View source for LIC-18-0028, Response to Request for Additional Information Regarding License Amendment Request, Uprate of Shutdown Cooling System Entry Conditions

{{Adams
| number = ML080850254
| issue date = 03/22/2008
| title = Response to Request for Additional Information Regarding License Amendment Request, Uprate of Shutdown Cooling System Entry Conditions
| author name = Clemens R
| author affiliation = Omaha Public Power District
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000285
| license number = 
| contact person = 
| case reference number = LIC-18-0028, TAC MD6993
| document type = Letter type:LIC
| page count = 77
| project = TAC:MD6993
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 March 22, 2008 LIC-08-0028 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==References:==

==SUBJECT:==
: 1. Docket No. 50-285 2. Letter from OPPD (D. J. Bannister) to NRC (Document Control Desk),"Fort Calhoun Station, Unit No. 1 License Amendment Request (LAR), Uprate of Shutdown Cooling (SDC) System Entry Conditions," dated October 12, 2007 (LIC-07-0054) (Accession No. ML072890192)
: 3. Letter from NRC (M. T. Markley) to OPPD (D. J. Bannister), 'Fort Calhoun Station, Unit 1 -Request for Additional Information Re: License Amendment Request, "Uprate of Shutdown Cooling System Entry Conditions," (TAC No. MD6993),'
dated February 27, 2008 (NRC-08-0025) (Accession No. ML080560007)
Response to Request for Additional Information Regarding License Amendment Request, Uprate of Shutdown Cooling System Entry Conditions (TAC No. MD6993)In Reference 2, the Omaha Public Power District (OPPD) requested changes to the Fort Calhoun Station (FCS), Unit No. 1, Renewed Operating License No. DPR-40, to modify the plant design and licensing basis to increase the shutdown cooling (SDC) system entry temperature from 300 degrees Fahrenheit (OF) to 350&deg;F (cold leg), and the SDC entry pressure from 250 pounds per square inch absolute (psia) to 300 psia (indicated at the pressurizer).
OPPD also requested changes to the Updated Safety Analysis Report (USAR) described design methodology applied to the SDC heat exchangers (HX). The Nuclear Regulatory Commission (NRC) staff reviewed the information provided in Reference 2 and determined that additional information is needed to complete their review. Reference 3, which was received on March 10, 2008, provides the NRC's request for additional information (RAI).Attachment 1 provides OPPD's responses to the NRC's RAI for FCS, Unit No. 1, Proposed LAR, "Uprate of Shutdown Cooling (SDC) System Entry Conditions" (TAC No.MD6993).40c)Employment with Equal Opportunity U. S. Nuclear Regulatory Commission LIC-08-0028 Page 2 This letter contains one regulatory commitment.
This one-time commitment is provided in Attachment 1, under OPPD's E.1. Response, as follows: "Therefore, the calculation is being revised to reflect this administrative limitation to show conformance with cooldown of the RCS within the time requirement.
This calculation revision will be completed by March 31, 2008. [AR 42008]" If you should have any questions regarding this submittal or require additional information, please contact Mr. Thomas C. Matthews at (402) 533-6938.I declare under penalty of perjury that the foregoing is true and correct. Executed on March 22, 2008 R' ard P. Clemens ivision Manager Nuclear Engineering RPC/dll 

==Attachment:==

OPPD's Responses to the NRC's RAI for FCS Unit No. 1 Proposed LAR,"Uprate of Shutdown Cooling (SDC) System Entry Conditions" (TAC No.MD6993)Appendices:
Appendix A -OPPD SDC RAI Calculation Excerpts Appendix B -Calculation FC07096, Evaluation of Temperature and Pressure Increase for [Pump Type] 6 UCL c: E. E. Collins, NRC Regional Administrator, Region IV M. T. Markley, NRC Senior Project Manager J. D. Hanna, NRC Senior Resident Inspector LIC-08-0028 Attachment 1 REQUEST FOR ADDITIONAL INFORMATION (RAI) FOR FORT CALHOUN STATION (FCS), UNIT No. 1 PROPOSED LICENSE AMENDMENT REQUEST (LAR),"UPRATE OF SHUTDOWN COOLING (SDC) SYSTEM ENTRY CONDITIONS" (TAC No. MD6993)A. SDC Pumps [Low Pressure Safety Injection (LPSI) Pumpsl Design Section 3.2.2 of the amendment request states that "[t]he SDC pumps, SI-1A and SI-1 B, also referred to as the LPSI pumps, have been analyzed for the new design pressure and temperature of 550 psig [pounds per square inch gauge] and 350&deg;F from a Code-stress standpoint.
In order to meet Code allowable stress limits, the existing pump hold-down bolting must be replaced with bolting composed of a stronger material." 1. Describe the pump analysis performed and provide a summary of the analysis results which determined that the existing pump hold-down bolting needs to be modified to meet Code-allowable stress limits. Provide current, revised and allowable values.OPPD A.1. Response: As a result of the proposed increases in pressure and temperature, revised thermal loading analysis was required.
The calculations of the pump hold-down bolting were completed using the revised nozzle loading as calculated in the piping reanalysis (calculation FC07234) and the plant design basis seismic accelerations (OPPD document DT-35639-02, "Seismic Accelerations").
The weight of the equipment was provided by the pump vendor (Flowserve).
Each LPSI pump (SI-1A & SI-1B) had unique load cases for each of the three nuclear load combinations:
Normal, Upset (OBE), and Faulted (SSE). LPSI pump SI-1B is the governing case due to the significantly higher loading observed on the nozzles.The pump hold-down bolts were originally carbon steel ASTM A307 Grade B (Oallowable 
-15000 psi). The new maximum calculated stress in the hold-down bolts is 19400 psi.These bolts are being upgraded to ASTM or ASME A193 Grade B7 with a O'allowable 
=25000 psi allowable. (Reference FC07096, page 7). The replacement at FCS will be with SA-193 Grade B7. Per OPPD general engineering instruction, PED-GEI-55, Section 5.4.4 and ASME Section X1 IWA-4224, ASTM A-type materials can be replaced with SA-type materials of the same grade, type, class, or alloy and heat condition (as applicable).
