ML20203J338
| ML20203J338 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/19/1999 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9902240010 | |
| Download: ML20203J338 (10) | |
Text
-
d CHARLES H. CRUSE BaltIrnore Cras and Electric Cornphny Vice President Cahen Chif s Nuclear Power Plant
}
Nuclear Energy 1650 Cah en Chff s Parkway Lusby, Maryland 20657 410 495-4455 February 19,1999 U. S. Nuclear Regulatory Commission Washington, DC 20555
~
ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Response to Request for Specific Information Needed for the NRC Evaluation of Environmental Oualificatica hr Licensa Renewal
REFERENCE:
(a)
Letter from Mr. C.1. Grimes (NRC) to Mr. C. H. Cruse (BGE),
January 7,1999," Specific Information Needed for the Staff Evaluation of Environmental Qualification for License Renewal" Reference (a) forwa ded four information requests from NRC Staff regarding the NRC evaluation of ens onmental qualification for license renewal. A conference call occurred on January 22,1999 to discuss clarification of the requests, specifically Request No. 4.
Responses to the four requests are provided in Attachment (1). Request No. 4a has been modified as agreed to during the telephone conference.
i I
m.s
.s 9902240010 990219 PDR ADOCK 05000317 P
PM NRC Distribution Code A036D
m,
^
Document Control Desk February 19,1999 Page 2 Should you have further questions regarding this matter, we will be pleased to discuss them with you.
Vcry truly yours, STATE OF MARYLAND
- TO WIT:
COUNTY OF CALVERT I, Charler H. 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 docuniesit are + rue and correct. To the extent that these statements are not based 'on my personal knowledge, they are based upon iaformation provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliabk.
a-
/
Subscribed and sworn before me, a Netary Public in and for the State of Maryland and County of AAL1/6/dT
.this /1 dayof Chruarw.1999.
.)
ALII 5a/
Ab WITNESS my Hand and Notarial Seal:
OAc d ' N'otary Public 'F
' i,
~
~My Commission Expires:
kF ru4 m I, d 861
'Date' CHC/JMO/ dim i
Attachment:
(1) Response to Specific Informelon Needed for the NRC Evaluation of Environmental Qualification for License Rer.ewal cc:
R. S. Fleishman, Esquire C. I. Grirr..:s, NRC J. E. Silberg, Esquite D. L. Solorio, NRC S. S. Bajwa, NRC Resident inspector, NRC A. W. Dromerick, NRC R. I. McLean, DNR H. J. Miller, NRC J. H. Walter, PSC i
l
^
i ATTACHMENT (1) i RESPONSE TO SPECIFIC INFORMATION NEEDED FOR THE NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant February 19,1999
ATTACIIMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR TIIE NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL NRC Reauest No. I Describe the procedures that are used to control the use of the " sound reasons to the contrary" alternative for equipment replacement, or commit to maintain such procedures in accordance with Regulatory Guide 1.89, Revision 1.
BGE Response 6
As described in Baltimore Gas and Electric Company's (BGE's) License Renewal Application (LRA)
Section 6.3.2.3 (6.3, Environmental Qualification), the 10 CFR 50.49(l) requirement to upgrade environmental qualification FQ) of replacement EQ equipment, unless there are sound reasons to the contrary, as provided in NRC Regulatory Guide 1.89, Revision 1, " Qualification of Class IE Equipment for Nuclear Power Plants," is part of our current licensing basis. Updated Final Safety Analysis Report (UFSAR) Section 7.12 provides a discussion of Calvert Cliffs Nuclear Power Plant's l
(CCNPP's) 10 CFR 50.49 (EQ) Program. As part of this discussion, the 10 CFR 50.49 requirement to upgrade equipment, unless sound reasons to the contrary exist, is already addressed. Internally, the procedures that control this requirement are EN-1-103, " Control of 10 CFR 50.49 Environmental Qualification of Electrical Equipment," and PM-1-101, " Procurement and Control of Items and Services for Calvert Cliffs."
As described in BGE's LRA Section 6.3.2.2, EN-1-103 is the control procedure for the EQ Program.
Requirements and responsibilities for upgrading replacement equipment are outlined in this
- procedure. The implementing procedure for upgrading replacement equipment is PM-1-101. This is a procurement procedure that controls the preparation and review of procurement requirements for replacement items, including EQ equipment replacements.
