ML17264A847

Application for Amend to License DPR-18,revising Spent Fuel Pool Storage Requirements

ML17264A847
Person / Time
Site:	Ginna
Issue date:	03/31/1997
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A848	List: ML17264A847 ML17264A849 ML17264A851 ML17309A614
References
NUDOCS 9704070038
Download: ML17264A847 (19)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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Rochester Gas and Electric Corporation ) Docket No. 50-244 (R.E. Ginna Nuclear Power Plant APPLICATION FOR AMENDMENT T PERATIN LI ENSE Pursuant to Section 50.90 of the regulations of the U.S. Nuclear Regulatory Commission (NRC), Rochester Gas and Electric Corporation (RG&E), holder of Facility Operating License No. DPR-18, hereby requests that the Technical Specifications set forth in Appendix A to that license, be amended. This request for change is to revise requirements associated with the spent fuel pool to reflect a planned modification to the storage racks.

A description of the amendment request, necessary background information, justification of the requested changes, and no significant hazards and environmental considerations are provided in Attachment I. This evaluation demonstrates that the proposed changes do not involve a significant change in the types or a significant increase in the amounts of effluents or any change in the authorized power level of the facility. The proposed changes do not involve a significant hazards consideration.

9704070038 9'7033k PDR ADQCK 05000244 P PDR

A marked up copy of the Ginna Station Technical Specifications which show the requested changes is set forth in Attachment II. The proposed revised technical specifications are provided in Attachment III. The Licensing Report supporting the proposed changes to the spent fuel pool storage racks is provided in Attachment IV.

WHEREFORE, Applicant respectfully requests that Facility Operating License No. DPR-18, and Attachment A to that license, be amended in the form attached hereto as Attachment Rochester Gas and Electric Corporation By Robert C. Mecr y Vice President Nuclear Operations on this3t'Jay of ~

Subscribed and sworn to before me 1997.

Notary Public DEBORAH A.PIPERN[

Notary Public rn the State of New York

~ ONTARIO COUNTY Commisrroo Expiree Nov. 23, I 9...2'

Attachment 1 R.E. Ginna Nuclear Power Plant License Amendment Request Revised Spent Fuel Storage Requirements This amendment provides the description of the license amendment request (LAR) and the necessary justifications to support an increase in the allowed spent fuel pool storage capacity.

This attachment is divided into six sections as follows. Section A summarizes all changes to the Ginna Station Technical Specifications. Section B provides background and history associated with the changes being requested. Section C provides the justifications associated with the proposed changes. A no significant hazards consideration evaluation and environmental consideration of the requested changes to Ginna Station Technical Specifications are provided in Sections D and E, respectively. Section F lists all references used in this attachment.

A. DESCRIPTION OF PROPOSED TECHNICAL SPECIFICATION CHANGES This LAR proposes to revise the Ginna Station Technical Specifications to reflect new spent fuel pool storage requirements. This change is summarized below and shown in Attachment II.

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1. The'requirement for the minimum boron concentration in the spent fuel pool is increased from 300 to 450 ppm.

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1. The requirements for storage of spent fuel in Region 1 of the storage racks is revised to include restrictions on initial enrichment and accumulated burnup as identified in new Figure 3.7.13-1.

2. The REQUIRED ACTION for not satisfying the storage requirements is revised to allow movement of the noncomplying fuel assembly to any acceptable storage location regardless of storage region.

3. Figure 3.7.13-1 is added to provide the initial enrichment and burnup restrictions for storage in specified locations in Region 1.

4. Previous Figure 3.7.13-1 is renumbered to 3.7.13-2 and revised to provide additional restrictions on acceptable storage locations for Region 2.

5. SR 3.7.13.1 is revised to remove the NOTE providing an exemption to the SR when moving a fuel assembly from Region 2 to Region 1.

6. SR 3.7.13.1 is revised to verify the additional restriction on burnup and accumulated burnup is satisfied prior to storing a fuel assembly in Region 1.

