ML20154D959
| ML20154D959 | |
| Person / Time | |
|---|---|
| Issue date: | 02/05/1988 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Jordan E Committee To Review Generic Requirements |
| Shared Package | |
| ML20153G441 | List: |
| References | |
| NUDOCS 8809160178 | |
| Download: ML20154D959 (12) | |
Text
- UNITED STATES
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NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
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FEB 5 1988 MEMORANDUM FOR: Edward L. Jordan, Chaiman Comittee To Review Generic Requirements FROM: Thomas E. Murley, Director Office of Nuclear Reactor Regulation
SUBJECT:
GENERIC LETTER ON BORIC ACID CORROSION OF CARBON STEEL REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS The subject dratt generic letter is included as Enclosure 1 for your attention.
We request that the Comittee To Review Generic Requirements (CRGR) be convened to consider the backfit of the proposed new requirements under the provisions of 10 CFR 50.109a(4). We have detemined that som* pressurized water reactor plants may fail to meet their licensing basis for .ne reactor coolant pressure boundary as defined by General Design Criteria 14. 30, and 31 of Appendix A to 10 CFR 50 relative to preventing degradation of the reactor coolant pressure boundary by boric acid corrosion. For the case under consideration, the provisions of 10 CFR rS.109a(2) and (3) are r.ot applicable, and, therefore, backfit analysis is not required.
The generic letter requires that PWR licensees provide assurances that they have implemented a program to detect, locate, evaluate, and repair borated water leakage that could degrade the reactor coolant pressure boundary. This will ensure that the reactor coolant pressure boundary will continue to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture as required by General Design Criterion 14 These require-ments also will ensure that the requirements of General Design Criteria 30 and 31 related to providing means for detecting and identifying the locations of the source of reactor coolant leakage and for ensuring that the reactor coolant pressure boundary shall have a sufficient margin to withstand the stresses under different operating conditions with only minimal probability of producing a rapidly ,ropagating failure are met.
Please perfom the fomal CRGR review of the issues prasented in this draf t generic letter. CRGR requirement item IV.B is provided as Enclosure 2.
Contact:
C. McCracken. DEST /NRR 49-23249
$$926017eoco434 HEET!QfNRCC
,' E. Jordan -2 FEB 5 193 This issue is sponsored by Mr. Lawrence Shao, Director, Division of Engineering and Systems Technology.
If you have any questions, please contact my staff.
~
nas E. Murl , ctor f ice of Nuclear R actor Regulation
Enclosures:
- 1. Draft Generic Letter
- 2. CRGR Item IV B.
cc: w/ enclosures:
W. Russell, Ifgion 1 J. N. Grace, kcgion II A. D. Davis, Region !!!
R. D. Martin, Region IV J. B. Martin, Region V
'. UNITED STATES
!" i NUCLEAR REGULATORY COMMISSION 7, j W ASHING TON, D. C. ?0555
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ENCLOSURE 1 I l
4 ALL LICENSEES OF OPERATING PWRS AND HOLDERS OF CONSTRUCTION PERMITS FOR PWRS I GENTLEMEN:
Subject:
BORIC ACID CORROSION OF CARBON STEEL REACTOR PRESSURE BOUNDARY COMPONENTS IN PWR PLANTS (GENERIC LETTER 88- )
Pu.suant to 10 CFR 50.54(f), the Nuclear Regulatory Cemission is requesting l infonnation to assess safe operation of pressurized water NMtors (PWRs) when reactor coolant leaks below technical specification lid t$ develop and the coolant containing dissolved boric acid comes in co%act with and degrades low alloy carbon steel components. The principal concern is whether the affected plants continue to meet the requirements of General Design Criteria 14, 30 and 31 of Appendix A to Title 10 of the Code of Federal Regulations (CFR) Part 50 when the concentrated boric acid solution or boric acid crystals, fonned by evaporation of water from the leaking reactor coolant, corrode the reactor coolant pressure boundary. Our concerns regarding this issue were prompted by incidents in PWR plants where leaking reactor coolant caused significant corrosion problems. In many of these cases, although the licensees had detected the existence of leaks, they had not evaluated their significance relative to the safety of the plant nor had they promptly taken appropriate corrective actions. Recently reported incidents are listed below.
(1) At Turkey Point Unit 4, leakage of reactor coolant from the lower instrument tube seal on one of the incore instrument tubes resulted in corrosion of various components on the reactor vessel head including three reactor vessel bolts. The maximum depth of corrosion was 0.25 inches.
