IR 05000317/1999002
| ML20205E564 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/08/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205E554 | List: |
| References | |
| 50-317-99-02, 50-317-99-2, 50-318-99-02, 50-318-99-2, NUDOCS 9904050225 | |
| Download: ML20205E564 (15) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos: 50-317;50-318 License Nos: DPR-53; DPR-69 Report Nos: 50-317/99-02;50-318/99-02 Licensee: Baltimore Gas and Electric Company Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Location: Lusby, Maryland Dates: February 8 - 12,1999 Inspectors: Michael C. Modes, Team Leader Fred Bower, Resident inspector Calvert Cliffs Douglas Dempsey, Reactor Engineer, RI Herman Graves, Senior Structural Engineer, RES c-2udie Julian, Technical Assistant, Rll Paul Kaufman, Sr Reactor Engineer, RI Robert Prato, Mechanical Engineer, NRR Approved by: Jimi T. Yerokun, Acting Chief Engineering Support Branch Division of Reactor Safety i
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9904050225 990326 PDR ADOCK 05000317 G PDR
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l EXECUTIVE SUMMARY Except for the apparent omissions caused by the Baltimore Gas and Electric (BGE)
methodology in applying the requirements of 10 CFR 54.4(a)(3), as noted in the fire pump house >
and station blackout diesel generator building analysis, the NRC inspection team concluded the applicant properly implemented the scoping and screening methodology approved by the NRC in the final safety evaluation titled, " Final Safety Evaluation (FSE) Concerning the Baltimore Gas
& Electric Company Report Entitled ' Integrated Plant Assessment Methodology'", dated
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April 4,1996. The onsite references and documentation supporting the information in the
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license renewal application was in an auditable and retrievable form with only one discrepancy noted in the Cathodic Protection system evaluation and an incorrect notation in Table 3.3E-1 for the Containment structure. The NRC inspection team did not evaluate the credibility of the applicant's screening of heat exchangers due to the indeterminate nature of the NRC's position on the function. The apparent omissions caused by the BGE methodology are being classified l
as an inspector follow up item (IFI) and along with the evaluation of the screening of heat transfer functions are being reviewed by the NRC Office of Nuclear Reactor Regulatio ii
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Reoort Details License Renewal Inspection Report 99-02 111. Enuineerina E8 Miscellaneous Engineering issues E8.1 System and Component Level Scooino and Screenina Insoection Scope To verify BGE, the applicant, implemented the scoping and screening process, including the identification of systems and structures and the applicable system-/ structural-level functions, consistent with their methodology, the information presented in their license ;
renewal application (LRA) and the results of the staffs review as documented in the final i safety evaluation report titled, " Final Safety Evaluation (FSE) Concerning the Baltimore Gas & Electric Company Report Entitled ' integrated Plant Assessment Methodology'",
dated April 4,1996. To verify on-site information and documentation required by, or otherwise necessary to document compliance with, the provisions of the rule are being maintained in an auditable and retrievable form, and are maintained consistent with NRC approved guidance for license renewal, quality assurance requirements, and site-approved procedures by reviewing onsite reference documents for the system scoping and screening results, the component scoping and screening results, the component pre-evaluation results, and selected supporting portions of the aging management review reports. In order to consistently focus NRC inspection resources the following systems and components were chosen for closer evaluation:
System Scoping 13 KVAC 250 VDC Make Up Demineralizer ,
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Turbine Building Ventilation Cathodic Protection Fuel Handling Structure Scoping Fire Pump House Diesel Generator Number 2 Building Bay Weir Well Transformer Foundations System Screening Auxiliary Feedwater Salt Water Con ponent Cooling s20VAC 125 VDC Main Control Room Ventilation Diesel Generator Building Ventilation Safety injection
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j Component Supports Piping Supports
Cablin Reactor Coolant System Fire Protection Structure Screening Intake Primary Building ~ )
Auxiliary Building ]
Safety Related Diesel Generator Building j i
A deterministic approach is used by BGE, in keeping with the license renewal rule .
requirements, to conduct an integrated plant assessment to ensure the aging of systems, .
