ML102250465
| ML102250465 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/13/2010 |
| From: | Sheikh A Division of License Renewal |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML102250464 | List: |
| References | |
| 50-282-LR, 50-306-LR, ASLBP 08-871-01-LR-BD01, RAS 18398 | |
| Download: ML102250465 (8) | |
Text
August 13, 2010 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
NORTHERN STATES POWER COMPANY ) Docket Nos. 50-282-LR/ 50-306-LR
)
(Prairie Island Nuclear Generating Plant, )
Units 1 and 2) )
NRC STAFF REBUTTAL TESTIMONY OF ABDUL H. SHEIKH AND DR. DAN J. NAUS CONCERNING THE SAFETY CULTURE CONTENTION AND REACTOR REFUELING CAVITY LEAKAGE Q1. Please state your name, occupation, and by whom you are employed.
A1(a). My name is Abdul H. Sheikh (Sheikh). I am employed as a Senior Structural Engineer in the Division of License Renewal (DLR), Office of Nuclear Reactor Regulation (NRR), U.S. Nuclear Regulatory Commission (NRC). A statement of my professional qualifications is attached to my July 30, 2010 pre-filed testimony.
A1(b). My name is Dr. Dan J. Naus (Naus).1 I am employed as a Distinguished Research Staff Member at Oak Ridge National Laboratory (ORNL) by UT-Battelle, LLC. A statement of my professional qualifications is attached to my July 30, 2010 pre-filed testimony.
Q2. What is the purpose of your testimony?
A2. The purpose of our testimony is to respond to testimony and exhibits submitted by the Prairie Island Indian Community (PIIC) on July 30, 2010.
Q3. Mr. Grimes criticizes NSP for the length of time it has taken to identify the source of the refueling cavity leakage and to stop the leakage. Do you believe that the criticism is 1
In this testimony, the sponsors of each numbered response are identified by their last name; no such designation is provided for paragraphs which are sponsored by all witnesses.
valid?
A3. (Sheik) No. The size of the holes, the intermittent nature of the leakage, and the inaccessibility of the areas affected make it difficult to identify the source of the leakage and stop it. Based on the leakage rate, the source of the leakage consists of small holes (equivalent to one hole of 1/60th of an inch diameter). Detection of defects of this size is very difficult, even using state-of-the-art equipment and test methods. In addition, the leakage only occurs when the refueling cavity is flooded. This only happens approximately once every 18 months and then only for a few days. Also, the path of the leakage and the areas where the water accumulates are generally inaccessible.
During the 1987-1998 period, PINGP personnel performed repairs to the Unit 1 reactor cavity stainless steel plate welds. In 1998, the PINGP personnel performed a non-destructive examination of the Unit 2 reactor cavity liner plate, identified three pinhole leaks, and repaired the welds at these three locations. PINGP personnel also performed an engineering evaluation in 1998 to determine the effects of borated water on the steel containment and concrete structures. In 2002-2003, PINGP personnel sprayed a coating on the reactor cavity stainless steel liner plate. Leakage was mitigated when the coating was applied properly. During the period from 2004-2008, PINGP personnel applied caulk at the reactor internal stand embeds and this stopped the leakage when applied properly.
Given the small size of the holes, the intermittent nature of the leakage, and the inaccessibility of the portions of containment affected, it is understandable that it has taken time for NSP to identify and stop the leakage.
Q4. Mr. Grimes testifies that [t]he allowable containment leakage for a design basis accident is equivalent to a 0.003 square-inch hole in the containment (about one-sixteenth of an inch in diameter). He cites Inspection Manual Chapter 0609, Containment Integrity Significance Determination Process, Appendix H, U.S. Nuclear Regulatory Commission (May 6, 2004). Do you agree with his testimony and reference to the Inspection Manual?
A4. (Sheikh) No. The Inspection Manual Chapter 0609, Appendix H (PIIC Exhibit 5),
does not contain any reference to design basis accident containment leakage equivalent to a 0.003 square-inch (about one-sixteenth of an inch in diameter) hole. Appendix H states, The guidance in this SDP [significance determination process] is designed to provide NRC inspectors, SRAs [senior resident analysts], and NRC management with a simple, probabilistic risk framework for use in identifying which findings are potentially risk-significant from a LERF
[large early release frequency] perspective.
Q5. In A19 of his direct testimony, Mr. Grimes states that [i]f leakage from the refueling cavity stays in contact with the steel liner and concrete structure for an extended period, corrosion could eat through the containment liner and weaken the concrete structure to such an extent that, should an accident occur, the containment leakage could result in radiological exposures in excess of 10 CFR Part 100. Is there evidence of corrosion at PINGP as a result of the refueling cavity leakage?
A5. No. There is no evidence of corrosion as a result of the refueling cavity leakage.
During the period 2008 to 2009, PINGP removed grout in both units at Sump B and performed visual and ultrasound examinations of the containment vessels, including the area around the RHR pump suction lines. In all instances, the containment steel wall thickness measurements were at or above the nominal thickness of the plate, and there were no indications of corrosion or pitting of the containment vessels.
While in theory leakage can give rise to corrosion that could weaken containment, there is no evidence that this is happening at PINGP. Corrosion is minimized by several factors.