7. Figure 3.7.13-1 is renumbered to 3.7.13-2 in SR 3.7.13.2 reflecting the addition of the figure for Region 1.

D I NFEA R 4

1. Specification 4.3.1.1 (c) is changed to remove the statement concerning RGAF2 fuel storage canister not satisfying the requirements for initial enrichment and burnup of LCO 3.7.13.

2. The specification of the spent fuel pool storage capacity is changed to reflect the maximum number of fuel assemblies and storage locations.

B. BACKGROUND The original spent fuel storage racks for Ginna Station had a storage capacity of 210 fuel assemblies. In 1977 these racks were removed and replaced with racks manufactured by US Tool and Die consisting of nine rack modules utilizing a checkerboard pattern alternating storage locations with water cells. This increased storage capacity to 595 storage locations. In 1985 six of the nine racks were removed from the storage pool and modified. This modification removed the lead-in funnels over the water box locations and added boraflex inserts to all locations to effectively double the storage capacity for the modified modules, This increased overall pool storage capacity to 1016 fuel assemblies and divided the storage racks into two regions. Region 1 consisted of three checkerboard pattern rack modules capable of storing all (fresh or recently discharged) fuel assemblies. Region 2 consisted of the six modified rack modules with higher density fuel storage and restrictions based on initial enrichment, minimum accumulated burnup, and minimum decay time after shutdown.

The proposed modification will replace the three Region 1 rack modules with seven new rack modules scheduled for implementation in 1998. Six new peripheral modules can be added at some time in the future. Two of the seven new modules planned to be installed in 1998 will be designated as part of Region 2 with similar restrictions on burnup versus initial enrichment (i.e., the Region 2 area will be effectively increased).

The other five new modules will compose Region 1. Higher density fuel storage in the new racks will be possible with the use of borated stainless steel and, for Region 1, the use of a checkerboard pattern alternating fresh fuel with burned fuel satisfying minimum burnup levels based on initial enrichment. Table 1 summarizes the proposed SFP

changes.

C. JUSTIFICATION OF CHANGES This section provides the justification for all changes described in Section A above and shown in Attachment II. The justifications are organized based on whether the change is: more restrictive (M), less restrictive (L), administrative (A), or the requirement is relocated (R). The justifications listed below are also referenced in the technical specifications which are affected (see Attachment II).

Reference 1 (enclosed as Attachment IV) provides a detailed analysis of the design and licensing basis of the proposed modification. This analysis is based upon the guidance established in Reference 2 and addresses the major areas of structural, criticality, thermal hydraulic, and radiological. This analysis is summarized below to the extent necessary to provide justification for the specific proposed changes to the Technical Specifications.

C.1 More Restrictive

1. LCO 3.7.12 specifies the minimum boron concentration required for the spent fuel pool. The proposed revision increases this minimum concentration to 450 ppm from 300 ppm.

The criteria for the criticality design of both Region 1 and 2 of the spent fuel pool is based upon maintaining the K,6 .95. NRC guidelines specify that due to the postulated accident condition where all soluble boron is lost, no credit for boron can be taken under normal conditions. However the double contingency principle discussed in ANSI N16.1-1975 (Reference 3) and Reference 2 allows credit for boron under accident or abnormal conditions since only a single accident need be considered at one time. The limiting accident condition, as discussed in Section 4 of Reference 1, is an incorrectly placed assembly in Region 2. The analysis shows that the criticality criterion for this, and any other abnormal event, is satisfied by a minimum boron concentration of 450 ppm during fuel movement.

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2. LCO 3.7.13 specifies the requirements for storage of fuel assemblies in Regions 1 and 2 of the spent fuel pool to maintain K,~ ~ .95. With the proposed modification the storage configuration for fuel in Region 1 changes to a checkerboard pattern alternating between storage locations for fresh or low burnup fuel, and locations for higher burnup fuel. The fresh/low burnup fuel locations are identified on the racks by using lead-in funnels for those locations only. The proposed change to the Technical Specification adds the reference to the new Figure 3.7.13-1 which provides the criteria for determining the

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acceptability of fuel assembly storage in either of two locations based upon the initial enrichment and accumulated burnup. The analysis providing the basis for this configuration is in Section 4 of Reference 1.

