Information Notice 2012-17, Certified Material Test Reports and Age Hardened Concrete Compressive Strength
| ML121840075 | |
| Person / Time | |
|---|---|
| Issue date: | 09/06/2012 |
| From: | Camper L, Laura Dudes, Mark Lombard, Mcginty T NRC/FSME/DWMEP, Office of Nuclear Material Safety and Safeguards, Division of Construction Inspection and Operational Programs, Office of Nuclear Reactor Regulation |
| To: | |
| Purnell, B A, NRR/DPR, 415-1380 | |
| References | |
| IN-12-017 | |
| Download: ML121840075 (6) | |
ML121840075 UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS
OFFICE OF FEDERAL AND STATE MATERIALS
AND ENVIRONMENTAL MANAGEMENT PROGRAMS
WASHINGTON, DC 20555-0001
September 6, 2012
NRC INFORMATION NOTICE 2012-17:
INAPPROPRIATE USE OF CERTIFIED
MATERIAL TEST REPORT YIELD STRESS AND
AGE-HARDENED CONCRETE COMPRESSIVE
STRENGTH IN DESIGN CALCULATIONS
ADDRESSES
All holders of an operating license or construction permit for a nuclear power reactor issued
under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of
Production and Utilization Facilities, including those who have permanently ceased operations
which have spent fuel in storage in spent fuel pools.
All holders of or applicants for combined licenses issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
All holders of and applicants for an independent spent fuel storage installation license under
10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of issues identified during recent NRC inspections regarding the design of seismic
Category I or safety-related structures. The NRC expects that recipients will review the
information for applicability to their facilities and consider actions, as appropriate, to avoid
similar problems. However, suggestions contained in this IN are not NRC requirements;
therefore, no specific action or written response is required.
DESCRIPTION OF CIRCUMSTANCES
Inappropriate Use of Certified Material Test Report (CMTR) Yield Stress
Monticello Nuclear Generating Plant
During an inspection at the Monticello Nuclear Generating Plant, NRC inspectors reviewed a
calculation for a service water pipe support that the licensee used to demonstrate seismic
Category I compliance. The safety-related design function of the service water pipe support
structure is to hold and maintain the service water discharge line in position during a seismic
Category I design-basis event to meet internal flooding license requirements. The licensees updated safety analysis report (USAR) requires seismic Class 1 structural steel to be analyzed
in accordance with the American Institute of Steel Construction (AISC) specification. The AISC
specification requires the allowable bending stress of the structural element to be based on an
AISC-specified numerical value multiplied by the specified minimum yield stress. Instead, the
calculation incorrectly evaluated acceptability of the baseplate component based on the material
yield stress documented in a CMTR.
Additional information is available in Monticello Nuclear Generating Plant, NRC Component
Design Bases Inspection Report 05000263/2009007, dated January 6, 2010, which can be
found on the NRCs public Web site using the Agencywide Documents Access and
Management System (ADAMS) at Accession No. ML100060183.
Fermi Power Plant
During an inspection at the Fermi Power Plant, NRC inspectors reviewed design calculations for
the reactor building steel structure. The licensees updated final safety analysis report (UFSAR)
specifies that the reactor building steel structure is to be designed to the acceptance limits
specified in the AISC specification to demonstrate seismic Category I compliance. The AISC
specification requires the allowable stress for axial compressive, bending, shear, and torsion to
be based on the specified minimum yield stress of the structural element material. Instead, the
licensee incorrectly used actual material yield stress from CMTRs for evaluating the reactor
buildings horizontal and vertical bracing steel members, girts, crane runway girder, building
columns, and column anchor plate.
Additional information is available in Fermi Power Plant, Unit 2 Integrated Inspection Report 05000341/2011002, dated May 2, 2011 (ADAMS Accession No. ML111220240).
Braidwood Nuclear Generating Station
During an inspection at the Braidwood Nuclear Generating Station, NRC inspectors reviewed a
design calculation for a safety injection pipe support that is supported from an embedment plate
inside the auxiliary building, which is a seismic Category I structure. The licensees UFSAR
specified that the auxiliary building steel structure is designed to the AISC specification. The
AISC specification requires the allowable bending stress of the structural element to be based
on a specified value multiplied by the material specified minimum yield stress. Instead, the
licensee incorrectly used a CMTR for design-basis input to the calculation of allowable bending
stress in the design-basis calculation.
Additional information is available in Braidwood Station, Units 1 and 2, Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications Baseline Inspection Report 05000456/2011008; 05000457/2011008, dated October 28, 2011 (ADAMS Accession
No. ML11301A260).
Inappropriate Use of Age-Hardened Concrete Compressive Strength
Monticello Nuclear Generating Plant
During an inspection at the Monticello Nuclear Generating Plant, NRC inspectors reviewed a
pipe support calculation for the main steam piping. The safety-related design function of the
pipe support is to hold a section of the main steam line in position during a seismic Category I
design-basis event and maintain primary containment isolation. The licensees USAR specified that concrete structures are designed in accordance with the American Concrete Institute
(ACI) 318 code. The design specification for anchor bolts requires the allowable load for anchor
bolts to be based on a concrete compressive strength established in accordance with the
ACI 318 code. Specifically, a concrete compressive strength of 3,000 pounds per square inch
(psi) or 4,000 psi, depending on the plant building structure where the anchorage is located, was to be used in design calculations. Instead, the licensee used an anchor bolt with an
allowed load incorrectly determined using a concrete compressive strength of 6,000 psi. This
value was determined by extrapolation using the principle that concrete continues to age harden
over time.
