ML20195J606
| ML20195J606 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/19/1998 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9811250029 | |
| Download: ML20195J606 (11) | |
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CucnLEs H. CHUSE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant
- Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 November 19,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATfENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Response to Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Reports for the Reactor Coolant System and for the Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System
REFERENCES:
(a)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated April 8,1998," Application for License Renewal" (b)
Letter from Mr. D. L. Solorio (NRC) to Mr. Charles H. Cruse (BGE),
September 3,1998," Request for AdditionalInformation for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment, Sections 4.1,4.2,5.2,5.7,4.1, and 5.16" (c)
Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE),
September 24,1998, " Renumbering of NRC Requests for Additional Information on Calvert Cliffs Nuclear Power Plant License Renewal Application Submitted by the Baltimore Gas and Electric Company" Reference (a) forwarded the Baltimore Gas and Electric Company (BGE) license renewal application (LRA).
Reference (b) forwarded questions from NRC staff on six sections of the BGE LRA.
Reference (c) forwarded a numbering system for tracking BGE's response to all of the BGE LRA requests for additional information and the resolution of the responses. Attachment (1) provides our responses to the Reactor Coolant System, and Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System questions contained in Reference (b). The questions are renumbered in accordance with Reference (c).
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9811250029 981119 PDR ADOCK 05000317 P
PDR' l
NRC Distribution Code A036D
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Document Control Desk November 19,1998 Page 2 J
Should you have further questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours, W
STATE OF MARYLAND
- TO WIT:
COUNTY OF CALVERT I, Charles 11. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal
~ knowledge, they are based upon information provided by other BGE employees and/or consultants. Such 4
information has been reviewed in accordance with company practice and I believe it to be reliable.
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WITNESS mylland and Notarial Seal:
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Attachment:
(1) Response to Request for Additional Information; Integrated Plant Assessment Reports for the Reactor Coolant System and for the Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System cc:
R. S. Fleishman, Esquire C. I. Grimes, NRC J. E. Silberg, Esquire D. L. Solorio, NRC S. S. Bajwa, NRC Resident Inspector, NRC A. W. Dromerick, NRC R. I. McLean, DNR
- 11. J. Miller, NRC J. H. Walter, PSC
4 ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR COOLANT SYSTEM AND FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRwE MECHANISMS / ELECTRICAL SYSTEM j
Baltimore Gas and Electric Company i
Calvert Cliffs Nuclear Power Plant November 19,1998
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACTOR COOLANT SYSTEM AND FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM NRC Ouestion No. 4.1.21 Discuss whether there are Reactor Coolant System (RCS) and reactor pressure vessel (RPV) components fabricated from inconel Alb other than Alloy 600, for example, Alloy 690 and Alloy 800. Discuss whether stress corroaion c-4mg (SCC) for these components is plausible, including basis for Baltimore Gas and Electric Company % (BGE's) determination [in the License Renewal Application (LRA)].
BGE Response In the Unit 2 RCS, four instrument taps located in the pressurizer upper head,119 pressurizer heater sleeves, and one welded pressurizer heater sleeve penetration plug are fabricated from thermally-treated Inconel 690. In the Unit i RCS, three welded plugs in pressurizer heater penetrations are fabricated from thermally-treated Inconel 690. These components are not considered susceptible to stress corrosion cracking. The basis for this is that these components are exposed only to the primary water environment. Researchers have been unable to produce stress corrosion cracking in thermally-treated or mill-annealed Alloy 690 in extensive tests in simulated pressurized water reactor primary side environments. To cite a sampling of these test programs:
Reverse U-bend testing of mill-annealed and thermally-treated Alloy 690 steam generator e
tubing material produced no cracking in over 25,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of testing in high purity water at 365"C with s'imilar hydrogen content to pressurized water reactor primary water (Reference 1).
A comprehensive test program of mill-annealed and thermally-treated Alloy 690 reverse U-bends exposed to a 360 C pure water plus hydrogen environment produced no cracking after more than 31,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (Reference 2).
in 400"C ' steam with 11 psia hydrogen, representative of an accelerated primary side e
environment, no cracking of Alloy 690 Reverse U-bends occurred atler 4000 hours0.0463 days <br />1.111 hours <br />0.00661 weeks <br />0.00152 months <br /> (Reference 3).