See OPPD Response to Question A.2 for discussion on code reconciliation.
: 2. Provide the Code of reference for evaluating the SDC pump anchorage.
If different than the design basis code of record, provide justification.
OPPD A.2. Response: The following code reconciliation was accepted by the Authorized Nuclear Inspector/
Authorized Nuclear Inservice Inspector (ANI/ANII) at FCS and is documented in the engineering change (EC) 35639 package.1 LIC-08-0028 Attachment 1 , Component Reconciliation:
This is a rerate in accordance with ASME XI, IWA-4330, RERATING.The original codes of record were ASME Section 111-1965 Winter 66 Addenda; ASME Section VIII-1965 Winter 66 Addenda; USAS B31.1.0-1967; ASA-B16.5-1961.
Only the design pressure-temperature rating is changed. The pressure temperature analysis for the rerate is per ASME Section VIII Div. 1, 1992 Edition and ANSI B16.5 1996 Edition. There are no changes to weight or configuration, fabrication, inspection or testing due to this engineering change. The changes to pressure-temperature rating are performed in calculation FC07096.* Material Reconciliation:
The new hold-down bolts are purchased under ASME Section 111-1989 Class I requirements.
The original code of record is USAS B31.1.0-1967 (ASTM-307).
The reconciliation between USAS B31.1.0-1967 (ASTM-307) and Section III, 1989 Edition Class 1 (ASME SA-193 Grade B7) is covered by quality procedure PED-QP-27, Attachment
: 4. The ASME SA-193 Grade B7 material is more stringent and stronger than ASTM-307.
Class 1 (1989) is a more stringent code class than Class C (USAS B31.1.0-1967).
: 3. Provide the schedule of completion for the hold-down bolting modification'of the SDC pumps.OPPD A.3. Response: The new hold-down bolts are scheduled to be installed during the 2008 refueling outage (RFO) contingent upon approval of the LAR.B. zSDC Piping Re-Analysis
: 1. Provide a quantitative summary of the SDC piping analysis results at the proposed increased design conditions (temperature and pressure) that shows conformance with the criteria of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section III, and any USAR commitments as applicable.
The evaluation should include maximum calculated stresses, fatigue usage factors and Code-allowable values. At critical locations, such as nozzles and penetrations, show that the allowable loads and movements have been satisfied.
OPPD B.1. Response: The attached Appendix A provides excerpts from calculations FC07234, FC07235, and FC01029. These calculations evaluate the following:
Calculation FC07234: Tables show the Maximum and Allowable Stress Levels for each Seismic Subsystem (except for SI-201A) that was evaluated.
2 LIC-08-0028 Attachment 1 Calculation FC07234: Penetration Loads.Calculation FCO0 029: Stress Levels and Usage Factor at Tee in Seismic Subsystem SI-201A.Calculation FC07235: Heat Exchanger Nozzle Load evaluation.
In addition, Calculation FC07096, which is provided as Appendix B, includes the Pump Nozzle Load Evaluation.
The NRC reviewer identified an error in the preliminary calculation for determination of the pressure (P) limits based on the thickness of the suction head bracket. After further review, OPPD determined that this was a typographical error in the thickness equation listed in the calculation.
The thickness equation (t) should include inputs raised to the 1/2 power not squared as shown in the OPPD calculation.
This typo was not carried through in the derived pressure (P)equation; therefore, the resultant pressure limit was not affected.
This typo has been corrected in the OPPD calculation.
: 2. Confirm whether a review of postulated pipe break criteria has been performed and provide justification that existing locations still meet the pipe-break criteria for the increased design conditions.
In addition, verify whether new postulated pipe-break locations were identified and provide justification.
OPPD B.2. Response: A pipe-break review was not performed.
The Shutdown Cooling (SDC) System is a fluid system in which the fluid operates at high energy conditions less than approximately 2% of the time. It is therefore considered a moderate energy system, and no high energy line breaks are postulated in the current design basis (USAR Appendix M and PLDBD-ME-1 1).Because of the new rerate conditions, the following conditions were considered:
* A review of the Stress Levels and Usage Factors in Calculations FC07234 and FC01029 shows that, with the exception of the tee in Seismic Subsystem SI-201A, the Stress Levels formed by a combination of the ASME III Class 2 (NC-3600)Equation 9 (Upset) and Equation 10 would remain below 80% of the combined allowable stress. This stress limit usually applies for determining intermediate break points." The tee in Seismic Subsystem SI-201A showed elevated stresses and was evaluated using a supplemental ASME Class 1 (NB-3600) analysis.
A Cumulative Usage Factor of 0.86 was determined.
