interaction on

lower head failure, and

ˇ determine the threats posed to steam generator tubes by

the natural circulation induced by the core degradation

processes.

5. Quantification of uncertainties is essential to riskinformed regulation. The NUREG-1150 effort contributed

significantly to the method for quantification of

uncertainties. Additional effort is needed to improve

understanding and quantification of phenomenological

uncertainties and their impact on Level 2 PRA results. We

plan to provide more specific recommendations in this area

in the future, as needed.

6. The assurance of the availability of specialized experts to

advise the Commission is sometimes a tacit motivation for

planning research programs. We believe that such assurance

is prudent and should be explicitly recognized as a

criterion in the funding of research.

Discussion

We believe it is important that the staff periodically perform

top down assessments of research to assure that the work supports

top level objectives, to review priorities, and to identify

research efforts that have reached maturity and perhaps should be

discontinued. In our view, severe accident research should have

the following top-level objectives:

 support assessments of severe accident risk from operating

plants,

 provide a technical basis for reviewing accident management

procedures,

 support the development of risk-informed regulation, and

 provide a technical basis for evaluating advanced plant

designs and operational features.

Better Level 2 PRAs are needed to reduce the uncertainties

associated with the assessment of the risk to public health and

safety. Severe accident research provides the bases for

improving Level 2 PRAs, many of which have used unnecessarily

simplistic models for severe accident behavior. Severe accident

research is needed to reduce the presently large uncertainties in

risk assessment results that are inimical to making sound

regulatory decisions.

The processes that lead to early failure of containment are of

particular importance to risk. Among such processes are DCH,
fuel coolant interactions, alpha-mode steam explosions, hydrogen

detonations, direct contact of core debris with containment

structures, and steam generator tube ruptures. Additional

assessment of DCH is needed for CE, B&W, and ice condenser

containments, and for BWRs. Although it appears that large dry

containments and containments with igniters can accommodate

hydrogen combustion without failing, we believe that

stratification and the potential for local detonation needs

additional investigation.

The extent to which debris can be cooled can be pivotal in

determining the likelihood of containment liner failure and longterm containment basemat melt-through. Viable criteria for

coolability of molten debris either in-vessel or ex-vessel have

not yet been developed.

A possible disadvantage of successful in-vessel debris cooling is

the potential failure of the reactor coolant system or steam

generator tubes caused by overheating from the convection of hot

gases. Steam generator tube ruptures that might occur as a

consequence of, or coincident with, a severe accident would

provide a direct path for radionuclide release from the reactor

core to the environment. The NRC and industry are addressing

this issue, but we believe additional thermal hydraulic and

radionuclide transport code development will be required for

resolution. The present NRC codes are not capable of assessing

this situation.

Currently, significant information in the severe accident area is

being developed in international cooperative programs. While we

fully support the bilateral agreements and the Cooperative Severe

Accident Research Program (CSARP), it is important for NRC that

its domestic contractors maintain capability in this area. Staff

and contractors who are knowledgeable of the physics and

technology of severe accident phenomena will be needed to resolve

complex issues in this area, to enhance the regulatory process,

and to provide technical support in the event of a real accident.

Dr. Dana A. Powers did not participate in the Committee's

deliberation regarding this matter.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

References:

1. Report dated August 18, 1992, from David A. Ward, Chairman,

ACRS, to Ivan Selin, Chairman, NRC, Subject: Severe

Accident Research Program Plan
2. U. S. Nuclear Regulatory Commission, SECY-95-004, dated

January 4, 1995, from James M. Taylor, Executive Director

for Operations, NRC, for the Commissioners, Subject: Status

of Implementation Plan for Closure of Severe Accident

Issues, Status of the Individual Plant Examinations and

Status of Severe Accident Research

3. U. S. Nuclear Regulatory Commission, NUREG/CR-6109, "The

Probability of Containment Failure by Direct Containment

Heating in Surry," May 1995

4. Nuclear Energy Institute, NEI 91-04, Revision 1, "Severe

Accident Issue Closure Guidelines," December 1994

5. Report (undated) by F. Cheung and K. Haddad, Pennsylvania

State University, Subject: Steady-State Observations and

Theoretical Modeling of Critical Heat Flux Phenomena on a

Downward Facing Hemispherical Surface

6. Sandia National Laboratories Letter Report, "Scaling and

Design Report for Lower Head Failure Experiments," May 1995

7. Secretary-General of the OECD Report, Senior Group of

Experts on Severe Accident Management (SESAM), "Severe

Accident Management Implementation," October 1995

8. Secretary-General of the OECD Draft Report, "Nuclear Safety

Research in OECD Countries, Areas of Agreement, Areas For

Further Action, Increasing Need For Collaboration," November

1995

9. Proceedings of the Specialist Meeting On Severe Accident

Management Implementation, held at Niantic, Connecticut, on

June 12-14, 1995, by the Committee on the Safety of Nuclear

Installations, OECD Nuclear Energy Agency
June 18, 1996

Mr. James M. Taylor

Executive Director for Operations

U. S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Mr. Taylor:

SUBJECT: DRAFT REGULATORY GUIDE DG-1047, "STANDARD FORMAT AND

CONTENT FOR APPLICATIONS TO RENEW NUCLEAR POWER PLANT

OPERATING LICENSES"

During the 432nd meeting of the Advisory Committee on Reactor

Safeguards, June 12-14, 1996, we discussed the subject draft

Regulatory Guide with representatives of the NRC staff and the

Nuclear Energy Institute. We also had the benefit of the

documents referenced.

We have no objection to the staff proposal to issue the draft

Regulatory Guide for public comment. We plan to review the

proposed final version of this Guide after reconciliation of the

public comments.

Dr. William J. Shack did not participate in the Committee's

deliberations regarding this matter.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

References:

1. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide

DG-1047, "Standard Format and Content for Applications to

Renew Nuclear Power Plant Operating Licenses," transmitted

by memorandum dated April 18, 1996, from Scott F. Newberry,

Office of Nuclear Reactor Regulation, NRC, to John T.

Larkins, ACRS

2. Nuclear Energy Institute, NEI 95-10 (Revision 0), "Industry

Guideline for Implementing the Requirements of 10 CFR Part

54The License Renewal Rule," March 1996

3. U.S. Nuclear Regulatory Commission, SECY-96-059 dated March

18, 1996, from James M. Taylor, Executive Director for

Operations, NRC, for the Commissioners, "Activities

Associated with the Implementation of 10 CFR Part 54"