There are no RCS or RPV components fabricated from Alloy 800.
NRC Ouestion No. 4.1.22 On pages 4.1-42 and 4.2-27 of the application, BGE indicated that the RCS and RPV components most susceptible to SCC have been or will be replaced. Identify the most susceptible Alloy 600 pressure boundary components and discuss the characteristics that render these components most susceptible to SCC. Describe what material has been or will be used in the replacement components, the schedule for l
replacement, and the basis for the schedule (i.e., how does the schedule ensure that the components will be replaced before the intended function (s) are compromised). Indicate if the replacement compownts l
are or will be within the scope of the Alloy 600 Program.
BGE Response Stress corrosion cracking of Alloy 600 in the RCS primary water environment (or primary water stress corrosion cracking [PWSCC]), is thought to be influenced by stress, operating temperature, operating time (in effective full power hours [EFPH]), and material heat treatment. The stress that gives rise to PWSCC is a combination of residual and operating stresses. Higher yield strength
[
. allows the material to have higher tensile residual stresses, which increases the probability of 1
4 ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACTOR COOLANT SYSTEM AND FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM PWSCC Factors such as additional cold working due to machining operations or rework during fabrication can play a role in increasing residual stresses. Alloy 600 material supplied in the mill-annealed (solution annealed) condition has been shown to be less resistant to PWSCC than thermally-treated material. Mill-annealed Alloy 600 with a higher final mill-annealing temperature is also generally more resistant to PWSCC than material with a lower final mill-annealing temperature. The yield strength provides an indirect indication of the final mill-annealing temperature, with higher yield strength correlating to a lower final-mill annealing temperature. The influence of other variables, such as material product form, have not been clearly defined and, therefore, cannot be accurately accounted for in susceptibility assessments. The influence of RCS chemistry variations, such as in hydrogen partial pressure and lithium concentration, are poorly defined but would tend to affect all RCS nozzles equally, it should be noted that probabilistic cracking models using the major variables of stress, temperature, and operating time have been developed for PWSCC, but actual times to failure show significant scatter around the best estimate model predictions. Industry experience with similar nozzles must also be taken into account when assessing the susceptibility of a given nozzle location.
All Alloy 600 RPV and RCS penetration materials are mill-annealed Alloy 600. All original Alloy 600, nozzles for Unit I have essentially the same number of operating hours with one exception. Similarly, virtually all Unit 2 Alloy 600 nozzles share the same number of EFPH, which is slightly less than that of Unit 1. Therefore, the major variables accounting for differences in susceptibility of nozzles within Calvert Cliffs are stress, temperature, and final mill-annealing temperature. Final mill annealing temperature, in most cases, was not recorded on material test reports and must be indirectly accounted for through the material yield strength, which also influences the stress level.
The RCS or RPV components most susceptible to SCC were the Unit 2 pressurizer heater sleeves.
These were considered most susceptible among the partial penetration welded RCS or RPV penetrations due to their relatively high yield strength for Alloy 600, relatively high operating i
temperature (650 F), and a reaming operation carried out prior to welding that cold worked the sleeve inner diameter Leakage of approximately 20 (out of a total population of 120) Unit 2 heater sleeves was discovered in 1989, prompting their replacement in 1989-1990 with thermally-treated Alloy 690 heater sleeves.