With the addition of new Figure 3.7.13-1 for Region 1, the figure for Region 2 is renumbered to 3.7.13-2. This new figure contains four possible locations for fuel storage within Region 2 (versus the previous two). With the proposed modification Region 2 is expanded from the original six Boraflex lined rack modules to include two new borated stainless steel rack modules. The acceptability for storage of fuel assemblies in Region 2 is expanded to include any fuel assembly provided specified initial enrichment, burnup, and storage configuration requirements are satisfied. These requirements are based upon maintaining the criticality criteria of K,S .95, and include an assumed amount of Boraflex degradation/shrinkage. Based'n the initial enrichment and accumulated burnup the fuel assembly is determined to fall within one of four burnup domains. Each domain has specific restrictions for the storage of fuel assemblies in adjacent storage locations to satisfy the basis for the criticality analysis. The analysis that demonstrates that fuel assemblies stored in conformance with this figure satisfy the criteria of K,s .95 is in Section 4 of Reference 1.

3. SR 3.7.13.1 specifies the surveillance required prior to placing a fuel assembly into a storage location in Region 1. Consistent with the proposed change to LCO 3,7.13, the proposed change to this SR adds the reference to the initial enrichment and accumulated burnup criteria identified on Figure 3.7.13-1. Prior to the proposed modification, any fuel assembly acceptable for storage in Region 2 could be stored in any location of Region 1. After the proposed modification and revisions to new Figure 3.7.13-2, the criteria should be verified prior to movement of any assembly regardless of its initial storage location. With the proposed change, the note exempting this surveillance when moving a fuel assembly from Region 2 is no longer appropriate.

C.2 s Restrictive

1. REQUIRED ACTION A.1 for LCO 3.7.13 specifies the action to be taken given a fuel assembly is identified in a location that is not in compliance with the requirements of the LCO. With the addition of requirements for storage of fuel assemblies within each region as identified on Figures 3.7.13-1 and 3.7.13-2, the action for non-compliance must allow movement of fuel within a region to an acceptable location. That is, it is now acceptable to allow movement of the noncomplying fuel assembly to a different location within the same spent fuel pool region and still meet the LCO requirements.

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2. Design Feature 4.3.1.1 (c) contains a statement noting that a fuel canister containing rods from a region RGAF2 fuel assembly that would not satisfy the initial enrichment and burnup specifications for the storage region. With the criticality'analysis documented in Section 4 of Reference 1, the subject fuel canister will satisfy the modified requirements of LCO 3.7.13 and the statement is no longer required.

3. Design Feature 4.3.3 currently specifies the spent fuel pool storage capacity in terms of the number of fuel assemblies. With the use of consolidated fuel canisters, fuel rods from multiple fuel assemblies (typically 2 assemblies or less) can be stored in one canister which occupies one storage location. With the proposed modification, the design and licensing basis will support the storage of up to 1879 fuel assemblies in 1369 storage locations. Therefore, this specification was revised to specify that no more than 1879 fuel assemblies be stored in no more than.1369 storage locations consistent with the analysis in Reference 1.

D. SIGNIFICANT HAZARDS CONSIDERATIONS EVALUATION The proposed changes to Ginna Station Technical Specifications as identified in Section A and justified in Section C have been evaluated with respect to 10CFR 50.92(c) and shown to not involve a significant hazards consideration as described below.

Operation of Ginna Station in accordance with the proposed changes does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The design basis events considered for the spent fuel pool include both external events and postulated accidents in the pool. The external events considered are tornado missiles and seismic events, The evaluation of the postulated impact of a tornado missile is detailed in Sections 3, 4, and 6 of Reference 1. The structural evaluation indicates that there are no gross distortions of the racks or any adverse effects upon plant structures or equipment. The radiological consequences of this event indicate that offsite doses are "well within" the 10CFR100 limits.