* The tee in Seismic Subsystem SI-201A is located approximately ten inches from an embedded floor sleeve (Anchor) which is considered as a terminal end in any pipe break reviews performed earlier.3 LIC-08-0028 Attachment 1 The tee in Seismic Subsystem SI-201A is located outside containment beyond the second isolation valve from the reactor coolant (RC) loop and is part of the SDC system. As such, it is considered as "Moderate Energy" per Standard Review Plan (SRP) 3.6.1.3. Identify any pipe support modifications required due to the increased design loads.OPPD B.3. Response: One pipe support modification is required:
support SIH-287 is scheduled to be removed via modification EC 35639 during the 2008 RFO. (Reference Calculation FC07234)C. Review of TS Changes to Increase the SDC Entry Temperature and Pressure and the Associated Low Temperature Overpressure Protection (LTOP) Analysis 1. Sections 3.2.2 and 3.3 indicated that the heat-transfer capacity of the SDC HX is adequate for the proposed range of the SDC temperature and pressure conditions, because a calculation verified that when a component cooling water (CCW) inlet temperature to the SDC HX is less than 110&deg;F, the SDC/CCW system has the capability to cool down from the new initiation reactor coolant temperature of 350&deg;F to 130&deg;F at nominal full-power of 1500 megaWatts thermal (MWt) and normal service fouling level in the original design basis time of 24 hours. Based on the results, Section 3.3 indicated that for a loss-of-coolant accident (LOCA) during plant shutdown, the period during which automatic initiation of the emergency core coolant system is not available during the shutdown is bounded by the analysis of record (AOR).a. Please discuss the computer codes and/or methods used in the SDC HX capacity calculation, and reference the associated NRC safety evaluation (SE) that approved the codes and/or methods. Address if there have been any changes to the NRC-approved codes and/or methods used in the SDC HX capacity calculation, and justify that the changes are acceptable.
Also, discuss the operating procedures that are used to control the CCW inlet temperature to the SDC HX within 110&deg;F.OPPD C. l.a. Response: With respect to the time the emergency core cooling system (ECCS) is not available, this change improves plant operations by enabling a more rapid normal cooldown, and reducing the amount of time that the ECCS is not available prior to cold shutdown.Cooldown with SDC is faster than with the steam generator (SG) at the lower reactor coolant temperatures.
The Technical Specification (TS) requirements as implemented in Technical Data Book (TDB)-111.42 for ECCS and containment cooling equipment operation in Mode 3, transition between Modes 3 and 4 and Modes 4 and 5 are not changed.4 LIC-08-0028 Attachment 1 There were no computer codes used in the SDC HX capacity calculation FC05694, Calculation of Minimum Reactor Coolant Cooldown Time Using Shutdown Cooling.Standard thermodynamic equations, formulas, and ASME Steam Tables were used in the preparation of this calculation.
This calculation was computed and tabulated using an Excel spreadsheet.
The calculation, which is a revision to an existing calculation, was performed and qualified under Stone & Webster's Quality Assurance (QA) Program as documented in the calculation.
This calculation did not involve any new computer codes or methods requiring prior NRC approval; thus, there is no associated NRC Safety Evaluation (SE).The CCW temperatures are administratively monitored and controlled in operating procedures OI-CC-1, "Component Cooling System Normal Operation," and OI-SC-1,"Shutdown Cooling Initiation." The existing temperature limitations are discussed in operating instruction OI-CC-1, Precautions, Item No. 5: In modes I or 2 nominal CCW temperature is 55 0 F to 110&deg;F. CCW temperature must remain below 120'F. CCW temperature may fall below 55&deg;F or exceed 110 0 F during testing periods, but the following parameters must be closely monitored to ensure the sudden temperature change does not induce an undesirable transient or violate a design limit:* Letdown -Maintain -120OF* Spent Fuel Pool -Normally maintain greater than or equal to 45 0 F and less than or equal to 100&deg;F. The design operating temperature for the Spent Fuel Pool and Storage Racks is 40&deg;F to 140 0 F.* Reactor Coolant Pumps -Maintain parameters within the limits specified in 0/-RC-9, Tables 1-4. Monitor pump parameters closely to ensure the temperature change caused by testing does not adversely affect [sic] pump performance.
* Control Room Air Conditioner Waterside Economizer 
-Maintain Control Room ambient temperature less than 105&deg;F.Consistent with the limitations on CCW temperature in procedure 01-CC-1 discussed above, the 100&deg;F temperature limitation to the LPSI pumps when the reactor coolant temperature is greater than 300&deg;F is addressed administratively.
OPPD determined that the appropriate procedure for this limitation is OI-SC-I. After NRC approval of this LAR, as part of implementation, the following precaution will be added to O-SC-I: "When the LPSI pumps are in operation in SDC mode with the RCS temperatures greater than 300TF, the CCW heat exchanger outlet temperature shall be limited to 98&deg;F." Note: The 98 0 F limitation accounts for a 2&deg;F water temperature increase across the CCW pumps, therefore limits the temperature to 1 00&deg;F at the CCW loads.The section entitled "LPSI Pump Cooler Performance Evaluation (for Seal and Bearing Oil Cooling)," in Question E.3 provides the discussion of the margins with respect to maintaining the CCW water temperature below 100'F during normal conditions.
Operator procedures and actions that would be used to maintain the CCW temperatures within the limits are also discussed.
5 LIC-08-0028 Attachment 1 The 100&deg;F limitation applies only to the normal operating conditions for the LPSI/SDC seal and bearing cooling. When the reactor coolant temperature is below 300'F, the pump vendor has indicated that no cooling from the CCW system is necessary for the seals and bearings.
The peak reactor coolant (sump water temperature) post-LOCA has been calculated to be 196.6&deg;F.Instruments TIC-493, TIC-494, TIC-495, and TIC-496 (located at the outlets of CCW heat exchangers AC-1A, AC-1 B, AC-1C, and AC-1 D, respectively) will alarm if the outlet temperature from the respective CCW heat exchanger exceeds 120 0 F. As these alarm settings are for accident conditions, and not normal shutdown, no adjustment to the alarm settings is required.
The operator actions in response to the alarms are in the Alarm Response Procedure (ARP) for the indicated instruments.