Additional information is available in Monticello Nuclear Generating Plant, NRC Component
Design Bases Inspection Report 05000263/2009007, dated January 6, 2010 (ADAMS
Accession No. ML100060183).
Fermi Power Plant
During an inspection at the Fermi Power Plant, NRC inspectors reviewed a calculation for
reactor building concrete structural members used to support spent fuel cask placement. The
licensees UFSAR specifies that the requirements for the reactor building concrete structural
members used to support spent fuel cask placement be designed to the acceptance limits
specified in ACI 318 to demonstrate seismic Category I compliance. The ACI 318 code requires
the concrete compressive strength to be based on the 28-day strength or another time as
indicated in design drawings or specifications. The design drawings and the UFSAR specified
the compressive strength 28-day value for the reactor building as 4,000 psi. Instead, the
licensee incorrectly used a concrete compressive strength of 5,900 psi for evaluation of the
reactor building refuel floor slab and beams, the spent fuel pool floor slab, and the reactor
building truck bay floor slab. This value was determined by extrapolation using the principle that
concrete continues to age harden over time.
Additional information is available in Fermi Power Plant, Unit 2 Integrated Inspection Report 05000341/2011003; 07200071/2010001, dated August 1, 2011 (ADAMS Accession
No. ML112140118).
DISCUSSION
General Design Criterion 2, Design Bases for Protection Against Natural Phenomena, of
Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, states, in
part, that structures, systems, and components important to safety shall be designed to
withstand the effects of natural phenomena, including earthquakes. In addition, the design
bases for structures, systems, and components shall reflect sufficient margin for the limited
accuracy, quantity, and period of time in which the historical data have been accumulated.
Licensees typically use industry standards with sufficient margin to establish a seismic design
basis. The AISC specifications use the specified minimum yield strength for stress evaluation of
structural components. CMTRs ensure that purchased structural steel has the minimum
strength specified in the design. The ACI 318 code uses a concrete compressive strength
based on the 28-day strength or another time as indicated on the design drawings and
specifications. Concrete cylinder tests provide assurance that the installed concrete will achieve
the minimum strength specified in the design. The AISC specification provides acceptance limits for structural steel design using the material
specified minimum yield strength to determine the allowable axial, bending, shear, and torsional
stresses for steel members. The NRC staff has approved the use of AISC in the design and
licensing bases of nuclear power plants. The use of actual yield stress is not consistent with
this specification.
The ACI 318 code provides acceptance limits for concrete design using the compressive
strength of the concrete specified at the time of construction. The concrete compressive
strength for installed concrete is typically 28 days or another time as indicated in design
drawings or specifications. The NRC staff has approved the use of ACI 318 in the design and
licensing bases of the plant. The use of concrete compressive strength based on the principle
that concrete continues to age harden over time is not consistent with this code.
The use of material yield stresses and concrete compressive strengths in structural design
calculations that are less conservative than specified in the plants design and licensing basis
result in inappropriate reduction in safety margins inherent in the associated specification and
code requirements. The safety implication of using CMTR values is the reduction in the AISC
factor of safety or safety margin. The safety implication of using age-hardened concrete
compressive strength is the reduction of the ACI 318 code factor of safety or safety margins.
Licensees should verify that their calculations for structural steel and concrete use allowables
that comply with the standards referenced in their licensing and design bases. Deviations from
these allowables may require a license amendment prior to use. Specifically, Criterion III,
Design Control, of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants, to 10 CFR Part 50 requires, in part, that measures shall be established to
include provisions to ensure that appropriate quality standards are specified and included in
design documents and that deviations from such standards are controlled.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or the appropriate NRC project manager.
/RA/
/RA/
Timothy J. McGinty, Director
Larry W. Camper, Director
Division of Policy and Rulemaking
Division of Waste Management
Office of Nuclear Reactor Regulation
and Environmental Protection
Office of Federal and State Materials and
Environmental Management Programs
/RA by JLuehman for/
/RA/
Laura A. Dudes, Director
Mark D. Lombard, Director
Division of Construction Inspection
Division of Spent Fuel Storage
and Operational Programs
and Transportation
Office of New Reactors
Office of Nuclear Material Safety and Safeguards
Technical Contacts: John V. Bozga, Region III
James Neurauter, Region III
630-829-9613
630-829-9828 E-mail: John.Bozga@nrc.gov
E-mail: James.Neurauter@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or the appropriate NRC project manager.
/RA/
/RA/
Timothy J. McGinty, Director
Larry W. Camper, Director
Division of Policy and Rulemaking
Division of Waste Management
Office of Nuclear Reactor Regulation
and Environmental Protection
Office of Federal and State Materials and
Environmental Management Programs
/RA by JLuehman for/
/RA/
Laura A. Dudes, Director
Mark D. Lombard, Director
Division of Construction Inspection
Division of Spent Fuel Storage
and Operational Programs
and Transportation
Office of New Reactors
Office of Nuclear Material Safety and Safeguards
Technical Contacts: John V. Bozga, Region III
James Neurauter, Region III
630-829-9613
630-829-9828 E-mail: John.Bozga@nrc.gov
E-mail: James.Neurauter@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
ADAMS Accession Number: ML121840075
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