ML20078M138
ML20078M138 | |
Person / Time | |
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Site: | Oyster Creek |
Issue date: | 11/25/1994 |
From: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
Shared Package | |
ML20078M141 | List: |
References | |
C321-94-2134, NUDOCS 9412010362 | |
Download: ML20078M138 (14) | |
Text
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4 GPU Nuclear Corporation NggIgf Post Office Box 388 Route 9 South Forked River. New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
C321-94-2134 November 25, 1994 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Facility Operating License No. DPR-16 Technical Specification Change Request No. 222 Pursuant to 10 CFR 50.90, GPU Nuclear Corporation (GPUN), operator of the Oyster Creek Nuclear Generating Station (OCNGS), Facility Operating License No. DPR-16, requests a change to Appendix A of that license.
As a result of the present unavailability of an off-site spent fuel storage facility and the need to load additional fuel assemblies to support operation during the next cycle, OCNGS will lose the capability to completely offload the reactor core after the current Cycle 15R refueling outage. We believe it is prudent to maintain core offload capability, when possible, to facilitate in-vessel inspections and repairs. GPU Nuclear is in the process of providing a dry storage facility on-site which is planned to be operational during the next operating cycle (Cycle 15). This will allow full core offloads beyond Cycle 15 until such time as an off-site spent fuel storage facility is operating. Therefore, in order to maintain full core offload capability until the dry storage facility is operational, we request that 45 existing fuel storage locations in the fuel pool be licensed for use.
The proposed Technical Specification change would allow 2645 fuel assemblies to be stored in the fuel pool. This is an increase of 45 fuel assemblies from the current limit of 2600 contained in Technical Specification 5.3.1.E. The 45 additional storage locations currently exist in the racks in the fuel pool. They were included in the re-racking project allowed by License Amendment No. 76 but were not incorporated in the Technical Specifications since, at the time, it was believed they would not be needed. These storage locations were to be used for the storage of other core components such as fuel channels and sources. The infonnation pro'.ided to the NRC staff in support of License Amendments 22,76 and 121 address most of the issues concerned with this request. Documentation previously forwarded to the staff far License Amendment 76 considered all 2645 installed storage locations.
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fR 941201o362 ADOCK h 9k h g PDR GPU Nuclear Corporahon is a subsidearv of General PuDoc UMbes Corporabon /
C321-94-2134 Page 2 This license amendment request has been reviewed in accordance with Section 6.5 of the OCNGS Technical Specifications, and using the standards in 10 CFR 50.92 we have concluded that the proposed changes do not constitute a significant hazards consideration.
The license amendment should be effective upon issuance.
This license amendment request is considered very important should a full core ofiload be necessary during Cycle 15 with the dry spent fuel storage facility not yet in service. Without the ability to fully offload the core, any inspection or repair activity will most likely result in higher personnel exposures and schedular delays. Full core offload capability, in particular, would facilitate any in-vessel repair which requires draining of the vessel.
GPU Nuclear has planned the current refueling outage to work within the constraints imposed by the inahility to utilize the additional 45 storage locations and still perform a full core offload. In order to maintain full core offload capability during Cycle 15, GPU Nuclear requests NRC approval of this license amendment request by January 30, 1995 which is a few months from restart from the Cycle 15 refueling outage.
Pursuant to 10 CFR 50.91 (b)(1), a copy of this change request has been sent to the State of New Jersey Department of Environmental Protection.
Sigerely, Barton V :e President and Director yster Creek cc: Administrator, NRC Region I NRC Senior Resident Inspector, Oyster Creek Oyster Creek NRC Project Manager
e ,
GPU Nuclear Corporation Nuclear o
- cn:r888 Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
C321-94-2134 November 25, 1994 The Ilonorable Theodore J. IIutler Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731
Dear Mayor Hutler:
Enclosed herewith is one copy of Technical Specification Change Request No. 222, for the Oyster Creek Nuclear Generating Station Operating License.
This document was filed with the United States Nuclear Regulatory Commission on November 25th ,1994.