Four Unit 2 pressurizer instrument nozzles located on the pressurizer upper head were also replaced in 1990, due to leakage of one of the four, at the 71/2 location. Examination of original fabrication records showed rework at the 7-1/2 nozzle location that may have increased residual stresses, thus increasing the likelihood of PWSCC. Thermally-treated Alloy 690 nozzles were installed as replacements for the four nozzles. One of these four nozzles was replaced again in August 1998 after evidence ofleakage was observed. A liquid penetrant test (PT) of the nozzle inner diameter detected no indications. Additionally, because the Alloy 690 nozzle material is known to be highly resistant to PWSCC, a leak path through the j-groove weld was suspected. The leaking nozzle was replaced by cutting the original nozzle a short distance from the j-groove weld, depositing a weld pad on the The Units 1 and 2 reactor vessel leakage monitor penetrations are not exposed to reactor coolant unless the RPV closure o-ring leaks. For PWSCC p'ediction purposes, its was conservatively assumed the leakage monitor penetrations are exposed to reactor coolant for 50% of the Unit's EFPH.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACTOR COOLANT SYSTEM AND FOR THE KEACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM vessel outer diameter using a temper bead weld technique, inserting a new Alloy 690 nozzle, and welding the new nozzle to the weld pad using Alloy 690 type filler (Alloy 52). It should be noted that the first susceptibility assessment of Alloy 600 nozzles was not performed until after the 1989 Unit 2 heater sleeve and instrument nozzle leaks occurred and PWSCC was confirmed by detailed metallurgical analysis.
All the original pressurizer instrument nozzles in both units were fabricated from the same heat of material and share the highest operating temperature among Calvert Cliffs Alloy 600 nozzles. The
. material heat is relatively low in yield strength, which tends to increase the predicted time for PWSCC. There are seven instrument nozzles per pressurizer, which vary in PWSCC susceptibility l
. dependent on their location. The early failure of the Unit 2 upper level instrument nozz!e is attributed to the rework during fabrication.
The Unit 1 pressurizer heater sleeves are the next most PWSCC-susceptible component because they were (with one exception) fabricated from the same heat of Alloy 600 material as the Unit 2 heater sleeves, and have the same configuration and have the same operating temperature as the original Unit 2 heater sleeves. The absence of the pre-weld reaming operation or significant rework made the Unit 1 sleeves slightly less susceptible to PWSCC. Nickel plating was applied in 1994 to the Unit I heater sleeve inner diameter in the top three to four inches of the sleeve as a PWSCC mitigative treatment. Replacement is, therefore, not planned for the Unit 1 pressurizer heater sleeves. The Unit I heater sleeves will continue to be managed under the Alloy 600 Program Plan.
The next most susceptible Alloy 600 components are the four Unit 1 pressurizer vapor space instrument nozzles located on the pressurizer upper head. These are fabricated from the same heat of material as the Unit 2 pressurizer upper head instrument nozzles replaced in 1990. The assessment of these nozzles as having high susceptibility is based on the high operating temperature of these nozzles (650*F), the number of operating hours, and an industry history of PWSCC in pressurizer vapor space nozzles. Baltimore Gas and Electric Company found no evidence of rework on these nozzles based on a review of original fabrication records, which is thought to have made Unit 2 upper instrument nozzles more susceptible to PWSCC than the otherwise identical Unit i vapor space instrument nozzles. These nozzles are scheduled to be repaired or replaced during the 2000 refueling outage.
The schedule for this replacement is intended to be prior to the predicted time of the first of these nozzles developing a through wall crack; however, actual results can show a large amount of scatter around the best estimate prediction. No predictive model can provide 100% assurance that a nozzle will not leak prior to the predicted time. Only axial PWSCC cracking is expected for these nozzles, and leakage of these nozzles from axial cracks does not constitute a threat to nuclear safety. Current plans are to weld replacement nozzles utilizing a design with the structural weld on the vessel outer surface, similar to the Unit 2 pressurizer instrument nozzle repair described above. However, a mechanical nozzle seal may be used.
The Units 1 and 2 pressurizer instrument nozzles located on the pressurizer bottom head are the next most susceptible nozzles to PWSCC. In BGE's judgment, these nozzles are less likely to experience PWSCC than the instrument nozzles on the pressurizer upper head due to an industry history of very low incidence of PWSCC in pressurizer bottom head nozzles. The Units 1 and 2 pressurizer mid-level nozzle (1 per unit) are somewhat less susceptible than the pressurizer bottom head instrument nozzles due to the geometry, which results in lower residual stresses. Replacement is not currently 3
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACTOR COOLANT SYSTEM AND FOR THE SEACTQR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM planned for the pressurizer bottom head or mid-level nozzles; however, contingency plans for replacement of these nozzles are scheduled to be in place by September 1999.
