ML20045D669
| ML20045D669 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/23/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045D668 | List: |
| References | |
| NUDOCS 9306290286 | |
| Download: ML20045D669 (7) | |
Text
'
a ano
..- yi 7,
UNITED STATES 5-NUCLEAR REGULATORY COMMISSION
\\'
j#
WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 62 AND 27 TO FACILITY OPERATING LICENSE NOS NPF-39 AND NPF-85 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION. UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353
1.0 INTRODUCTION
i By letter dated March 8, 1993, as supplemented by letter dated June 2, 1993, the Philadelphia Electric Company (the licensee) submitted a request for changes to paragraph 2.B.(5) to the Operating License Nos. NPF-39 and NPF-85 for the Limerick Generating Station (LGS), Units 1 and 2.
The requested changes would allow the receipt, possession and use of the fuel assemblies and fuel channels previously irradiated in the She-eham Nuclear Power Station (SNPS).
The fuel was fabricated by General Electric Company (GE) and consists of 560 GE6-(P8X8R) pressurized, C-lattice, non-barrier fuel assemblies. The 560 fuel assemblies include 340 enriched to 2.19 w/o U-235,144 enriched to 1.76 w/o U-235, and the remaining 76 are natural uranium (i.e., 0.711 w/o U-235).
These fuel assemblies are similar to those uti'.' zed in the LGS, Unit 1 initial core loading.
The supplemental letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
The fuel was used at SNPS in a limited testing program at 5% power.
It has been irradiated to a core average exposure of approximately 48 megawatt days per metric ton (MWD /MT).
The estimated core fission inventory is less than 0.02% of the source term, and its decay heat rate is approximately 265 watts (i.e., 900 Btu /hr) as of June 1992. The fuel transport between the two sites will utilize the GE IF-300 Series spent fuel cask. The GE IF-300 has received an NRC Certificate of Compliance (No. 9001), that has been amended to address the specific pay load to be utilized for the proposed transport of'the SNPS fuel to the LGS site.
The staff has confirmed that 1) 'a current amendment to the NRC Certificate of Compliance No. 9001 has been issued for the spent fuel cask; 2) a security plan has been established for the transport of the subject fuel; 3) an Environmental' Assessment and Finding of No Significant Impact has i
been issued; and 4) a complete technical evaluation of all aspects affecting j
the receipt, possession and use of the subject fuel at the LGS site has been-performed.
2.0 EVALUATION The staff has addressed all pertinent. issues associated with the proposed fuel transfer, as applicable to the loading and transport from the SNPS to the LGS, and the unloading, storage, and use of the fuel assemblies and the fuel channels at the LGS. The specific issues addressed by the staff in this i
9306290286 930623 ADOCKOSOOg2 DR.
.. evaluation include the determination o' the applicability of the Price-Anderson Rule; the evaluation of the criticality aspects of receiving, storing, and using the slightly irradiated fuel; the radiological assessment; and the handling of the heavy loads and cooling of the subject fuel and components.
2.1 Price-Anderson There are no unresolved financial protection issues involved in the use of Shoreham spent fuel at Limerick.
Price-Anderson coverage would cover the fuel from the SNPS to the LGS and would also extend to the fuel while it is being used at the LGS and to the natural uranium fuel assemblies that would be used to test for damage.
See Section 170 of the Atomic Energy Act of 1954, as amended.
2.2 Storace and Use of the Irradiated Fuel Storace of Irradiated Fuel The criticality analysis for the LGS spent fuel pool, as described in the Updated Final Safety Analysis Report (UFSAR) Section 9.1.2.3.1, assumed fuel assemblies with a uniform 3.5 w/o U-235 enrichment. The analysis also assumed the presence of zircaloy channels.
The resulting worst case k'.,95.was 0.933, nogreaterthanD The which meets the NRC limiting criterion of k highest average assembly enrichment of the $NPS fuel -is 2.19 w/o U-235 and the maximum planar enrichment is 2.33 w/o U-235.
Based on the lower enrichment, the reactivity of the storage array of the SNPS fuel in the LGS storage pool will result in a lower value of k,,, than was calculated _ for the LGS fuel.
The SNPS fuel will be packaged for transportation to the LGS with polyethylene spacers and a protective stainless steel channel.
GE, therefore, evaluated the effect of these spacers and channels on the spent fuel storage pool k,,,.
The stainless steel channels were found to lower the reactivity of.the spent fuel pool k in all cases. However, the increased neutron moderation due to the hydroge,n,,in the polyethylene spacers tends to cause a reactivity increase.
GE has determined that the lower enrichment of the SNPS fuel, compared to the enrichment used in the LGS criticality analysis, causes a much greater negative reactivity effect than the positive reactivity addition caused by the polyethylene spacers. Therefore, the storage of the SNPS fuel in th'e LGS spent fuel pool is acceptable since it results in a k,,, of less than 0.933, thus meeting the NRC limit of no greater than 0.95.
Use of the SNPS Fuel in the LGS Core A detailed inspection of two of the irradiated SNPS fuel assemblies was performed by GE in August 1990.