Sincerely, j
0 Barton ice President and Director Oyster Creek JJB/PFC/ pip Attachment 1
GPU Nuc; ear Corporanco ts a subscaay of Gene'a Pubbc Ut.hties Corporation i
GPU Nuclear Corporation Nuclear :::eSNI **
Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
C321-94-2134 November 25, 1994 Mr. Kent Tosch, Director Bureau of Nuclear Engineering Department of Environmental Protection CN 411 Trenton, NJ 08625
Dear Mr. Tosch:
Subject:
Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 Technical Specification Change Request No. 222 Pursuant to 10 CFR 50.91(b)(1), please find enclosed a copy of the subject document which was filed with the United States Nuclear Regulatory Commission on 11/25 ,1994.
Sincerely, f
/
J Barton ill President and Director Oyster Creek Attachment JJB/PFC/ pip GPU Nuclear Corporation is a subsic.ary Of General Puchc Utnties Corpo<at on l
a . .
GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION Facility Operating License No. DPR-16 Technical Specification Change Request Request No. 222 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No. 222 to the Oyster Creek Nuclear Generating Station Operating License, a change to Section 5.3.
, J B
J. I a/tgn ice 'rddent and Director Creek el 1)
Sworn and Subscribed to before me this i25' day of pexpo 1994.
- 11 1' -
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\ Notary Public of NJ JUDrni M. CROWE Notary Public of New Jersey My Commission Expires *//'9 'I/ Mol) i l
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b j UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
) Docket No. 50-219 GPU Nuclear Corporation )
CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 222, for Oyster Creek Nuclear Generating Station Technical Specifications, filed with the U.S.
Nuclear Regulatory Commission on 11/25 1994, has this day of 11/25/94 been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:
The Honorable Theodore J. Hutler Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 o /
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Bv I
. J. I Vice jre frt>n ident and Director I
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OYSTER CREEK NUCLEAR GENERATING STATION <
FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 TECIINICAL SPECIFICATION CIIANGE REQUEST NO,222 Applicant hereby requests the Commission to change Appendix A to the above captioned license as below, and pursuant to 10 CFR 50.92, an evaluation concerning the determination of no significant hazards considerations is also presented: -
PROPOSED CIIANGE TO TECIINICAL SPECIFICATIONS SPECIFICATION TO BE CIIANGED Specification 5.3.1.E EXTENT OF CIIANGE Revise the number of spent fuel assemblies allowed in storage in the fuel pool to 2645 from 2600. In addition, revise bases associated with Specification 5.3.1.A to reflect updated criticality analyses. The changes are contained on Technical Specification pages 5.3-1 and 5.3-2.
BASIS FOR TIIE CIIANGE This evaluation establishes the basis to increase the licensed capacity of the Oyster i Creek Nuclear Generating Station (OCNGS) spent fuel pool to 2,645 fuel assemblies, 45 fuel assemblies more than previously approved by License Amendment No. 76.
The increased capacity allows full core discharge capability (560 assemblies) to the end of Cycle 15 by utilizing existing storage locations. The licensed capacity of 2600 will permit a full core offload for the current (Cycle 15R) refueling outage, however, all the new fuel to be loaded (172 assemblies planned) could not be staged in the fuel pool. With 172 assemblies discharged during the outage, the total number of spent assemblies in the pool will be 2048 which does not allow future full core offloads without an increase in licensed capacity.
GPU Nuclear submitted an application for license amendment on August 20,1982, as supplemented September 2 and December 20,1983, to increase the storage capacity .
of the fuel pool from 1800 fuel assemblies to 2600 fuel assemblies. GPU Nuclear letters dated May 30, June 4, and June 13, 1984 provided additional information in I
response to NRC staff requests. The staff issued License Amendment No. 76 on September 17,1984 approving the increased storage capacity The information considered by the staff in issuing Amendment No. 76 also supported the use of the :
additional 45 norage locations which GPU Nuclear elected not to license at the time. )
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The additional 45 storage locations were not licensed with License Amendment No.