structures and components (SSC) are appropriately ma'1 aged. BGE uses evaluation tools to determine, on a system-by-system, whether the function of the system meets 1 any criteria of the rule.-- However, a plant specific Probabilistic Risk Assessment (PRA)
can be used as an effective tool to provide integrated insights into the plant design, resulting in an additional relative measure of overall plant safety. The systems, structures, and components chosen by the NRC team, for closer inspection, used insights gained from Calvert Cliffs plant specific PRA in establishing the importance of the SSC to plant safet In order to inspect the applicant's scoping process the NRC team selected five systems and three structures the applicant determined were not with in the scope of license renewal. The NRC team reviewed the design, function, and documentation of the selected systems to determine, independently, if there was reasonable assurance that systems and structures meeting the criteria of 10 CFR 54.4 were identifie I in order to inspect the applicant's screening process the NRC team selected seventeen l
SSCs the applicant determined were in the scope of license renewal. The NRC team i reviewed the evaluation boundaries, intended functions, and active / passive or short/long lived characteristics of the seventeen SSCs. The NRC team reviewed the design, function, documentation, and the applicants screening of each SSC to determine if the process was implemented consistent with the rule and BGE's methodolog b. Observations and Findinas System and Structure Scooina The applicant used engineering administrative procedure EN-1-301, " System Level
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Scoping," to identify those systems and structures within the scope of license renewa The applicant documented a summary description of each system and structure in the system and structure screening results. The NRC inspection team reviewed the l Updated Final Safety Analysis Report (UFSAR) and system and structure descriptions to j verify that the descriptions were appropriately summarize l i
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As defined by the applicant's scoping methodology the 13.8 KV system must reliably supply power to plant auxiliaries during normal and accident operation, and supply power directly to the Reactor Coolant Pump (RCP) buses and to the 4 KV system through voltage regulators and unit service transformers. A review, by the NRC inspection team, of electrical loads and the UFSAR verified that essential 4 KV loads are supplied by the emergency diesel generators and therefore 13.8 KV power is not required under the current licensing basis to fulfill the requirements of 10 CFR 54.4(a)(1) and (a)(2).
However a potential issue was identified la a Design Basis Event (DBE) in Cnapter 14,
" Safety Analysis," of the UFSAR, Section 14.4, " Excessive Load Event," Subsection 14.4.3 (a.k.a. DBE 4), " Core and System Performance," because " . non-emergency AC power is still available in the Excess Load Event, steam may be directed to the condenser after 10 minutes for controlled plant cool down." This indicates the RCPs are energized by the 13.8 KV non-emergency attemating current (AC) system during DBE 4 in contradiction to the previous informatio Further review showed the 13.8 KV non-emergency AC system is assumed to be available in DBE 4 in order to establish a bounding worst case condition in the loss of Non-emergency AC Power event (DBE 10). If the 13.8 KV non-emergency AC system is available in DBE 4 the steam generator motor driven auxiliary feedwater, condensate, condensate booster, heater drain, and circulating water system pumps remain to facilitate steam generator blowdown to the condenser by maintaining condenser vacuum. This potentially reduces offsite dose in DBE 4 and makes DBE 10 the bounding event for offsite dose calculations. The assumption made about 13.8 KV non-emergency AC availability only secondarily makes the 13.8 KV non-emergency AC system available to the RCPs. During analysis of DBE 4 no credit is given to the RCPs. The NRC inspection team could find no other intended function of the 13.8 KV non-emergency AC system that puts the system within the scope of license renewal. The applicant appropriately excluded the system from the scope of license renewal The 250 VDC system, as defined by the applicant's scoping process, is designed to supply power "to various plant backup lube oil and seal oil emergency pumps in case of )
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loss of auxiliary AC power or failure of the normal AC pumps." The UFSAR, Subsection 6.3.6.1, " Design Basis," states that "[t]here are no loads connected to the 250 VDC bus that are related to the functioning of ESF." The UFSAR, Figure No. 8-5, identifies the turbine emergency lube oil pumps, turbine emergency seal oil pumps and the steam generator feedwater turbine emergency lube oil pumps as the only loads supplied by the 250 VDC bus. Based on the intended functions of the 250 VDC system. the applicant ,
appropriately excluded the system from the scope of license renewa l
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The inspection team reviewed a common water treatment system description document for systems 8,21,22, and 37, Revision 0, dated 11/21/97. The makeup demineralizer system receives pretreated water from the plant supply wells and purifies it for use in various plant applications. The originally installed ion exchangers have been retired in place and the main purification system in use is a vendor supplied portable system. The system supplies pure water for reactor plant and secondary plant makeup. Based on the intended functions of the common water treatment system, the applicant appropriately excluded the system from the scope of license renewal l