The susceptibility of Units 1 and 2 control element drive mechanism nuzzles will not be discusse' in d
this response since they will be addressed in the reply to the request for additional information for GL 97-01, " Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations," forwarded by Reference (4).
Other Alloy 600 RPV and RCS penetrations are not predicted to experience PWSCC during the current or proposed extended license for Calvert Cliffs Units 1 and 2.
l All nozzle replacements will utilize Alloy 690 thermally-treated material or other SCC-resistant materials, excluding Alloy 600. Weldments exposed to the RCS environment will use Alloy 690 type weld fillers or other SCC-resistant materials. These replacement nozzles will, therefore, not be managed under the Alloy 600 Program Plan, which was designed to specifically address PWSCC of Alloy 600 and Alloy 600 -type weld fillers.
Although this response provides currently scheduled milestones for certain replacement activities, these dates are subject to change. The schedule for implementing Alloy 600 Program activities is based on plant availability, safety considerations, and site priorities. It should be noted, however, that corrective actions ander the Alloy 600 Program will be implemented in a timely manner to ensure the 1
safety functions of equipment concern are maintained in accordance with the current licensing basis at all times.
NRC Ouestion No. 4.1.23 Describe the specific inspection activities for the most susceptible RCS and RPV components under the Alloy'600 program. Include a description of and the bases for the included components, inspection schedules, inspection techniques, inspection procedures, inspection personnel qualification, acceptance criteria, and sample expansion criteria.
BGE Response Those remaining Alloy 600 components that are highest in susceptibility will be replaced rather than periodically inspected with an augmented technique. Primary water stress corrosion cracking of Alloy 600 penetrations is considered an economic risk rather than a nuclear safety risk, and augmented inspections would only be cost justified for large populations of similar nozzles that are judged highly susceptible and are costly to replace.
All Alloy 600 nozzles are examined, at a minimum, every 24 months. The inspections are conducted under the Boric Acid Corrosion Inspection Program as stated on page 4.1-46 of the BGE LRA. The inspections consist of VT 2 examinations (a type of visual examination described in American Society of Mechanical Engineers-(ASME) XI, IWA-2212), with the exception of the Unit 1 pressurizer h:ater sleeves, for which the following inspections are performed:
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A'ITACHMENT (1) s REST ONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REI' ORTS FOR Tile REACTOR COOLANT SYSTEM AND FOR TiiE 1(EACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECllANISMS/ ELECTRICAL SYSTEM Two types of examination are currently performed on the pressurizer heater sleeves under the Boric Acid Corrosion Inspection Program:
During each reactor shutdown where a hot standby (Mode 3) condition is reached, a VT-2 examination (as described in ASME Section XI, IWA-2212) is performed to detect evidence ofleakage. The exam must be performed from a location that will directly view the visible portion of the pressurizer heater sleeve, the insulation and insulation seams around the penetration point. This exam need not be performed if a similar examination has been performed in the past 30 days.
During each refueling outage (Mode 6), a visual examination (VT-1) (as described In ASME Section XI, IWA-2211) shall be performed to detect evidence of leakage by viewing each penetration region for boron deposits. This is a " bare metal" inspection.
The inspection requirement originally required the VT-1 exam be performed at a maximum interval of 18 months. The requirement was modified so the plant would not have to shut down to perform the inspection, since Calvert Cliffs Unit 1 is now on a 24-month refueling cycle. The basis for the inspection interval is that it is sufficient to allow detection of cracked penetrations via boric acid leakage prior to the cracks attaining critical size. Fracture mechanics calculations have also been performed that show the critical crack length for axial cracks in Alloy 600 penetration are very large.
Reference (5) describes one such calculation. Axial cracks would be incapable of attaining a length that could cause unstable crack propagation prior to becoming through wall and being detected via boric acid leakage. Several evaluations considering boric acid corrosion of carbon or low-alloy steel components due to leaking Alloy 600 nozzles have concluded that boric acid corrosion would not violate code requirements for wall thickness prior to the leakage being detected by the routine visual inspection. Reference (6) is an example of one such evaluation. Industry experience, along with stress analyses (References 7 and 8), have shown that circumferential cracks are not expected in Alloy 600 penetrations. There are no expansion criteria for the routine visual examinations since 100% of the penetrations are inspected.