This inspection, which included eddy current testing of individual fuel and water rod's as well as a visual inspection of the entire feel assembly, verified that the SNPS fuel was suitable for future use.
In addition, an evaluation of the water chemistry history of both the SNPS reactor and spent fuel pool determined that the fuel has not been exposed to an adverse environment that would preclude its future use.
A
.. - PEco will ensure that the SNPS fuel assemblies arrive in a condition suitable for future use by inspecting a dummy test assembly after it has been subjected to accelerations and loadings at least as great as those expected during shipping and handling.
The acceptance criteria will be the same as applied to the shipment of new fuel, as specified in NEDE-23542-P, " Fuel Assembly Evaluation of Shipping and Handling Loads," dated March 1977.
In addition to disassembling and inspecting at least one fuel assembly from the first shipment, all assemblies shipped from the SNPS to the LGS will be visually inspected before and after packaging as well as upon arrival at LGS. Any assembly that does not meet the acceptance criteria used for the receipt inspection of new fuel will be excluded from future use in the LGS cor -
unless it is appropriately repaired.
The staff finds the acceptance c-c ria as well as the tests and inspections used to determine the suitability
'he SNPS fuel for future use at the LGS acceptable.
Before operation with the SNPS fuel, a cycle-specific core nuclear analysis will be performed based on the latest NRC-approved version of NEDE-240ll-P-A,
" General Electric Standard Application for Reactor Fuel GESTAR II."
The effect of the SNPS fuel on the thermo-hydraulic stability of the core will also be evaluated based on NRC Generic Letter 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors (BWR)." These are the same evaluations performed for all the LGS reload cores and are acceptable.
In addition, an evaluation was performed to determine if any analysis changes are required to account for the prior operating history, handling, and transportation of the SNPS fuel. The SNPS fuel was found to meet all the licensing bases documented in NEDE-240ll-P-A and, therefore, no exceptions to GESTAR II will be needed when the SNPS fuel is analyzed for use in the LGS cores.
PEco has stated that only a limited number of the SNPS fuel assemblies will be used each cycle.
These assemblies will only be placed in low duty core locations.
The staff finds this limited use in low power locations acceptable.
Conclusion of Storaae and Use of the SNPS Fuel The staff has reviewed the criticality aspects of storage of the irradiated SNPS Fuel in the LGS spent fuel pools and the suitability of this fuel for future use in the LGS cores.
The impact of the SNPS fuel and its packaging material on the LGS spent fuel pool criticality was found to be bounded by the fuel pool criticality analysis presented in Section 9.1.2.3.1 of the LGS UFSAR.
In addition, before the SNPS fuel is used in an LGS core, a cycle-specific analysis, which will include the effect on the thermal-hydraulic stability, will be performed in accordance with NRC-approved methods to determine its acceptability.
W e
.. 2.3 Radioloaical Assessment In the submittal from PEco, it was stated that SNPS fuel had been irradiated to a core average exposure of approximately 48-Megawatt-days-per-metric-ton and that the fuel had been removed from the reactor and placed in the SNPS spent fuel pool in August 1989.
The submittal indicated that the slightly irradiated fuel contains 0.02% of the source term assumed in the design basis loss of coolant accident described in the LGS UFSAR.
PEco also stated that the radiological consequences of a dropped fuel assembly involving the SNPS fuel are bounded by the fuel handling accident involving highly irradiated spent fuel described in the LGS UFSAR Section 15.7.4, " Fuel Handling Accident." They stated further that while handling the IF-300 cask, which weighs 85 tons including the basket,17 fuel assemblies, and a redundant cask-lifting yoke, the requirements of NUREG-0554, " Single-Failure-Proof Cranes for Nuclear Power Plants" and NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," would be met by the use of a single-failure-proof redundant yoke and by restricting the critical load of the reactor enclosure main hoist to 110 tons.
PECo also stated that restricting the reactor enclosure main hoist critical load to 110 tons and the use of single-failure-proof equipment precludes a cask drop due to single-failure.
Therefore, an analysis of the spent fuel cask drop is not required.
The staff has assessed the consequences of a fuel handling accident involving the SNPS fuel.
The staff is in agreement with PECo that existing analysis for fuel handling accident involving highly irradiated fuel at the LGS, which is described in the LGS UFSAR Section 15.7.4, bounds any potential fuel handling accident associated with the SNPS fuel.
In addition, such a postulated accident is also bounded by the staff's analysis of the consequences of a fuel handling accident, which was presented in NUREG 0991, " Safety Evaluation Report Related to the Operation for the Limerick Generating Station, Units 1 and 2."
The staff has also concluded that, as a result of the steps PEco is taking to meet the requirements of NUREG-0554 and NUREG-0612, an analysis of a spent fuel cask drop accident is not required for this licensing action.
2.4 Fuel Handlina and Coolina The licensee plans to move the fuel from the SNPS via barge to a PEco site along the Delaware River and then to the LGS by rail.