For the turbine building ventilation the inspection team reviewad SD 033 Rev 0, UFSAR l section 9.8.2.4, UFSAR figure 9-208, and plant drawing 64-317, Rev 1. The Turbine ,
Building Ventilation system provides year round air conditioning to maintain the air i
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temperature between 60*F and 110*F in the turbine building. The ventilation system consists of 12 air supply fans,4 retum fans, and associated equipment and based on its intended function is correctly excluded, by the applicant, from the scope of license renewa The inspection team reviewed section 5.1.7 of the UFSAR and applicant specification 61-406-A SEC.101.1 SH.1 describing the cathodic protection system. The system uses a group of four rectifiers to apply direct current voltage to anodes buried in the ground around structures and buried pipe. The applied voltage prevents electrochemical corrosion of buried metallic structures and components. A member of the NRC inspection team toured the plant yards with the applicant's system engineer, inspected accessible portions of the cathodic protection system, and discussed the system history, performance and maintenance with the system engineer. The system engineer was quite knowledgeable about the system and was involved in its performance and maintenance. Many of the anodes have been recently replaced and several new anodes added. The applicant has excavated some buried pipe, performed visualinspection and found the pipe to be in operabie conditio I As part of the document review for this system the inspection team discovered that a foot note to Table 1, Revision 6, page 41 of the system and structures screening results tabulation refers to a confusing statement that appears to place 29 components from the Cathodic Protection System within the integrated plant assessment (IPA) for the Main Steam System. Because the Main Steam System and the Cathodic Protection System have no interaction the inspection team brought this to the attention of the applicant. The applicant investigated the basis for the note and reported the note was an error. Two of the components were fuses associated with Auxiliary Feedwater, three were fuses associated with Radiation Monitoring, and the remaining twenty four were a mixture of fuses and indicating lights associated with Main Steam. These mis-classifications were caused by errors in the system coding for the components in the applicants NUCLEIS computerized equipment data base. Thus components were coded as being in system 063 cathodic protection, when they were actually components for other systems. The applicant initiated corrective action Engineering Service Package (ESP) ES199800475 on April 2,1998 which documents these problems. At the conclusion of the inspection the corrective actions remained unfinished. The NRC inspection team did not find other
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examples of this type of mis-classification and concluded the error was isolated to this example. Except for the confusion caused by the incorrect component classification the applicant appropriately excluded the Cathodic Protection System from the scope of license renewa The applicant's system level scoping results, Table 2, page 11, concludes the Fuel Handling system is not within the scope of license renewal. The result did, however, l identify five plant systems related to the handling of new and irradiated fuel. They are l Cranes / Test Equipment, Fuel Handling, New Fuel Storage and Elevator, Refueling Pool,
! and Spent Fuel Storage. The function of this equipment is to receive new fuel and place I it in the refueling pool, remove spent fuel from the reactor and place it in storage in the
- spent fuel storage pool, and replace it with the new fuel. Because the only intended l passive functions subject to license renewal are structural in nature the applicant
! grouped these five systems into a commodity evaluation. The applicant determined the l
intended functions of the fuel handling equipment heavy load handling cranes were to provide structural and/or functional support to safety related equipment; provide structural and/or functional support to non-safety-related equipment whose failure could directly prevent success of safety-related functions; and support single-failure-proof '
l criteria for lifting heavy loads over the spent fuel pool. As a result of this analysis, the l applicant comp!!ed a list of 48 components that ne subject to an aging management i review to support license renewal. Although th 4 applicant appropriately excluded the l Fuel Handling System from the scope of the lic ense renewal the NRC inspection team 1
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agrees the 48 components should be subject t ) aging management revie The function of the fire protection pump house, as defined in the applicant's screening j tools, is "[t]o provide protection of the fire and jockey pumps and their control cabinets i from weather." In addition, the UFSAR, Subsection 9.9.7.2, " Fire Pumps and Distribution Piping," states:
"The diesel fuel oil tank is provided with a dike, sized to contain the entire content of the tank. Therefore, the layout and fire protection features installed in the pump house will provide ;
protection for the electrically driven fire pump from a fire involving l fuel oil." i During a plant tour, members of the NRC team observed that the fuel oil protection dike was an integral part of the pump house floor and the pumps were mounted to support ,