Engineering Standard ES-024, "10 CFR 50.49 Environmental Qualification Program," provides the sound reasons to the contrary criteria, extracted from NRC Regulatory Guide 1.89, Revision 1, utilized in the PM-1-101 process.
i NRC Reguest No. 2 Describe the processfor refurbishment as it applies to EQ. Describe the procedural controls that BGE relles upon to ensure that replacement components or groups ofdevices are returned to their original quahfiedcondition.
BGE Response The purpose / process of refurbishment is to restore designated (qualified) life to the EQ component, by replacement of materials and/or sub-components, of the EQ component, which have been identified in the EQ documentation as life limiting / susceptible to significant degradation due to aging.
For all qualified equipment that have established qualified lives, based on preconditioning, refurbishment, by replacing certain materials of construction and/or sub-components, is based on the level of preconditioning performed on the equipment. For certain Division of Operating Reactors I
(DOR) Guidelines and NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," Category Il qualified equipment that were not pre-conditioned, DOR Guidelines Sections 5.2.4 and 7.0; NUREG-0588 Section 4(2) and NRC Generic Letter 82-09,
" Environmental Qualification of Safety Related Electrical Equipment," provide NRC l
criteria / guidance for refurbishment, based on known susceptibility to aging degradation, results of inspections, or manufacturer's recommendations.
l J
i ATTACHMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR Tile NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL The refurbishment option can best be applied to, but not limited to, large EQ components, such as motors, which can be rewound with a new qualified insulation system, thus re-establishing the motor's designated (qualified) life.
The procedural controls to ensure that components or groups of devices are returned to their original qualified condition, are discussed in our LRA Section 6.3.2.2.
Specifically, the qualification maintenance requirement sheets are utilized in the planning of the replacement of EQ components to ensure that the replacement components are installed in accordance with EQ requirements.
NRC Reauest No. 3 Describe how BGE would intend to incorporate such an ongoing quahfication program [1EEE 323-1974, Section 6.6 (1) or (2)] into the licensing basisfor Calvert Chffs at some time in thefuture.
]lGE Response The option of ongoing qualification is part of our current licensing basis. 10 CFR 50.49(e)(5) stipulates that, "The equipment must be replaced or refurbished at the end of this designated life i
unless ongoing qualification demonstrates that the item has additional life." Although we have not previously used ongoing qualification, we may choose to use it in the future.
Regulatory Guide 1.89, Revision I describes processes and methods acceptable to the NRC for l
complying with 10 CFR 50.49. This regulatory guide provides NRC endorsement of the processes l
and methods described in IEEE 323-1974 for satisfying 10 CFR 50.49 requirements. This regulatory l
guide provides supplemental information, to that contained in IEEE 323-1974, which idemifies additional processes and methods for compliance with 10 CFR 50.49. Baltimore Gas and Electric Company believes that ongoing qualification as defined in IEEE 323-1974 Section 6.6, and in NRC Regulatory Guide 1.89, Revision 1, Section C.S.d can be utilized.
i Updated Final Safety Analysis Report Section 7.12 provides a description of the EQ Program, along with identifying the pertinent NRC EQ requirement and guidance documents. Ongoing quali5 cation, as described in 10 CFR 50.49 (e)(5) and expanded upon in NRC Regulatory Guide 1.89, Revision 1, is part of our current licensing basis. No changes to our licensing basis are necessary to support j
utilizing these options now or in the renewal period.
Additionally, equipment condition monitoring is another area of ongoing qualification that is acceptable per both 10 CFR 50.49 and Regulatory Guide 1.89, Revision 1. As you are aware, there is international activity in this area, especially at it relates to cable aging. The NRC is presently conducting research in this area to determine techniques for cable condition monitoring. As the technologies / techniques become viable, their use, as a method of ongoing qualification, could be
[
pursued.
NRC Reguest No. 4a Identify tf any changes have been made to the underlying assumptions affecting equipment quahfied hfe related to the normal thermal and radiation environment in which the equipment operates. Describe the process by which such changes will be made in thefuture. (This questions has been modified as agreed to at the January 22,1999 conference call. Baltimore Gas and Electric Company, by NRC direction, has 2
ATTACHMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR THE NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL not addressed the issue of changes to the mechanical, electrical, or chemical environment in this response.]
BGE Respanac Normal Thermal Environn.lcul The establishment of expected plant operating ambient temperatures, either design maximums or conservative bounding values, should be representative of that which is expected to be seen by the component during its installed life. NUREG-0588 Section 4(10), and Regulatory Guide 1.89, Revision 1, Section C.S.b specify this criteria.