76 even though they were considered in licensing documentation because it was 1
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e believed that they would be used for the storage of miscellaneous equipment such as fuel channels and not needed for fuel storage. Since it is important to maintain full core ofHoad capability, it is necessary to license these storage locations for fuel ,
storage.
The evaluation below addresses the pertinent aspects of criticality prevention, spent fuel pool cooling, structural adequacy and radiological and environmental impact.
Criticality Considerations The criticality analysis performed in support of Amendment No. 76 was superseded by an analysis submitted in support of License Amendment No.121 which removed the U-235 enrichment limitation in the Technical Specifications. The analysis considered fuel of up to 3.8% U-235 with burnable poison in the form of at least seven fuel pins containing at least 3 weight percent gadolinia (Gd203).
Blackness testing performed at Oyster Creek in October of 1993 revealed that gaps .
have developed in Boraflex sheets contained within the spent fuel pool racks.
Horaflex is a rubber-like material containing Baron which is used in the high density fuel storage racks to provide adequate neutron absorption to maintain the neutron multiplication factor (K,n) below 0.95. When Boraflex is subjected to gamma radiation from spent fuel it becomes brittle, shrinks and in combination with mechanical restraint induced during manufacture can develop gaps.
In conjunction with Boraflex gap formation, a new fuel design of higher U-235 enrichment has been loaded in the fuel pool for cycle 15 operation. A comprehensive in-house analysis was performed to determine the effect of Boraflex shrinkage in .
combination with the loading of 4.0% enriched fuel on fuel pool K,u. This analysis combines an updated criticality analysis on 4.0% enriched fuel with EPRI research data, industry findings and plant-specific surveillance test results on Boraflex. This evaluation has determined that fuel up to 4.0% U-235 enrichment with 3 weight percent gadolinia (Gd2O3 ) in at least 7 pins can be loaded in the fuel pool with the consideration of a bounding projection of Boraflex gaps and still provide margin to the Technical Specification K,y limit of 0.95.
Fuel pool storagi consists of 10 free standing rack.s to house up to 2645 spent and new fuel assemblies. Due to fuel spacing, neutron absorbing Boraflex is sandwiched hereca the stainless steel support material during rack fabrication. The Boraflex l pr.6 niend nearly the full length of the active fuel and are contained in every panel of every cell of the fuel racks. The BoraDex panels combined with cell spacing of 6.198 inches provide a margin to criticality.
A spent fuel pool criticality analysis was performed for fuel up to 4.0% U-235 enrichment with seven 3% gadolinium pins at peak reactivity. This analysis used computer codes and methodology previously approved by the NRC staff for lower enriched fue! of the same type. The fuel pool K ye including all biases and uncenainties was determined to be 0.9174. This analysis, however, did not consider 2 l l
gaps in the Bouinx panels. In order to ensure that current and future Borallex gaps do not cause an increase in fuel pool K<.y above 0.95 with 4.0% enriched fuel, an Oyster Creek specific analysis was performed. This analysis detennined a delta K corresponding to the maximum expected gap formation due to Boraflex shrinkage.
The maximum expected gap is 3.9 inches. With the conservative assumption that the gaps are coplanar this will result in a delta K of 0.028. If this value is added to the base K,y for 4.0% enriched fuel given above, the resulting fuel pool Ke n with 95%
confidence is 0.945, which is less than the 0.95 Technical Specification limit. Based on blackness testing, the maximum total gap in a Boraflex panel at this time is 2.42 inches. Although this gap is expected to increase with gamma exposure, shrinkage will saturate and gap growth will stop at a point less than 3.9 inches.
Since the criticality analysis was performed with an infinite lattice, it is valid for a spent fuel pool capacity of 2,645 fuel assemblies. ,
Spent Fuel Pool Cooline and Makeup
- The spent fuel pool cooling system consists of the original system comprised of two spent fuel pool cooling pumps and heat exchangers and an augmented system comprised of two augmented fuel pool cooling pumps and one heat exchanger, a filter, a demineralizer, two surge tanks, associated piping and valves, and interconnections to the condensate demineralizers and the condensate system.
