ML15026A538
| ML15026A538 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/07/2015 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML15023A514 | List:
|
| References | |
| Download: ML15026A538 (48) | |
Text
CHAPTER 9 CONDUCT OF OPERATIONS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION LEP 9-1 23 9.8-2 23 9-i 23 9-ii 23 9-iii 23 9-iv 23 9.1-1 15 9.1-2 15 9.2-1 8 9.2-2 8 9.2-3 8 9.3-1 22 9.3-2 8 9.4-1 20 9.4-2 22 9.4-3 20 9.4-4 20 9.4-5 22 9.4-6 22 9.4-7 20 9.4-8 22 9.5-1 8 9.6-1 23 9.6-2 23 9.6-3 23 9.6-4 23 9.6-5 23 9.6-6 23 9.6-7 23 9.6-8 23 9.6-9 23 9.6-10 23 9.6-11 23 9.6-12 23 9.6-13 23 9.6-14 23 9.6-15 23 9.6-16 23 9.6-17 23 9.6-18 23 9.6-19 23 9.6-20 23 9.6-21 23 9.6-22 23 9.6-23 23 9.6-24 23 9.7-1 23 9.8-1 23 CALVERT CLIFFS ISFSI USAR LEP 9-1 Rev. 23
CHAPTER 9 CONDUCT OF OPERATIONS TABLE OF CONTENTS PAGE 9.0 CONDUCT OF OPERATIONS 9.1-1 9.1 ORGANIZATIONAL STRUCTURE 9.1-1 9.1.1 CORPORATE ORGANIZATION 9.1-1 9.1.1.1 Corporate Functions, Responsibilities, and 9.1-1 Authorities 9.1.1.2 Applicants In-House Organization 9.1-1 9.1.1.3 Interrelationship with Contractors and Suppliers 9.1-1 9.1.1.4 Applicants Technical Staff 9.1-1 9.1.2 OPERATING ORGANIZATION, MANAGEMENT, AND 9.1-2 ADMINISTRATIVE CONTROL SYSTEM 9.1.2.1 On-site Organization 9.1-2 9.1.2.2 Personnel Functions, Responsibilities, and 9.1-2 Authorities 9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS 9.1-2 9.1.4 LIAISON WITH OTHER ORGANIZATIONS 9.1-2 9.2 PREOPERATIONAL TESTING AND OPERATION 9.2-1 9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING 9.2-1 TEST PROGRAM 9.2.2 TEST PROGRAM DESCRIPTION 9.2-1 9.2.2.1 Physical Facilities and Operations 9.2-1 9.2.3 TEST DISCUSSION 9.2-2 9.3 TRAINING PROGRAM 9.3-1 9.3.1 PROGRAM DESCRIPTION 9.3-1 9.3.1.1 Training for ISFSI Operations Personnel 9.3-1 9.3.1.2 Training for Maintenance Personnel 9.3-1 9.3.1.3 Training for Health Physics Personnel 9.3-1 9.3.1.4 Training for Security Personnel 9.3-1 9.3.2 RETRAINING PROGRAM 9.3-1 9.3.3 ADMINISTRATION AND RECORDS 9.3-2 9.4 NORMAL OPERATIONS 9.4-1 9.4.1 ADMINISTRATIVE CONTROLS 9.4-1 9.4.1.1 Qualification of Spent Fuel 9.4-1 9.4.1.2 Spent Fuel Identification 9.4-3 CALVERT CLIFFS ISFSI USAR 9-i Rev. 23
CHAPTER 9 CONDUCT OF OPERATIONS TABLE OF CONTENTS PAGE 9.4.2 RECORDS 9.4-3 9.5 EMERGENCY PLANNING 9.5-1 9.6 ISFSI LICENSE RENEWAL ACTIVITIES 9.6-1 9.6.1 AGING MANAGEMENT REVIEW 9.6-1 9.6.2 TIME-LIMITED AGING ANALYSIS 9.6-1 9.6.2.1 DSC Time-Limited Aging Analysis 9.6-1 9.6.2.2 HSM Time-Limited Aging Analysis 9.6-3 9.6.2.3 Transfer Cask Fatigue Evaluation 9.6-5 9.6.2.4 Time-Limited Aging Analysis of the Transfer 9.6-6 Cask Lifting Yoke 9.6.3 AGING MANAGEMENT PROGRAMS 9.6-6 9.6.3.1 HSM Aging Management Program 9.6-6 9.6.3.2 Transfer Cask Aging Management Program 9.6-7 9.6.3.3 Transfer Cask Lifting Yoke Aging Management 9.6-7 Program 9.6.3.4 Cask Support Platform Aging Management 9.6-7 Program 9.6.3.5 DSC External Surfaces Aging Management 9.6-7 Program 9.6.3.6 High Burnup Fuel Aging Management Program 9.6-8 9.7 DECOMMISSIONING PLAN 9.7-1
9.8 REFERENCES
9.8-1 CALVERT CLIFFS ISFSI USAR 9-ii Rev. 23
CHAPTER 9 CONDUCT OF OPERATIONS LIST OF TABLES TABLE PAGE 9.4-1 POST-DISCHARGE COOLING TIME 9.4-5 9.4-2 NUHOMS-24P BURNUP CURVE DATA 9.4-7 9.4-3 ADDITIONAL COOLING TIME REQUIREMENTS FOR 9.4-8 LOADING ASSEMBLIES WITH IRRADIATED STAINLESS STEEL INERT REPLACEMENT RODS IN A 32P DSC 9.6-1 AGING MANAGEMENT REVIEW RESULTS FOR THE 9.6-9 IRRADIATED FUEL ASSEMBLIES 9.6-2 AGING MANAGEMENT REVIEW RESULTS FOR THE DSCS 9.6-11 (NUHOMS-24P AND 32P) 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM 9.6-13 9.6-4 AGING MANAGEMENT REVIEW RESULTS FOR THE 9.6-18 TRANSFER CASK 9.6-5 AGING MANAGEMENT REVIEW RESULTS FOR THE 9.6-22 TRANSFER CASK LIFTING YOKE 9.6-6 AGING MANAGEMENT REVIEW RESULTS FOR THE CASK 9.6-24 SUPPORT PLATFORM CALVERT CLIFFS ISFSI USAR 9-iii Rev. 23
CHAPTER 9 CONDUCT OF OPERATIONS LIST OF ACRONYMS ASME American Society of Mechanical Engineers BGE Baltimore Gas and Electric Company CCNPP Calvert Cliffs Nuclear Power Plant CFR Code of Federal Regulations DSC Dry Shielded Canister ERP Emergency Response Plan HSM Horizontal Storage Module HSM-HB High Burnup Horizontal Storage Module IFA Irradiated Fuel Assembly ISFSI Independent Spent Fuel Storage Installation NFSS Nuclear Fuel Services Section NUHOMS Nutech Horizontal Modular Storage SFSP Spent Fuel Storage Project SNM Special Nuclear Material TLAA Time-Limited Aging Analyses UFSAR Updated Final Safety Analysis Report CALVERT CLIFFS ISFSI USAR 9-iv Rev. 23
9.0 CONDUCT OF OPERATIONS 9.1 ORGANIZATIONAL STRUCTURE 9.1.1 CORPORATE ORGANIZATION The Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI) is operated under the same corporate management organization responsible for operation of the Calvert Cliffs Nuclear Power Plant (CCNPP). This organization is described in Section 12.1 of the CCNPP Updated Final Safety Analysis Report (UFSAR).
9.1.1.1 Corporate Functions, Responsibilities, and Authorities The corporate organization described in UFSAR Section 12.1, provides line responsibility for operation of Constellation Generation Group, LLC.
Various departments within the Company have responsibility for design, construction, quality assurance, testing, and operation of CCNPP as well as the ISFSI. Constellation Generation Group, LLC corporate functions, responsibilities, and authorities for quality assurance, as described in Chapter 11, are applicable for appropriate portions of the ISFSI.
9.1.1.2 Applicant's In-House Organization Constellation Generation Group, LLC's, Nuclear Fuel Services Section (NFSS) has the specific responsibility for design of structures and systems, specifications, and procurement of materials and equipment, and preparation of construction and installation drawings for the ISFSI. The Director-NFSS has overall responsibility for the design of CCNPP including the ISFSI. The Nuclear Fuel Management Unit of this section maintains responsibility for management of spent fuel.
Calvert Cliffs Nuclear Power Plant Department is responsible for operation and maintenance of CCNPP. This department provides general supervision and technical management services for the plant and is responsible for ISFSI maintenance.
9.1.1.3 Interrelationship with Contractors and Suppliers The prime contractor for design and analysis of the Calvert Cliffs ISFSI is Transnuclear (formerly Nutech Engineers, Inc.). The ISFSI is owned and operated by CCNPP. Construction of the ISFSI was the responsibility of an approved construction contractor. Licensing support, Geotechnical Engineering and Quality Assurance Program revisions were performed by Duke Engineering and Services utilizing Duke Power Company personnel experienced on the Oconee Nuclear Station ISFSI. Subsurface investigations at the ISFSI site were performed by Law Engineering Testing Company.
9.1.1.4 Applicant's Technical Staff The Corporate technical staff supporting the ISFSI is described in UFSAR Section 12.1.
CALVERT CLIFFS ISFSI USAR 9.1-1 Rev. 15
9.1.2 OPERATING ORGANIZATION, MANAGEMENT, AND ADMINISTRATIVE CONTROL SYSTEM 9.1.2.1 On-site Organization The on-site organization of CCNPP is responsible for operation of the ISFSI. The Nuclear Fuels Management Unit of the NFSS maintains primary responsibility for spent fuel storage. The organization for CCNPP is fully described in UFSAR Section 12.1.
9.1.2.2 Personnel Functions, Responsibilities, and Authorities The functions, responsibilities, and authorities of major personnel positions, including discussions of specific succession of responsibility for overall operation of CCNPP are described in UFSAR Section 12.1. These functions, responsibilities, and authorities extend to the Calvert Cliffs ISFSI.
9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS The minimum qualification requirements for major operating, technical and maintenance supervisory personnel, as well as the qualifications of persons assigned to managerial and technical positions, are as stated in UFSAR Section 12.1.
9.1.4 LIAISON WITH OTHER ORGANIZATIONS Arrangements made with outside organizations are as described in Section 9.1.1.3 of this Chapter.