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pads which are independent from the overall structural foundation. Review of design documents verified these structural and support functions and the NRC inspection teams observations. As a consequence the NRC inspection team concluded the portions of the pump house supporting the dike wall, the dike wall itself, and the independent pads supporting the pumps should be added to the scope of license renewal under 10 CFR 54.4(a)(3) because "[a]Il systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48)," should be within the scope of the rule. In addition the dike, floor, foundation and pump pads perform functions specified under 10 CFR 54.4(a) without any moving parts or without a change in L
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configuration or properties. For this reason the NRC inspection team concluded the portions of the pump house that support the dike wall, the dike wall itself, and the independent pads that support the pumps should be added to the scope of components that require an aging management review. The NRC inspection team determined l however that the portion of the pump house providing protection against the weather does not meet any of the scoping criteria under 10 CFR 54.4(a), because the function of the pumps are not weather dependen Discussion with the applicant indicated they had included only the pump support pads within the scope of the application under the general category of component supports j and were requiring an aging management review. The applicant did not include the l pump house dike, floor and foundation within the scope of the application. The applicant l determined that the pump house did not meet the seven structural functions under l subsection 4.2.2 of their methodology, which include the following:
l Provide structural and/or functional support to SR equipment.
i Provide shelter / protection to SR equipment.
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' Serve as a PB or a fission product retention barrier to protect public health and safety in the event of any postulated DBE . Serve as a missile barrier.
, Provide structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions.
! Provide flood protection barrie . Provide a rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plan Because fire protection is included within the scope of the rule based on 10 CFR 54.4(a)(3), the NRC regulation for fire protection (10 CFR 50.48), the only structural j function identified in the applicant's methodology, that could have identified the dike wall, l flooring and foundation as being within the scope of the application is number 7.
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Exclusion of the dike, flooring, and foundation from the scope of the rule is consistent with the applicant's methodology and logic that these structural components did not strictly perform a fire barrier function, but the exclusion is not consistent with the rul The NRC, in originally reviewing BGE's rpplication, did not identify this inconsistenc The applicant is currently re-evaluating their position on the fire pump house and tentatively agreed to include the referenced structures in the scope of an aging j management revie A similar problem was identified with Diesel Generator Building 2. The Diesel Generator Building Number 2 (DGB2) functions, as defined by the applicant's scoping process, include the following: (1) provide housing for the station black-out (SBO) diesel generator p
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(D/G), (2) provide safe access to equipment for maintenance and operation requirements, (3) maintain DGB2 integrity and protect the SBO D/G from damage under i various design loads and missiles, and (4) . minimize inteaction between the SBO diesel !
generator building and the building containing the fourth safety related diesel generato In addition, the UFSAR, Subsection 5.6.6.2, " Design," states that the design code of {
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7 i record for DGB2 requires the consideration of likely weather-related external events (i.e.,
tomados). Because none of these functions are consistent with the seven structural i functions under subsection 4.2.2 of the BGE scoping and screening methodology, the i applicant did not include DGB2 in the scope of the rul !
l The building itself and the components within the building do not provide a safety related !
function or support such a function as described in 10 CFR 54.4(a)(1) and (a)(2),
respectively. However, the SBO D/G is included within the scope of the rule under 10 CFR 54.4(a)(3) because it performs functions necessary to demonstrate compliance with i the Fire Protect:on Rule (10 CFR 50.48) and plays an important part in the Station Blackout Rule (10 CFR 50.63).
i BGE has a dual SBO coping analysis. One analysis considers coping with a station black out for four hours and the second analysis considers coping for the less restrictive i one hour. The " Station Blackout Analysis" discusses reliance on the SBO diesel {
generator for the one-hour coping analysis when five DGs are available. However, !
i based on Regulatory Guide 1.155 and the NRC endorsed guidance in NUMARC 87-00 j which were used to determine the "P2" offsite power design characteristic, BGE is !
required to have a two-out-of-four D/G configuration and a more limiting four-hour coping analysis that does not rely on the fifth D/G. For these reasnns, the SBO D/G is not required under 10 CFR 50.63. It is important to note the license renewal scoping I documents were developed prior to installation of the fifth D/G (during the engineering j stage), and prematurely included the fifth D/G within the scope of the rule under 10 CFR j 50.63. This confusion adds to the complexity of the analysis of the SBO D/ !