The underlying assumption for this parameter is that the ambient temperature (s) utilized in the 3
qualified life calculation are representative. As discussed in BGE's LRA Section 6.3.3, design maximum temperatures, for both inside and outside Containment, were generally utilized to perform the current qualified life calculations / evaluations for the EQ time-limited aging analyses (TLAAs).
Maximum design temperatures of various plant ventilation systems, both inside and outside Containment, are described in UFSAR Section 9.8. To ensure that the general areas covered by these ventilation systems are maintained at or below their respective design maximum values, the Containment and Auxiliary Building are instrumented with temperature sensors, installed in locations representative of the general areas. Corresponding a!sms/ indication exist, which have alarm setpoints and proceduralized operator actions established to ensure that actions are taken to prevent / minimize temperature excursions above the design maximums. These monitoring / alarm systems have been installed since initial operation. We will continue to rely on them to ensure continued compliance with CCNPP's design basis, maximum ambient temperatures, into the renewal period.
Containment normal operating temperatures are monitored and maintained at or below design maximums to ensure continued compliance with CCNPP Technical Specification 3.6.5, and indirectly maintaining qualified life underlying assumptions.
Monitoring frequencies of containment temperatures are dictated by the Technical Specifications.
Excursions above the Technical Specification containment maximum temperatures are stringently controlled within the corresponding Technical Specification Limiting Condition for Operation.
Auxiliary Building (outside Containment) normal operating temperatures are maintained at or below design maximums to ensure compliance with design maximum temperatures, identified in UFSAR Section 9.8, and indirectly maintaining qualified life underlying assumptions. Excursions above these design maximum temperatures are controlled within the CCNPP corrective action process, and as discussed above.
In response to NRC Information Notice 89-30, "High Temperature Environments at Nuclear Power Plants," and Information Notice 89-30, Supplement 1, utilizing CCNPP's corrective action process, BGE performed a review of plant areas to identify high temperature environments, as described in the Notices. As part of this review, periodic temperature surveys were conducted and results recorded.
The results of this review and past plant surveys and operating experiences identified specific plant areas, both inside and outside Containment, where there are high temperature conditions greater than l
the general area temperatures. The temperatures in these areas have been quantified and incorporated i
3
r ATTACIIMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR T11E NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL into the EQ Program as well as the design change process. Engineering Standard ES-014 lists the normal operating temperatures for these specific plant areas.
In many cases, we concluded that design maximum temperatures are conservative representations of expected ambient operating temperatures. Various areas of the plant, both inside and outside Containment, have been periodically monitored during the last 10 years, with temperature data being collected to identify and quantify these conservatisms. This effort was undertaken primarily to support lengthening of short-lived EQ component qualified lives, to coincide with refueling / system outage schedules, to minimize impact on normal plant operations.
Normal plant operating temperature data collection methods employed over the years have oeen evolutionary. Data colletion has been performed by a number of techniques including, temperature dots installed in plant areas /on equipment, hand-held temperature devices, infrared surveys, automatic temperature data collection equipment ( i.e., Logic Beach Bitlogger) and plant installed components (i.e., Technical Specification containment temperature monitors / Operations logs). Typically, data reduction was done graphically, overlaying all the collected data and developing yearly plant area profiles for various plant area both inside and outside Containment. Each plant area profile consists of one maximum temperature established for each of the 12 months (12 total).
The highest temperature observed / collected at any time during each month was selected as the maximum temperature for that month. Additional margin was added to each of these monthly maximum temperatures. These 12 con >ervatively established monthly temperatures were then assembled into a yearly plant area profile. These yearly profiles were conservatively developed and would envelop minor fluctuations in temperature caused by operational changes in the plant.
The resultant plant area profiles are identified in ES-014. These conservatively established yearly profiles were input into the thermal qualified life calculations of various short lived EQ components, utilizing the degradation-weighted average temperature methodology', to generate revised qualified lives. The application of the equipment (i.e., de-energized versus energized) was also accounted for.
Utilizing these yearly profiles, along with the degradation-weighted average temperatme methodology to reassess EQ TLAAs, will be a primary method of extending their existing 40-year thermal qualified lives.