The maximum allowed CCW temperature post-LOCA is 160 0 F. The maximum calculated temperature post-LOCA is 156.4&deg;F.2. Sections 3.4 and 4.1.3 indicated that the RELAP5 Mod 3.2d model was used to reanalyze mass addition and heat addition cases in support of the existing LTOP setpoint curve.a. Please list the NRC SE that approved the use of the RELAP5 Mod 3.2d model for the LTOP reanalysis and show how the restrictions or conditions in the NRC SE approving the use of the model have been met. Address if there have been any model changes including the nodal scheme in the NRC-approved code used in the reanalysis, and justify the changes. If the RELAP5 Mod 3.2 model was not previously approved by the NRC, provide a discussion of the model with the code verification applicable to the SDC conditions for the NRC staff to review and approve.OPPD C.2.a. Response: RELAP5 mod 3.2d was used in support of an October 8, 2002, license amendment request (LAR) to revise the Fort Calhoun Station (FCS) Technical Specifications (TS)related to the low temperature overpressure protection (LTOP). The NRC's approval of the use of the RELAP5 code is contained in the SE related to Amendment No. 221, Facility Operating License No. DPR-40, dated August 15, 2003 (ML032300305).
Section 3.2.7 discusses the RELAP5 model and identifies "RELAP/Mod3.2.d
[sic] as.the code of record for the LTOP analysis." There have been no changes in the nodal scheme of the model; however, there were changes made in 2005 to reflect the steam generator and pressurizer replacement conducted in 2006. The specific changes at that time did not affect the description of the model as it existed in the SE to Amendment No. 221. The changes were in head losses through the SG, SG heat transfer area, SG volumes, pressurizer volumes, and decay heat. The specific changes are shown in the following table. The net impact was minor but beneficial (greater margin to the P/T limit) due to the larger pressurizer volume.Note: the decay heat was conservatively increased to address potential unit uprating, which has not yet occurred.6 LIC-08-0028 Attachment 1 The current application does change the RELAP5 model descriptions in the SE to Amendment No. 221, in that the SE identifies the maximum SG temperature as 314'F, and in this analysis, the maximum SG temperature is 364 0 F. This does not impact mass addition events significantly, since these events are limited by the lower temperature events, and considerable margin exists at the high temperature limit. However, the additional heat energy impacts the heat addition events in a negative fashion. The remaining pressurizer bubble is smaller, but still adequate to mitigate the event without a significant pressure transient.
The justification for this change in the model is that the limit of 314&deg;F previously mentioned in the SE was a limit related to operational restrictions, not model restrictions.
Changing this value to 364&deg;F in order to justify operational changes does not impact the validity of the RELAP5 model.Table 1 -Changes Made to the FCS LTOP RELAP5 Model Due to SG and Pressurizer
('K- and Kf are reverse and forward head loss Replacement Comiponent 
~Description 0 O SG Value New RSG Value 201,303 SG Inlet Nozzle Pressure Loss Kf = 0.316 Kf 0.327 Kr = 0.156 Kr 0.121 210,310 SG Inlet Plenum Volume 124.155 ftW 134 ftW 211,311 SG Tubes Inlet Pressure Loss (heat addition/mass Kf = 1.11/0.76 Kf = 0.23/0.23 addition cases) K, = 0.55/0.38 Kr=0.43/0.477 220,320 Number of SG Tubes (heat addition assumes 0% A: 4838 5200 unplugged plugging, mass addition assumes 10% plugged) B: 4848 4680 at 10%min 4004 plugged 220,320 SG Tube Area (heat addition/mass add) 10.475/8.651 ft; 12.50/11.25 ft 220,320 SG Tube Volume A: 546.2 ftJ 662 ftW hot B: 547.3 ft 3  654 ft 3 cold (plugged min 452 ft 3  is reduced by 10%)220,320 SG Tube Length (unplugged is based on hot 52.25 ft 52.95 ft unplugged, volume for conservatism in heat addition cases; 52.31 ft plugged plugged is based on cold volume for conservatism in mass addition)220,320 SG Tube Pressure Loss 32.4 psi at 22.5 psi unplugged, 3.578e7 4.013e7 Ibm/hrSG;Ibm/hrSG, 26 plugged, currently plugging 3.91 1e7 Ibm/hrSG 221,330 SG Tubes Exit Loss (heat addition/mass add) Kf = 2.24/1.53 Kf =0.43/0.477 Kr = 1.65/1.12 Kr = 0.23/0.23 230,330 SG Outlet Plenum Volume 121.745 ft" 135 ft cold 231,330 SG Outlet Nozzle Pressure Loss Kf = 0.41 Kf= 0.176 Kr = 0.41 Kr = 0.278 410 Pressurizer Volume 900 940 ftW 410 Pressurizer Length 24.364 ft 25.364 ft 410 Pressurizer Area 36.94 ft' 37.06 ft 2 400 Surge Line Length 69.467 ft 68.614 ft 400 Surge Line Vertical Rise 16.042 ft 15.189 ft 510, 520 SG Secondary Side Volume 4549.96 ft" 4722 ftJ Heat Slab SG Heat Transfer Area 47,660 ft 2  48,980 ftW 200 Heat Slab SG tube thickness 0.0484" 0.043" 200 (OD -ID)/2 (0.75-0.6532)/2 (0.75-0.664)/2 Heat Slab Decay Heat 25.7 MW 30.3 MW 110 7 LIC-08-0028 Attachment 1 3. Section 3.7 indicated that the boron dilution event was reanalyzed based on the revised SDC conditions.
The results of the reanalysis showed that the available operator time to terminate the event was reduced by 0.55 minutes as compared to the results in the AOR.a. Please identify any model and assumptions used in the reanalysis that are different from those used in the AOR, and justify the differences.
Discuss the assumptions and the associated effects used in the analysis that result in a reduction of the operator time by 0.55 minutes.OPPD C.3.a. Response: There were no changes to the model or assumptions for the re-analysis from those used in the AOR. Two models are used to analyze this event, an instantaneous mixing model for when the reactor coolant pumps (RCPs) are operating and a dilution front model used when cooling is being provided by the SDC system. Changing the maximum temperature and pressure at which the SDC system can operate increases the range that the dilution front model is applied. The higher temperature when the SDC system is operating decreases the starting mass of the RCS and the SDC system, effectively lowering the starting amount of boron present in the core and leading to a decreased time to criticality.