BGE does, however, rely on the SBO D/G to supply loads necessary for safe-shutdown I under their fire protection program. For example, if a fire occurs in the Unit 1 Service j Water room, the SBO D/G may be required to power the Saltwater, Service Water, and Component Cooling pumps, which, in tum, are required to bring the plant to a safe shutdown condition. Because the SBO D/G is needed for safe-shutdown under the fire protection plan, it appears DGB2 should be included within the scope of the license 1 renewal rule because it performs the functions of structural support, missile protection, i and weather related protection against external events. This omission by the applicant follows the same logic discussed previously for the fire pump house. The applicant's scoping tools do not include support functions for implementing 10 CFR 54.4(a) (3).
Unlike the fire pump house components however the applicant disagrees with the NRC inspection team's conclusion the DGB2 should be within the scope of the license renewal rule. This matter will be canied forward as an inspector follow-up item (IFl 50-317-318/99-002-01). l The intake Baffle Wall, as defined by the applicant's scoping methodology, is designed to draw a large volume of water from the bottom strata of the bay at a low velocity and with minimal ecological impact. No additional functions were identified. The NRC team l
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evaluated this function and concluded it did not meet any of the criteria under 10 CFR 54.4(a) and the applicant appropriately excluded it from the scope of license renewal.
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l The NRC inspection team reviewed the applicant's scoping of alte transformers having separate foundations and conducted a physicalinspection of some of the transformer The NRC inspection team concluded no BGE safety related power supplies were directly connected to transformers that were supported by independent transformer foundation Therefore, BGE independent transformer foundations do not meet the scoping criteria r under 10 CFR 54.4(a) and the applicants methodology appropriately eliminated them
) from the' applicatio System and Structure Screenina l The applicant used engineering administrative procedure EN-1-302, " Component Level Scoping For Systems," to identify those components of a given system that were within the scope of license renewal. The intended functions identified in the system level scoping were used to determine the component level intended functions. The component level intended functions were in turn used to determine which components required an aging management review. In accordance with EN-1-302, function catalogs l were developed to identify the structures and components needed to perform each of the
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intended functions for the system or structure. These intended function catalogs were then sorted in alphanumeric order to create a list of each of the components and structures in the syste The NRC inspection team reviewed the system piping and instrumentation drawings l (P&lD) and the list of safety related equipment (Q-List) in the master equipment list
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(NUCLEIS) database for the selected systems. The NRC inspection team selected a
! sample of components from the P&lDs and Q-List and a sample of components from the l NUCLEIS database and verified all the sampled components were properly included in ;
l the component level scoping results. No concerns were identified. The component level l scoping for the selected systems (Auxiliary Feedwater, Salt Water, Component Cooling, i
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120 VAC,125 VDC, Main Control Room Ventilation, Diesel Generator Building Ventilation, Safety injection, Component Supports, Piping Supports, Cabling, Reactor Coolant System, Fire Protection) was acceptably implemented in accordance with the methodology approved in the applicant's FS I Structures reviewed during the screening portion of this site inspection include: the l Intake, Primary Containment, Auxiliary Building, and the Safety-Related Diesel l Generator Building. In addition to the seven screening criteria discussed previously,
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BGE established four structural component categories based on their design and materials. The categories are: a) concrete components, b) structural steel components, c) architectural components, and d) unique components. Within each of the above four categories individual structural component types are assigned to a category as appropriate, i.e. a) concrete components - columns, walls, and floor slabs, et BGE determined six out of seven of the structural intended functions listed above apply to the intake Structure, with number three not applicable, and all the functions are passive; perform 6d without moving pads or without a change in configuration or properties. Within the four structural categories, the applicant determined 27 structural component types contribute to at least one of the intake Structure intended functions.