CALVERT CLIFFS ISFSI USAR 9.1-2 Rev. 15
9.2 PREOPERATIONAL TESTING AND OPERATION Prior to operation of the ISFSI, complete functional tests of the in-plant operations, transfer operations, and horizontal storage module (HSM) loading and retrieval were performed. These tests verified that the storage system components [e.g., dry shielded canister (DSC), transfer cask, transfer trailer, etc.] can be operated safely and effectively (Reference 9.12).
9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAM Preoperational testing was governed by Calvert Cliffs administrative procedures for conducting testing.
9.2.2 TEST PROGRAM DESCRIPTION The testing program required use of a DSC, DSC mock-up, transfer cask and associated handling equipment, transfer trailer, and an HSM. The tests simulated, as nearly as possible, the actual operations involved in preparing a DSC for storage and demonstrated that they can be performed safely during actual emplacement of irradiated fuel assemblies (IFAs) in the ISFSI. Shielding verification, which is not completely achievable during dry runs, was accomplished during the initial IFA loadings.
9.2.2.1 Physical Facilities and Operations 9.2.2.1.1 Dry Shielded Canister and Associated Equipment An actual DSC and a part-length mock-up of a DSC were used for preoperational testing. The DSC was loaded into the transfer cask to verify fit and suitability of the DSC lift rig.
Additionally, the DSC was used in operational testing of the transfer equipment and HSM.
The part-length mock-up was configured exactly as the top end of the DSC with lead shield plug and covers. The mock-up was used for checkout of the automated welding equipment including actual welding of the simulated lead shield plug and top cover plate. Emphasis was placed on acceptability of the weld, as well as compliance with approved as low as reasonably achievable practices.
9.2.2.1.2 Transfer Cask and Handling Equipment Functional testing was performed with the transfer cask and lifting yoke. These tests demonstrated that the transfer cask can be safely transported from the Auxiliary Building truck bay to the cask washdown pit. From there, it was placed into the spent fuel pool to verify clearances and travel path.
9.2.2.1.3 Off-Normal Testing of the DSC and Transfer Cask In the unlikely event that a problem arises during actual loading of the IFAs into the DSC, seal welding of the DSC, or during emplacement of a loaded DSC into an HSM, no CALVERT CLIFFS ISFSI USAR 9.2-1 Rev. 8
immediate action would be required since the fuel assemblies would be in a safe condition. The pre-operational testing program confirmed that the IFAs can be safely removed from the DSC by demonstrating that the DSC lids can be removed.
9.2.2.1.4 Transfer Trailer and HSM The transfer cask was placed on the transfer trailer, and then transported to the ISFSI and aligned with an HSM.
Compatibility of the transfer trailer with the transfer cask, negotiation of the travel path to the ISFSI, and maneuverability within the confines of the ISFSI were verified.
The transfer trailer was aligned and docked to the HSM. The hydraulic ram was used to emplace a DSC loaded with test weights in the HSM and remove it. Loading of the DSC into the HSM verified that the transfer skid alignment system, hydraulic positioners, and ram grapple assembly all operate safely for both emplacement of a DSC into, and removal from, a HSM.
9.2.2.1.5 Off-Normal Testing of the Transfer Trailer and HSM In the unlikely event that a problem should occur that prevents loading the DSC into the HSM, no immediate remedial action will be required. Irradiated fuel assemblies may be stored in the transfer cask while corrective action is taken.
The most severe condition would occur if a failure of the hydraulic ram, after partial insertion of a DSC into an HSM, were to prevent complete emplacement of the DSC.
(Radiological shielding and decay heat removal are not compromised by this condition, but the transfer trailer may not be moved away until the DSC is completely within the confines of either the transfer cask or the HSM.) Pre-operational testing verified that reversal of DSC movement can be completed by the operator of the hydraulic ram.
9.2.3 TEST DISCUSSION The purpose of the preoperational tests was to ensure that a DSC can be properly and safely placed in the spent fuel pool, loaded with IFAs, transported to the ISFSI, emplaced in the HSM, and removed from the HSM. Proper operation of the DSC, transfer cask, and transfer trailer, as well as the associated handling equipment (e.g., lifting yoke, welding equipment, vacuum drying equipment), provided such assurance.
Preoperational test requirements were specific. Detailed procedures were developed and implemented by Calvert Cliffs personnel who were responsible for ensuring that the test requirements were satisfied.
CALVERT CLIFFS ISFSI USAR 9.2-2 Rev. 8
The expected results of the preoperational tests were the successful completion of the following: loading of a DSC into the transfer cask, seal welding of the mock-up DSC, placement of a DSC into the transfer cask into and out of the spent fuel pool, transporting the transfer cask loaded with a DSC and test weights to the ISFSI, and emplacement in an HSM and removal from an HSM. The tests were deemed successful since the expected results were achieved safely and without damage to any of the components or associated equipment.
Any equipment or components which required modification in order to achieve the expected results were retested to affirm that the modification was sufficient. If any pre-operational procedures were changed in order to achieve the expected results, the changes were incorporated into the appropriate operating procedures.
Power operation of CCNPP was not affected by testing of the storage system, and in-plant testing was conducted concurrently with plant operation. In-plant testing was conducted entirely within the Auxiliary Building, and was scheduled so that there was no conflict with refueling. All normal prerequisites for safe handling of components in, or near, the spent fuel pool were satisfied, and normal safety and radiological practices were employed.
CALVERT CLIFFS ISFSI USAR 9.2-3 Rev. 8
9.3 TRAINING PROGRAM All personnel working at the Calvert Cliffs ISFSI receive training and indoctrination geared toward providing and maintaining a well-qualified work force for safe and efficient operation of the ISFSI. The existing Calvert Cliffs training program, as described in CCNPP UFSAR Section 12.2, is used to provide this training and indoctrination. Additional sections have been added to this program to include information specific to the ISFSI.
9.3.1 PROGRAM DESCRIPTION 9.3.1.1 Training for ISFSI Operations Personnel Generalized training is provided to operations personnel in the applicable regulations and standards and in the nuclear engineering principles of cooling, radiological shielding, and structural characteristics of the DSC/HSM.
Detailed operator training is provided for DSC preparation and handling, fuel loading, transfer cask preparation and handling, and transfer trailer loading.
9.3.1.2 Training for Maintenance Personnel Generalized training is provided to maintenance personnel on the applicable regulations and standards and on the nuclear engineering principles of cooling, radiological shielding, and structural characteristics of the DSC/HSM/HSM-HB (high burnup horizontal storage module).
Specific training is provided for use of the automated seal welding equipment for the top end shield plug and top cover plate, operation of the transfer trailer, alignment of the cask skid with the HSM/HSM-HB, alignment of the hydraulic ram assembly, and normal and off-normal operation of the hydraulic ram. Specific training is also provided for cleaning of the HSM/HSM-HB air inlets and outlets.
9.3.1.3 Training for Health Physics Personnel Generalized training is provided to Health Physics personnel on the applicable regulations and standards and on the nuclear engineering principles of cooling, radiological shielding, and structural characteristics of the DSC/HSM/HSM-HB.
Specific training is provided in radiological shielding design of the system, particularly the top end shield plug, DSC/transfer cask and the DSC/HSM/HSM-HB.
9.3.1.4 Training for Security Personnel Details of the training program for security personnel are provided in the Security Plan which is withheld from public disclosure in accordance with Title 10, Code of Federal Regulations (CFR) 2.790(d) and 10 CFR 73.21.
9.3.2 RETRAINING PROGRAM Retraining is consistent with retraining requirements in effect at CCNPP for personnel involved in fuel handling operations.
CALVERT CLIFFS ISFSI USAR 9.3-1 Rev. 22
9.3.3 ADMINISTRATION AND RECORDS The organization responsible for training programs and for maintaining up-to-date records on the status of personnel training is the existing Nuclear Training Section at Calvert Cliffs.
CALVERT CLIFFS ISFSI USAR 9.3-2 Rev. 8
9.4 NORMAL OPERATIONS The Calvert Cliffs ISFSI utilizes the Nutech Horizontal Modular Storage (NUHOMS) system, which is completely passive during storage. Therefore, no monitoring instruments or limiting control settings are utilized at the ISFSI. Other limits and controls that are applied to the system during fuel loading and DSC transfer to the ISFSI are the fuel selection criteria, DSC surface contamination limits and DSC vacuum and helium backfill pressures, DSC closure weld examination requirements, and cask height restrictions during transport.
The components of storage, the DSC, the HSM, and (during transfer) the transfer cask, have been analyzed for all credible equipment failure modes and extreme environmental conditions.
No postulated event results in damage to fuel, release of radioactivity, or danger to the public health and safety. All operational equipment will be maintained, tested, and operated according to the implementing procedures developed for the ISFSI. The failure or unavailability of any operational component can result in delay in transfer of the DSC to the HSM, but will not result in an unsafe condition.
Under normal operations, the ISFSI provides for independent storage of spent fuel away from the CCNPP facilities. With the exception of some limited physical and continuous electronic security surveillance, the facility functions as a passive system once fuel has been loaded.
Loading of fuel assemblies into the facility, which occurs periodically, requires specific procedures that are separate from those of normal plant operations.
9.4.1 ADMINISTRATIVE CONTROLS Existing and proposed CCNPP organizational and administrative systems and procedures, record keeping, review, audit, and reporting requirements are used to ensure that the operations involved in the storage of spent fuel at the Calvert Cliffs ISFSI are performed in a safe manner. This includes both the selection of assemblies qualified for ISFSI storage and the verification of assembly identification numbers prior to and after placement into individual storage canisters.
9.4.1.1 Qualification of Spent Fuel Fuel assembly qualification is based on the requirements for criticality control, decay heat removal, radiological protection, and structural integrity.
Fuel assembly reactivity, radiological source strength, and decay heat removal capabilities are defined by three variables: (1) the initial enrichment of the unirradiated fuel assembly, (2) the final assembly burnup at discharge, and (3) the out-of-reactor cooling time. Table 9.4-1 (Reference 9.15) presents the minimum cooling time for each fuel batch to achieve the 0.66 kW decay heat limit. Table 9.4-2 represents the acceptance criteria for minimum spent fuel burnup as a function of initial enrichment for the NUHOMS-24P DSC. The NUHOMS-32P DSC has no limit on minimum spent fuel burnup, but assemblies containing irradiated stainless steel replacement pins have additional acceptance criteria presented in Table 9.4-3. The administrative procedures controlling these variables are as described below.