D. RAI for the review of the LTOP Analysis (OPPD Calculation FC07187, Revision 0)1. The operational restrictions assumed in the LTOP reanalysis are provided in Table 2.a. Please list the corresponding TS sections that include the operational restrictions specified in Table 2 relating to reactor coolant pumps (RCPs), high-pressure safety injection pumps (HPSIs), SDC, pressurizer steam void, and reactor coolant system (RCS) pressure.
If the operational procedures were used to implement the operational restrictions, discuss the operator actions in the procedures and address the compliance with the requirements of paragraph 50.36.c(2)(ii) of Title 10 of the Code of Federal Regulations (10 CFR) that specify criteria for each item to be included in the Technical Specifications (TS).Also, the last item (RCS pressure) in Table 2 requires that when starting the first RCP, the RCS pressure be at least 61 pounds per square inch (psi) below the LTOP setpoint pressure at a given RCS temperature, in order to prevent a power-operated relief valve (PORV) lift.b. Please discuss how the value of 61 psi was determined.
OPPD D.1.a. Response: a. The Table 2 operational restrictions provided in FC07187, Rev. 0, are repeated here for reference:
8 LIC-08-0028 Attachment 1 Table 2 -Operational Restrictions ComppopnnIts, Restrictions RCPs Only 3 or fewer RCPs are allowed to be operating once LTOP is enabled; only 2 are allowed below 2240F indicated RCS temperature (based on factors used to develop the P/T curve)HPSIs Only the equivalent of 2 HPSIs and 3 CCPs can be operational once LTOP is enabled; only the equivalent of 1 HPSI and 3 CCPs are enabled below 320&deg;F indicated RCS temperature; and only the equivalent of 3 CCPs are enabled below 270&deg;F indicated RCS temperature.
Shutdown The unit cannot be put on Shutdown Cooling until the RCS has Cooling cooled to 350 F indicated RCS temperature.
Pressurizer When starting the first RCP, there must be an indicated steam void Steam Void of 50% in the Pressurizer.
RCS When starting the first RCP, the RCS pressure should be at least pressure 61 psi below the LTOP setpoint pressure at the given RCS temperature, in order to prevent a PORV lift.* The LTOP enable temperature and RCP operations shall be maintained in accordance with the pressure temperature limits report (PTLR) per TS 2.1.1(11)a.
The restrictions regarding RCPs are standard plant operation requirements that exist in Procedure OP-2A, "Plant Startup." The requirement for 2 RCPs maximum below 224&deg;F is in Attachment 1, Step 37. The requirement that no more than 3 RCPs can be running while below 500'F (LTOP is not enabled until 350'F) is in Precautions Step 19. During plant shutdown, only one RCP is running in order to decrease heat input to the RCS." The restrictions regarding HPSIs are in TS 2.3(3)." The restriction regarding SDC initiation is in TS 2.1.1(11)b.
* The restriction regarding the Pressurizer void is in TS 2.1.1(11)c." The LTOP enable temperature and RCP operations shall be maintained in accordance with the PTLR per TS 2.1.1(11)a.
The restriction regarding RCP starts with 61 psi margin to the LTOP setpoint is in Curve 3 in TDB-III.7a.
OPPD D.1.b. Response: The 61 psi restriction was originally developed in the LTOP analysis done in support of the October 8, 2002, LAR which was approved via Amendment No. 221 in NRC-03-0157 dated August 15, 2003. Specifically, Case 11 demonstrated that the RCS pressure rise following an RCP start would be 61 psi under the conservative conditions of 390 psia RCS pressure, 50'F RCS temperature, 314&deg;F SG temperature, and a decay heat appropriate to 2.18 hours after shutdown.
That is, starting an RCP under these extreme conditions could cause a 61 psi pressure spike; therefore it is advisable to start the first RCP with at least 61 psi margin to the PORV setpoint. (The decay heat enters into consideration because the analysis conservatively assumes loss of shut down cooling concurrent with an RCP start.)9 LIC-08-0028 Attachment 1 In the current application, the SG temperature may be as high as 364&deg;F so that the equivalent Case 11 results for a 50'F RCS and a 364 0 F SG temperature, predicted a 72 psi pressure rise. However, it was recognized that the Case 11 scenario assumed a conservative but unrealistic high initial RCS pressure of 390 psia. Upon review, it was determined that since the RCS pressure would be limited to 300 psia indicated (350 psia with the 50 psi error associated with the relevant indication) due to the maximum allowable SDC system pressure; it was acceptable to rerun Case 11 at an initial pressure of 350 psia. Using this assumption, the revised pressure overshoot amounted to 61 psi. It was decided to maintain the 61 psi value to avoid unnecessary TDB and procedure revisions and operational restrictions, so this revised Case 11 calculation was used to justify retention of the 61 psi value. This is discussed in Section IX of the LTOP analysis (calculation FC07187).
An additional conservatism -in Case 11 is the large temperature differential between the RCS and the SGs (by the time that the RCS cools to 50 0 F, the SGs will also cool). Also, the limitations of the SDC system prevent the RCS from reaching 50&deg;F with decay heat associated with 2.18 hours after shutdown.
Therefore, the 61 psi value is conservatively bounding.2. Page 8 indicated that the LTOP setpoints were established to limit pressure transients to below the pressure-temperature (P/T) limits as shown in Figure 1 and Table 6.a. Please list the NRC SE that approved the PIT limits. If the limits were not previously reviewed and approved by the NRC, please provide a derivation of the limits, and justify the acceptance of the P/T limits for the licensing application.
