ML061240217

2006/05/01-Response to NRC Request for Additional Information, Dated March 30, 2006, Related to Oyster Creek License Renewal Application

ML061240217
Person / Time
Site:	Oyster Creek
Issue date:	05/01/2006
From:	Gallagher M AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
%dam200606, 2130-06-20300, TAC MC7624
Download: ML061240217 (46)

Text

AmerGen SM Michael P. Gallagher, PE Vice Presilent License Renewal Projects AmerGen 200 Exelcn Way KSA/2-E Kennett Square, PA 19348 Telephone 610.765.5958 www.exeloncorp.com michaelp.gallagher@exeloncorp.com An Exelon Company 1 0 CFR 50 10 CFR 51 1 0 CFR 54 2130-06-20300 May 1,2006 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Reference:

Response to NRC Request for Additional Information, dated March 30, 2006, Related to Oyster Creek Generating Station License Renewal Application (TALC No. MC7624)"Request for Additional Information for the Review of the Oyster Creek Nuclear Generating Station, License Renewal Application (TAC No. MC7624)," dated March 30, 2006 In the referenced letter, the NRC requested additional information related to Sections 4.3, 4.6, and 4.7 of the Oyster Creek Generating Station License Renewal Application (LRA). Enclosed are the responses to this request for additional information.

If you have any questions, please contact Fred Polaski, Manager License Renewal, at 610-765-5935.

I declare under penalty of perjury that the foregoing is true and correct.Respectfully, Executed on OSr-o, -2oo I Michael P. Gallagher Vice President, License Renewal AmerGen Energy Company, LLC

Enclosure:

Response to 03/30/06 Request for Additional Information A((+

May 1, 2006 Page 2 of 2 cc: Regional Administrator, USNRC Region 1, w/o Enclosure USNRC Project Manager, NRR -License Renewal, Safety, w/Enclosure USNRC Project Manager, NRR -License Renewal, Environmental, w/o Enclosure USNRC Project Manager, NRR -OCGS, w/o Enclosure USNRC Senior Resident Inspector, OCGS, w/o Enclosure Bireau of Nuclear Engineering, NJDEP, w/Enclosure File No. 05040 Enclosure Response to 3/30/06 Request for Additional Information Oyster Creek Generating Station License Renewal Application (TAC No. MC7624)RAI 4.3-1 RAI 4.3-2 RAI 4.3-3 RAI 4.3-4 RAI 4.6-1 RAI 4.7.3-1 RAI 4.7.3-2 RAI 4.7.3-3 I of 16 RAI 4.3-1 Section 4.3.1 of the license renewal application indicates that the fatigue usage (based on the use of projected cycles for 60-years) for the reactor vessel closure studs, the vessel support skirt and the basin seal skirt to vessel flange junction was predicted to exceed the Oyster Creek acceptance limit of 0.8. The application also indicates that the fatigue usage Of these components was shown to be acceptable by using more refined analysis methods. Describe the more refined analyses that were performed for these components.

Respoise RPV Closure Studs: The original design analysis for the RPV and closure bolting predicted a cumulative fatigue usage factor for the head closure studs of 0.796 for 40 years of operation.

When projected to 60 years of operation the fatigue usage for the head closure bolts would exceed the original acceptance limit of 0.8. In August 2002, a revised analysis for the entire RPV fle.nge/stud region, including the RPV closure studs, was prepared to support a revised head tensioning sequence.

The new analysis included a fatigue usage calculation for the studs that used the 1995 Edition of Section III of the ASME Code, including the 1996 Addenda. As a part of The License Renewal process, the closure stud fatigue calculation was revised to use the number of actual cycles projected for 60 years of plant operation, which produces a fatigue usage value of 0.196.RPV SuDoort Skirt: A revised analysis was performed for the RPV support skirt using Section III of the 1995 Edition of the ASME Code, including the 1996 Addenda. The fatigue calculation' used the original RPV Stress Report stresses as input, but was reconciled to use the more modern ASME Code Section III methodology by modifying those stresses, where appropriate, to account for the Young's Modulus ratio, adjust for K., and to use the updated fatigue curve. In addition, the analysis used the number of cycles currently projected for 60 years, based on plant operating history. A revised fatigue usage value of 0.710 for 60 years of plant operation was obtained from this evaluation.

RPV Basin Seal Skirt: A revised analysis was performed for the RPV basin seal skirt as a part of license renewal efforts. The stress and fatigue evaluation in the original RPV Stress Report was maintained for this component with two exceptions:

(1) a finite element analysis was performed to determine refined geometric stress concentration factors for the critical location, and (2) projected cycles for 60 years of plant operation were used. To accomplish Item (1), a finite element model (FEM) was constructed that included the vessel flange, a portion of the vessel shell, the basin seal skirt, the basin seal skirt plate, and the fillet attachment weld between the basin seal skirt and the RPV flange. Tension and bending loads were applied t3 the FEIM so that revised tension and membrane stress concentration factors could be determined.

The original fatigue analysis for the basin seal skirt component was revised to incorporate the refined geometric stress concentration factors, as well as projected cycles for 60 years cf plant operation.

A fatigue usage value of 0.270 for 60 years of plant operation was 2 of 16 obtained using the 1963 Edition of the ASME Code (consistent with the original RPV Stress Report for this location).

RAI 4.3-2 Section 4.3.1 of the license renewal application indicates that the reactor vessel feedwalter nozzles were reanalyzed to account for the effects of rapid thermal cycling.The application also indicates that the analysis satisfied the original Oyster Creek reactor vessel design limits. However, Table 4.3.1-2 of the license renewal application indicates that the 40-year fatigue usage of the feedwater nozzle was 0.952. Clarify whether the reanalysis of the feedwater nozzle for the rapid thermal cycling satisfied the original Oyster Creek reactor vessel design fatigue limit of 0.8. Also, indicate when the analysis that calculated the fatigue usage of 0.952 was performed and provide the basis for its acceptance.

Respoise: In the criginal RPV Stress Report, fatigue was analyzed in accordance with GE specification 21A11C)5.

The GE specification provided a fatigue curve and established a conservative fatigue usage lactor acceptance limit of 0.8. The original RPV Stress Report fatigue analysis was computed for a 40-year period for the nozzle blend radius region, and had a predicted cumulative fatigue usage value of 0.1.As a result of the repair of crack indications found in the feedwater nozzles in 1977, the feedwater nozzles were reanalyzed, as documented in MPR Associates, Inc. Report No. MPiR-568, "Design Report for Replacement FW Sparger," December 1977. The 1977 MPR analysis did not account for rapid thermal cycling at the feedwater nozzle resulting from bypass leakage behind the thermal sleeve. However, the analysis contains one significant conservatism; the analysis assumed 60 cycles per year of on/off feedwater flow at low power conditions.

The feedwater control system was modified following the repair of the feedwater nozzles to prevent such cycling. The MPR report concluded that the feedwater nozzles would reach an allowable usage factor of 1.0 after approximately 36 additional years of plant operation (or 42 years after initial p'ant startup), including the 60 cycles/year of hot standby (on/off) flow injection cycling.The allowable value of 1.0 used in the re-evaluation was consistent with the ASME Code Section III edition that was used as a basis to perform the fatigue analysis, but was inconsistent with the acceptance criterion cited in the original RPV Stress Report. Oyster Creek has recently changed the cumulative usage factor acceptance limit for the RPV from 0.8 to 1.0, using the 10 CFR 50.59 process, making it consistent with the ASME Section III fatigue usage acceptance limit.The first step as part of the license renewal process was to establish a 40-year projected fatigue usage for the feedwater nozzles. The 1977 MPR analysis was re-visited, and a 40-year usage factor of 0.952 was developed using the thermal cycles assumed in the MPR analysis.

The evaluation was then revised to properly account for modification to prevent low power feedwater thermal cycling, as well as to account for the number of plant transients projected for 60 years of 3 of 16 operation.

From this revised evaluation, the projected 60-year cumulative fatigue usage for the feedwater nozzles was determined to be 0.368.As the final step in the license renewal effort, a refined fatigue calculation was performed usillg Section III of the 1995 Edition of the ASME Code, including the 1996 Addenda to develop detailed and up-to-date input for stress based fatigue (SBF) monitoring of the nozzles in the Fatigue. Monitoring Program (FatiguePro).