First, borated water leakage at PINGP has the potential to collect at the bottom head of the steel containment, which is located between thick concrete layers above and below it. The concrete above the containment steel ranges from 34 to 140 inches thick. The containment steel has a minimum thickness of 1.50 inch. The concrete layer below the steel containment is more than 60 inches thick. The concrete was poured so as to lie immediately next to the steel. The
containment steel is thus abutted by concrete, above and below. In the event the borated water leakage from reactor cavity is trapped and remains in contact with the steel containment, there is a potential for degradation of steel. However, the degradation (corrosion) is not likely to cause significant loss of steel containment thickness since it is not likely to be exposed to oxygen. The thick concrete layers above and below the steel containment at this location prevent the steel containment from exposure to air and thus prevent significant corrosion or pitting. Also, while the borated water is slightly acidic, it is in contact with the concrete (which is alkaline) and this contact reduces the acidity of the water, moving it toward a more neutral pH.
The effect of the concrete on the borated water thus further reduces the corrosive effect of the water.
Q6. In A19 of his direct testimony, Mr. Grimes asserts that the leaks could have potentially disastrous consequences for the Community and the populace of the surrounding area. Have you considered the worst case scenario involving the refueling cavity leakage?
A6. Yes. Even under the worst case scenario, the maximum projected loss of thickness in the steel containment due to corrosion is much less than the design corrosion allowance. Boric Acid Corrosion Guidebook, Revision 1: Managing Boric Acid Corrosion Issues at PWR Power Stations, EPRI, Palo Alto, CA (2001) 1000975, pp. 4-25 to 4-26 (NRC Staff Exhibit 63),2 has corrosion rate research data for carbon steel immersed in deaerated and aerated borated water. According to this report, the corrosion rate of steel immersed in borated water with boron concentration of 2500 ppm at 100OF is 0.00005 inch per year. Id. Water in the reactor cavity has a similar boron concentration and temperature range. Assuming the worst case scenario, that leakage has been present from the beginning of operation and continues throughout the period of extended operation, for a total of 60 years, the maximum projected loss 2
NRC Staff Exhibit 63 consists of the title page of the guidebook and pages 4-25 and 4-26. The complete guidebook may be accessed at http://my.epri.com/portal/server.pt?Abstract_id=000000000001000975.
of thickness in the steel containment due to corrosion will be 0.003 inch. The PINGP containment steel design has a corrosion allowance of 0.25 inches, or more than 80 times the maximum projected corrosion loss of 0.003 inches over a period 60 years. This worst case scenario estimate is also conservative because it does not take into account the effect of the concrete on the borated water, which will further reduce the corrosion rate.
Furthermore the allowable leakage rate, as specified in the Prairie Island Technical Specifications in compliance with the 10 CFR Part 100 requirements, is for the whole containment system and not specific to the containment steel plate. The concrete adjacent to both sides of the 1.5-inch-thick steel containment below grade contributes significantly to leak tightness of the containment in the event the steel containment is perforated. The NRC Staff has determined that a hole in excess of one-twelfth of an inch diameter in the above grade portion of the PINGP steel containment (i.e., no adjacent concrete) is required in order to exceed the allowable leakage rate as specified in the plant technical specification. However, as shown below, the hole diameter has to be significantly larger in the area where the steel containment is bounded on each side by thick layers of concrete.
A containment leakage rate test was performed at 45 psig at the North Anna 2 nuclear power plant with a 1/4 inch diameter hole present in the 1/4 inch thick steel liner plate. The liner plate was attached and in contact with the 54-inch-thick concrete containment wall. Despite the presence of the hole in the liner, the total leakage from the containment was significantly less than the allowable leakage rate specified in the plant technical specification. Similarly, corrosion of the 6-mm (1/4 inch)thick liner (i.e., pressure boundary) of concrete containments has occurred at nuclear power plants in France. The corrosion occurred over a 20-cm (8 inch) section of the liner extending from the location of the seal at the junction of containment vertical wall and the 1-meter (39 inch) thick concrete slab over the liner toward the base mat. Several 1 centimeter (0.375 inch) diameter holes penetrated the liner in this region. Despite the presence of the holes, the containments passed their integrated leakage-rate tests. In both of the above
examples, concrete adjacent to the steel liner (i.e., pressure boundary) provided significant resistance to leakage through the holes. PINGP also has concrete immediately adjacent to the containment steel. Thus, the concrete at PINGP is expected to provide significant resistance to leakage through the holes.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
NORTHERN STATES POWER COMPANY ) Docket Nos. 50-282-LR/ 50-306-LR
)
(Prairie Island Nuclear Generating Plant, )
Units 1 and 2) )
AFFIDAVIT OF ABDUL H. SHEIKH I, Abdul H. Sheikh, do hereby declare under penalty of perjury that my statements in the foregoing testimony are true and correct to the best of my knowledge and belief.
Executed in Accordance with 10 CFR § 2.304(d)
Abdul H. Sheikh Senior Structural Engineer Aging Management of Structures, Electrical, and Systems Branch Division of License Renewal, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mailstop O-11F1 Washington, DC 20555-0001 (301) 415-6004 Abdul.sheikh@nrc.gov Dated at Rockville, Maryland this 13th day of August, 2010
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
NORTHERN STATES POWER COMPANY ) Docket Nos. 50-282-LR/ 50-306-LR
)
(Prairie Island Nuclear Generating Plant, )
Units 1 and 2) )
AFFIDAVIT OF DR. DAN J. NAUS I, Dan J. Naus, do hereby declare under penalty of perjury that my statements in the foregoing testimony are true and correct to the best of my knowledge and belief.
Executed in Accordance with 10 CFR § 2.304(d)
Dr. Dan J. Naus Distinguished Research Staff Member Oak Ridge National Laboratory Post Office Box 2008 MS6069 Oak Ridge, TN 37831-6069 (865)-574-6098 nausdj@ornl.gov Dated at Oak Ridge, TN this 13th day of August, 2010