OPPD D.2.a. Response: The P/T limits were approved in the SE related to Amendment No. 221, Facility Operating License No. DPR-40, dated August 15, 2003 (ML032300305).
The SE which approved these limits also approved the methodology to be used for future revisions of the P-T curves by OPPD without prior NRC approval for implementation.
Changes in methodology would require NRC approval prior to use and implementation.
: 3. Page 9 indicated the pressure correction factor (PCF) ranging from 61 psi below 210&deg;F to 67 psi above 210&deg;F was used to account for the elevation and flow effects on the P/T limits that are based on the pressurizer pressure.a. Please provide a derivation of the PCF values of 61 psi and 67 psi based on the pressure loss and elevation difference between the reactor vessel beltline and pressurizer, and show that the PCF values are conservative and applicable to the replacement steam generators and pressurizer in determination of the P/T limits.OPPD D. 3.a. Response: Note: The value of 210'F as opposed to the 224 0 F restriction in Table 2 described in RAI 4.a. is due to an assumed 14'F instrument error in the control room temperature 10 LIC-08-0028 Attachment 1 indication.
That is, the operators are not allowed to run three RCPs below 224 0 F indicated, but the analysis assumes three RCPs may be running at a real temperature as low as 210&deg;F.The PCF values of 61 psi and 67 psi were originally developed by Combustion Engineering (CE), which has since become part of Westinghouse Electric Company (WEC). The PCF values were approved by the NRC in "Safety Evaluation of Topical Report CE NPSD-683, Rev. 6, Development of a RCS Pressure and Temperature Limits Report (PTLR) for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications (TAC No. MA9561);" and for use by OPPD in the SE for Amendment No. 221.The PCFs were verified to remain conservative for the new steam generators and pressurizer in Framatome ANP document 32-5038452-01, 8/17/04.Note that the PCFs are a combination of elevation head and flow pressure drop. They assume either two or three running RCPs (61 and 67 psi, respectively) and a conservative density. The scenario-specific pressure drop from the beltline to the pressurizer can be seen in the report's pressure graphs. With the reduced number of RCPs running (only one in the heat addition cases), the PCFs would be on the order of 35 psi. This is discussed in Section 3 of the report. However, it should be noted that the 61 and 67 psi values are described in the SE to the original CE LTOP analyses and the SE to license amendment No. 221; so it would be a change of methodology to use more realistic, less conservative scenario-specific PCFs.4. Table 7 listed the analytical and actual values for the LTOP setpoints.
For example, at 220 0 F, the analytical LTOP pressure is 690 psia.a. Please provide a derivation of the actual setpoint of 587.75 psia and the analytical value of 690 psia, and show that the PORV actuation system uncertainties of 16.3&deg;F and 66.9 psi are adequately considered in determining the LTOP actual and analytical setpoints.
Also, page 10 (last paragraph) indicated that the pressure uncertainty used for the PORV actuation pressure adjustment is 66.7 psi, which is different from the value of 66.9 psi specified in Tables 3 (page 14) and 7 (page 17).b. Please clarify the inconsistencies.
OPPD D.4.a. and D.4.b. Responses:
Note: Use of the 66.7 psi on page 10 (last paragraph) of calculation FC07187 (LIC-07-0054, Enclosure, Attachment 3), was erroneous.
The previous LTOP analyses were consulted, and the pressure error is 66.9 psi in all other uses. The only appearance of the 66.7 psi was in justifying a value that had additional margin;therefore, there was no impact of this mistake on the report results or conclusions.
11 LIC-08-0028 Attachment 1 The derivation of the actual setpoints (including consideration of the analysis curve and the 16.3 0 F and 66.9 psi measurement errors) was described in the supporting documentation for Amendment No. 221. The uncertainty is complicated because any change in temperature measurement also affects the pressure setpoint.
Therefore, the shift from the analysis curve to the dialed-in plant setpoint curve must be based on both the temperature-pressure point and the slope of the pressure vs. temperature curve.Working from the analysis curve (meaning the values assumed in the RELAP5 analysis)to the plant dialed-in setpoint curve, the curve is offset by -66.9 psi where flat because the temperature error has no impact on the pressure setpoint.
Similarly, if the setpoint curve went vertical so that pressure errors had no impact, the offset would be 16.3 0 F.For all intermediate points, the offset for temperature is 6T*sin(slope) and the offset for pressure is -8P*cos(slope) where 8T is 16.3 0 F and 5P is 66.9 psi. Here the sine of the slope between two points separated by values AP and AT is AP/SQRT(AT 2+Ap 2), and the cosine is AT/SQRT(AT 2+Ap 2). The figure below shows conceptually how this works.Figure 1 -Conceptual Representation of the LTOP Setpoint Curve Development P (psia)I TAs slope approaches vertical, offset approaches 16.3 0 F horizntal Z/At 45 degree slope, offset = 16.3/SQRT(2), and -66.9/SQRT(2), which is identical to SRSS bv Offset = full pressure error of 66.9 psi/ Pythagore----At all points: Offset= 1 6.3*sin(slope) horizontal and -66.9*cos(slope) vertical an Theorem Temperature (OF)For the specific analysis point at T=220&deg;F and P= 690 psia, the adjacent point is 221&deg;F and 696 psia. So the setpoint curve point becomes: Ts = T + 16.3*6/SQRT(1 2+6 2) = 236.1 Ps= P -66.9*1/SQRT(1 2+6 2) = 679.0 This explains the development, but does not justify it. In order to prove that the approach provides a roughly 95% confidence value, a Monte Carlo analysis was 12 LIC-08-0028 Attachment 1 performed as part of the report supporting Amendment No. 221. A figure from that report is replicated below showing just 1000 trials. The actual result of 100,000 trials was that 96.8% were below the analysis curve. The higher percentage is believed to be the result of the flat portion of the curve, where 100% of the points (offset by 66.9 psi here) are below the analysis curve.As a final note, the LTOP setpoint curve derived from the Amendment No. 221 report was further reduced by a slight amount when developing the plant values, in part because it was not convenient to input a different pressure value manually for every temperature and it was desired to envelope the analysis result in a conservative direction.