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For efficiency in presenting the results of these evaluations BGE grouped the 27 individual component types into four groups. Group 1 includes caulking, sealants, and expansion joints; group 2 is fluid-retaining walls and slabs subject to aggressive chemical attack on concrete; group 3 is steel components subject to corrosion; and group 4 includes sluice gates subject to corrosion. A review of BGE's screening documents and a walk down of the Intake Structure reveals the screening activities were implemented consistent with the applicant's methodology. The list of structural components types considered in the evaluation boundary of the intake Structure appears to be complete and no deficiencies in the screening process were identifie BGE determined that Primary Containment consists of two categories of components, the Containment Structure category (beams, columns, walls and liners) and the l i Containment System category (penetrations, hatches, air locks, and instrumentation). l l BGE concluded all seven structural intended functions listed above apply to the Primary Containment Structure. In addition the Primary Containment serves two additional functions: 1) provide closure of containment air lock and access / egress hatches during a station blackout, and 2) maintain the functionality of electrical components as addressed by the EQ program. All the intended functions of the Containment Structure / System are !
considered by the applicant to be passive except the closure of the manually operated
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containment air lock. The NRC inspection team concurs with the applicant's screening of the Primary Containment categories, which require an aging management review, and the applicant's exclusion of the manually operated containment air loc BGE determined that all seven intended structural functions are applicable to the Auxiliary Building and Safety-Related Diesel Generator Building Structures (SRDG).
However, the NRC inspection team observed that in Table 3.3E-1 of the " Application for .
License Renewal" intended function number three which " serve as a pressure boundary l or a fission product retention barrier to protect public health and safety in the event of '
any postulated DBEs" was not applicable to the SRDG. This exception should be noted in the application. All of the intended functions for the Containment and SRDG i structures are considered passive. A total of 47 structural component types were determined to contribute to at least one of the structural intended functions. One unique ,
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component type, Pipe Encapsulations, was evaluated in the aging management review for the Main Steam System not the Containment Structure. For the 47 structural
' component types four general groups of designed / material components were established. The NRC inspection team concurs with the applicants conclusions for the i Auxiliary Building and SRDG.
! Conclusions Except for the apparent omissions caused by BGE's methodology in applying the requirements of 10 CFR 54.4(a)(3) for "[a)ll systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48)," as noted in the discussions on the fire pump house and SBO D/G, the NRC inspection team concluded the applicant properly implemented the scoping and screening methodology approved by the NRC in the final safety evaluation titled, " Final Safety Evaluation (FSE)
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l 10 l Concerning the Baltimore Gas & Electric Company Report Entitled ' Integrated Plant l Assessment Methodology'", dated April 4,1996. The onsite references and l documentation supporting the information in the license renewal application were in an auditable form with only one discrepancy noted in the Cathodic Protection sys'am evaluation and an incorrect notation in Table 3.3E-1. The NRC inspection team did not i evaluate the credibility of the applicant's screening of heat transfer functions due to the l indeterminate nature of agency position. This matter and the IFl for the SBO D/G are l under review by the NRC Office of Nuclear Reactor Regulatio l V. Mananoment Meetinas )
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l The NRC inspection team presented the results of the inspection to members of the a.oplicant's management at the conclusion of the inspection on February 12,1999. The applicant acknowledged the findings presented and restated their disagreement with the team's conclusions about the SBO D/G.
l The inspection team asked the applicant whether any materials examined during the inspection
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should be considered proprietary. No proprietary information was identifie ,
PARTIAL LIST OF INDIVIDUALS CONTACTED BGE B. Montgomery Director NRM B. Doroshuk Project Director LR Richard Heibel Manager NP P.Penn Sr. Engineer LR D.Shaw Sr. Engineer LR J. Osborne Regulatory Analyst NRM C.Yodes Sr. Engineer LR J. Rycyna Sr. Engineer LR M. Bowman Sr. Engineer LR NRC P.Kuo Section Chief, NRR C. Grimes Branch Chief, NRR W. Ruland Acting Deputy Director, RI
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INSPECTION PROCEDURES USED IP 71002 License Renewal Inspection Procedure PARTIAL LIST OF DOCUMENTS REVIEWED Calvert Cliffs Nuclear Power Plant Units 1 and 2 License Renewal Application
' License Renewal (ITLR) Screening Results Life Cycle Management / License Renewal Program, Component Level Scoping Results for the Component Cooling System (015)
Life Cycle Management / License Renewal Program, Component Level ITLR Screening Results for the Salt Water Cooling System (012)
Calvert Cliffs Nuclear Power Plant License Renewal Project, Aging Management Review for the .