The structural evaluation is detailed in Section 3 of Reference 1. Current state of the art methods are used in the structural analysis. The evaluation of the storage racks is based on a conservative interpretation of the ASME Boiler and Pressure Vessel Code. The evaluation of the spent fuel pool is based on a conservative interpretation of requirements set forth in the American Concrete Institute, Code Requirements for Nuclear Safety Related Concrete Structures, and American Institute of Steel Contruction, Specification for Structural Steel Buildings. The

spent fuel storage system was designed to meet all applicable structural criteria for normal (Level A), upset (Level B), and faulted (Level D) conditions as defined in NUREG-0900, SRP 3.8.4, Appendix D. The following loadings were considered:

dead weight, seismic, thermal, stuck fuel assembly, drop a fuel assembly, and tornado missile impact. Load combinations were performed in accordance with SRP 3.8.4, Appendix D. Given the evaluated seismic events, the changes in the final position of the racks are small as compared to the initial position prior to the seismic event. The maximum closure of gaps is such that no significant changes in gaps result during any single seismic event. Furthermore, the combined gap closures resulting from a combination of 5 OBEs and 1 SSE show that there are no rack-to-rack or,.rack-to-wall impacts. These evaluations conclude that under these postulated events the stored fuel assemblies are maintained in a stable, eoolable geometry, and a subcritical configuration.

As described in the bases for LCO 3.7.12 and 3.7.13, the postulated accidents in the spent fuel pool are divided into two categories. The first are those involving a loss of cooling in the spent fuel pool. The thermal-hydraulic analysis for the maximum expected decay heat loads is described in Section 5 of Reference 1. The proposed modification does not change the configuration of the available spent fuel cooling systems, the limiting design conditions for maximum decay heat load which occurs during a full core offload, or the existing requirement to maintain pool temperature below 150'F. Utilizing the three available spent fuel cooling systems, Ginna Station maintains full redundancy during high heat load conditions.

The decay heat load to the spent fuel pool is maintained within the capacity of the operating cooling system by appropriately delaying fuel offload from the reactor.

Should a failure occur on the operating cooling system, the resulting heat rates allow sufficient time to place a standby cooling system in service before the pool design limit temperature is exceeded. Increases in spent fuel pool temperature, with the corresponding decrease in water density and void formation from boiling, will result in a decrease in reactivity due to the decrease in moderation effects. In addition, the analysis demonstrates that the storage rack geometry and required fuel storage configurations result in a K,~ S .95 assuming no soluble boron allowing for the potential of makeup to the pool with unborated water.

The second category is related to the movement of fuel assemblies and other loads above the spent fuel pool. The limiting accident with respect to reactivity is the fuel handling accident which is analyzed in Section 4 of Reference 1. For both the incorrectly transferred fuel assembly (placed in an unauthorized location) or a dropped fuel assembly, the positive reactivity effects resulting are offset by the negative reactivity from the required minimum soluble boron concentration. The resulting K, is shown to be less than 0.95. The radiological consequences of a fuel assembly drop remain as described in Section 15.7.3 of the UFSAR and as discussed in Section 6 of Reference 1. Loads in excess of a fuel assembly and its

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handling tool are administratively prohibited from being carried over spent fuel.

There are no changes anticipated for either the fuel handling equipment or the auxiliary building overhead crane due to the proposed modification to the fuel storage racks. The modification is scheduled for the Year 1998 to be performed while Ginna Station is operating. Movement of heavy loads around the spent fuel pool are controlled by the requirements of NUREG-0612 and the regulatory guidelines set forth in NRC Bulletin 96-02 (see Section 3 of Reference 1). Spent fuel casks and storage racks (during removal and installation) will be moved using the auxiliary building crane and lifting attachments satisfying the single failure proof criteria of NUREG-0554, obviating the need to determine the consequences for this accident.

Based on the above, it is concluded that the proposed changes do not significantly increase the probability or consequences of any accident previously analyzed.

2. Operation in accordance with the proposed changes does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed modification does not alter the function of any system associated with spent fuel'handling, cooling, or storage. The proposed changes do not involve a different type of equipment-or changes in methods governing normal plant operation. The additional restrictions placed on the acceptable storage locations for spent fuel are consistent with the type of restriction that previously existed. The potential violation of these restrictions (incorrectly transferred fuel assembly) are analyzed as discussed above. The design, analysis, fabrication, and installation meet all the appropriate NRC regulatory requirements, and appropriate industry codes and standards.

Based on the above, the change does not create the possibility of a new or different kind of accident from any previously analyzed.