As described in BGE's LRA Section 6.3.2.2, EN-1-103 is the controlling procedure for the CCNPP EQ Program. Responsibilities outlined in this procedure include the maintenance of ambient er.vironmental monitoring and collection program to support qualified life re-evaluations, including the EQ TLAAs. Specifically, Plant Engineering Sectinr Lundeline (PEG)-11, "EQ Temperature Monitoring Program," controls this program. Nriodic monitoring, in accordance with PEG-11, will continue to, as a minimum. r -vniate that the conservatively established yearly area profiles discussed above remain valid.
1 Based on the above discussion, coupled with the fact that CCNPP has not implemented any major modifications nor experienced any events of sufficient duration that revised these maximum design temperatures, it is concluded that: 1) the design maximum temperatures, as described in the UFSAR Section 9.8, have been maintained since initial licensing; 2) as discussed above, the yearly plant area j
Describcd in EPRI Nuclear Power Plant Equipment Qualdication Reference Manual. Copyright 1992 4
l
- 1 ATTACHMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR TIIE NRC EVALUATION OF l
ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL profiles established are valid and applicable, since initial operation; and 3) that localized area hot spots have been identified. The use of these temperatures in qualified life calculations / evaluations are representative of that which is expected to be seen by the component during its installed life.
The above provides examples of current BGE practices that are envisioned to be utilized to reassess the EQ TLAAs. Other technicallyjustifiable analytical and data collection methods could be utilized in the future, depending on specific needs. Typical analytical methods and data collection and reduction methods for reanalyzing equipment thermal qualified life are outlined in Electric Power Research Institute (EPRI) TR-104873, " Methodologies and Processes to Optimize Environmental Qualification Replacement Intervals." The examples and discussions in this EPRI document are representative of the various approaches currently available to BGE and the ind'istry to extend thermal qualified lives.
Normal Radiation Environment 10 CFR 50.49.c(4) stipulates,"The radiation environment must be based on type of radiation, the total dose expected during normal operation over the installed life of the equipment..." Similar regulatory language exists in NUREG-0588 Section 1.4, with clarification in NRC Staff response to public comment No. 22 (NUREG-0588, Revision 1), Section 4.1.2 of DOR Guidelines and Regulatory Guide 1.89, Revision 1, Section C.2.c. The underlying assumption for this parameter is that the normal radiation doses used in the radiation qualified life calculations / evaluations are normally expected or representative.
Normal operating,40-year inside Containment, design maximum radiation doses were originally used in the EQ Program, developed in response to NRC Inspection and Enforcement Bulletin 79-01B,
" Environmental Qualification of Class IE Equipment." These normal operating design radiation doses were obtained from the Nuclear Steam Supply System vendor's (Combustion Engineering) design specifications.
In response to NRC Information Notice 93-39, " Radiation Beams from Power Reactor Biological Shields," utilizing the CCNPP corrective action process, BGE performed periodic (one fuel cycle) monitoring of the actual containment doses. Based on this monitoring, revised maximum,40-year, radiation doses, greater than the original design values, were established and integrated into the EQ Program, as well as the design change process. Engineering Standard ES-014 lists the revised 40-year, normal inside Containment doses for these specific plant areas.
Normal operating,40-year outside Containment, design maximum radiation doses were not originally specified/ calculated for Calvert Cliffs during its initial licensing. Normal operating,40-year outside Containment, radiation doses were established for the EQ Program based on radiological survey records. Auxiliary Building monthly radiological surveys from 1985 were reviewed and the highest dose rate value recorded each month, for each plant aica, regardless of where in the room the dose i
was determined. These maximum monthly dose rate values were averaged, yielding an average dose rate value for each room. This average dose rate value was multiplied by 350,400 (the number of hours in 40 years) to obtain the total 40-year normal radiation dose. Engineering Standard ES-014 lists the revised 40-year, normal outside Containment doses for these specific plant areas.
Based on the above discussion, coupled with the fact that CCNPP has not implemented any major modifications that would increase these conservatively established normal 40-year radiation doses, it 5
l i
e..
~
ATTACHMENT (1) 1 RESPONSE TO SPECIFIC INFORMATION NEEDED FOR TIIE NRC EVALUATION OF l
ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL is concluded that: 1) the normal radiation doses derived, as discussed above, are valid and applicable since initial operation; and 2) that localized inside Containment radiation doses have been identified.
'Ihe use of these radiation dose values in qualified life evaluations are representative of that which is expected to be seen by the component during its installed life.