The heat removal capacity of the original spent fuel pool cooling system is approximately 5.5 nullion BTU / hour while the capacity of the augmented system is approximately 19 million BTU / hour. During the period between refueling outages, the original cooling system is adequate to assure the Technical Specification limit for pool water temperature of 125 degrees F is not exceeded. The maximum heat load <
demand on the cooling system occurs when the reactor core (560 fuel assemblies) is completely offloaded to the fuel pool with prior discharges already in storage.
The decay heat load generated by an additional 45 fuel assemblies will not prevent the spent fuel pool cooling system from adequately maintaining fuel pool water temperature at less than the Technical Specification limit of 125 degrees F. The combined cooling capacity of the original and augmented spent fuel pool cooling systems is in excess of 20 million BTUs per hour. In response to an NRC staff request for information in support of License Amendment No. 76, GPU Nuclear letter to the NRC, dated June 4,1984, documented that the spent fuel pool can be adequately cooled for a maximum heat load of 2,732 spent fuel assemblies which includes a full core discharge. The conservative analysis shows sufficient cooling capability for 2,645 fuel assemblies to ensure that excessive thermal stresses on the pool liner are avoided.
The spent fuel pool makeup provisions remain the same as in our application for !
License Amendment No. 76. Failure of the fuel pool cooling system will not cause the fuel to be uncovered. Normal makeup to the pool is provided from the 525,000 gallon (nominal capacity) condensate storage tank at a rate of 250 gpm by a single 3
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, i condensate transfer pump. The makeup capability from this system is increased to I
about 420 gpm if both condensate transfer pumps are used. Additional makeup, at a rate of 150 gpm, can be provided from the (nominal) 30,000 gallon demineralized l water storage tank by the demineralized water transfer pumps through the use of hoses. Other sources of water are also available through the use of fire hoses or portable pumps. The 2000 gpm diesel driven fire pumps for the fire protection system can be used to provide makeup water from the fire pond to the condensate storage tank through a permanent connection. In addition, the two skimmer surge tanks which handle pool level surges contain a total of about 3500 gallons which can be pumped into the pool by the spent fuel pool cooling pumps.
Spent fuel pool water level is monitored, and high or low level is alarmed in the control room. Since the pool has no installed drains, level cannot be lowered by the cooling system below the level of the surge tank weirs.
Structural Adequacy The additional 45 storage locations were part of the fuel pool expansion of which only 2,600 fuel assemblies were licensed for fuel storage. The storage racks are designed to maintain the spent fuel assemblies in a safe configuration through all environmental and abnormal loadings, such as an SSE or impact due to spent fuel assembly drop.
Structural and seismic analyses of the racks have established margins against tilting ;
and deflection or movement to preclude impact of the racks with each other or with the pool walls. It is shown that the rack modules will undergo inf~mitesimal rotations if seismic excitation 50% over the SSE loading are imposed. The threshold of kinematic instability is not even approached.
Analyses performed to arrive at the above conclusions indicate that margins in all areas of structural concern exist. In particular, the racks are placed in the pool as individual stand-alone structures, do not load pool walls directly, and are uncoupled from pool liner temperature rise. The analyses were reviewed by the NRC staff as part of the review of the application for License Amendment No. 76.
To limit the out-of-phase motion of adjacent racks due to non-symmetric loading of the racks. Oyster Creek procedures for loading spent fuel pool racks currently require the racks to be loaded symmetrically, i.e. the total fuel assemblies stored in any one quadrant of a rack will not deviate by more than 10% of the average of the four quadrants. This limitation will remain in effect for storage of 2,645 fuel assemblies.
In summary, the additional 45 fuel bundles in storage will not decrease structural margins since 2,645 fuel assemblies were considered in the original analysis which demonstrated that the acceptance criteria were met.