The refined license renewal fatigue calculation used the stresses from the 1977 re-evaluation as input, but modified those stresses where approp iate to use the more modern ASME Code Section III methodology (i.e., Young's Modulus ratio, K,, updated fatigue curve, etc.). The projected cycles for 60 years of plant operation were again used. A revised fatigue usage value of 0.389 for 60 years of plant operation was obtained from this evaluation.

The 1977 analytical results did not include any usage factor from rapid thermal cycling effects.The impact of rapid cycling was evaluated by MPR in 1983 (MPR-783) and found to have a small impact of cumulative fatigue usage. As a part of the license renewal process, the rapid cycling calculations were re-evaluated for bounding reactor conditions, and it was confirmed that the design of the thermal sleeve/sparger assembly and associated flow baffle installed in 1977 is effective in reducing the fatigue due to rapid cycling to a negligible value.RAI 4.C-3 Section 4.3.3.2 of the license renewal application discusses the fatigue evaluation of tile isolation condenser.

Provide the following information regarding the evaluation:

a. The application indicates that a fatigue analysis was not performed as part of the original component design. The application also indicates that a later evaluation was performed for the tube bundle replacement in 1998. The application further indicates that the design life of the tube bundle replacement is 1500 cycles. Explain how the design life of 1500 cycles was determined.

Provide the fatigue usage based on the peak stresses calculated for the Oyster Creek tube bundle replacement.

b. The application references the fatigue analysis of the Nine Mile Point Unit 1 isolation condenser.

The application indicates that the Nine Mile Point, Unit 1 isolation condenser stress and fatigue results are considered bounding for Oyster Creek. Provide a detailed discussion of how it was determined that the Nine Mile Point, Unit 1 analysis was bounding for Oyster Creek.The discussion should include a comparison of the isolation condenser sizes and the sub-component materials, geometries and thicknesses.

The discussion should also address the tube and shell thermal transients and flow rates.c. The application indicates that the isolation condenser piping outside of the containment was evaluated for fatigue as part of a leak-before-break (LEIB)analysis completed in 1991. The application also indicates that the piping 4 of 16 outside the drywell was replaced in 1992. Provide the design criteria that was used to evaluate the replacement piping, including the number and types of thermal transients analyzed.

Provide the maximum calculated fatigue usage for the replacement piping.Resnonse: (a) At the time of tube bundle replacement in 1998, stress and fatigue analyses were performed for the Oyster Creek Isolation Condensers (

Reference:

Holtec International Report No. HI-982027, Revision 0, "Oyster Creek Nuclear Plant Emergency Isolation Condenser Tube Bundle Head Stress Analysis," 9/15/98, HOLTEC proprietary).

As a part of this assessment, six cross sections were selected for stress analysis: Cross Section Description 1 Inlet pipe to nozzle junction 2 Hemispherical head attachment weld 3 Tubesheet 4 Cylinder flange 5 Tubesheet-outer casing attachment weld 6 Outer casing attachment weld A maximum alternating stress intensity of 189.4 ksi was reported for the limiting cross section based on the thermal and pressure transients evaluated.

Specific stresses for each cross-section were not documented.

Based on a bounding alternating stress intensity of 195 ksi, 1,500 allowable cycles were established for the Isolation Condensers.

Based on the above allowed cycles and the number of projected cycles presented in Table 4.3.1-1 of the LRA the following fatigue usage for each Isolation Condenser is determined to be: I40-years 60-years Isolation Condenser A 0.252 0.321 Isolation Condenser B 0.277 0.347 Note that the above usage factors are very conservative since they utilize the total cycle counts for all years of operation for OCGS even though all of the evaluated cross sections have been replaced well after plant startup.(b) At the time that the Oyster Creek License Renewal Application was prepared, a plant-specific stress and fatigue analysis had not been located for the Oyster Creek Isolation Condensers.

To be prudent and based on field experience, it was decided to add the Oyster Creek Isolation Condensers to the Fatigue Management Program based on use of the Nine Mile Point analyses.The intent was to use the Nine Mile Point evaluation as a gauge to estimate fatigue and to identify when corrective action might be necessary.

Since that time, the Holtec Report identified 5 of 16 in the response to RAI 4.3-3(a) was located. AmerGen intends to remove all references to the Nine Mile Point analysis, and utilize the Oyster Creek plant-specific analysis as a basis for including the Isolation Condensers into the Fatigue Management Program.(c) The Isolation Condenser piping outside of the drywell was replaced in 1992. As a part of that replacement effort, a fracture mechanics evaluation was performed, as well as a full ASME Code Section II (1989 Edition) fatigue evaluation on the Isolation Condenser system piping outside the drywell (

Reference:

MPR Associates, Inc. Document No. MPR-1226, Volumes I and 11, April 1991, "Oyster Creek Nuclear Generating Station Leak Before Break Evaluation of Isolation Condenser System Piping Outside Containment').

A conservative number of system actuations was assumed (10 per year for 40 years), during which the condensate return piping was conservatively assumed to undergo a step change in temperature from 70 0 F to 575 0 F (tile system design temperature).

In addition, conservative stress intensity factors were assumed for weld joints and discontinuities between pipe and fittings and within fittings.

The calculated usage factors for 40 years of operation were all found to fall below 0.2 (maximum value = 0.174 in the Condensate line), with a Code allowable of 1.0. Therefore, fatigue failure of the piping was not considered to be a concern. Based on these results, the usage factor screening criterion of 0.4, and the fact that the more bounding Isolation Condenser assembly is monitored, the Isolation Condenser piping outside of containment was excluded from the Oyster Creek Fatigue Management Program.RAI 4.',-4 Section 4.3.4 of the license renewal application discusses the evaluation of the effects; of the reactor coolant environment on the fatigue life of components and piping. Table 4.3.4-1 provides the overall environmental fatigue multipliers for the components analyzed.

Provide the calculation of the environmental factors for the RPV inlet and outlet nozzles and the feedwater nozzle. Explain how each parameter used in the calculation was determined.

Response: The environmental fatigue calculations for the recirculation inlet and outlet nozzles and the feedwater nozzle are contained in Structural Integrity Associates Calculation No. OC-05Q-314, Revision 0, "Environmental Fatigue Calculations for RPV Locations" (proprietary).

The calculations for all three of these locations are performed in accordance with NUREG/CR-65B3 (ANL-97/18), "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon arid Low-Alloy Steels," March 1998, as the limiting locations for all three components are low alloy steel material.

All three locations were evaluated in a similar fashion, based on the governing fatigue calculation for each component, with the following specifics:

Recirculation Inlet Nozzle: Bounding Fen multipliers for hydrogen water chemistry (HWC) and normal water chemistry (NWC) were determined based on maximum transient temperature, minimum (saturated) assumed strain rate, and oxygen values estimated for the recirculation 6 of 16 system. An overall usage factor was computed considering the NWC Fen value for the time period prior to HWC implementation (41 %) and the HWC Fen value for the time period after HWC implementation until the end of the 60-year operating period (59%).Recirculation Outlet Nozzle: The fatigue usage at the outlet nozzle is greater than at the inlet nozzle, primarily because the outlet nozzle experiences the added thermal transients associated with operation of the Isolation Condenser.

As a result, the fatigue usage for the outlet nozzle was calculated using a more detailed approach.

Load pair specific Fen multipliers for hydrogen water chemisl:ry (HWC) and normal water chemistry (NWC) were determined based on the maximum load pair temperature, average computed (tensile) strain rate, and oxygen values estimated for the recirculation system. Because load-pair specific Fen multipliers were determined based on load-pair specific strain rates and temperatures, the overall Fen multiplier for the recirculation outlet nozzle was determined to be significantly lower than the bounding value described above for the recirculation inlet nozzle. An overall usage factor was computed considering the NW`G Fen value for the time period prior to HWC implementation (41 %) and the HWC Fen value for the time period after HWC implementation until the end of the 60-year operating period (59%).Feedwater Nozzle: Similar to the recirculation outlet nozzle, load pair specific Fen multipliers For hydrogen water chemistry (HWC) and normal water chemistry (NWC) were determined based on the maximum load pair temperature, average computed (tensile) strain rate, and oxygen values estimated for the feedwater system. An overall usage factor was computed considering the NWC Fen value for the time period prior to HWC implementation (41 %) and the HWC Fen value for the time period after HWC implementation until the end of the 60-year operating period (59%).As a result of a recent review of our documents, it was noticed that some of the fatigue usage factors cited in Section 4.3 of the OC LRA do not correspond to the values in our latest fatigue calcula :ions. Some fatigue values changed slightly as a result of incorporation of comments during the finalization of these calculations.