Procedures currently in place for special nuclear materials (SNMs) accountability and record keeping are used to verify initial fuel assembly enrichment and burnup levels at discharge. New fuel enrichments and initial uranium isotopics are recorded from the Department of CALVERT CLIFFS ISFSI USAR 9.4-1 Rev. 20
Energy/Nuclear Regulatory Commission Form 741s and stored in both a database file and on duplicate paper copies of the Form 741s. Individual fuel assembly burnups are also stored in the SNMs database. These values are generated by utilizing thermal energy production data determined by in-core flux mapping. Burnup and initial enrichment values from the SNM accountability records of each IFA are compared to the applicable limits to verify that the reactivity level is acceptable for DSC loading and storage. The enrichment vs. burnup method for reactivity verification will routinely be used, and required by procedures, for the NUHOMS-24P DSC. Calvert Cliffs Nuclear Power Plant reserves the right to rely on other Nuclear Regulatory Commission accepted analytical methods to qualify fuel assemblies in special cases.
Subcriticality in the NUHOMS-32P DSC is assured by limiting the initial enrichment of unirradiated fuel assemblies to 4.5 wt% U-235, by the presence of fixed neutron absorbing plates in the basket assembly, and by the presence of soluble boron in the spent fuel pool water.
For decay heat control, only those irradiated assemblies which do not exceed a decay heat level of 0.66 kW qualify for loading into the DSC. Due to Co-60 production, assemblies with stainless steel replacement rods require additional cooling time beyond the time at which they reach 0.66 kW as shown in Table 9.4-3. Decay heat loadings at or below this level ensure that peak fuel rod cladding temperatures are maintained within acceptable levels. Since individual fuel assembly decay heat levels are a function of both the discharge burnup and the cooling time, procedural controls are used to verify these parameters prior to fuel assembly loading.
For the Calvert Cliffs fuel design and operating histories, the cooling time necessary to achieve a 0.66 kW decay heat level is between 4 and 17 years. The variation in required cooling time is a very strong function of discharge burnup and a very weak function of initial enrichment. The cooling times of fuel discharged is presented in Table 9.4-1.
Specific qualification of the fuel assembly radiological source term is not necessary prior to fuel loading. Analysis shows that the reference source term used to generate the surface dose rate values found in Chapters 7 and 12 is not exceeded by any fuel assembly meeting the limiting conditions for cooling times specified in Table 9.4-1 and Table 9.4-3.
Therefore, assemblies that fall into the acceptance region of Table 9.4-2 qualify as candidates for ISFSI storage with the appropriate minimum cooling. Additional calculations relating reactivity (i.e., initial enrichment and discharge burnup) with decay heat and the required cooling time may be performed, as needed, to qualify future assemblies.
To ensure the structural integrity of the spent fuel to be loaded into the DSC, plant records of all known damaged assemblies are reviewed. A fuel assembly and component database has been compiled which incorporates previous sipping, ultrasonic testing, eddy current, and visual observation.
This database is examined as a part of the dry storage qualification process to verify that assemblies with known cladding breaches are not included.
CALVERT CLIFFS ISFSI USAR 9.4-2 Rev. 22
Fuel assemblies are also screened to ensure they conform to the requirements of Reference 9.14 to ensure the fuel assemblies with burnup
< 47,000 MWD/MTU contain no more than two vacancies in any location within a column or row. This allows 28 vacancies per assembly with burnup < 47,000 MWD/MTU. The vacancies do not need to be adjacent to one another. The analysis finds fuel assemblies that conform to this configuration to be structurally sound under all anticipated conditions.
Vacancies are restricted for fuel with burnup between 47,000 MWD/MTU and 52,000 MWD/MTU pending further structural analysis.
If the reactivity, decay heat, and structural integrity criteria are all met, then approval for dry storage for a given assembly is granted. This qualification is documented and subsequently referenced through ISFSI operating procedures prior to loading fuel into the DSC.
9.4.1.2 Spent Fuel Identification Administrative controls will be utilized to avoid fuel misplacement.
Information on fuel assembly qualification for dry storage will be documented and transmitted to fuel handling personnel. Prior to any transfer of a fuel assembly to the DSC, specific DSC loading procedures will require a review of assembly documentation. This will be followed by an independent visual verification of the assembly identification number.
These procedures ensure that the correct (approved) fuel assembly is being accessed and loaded into the DSC. As a final check, all assembly identification numbers will be visually checked and recorded after the DSC has been fully loaded.
9.4.2 RECORDS The ISFSI records are maintained in accordance with the requirements of 10 CFR Part 72. Procedures have been developed for use by the Spent Fuel Storage Project (SFSP) which meet the requirements of 10 CFR Part 72 for records retention during the construction phase of the project (Reference 9.5). Additional procedures have been developed to encompass the fuel loading and storage phases of the project.
For SNM accountability, the management system in place for Calvert Cliffs Units 1 and 2 has been expanded to allow record-keeping relative to storage of fuel at the ISFSI.
The requirements of 10 CFR 72.72, 10 CFR 72.74, 10 CFR 72.76, and 10 CFR 72.78 have been met by adding the ISFSI to our current system, which meets the equivalent requirements of 10 CFR 70.51, 10 CFR 70.52, 10 CFR 70.53, and 10 CFR 70.54, respectively. Horizontal storage module and DSC identification numbers, along with individual assembly locations within a DSC, are maintained in our SNM database consisting of core locations, spent fuel pool rack locations, etc. In this way, ISFSI SNM accountability requirements are met. Periodic physical inventory requirements are met by verifying that HSMs have not been tampered with since the previous inventory (References 9.9 and 9.11).
While 10 CFR 70.51 imposes a three year duration of records storage, by maintaining the ISFSI records for ISFSI lifetime plus five years the duration requirement of 10 CFR 72.72 is met. It is the intention of CCNPP to use the existing system for maintaining CALVERT CLIFFS ISFSI USAR 9.4-3 Rev. 20
records, ensuring that the stricter of the requirements of the various Parts of 10 CFR are met.
CALVERT CLIFFS ISFSI USAR 9.4-4 Rev. 20
TABLE 9.4-1 POST-DISCHARGE COOLING TIME
- NUHOMS-24P COOLING TIME TO MEET 0.66 kW DECAY HEAT LIMIT INITIAL ENRICHMENT BURNUP COOLING TIME**
(W/O U235) (MWD/MTU) (Years) 4.00 < 4.50 B.U. 47,000 10 3.50 < 4.00 45,000 < B.U. 47,000 11 42,000 < B.U. 45,000 10 B.U. 42,000 8 3.00 < 3.50 45,000 < B.U. 47,000 12 42,000 < B.U. 45,000 11 B.U. 42,000 9 2.50 < 3.00 45,000 < B.U. 47,000 13 42,000 < B.U. 45,000 12 39,000 < B.U. 42,000 10 B.U. 39,000 8 2.00 < 2.50 45,000 < B.U. 47,000 15 42,000 < B.U. 45,000 13 39,000 < B.U. 42,000 11 B.U. 39,000 9
- All assemblies loaded into DSC must meet the source spectra requirements of Technical Specification 2.1.
- These bounding cooling times may be superceded with bundle specific cooling times via explicit bundle specific decay heat calculations.
CALVERT CLIFFS ISFSI USAR 9.4-5 Rev. 22
TABLE 9.4-1 POST-DISCHARGE COOLING TIME
- NUHOMS-32P COOLING TIMES** (Years)
Burnup (GWd/
MTU) 38 < 39 < 40 < 41 < 42 < 43 < 44 < 45 < 46 < 47 < 48 < 49 < 50 < 51 <
B
B B B B B B B B B B B B B B 38
39 40 41 42 43 44 45 46 47 48 49 50 51 52 Enrichment 2.00E<2.10 8.4 8.9 9.4+ 9.9+ 10.5+ 11.2+ 11.9+ 12.6+ 13.4+ 14.3+ 15.2+ 16.2+ 17.3+ 18.5+ 19.7+
2.10E<2.20 8.3 8.7 9.2 9.8+ 10.4+ 11.0+ 11.7+ 12.5+ 13.3+ 14.1+ 15.1+ 16.0+ 17.1+ 18.2+ 19.4+
2.20E<2.30 8.1 8.6 9.1 9.6 10.2+ 10.8+ 11.5+ 12.3+ 13.1+ 13.9+ 14.9+ 15.9+ 16.9+ 18.0+ 19.2+
2.30E<2.40 8.0 8.5 9.0 9.5 10.1+ 10.7+ 11.4+ 12.1+ 12.9+ 13.8+ 14.7+ 15.7+ 16.7+ 17.8+ 19.0+
2.40E<2.50 7.9 8.4 8.8 9.4 9.9 10.6+ 11.2+ 12.0+ 12.8+ 13.6+ 14.5+ 15.5+ 16.5+ 17.7+ 18.8+
2.50E<2.60 7.8 8.2 8.7 9.3 9.8 10.4 11.1+ 11.8+ 12.6+ 13.5+ 14.4+ 15.3+ 16.4+ 17.5+ 18.7+
2.60E<2.70 7.7 8.2 8.6 9.1 9.7 10.3 11.0 11.7+ 12.5+ 13.3+ 14.2+ 15.2+ 16.2+ 17.3+ 18.5+
2.70E<2.80 7.6 8.1 8.5 9.0 9.6 10.2 10.9 11.6 12.4+ 13.2+ 14.1+ 15.0+ 16.1+ 17.2+ 18.3+
2.80E<2.90 7.6 8.0 8.4 9.0 9.5 10.1 10.8 11.5 12.2 13.1+ 13.9+ 14.9+ 15.9+ 17.0+ 18.2+
2.90E<3.00 7.5 7.9 8.4 8.9 9.4 10.0 10.7 11.4 12.1 12.9 13.8+ 14.8+ 15.8+ 16.9+ 18.0+
3.00E<3.10 7.4 7.9 8.3 8.8 9.3 9.9 10.6 11.3 12.0 12.8 13.7 14.6+ 15.7+ 16.7+ 17.9+
3.10E<3.20 7.4 7.8 8.2 8.7 9.3 9.8 10.5 11.2 11.9 12.7 13.6 14.5 15.5+ 16.6+ 17.7+
3.20E<3.30 7.3 7.7 8.2 8.7 9.2 9.8 10.4 11.1 11.8 12.6 13.5 14.4 15.4 16.5+ 17.6+
3.30E<3.40 7.3 7.7 8.1 8.6 9.1 9.7 10.3 11.0 11.7 12.5 13.4 14.3 15.3 16.4 17.5 3.40E<3.50 7.3 7.6 8.1 8.5 9.1 9.6 10.2 10.9 11.6 12.4 13.3 14.2 15.2 16.2 17.4 3.50E<3.60 7.2 7.6 8.0 8.5 9.0 9.6 10.2 10.8 11.6 12.3 13.2 14.1 15.1 16.1 17.2 3.60E<3.70 7.2 7.6 8.0 8.4 8.9 9.5 10.1 10.8 11.5 12.3 13.1 14.0 15.0 16.0 17.1 3.70E<3.80 7.2 7.5 7.9 8.4 8.9 9.4 10.0 10.7 11.4 12.2 13.0 13.9 14.9 15.9 17.0 3.80E<3.90 7.1 7.5 7.9 8.3 8.8 9.4 10.0 10.6 11.3 12.1 12.9 13.8 14.8 15.8 16.9 3.90E<4.00 7.1 7.5 7.9 8.3 8.8 9.3 9.9 10.5 11.2 12.0 12.8 13.7 14.7 15.7 16.8 4.00E<4.10 7.1 7.4 7.8 8.2 8.7 9.3 9.8 10.5 11.2 11.9 12.7 13.6 14.6 15.6 16.7 4.10E<4.20 7.0 7.4 7.8 8.2 8.7 9.2 9.8 10.4 11.1 11.8 12.7 13.5 14.5 15.5 16.6 4.20E<4.30 7.0 7.4 7.7 8.2 8.6 9.1 9.7 10.3 11.0 11.8 12.6 13.4 14.4 15.4 16.5 4.30E<4.40 7.0 7.3 7.7 8.1 8.6 9.1 9.7 10.3 10.9 11.7 12.5 13.3 14.3 15.3 16.4 4.40E<4.50 7.0 7.3 7.6 8.1 8.5 9.0 9.6 10.2 10.9 11.6 12.4 13.2 14.2 15.2 16.2
+ indicates that additional cooling time beyond that shown must be determined through an assembly specific source term calculation to ensure compliance with Technical Specification 2.1.