Figure 2 -LTOP Curves (Figure 27) from Amendment No. 221 Report Figure 27: LTOP Analysis, LTOP Setpoint, and LTOP Pre-Trip Curves 1800 _._ 1600 _1400 ".1200 ")1000 _ LTOP Am 0."Curve 0. 800 'LTOP Se-600 , 'N Monte Ca" 400.........
LTOP PrE U)200 a. 0 0 50 100 150 200 250 300 350 RCS Temperature (degree F)5. Page 22 (last paragraph) indicated that the decay heat used in the LTOP reanalysis is 20 percent greater than the 1.4 percent value shown in the CESEC code based on a cooldown time of 2.18 hours.a. Please discuss the decay heat model (decay heat level versus time) used in the CESEC code and justify the acceptance of the model used for the LTOP reanalysis.
OPPD D.5.a. Response: The decay heat model in the LTOP report is simple but conservative.
A constant value of 30.3 MW was used for all scenarios.
This value was generated from the decay heat 13 LIC-08-0028 Attachment 1 at the minimum time to reach 314'F (7,848 seconds based on an initial temperature of 532&deg;F and a maximum cooldown rate of 100&deg;F/hr) which was 21.4 MWt based on the CESEC code, which in turn uses the 1971 ANS 5.1 standard with 1997 modifications from ABB/CE Letter LD-97-029 and summarized in the C3M4 code certificate package, specifically accounting for U239 and Np239 (the actinides).
The value of 21.4 MW is 1.425% of the full power of 1500 MWt. This value was taken from the graph below, which is Figure 2 of the Software Release Description, Transmittal of CESEC-I11 89300 Mod 4 CA-FE-0922-CT Rev 0, 9/10/1997.
It is noted that the value of 1.425% is intentionally chosen to bound all test and model data.Figure 3 -Decay Heat Curves from the Software Release Description, Transmittal of CESEC-I11 89300 Mod 4 CA-FE-0922-CT Rev 0 (Figure 2)DECAY HEAT CURVES 0.03 C 0 0 (L 0.02 0.01 0.03 0.02 0.01*" 0.00 10000 0.00 0 2000 4000 6000 8000 Time (s)The 21.4 MW decay heat was increased by 20% to 25.7 MWt for conservatism in the first LTOP study (2002), and then conservatively increased by the ratio of 1765/1500 to 30.3 MWt in anticipation of a potential future uprate from 1500 MWt to 1765 MWt.It is noted in the LTOP report (calculation FC07187) that one scenario, assumes an RCS temperature of 364 0 F, implying that the time since shutdown less than 7,848 seconds. However, as a practical matter, it is highly unlikely cooldown would proceed with an instantaneous maximum cooldown rate until SDC initiation.
For all other cases, the RCS temperature has been cooled Case 9, could be that any reaching to lower 14 LIC-08-0028 Attachment 1 values, implying longer post-shutdown periods and lower decay heats. (Note: Case 9 is not a limiting case in terms of margin to the P/T limit.)That the 30.3 MWt net value is conservative is satisfied by noting that this is a decay heat/full power ratio of (30.3/1500) 
= 2.02% at 7,848 seconds. By comparison, sample cores in ANSI/ANS 5.1-1979, show power ratios of around 1% at this time (Tables A-1 and A-2 in the Standard).
In summary, the reasons for high confidence in decay heat conservatism are: 1. A conservative assumption about decay heat time based on the fastest cooldown to 314&deg;F RCS temperature;
: 2. Application of this cooldown time to scenarios with colder RCS temperatures which logically must have experienced longer decay times;3. An addition of 20% over the CESEC decay heat estimate; and 4. Comparison to sample decay heats for similar times in ANSI/ANS-5.1-1979.
: 6. Table 10 (page 28) listed the cases used to support the LTOP setpoints.
Cases 1 through 8 are not changed from Reference 1, Low Temperature Overpressure Protection (LTOP) Analysis in Support of Steam Generator Replacement, OPP006-REPT-001.
Page 29 indicated that Reference 1 provided a detailed description of each case.a. Discuss the analysis, test data and/or procedures that are used to assure that the PORVs will close in the steam, two-phase or liquid conditions applicable to the assumptions used in the LTOP analysis for the mass-addition events. Closure of the PORVs will avoid occurrence of a small break LOCA resulting from a stuck-open PORV.b. Reference the NRC SE that approved the use of the void in the pressurizer for consequence mitigation of the heat-addition events considered in the LTOP analysis.c. Please list the NRC SE that approved Reference
: 1. If the reference was not previously approved by the NRC, provide the reference for the NRC to review and approve.OPPD D.6.a., D.6.b., and D.6.c. Responses:
: a. Each PORV is downstream of a motor-actuated block valve (HCV-150 or HCV-151)that is provided to permit isolating the PORVs in case of PORV failure or leakage.The block valve is capable of remote closure (from the Control Room) under accident conditions to avoid a small break LOCA resulting from a PORV failure. The topic of the capability of the PORV and PORV block valve combination to mitigate pressure relief events, including LTOP events, is the subject of NUREG 0933, Issue 70, which was considered resolved with Generic Letter 90-06.15 LIC-08-0028 Attachment 1 b. Previously approved LTOP system submittals included the use of pressurizer voids (i.e., steam space) for consequence mitigation.