3. Operation of Ginna Station in accordance with the proposed changes does not involve a significant reduction in the margin of safety.

The Licensing Report enclosed as Reference 1 addresses the following considerations: nuclear criticality, thermal-hydraulic, and mechanical, material, and structural. Results of these evaluations demonstrate that the changes associated with the spent fuel reracking'do not involve a significant reduction in the margin of safety as summarized below:

N cl r ri icalit The established regulatory acceptance criterion is that K,~ be less than or equal to

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0.95, including all uncertainties at the 95/95 probability/confidence level, under normal and abnormal conditions. The methodology used in the evaluation meets NRC requirements, and applicable industry codes, standards, and specifications. In addition, the methodology has been reviewed and approved by the NRC in recent nuclear criticality evaluations. Specific conditions which were evaluated include misloading of a fuel assembly, drop of a fuel assembly (shallow, deep drops, and side drops), pool water temperature effects, and movement of racks due to seismic events. Results described in Section 4 of Reference 1 document that the criticality acceptance criterion is met for all normal and abnormal conditions.

Thermal-H d ulic Conservative methods and assumptions have been used to calculate the maximum temperature of the fuel and the increase of the bulk pool water temperature in the spent fuel pool under normal and abnormal conditions. The methodology for performing the thermal-hydraulic evaluation meets NRC regulatory requirements.

Results from the thermal-hydraulic evaluation show that the maximum temperature of the hottest fuel assembly, intact or consolidated canister, is less than the temperature for nucleate boiling condition. The effects of cell blockage on the maximum temperature of intact fuel and consolidated canisters were evaluated.

Results described in Section 5 of Reference 1 show that adequate cooling of the intact or consolidated fuel is assured. In all cases the existing spent fuel pool cooling system will maintain the bulk pool temperature at or below 150'F by delaying core offload from the reactor.

Mechanical Material nd r u 1 The primary safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal and abnormal loads., Abnormal loadings which have been considered in the evaluation are:

seismic events, the drop of a fuel assembly, the impact of a tornado missile, a stuck assembly, and the drop of a heavy load. The mechanical, material, and structural design of the new spent fuel racks is in accordance with NRC regulatory requirements (including the NRC OT Position dated April 14, 1978, and addendum dated January 18, 1979), and applicable industry standards. The rack materials are compatible with the spent fuel pool environment and fuel assemblies. The material used as a neutron absorber (borated stainless steel) has been approved by the American Society for Testing and Materials (ASTM), and licensed previously by the NRC for use as a neutron absorber at Indian Point 3, Indian Point 2, and Millstone 2. The structural evaluation presented in Section 3 of Reference 1 documents that the tipping or sliding of the free-standing racks will not result in rack-to-rack or rack-to-wall impacts during seismic events. The spent fuel assemblies will remain intact and the criticality criterion of k,z less than or equal to

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0.95 is met, Based upon the above information, it has been determined that the proposed changes to the Ginna Technical Specifications do not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident previously evaluated, and does not involve a significant reduction in the margin of safety. Therefore, it is concluded that the proposed changes meet the requirements of 10CFR 50.92(c) and do not involve a significant hazards consideration.

E. ENVIRONMENTALCONSIDERATION RG&E has evaluated the proposed changes and determined that:

1. The changes do not involve a significant safety hazards consideration as documented in Section D above;

2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, since:

a. the consequences following a fuel handling accident or tornado missile event remain within accepted limits (Section 6 of Reference 1),

b. it is not expected that the waste generated from demineralizer resin replacement will significantly increase (Section 6 of Reference 1),

c. anticipated waste generated from the rack replacement will be substantially decreased through decontamination prior to disposal (Section 6 of Reference 1), and

d. it is not expected that a significant activity will be released to receiving waters

- or to the atmosphere as a result of the reracking (Section 6 of Reference 1).

3, The changes do not involve a significant increase in individual or cumulative occupational radiation exposure. The occupational exposure limits'are limited by 10CFR 20 and controlled to as low as reasonably achievable by plant procedures and practices. Expected dose rates in accessible spaces adjacent to the spent fuel pool are calculated and documented in Section 6 of Reference 1. Increased storage capacity of the spent fuel pool is not expected to result in any significant increase in the radiation dose levels at the pool surface or other locations of accessibility.