- All assemblies loaded into DSC must meet the source spectra requirements of Technical Specification 2.1.
- These bounding cooling times may be superceded with bundle specific cooling times via explicit bundle specific decay heat calculations.
CALVERT CLIFFS ISFSI USAR 9.4-6 Rev. 22
TABLE 9.4-2 NUHOMS-24P BURNUP CURVE DATA 4th ORDER CURVE INITIAL ENRICHMENT ACTUAL RESULTS FIT DATA(a)
(W/O U235) (GWD/MTU) (GWD/MTU) 1.8 0.00 0.25 1.9 2.83
- 2. 5.17 2.1 7.28 2.2 9.20 2.3 10.73 10.95 2.4 12.56 2.5 14.05 2.6 15.45 2.7 16.77 2.8 17.65 18.03 2.9 19.25
- 3. 20.44 3.1 21.61 3.2 22.78 3.3 23.96 23.96 3.4 25.14
3.5 26.33 3.6 27.55 3.7 28.79 3.8 29.55 30.04 3.9 31.32
- 4. 32.61 4.1 33.91 4.2 35.21 4.3 36.37 36.50 4.4 37.77 4.5 39.01 4.6 40.20 4.7 41.33 4.8 42.11 42.38 NOTE:
Equation BU = A*X 4+B*X 3+3+C*X 2+D*X+E Where A= -0.7521212 B = 10.9435555 C = -58.678357 D = 149.626326 E = 134.88247 (a)
Fuel burnup in excess of the curve fit data as shown in the table or as calculated by the fourth-order polynomial lie within the acceptance region illustrated in Figure 3.3-1.
This table is not used for assemblies loaded into a NUHOMS-32P DSC.
CALVERT CLIFFS ISFSI USAR 9.4-7 Rev. 20
TABLE 9.4-3 ADDITIONAL COOLING TIME REQUIREMENTS FOR LOADING ASSEMBLIES WITH IRRADIATED STAINLESS STEEL INERT REPLACEMENT RODS IN A 32P DSC Max SS Pin # SS Pins Allowable by Years after 660W Exposure MWD/MTU 0 Year 1 Year 2 Years 3 Years 20000 0 12 21 30 30000 0 7 13 18 40000 0 5 9 13 CALVERT CLIFFS ISFSI USAR 9.4-8 Rev. 22
9.5 EMERGENCY PLANNING The Emergency Response Plan (ERP) for CCNPP has been determined to be adequate for events which might occur involving the ISFSI. The ERP has been prepared in accordance with the requirements of 10 CFR 50.47, and therefore, satisfies the requirements of 10 CFR 72.32.
The CCNPP ERP has been developed to protect the general public and site personnel from possible consequences of an emergency condition. This plan, combined with its implementation procedures and the Radiological Emergency Plans of the state and local agencies, allows for (a) early recognition and classification of a possible emergency condition; (b) prompt notification, via reliable communication channels, of agencies and personnel to augment the normal operating personnel; (c) planned actions to be taken to protect the population-at-risk.
The CCNPP staff is trained to cope with emergencies. Written agreements with Federal agencies, private contractors, and coordinated state and local agency emergency plans (required by law) provide assistance to ensure resources can be readily available in as short a time as possible to cope with emergencies and protect the population-at-risk. The agencies, and the resources they will provide, are described in the ERP and the "Maryland Disaster Assistance Plan, Annex Q, Radiological Emergency Plan." Both plans describe the roles of the various state and local agencies and their interfaces for carrying out protective and parallel actions in a 10-mile-radius plume zone and a 50-mile-radius ingestion zone.
The ERP describes (1) the emergency classification system used at the plant; (2) the organizational control of emergencies, including on-site, off-site, and augmentation organizations; (3) the emergency measures to be taken; and (4) available emergency facilities and equipment.
Procedures for implementation of the CCNPP ERP are contained in the Emergency Response Plan Implementation Procedures.. These procedures are distributed to those individuals, and/or facilities where immediate availability of such procedures would be required during an emergency. The Emergency Response Plan Implementation Procedures provide the following information:
A. Means of classifying emergencies; B. Lists of available equipment; C. Directions for meeting notification requirements; D. Directions for seeking emergency assistance; E. Detailed instructions to individuals responsible for (a) assessing emergency conditions and (b) providing steps to be taken to mitigate the consequences of the accident.
The Emergency Response Plan Implementation Procedures are used in conjunction with applicable plant operating, radiological control, and security procedures to correct the emergency condition and to mitigate the consequences of the accident. Further details of the CCNPP ERP are contained in UFSAR Section 12.6.
CALVERT CLIFFS ISFSI USAR 9.5-1 Rev. 8
9.6 ISFSI LICENSE RENEWAL ACTIVITIES As part of ISFSI license renewal application an aging management assessment of the HSMs, DSCs [Nutech Horizontal Modular Storage] (NUHOMS-24P and NUHOMS-32P), transfer cask, transfer cask lifting yoke, the cask support platform, and high burnup fuel was performed. The assessment identified existing activities necessary to provide reasonable assurance that ISFSI and transfer cask components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis for the renewal period. This section describes these aging management activities that were performed.
9.6.1 AGING MANAGEMENT REVIEW The purpose of the aging management review is to address the aging effects that could adversely affect the ability of the components and subcomponents to perform their intended function during the period of extended operation. The aging management review involves the following four major steps:
- Identification of in-scope subcomponents requiring aging management reviews
- Identification of materials and environments
- Identification of aging effects requiring management
- Determination of the activities/programs required to manage the effects of aging The results of this aging management review of the in-scope components are contained in Tables 9.6-1 thru 9.6-6. These tables are the tables developed and submitted in Attachment (3) of Reference 9.21 as part of the license renewal application.
9.6.2 TIME-LIMITED AGING ANALYSIS Time-Limited Aging Analyses (TLAA) were conducted to identify and evaluate the effect of time-limited aging in order to demonstrate safe operation of the applicable components over the entire period of extended operation. This section discusses the results for each of the TLAAs evaluated for license renewal. The evaluations have demonstrated that the analyses have been projected to the end of the renewed license period.
9.6.2.1 DSC Time-Limited Aging Analysis AREVA Technical Report 10955-0101 (Reference 9.16) was prepared to identify and evaluate the effect of TLAA, to demonstrate safe operation over the extended service life of the ISFSI. The report evaluated the DSCs stored in the HSMs. This section describes the finding of the TLAA.
DSC Materials:
Stainless Steel The NUHOMS-24P and NUHOMS-32P DSC construction uses Stainless Steel 304 material and compatible weld metal. Since the DSC is filled with the inert helium gas there is no significant corrosion of the DSC shell and other components. Neutron fluence can affect mechanical properties of steels. However, studies on fast neutron damage in stainless steel and low alloy steels rarely evaluate damage at fluence levels below 1017 neutrons/cm2 because they are not significant (Reference 9.17). For the CALVERT CLIFFS ISFSI USAR 9.6-1 Rev. 23
DSC, the neutron fluence (1014 neutrons/cm2) is much less than this level for the intended storage period and hence, a TLAA is not required.
The DSC and weld stresses due to temperature and pressure inside the DSC are an important aspect of the design. Per Reference 9.16, the NUHOMS-24P DSC normal operating design temperature and pressure are 400°F and 10 psig and the accident temperature and pressure are 460°F and 50 psig.
Per Reference 9.16, the NUHOMS-32P DSC normal operating design temperature is 460°F for the DSC shell and 380°F for the welds, top and bottom shield plug/cover plate assembly and the design pressure is 30 psig. The design temperatures for the accident conditions for the NUHOMS-32P DSC shell is 575°F and 475°F for the welds, top and bottom shield plug/cover plate assembly and the design pressure is 100 psig. The accident pressure values are established using 100% cladding breach.
The design temperatures are calculated at the beginning of NUHOMS storage. This calculation is bounding since the DSC temperatures are shown to monotonically decrease as a function of time.
Therefore, the heating effect (and hence, the internal pressure effect) on the DSC for the future 40 years of service will be much less severe than that during the past 20 years of service. Hence, the stresses in the DSC components will be acceptable for the extension period.