As part of the corrective actions for this finding a complete review of all of the values cited in LRA Section 4 was performed to ensure the fidelity of the information provided.

The changes are shown in bold face on the new Table 4.3.4-1 (below). The changes are relatively small and do not impact the conclusions discussed in the application.

7 of 16 Table 4.3.4-1 Environmental Fatigue Results for Oyster Creek for NUREG/CR-6260 Components 60-Year 60-Year Fatigue Ov erall Fatigue Usage Factor Environmental NUREG/CR-6260 Equivalent OCGS Material Usage Factor with Fatigue Location Location Environmental Multiplier Effects (2)Reactor Vessel (Lower Reactor Vessel (Vessel- Low Alloy Steel 0.0004 0.0042 10.28 Head to Shell Head Junction)Transition)

Feedwater Nozzle Feedwater Nozzle Low Alloy Steel 0.3889 0.8433 2.17 Recirculation System (RHR Return Line Tee Lin Condenser Return and the RPV inlet and Line Tee into SDC Line Stainless Steel 0.1205 0.43 3.57 outlet nozzles)RPV inlet nozzle Low Alloy Steel 0.0151 0.1554 10.28 RPV outlet nozzle Low Alloy Steel 0434- 0.978 5.34.1832 Core Spray System Core Spray Nozzle Low Alloy Steel 0.0013 0.0129 10.28 (Nozzle and Sare End) Core Spray Nozzle Safe End Stainless Steel 0.0006 0.0072 12.48 Residual Heat Bounded by Isolation Stainless Steel N/A N/A N/A Removal Line Condenser Return Line Tee (Tapered Transition)

Location Above Feedwater Line Limting Class I Location in Carbon Steel 0.0789 0.48 (Feedwater/RCIC Tee the Feedwater Line 0.0245 0.0767 .3.13 Connection)

Notes: I. Revised fatigue usage factors were computed for all of the NUREG/CR-6260 components based on projected cycles for 60 year.; of plant operation and updated ASME Code fatigue methodology.

2. Environmental fatigue usage was computed using the methodology of NUREG/CR-6583 (for carbon/low alloy steels) and NUREG/CR-5704 (for stainless steels), as appropriate for the material for each location.8 of 16 RAI 4.6-1 Section 4.6 of the license renewal application discusses the fatigue of the primary containment.

The application indicates that a structural evaluation of drywell thinning at various locations was performed in 1986 and 1987. Describe the structural evaluation that wais performed and indicate whether the evaluation involved any TLAAs.Resporise:

As stated in the Section 4.6 of the LRA, drywell shell plates were not evaluated for fatigue. The structural evaluation of drywell wall thinning cited in Section 4.6 of LRA refers to the statement made in 3.8 of the UFSAR. These evaluations are discussed in Section 3.8.2.8 of the UFSAR..LRA section 4.7.2 discusses drywell corrosion as a TLAA. Updated information regarding the evaluation of drywell wall thinning is provided in the response to RAI 4.7.2-1, which was transmitted to the NRC on April 7, 2006 in AmerGen letter 2130-06-20289.

RAI 4.7.3-1 The staff needs the following additional information to complete its review of this TLAA: a. An explanatory figure of the equipment pool and the reactor cavity wall areas affected by the rebar corrosion and leakages.b. The extent of areas of walls affected by the corrosion and leakages.c. Calculated maximum stresses in the affected rebars during (1) normal operating condition, (2) the postulated accident condition, and (3) during the postulated seismic event for which the walls are designed.d. The effect of the 60-year corrosion on the stresses calculated in Item c.above.Response: a. The attached explanatory figure (Figure 1) provides a plan of the area affected by rebar corrosion and water leakage. Locations affected by corrosion and leakage are localized and noted on the figure. Water and rust stains were observed around hairline cracks on the exterior surfaces of these walls in 1986. As a result, these areas were considered suspect for rebar corrosion.

The walls are also affected by the elevated temperature in the upper region of the drywell, evaluated under Integrated Plant Assessment Systematic Evaluation Report, SEP Topic 111-7B.9 of 16

b. Areas of reactor cavity and equipment pool walls affected by rebar corrosion are limited to localized portions of the walls between elevation 95' and 119'. These local areas were documented in a Material Nonconformace Report MNCR #86-870 in 1986 when a reddish brown deposit (rust like) was observed in and around hairline cracks in the walls. Later, it was determined that these deposits were from iron oide corrosion products from the embedded reinforcement steel and the corrosion was most likely a product of water leakage during refueling outages. It was considered probable that treated water entered the pre-existing cracks in the concrete wall, which wetted the surface of the rebar (Ref. 1). Based on these determinations GPU concluded that a corrosion damage assessment was necessary to establish the degree of rebar corrosion.

To accomplish the corrosion assessment, concrete core samples were taken and tested as described in response to RAI 4.7.3-2 (1) below to determine if water intrusion into the cracks created an environment that is aggressive to the rebar. lThe test results show that the environment is not aggressive and only minimal rebar corrosion, if any, should be expected.

However since rust was observed in and around the hairline cracks on the walls, GPU conservatively used 0.001 inch/year corrosion rate, as discussed in response to RAI 4.7.3-2 (1) below. Using this corrosion rate, Oyster Creek estimated that the rebar diameter will be reduced by 0.002 inch/year and that the diameter of the affected #8 and #11 rebar will be reduced by approximately 8% and 6% respectively over a 40-year period. This information was submitted to the NRC Staff in a letter dated December 5, 1990 (Flef.2). Engineering evaluation and subsequent analysis determined that the reduction in rebar diameter has no impact on structural integrity of the affected reinforced concrete walls as discussed in item c below.In 1993 GPU conducted additional evaluations to assess the condition of the rebar using a corrosion rate that is based on plant operating experience.

For this evaluation GPU reviewed the loss of metal in the upper region of the drywell shell thickness that are based on actual UT measurements.

The review indicated that drywell shell thickness in this area was reduced by approximately 0.020 inches.Conservatively assuming the affected #8 and #11 rebar experienced the 0.020 inches corrosion all around; the rebar diameter would be reduced by 0.040 inches.This represents approximately 8% reduction in cross section area of #8 rebar and 6%reduction in the cross section area of the #11 rebar.The GPU evaluation also noted that given the minimal amount of time the rebar is exposed to moisture, the fact that concrete provides an alkaloid environment which limits corrosion of reinforcing, and the fact that no additional indications of corrosion have been observed, GPU believes significant corrosion has not occurred and will not occur in the future. This information was transmitted to the NRC Staff in a letler dated November 19, 1993.The NRC review of information submitted in the November 19, 1993 letter was documented in the drywell shield wall Safety Evaluation Report dated May 11, 1994.The Staff found GPU's evaluation acceptable but was concerned about concrete 10 of 16 cracks on the outside surfaces of the drywell shield wall and the potential of additional corrosion that could occur because of water leakage or spills. The Staff recommended the upper portion of the drywell shield wall be monitored and repair, as necessary, cracks greater than 0.02 inches in width.As recommended by the Staff, the upper portion of the drywell shield wall is monitored for cracks during refueling outages. To date, cracks greater than 0.02" have not been identified and no repairs have been made. In addition, exterior surfaces of the wall are observed for water and rust stains. Inspections conducted in 1994,1996,1998, 2002, and 2005 identified no indications of water stains or rust stains and no concrete spalling or additional significant cracking was observed.

This provides reasonable assurance that significant rebar corrosion is not occurring.

Based on the information above and as discussed in more detail in response to RAI 4.7.3-2 item (2) below, AmerGen concurs with GPU's conclusion that significant corrosion has not occurred in the current term. AmerGen evaluated and concluded that significant corrosion will not occur during the period of extended operation.

However, because the rebar is inaccessible for direct visual examination, AmerGen is conservatively assuming an additional 0.010 inches loss of metal all around the #8 and the # 11 rebar. This represents a total rebar corrosion of 0.030 inches (0.020 inches during the current term plus 0.010 inches during the extended period of operation).