(!E Infonr.ation Notice No.86-108, Supplement 1)
(2) At Salem Unit 2, leakage occurred from the seal weld o'1 one of the instrument penetrations in the reactor vessel head, and the leaking coolant corroded the head surface. The maximum depth of corrc2 ion was 0.36 inches.
(!E Infonnation Notice No.86-108 Supplement 2)
(3) At Saa Onofre Unit 2, boric acid solution corroded nearly through the bolts holding the valve packing follow plate in the shutdown cooling system isolation valve. During an attempt to operate the valve, the bolts failed and the valve packing follow plate became dislodged causing leakage of approximately 18,000 gallons of reactor coolant into the containment.
(IE Infonnation Notice No.86-108. Supplement 2)
(4) At Arkansas Nuclear One Unit 1, leakage from a high pressure injection valve dripped onto the high pressure injection nozzle. The maximum depth of corrosion was 0.5 inches, which represented a 67 percent penetration of the pressure boundary. (IE Infonnation Notice No.86-108)
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(5) At Fort Calhoun, seven reactor coolant pump studs were reduced by baric acid corrosion from a nominal 3.5 inches to between 1.0 and 1.5 inches.
(IE Infonnation Notice 80-27)
Additionally, corrosion rates of up to 400 mils / month have been reported from an er.perimental program. (IE Infomation Notice No.86-108, Supplement 2)
Although failure of the reactor coolant pressure boundary did not occur in every instance, all of these incidents demonstrated the potential adverse consequences of boric acid corrosion.
The corrosion caused by the leaking coolant containing dissolved boric acid has been recognized for some time. Since 1979, the NRC has issued five infomation notices (80-27; 82-06;86-108; and 86-108, Supplements 1 and 2) and Bulletin 82-02 addressing this problem. In June 1981, the Institute for Nuclear Power Operations issued a report discussing the effect of low level leakage from the gasket of a reactor coolant pump and concluded that significant corrosion of the pump studs could occur during all modes of operation. In December 1984, the Electric Power Research Institute issued a sumary report on the corrosion of low alloy steel fasteners which, among other things, discussed .
boric acid-induced corrosiun. The infomation contained in these documents !
clearly indicated that boric acid solution leaking from the reactor coolant system can cause significant corrosion damage to carbon steel reactor coolant pressure boundaries.
Office of Inspection and Enforcement (IE)Bulletin 82-02 requested licensees '
to identify all of the bolted closures in the reactor coolant pressare boundary that had experienced leakages and to infom the NRC about the inspections to be made and the corrective actions to be taken to eliminate that problem.
However, the bulletin did not require the licensees to institute a systematic program for monitoring small primary coolant leakages and to perform maintenance before the leakages could cause significant corrosion damage.
In light of the above experience, the NRC believes that boric acid leakage potentially affecting the integrity of the reactor coolant pressure boundary should be procedurally controlled to ensure continued compliance with the licensing basis. We therefore request that you provide assurances that a program has been implemented consisting of systematic measures to ensure that boric acid corrosion does not lead to degradation of the assurance that the reactor coolant pressure boundary will have an extremely low probability of abnomal leakage, rapidly propagating failure, or gross rupture. The program should include the following:
(1) A detemination of the principal locations where leaks that are smaller than the allowable technical specification limit can cause degradation of the primary pressure boundary by boric acid corrosion. Particular consideration should be given to identifying those locations where conditions exist that could cause high concentrations of boric acid on pressure boundary surfaces.
(2) Procedures for locating small coolant leaks (i.e., leakage rates at less than technical specification limits). It is important to establish the potential path of the leaking coolant and the reactor pressure boundary components it is likely to contact. This information is important in determining the interaction between the leaking coolant and reactor coolant pressure boundary materials.
(3) Methods for conducting examinations and performing engineering evaluations to establish the impact on the reactor coolant pressure boundary when leakage is located. This should include procedures to promptly gather the necessary infomation for an engineering evaluation before the removal of evidence of leakage, such as boric acid crystal buildup.
(4) Corrective actions to prevent recurrences of this type of corrosion. This should include any modifications to be introduced in the present design or operating procedures of the plant that (a) reduce the probability of primary coolant leaks at the locations where they may cause corrosion damage and (b) entail the use of suitable corrosion resistant materials or the application of protective coatings / claddings.
Additional insight into the phenomena re'ated to boric acid corrosion of carbon steel corrponents is provided in the attachment to this letter.