_ Environmental and Radiological Immc_1 GPU Nuclear's license amendment application (!ctter dated September 2,1983) considered 2,600 storage locations in the environmental evaluation. It concluded that 4
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l an increase in storage of fuel assemblies from 1800 to 2600 created no significant l environmental or radiological impact for the following reasons: j l
- 1. The fuel pool cleanup system (filters and demineralizer) can adequately j accommodate the minor increase in the radiological burden resulting from the i expansion. l
- 2. No appreciable increase in solid radioactive waste (i.e., filters and demineralizer resins) will be generated.
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- 3. No increase in the release of radioactive gases is expected. The current levels of airborne activity are considerably lower than allowed by Table 1 of 10CFR Part 20, Appendix B.
- 4. The additional crud resulting from the expansion will have an insignificant effect on personnel exposure since the loose crud is removed by the fuel pool filters and demineralizers.
- 5. Existing radiation protection monitor systems and programs are adequate to detect and warn of any expected or abnormal increases in radiation levels.
- 6. Zircaloy cladding corrosion will not increase.
- 7. Total exposure to personnel in the vicinity of the fuel pool will not increas as a result of the increased storage.
i Considering the smaller incremental addition to the licensed storage capacity, the environmental and radiological conclusions stated above are not altered by the storage of 45 additional spent fuel assemblies. In addition, the conclusions reached in the NRC staff SER dated September 17, 1984 and the Environmental Assessment and Finding of no Significant Impact - Spent Fuel Pool Expansion dated September 13, 1984 remain applicable.
No Sienificant liarards Consideration Evaluation
- 1. The operation of the Ovster Creek Nuclear Generatine Station. in accordance with the proposed amendment. will not involve a sinnificant increase in the probability or consequences of an accident previously eyaluated.
There are no changes in the existing provisions for load handling in the vicinity of the spent fuel pool associated with the proposed increase in licensed storage capacity. OCNGS Technical Specification 5.3.1.B limits the loads carried over the spent fuel pool to no greater than the weight of one fuel assembly.
Therefore, accidents involving the mispositioning or drop of a fuel assembly establish the extent of accident probability or consequences. The Abnormal j Positioning of a Fuel Assembly Outside the Storage Rack and the Dropped Fuel Assembly accident scenarios are addressed as follows:
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- a. The probability of occurrence of the above accidents is not affected by the racks themselves or the stored fuel. Since no physical changes are being made to the racks, an increase in licensed storage capacity cannot increase the probability of these accidents,
- b. The consequences of abnormal positioning of a fuel assembly outside the storage rack were evaluated. Since the storage rack criticality calculations were made using an infinite array of storage cells with no neutron leakage, positioning a fuel assembly outside and adjacent to the actual finite rack can add reactivity, but would, because of neutron leakage, result in a lower Km than the Koo calculated for the infinite array. Thus, additional stored fuel assemblics will not increase consequences of this type of accident than those previously evaluated.
- c. The consequences of a dropped fuel assembly striking either the base of the rack or the top of a storage location and the reactivity effects were also evaluated in the licensing report supporting Amendment 76. In all cases, the mechanical integrity of the racks was not exceeded. Also, the dropped fuel assembly did not constitute a criticality hazard because the infinite multiplication factor of the fuel storage racks was not materially altered. An ,
increase in fuel enrichment does not increase consequences since the GE-9 assemblies' mechanical specifications are bounded by previous designs and consequences are not dependent on U-235 enrichment. Thus, since no physical alteration of the storage racks is necessary to store 45 additional fuel assemblies the consequences of this type of accident are not increased.
- 2. The operation of Oyster Creek Nuclear Generatine Station. in accordance with the proposed amendment. will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The increase in licensed spent fuel pool storage capacity involus the a.idition of 45 fuel assemblies. The increased structural loading has already been accounted for in the analyses reviewed by the NRC staff in support of Amendment 76.
There are no physical changes to the fuel pool cooling, makeup, filtering, handling and monitoring systems which support fuel storage.
These systems are capable of handling the additional duty originating from the additional fuel. Criticality accidents or malfunctions also do not change because the analysis assumes an infinite array of fuel and BoraHex gaps have been conservatively accounted for. Therefore, there is no possibility for an accident or malfunction of a different type than previously evaluated.