This results in approximately 13% reduction of cross section area of the#8 rebar and 8% reduction in cross section area of # 11 rebar. The impact of this reduction in rebar stress is addressed in item c below.c. The calculated maximum stress used to evaluate the affected rebar by corrosion is 32.8 kips per square inch (ksi). This maximum stress is based on the comprehensive analysis conducted by GPU to assess the impact of observed cracking and elevated temperature on the spent fuel pool structure and the drywell shield wall. The analysis was conducted using a finite element ANSYS model of the north side of the reactor building, which includes half of the drywell shield wall (DSW) and the spent fuel storage pool (SFP). The results of the analysis were summarized in ABB Impell Report No. 03-0370-1341 and transmitted to the NRC Staff in a letter dated September 1992. The results of the analysis show that for load combinations that involve operating and seismic loads (combinations 3.3.2c and 3.3.2d in ABB Impell Report No. 03-0370-1341, page 39), the maximum calculated stress is 32.8 kip per square inch (ksi). This maximum stress is only in a few elements in the area of the fuel transfer canal on the south wall of the spent fuel pool. The areas affected by rebar corrosion are in the south side of the reactor building, away from the transfer canal and from the heavily loaded spent fuel pool area. Thus, using 32.8 ksi stress for areas affected by rebar corrosion is very conservative because loads in the north side half of the reactor building are significantly higher than the south half of the building due to the fuel pool structure weight and the weight of the high density spent fuel racks.In a letter from Alexander W. Dromerick (NRC) to John J. Barton (GPU), "Request for Additional Information -SEP Topic 111-76, Shield Wall Temperature", dated July 26, 1993, NRC requested GPU provide numerical values of stresses under load II of 16 combinations 3.3.2c and 3.3.2d in the concrete and reinforcing bars in the drywell shield wall, above elevation 95 ft. The NRC also requested GPU to discuss the measures taken (if any) to prevent the migration of moisture through the cracks to alleviate rebar corrosion.

In response to the request, GPU transmitted to the NRC: in a letter dated November 19, 1993, ABB Impell Report No. 0037-00196-01 (see Attachment 1). The report summarizes rebar and concrete stresses at locations used to evaluate the capacity of the drywell shield wall above elevation 95 ft. for load combinations 3.3.2c and 3.3.2d. These load combinations include normal operating loads, and design basis seismic loads. The maximum calculated reinforcement stresses in critical locations (worst area) are 32.8 ksi for load combination 3.3.2c cand 31.4 ksi for load combination 3.3.2d.As discussed in response to item b of this RAI, the estimated reduction in rebar cross section area, in locations affected by this rebar corrosion, through the period of extended operation is 13% for #8 rebar and 8% for #11 rebar. This results in a str.ss increase of 14.5% for the #8 rebar and 9.1% for the #1 1 rebar.Based on stress increases discussed above and using the maximum calculated stress of 32.8 ksi, the maximum stress in the #8 rebar affected by corrosion is 37.6 ksi and 35.8 ksi for the #11 rebar. These stresses remain below ACI yield stress of 40 ksi. The calculated 32.8 ksi stress is overly conservative for the DSW affected by the rebar corrosion because it is based on the highly loaded spent fuel pool (high density racks) and for the highly stressed area around the slot in the south wall of the fuel pool.d. As discussed in TLAA 4.7.3 analysis, periodic inspections of the reactor cavity and equipment pool walls conducted since the mid 1 990s show no signs of water intrusion or indications of further deterioration of the rebar. The TLAA was based on the corrosion rate of 0.001 inch/year reported to the NRC in the December 5, 1990 letter. However after submittal of the LRA, the more recent corrosion information identified to the NRC in the November 19, 1993 was discovered.

This 1993 letter informed NRC that GPU believes that corrosion is not ongoing.Although there is no evidence of continuing rebar corrosion, AmerGen is conservatively assuming a corrosion of 0.010 inches all around the rebar during the period of extended operation, in addition to the assumed corrosion of 0.020 inches all around for the current term. This results in a total assumed corrosion of 0.030 inches, yielding a reduction of cross section area of 13% for #8 rebar and 8% for #11 rebar. The maximum tensile stress in rebar affected by corrosion is found to be 3,7.6 ksi, for the reinforcing steel having the minimum yield strength of 40 ksi. Since the corrosion continues to be localized there is no significant impact on structural integrity of the reinforced concrete walls. See response to RAI 4.7.3-2, (2) for additional rationale why the corrosion rate is conservative.

12 of 16 RAI 4.7.3-2 The staff requests the applicant to provide (1) the bases for the corrosion rate established in the Analysis, (2) assertions that these rates will not be exceeded during the period of extended operation, and (3) a summary of the program for monitoring the actual corrosion of the rebar during the period of extended operation.

Respooise:

(1). The corrosion rate was derived based on chemical analysis of concrete core samples taken on a location that is representative of the drywell shield wall and the equipment pool walls concrete.

The samples were analyzed via standard gravimetric, titrimetric, EDAX (Energy Dispersive X-ray), and leachate ion chromatography techniques.

In addition, a pH determination was derived from the leachate sample. The samples were analyzed for total composition, chlorides, and sulfates.

The test results indicated that rebar is exposed to a non-aggressive environment that contains 10 ppm chlorides, 890 ppm sulfates, and a pH of 11 .6.Based on the results of these analyses it was concluded that only a mild corrosion environment would exist due to an absence of aggressive levels of contaminants within an alkaline environment.

Under this type of environment and considering that this rebar is not continuously wetted, it is estimated the rate of corrosion would be approximately 0.001 inch/year.

Published corrosion data in Reference 4, for carbon steel (not rebar) in alkaline environment, was used as input to establish the corrosion rate of 0.001 inch/year.

This was considered appropriate since the environmental conditions within the crack annulus are PH controlled rather than oxygen controlled.

For evaluation of the walls, Oyster Creek conservatively estimated that the rebar diameter will be reduced by 0.002 inch/year as reported to the NRC in a letter from GPU to NRC dated December 5, 1990.In 1993 GPU conducted additional evaluations to assess the condition of the rebar using a corrosion rate that is based on plant operating experience.

For this evaluation GPU reviewed the loss of metal in the upper region of the drywell shell thickness that are based on actual UT measurements.

The review indicated that drywell shell thickness in this area was reduced by approximately 0.020 inches. Conservatively assuming the affected #8 and#11 rebar experienced the 0.020 inches corrosion all around; the rebar diameter would be reduced by 0.040 inches. This represents approximately 8% reduction in cross section area of 48 rebar and 6% reduction in the cross section area of the #11 rebar.The GPU evaluation also noted that given the minimal amount of time the rebar is exposed to moisture, the fact that concrete provides an alkaloid environment which limits corrosion of reinforcing, and the fact that no indication of corrosion has been observed, GPU believes significant corrosion has not occurred and will not occur in the future.Based on this information and as discussed in more detail in response to RAI 4.7.3-2 itern (2) below, AmerGen concurs with GPU's conclusion that significant corrosion has not 13 of 16 occurred in the current term. AmerGen evaluated and concluded that significant corrosion will not occur during the period of extended operation.

However, because the rebar is inaccessible for direct visual examination, AmerGen is conservatively assuming this rebar would be subject to additional corrosion of 0.010 inches all around the rebar during the period of extended operation.

(2). Oyster Creek asserts that the corrosion rate used to evaluate rebar corrosion is conservative and the rebar yield stress of 40 ksi will not be exceeded during the period of extended operation.

First, the estimated corrosion of 0.020 inches for the current term is based on carbon steel in a slightly corrosive environment.

The rebar is not subject to a corrosive environment as shown by concrete test samples. The assumed 0.010 inches for the period of extended operation is also conservative because there is no evidence of ongoing corrosion based on the existing monitoring activities in accordance with the Structures Monitoring Program (B.1.31).Secondly, rebar embedded in concrete is passivated by the alkalinity of the concrete mix by forming a protective hydrous ferrous oxide on their exposed surfaces.

Even when portions of the reinforcements are exposed via cracks in the concrete, which acts as a passageway for environmental contact, the rate of corrosion is generally low due to the barrier effect of the pre-existing oxide film. The limited corrosivity under these conditions within a crack annulus is a product of the alkaline leachant from the concrete and the slow diffusion of oxygen within the annulus and through the protective oxide layer. This type of condition would promote a weak electrochemical corrosion cell, precluding dissolution of the protective film.Thirdly, the cause of corrosion was attributed to water leakage from the reactor cavity and equipment pool during refueling outages. The source of leakage has been investigated extensively and determined to be due to cracks in the stainless liner of the wall. The cracks are now sealed with a strippable coating prior to filling the reactor cavity and the equipment pool with water. The strippable coating has been found effective in minimizing water leakage. AmerGen has made a commitment (see AmerGen letter to NRC dated April 4, 2006) to continue applying the strippable coating during the period of extended operation.