The request that licensees provide assurances that a program has been implemented to address the corrosive effects of reactor coolant system leakage at less than technical specification limits constitutes a new staff position. Previous staff positions have not considered the corrosion of external surfaces of the reactor coolant pressure boundary. Based on the frequency and continuing pattern of significant degradation of the reactor coolant pressure boundary that was discussed above, the staff now concludes that in the absence of such a program compliance with General Design Criteria 14, 30 and 31 cannot be ensured. This request is consistent with applicable requirements of 10 CFR Part 50 109(a)(4).
You are required to submit your response signed under oath or affimation, as specified in 10 CFR 50.54(f), within 60 days of receipt of this letter. Your response will be used to determine whether your license should be modified, suspended, or revoked. Your response should provide assurances that such a program is in place or provide a schedule for implementing such a program if one is not in place.
This information is required pursuant to 10 CFR 50.54(f) to assess confomance of PWRs with their licensing basis and to datermine whether additional NRC action is necessary. The staff does not request submittal and will not conduct a detailed review and a fom61 approval of your program. You shall maintain, in auditable fom, records of the program and results obtained from implementation of the program and shall make such records available to NRC inspectors upon request.
This requee.t for infomation is covered by the Office of Management and Budget ur,L r Clearance Number 3150-0011, which expires December 31, 1989.
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Consnents on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, !
Washington, D.C. 20503. ;
Sincerely, ,
Frank J. Miraglia, Jr.
Associate Director for Projects '
Office of Nuclear Reactor Regulation '
Enclosure:
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ATTACHMENT BORIC ACID CORROSION OF CARBON STEEL REACTOR COMPONENTS IN PWR PLANTS Boric acid is used in PWR plants as a reactivity control agent. Itt concentration it, the reactor coolant ranges between 0 and approximately I weight percent. At these concentrations boric acid solutionc will n't cause significant corrosion even if they come in contact with carhon stol components. In many cases, however, coolant that leaks out of the 'esctor coolant system loses a substantial volume of its water by evaporation, resulting in the fonnation of highly concentrated boric acid solutiens or deposits of boric acid crystals. These concentrated solutions of boric acid may be very corrosive for carbon steel. This is illust ated by recent test data, tabulated below, which were referenced in NRC Information Notice No.86-108 Supplement 2.
Concentration of boric acid Temperature Corrosion rate (percent) Condition (*F) mils / month 25 Aerated 200 400 25 Deaera ted 200 250 15 Aera ted 200 350-400 15-25 Dripping 210 400 If all of the water evaporates and boric acid crystals are fonned, the corrosion is less severe. However, boric acid crystals are not completely benign toward carbon steel, and at a temperature of 500'F, corrosion rates of 0.8 to 1.6 mils /menth were obtained in the Westinghouse tests referenced in the generic letter. Corrosion by boric acid crystals was observed in Turkey Point Unit 4 where more than 500 pounds of boric acid crystals were found on the reactor vessel head. After these crystals were removed, corrosion of various corrporents on the reactor vessel head was observed.
The most effective way to prevent boric acid corrosion is to minimize reactor coolant leakages. This can be achieved by frequent monitoring of the locations where potential leakages could occur and repairing the leaky components as soon as possible. Review of the locations where leakages have occurred in the past indicates that the most likely locations are (1) valves; (2) flanged connections in steam generator manways, reactor head closure, etc.; (3) primary coolant pumps where leakages occur at cover-to-casing connections as a result of defective gaskets; and (4) defective welds.
In many of these locations the components exposed to boric acid solution are covered by insulation and the leaks may be difficult to detect. If leak detection systems have been installed in the components (e.g., reactor coolant pumps from certain venders), they should be used to monitor for leakage.
It is important to detemine not only the source of the leakage but also the path taken by the leaking fluid by evaluating the mechanism by which leaking boric acid is transported. In some cases boric acid may be entrained in the steam emerging from the opening in the pressure boundary .
- that subsequently condenses inside the insulation thus carrying boric acid I to locations that are remote from the source of leakage. ,
Boric acid corrosion can be classified into two distinct types: (1) i corrosion that actually increases the rate of leakage and (2) corrosion that occurs some distance from the source of leakage and hence does not significantly affect the rate of leakage. An example of the first type
, is the corrosion of fasteners in the reactor coolant pressure bounury,
, for example, in reactor coolant pumps. This type of corrosion can lead to excessive corrosion of studs. The second type of corrosion can contribute significantly to the degradation of the reactor coolant pressure boundary.