Doses to the workers will be constantly monitored during the reracking. The use

of electronic dosimetry, in-pool radiation area monitors, as well as the presence of radiation protection staff will provide a high degree of assurance that doses to workers will be minimized in accordance with ALARA principles. The Radiation Protection Staff will be an integral part of this operation, and therefore will be available to support emerging requirements. The estimated total exposure for this operation will be between 8 and 12 Person-Rem. Reviews of the rerack will be conducted formally as part of the ALARA process, and documented as part of the project work scope. Additional radiological considerations are detailed in Section 6 of Reference 1.

Accordingly, the proposed changes meet the eligibility criteria of. categorical exclusion set forth in 10CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement of the proposed changes is not required.

F.. REFERENCES

1. Framatome Technologies, Inc., "R. E. Ginna Nuclear Power Plant, Spent Fuel Pool Re-reracking Licensing Report", February 1997.

2. Letter from B. K. Grimes, NRC to All Reactor Licensees,

Subject:

"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and modified January 18, 1979.

3. ANSI N16.1-1975, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

TABLE 1 PROPOSED CHANGES TO T.S.

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9g iixREPORT,~/y/;.

Sections 433 and Sections 1, 2, 3, 0 CAPACITY 1016 FUEL ASSEhIBLIES 1879 FUEL ASSEi~LIES IN 1369 STORAGE B 3.7.13 4, 5, 6, and 8 LOCATIOiNS (ASSUMES CONSOLIDATION)

Figure B 3.7.13-1 Not described Section 1 0 NUMBER OF RACKS REGIOiN 1 (176 LOCATIONS) REGIOiN 1 (294 LOCATIONS)

AND TYPE OF POISOi( 3 S.S. IVITHNO POISONS 5 S.S. iVITHBORATED STAINLESS STEEL MATERIAL REGIOiN 2 (1075 LOCATIONS)

REGION 2 (840 LOCATIONS) 2 S.S. iVITHBORATED STAINLESS STEEL 6 S.S. IVITHBORAFLEX 6 S.S iVITHBORAFLEX 6 S.S. IVITHBORATED STAINLESS STEEL Sections 3.7.13 and Section 4 0 LOADING REGION 1 REGIOiN 1 B 3.7.13 REQUIREI,KNTS NO ENRICILmm/BURNUP EiiiRICHMENT/BURNUP(TiVO DOMAINS) AND REQUIRIMFKI'S STORAGE CONFIGURATIONS Figure 3.7.13-1 REGIOiN 2 REGION 2 EiNRICI&KÃf/BURhVP ENRICIBKNT/BURi&JP(FOUR DOMAINS) AM)

REQUIREMENTS (TiVO DOMAINS) STORAGE COi IFIGURATIONS TRANSFER OF NOiN-COMPLYIiNG TRANSFER OF NON-COMPLYING FUEL ASSEMBLIES FROM REGIOiN 2 TO ASSEMBLIES IVITHINA REGIOiN REGIOiN 1 Sections 43 and Section 4 0 CONSOLIDATED ROD iN&iIBEROF RODS IN COiNSOLIDATED NO RESTRICTIONS ON THE NUMBER OF RODS IN B 3.7.13 STORAGE CANISTERS STORAGE CANISTER SHALL BE ( 144 COiNSOLIDATED ROD STORAGE CANISTER RODSOR ) 256RODS Section 3.7.12 and Section 4 0 BORON 300 PPM 450 PPiVI B 3.7.12 COiNCENTRATIOiN

Attachment II Marked up copy of R.E. Ginna Nuclear Power Plant Technical Specifications Included pages:

3.7 27 3.7-29 3.7-30 3.7-31 4.0-2 4.0-3 B 3.7-87*

B 3.7-89*

B 3.7-90~

B 3.7-91*

B 3.7-92+

B 3.7-93*

B 3.7-94*

B 3.7-96*

"'hese bases pages are under the control of RG&E and are being provided for information only.