Fourth, the water used to fill the reactor cavity and the equipment pool is treated in accordance with BWRVIP-130 guidance as described in Oyster Creek Water Chemistry aging management program (B.1.02).

The treated water maintains an environment that is non-aggressive consistent with concrete sample test results described in item (1) above.Also as discussed in NUREG-1801 Rev. 1, and EPRI Report #1002950, corrosion of embedded steel in concrete is not significant if the steel is not exposed to an aggressive environment defined as concrete pH<1 1.5 or chlorides

>500 ppm. Oyster Creek concrete samples test, described in response to RAI 4.7.3-2 (1) above indicate that concrete pH=1 1.6, and chlorides=10 ppm. Thus the reinforcement is exposed to a non-aggressive environment and the corrosion is expected to be insignificant.

On the technical basis described above, AmerGen asserts that the estimated total corrosion of 0.020 inches all around the rebar diameter and the assumed of corrosion of 0.010 inches during the period of extended operation is bounding and will not be exceeded during the 14 of 16 period of extended operation.

Visual inspections conducted in 1994,1996,1998, 2002, and 2005, in accordance with the Structures Monitoring Program (B.1.31), identified no indications of water stains, or rust stains. This provides objective evidence and reasonatle assurance that significant rebar corrosion is not occurring and that the walls will continue to penorm their intended function during the period of extended operation.

(3). The walls affected by rebar corrosion are in the scope of the Oyster Creek Structures Monitoring Program (B.1.31).

The walls will be inspected every refueling outage while the reactor cavity and equipment pool are full of water to ensure that water leakage during refueling is detected.

The walls will be visually inspected for new cracks, crack growth, water stains, and rust stains. Monitoring these parameters provides reasonable assurance that significant rebar corrosion will be detected before a loss of an intended function.RAI 4.7.3-3 The staff requests the applicant to provide the quantitative aspect (i.e., corrosion rate and amount of corrosion predicted) in Section A.5.3 of the UFSAR Supplement.

Respo ise: The license renewal application (LRA) Section A.4.5.3 will be revised to reflect the following paragraph:

Corrosion of reinforcing bar in localized areas of the reactor cavity and equipment pool walls was suspected as a result of observed rust in and around cracks in the walls between elevation 95'and 119'. To assess the condition of the reinforcing bars, concrete core samples were takern in 1988 and chemically analyzed to determine if water intrusion into concrete cracks created an enviror ment that is aggressive to rebar. These analyses showed that the environment is nol aggressive and thus corrosion should not be significant.

However because of the observed rust like substance in and around the cracks, the affected rebar were conservatively assumed to be subject to corrosion of 0.020 inches all around the rebar during the current term. Engineering analysis concluded the corrosion amount of reinforcing bars would not impact structural integrity of the affected walls during the current period of operation.

For the period of extended operation, corrosion of the reinforcing bars and the rate of corrosion is a TLkA. Although there is no evidence of continuing rebar corrosion, AmerGen is conservatively assuming additional corrosion of 0.010 inches all around the rebar during the period Df extended operation.

Corrosion of.the reinforcing bar has been projected to the end of the extended period in accordance with 10 CFR 54.21(c)(1)(ii), and determined that the intended function of the drywell shield wall and the equipment pool wall will be maintained through the period of extended operation.

15 of 16

References:

1. Letter from R. W. Keaten (GPU) to U.S. Nuclear Regulatory Commission, SEP Topic 111-71, Drywell Shield Wall Integrity, dated April 19, 1994 2. Letter from J. C. DeVine, Jr. (GPU) to NRC, Oyster Creek Drywell Containment, dated December 5, 1990 3. Letter from Alexander W. Dromerick (NRC) to John J. Barton (GPU), Evaluation of Effects of High Temperature on Drywell Shield Wall and Biological Shield Wall, SEP Topic 111-7.1 "Design Codes, Design Criteria, Load Combinations, and Reator Cavity Criteria", dated May 11, 1994 4. Uhlig's Corrosion Handbook, 2 nd Edition 5. Letter from R. Keaten (GPU) to NRC, Request for Additional Information-SEP Topic 111-71, Shield Wall Temperature, dated November 19, 1993 16 of 16 RAI 4.7.3-1 Figure 1 Local Areas Affected by Rebar Corrosion I I I I I 41AMR AM80=S' 'O G I 2 7 '.'gt .4 .I:a ' , -.9A .-I -- -* ,! .3 IL \4.= 1.I, .__. .._,_ / / \ \ p. 957'Areal Lca Areas AfcebyRaReactoson(re Building3 Note:Lclaesbtene.9'&e.19 hr ln @ 9 wae an-utsan wr-bevd Ii.1 Loa Ara fetdb ea oroin(ra1.,&3 RAI 4.7.3-1 Attachment 1 ABB Impell Report No. 0037-00196-01 GPU Nuclhr Corporatlon AlpOne Upper Pond Road lNuc e rPars ippany, New Jersey 07054 201-316.7000 TELEX 136-482 Write's Direct Dial Number: November 19, 1993 U. S. Nuclear Regulatory Comnission Att: Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)Docket No. 50-219 Request for Additional Information

-SEP Topic III-7B, Shield Wall Temperature (TAC No. M76879)Your letter dated July 26, 1993 requested additional information relating to SEP Topic 111-78.As requested, we have calculated stresses in the concrete and reinforcing bar;in the drywell shield wall above elevation 95 ft. We have also determined that the stresses are below allowables considering the existing (cracked)condition of the shield wall. Attached is a report (ABB Impel1 Corporation Report No. 0037-00196-01, Rev. 0, dated September 20, 1993) which provides stresses in the concrete and steel and percent margin compared to allowables at each. critical section. As you requested, all values are provided for load combinations 3.3.2c and 3.3.2d from the original Impell report (p. 39).The analysis predicts cracking of the outside surface of the drywell shield wall ;above elevation 95'0". However, the analysis does not predict cracking of the) inside surface except locally around the notch in the south wall of the spent fuel pool. A visual inspection of the outside surface of the wall was conducted and no indication of reinforcing corrosion was observed.During normal plant operation, very little moisture is present in the vicinity of the drywell shield wall due to the relatively high temperatures.

During refueling activities, the reactor cavity is flooded and the inside surface of the wall is exposed to the water. However, a steel plate covers this surface and prevents the water from directly contacting the concrete.

While water leaks past this plate, it is not expected to cause substantial corrosion of the reinforcement due to the small percentage of time the cavity is flooded.GPU Nuclear Corporation Is a subsidiy of General Public Utilities Corporation C321-93-2300 Page 2 To obtain an estimate of the amount of corrosion that might have occurred, we reviewed loss of metal in the drywell shell in this area of the plant. In the upper regions of the drywell, approximately

.020 inches In thickness of the drywell shell has been lost. If the No. 8 and No. 11 reinforcing bars present.in the vicinity of the notch lost this amount of metal all around, the reduction in steel area would be only 10% and 4% respectively.

Given the minimal amounts of time the reinforcing is exposed to moisture, the fact that concrete provides an alkaloid environment which limits corrosion of reinforcing, and the fact that no indication of corrosion has been observed, GPU Nuclear is confident that significant corrosion has not occurred and will not occur in the future. Since significant margin exists between calculated reinforcing stresses and allowable stresses, and since the analysis indicates the spent fuel pool does not require any structural support from the drywell shield wall, GPU Nuclear believes that the adequate structural integrity of the shield wall is maintained.