The service life for the NUHOMS-24P DSC system is documented in Table 1.2-2, Reference 9.18 as 50 years. Sufficient clearances are provided in both the radial and axial direction between the DSC internal components to permit free thermal expansion for NUHOMS-24P and NUHOMS-32P DSCs. This design feature acts to minimize the thermal cycling and fatigue on the DSC. There will be more room for free thermal expansion as the decay heat from the fuel decreases causing the DSC internal component temperatures to decrease as the storage time is increased from 50 years to 60 years. Therefore, thermal cycling and fatigue on the DSC will not be impacted when the storage period is increased from 50 years to 60 years.
Lead in the Shield Plugs The DSC uses lead in the shield plugs. The lead shielding function is not affected by the radiation level in the DSC.
Poison Plates (NUHOMS-32P DSC Only)
Since the NUHOMS-32P DSC uses fixed neutron absorbers in the DSC basket, calculations were performed to assure criticality safety. Time dependency for criticality calculations may result due to depletion of boron in the poison plates utilized in the NUHOMS-32P DSC baskets.
Reference 9.19, Section 6.3, states that The continued efficacy of the neutron absorber materials over a 20-year storage period is assured by the design of the NUHOMS-32P canister which ensures that the neutron CALVERT CLIFFS ISFSI USAR 9.6-2 Rev. 23
absorbers will remain in place during accident conditions. Additionally, the neutron flux from the irradiated fuel will result in negligible depletion of the Boron 10 content in the neutron absorber materials over the life of the storage system.
Per Reference 9.16, the total neutron activity in the NUHOMS-32P DSC is 4.175x108 n/s/assembly. To estimate the total flux a conservative final assembly surface area of 25,000 cm2 is considered in Reference 9.16. The total scalar flux is estimated to be 8.65x104 n/cm2-s. Using the thermal cross section for Boron 10, (3,837 barn), the fraction of the original Boron 10 depleted after 60 years is 2.3E-6, which is negligible. The actual neutron flux is mostly fast and epithermal, and will be declining with time, so the actual depletion during dry storage will be less than the depletion calculated in Reference 9.16. Therefore, continued efficacy of the neutron poison is assured for an additional 40 years of fuel storage.
DSC Fatigue Evaluation The fatigue evaluation of the NUHOMS-24P DSC is documented in Appendix C.4.1 of Reference 9.16 for a 50 year operational life of the ISFSI. The fatigue effects on the DSC were addressed using the criteria contained in Section III NB-3222.4 of the American Society of Mechanical Engineers (ASME) Code (Reference 9.20, The 1983 Edition). The analysis evaluated the DSC under six criteria and concluded that the DSC and other components satisfy these criteria and that no consideration of fatigue is required for a service life of 50 years.
In order to extend the operational life by another 10 years, it is necessary to re-evaluate the DSC against these six criteria using an approach that is consistent with that utilized in Reference 9.18. A fatigue analysis for the NUHOMS-24P and NUHOMS-32P DSCs was performed for a 60 year service life. This analysis uses the six criteria in NB-3222.4(d) of ASME Code (Reference 9.20, the 1998 Edition including the 1999 Addenda) and determines that the DSC service loads of the NUHOMS-24P DSC and NUHOMS-32P DSC systems do not create any potential risk of the DSC design fatigue failure and that no detailed fatigue evaluation is necessary.
9.6.2.2 HSM Time-Limited Aging Analysis Reference 9.16 was prepared to identify and evaluate the effect of TLAA, to demonstrate safe operation over the extended service life of the ISFSI.
The report evaluated the HSMs. This section describes the finding of the TLAA.
HSM Concrete The HSM is a reinforced concrete structure. The effect of radiation on the HSM concrete is evaluated in Section 8.1.1.5.D and Section 12.8.1.1.5.D.
These evaluations demonstrate that the magnitude of the neutron fluence incident on the concrete is low enough to not affect the properties of the concrete. These evaluations also demonstrate that the magnitude of the gamma-ray energy deposition on the HSM concrete is not sufficient to cause any radiation heating in the concrete of the HSM. Therefore, the CALVERT CLIFFS ISFSI USAR 9.6-3 Rev. 23
thermal analyses documented in Sections 8.1.1.5D and 12.8.1.1.5D implicitly considered the radiation heat effects adequately for the HSM concrete.
NUHOMS-24P DSC The maximum predicted temperature of concrete at the beginning of storage was estimated to be below 150°F in Reference 9.16 using a bounding decay heat at the beginning of storage life.
NUHOMS-32P DSC The maximum temperature of the concrete at beginning of storage with a NUHOMS-32P DSC is 157°F per Reference 9.16.
The maximum concrete temperatures for the additional 40 years of service (as shown in Figure 8.1-27 of Reference 9.18) will be lower because the decay heat reduces monotonically as a function of time. Hence, the heating effect on the HSM concrete, for an additional 40 years of service, will be much less severe than the past 20 years of service.
The environmental degradation of reinforced concrete will not be significant, as proper concrete cover has been provided to the reinforcing bars made of carbon steel.
Reference 9.18, Section 8.2.10.6 documents the analysis of thermal cycling of the HSM based on the 50 year storage life. The number of cycles will increase from 18,250 to 21,900 when the design life is extended from 50 years of storage to 60 years of storage. These are still significantly below the limit of 10,000,000 (See Section 8.2.10.6 of Reference 9.18).
Therefore, thermal cycling will have negligible impact on the HSM reinforced concrete for an additional 40 years of service.
DSC Support Rail Steel in the HSMs The DSC support structure inside the HSM is designed to support the DSC during normal, off-normal and accident conditions. Since the DSC support rails are fabricated from Nitronic 60 austenitic stainless steel, it is expected that there would be no corrosion of the rail material and is expected to maintain its function for the additional 40 years of service.
DSC Support Rail Lubricant in the HSMs The HSM and transfer cask support rails are coated with a dry film lubricant Perma-Slik to minimize friction during insertion and retrieval of the DSC.
The material specification of the lubricant indicates that it is suitable for very high and cryogenic temperature applications. The presence of a non-corrosive environment due to the absence of a formal sea breeze and relatively milder temperature fluctuations at ISFSI site ensure that the lubricant does not degrade with age. The effect of radiation on these lubricants is not specified, however, it is expected that it is minimal since these are inorganic and consist entirely of graphite, a moderating material.
As stated above once the DSC is in place within the HSM, the lubricant performs no function during storage of the DSC.
CALVERT CLIFFS ISFSI USAR 9.6-4 Rev. 23
NUHOMS-24P DSC The coefficient of friction associated with these lubricants is below 0.05 while the design basis calculations employed a coefficient of friction of 0.25 (Section 8.1.1.1 D of Reference 9.18). The mechanical system to be used for DSC transfer is capable of exerting a force equal to the loaded weight of a DSC and this condition has been evaluated in Section 8.1.2.1 of Reference 9.18 for the NUHOMS-24P DSC. A coefficient of friction of 1.0 has been used (for these jammed DSC analyses) without relying on the solid film lubricant. The support structure is designed for this loading.
Hence, no further analysis is required.
NUHOMS-32P DSC The coefficient of friction associated with these lubricants is below 0.05 while the design basis calculations employed a coefficient of friction of 0.25 (Reference 9.16). The mechanical system to be used for DSC transfer is capable of exerting a force equal to the loaded weight of a DSC and this condition has been evaluated for the NUHOMS-32P DSC. A coefficient of friction of 1.0 has been used (for these jammed DSC analyses) without relying on the solid film lubricant. The support structure is designed for this loading. Hence, no further analysis is required.
9.6.2.3 Transfer Cask Fatigue Evaluation The fatigue evaluation for the transfer cask is performed in accordance with ASME code criteria listed in Section NC-3219.2 to determine whether the transfer cask service loads of NUHOMS-24P DSC and NUHOMS-32P DSC systems create potential risk of the design fatigue failure. The criteria evaluation shows that transfer cask service loads of NUHOMS-24P and NUHOMS-32P DSC systems do not create potential risk of the transfer cask design fatigue failure, for 600 cycles in 60-year transfer cask life, and that detailed fatigue evaluation is not necessary.
Transfer Cask Trunnions Fatigue Evaluation The fatigue evaluation of transfer cask trunnions shows that the transfer cask operations do not pose potential risk of fatigue failure of trunnion or trunnion sleeve throughout the planned 60-year service time.
NS-3 in Transfer Cask The transfer cask contains 3 inches of NS-3 neutron shielding sandwiched between the cask outer shell material and neutron shield jacket. Per Reference 9.16, the gamma and neutron dose at 1 inch from the cask surface for the accident conditions is 135 mrem/hr and 1000 mrem/hr, respectively for the NUHOMS-24P DSC and 85 mrem/hr and 1433 mrem/hr, respectively for the NUHOMS-32P DSC. Also, the dose rates at 1 inch from the transfer cask surface for the accident conditions with the NS-3 at the side of the transfer cask replaced with air bounds the dose rates at the inner surface of NS-3 in the transfer cask during normal conditions and that the transfer cask is only subjected to this gamma exposure when a fuel-loaded DSC is in the transfer cask during loading CALVERT CLIFFS ISFSI USAR 9.6-5 Rev. 23
and transfer operations which are short term durations. This results in a gamma dose of approximately 3.0x105 Rads over the service life of 60 years. This is based on an assumption that 1 Rad = 1 Rem and is considered reasonable for gamma radiation for hydrogenous materials.
This is significantly below the exposure limit of 1.5x1010 Rads for the material as stated in Reference 9.16.
Per Reference 9.16, to estimate the neutron fluence, a neutron dose to flux factor of 1 mrem/hr = 100 n/cm2-s is used. The dose to flux factor for neutrons is based on dose rate spectra results from various NUHOMS ISFSI evaluations. The integrated fluence is estimated to be approximately 3.16 x 1014 neutrons/cm2 over the service life of 60 years for the NS-3 in the transfer cask. Reference 9.16 noted that the thermal neutron exposure limit 1.5x1019 neutrons/cm2 for the NS-3 material. Therefore, it is concluded that there is no significant degradation to the NS-3 material for the additional 40 years of operations of the transfer cask.
The exposure to radiation sources for an additional 40 years of service is shown to have no significant impact on the shielding capability of the NS-3 in the transfer cask. No significant hydrogen loss in the NS-3 material is expected due to radiation exposure.
9.6.2.4 Time-Limited Aging Analysis of the Transfer Cask Lifting Yoke Lifting Yoke Fatigue Evaluation The fatigue evaluation of transfer cask trunnions and lifting yoke system shows that the transfer cask operations do not pose potential risk of fatigue failure of trunnion or trunnion sleeve throughout the planned 60-year service time.