All non-destructive examination (NDE) personnel who perform examinations on Alloy 600 penetrations at Calvert Cliffs are qualified and certified using a written practice prepared in accordance with the American Society for Non Destructive Testing (ASNT) Recommended Practice No. SNT-TC-1 A -1980, " Recommended Practice for Nondestructive Testing Personnel Qualification i
and Certification," in accordance with ASME Section XI 1983 Edition through Summer 1983 addenda. Non-destructive examination personnel are certified at a minimum as Level 11, in the applicable technique, and all results were and will be reviewed and accepted by an ASNT-qualified Level Ill examiner in the applicable technique. In the future, NDE examiners may be qualified to different standards if mandated by ASME Section XI.
Several examinations using tedmiques other than visual examination of the component outer surface have been performed at Calvert Cliffs. After visual detection of leakage in 1989 on the Unit 2 pressurizer heater sleeves, extensive NDE was performed on both the Units 2 and I pressurizer penetrations, as summarized below (the examinations are described in more detail in Reference 5):
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4 ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR Tile REACTOR COOLANT SYSTEM AND FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECilANISMS/ ELECTRICAL SYSTEM Unit 2 Pressurizer Heater Slecyss Visual (VT), PT, and eddy current (ET) inspections were perfonned on the inner diameter of 28 Unit 2 heater sleeves (20 identified as definitely leaking by the visual examination and 8 possible leakers). Indications were detected by at least one NDE technique in 23 of the 28 heater sleeves.
No expansion criteria were documented for these examinations. The heater sleeves were examined to help determine the nature of the problem causing the leakage, which was unknown at the time. The initial visual identification ofleakage was sufficient to establish the generic nature of the problem, thus it was known that repair or replacement of all the Unit 2 heater sleeves would be necessary.
Pressurizer Instrument Nozzles Upper or Vapor Space: Two nozzles were inspected with PT and ET, including the one determined by visual inspection to be leaking. Indications were detected in the visually leaking nozzle.
Lower or Bottom Head: Two nozzles were inspected using PT and ET, with no indications detected.
Unit 1 Pressurizer Heater Sleeves Twelve heater sleeves were inspected using VT, PT and ET. No indications were detected. The inspection scope would have expanded to 100% of the sleeves had any indications been detected in the sample.
Pressurizer Instrument Nozzles Upper or Vapor Space Nozzles: VT, PT, and ET examinations of two nozzles revealed no indications, in 1994, an ET inspection was performed on 31 pressurizer heater sleeves in Unit 1. The exam was performed after nickel plating had been applied to the top 3-4 inches of all but one of the sleeves.
l The objective of the exam was to ensure no cracking existed in the non-plated regions of the sleeves; in light of the fact that circumferential cracking had been found in sleeve FF-1 below the pressurizer shell(Reference 9). This exam also utilized motorized rotating nancake ET probes. No indications were found. This examination was a one-time examination. l{ad the ET inspections detected any indications in the sample, the inspection scope would have been expanded to examine 100% of the sleeves.
NRC Ouestion No. 4.1.24 Describe the most recent example ofimplementation of BGE's corrective action program initiated by, or related to, the Alloy 600 Program, include a description of the initiating event, the corrective action (s) taken, and how the issue was resolved.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACI'OR COOLANT SYSTEM AND FOR T!!E 1(EACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECIIANISMS/ ELECTRICAL SYSTEM BGE Response On July 25,1998, with Unit 2 in Mode 3 (flot Standby), a steam leak was discovered at an upper level tap on the pressurizer. This leak was considered RCS pressure boundary leakage and Operations promptly shut down the Unit to Mode 5 (Cold Shutdown). This event resulted in no significant consequences to the health and safety of the public. There were no personnel injuries or associated i
equipment damage as a result of the steam leak.
The cause of the leak is postulated to be a crack in the Inconel Alloy 600-type weld Oller material of the nozzle caused by PWSCC. The leaking penetration was repaired from the outside of the pressurizer. Visual inspections were completed on the three other upper level taps and heater sleeves that use the same weld filler material and nozzle material as the leaking upper level tap. No evidence ofleakage was discovered during these inspections.