Sincerely, R. Keaten Director, Technical Functions Attachment RK/YN/pl p cc: Administrator, Region I NRC Resident Inspector Oyster Creek NRC Project Manager

.S .OYSTER CREEK NUCLEAR GENERATING STATION ADDI1IONAL STRESS NFORMATION FOR THE DRYWELL SHIEL WALL FROM THE STRUCTURAL EVALUATION OF THE SPENT FUEL POOL Prepared for: GPU Nuclear Corporation One Upper Pond Road Parsippany, NJ 07054 I1 Prepared by.ABB Impel! Corporation 225 Scientific Drive Technology Park / Atlanta Norcross, GA 30092 Report No. 0037-00196-01 Revision 0 September 20, 1993 ABOl ASEA BROWN BOVMI ABB Impell Corporation G;PU Nuclear Corporation Oyster Cmeek Nuclear Generating Station Additional Stress Infom uation for the Drywdl Shield Wall from the ,tructural Evalustion of the Spent Fud Pool ABB Impell Report No. 0037-00196-01 Revision:

0 Page 2 o 23 T~AlLE OF CONTENT ageo.1.0 UDTMODUCTION 3 2.0 METHODS OF ANALYSIS 5 3.0

SUMMARY

OF RESULTS

4.0 CONCLUSION

S 6 22

5.0 REFERENCES

23 ABB ASEA OROWN BOVEAi ABB Impell Corporation GPU Nuclear Corporation ABB Impell Report No. 0037-00196-01 Oyster Croedc Nuclear Generating Station Revision:

0 Additional Sxtres Intornmaion for the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool Page 3 of 23 1.0 INRODUCTION 1.1 _kjectiys The objective of the structural evaluation of the Spent Fuel Pool (SFP) at the Oyster Creek Nuclear Generating System (OCNGS), initiated under GPUN Contract No. PC-082008, was to evaluate the SFP concrete structure for consolidated and unconsolidated fuel loads with other design basis loads. The general technical requirements for this evaluation are defined in GPUN Specification SP-1302-53-047, Revision 1, [Ref. 8].Tlie analysis considers the effects of dead load, live load, thermal gradients, seismic load and cask drop accident using prescribed loads and load combinations.

The specific evaluation of section capacities and stresses are performed in accordance with ACI-349 JRlef. 6]. A detailed finite element model of the SFP concrete structure including all connecting and supporting members was generated to consider the effects of internal force redistribution and to obtain forces on the Reactor Building structural elements.

The results and conclusions of the analyses are reported in ABB Impell report no. 03-0370-1341, Revision 0, [Ref. 1].(OPUN has requested, under contract no. PC-0443867, [Ref. 7], numerical values of stresses in the concrete and reinforcing steel in the OCNGS drywell shield wall above elevation 95 ft. These stresses are to be provided for load combinations 3.3.2c and 3.3.2d as contained on page 39 of ABB Impell report no. 03-0370-1341, revision 0,[Ref. 1]. Additionally, GPUN requested that the stresses be evaluated per ACI 349,[Ref. 6].This report provides a summary of the stresses in the drywell shield wall above elevation L'S ft., and demonstrates that the stresses in the concrete and the reinforcing steel are within allowables.

1.2 'coz The scope of work for this project is as follows: 1.2.1 :Retdgye backut, files-for SfP analysis The backup tapes for the Spent Fuel Pool Analysis are obtained from storage. TIe ANSYS postprocessing files for iteration 4S, Analysis Case C, load combinations 3.3.2c and 3.3.2d from the previous analysis are reloaded on to the computer.

This iteration ABBl ASEA BROWN KNIO ABB Impell Corporation CPU Nucleir Corporation Oyster Creek Nuclear Generating Station Additional Stress Information for the Drywell Shicld Wall from the Structural Evaluation of the Spent Fuel Pool ABB Impell Report No. 0037-0019601 Revision:

0 Page 4 of 23 is chosen because it represents the most realistic estimation of the present stiffness of the is chosen because it represents the most realistic estimation of the present stiffness of the Roactor Building.1.2.2 Mirzification of retrieved data The results for load combinations c and d are examined closely to ensure that the appropriate load combinations are loaded. This is accomplished by comparing smess results to those reported in the SFP Report and Calculations.

1.2.3 generate color contour gtress plots and tables These postprocessing files are then used to produce plots and lists of the vertical and hoop stress levels in the drywell shield wall above elevation 95 feet for load combinations 3.3.2c and 3.3.2d. Both the outer and inner surfaces of the drywell shield wall are examined.

The load capacities of the concrete and reinforcing steel at the highest stressed area in this region are determined and reported.1.3 E;rief Summary of Reslits The results of the review of stresses in the drywell shield wall above elevation 95 ft., demonstrate that all sections meet the requirements of ACI 349, [Ref. 6], for load combinations 3.3.2c and 3.3.2d.ABSt&SWA BROWNP RI ABB Impell Corporation OPU Nuclear Corporation Oyster Crea Nuclear Generating Station Additional Stress Infonnation for the Dryvell Shield Wall from the Structural Evaluation of the Spent Fuel Pool ADB Impell Rcport No. 0037-00196-01 Revision:

0 Page 5 of 23 2.0 METHODS OF ANALYSIS Ihe following two load combinations are loaded to obtain stresses in the concrete shield wall: c. 0.75 (1.4D + 1.7L + 1.4T, +/- 1.9E)d. D+L-+ To+/-E'Where: DU= dead load as specified in Section 3.1.1 of [Ref. 1], Rack Conditions 2 and 3= design live load as specified in Section 3.1.2 of (Ref. 1]I.0-thermal load due to temperature differential across the slab or wall. Two critical cases were considered as specified in Section 3.1.4 of [Ref. 1]I OBE seismic load as specified in Section 3.1.3 of (Ref. 1]I' = SSE seismic load as specified in Section 3.1.3 of [Ref. 1]loC -Cask drop load as specified in Section 3.1.5 of [Ref. 1]Verification that appropriate load combinations are loaded is contained in ABB Impell Calculation no. 0037-00196-C002, revision 0, [Ref. 3]. Contour stress plots and numerical values of stresses in the drywell shield wall above elevation 95 ft., are provided for the above load combinations.

An ACI evaluation is also performed to demonstrate that the drywell shield wall and reinforcing steel are within allowables.

These are documented in ABB Impell Calculation no. 0037-00196-C003, revision 0, (Ref. 4].Al!:&SEA BRSWW sot ABB Impell Corporation CPU Nuclear Corporation ABB Impell Report No. 0037X00196-01 Oyster Crtek Nuclear Generatung Station Revision:

0 Additional Stress Information for the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool Page 6 of 23 3.0

SUMMARY

OF RESULTS Figure 3.0-1 shows a view of the drywell shield wall for which stresses are extracted, (i.e. above elevation 95 ft.). Figures 3.0-2, 3.0-3 and 3.0-4 show the vertical stress (S)Ivels in the drywell shield wall above elevation 95 feet for load combination c, viewed in the Southwest, Southeast and North directions.

Figures 3.0-5, 3.0-6 and 3.0-7 show the circumferential (hoop) stress (Sy) levels in the drywell shield wall above elevation 95 filet for load combination c, viewed in the Southwest, Southeast and North directions.

Figures 3.0-8, 3.0-9 and 3.0-10 show the vertical stress (S.) levels in the drywell shield vwall above elevation 95 feet for load combination d; viewed in the Southwest, Southeast and North directions.

Figures 3.0-11, 3.0-12 and 3.0-13 show the circumferential (hoop)stress (SY) levels in the drywell shield wall above elevation 95 feet for load combination el, viewed in the Southwest, Southeast and North directions.

These figures show that the highest stress levels in the shield wall and SFP occur in the SFP south wall at the fuel transfer opening. This location is a natural stress riser due to the stress concentrations around the slot. Figures 3.0-5 and 3.0-6, indicate a stress uiser in the cast and west shield wall a few feet below eevation 119 ft. However, the k.ss levels here are lower than those around the fuel transfer opening and this area of the wall contains similar reinforcing steel patterns as in the area around the fuel transfer opening, [Ref.9]. Therefore, the critical section is the area around the fuel transfer opening for all stresses and hence an ACI evaluation of this area is performed to demonstrate that the drywell shield satisfies the requirements of ACI 349, [Ref. 6].'Table 3.0-1 shows linearized hoop (S.) stresses around the fuel transfer opening, for locations see Figure 3.0-14. See Section 5.2.4 of [Ref. 1], for further discussion on linearized stresses.