In the case of the transfer cask lifting yoke assembly, the structural adequacy against fatigue failure is secured for up to 286 transfer cask loading/unloading operations in the planned 60-year service time.
9.6.3 AGING MANAGEMENT PROGRAMS Aging management programs are developed to identify the activities to be implemented to address the possible aging effects such that no aging effect results in the loss of intended function of the in-scope components. An aging management program is considered effective if it meets one of the following conditions:
- Provides for timely discovery of the effects of aging to be managed
- Mitigates the effects of aging to be managed Calvert Cliffs will implement and maintain the aging management programs submitted in Attachment 2 of Reference 9.21 for the duration of the ISFSI renewed license operating period.
9.6.3.1 HSM Aging Management Program The HSM aging management program credits the Calvert Cliffs Part 50 programs credited for managing the effects of aging in the Calvert Cliffs Nuclear Power Plant as described in Chapter 16 of Reference 9.1. The CALVERT CLIFFS ISFSI USAR 9.6-6 Rev. 23
HSM aging management program involves monitoring the exterior surfaces of the HSMs, including visual inspection of the accessible concrete; any exposed steel subcomponents, embedments, and attachments; and the lightning protection system. Interior inspections are conducted upon loading of a cask. Exterior inspections are conducted annually.
The HSM monitored conditions include, but not limited, to the following:
Concrete - spalling, cracking, delaminations, honey combs, leaching, discoloration, loss of material, or any other property that would be noted by visual inspection Structural Steel - corrosion, peeling paint, deflection, lost or missing anchors/fasteners, missing or degraded grout under base plates, twisted beams, cracked welds Equipment Foundations - settlement, cracked concrete Equipment Supports - cracked concrete, loose connections, corroded steel Roof Systems - structural integrity, deteriorated penetrations (i.e.,
drains, vents, etc.), signs of water infiltration, cracks, ponding, and flashing degradation Seismic Gaps - gaps or loss of joint filler material Lightning Protection System (above grade) - corrosion 9.6.3.2 Transfer Cask Aging Management Program The transfer cask aging management program credits periodic inspections performed on the transfer cask. The procedure includes visual and penetrant test inspections of the carbon steel subcomponents. Monitored aging effects by the aging management program include loss of material due to various forms of corrosion.
9.6.3.3 Transfer Cask Lifting Yoke Aging Management Program The transfer cask lifting yoke aging management program credits the transfer cask lifting yoke annual inspection. This procedure includes visual and magnetic particle test inspections of the transfer cask lifting yoke carbon steel subcomponents. Monitored aging effects by the aging management program include loss of material due to various forms of corrosion and cracking of material due to stress/strain from lifting.
9.6.3.4 Cask Support Platform Aging Management Program The cask support platform aging management program credits the Calvert Cliffs power plant Chemistry Control Program, as described in Chapter 16 of Reference 9.1. Loss of material and cracking are prevented through control of specified limits on chloride in the spent fuel pool water.
9.6.3.5 DSC External Surfaces Aging Management Program The DSC external surfaces AMP consists of condition monitoring activities to confirm there is no degradation of the DSC shell or cover plates that would result in a loss of the pressure/confinement boundary function.
CALVERT CLIFFS ISFSI USAR 9.6-7 Rev. 23
9.6.3.6 High Burnup Fuel Aging Management Program The high burnup fuel AMP will rely upon joint Electric Power Research Institute and Department of Energy research program as a surrogate program to monitor the condition of high burnup spent fuel assemblies.
The results from the research programs will be relied upon to provide reasonable assurance that the high burnup fuel assemblies at Calvert Cliffs would not experience degradation that would result in a loss of intended functions.
CALVERT CLIFFS ISFSI USAR 9.6-8 Rev. 23
TABLE 9.6-1 AGING MANAGEMENT REVIEW RESULTS FOR THE IRRADIATED FUEL ASSEMBLIES Aging Effects Aging Intended Subcomponent Material Group Environment Requiring Management Function Management Activities Fuel Rod (Cladding Air and Gas1,2 CC, HT, PB, SS Zircaloy-4, Zirlo, M54 None Identified None Required4 and End Caps) Residual Boron Coating3 Zircaloy-4 Guide Tubes SS Stainless Steel Air and Gas1 None Identified None Required4 (Chrome Plated)
Spacer Grid CC, SS Zircaloy-4 Air and Gas1 None Identified None Required4 Assemblies Lower End Fitting SS Stainless Steel Air and Gas1 None Identified None Required4 (and Connectors) (CC, SS) Inconel Upper End Fitting Stainless Steel (Connectors and SS Air and Gas1 None Identified None Required4 Inconel X-750 Holddown Spring)
Holddown Spring Retainer and None N/A N/A N/A N/A Upper End Plugs Fuel Assembly Control None N/A N/A N/A N/A Components Fuel Rod Pellets and Other Internal None N/A N/A N/A N/A Portions Notes:
1 Air and gas environment outside the fuel rods (inside the DSC) is helium at atmospheric pressure with trace amounts of air and water vapor. Minimal amounts of fission product gases may also be present. Temperature and radiation have been considered as described in Section 3.2.3, Environments for the Irradiated Fuel Assemblies.
2 Air and gas environment inside the fuel rods is pressurized helium and fission product gases. Temperature and radiation have been considered as described in Section 3.2.3, Environments for the Irradiated Fuel Assemblies.
CALVERT CLIFFS ISFSI USAR 9.6-9 Rev. 23
TABLE 9.6-1 AGING MANAGEMENT REVIEW RESULTS FOR THE IRRADIATED FUEL ASSEMBLIES 3
Residual boron may coat the irradiated fuel assemblies surfaces since they were exposed to a borated water environment in the SFP prior to storage. Any boric acid residue remaining on the irradiated fuel assemblies will have no deleterious effects due to the minimal amount of water remaining on the irradiated fuel assemblies and the materials of construction for the irradiated fuel assemblies.
4 A confirmatory program for high burnup fuel is being performed by the Department of Energy which will monitor the condition of high burnup fuel assemblies in dry storage. This confirmatory program includes Zircaloy-4, Zirlo, and M5 cladding materials that have been used at Calvert Cliffs but may not yet have been placed into dry storage at the time of ISFSI license Renewal.
CC Provides criticality control of spent fuel HT Provides heat transfer PB Directly or indirectly maintains a pressure boundary (confinement)
SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
N/A Not applicable CALVERT CLIFFS ISFSI USAR 9.6-10 Rev. 23
TABLE 9.6-2 AGING MANAGEMENT REVIEW RESULTS FOR THE DSCs (NUHOMS-24P AND 32P)
Intended Aging Management Subcomponent1 Materials Environment Aging Effect/Mechanism Function Activities Basket SS, CC Stainless Steel Air and Gas2,4 None Identified None Required Guide Sleeves (Basket) SS, CC Stainless Steel Air and Gas2,4 None Identified None Required Stainless Steel /
Spacer Disks (Basket) SS,CC Aluminum Coated Air and Gas2,4 None Identified None Required Carbon Steel Stainless Steel /
Support Rods (Basket) SS, CC Aluminum Coated Air and Gas2,4 None Identified None Required Carbon Steel Rails (Basket) SS Stainless Steel Air and Gas2,4 None Identified None Required Rail Inserts (Basket) SS Aluminum Air and Gas2,4 None Identified None Required Fixed Neutron Absorbers Borated CC Sheltered3 None Identified None Required (Basket) Aluminum Alloy Loss of Material due to DSC External Surfaces DSC Shell w/ Bottom PB, SH, Stainless Steel Air and Gas2 Crevice and Pitting Aging Management shield Plug SS, HT and Lead Sheltered3 Corrosion Cracking from Program Stress Corrosion Cracking PB, SH, Stainless Steel Air and Gas2,4 Top Shield Plug None Identified None Required SS, HT and Lead Sheltered3 Loss of Material due to DSC External Surfaces Cover Plates (Top and PB, SH, 3 Crevice and Pitting Stainless Steel Sheltered Aging Management Bottom) SS, HT Corrosion Cracking from Program Stress Corrosion Cracking Siphon and Vent Ports PB, SH Stainless Steel Air and Gas2,4 None Identified None Required Loss of Material due to DSC External Surfaces Crevice and Pitting Ram Grapple Ring SS Stainless Steel Sheltered3 Aging Management Corrosion Cracking from Program Stress Corrosion Cracking Dry Film Lubricant none N/A N/A N/A none Swagelok Quick none N/A N/A N/A none Disconnects Siphon Tube none N/A N/A N/A none CALVERT CLIFFS ISFSI USAR 9.6-11 Rev. 23
TABLE 9.6-2 AGING MANAGEMENT REVIEW RESULTS FOR THE DSCs (NUHOMS-24P AND 32P)
Intended Aging Management Subcomponent1 Materials Environment Aging Effect/Mechanism Function Activities Aluminum coating (Carbon Steel Spacer none N/A N/A N/A none Discs and Top Shield Plug)
Nickel Based Thread Lubricant; thread tape none N/A N/A N/A none or sealant Stainless steel plugs/bolts (non- none N/A N/A N/A none Structural)
DSC Lifting Lugs none N/A N/A N/A None Notes:
1 Each individual DSC may not contain all of the listed subcomponents.
2 Air and gas environment is helium inside DSC cavity, with possible trace amounts of air, water vapor and fission product gases.
Temperature and radiation have been considered as described in Section 3.3.3, Environments for the DSCs.
3 Sheltered environment for DSC interior/exterior surfaces that are not part of helium filled DSC cavity.
4 One time short exposure to borated water during loading operations is not considered an environment that impacts long term aging management.