- 3. The operation of the Ovster Creek Nuclear Generatine Station. in accordance with the proposed amendment. will not involve a sienificant reduction in a marcin of safety.
The margin of safety, when applied to a storage expansion, needs to address nuclear criticality, thermal-hydraulic, mechanical, material and structural 6
adequacy.
Nuclear Criticality The acceptance criterion for criticality as established in Technical Specification 5.3.1.A, is that the neutron multiplication factor shall be less than or equal to 0.95, including all uncertainties.
Since the increase in licensed capacity to 2,600 storage locations, the maximum allowable average enrichment was increased twice. The original analysis was for 3.01% U-235 enriched fuel with no credit for Gd 20.3Subsequent analyses increased the maximum allowable enrichment to 3.8% and then 4.0% U-235.
Both analyses take credit for Gd2 03 requiring a minimum of 7 (seven) rods containing 3.0% or more Gd2 0. 3 Subsequent to the rack installation, an industry concern was raised with the discovery of the formation of gaps in Boraflex panels. The problem of gap formation in the BoraRex and its impact on criticality has been addressed. The criticality analysis was updated to take into account the presence of gaps, including projected gap formation and growth with time. The analysis used a conservative assumption that the gap formation is coplanar. The fuel pool K err for the 4.0% U-235 enriched fuel with at least 7 (seven) Gd2 03 rods at peak reactivity is 0.9174 and increases to 0.945 with 3.9 inch coplanar gaps in the BoraHex which is below the 0.95 limit. Oyster Creek maintains a Boranex surveillance program to ensure the assumptions used in the analysis remain valid.
Since all criticality analyses were performed with an infinite lattice, it is valid for a spent fuel pool capacity of 2,645 fuel assemblies. Therefore, there is no decrease in the margin of safety.
Thermal-Ilydraulic The heat load analysis performed for the expansion to 2600 licensed storage locations considered all 2.645 actual storage locations filled. Therefore, the initial conclusions are not changed and no re-analysis is required. The thermal-hydraulic calculations, which used 125 F pool water temperature, have shown that the cladding temperatures (<219 F) will be well below the local fuel pool water saturation temperature of approximately 240 F. The maximum cladding temperatures will be low enough to preclude nucleate boiling.
Analysis has demonstrated that with an abnormal heat load from 2,732 fuel assemblies in the spent fuel pool, the temperature of the pool will be maintained within the Technical Specification limit of 125 F. Therefore, since this limit will i be maintained, other restrictions such as the temperature differential of the spent l fuel pool liner will also be maintained. Thus, there is no reduction in the margin of safety from a thermal-hydraulic point of view.
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l Mechanical and Structural The additional 45 storage locations were part of the fuel pool expansion of which only 2,600 fuel assemblies were licensed for storage. The fuel storage racks are designed to maintain the spent fuel assemblies in a safe configuration through all environmental and abnormal loadings, such as an SSE or impact due to spent fuel assembly drop. Structural and seismic analyses of the racks t. ave established I margins against tilting, deflection or movement to preclude impact of the racks )
with each other or with the pool walls. It is shown that the rack modules will undergo infinitesimal rotations if seismic excitation 50% over the SSE loading are 1 imposed. The threshold of kinematic instability is not even approached.
1 Analyses performed to arrive at the above conclusions indicate that margins in all areas of strucmral concern exist. The racks are placed in the pool as individual stand-alone structures, do not load pool walls directly, and are uncoupled from pool liner temperature rise.
To limit the out-of-phase motion of adjacent racks due to non-symmetric loading of the rack.s. Oyster Creek procedures for loading spent fuel pool racks require the racks to be loaded symmetrically, i.e. the total fuel assemblies stored in any one quadrant of a rack will not deviate by more than 10% of the average of the four quadrants. This limitation will remain in effect for storage of 2,645 fuel assemblies.
In summary, the additional 45 fuel bundles in storage will not d: crease structural margins since there is no associated physical change to the storage facility and the 2,645 fuel assemblies were considered in the original analysis which demonstrated that the acceptance criteria were met.
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