Normal operating,60-year doses, for both inside and outside Containment, will be developed, based on radiological survey records, which are procedurally controlled by Radiation Safety and additional i
confirmatory monitoring, as necessary. The 40-year radiation doses will be multiplied by 1.5 to reflect the additional 20 years or extended operation. These 60-year projections will then be utilized to evaluate the EQ TLAAs, in category 10 CFR 54.21(c)(1)(iii), to determine if existing EQ radiation qualification documentation envelops these projected 60-year doses. Revised qualified lives, based on these 60-year doses, added to the applicable accident doses, will then be defined.
The above provides examples of current BGE practices that are envisioned to be used to reassess the i
EQ TLAAs. Other technically justifiable analytical and data collection methods could be used in the future, depending on specific needs. Typical analytical methods for reanalyzing equipment radiation qualified life are outlined in EPRI TR-104873. The examples and discussions in this EPRI document are representative of the various approaches currently available to BGE and the industry to extend radiation qualified lives.
Plant Environmental Changes As discussed in LRA Section 6.3.2.2 and in greater detail above, plant design ambient environmental service condition information is identified in ES.014. This Engineering Standard identifies the harsh environment areas of the plant for loss-of-coolant accident and high energy line breaks, including accident radiation doses and chemical spray composition. In addition, this Engineering Standard identifica the design maximum normal operating ambient service conditions of these same harsh environment areas, including thermal and radiation hot spots.
Engineering Standard ES-014 is utilized, by Calvert Cliffs engineering personnel, to identify both normal operating and accident design ambient environmental service conditions. Changes in these environmental service conditions, either due to planned plant modifications or as a result of corrective I
actions to address as-found, non-conforming conditions, found in the plant, are controlled within the i
design change process, as discussed in BGE's LRA Section 6.3.2.2. Engineering Standard ES-014 is revised as part of this design change process. These environmental service condition changes are reviewed for impact on the EQ Program, including the EQ TLAAs, and required changes to the EQ documentation (i.e., EQ Files) are performed.
NRC Reguest No. 4b If BGE has made any changes to analytical methods or acceptance criteria, describe those changes so that they may evaluated as they were at the time ofinitiallicensing.
BGE Response Nuclear Regulatory Commission requirements and guidance regarding EQ have evolved since initial licensing of CCNPP. Baltimore Gas and Electric Company has followed arse requirements and guidance as they have been implernented by the NRC over time. In particular, the following EQ 6
i
4 ATTACHMENT (1)
RESPONSE TO SPECIFIC INFORMATION NEEDED FOR THE NRC EVALUATION OF ENVIRONMENTAL QUALIFICATION FOR LICENSE RENEWAL requirements and guidance contain the currently approved methodologies and acceptance criteria applicable to EQ:
NRC Bulletin 79-01B (DOR guidelines), including Supplements 1,2, and 3; NUREG-0588 (for Comment) along with staff responses to public comments, issued as e
NUREG-0588, Revision 1; 10 CFR 50.49 along with the Statements of Consideration and resolution of public comments, published with the issuance of 10 CFR 50.49, on January 21,1983; Regulatory Guide 1.89, Revision 1; e
various NRC Generic Letters and Information Notices; and e
individual NRC approvals of case-by-case issues have been granted within NRC Safety e
Evaluation Reports on EQ.
Our current licensing basis for EQ is listed in UFSAR Section 7.12.
Maintaining compliance with the requirements 10 CFR 50.49 and other ancillary NRC requirement and guidance documents (i.e., DOR Guidelines), as discussed above, ensures that methodologies and acceptance criteria, either stipulated or endorsed by the NRC, are complied with.
NRC Request No. 4c l
Describe the procedures that will be used to make such changes in thefuture. Ifno changes have been made or are planned, please so state.
l BGE Response i
In order to make changes to NRC EQ requirements and guidance, including methodologies and l
acceptance criteria contained within them, changes to UFSAR Section 7.12 may be required. Such l
changes require a 10 CFR 50.59 safety evaluation be performed to evaluate the change in i
methodology or acceptance criteria. Changes to the UFSAR are controlled by Administrative Procedure RM-1-104," Updating the Safety Analysis Report (UFSAR, USAR, TSB)."
No changes to currently approved methodology and/or acceptance criteria, including the use of Arrhenius methodology to evaluate thermal aging, have been made by BGE nor are any planned in the future, including the period of extended operation.
i l
1 7