This event and corrective actions were reported to the NRC in BGE's Licensee Event Report No. 318/98-005, forwarded by Reference (10).
NRC Ouestion No. 4.125 The application indicates that the Alloy 600 program will be modified. Describe the reason for the program changes, schedule, and proposed content related to this program modification to include all Alloy 600 RCS and RPV components, including RCS nozzle thermal sleeves.
BGE Response The reason for the changes to the Alloy 600 Program Plan is to add non-pressure boundary Alloy 600 components to the program. These non-pressure boundary components include all Alloy 600 thermal sleeves in the RCS, and all Alloy 600 internal attachments to the RPV. The content for these additions will consist of an assessment of the PWSCC susceptibility of these components based on fabrication history, material properties, and operating environment; an assessment of the failure consequences of these components; a recommendation for corrective actions if any are deemed necessary; and a recommended time frame for completion of any corrective actions. Corrective actions may include, but are not limited to, inspection, replacement, or a mitigation treatment. The program plan changes will be completed by March 1999.
Additional changes are made as necessary to the Alloy 600 Program Plan, which, by procedure, is reviewed annually.
NRC Ouestion No. 4.1.26 Provide the results of BGE's most recent internal audit of the Alloy 600 program; including areas of strengths and weaknesses, safety implication of Gndings, and corrective action plans and schedule for implementation.
BGE Response Baltimore Gas and Electric Company has requested clari0 cation from NRC on this item and has agreed to work toward clari0 cation through forthcoming interaction, most likely in the form of a 7
A'ITACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORTS FOR THE REACTOR COOLANT SYSTEM AND FOR THE I{EACTQR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM BGE Response Baltimore Gas and Electric Company has requested clarification from NRC on this item and has agreed to work toward clarification through forthcoming interaction, most likely in the form of a public meeting. Baltimore Gas and Electric Company may supplement this response, based on the outcome of that interaction.
References 1.
Norring, K, Engstrom, J., Tornblom, H., "Intergranular Stress Corrosion Cracking of Steam Generator Tubing. 25000 hours Testing of Alloy 600 and Alloy 690," in Proceedings of the Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 6-10,1989, NACE International l
2.
Santarini, G., Blanchet, J., Rouillon, Y., Duysen, J.C., Slama, G., and Gimond, C., " Alloy 690:
Recent Corrosion Results", In Proceedings: 1989 EPRI Alloy 690 Workshop" EPRI NP-6750-M, April 1990
)
3.
Aspen, R. G., Grand, T. F., and Harrod, D.L., " Corrosion Performance of Alloy 690 - In Proceedings,1989 EPRI Alloy 690 Workshop," EPRI NP-6750-M, April 1990 4.
Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (DGE), dated August 24,1998,
" Generic Letter (GL) 97-01, ' Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations' Responses for Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 and the Relationship of the Responses to Topical Report No. CE NPSD-1085" 5.
Letter from Mr.
G.
C.
Creel (BGE) to NRC Document Control Desk, dated September 20,1989, " Submittal of Basis for Determination"(Unit 1 Safe Operation Relative to Unit 2 Pressurizer Heater Sleeve Leakage) 6.
Combustion Engineering Owners Group Task 700 Report, " Evaluation of Pressurizer Penetrations and Evaluation of Corrosion after Unidentified Leakage Develops," 1992 7.
ABB-Combustion Engineering Report CEN-607, " Safety Evaluation Of The Potential for and i
Consequence of Reactor Vessel Head Penetration Alloy 600 ID Initiated Nozzle Cracking,"
May 1993 8.
Combustion Engineering Owner's Group Report del 357, " Stress Analysis - Alloy 600 CEDM And ICI Nozzles," May 1993 (prepared by Dominion Engineering) 9.
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated April 20,1994,
" Licensee Event Report 94-003, ' Pressurizer Heater Sleeve Cracking'"
10.
Letter from Mr. P. E. Katz (BGE) to NRC Document Control Desk, dated August 24,1998,
" Licensee Event Report 95-005, Plant Cooldown Due to Reactor Coolant System Pressure Boundary Leakage" I
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