CC Provides criticality control of spent fuel HT Provides heat transfer PB Directly or indirectly maintains a pressure boundary (confinement)
SH Provides radiation shielding SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
N/A Not applicable CALVERT CLIFFS ISFSI USAR 9.6-12 Rev. 23
TABLE 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM Intended Aging Aging Management Materials Environment Subcomponent Function Effect/Mechanism Activities Reinforced Concrete Walls, Roof and Foundation; Shielded Cracking due to HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air freeze-thaw Management Ventilation Air Outlet Shielding degradation Program Blocks Reinforced Concrete Walls, Loss of material Roof and Foundation; Shielded HSM Aging (spalling, scaling)
Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air Management due to freeze-thaw Ventilation Air Outlet Shielding Program degradation Blocks Reinforced Concrete Walls, Roof and Foundation; Shielded Cracking due to HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air moisture, chemical Management Ventilation Air Outlet Shielding attack, or leaching Program Blocks Reinforced Concrete Walls, Loss of material Roof and Foundation; Shielded (spalling, scaling) HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air due to moisture, Management Ventilation Air Outlet Shielding chemical attack, or Program Blocks leaching Reinforced Concrete Walls, Roof and Foundation; Shielded Corrosion due to HSM Aging Embedded Ventilation Air Inlet Plenum; SS Yard, Air moisture, chemical Management Steel Ventilation Air Outlet Shielding attack, or leaching Program Blocks Reinforced Concrete Walls, Increase in Roof and Foundation; Shielded HSM Aging porosity/permeability Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air Management due to leaching of Ventilation Air Outlet Shielding Program Ca(OH)2 Blocks CALVERT CLIFFS ISFSI USAR 9.6-13 Rev. 23
TABLE 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM Intended Aging Aging Management Materials Environment Subcomponent Function Effect/Mechanism Activities Reinforced Concrete Walls, Roof and Foundation; Shielded Loss of strength due HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air to leaching of Management Ventilation Air Outlet Shielding Ca(OH)2 Program Blocks Reinforced Concrete Walls, Increase in Roof and Foundation; Shielded HSM Aging porosity/permeability Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air Management due to cement Ventilation Air Outlet Shielding Program aggregate reactions Blocks Reinforced Concrete Walls, Roof and Foundation; Shielded Loss of strength due HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air to cement aggregate Management Ventilation Air Outlet Shielding reactions Program Blocks Reinforced Concrete Walls, Cracking due to Roof and Foundation; Shielded HSM Aging settlement or loss of Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air Management bond with embedded Ventilation Air Outlet Shielding Program steel Blocks Reinforced Concrete Walls, Roof and Foundation; Shielded HSM Aging Cracking due to Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air Management irradiation Ventilation Air Outlet Shielding Program Blocks Reinforced Concrete Walls, Roof and Foundation; Shielded Loss of material HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air (spalling, scaling) Management Ventilation Air Outlet Shielding due to irradiation Program Blocks CALVERT CLIFFS ISFSI USAR 9.6-14 Rev. 23
TABLE 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM Intended Aging Aging Management Materials Environment Subcomponent Function Effect/Mechanism Activities Reinforced Concrete Walls, Roof and Foundation; Shielded Cracking due to HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air cement aggregate Management Ventilation Air Outlet Shielding reaction Program Blocks Reinforced Concrete Walls, Roof and Foundation; Shielded Loss of material due HSM Aging Ventilation Air Inlet Plenum; HT, SH, SS Concrete Yard, Air to cement aggregate Management Ventilation Air Outlet Shielding reaction Program Blocks Cracking due to HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete freeze-thaw Management Roof and Foundation Underground degradation Program Loss of material HSM Aging Reinforced Concrete Walls, Embedded/ (spalling, scaling)
HT, SH, SS Concrete Management Roof and Foundation Underground due to freeze-thaw Program degradation Cracking due to HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete moisture, chemical Management Roof and Foundation Underground attack, or leaching Program Loss of material due HSM Aging Reinforced Concrete Walls, Embedded/ to moisture, HT, SH, SS Concrete Management Roof and Foundation Underground chemical attack, or Program leaching Corrosion due to HSM Aging Reinforced Concrete Walls, Embedded Embedded/
SS moisture, chemical Management Roof and Foundation Steel Underground attack, or leaching Program Increase in HSM Aging Reinforced Concrete Walls, Embedded/ porosity/permeability HT, SH, SS Concrete Management Roof and Foundation Underground due to leaching of Program Ca(OH)2 CALVERT CLIFFS ISFSI USAR 9.6-15 Rev. 23
TABLE 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM Intended Aging Aging Management Materials Environment Subcomponent Function Effect/Mechanism Activities Loss of strength due HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete to leaching of Management Roof and Foundation Underground Ca(OH)2 Program HSM Aging Reinforced Concrete Walls, Embedded/ Cracking due to HT, SH, SS Concrete Management Roof and Foundation Underground irradiation Program Loss of material HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete (spalling, scaling) Management Roof and Foundation Underground due to irradiation Program Cracking due to HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete cement aggregate Management Roof and Foundation Underground reaction Program Loss of material due HSM Aging Reinforced Concrete Walls, Embedded/
HT, SH, SS Concrete to cement aggregate Management Roof and Foundation Underground reaction Program HSM Aging Carbon Loss of material due DSC Structural Steel Assembly SS Sheltered Management Steel to corrosion Program Nitronic 60 DSC Structural Steel Assembly SS Stainless Sheltered None Identified None Required Steel HSM Aging Carbon Loss of material due DSC Seismic Retainer SS Sheltered Management Steel to corrosion Program HSM Aging Cask Docking Flange and Tie Carbon Loss of material due SS Sheltered Management Restraints Steel to corrosion Program HSM Aging Cask Docking Flange and Tie Carbon Loss of material due SS Yard Management Restraints Steel to corrosion Program Stainless Heat Shield HT Sheltered None Identified None Required Steel CALVERT CLIFFS ISFSI USAR 9.6-16 Rev. 23
TABLE 9.6-3 AGING MANAGEMENT REVIEW RESULTS FOR THE HSM Intended Aging Aging Management Materials Environment Subcomponent Function Effect/Mechanism Activities Shielded Front Access Door and SH, SS Concrete Embedded None Identified None Identified Door Supports HSM Aging Shielded Front Access Door and Carbon Loss of material due SH, SS Yard Management Door Supports Steel to corrosion Program Ventilation Air Openings (One Stainless HT Yard None Identified None Required Inlet/ Two Outlets) Steel Shielded Ventilation Air Inlet Stainless HT Yard None Identified None Required Plenum Steel Ventilation Air Outlet Shielding Stainless HT Yard None Identified None Required Blocks Steel Lighting Protection System SS Copper Yard None Identified None Required Threaded Fasteners and Stainless Yard Embedded/
HT, SS None Identified None Required Expansion Anchors Steel Yard Carbon Handrail SS Yard None Identified None Required Steel HT Provides heat transfer SH Provides radiation shielding SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
CALVERT CLIFFS ISFSI USAR 9.6-17 Rev. 23
TABLE 9.6-4 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK Intended Aging Aging Management Subcomponent Function Materials Environment1,2 Effect/Mechanism Activity Loss of Material due to General Corrosion Structural Shell SS, HT, Embedded/Borated Transfer Cask Aging Carbon Steel Loss of Material due to (Cask Body) SH Water Management Program Pitting or Crevice Corrosion Bottom Support Sheltered/Borated SS Stainless Steel None Identified None Required Ring (Cask Body) Water Bottom Cover Plate Sheltered/Borated SS Stainless Steel None Identified None Required (Cask Body) Water Top Flange (Cask Sheltered/Borated SS Stainless Steel None Identified None Required Body) Water Inner Shell (Cask SS,HT, Sheltered/Borated Stainless Steel None Identified None Required Body) SH Water Lead (Cask Body) SS,HT Lead Embedded None Identified None Required Rails (Cask Stainless Steel SS Sheltered None Identified None Required Attachments) (Nitronic 60)
Upper Trunnions Cracking of material due Sheltered/Borated Transfer Cask Aging (Cask SS Stainless Steel to stress/strain from Water Management Program Attachments) lifting Upper Trunnion Cracking of material due Embedded/Borated Transfer Cask Aging Sleeves (Cask SS Stainless Steel to stress/strain from Water Management Program Attachments) lifting Upper Trunnion Cracking of material due Sheltered/Borated Transfer Cask Aging Nickel Alloy (Cask SS Inconel to stress/strain from Water Management Program Attachments) lifting Upper Trunnion Neutron Shielding HT, SH Bisco NS-3 Embedded None Identified None Required (Cask Attachments)
CALVERT CLIFFS ISFSI USAR 9.6-18 Rev. 23
TABLE 9.6-4 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK Intended Aging Aging Management Subcomponent Function Materials Environment1,2 Effect/Mechanism Activity Upper Trunnion Cracking of material due Sheltered/Borated Transfer Cask Aging Cover Plate (Cask SS, SH Stainless Steel to stress/strain from Water Management Program Attachments) lifting Lower Trunnions Cracking of material due Sheltered/Borated Transfer Cask Aging (Cask SS Stainless Steel to stress/strain from Water Management Program Attachments) lifting Lower Trunnion Cracking of material due Embedded/Borated Transfer Cask Aging Sleeves (Cask SS Stainless Steel to stress/strain from Water Management Program Attachments) lifting Lower Trunnion Sleeve Nickel Cracking of material due Embedded/Borated Transfer Cask Aging Alloy Weld SS Stainless Steel to stress/strain from Water Management Program Overlay (Cask lifting Attachments)
Lower Trunnion Neutron Shielding Embedded/Borated HT,SH Bisco NS-3 None Identified None Required (Cask Water Attachments)
Ram Access Penetration Ring Sheltered/Borated SS Stainless Steel None Identified None Required (Cask Water Penetration)
Upper and Lower Rings, Outer Shell, Relief Valve Sheltered/Borated SH,HT Stainless Steel None Identified None Required Support Plates Water (Cask Neutron Shield)
Inner and Outer Support Angle Sheltered/Borated SH, HT Stainless Steel None Identified None Required (Cask Neutron Water Shield)
CALVERT CLIFFS ISFSI USAR 9.6-19 Rev. 23
TABLE 9.6-4 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK Intended Aging Aging Management Subcomponent Function Materials Environment1,2 Effect/Mechanism Activity Shielding Material (Cask Neutron SH, HT Bisco NS-3 Embedded None Identified None Required Shield)
Inner, Outer, and Side Plates (Top SS Stainless Steel Sheltered None Identified None Required Cover Assembly)
Ring; Eye Bolt 24 Hot Stand-offs (Top SS Sheltered None Identified None Required Galvanized Finish Cover Assembly)
Neutron Shielding (Top Cover SH Bisco NS-3 Embedded None Identified None Required Assembly)
Inner, Outer, and Side Plates Sheltered/Borated SH Stainless Steel None Identified None Required (Bottom Cover Water Assembly)
Polymer Bottom Cover Materials Property N/A - subject to routine HT, SH (Ethylene Sheltered O-ring Seals Change replacement Propylene)
Neutron Shielding (Bottom Cover SH Bisco NS-3 Embedded None Identified None Required Assembly)
Cask Bottom Cover SH Stainless Steel Sheltered None Identified None Required Plate Neutron Shielding SH Bisco NS-3 Embedded None Identified None Required (Cask Bottom)
Bolts, Washers, Loss of Material due to and Threaded General Corrosion Transfer Cask Aging Fasteners for Top SH Carbon Steel Sheltered Loss of Material due to Management Program Cover Plate and Pitting or Crevice Ram Access Plate Corrosion CALVERT CLIFFS ISFSI USAR 9.6-20 Rev. 23
TABLE 9.6-4 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK Intended Aging Aging Management Subcomponent Function Materials Environment1,2 Effect/Mechanism Activity Misc none N/A N/A N/A N/A Subcomponents Notes:
1 Sheltered environment represents ambient conditions on the interior of the transfer cask, conservatively including connecting and embedded surfaces. Some subcomponents may have interior surfaces that are considered embedded. No aging effects are identified for the embedded surfaces and no aging management is required. Temperature and radiation were considered, as described in Section 3.5.3, Environments for the Transfer Cask.