At Arkansas Nuclear One Unit 1, a leak developed in a high pressure injection 15o14 tion valve located 8 feet above the high pressure iajection nozzle which was made of carbon steel. Accumulation of boric acid resulted in an approximately 1/2-inch-deep corrosion wastage adjacent to the stainless-to-carbon steel weld. Other locations of the nozzle exhibited corrosion to a lesser degree. Corrosion of the reactor vessel head was observed at Salem Unit 2. Corrosion pits were 1 to 3 inches in diameter and 40 to 300 mils deep. The source of this corrosion was a defective seal weld in one of the instrument penetrations. These examples indicate that the corrosion produced by boric acid could degrade even relatively bulky components. At Fort Calhoun, the diameter of a reactor coolant pump closure bolt was reduced from 3.5 inches to 1.1 inches by boric acid corrosion. At San Onofre Unit 2, boric acid corrosion of the valve bolts was responsible for the failure of the valve and the discharge of 18,000 gallons of primary coolant into the containment.
Because of the nature of the corrosion produced by boric acid, the most reliable method of inspection of components is by visual examination.
Ultrasonic testing perfomed in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code may not be sensitive enough to detect the wastage. At Fort Calhoun, two successive ultrasonic tests failed to detect corrosion of the reactor pump closure studs. When ultrasonic testing is used, the licensee should provide assurances that the results are reliable.
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,' ENCLOSURE 2 CRGR Item IV.8. Contents of Packages Submitted to CRGR (Rev. 4 Stello to List 042387, des 41860 342 ff)
The following requirements apply for proposals to reduce existing requirements or (regulatory) positions as well as proposals to increase requirements ori (regulatory) positions. Each package submitted to the CRGR for review shall include fifteen (15) copies of the following infonnation:
Subject:
BORIC ACID CORROSION OF CARBON STEEL REACTOR PRESSURE BOUNDARY COMPONENTS IN PWR PLANTS
!. The proposed generic requirement or staff position as it is proposed to be sent out to licensees.
A. The proposed generic requirements are spelled out in the proposed generic letter enclosed with this review package. A brief sumary of these requirements is provided below.
B. It is required that the licensees provide assurances that a program is in place to detect, locate, evaluate, and repair leakage that can degrade the reactor coolant pressure boundary.
II. Draft staff papers or other underlying staff documents supporting the requirements or staff positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any comittee member may request CRGR staff to obtain a copy of any referenced material for his or her use.)
A. V. S. Nuclear Regulatory Comission, IE Bulletin No. 82-02, "Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants," June 2, 1982.
B. U. S. Nuclear Regulatory Comission, IE Information Notice No. '
80-27, "Degradation of Reactor Coolant Pump Studs," June 11, 1980 (FortCalhoun).
C. V. S. Nuclear Regulatory Comission, IE Information Notice No. 82-06, "Failure of Steam Generator Primary Side Manway Closure Studs," Harch 12, 1982 (Maine Yankee).
D. U. S. Nuclear Regulatory Comission, IE Infonnation Notice No.86-108, "Degradation of Reactor Coolant System Pressure Boundary Resulting From Boric Acid Corrosion," December 29, 1986 (Arkansss Nuclear One, Unit 1); Supplement 1. April 20, 1987 (Turkey Point Unit 4); Supplemont 2 November 19,1987(SalerUnit2).
!!!, Each proposed requirement or staff position shall contain the sponsoring i office's position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirements or staff positions.
A. The proposed requirements would ensure compliance with General Design Criteria 14, 30 and 31 of Appendix A to 10 CFR Part 50 relative to preventing degradation of the reactor pressure boundary by boric acid corrosion. It is expected that most plants already have such a program, so this will only result in increased requirements for plants which do not provide assurances that a program is already in place.
B. Staff regulatory positions are increased by this proposed generic letter.
IV. The proposed method of implementation along with the concurrence (and any coments) of OGC on the method proposed.
The staff requires that within 60 days of receipt of this letter the licensees provide assurances that such a program is in place or provide a schedule for implementing such a program if one is not in place. Plants which do not have a program in place will be directed to do so by confirmatory order. The staff does not request submittal and will not conduct a detailed review and a fonnal approval of licensees' programs. However, licensees shall maintain, in auditable fom, records of the program and results obtained from implementation of the program and shall make such records available to NRC inspectors upon request.