The results shown in Table 3.0-1 indicate that circumferential tensile stresses rapidly decay away from the bottom of the fuel transfer opening. At a vertical distance of 90 inches away from the bottom of the fuel transfer opening, the tensile stress is shown to have decreased to 41 % of the maximum tensile stress at the corner of the fuel transfer opening.Table 3.0-2 shows linearized vertical (Sj stresses around the fuel transfer opening, for locations see Figure 3.0-14. The results shown in Table 3.0-2 indicate that vertical tensile stresses rapidly decay away from the bottom of the fuel transfer opening. At a horizontal distance of 79 inches away from the bottom of the fuel transfer opening, the vertical tensile stress is shown to have decreased to 22% -42% of the maximum tensile stress at the side of the fuel transfer opening.ABly ASU BfROWN Scott ABB Impell Corporation GPU Nuclear Corporation ABB Impell Report No. 0037-00196-01 Oyster Credk Nuclear Generating Station Revision:

0 Additional Stress Inforination for the Drywell Shield Wall from the Stnictural Evaluation of the Spent Fuel Pool Page 7 of 23 In evaluating the capacity of the shield wall, two types of sections are examined.

Figure 3.0-15 shows a typical section representing the bending stress acting in the circumferential direction , which corresponds to the section A-A shown in Figure 3.0-14.'Te depth of beam is taken to be the thickness of the shield wall below elevation 95 ft., and above the bottom of the spent fuel pool (i.e., 90 inches). The streses from Table 30-1 are averaged and the corresponding moment and axial forces are evaluated considering moment-axial force interaction for ACI 349, (Ref. 61. See Section 5.2.4 of Ptef. 1] for further discussion on computing moments and axial forces from averaged linearized stresses using RCBEAM.For the vertical stresses (parallel to the side of the fuel transfer opening), the re.presentative beam cross section is shown in Figure 3.0-16. For this cross section, the depth of the beam is taken to be 60 inches which corresponds to the thickness of the shield wall above elevation 95 ft. Two sections (Section A-B and section C-C as shown in Figure 3.0-14) are used to determine two sets of averaged stresses and associated moments and axial forces acting on the typical beam cross section shown in Figure 3.0-16. The resulting moments and axial forces are evaluated in the same fashion as for Seion A-A.The results of the ACT evaluation are presented in Table 3.0-3 which demonstrates that the concrete around the fuel transfer opening is within allowables and in compliance with AiCT 349, (Ref. 6]. Table 3.0-4 demonstrates the maximum tensile stresses (S) in the reinforcing steel around the fuel transfer opening, as shown in Figure 3.0-14, are within.Olowables.

ABBS ASEA BROWN M ERM ABB Impell Corporation GPU Nucleatr Corporation Oyster Creek Nuclear Generating Station Additional Stress Information for the Drywell Shield Walt from the Siructural Evaluation of the Spent Fuel Proo ABB Impcll Report No. 0037?00196-01 Revision:

0 Page 8 of 23 FI;gure 3.0-1: Model D drywell sh1ield wall afxve elevation-

95ft, GPU Nuclear Corporation Oyster Creel Nuclear Gencrating Station Additional Stress Infonnation for the Drywell Shield Wall from the Structural Evaluation of the Spent Fud Pool ABB .Impall Report No. 0037-00196-01, Revision:

0 Page 9 of 23 Figure 3.0-2: Ver-tical Stress Mgz) in the drywelI shiled-wall above elevation 95 ft.Looking Southwest

-Load combination Figure 3.0-3: Vertical Stress (Sz) in the drywell shield wall above elevation 95 ft.Lookina Southeast

-Load com-bination c

GPU Nuclear Corporation Oyster Cre*k Nuclear Generating Statioi Additional Stress Information for the Drywell Shield Wall from the Structural Evaluation or the Spent Fuel Pool ABB3 Impell Report No. 0037-00196-01 Revision:

0 Page 10 of 23 Figure 3.0-4: Vertical Stress (Sz) in the drywell shield wall above elevation 95 ft..ookine North -Load cormbination c Figure 3.0-5: Circumferential (hooD) Stress (Sv) in the drvwell shield. wall above elevation 95 ft.-tookine Southwest

-Load combination c

GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Additional Stress Infonnation for the Drywell Shield Wall rrom the Structural Evaluation or the Spent Fuel Pool ABB Impeli Report No. 0037-00196-01 Rtvision; 0 Page 11 of 23 Figure 3.0-6: Circumfeoent9al (ioop)n Strhes (SY to the drywel shield wall abova elevation 2 ft. Loolking Sguthesast

-XIgad combinatioi -c Figure 3.0-7: Circumrerential (hoop) Stress (Sy) In the drywell shiield wall above elevation 9S ft. Looking North -Load combination c

GPU Nuclear Corporation Oyster CrGek Nuclear Generating Station Additional Stress Jnformation for the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool APB Impell Report No. 0037-00196-01 Revision:

0 Page 1I of 23 Figure .i.048: il Soint ieldwall above elevation 95 ft.Looking Southwest--

Loa -obnationd Figure 3.0-9: Vertical Stress (bz) In the Orvwell shield wall above elevation Lnnkinc rtithonst

-Lnoad cnmhination d

GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Additional Stress Infomrnation for the Drywell Shield Wall from the St:ructural Evaluation of the Spent Fuel Pool ABB Impell Report No. 003700196-01 Revision:

0 Page 13 of 23 Figure 3.0-10: Vertical Stre (Si) in the dvwell Shield wall above elevatin 95 ft.Lookinz NodrtW-Lood combination d Figure 3.0-11: Circumferential (hooDL Stress (SvY) In the drvnwll shield wall above elevation 95 ft. Looking Southwest

-Load combinatIon d

GPU Nuclear Corporation Oyster Credc Nuclear Generating Station Additional Stress Infonnation for the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool ADB Impcll Report No. 0037-00196-01 Revision; 0 Page 14 of 23 Figure 3.0-12: Circumferential (hoop) Stre (Svl in the drywell shield wall above elevation 95 ft. ooking Southeast

-Load combination d Figure :1.0-13: ercumtenntial ftLhool Sitrs (Sy in the drywell snin eld wall above elevgution 9S5 ft, Looking North -Load comnbination d

GPU Nucleir Corporation Oyster Creek Nuclear Generating Station Additional Stire Information ror the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool ABB Impell Report No. 0037-00196-01 Revision:

0 Page I5 of 23 FIgure 3.0-14: Location of sections used to evaluate the cgptitv of the drvwell shield wall above elevation 95 ft.z Ly FUEL IRANSVER 39.6" 39.6" .39.6" I -,I-'--- -a------ ----I-_ -_ _.--_-_- -4211 4212 ,4213 I I Ifl E -I -I -- U -V C C Ii C'J U,2'4111 4+112 4113 4011 _ _U)_ _ __ _ _ _3 1 1 a .,3611_ I __AISh EElA BROWN BOVRI ABB Impell Corporation GPU Nuclwkr Corporation Oyster Creek Nuclear Generating Station Additional Stress lntonnoion for the Drywell Shield Wall from the Smructural Evaluation or the Spent Fuel Pool ABB Impell Report No. 0037'00196-01 Revision:

0 Page 16 of 23 F;igure 3.0-15: teprenejatin of a tynical beam cross scton correspondin?

to the averaPge tres on RC~eam Sectionl A.A I COMPRESSON EDGE (INNER SURFACE OF VEIRECAL SHIELD WALL)'Sy (COMPRESSION)

TENJSION EDGE /A (OUTER SURfACE Of SHIELD WI)// A NODE No.. D 4111 * -A--- -- CIRCUMFERENTIAL 4011 DRCTION (HOW)401 -j HOOP REINFORCEmENT

(#10)3911 _ --4 jSy (TENSION) j 3811 A--------NOTES, 1. OEPTH OF BEAM SECTION: 90s 2. THE HOOP STRESS (Sy) SHOWN ABOVE WAS OBTAINED BY AVERAGING THE HOOP STRESSES FROM THE UNEARIZED STRESSES AT 4111. 4011. 3911, 3811 SHOHN ON FIGURE 3.0-14.3. RCBEAM SECTION A-A IS IDENTIFIED ON FIGURE 3.0-14.ABJI ASIA 3140WN ABB Impell Corporation GPU Nucli:ar Corporation Oyster Creek Nuclear Gecenzing Station Additionail Stress Infornatlon for the Drywell Shield Wall from the S;rudcural Evaluation or the Spent Fuel Pool ABB Impel) Report No. 0037.00196-0O Revision; 0 Page 17 of 23 Figure 3.0-16: Representation of a typical btqm cross section coryesponding to the averaLe stress on RCBeam Section A-B and C-C VERTICA DIRECTION COMPRESSION EDGE (INNER SURFACE OF SHIELD WALL)Sz (COMPRSSION)