The shipping container will be the GE IF-300 series spent fuel cask with a basket design that can hold 17 fuel assemblies. The railcar will be moved into the reactor building under the refueling hoist-way. The reactor enclosure crane will lift the cask from the railcar through the open hoist-way via the yoke designed for lifting the IF-300 cask.
The cask will then be moved to the cask pool, located.
between the Unit I and Unit 2 spent fuel pools.
The cask top will then be removed and individaal fuel assemblies will be moved from the cask to the spent fuel pool for Unit 1 or Unit 2 thrbugh open slot B in either pool.
)
The licensee plans to inspect the shipped fuel sometime after it arrives once the cask is in the LGS cask pool.
Note that this safety evaluation is only i
concerned with the movement of the fuel handling cask within the reactor building and cooling of the SNPS fuel after removal from the transfer cask.
.. This evaluation addresses two aspects considared in the licensee's submittal and does not address the transfer process from Shoreham to Limerick.
The two issues considered in this evaluation are:
(1) Heavy loads handling which involves the movement of cask containing the SNPS fuel within the confines of the LGS reactor building, and (2) The capability of the LGS spent fuel storage pool cooling system as regards cooling the SNPS fuel assemblies stored in the spent fuel pool.
Heavy Load Handlina The reactor enclosure crane, with which the licensee plans to move the IF-300 series cask, has been found acceptable for use as a single-failure-proof crane.
The specified maximum critical crane load is 110 tons, while the IF-300-type cask with basket,17 assemblies and yokes, weigh about 85 tons. The crane bridge and trolley have travel limit switches ti prevent movement of the crane over spent fuel.
The special lifting device, or yoke, has 2 independent components; the standard lifting yoke and a redundant yoke. The standard yoke engages the cask trannions with the standard yoke's J-hooks; the yoke cross-members hold cables which are used to remove the cask head. The redundant yoke has a cradle into which the cask is lowered before moving.
Each yoke is designed in accordance with the criteria of ANSI 14.6-1977; each is designed with a safety factor af 3 to minimum component yield stress and 5 to minimum component ultimate stress, thus complying with the criteria of a single-failure-proof lifting device.
The licensee will follow the same load path that would be encountered in moving highly irradiated fuel from the plant except in reverse, i.e., movement will be from hoist-way to cask pool instead of reverse.
The head of the cask containing the Shoreham fuel will not be removed until the cask is in the cask pit, under water. After the head is removed, the SNPS fuel may be moved into either the LGS, Units 1 or 2 spent fuel pool or may be removed for examination, at the licensee's discretion.
Removal and subsequent examination is to be conducted in accordance with applicable safety requirements.
The load path from the hatch-way to the cask storage pit has been determined to be a safe load path, i.e., a path which avoids spent fuel and redundant safety shutdown equipment in the.unlikely event of a lead drop.
Coolina of the Fuel Assemblies There are no thermal / hydraulic concerns because of the extremely low heat generation rate for the irradiated core,'900 BTU /HR. This value may be contrasted to the capability of one of the Limerick fuel pool cooling systems.
. Each unit has 3 pumps and 3 heat exchangers.
With 2 pumps and 2 heat exchangers operating and a pool filled with spent fuel assemblies generating up to 16,320,000 BTU /HR, the fuel pool water is maintained below 140*F.
Conclusion for Fuel Handlina and Coolina The staff finds that movement of the series IF-300 cask from its entrance into the reactor building to the cask pool to present no handling problems since the reactor enclosure crane and yoke constitute a single-failure-proof handling system, in accordance with the provisions of Section 5.1.6 of NUREG--
0612, " Control of Heavy Loads." Such compliance assumes the possibility of a load drop to be negligibly low.
In addition, the path of the cask, from entrance into the fuel handling building to the cask pool bypasses irradiated fuel and dual or redundant safe shutdown systems so that the cask, even were a load drop to occur, would have no effect upon spent fuel or the capability of the plant to shut down safely.
The movement of individual fuel elements into either spent fuel pool from the cask also presents no problem beyond that normally encountered, and provided for, when moving irradiated fuel from either pool into a cask when such fuel has been irradiated as part of an operational core.
As noted above, in Section 2.2, there are no thermal / hydraulic concerns because the Shoreham fuel elements are generating very little heat as compared to the capability of the spent fuel pool cooling system.
Therefore, the staff finds the movement of tLa Shoreham fuel inside the LGS and subsequent storage in the spent fuel storage pools to be acceptable in that such movement and storage will be in accordance with applicable criteria, from a heavy loads and fuel handling aspect and from a thermal / hydraulic aspect. All other concerns, including that of spent fuel pool storage criticality and movement of fuel from the SNPS to the LGS are addressed el sewhere.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published (58 FR 29010) in the Federal Reaister on May 18, 1993.
Accordingly, based upon the environmental assessment, the Commission ~has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.
e e
... ' 5.0 CONCLUS!0N The Commission has concluded, based on the considerations discussed above, that:
(1) there is reascnable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such-activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
F. Rinaldi L. Kopp J. Hayes N. Wagner I. Dinitz Date: June 23, 1993 l
- t P
O P
4