OGC has reviewed the proposed generic letter and their connents were incorporated.
V. Regulatory analyses generally conforming to the directives and guidance of NUREG/BR-0058 and NUREG/CR-3568.
This request for infomation was approved by the Office of Management and Budget under clearance number 3150 0011, which expires December 31, 1989.
Yl. Identification of the category of reactor plants to which the generic reqeirement or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only. Ols after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants, all water reactors, all PWRs only, some vendor types, some vintage types such as BWR 6 and 4. jet pump and nonjet pump plants, etc.).
The proposed generic requirements are applicable to all PWRs.
V!!. For each such category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The evaluation shall document for consideration infomation available concerning any of the factors as may be appropriate and any other infomation relevant and material to the proposed action.
Response to this item is not required pursuant to Revision 4 of the CRGR Charter, Sectio 9 111.0, since the requirements of the proposed generic letter are intended to bring facilities into confor: nance with the rules of the Comission.
VIII. For each evaluation conducted pursuant to 10 CFR 50.109, the proposing office director's determination, together with the rationale for the determination based on the considerations of paragraphs (i) through (vii)above,that A. there is a substantial increase in the overall protection of public health and safety or the common defense and security to be Jerived from the proposal.
- 1. Boric acid corrosion of carbon steel components could constitute a major degradation of essential safety-related equipment; namely, the primary coolant boundary. Although such events, like LOCAs in general, are within the design basis, such events constitute a reduction in the degree of protection 4
of public health and safety. Such events have been considered reportable to Congress as at, normal occurrences pursuant tu a NRC policy statement published in the Federal Register on February 24, 2977 (Volume 42, No. 37, pp. 10950-10952).
- 2. The proposed generic letter is intended to identify and remedy conditions that may occur when a leakage of borated reactor coolant develops. Thus, the staff concludes that the *,ineric lett' may result in a substantive increase in the overall protection of public and safety by ensuring that the primary pressure boundary is not degraded below its licensing / design basis.
B. the direct and indirect costs of implementation, for the facilities '
affected, are justified in view of this increased protection.
- 1. Direct and indirect costs associated with the actions required by the generic letter involve primarily setting up the program if one is not already in place and developing suitable procedures for detection and identification of the leakages and for examination of the affected components. In many instances the licensees may have already established procedures for perfoming these functions.
- 2. Although the costs for implementing the actions required by the generic letter have not been estimated by the staff, the costs are expected to be modest and justifiable in view of the ;
increased protection to public health and safety. Furthemore ,
these costs should be offset by avoided costs that could be expected to accompany any replacement of the corroded components.
IX. For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing office director's detemination, together with the rationale for the detemination based on the considerations of paragraph (1) through (vii) above, that this is not applicable to the proposed generic letter since no relaxation or decrease in current requirements is being proposed.
7 References (1) Electric Power Research Institute, EPRI NP-3784 Topical Report, "A Survey of the Literature on Low-Alloy Steel Fastener Corrosion in PWR Plants,"
Palto Alto, CA, December, 1984.
(2) Institute for Nuclear Power Operations, INP0/NSAC Significant Operating Report 81-12. "Reactor Coolant Pump Closure Stud Corrosion," Atlanta, GA, June 24, 1981.
(3) Johnson, W. J. (Westinghouse). Letter NS-NRC-87-3200 to C. Berlinger (NRC),
"Degradation of Reactor Coolant System Pressure Boundary Resulting From Boric Acid Corrosion," October 15, 1987.
I (4) U.S. Nuclear Regulatory Comission, IE Bulletin No. 82-02, "Degradation I
of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants," June 2, 1982.
(5) U.S. Nuclear Regulatory Comission, IE Information Notice No. 80-27 "Degradation of Reactor Coolant Pump Sturis," June 11,1980(FortCalhoun).
(6) U.S. Nuclear Regulatory Comission, IE Infomation Notice No. 82-06, i "Failure of Steam Generator Primary Side Manway Closure Studs," March 12, 1 1982 (Maine Yankee). l l
(7) U.S. Nuclear Regulatory Comission, IE Infomation Notice No.86-108, "Degradation of Reactor Coolant System Pressure Boundary Resulting From Boric Acid Corrosion," December 29,1986(ArkansasNuclearOne, Unit 1);
Supplement 1. April 20,1987 (Turkey Point Unit 4); Supplement 2 November 19, 1987 (Salem Unit 2).
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