IVCAL REINFORCEMENT (f#0)TENSION EDGE (OUTER SURFACE SHIELD WALL)CIRCUuFEREN11AL DIREClION Sz (TENSON)NODE No.NOTES, 1. DEPTH OF BEAU SECTION: 60'Z THE VERTICAL STRESS (Sz) SHOWN ABOVE WAS OBTAINED BY AVERAGNG THE VERTICAL STRESSES FROM THE LINEARIZED STRESSES SHOW ON FIGURE 3.0-14.FOR RCBEAU SECTiON A-B, RESULTS FOR NODES 4111, 4112. 4113 WERE USED TO COMPUTE THE AVERAGE Sz FOR RMA.FOR RCSEAU SECTION C-C. RESULTS FOR NODES 4211, 4212 4213 WERE USED TO COMPUlTE THE AVERAGE Sz FOR RCBEAM.ABII ASIA BROWN "'CR ABB Impell Corporation GPU NuCigar Corporation Oyster Creek Nuclear Generating Station Additional Stress Inrormation for the Drywell Shield Wall from the Structural Evaluation or the Spent Ful Pool ABB Impell Report No. 0037-00196-01 Revision:

0 Page 18 of 23 Table 3.0.1: LinenrI2ed Hoot, Stresses S (isl) 3n the South Wall of the Srent Fuel Table 3.0-1: U~nearized H~o[ Stmse psi) in the South Wall-of the Spent Fuel Pool for Load Combinatilon c and Load Combinat-on

d. " rReL 21 RCBEAM SECTION: A-A M Nodal EDO) Load Combination c Load Combination d Outer Inner Outer Inner___ Surface Surface Swface Surface 4111 331 22 332 21 4011 157 -112 154 -122 3911 137 -110 136 -115 3811 116 -113 114 -117 hots: (I) The stress S, corresponds to the stress in the circumferential direction in the shield wall.(2) The sections employed in RCBEAM to determine the capacity of the concrete and the tension in the reinforcing steel are identified in Figures 3.0-14, 3.0-15 and 3.0-16.(3) IL series of linearized stresses were computed to determaine the average stress acting on the section for which RCBEAM was used to perform the AC! evaluation.

The node number Indicated in this column orresponds to the node on the outer surface of the south wall of the spent fuel pool model of the pair of ziodes used to determine the linearized stresses at this location.

Figures 3.0-14, 3.0-15 and 3.0-16 show the location of the nodes shown in this table.ABI ASEA MOWN BYE ABB Impel( Corporation

• GPU Nuclesr Corporation Oyster Creedc Nuclear Generating Station Additional Strews Information for the Drywdl Shield Wall from the Structural Evaluation of the Spent Fuel Pool ABB Impell Report No. 0037-00196-01 Revision; 0 Page 19 of 23 Table 3.0-2: Linearized Vertical Stre S (pli) in the South Wall ofthe Spynt Fuel Pool for Load Combination c and Load Combination d.("{Ref. 21 RCBEAM SECTION: A-B 0 Nodal Mal Load Co _ination _ Load Combination d Outer Inner Outer Inner Surface Surface Surface Surface 4111 261 123 216 96 4112 _ 111 -12 108 -14 4113 57 S .70 74 -61 RCBEAM SECTION: C-C Nodal ID"' Load Combination c L.Ad Combination d Outer Inner Outer Inner__ _ Surface Surface Surface Surface 4211 232 9 282 108 4212 147 -45 154 8 4213 9: -79 74 -58 Notes: (1) The stress St corresponds to the stress in the vertical or axial direction in the shield wall.(2) 7he sections employed in RCBEAM to determine the capacity of the concrete and the tension in the reinforcing steel are identified in Figures 3.0-14, 3.0-15 and 3.0-16.(3) A series of linearized stresses were computed lo deternine the average strezs acting on the section for which RCBEAM was used to perform the ACI evaluation.

The node number indicated in this column corresponds to the node on the outer surface of the south wall of the spent fuel pool model of the pair of rodes used to determine the linearized stresses at this location.

Figures 3.0-14, 3.0-15 and 3.0-16 show the location of the nodes shown in this table.ABIB ASIA SMOWN El Aa ABB Impell Corporation

• GPU Nucb:ar Corporation Oyster Creek Nuclear Generating Station Additional.

Strcss Inrormation for the Drywell Shield Wall from the S;tiucural Evaluation of the Spent Fuel Pool ABB Impell Report No.0037.00196-01 Revision:

0 Page 20 of 23 TABLE 3.0-3: Moment Capacity Margin In the SFPWtawgil Shield wall Analvsis Case C. Ref. 2 ITER. EQ. SECr7ON Of AXIAL MOMENT MOMENT MARGIN WALL FORCE", M. (ftrkips) cAPACrrYm X= ._ (kips) j__ M. (ft-kips)South-WaflW

__ _4S c A-A -38 173 192 10 4S_ d A-A -3S 176 206 15 4S c C-C 48 47 10S SS 4S d C-C -54 44 91 S2-- _ __ __(1) A negative axial force produces a tensile axial siress on the slab cross section. The moment capacity is based ot the same axial force.(2) Axiai Force, Moment, and Moment Capacity are for a 12' wide zection of wall.(3) The sections employed in RCBEAM to determine the capacity of the concrete section are Idenmified in Figures 3.0-14, 3.0-15 and 3.0-16.AB15 ASEA OA$H ABB Impell Corporation

• GPU Nuclear Corporation Oyster Creik Nuclear Generating Station Additional Strhe Intonnation for the Drywell Shield Wall from the Structural Evaluation of the Spent Fuel Pool ABB Impell Report No. 0037-00196-01 Revision:

0 Page 21 of 23 Tiable 3.0-4: Maximum Tensile Stresses (S ) In the Reinforcement in the South Wall of the Suent Fuel Pool for Load Combinafinn e anti TIadl Combination-d.

rRet. 41 RCBEANI Load Combination c Load Combination d Sectiionfi S, (ksi) S,!4ksl) A, (ksi) S,,,(ksi)A-A 32.8 40 31.4 40 A-B 22.7 40 24.1 40 C-C 27.5 40 3040 O0 I Notes;(I) Tic secvions used to detenrnine the averag streseu on a typical beum cross Sction are identified in Figure 3.0-14, 3.0-15 anid 3.0-16.ABJ AM "owN'R ABB Impell Corporation GPU Nucicar Corporation Oyster Cr~ak Nuclear Generating Station Additionsl Stress Information for the Drywell Shield Wall from the 5tzRctural Evaluation of the Spent Fuel Pool ABB Impell Report No. 0037-00196-01 Revision:

0 Page 22 of 23

4.0 CONCLUSION

S TMe review of the stresses in the drywell shield wall above elevation 95 ft. indicates that both the concrete and reinforcing steel are within allowables and in compliance with ACI 3419, [Ref.6J.AB11*S, A% OWN DOE ABB Impell Corporation CPU Nuclear Corporation Oyster Crmik Nuclear Generating Station Additional Stress Inortnation for the Drywedl Shield Wall from the Structural Evaluation of the Spent Fuel Pool ABB Inmpell Report No. 0037.00196-01 Revision:

0 Page 23 of 23

5.0 REFERENCES

[1] ABB Impell, "OCNGS Structural Evaluation of the Spent Fuel Poolw, Report No. 03-0370-1341, revision 0.[2] ABB Impell Calculation No. 0370-187-007, "OPUN SFP Analysis Case C Using Cracked Transformed Properties.", Revision 0.(3] ABB Impell Calculation No. 0037-00196-C002, 'GPUN SFP Analysis -Verification of Retrieved Data", Revision 0.[4] A.BB Isnpell Calculation No. 00037-00196-C003, "GPUN SEP Analysis Case C Additional Stress Information for the Drywell Shield Wall", Revision 0.[5] ANSYS, Version 4.4, Swanson Analysis Systems, Inc.[6] ACI 349 " 'Code Requirements for Nuclear Safety Related Concrete Structures," 1980.[71 (iPUN Contract No PC-0443867, 'Engineering Services for the Oyster Creek Drywell Shield Wall", August 30, 1993.(8 GPUN, "Specification for Oyster Creek Nuclear Generating Station Spent Fuel Pool Structure Qualification for Consolidated Spent Fuel Storage," ,'P-1302-53-047, Revision 1.19] Burns & Roe Drawing, 4066-3 (As-Built), Revision 3, Reactor Building Cross Section Details.ABJI ASMA fROWN IERI ABB Impell Corporation