Component Cooling (015)
Calvert Cliffs Nuclear Power Plant License Renewal Project, Aging Management Review for the Salt Water Cooling (012)
Calvert Cliffs Nuclear Power Plant License Renewal Project, Component Pre-Evaluation for the Component Cooling Water System (015)
Calvert Cliffs Nuclear Power Plar! License Renewal Project, Component Pre-Evaluation for the )
Salt Water System (012) - 1 Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure iTLR Scoping Results," Table 1, System 003 ,
Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR Scoping Results," Screening Tools Calvert Cliffs UFSAR Electrical Main Single Line Diagram FSAR FIG. No. 8-1 Diesel Generator Project Electrical Main Single Line Diagram Meter & Relay Diagrams 4KV Systems Unit Buses 12,13,15,16 Meter & Relay Diagrams 4KV Systems Unit Buses 22,23,25,26 Single Line Meter & Relay Diagrams 4KV Systems Unit Buses 12A,128,13A,13B,15 Single Line Meter & Relay Diagrams 4KV Systems Unit Buses 22A,228, Single Line Diagram Water Treatment 480V MCC 106WT Single Line Diagram Water Treatment 480V MCC 208WT l . Single Line Diagram Intake Structure 480V MC c a 107SW
" Accident flow Sheet Event 4, Excessive Load,' Drawing Number 61403SH0004 Single Line Diagram Intake Structure 480V MCC's 207SW & 222 Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR l_
Scoping Results," Table 1, System 016
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Calvert Cliffs Nuclear Power Plant Licenw Renewal Project, " System and Structure ITLR l - Scoping Results," Screening Tools l Single Line Diagram Vital 120 VAC & 125 VDC, Emergency 250 VDC, FSAR FIG. No. 8-5 l Table 8-6, " Rating and Construction of 250 Volt DC Emergency Pump System Components"
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Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR Scoping Results," Table 1, Fire Protection Pump House Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR Scoping Results," Screening Tools i
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Specification for Plant Fire Protection Pumping Equipment l Civil Drawings: Yard Building Foundation, Well Water Pretreatment Bldg. & Fire Pump House l Piping Plan & Sections, Miscellaneous Yard Foundation and Structures Sheet.1, Fire Pump l House Plans, Elevations, Details, Civil Standards l ' Appendix A to Branch Technical Position APCSB 9.5-1, " Guidelines for Fire
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Protection for Nuclear Power Plants Docketed Prior to July 1,1976
- Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR Scoping Results," Table 1, DGB2 Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR Scoping Results," Table 1, System 24 Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR
- Scoping Results," Screening Tools CCNPP Station Blackout Analysis Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR l Scoping Results," Table 1, System 108 l Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR j Scoping Results," Screening Tools -
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CCS-91-B75, ASEA Brown Boveri Calvert Cliffs Site Office letter on System 108 CCNPP Circulating Water System Description Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR l Scoping Results," Table 1, System 003 l Calvert Cliffs Nuclear Power Plant License Renewal Project, " System and Structure ITLR
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Scoping Results," Screening Tools Electrical Main Single Line Diagram FSAR FIG. No. 8-1 CCS-91-B75, ASEA Brown Boveri Calvert Cliffs Site Office letter relating to transformer foundations MN-1-319, Revision 2, Structure and System Walk downs Procedure NO-1-100, Revision 14, Conduct of Operations Fire Protection Aging Management Review, Revision 1
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, LIST OF AGRONYMS USED BGE Baltimore Gas and Electric DBE Design Basis Event D/G Diesel Generator DGB2 Diesel Generator Building Number Two ESF Engineered Safety Feature ESP _ Engineering Service Package FSE Final Safety Evaluation IFl Inspector Fellow Up item IPA - Integrated Plant Assessment LRA License Renewal Application P&lD Piping and Instrument Drawings PRA Probabilistic Risk Assessment RCP Reactor Coolant Pump .
SBO Station Black Out SRDG Safety Related Diesel Generator SSC Systems, structures, and Components UFSAR Updated Final Safety Analysis Report l
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