2 All subcomponents that are immersed in the borated water of the SFP are rinsed off with deionized water after use.
HT Provides heat transfer SH Provides radiation shielding SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
N/A Not applicable CALVERT CLIFFS ISFSI USAR 9.6-21 Rev. 23
TABLE 9.6-5 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK LIFTING YOKE Intended Aging Aging Management Subcomponent Materials Environment1 Function Effect/Mechanism Activity Loss of Material due to General Corrosion Loss of Material due to Transfer Cask Lifting Yoke Lifting Hook Sheltered/Borated Pitting or Crevice SS Carbon Steel Aging Management Plates Water Corrosion Program Cracking of material due to stress/strain from lifting Loss of Material due to General Corrosion Loss of Material due to Transfer Cask Lifting Yoke Lifting Beam Sheltered/Borated Pitting or Crevice SS Carbon Steel Aging Management Plates Water Corrosion Program Cracking of material due to stress/strain from lifting Loss of Material due to General Corrosion Transfer Cask Lifting Yoke Later Brace Sheltered/Borated SS Carbon Steel Loss of Material due to Aging Management Plates Water Pitting or Crevice Program Corrosion Loss of Material due to General Corrosion Transfer Cask Lifting Yoke Support Brace Sheltered/Borated SS Carbon Steel Loss of Material due to Aging Management Plates Water Pitting or Crevice Program Corrosion Loss of Material due to General Corrosion Transfer Cask Lifting Yoke Stainless Sheltered/Borated Pin (Round Bar) SS Loss of Material due to Aging Management Steel Water Pitting or Crevice Program Corrosion Pin Handle None N/A N/A N/A N/A CALVERT CLIFFS ISFSI USAR 9.6-22 Rev. 23
TABLE 9.6-5 AGING MANAGEMENT REVIEW RESULTS FOR THE TRANSFER CASK LIFTING YOKE Intended Aging Aging Management Subcomponent Materials Environment1 Function Effect/Mechanism Activity Pin Cradle Pipe None N/A N/A N/A N/A Rear Pin Stop None N/A N/A N/A N/A Pin Lock None N/A N/A N/A N/A Loss of Material due to Main Assembly General Corrosion Transfer Cask Lifting Yoke Bolts, Nuts, SS Carbon Steel Sheltered Loss of Material due to Aging Management Washers Pitting or Crevice Program Corrosion Support angles and misc None N/A N/A N/A N/A hardware Hook Bearing SS Bronze Sheltered None Identified None Required Plate Notes:
1 All subcomponents that are immersed in the borated water of the SFP are rinsed off with deionized water after use.
SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
N/A Not applicable CALVERT CLIFFS ISFSI USAR 9.6-23 Rev. 23
TABLE 9.6-6 AGING MANAGEMENT REVIEW RESULTS FOR THE CASK SUPPORT PLATFORM Intended Aging Management Subcomponent Materials Environment Aging Effect/Mechanism Function Activity Stainless Loss of material due to pitting Cask Support Aging Base Plate SS Borated Water Steel or stress corrosion cracking Management Program Stainless Loss of material due to pitting Cask Support Aging Web Plates SS Borated Water Steel or stress corrosion cracking Management Program Stainless Loss of material due to pitting Cask Support Aging Mid Plate SS Borated Water Steel or stress corrosion cracking Management Program Stainless Loss of material due to pitting Cask Support Aging Top Plate SS Borated Water Steel or stress corrosion cracking Management Program Honeycomb Energy SS Aluminum Compressed Air None Identified None Required Absorber Honeycomb Base Stainless Loss of material due to pitting Cask Support Aging SS Borated Water Plate Steel or stress corrosion cracking Management Program Honeycomb Casing Stainless Loss of material due to pitting Cask Support Aging SS Borated Water Plate Steel or stress corrosion cracking Management Program Honeycomb Outer Stainless Loss of material due to pitting Cask Support Aging SS Borated Water Plate Steel or stress corrosion cracking Management Program Bottom Location Stainless Loss of material due to pitting Cask Support Aging SS Borated Water Plates Steel or stress corrosion cracking Management Program Stainless Loss of material due to pitting Cask Support Aging Lifting Lugs SS Borated Water Steel or stress corrosion cracking Management Program Tubing Manifold, Relief Valve, none N/A N/A N/A N/A Pressure Gauge, Quick-Connect SS Provides structural support and/or functional support of important to safety equipment (structural integrity)
N/A Not applicable CALVERT CLIFFS ISFSI USAR 9.6-24 Rev. 23
9.7 DECOMMISSIONING PLAN Decommissioning of the ISFSI will be performed in a manner consistent with decommissioning of the CCNPP. It is anticipated that the DSCs will be transported to a Federal repository when such a facility is operational. However, should the storage facility not accept the DSCs intact, the NUHOMS system allows the DSCs to be brought back into the spent fuel pool and the fuel repositioned into the racks for loading into transport casks provided by the Department of Energy.
All components of the NUHOMS system are manufactured of materials similar to those found at the existing CCNPP (e.g., reinforced concrete, stainless steel, lead). These components will be decommissioned by the same methods in place to handle those materials within the plant. Any of the components that may be contaminated will be cleaned and/or disposed of using the decommissioning technology available at the time of decommissioning.
Removal of fuel assemblies from the DSC can be done in the plant's spent fuel pool, as described in Chapter 5, or the DSC could also be qualified for off-site shipment in a suitable transportation cask licensed to 10 CFR Part 71. If such transport is made, the DSC could be disposed of as-is at the final spent fuel repository. If the DSC is not compatible with the final repository handling systems, fuel transfer to a suitable container can be performed in any suitable large hot cell or off-site fuel pool.
The detailed decommissioning plan for Calvert Cliffs ISFSI is provided in Reference 9.13.
CALVERT CLIFFS ISFSI USAR 9.7-1 Rev. 23
9.8 REFERENCES
9.1 Calvert Cliffs Nuclear Power Plant, Updated Final Safety Analysis Report, Docket Nos. 50-317 and 50-318, Baltimore Gas and Electric Company 9.2 SFSP Procedure SFSPP-1, Preparation and Control of Spent Fuel Storage Procedures 9.3 SFSP Procedure SFSPP-2, Control of Changes and Deviations Found During Construction 9.4 SFSP Procedure SFSPP-4, Procurement 9.5 SFSP Procedure SFSPP-13, Records Retention 9.6 SFSP Procedure SFSPP-14, Nuclear Related Indoctrination, Training and Qualification 9.7 QAP-36, Independent Spent Fuel Storage Installation 9.8 10 CFR Part 72 Quality Assurance Program for the Spent Fuel Storage Project 9.9 Letter from Mr. G. C. Creel (BGE) to Director, Office of Nuclear Material Safety and Safeguards (NRC), dated December 20, 1990, Response to NRC's Comments on the Safety Analysis Report (SAR) for BGE's License Application for Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI) 9.10 Letter from Mr. G. C. Creel (BGE) to Director, Office of Nuclear Material Safety and Safeguards (NRC), dated September 30, 1991, Response to NRC's Follow Up Comments on the Safety Analysis Report (SAR) for BGE's License Application for Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI) 9.11 Letter from Mr. G. C. Creel (BGE) to Director, Office of Nuclear Material Safety and Safeguards (NRC), dated December 27, 1991, Response to Requests for Additional Information (RAI), Dated December 12 and 19, 1991, on the Safety Analysis Report (SAR) for BGE's License Application for Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI) 9.12 Letter from Charles H. Cruse (BGE) to Region I, Regional Administrator (NRC), dated October 19, 1993, Calvert Cliffs Nuclear Power Plant Independent Spent Fuel Storage Installation; Docket No. 72-8150-317/318, Preoperational Test Acceptance Criteria and Test Results 9.13 Letter from Mr. G. C. Creel (BGE) to Director, Office of Nuclear Material Safety and Safeguards (NRC), dated August 18, 1992, Revision to the ISFSI Decommissioning Plan 9.14 CCNPP Calculation No. CA06354, "Accidental Drop Loading Evaluation of 14x14 Fuel Assembly with Missing Fuel Rods" 9.15 CCNPP ES200600043, Implementation of ISFSI License Amendment No. 9 Prior to 2010 Loadings 9.16 AREVA Technical Report 10955-0101, Revision 1, May 21, 2010, ISFSI Time-Limited Aging Analysis Report 9.17 U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, May 1988, Radiation Embrittlement of Reactor Vessel Materials 9.18 NUH-002, Revision 2, March 1990, Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUHOMS-24P, Submitted to the United States Nuclear Regulatory Commission by Pacific Nuclear Fuel Services, Inc., San Jose, California CALVERT CLIFFS ISFSI USAR 9.8-1 Rev. 23
9.19 Letter from R. J. Lewis (NRC) to G. Vanderheyden (CCNPP), dated June 10, 2005, Amendment 6 to Material License No. SNM-2505 for the Calvert Cliffs Independent Spent Fuel Storage Installation, Safety Evaluation Report 9.20 American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code,Section III, Division I, Subsection NB, July 1, 2009, Class 1 Components, Rules for Construction of Nuclear Facility Components 9.21 Letter from G. H. Gellrich (Exelon) to Document Control Desk (NRC), dated September 18, 2014, Response to Fourth Request for Additional Information for Renewal Application to Special Nuclear Materials License No. 2505 for the Calvert Cliffs Site Specific Independent Spent Fuel Storage Installation (TAC No. L24475)