The NRC's approval of this approach is contained in the Safety Evaluation related to Amendment No. 221, Facility Operating License No. DPR-40, dated August 15, 2003. (Section 3.2.7 discusses the LTOP analysis.)
: c. OPPD did not receive an NRC SE on the LTOP analysis for the replacement SGs (i.e., Reference 1, "Low Temperature Overpressure Protection (LTOP) Analysis in Support of Steam Generator Replacement, OPP006-REPT-001"), as FCS remained within the bounds of the PTLR. Revision 3 of the FCS PTLR was, transmitted to the NRC in accordance with FCS TS 5.9.6.c via letter LIC-06-0138 dated November 29, 2006. The PTLR was updated to incorporate changes resulting from the installation of replacement SGs during the 2006 RFO.7. Section 3.2.1 of Enclosure to an October 12, 2007, letter indicated that the SDC suction-to-RCS valves (HCV-347 and HCV-348) interlock setpoint was changed from 250 psia to 300 psia at pressurizer.
We found that USAR-9.3 (page 6)discussed the valve interlock functions.
Specifically, it stated that... if the breaker is closed and the operator attempts to open either of these valves when pressure in the RCS is above 250 psia [300 psia for the new SDC entry pressure], an inhibit will prevent opening the valve, an alarm will sound and both valves will shut automatically...
It is not clear from the above statement whether this SDC interlock feature will automatically shut the valves (HCV-347 and HCV-348) or not if the RCS pressure increases above the interlock setpoint (300 psia) when a mass or heat-addition event occurs during SDC operating conditions.
: a. If the valves are not automatically closed when the RCS increases above the interlock setpoint, discuss the design features and procedures used to prevent the SDC from over-pressurization for a mass or heat-addition event.OPPD D.7.a. Response: The SDC system piping and components are currently protected from over-pressurization due to RCS pressure by redundant isolation valves HCV-347 and HCV-348. Each of the valves is equipped with two redundant interlocks to the pressurizer pressure.
The interlocks are tied to the 115 and 118 pressurizer pressure loops.The overpressure protection is provided by these interlocks per two scenarios:
: 1. Prevent opening of the valves (thereby preventing initiation of SDC) until the reactor pressure is below the nominal design pressure of the SDC system piping and components.
: 2. Close the valves should the RCS pressure increase above the nominal design pressure of the SDC system piping and components after SDC is in operation.
16 LIC-08-0028 Attachment 1 The interlock has no function during operating Modes 1 through 3 since the valve is closed and electrically disabled.
Additionally, this interlock fulfills a non-safety related design basis function and has no design basis accident function.The interlocks are currently set at 250 psia, which is the maximum allowable pressurizer pressure for initiation of the SDC system and is being changed to 300 psia. There are no changes to the valve closure logic or function except for the setpoint change. The impact of the instrument uncertainty on the system pressure boundary is addressed in the response to Question B.6.8. Section 1.0 of Enclosure to an October 12, 2007, letter indicated that the SDC entry pressure would increase from 250 psia to 300 psia (indicated at the pressurizer).
In support of the SDC entry pressure change, the related TS and the SDC design pressure were changed.Section 2.4 discussed the proposed TS 3.16(1)a, which stated that "the portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 300 psig..." Table 1 of Section 3.2.1 indicated that the new design pressure for the SDC pumps suction piping is 350 psig.Since the SDC entry pressure of 300 psia (indicated at the pressurizer) is based on the pressure measurement at a high elevation of the pressurizer, the corresponding pressure at the SDC system would be the pressurizer pressure plus the gravitational head and pressure loss between the pressurizer and the SDC system, and the pressure measurement uncertainty of +50 psi indicated (Table 3 of Attachment 3 to an October 12, 2007, letter). Therefore, it is likely that the maximum SDC operating would be greater than the SDC entry pressure of 300 psia.a. Please justify that the proposed leakage testing pressure of 300 psig in TS 3.16(1)a and design pressure of 350 psig for the SDC pumps suction piping are high enough and adequate to support the proposed SDC entry pressure of 300 psia with consideration of the pressure measurement uncertainty, the pressure difference due to the gravitational head and pressure loss between the pressurizer where the pressurizer pressure is measured and the SDC system where the leakage tests are performed and the SDC piping design pressure is established.
OPPD D.8.a. Response: Technical Specification Basis for Pressure The Basis for TS 3.16(1 )a, which specifies a test pressure of 250 psig (now 300 psig) is: "The limiting leakage to the atmosphere from the RHR (for FCS SDC) system (3800cc/hr) is based upon a plant specific leak rate analysis for the RHR system 17 LIC-08-0028 Attachment 1 components operating after a design basis accident." The test pressure for TS Sections 3.16(1)a and 3.16(1)b and the pressure correction factors in sections 3.16(1)c give adequate margins over the highest pressures within the lines after a design basis accident (Large Break -LOCA, USAR Section 14.15.8).
Since the SDC uprate has no impact on the design basis accident conditions (see response to question E.3.a for more details) and the 3800 cc/hr acceptance criteria is not changed, this change provides a more conservative leak test than previously specified.
Maximum System Pressure Determination The following conservative assessment has been made of maximum pressures in the system piping and components.
The pressure computed here will be present only during initiation of SDC for cooldown and termination of SDC for restart.For the new SDC initiation conditions both the piping and components upstream of the LPSI pumps (pump suction side) and piping and components downstream of the LPSI pumps (pump discharge side) are evaluated.
The peak operating pressures for the upstream and downstream piping are calculated for two operating points: at SDC initiation and during SDC system operation after the RCS has cooled significantly.
These peak pressures are calculated using the following equations, which conservatively neglect friction losses: Equation -1 PMaxUp = ARCS + (ZPZR -ZLowUp) 144 in 2j Equation -2 PM.oow.n PRcs + (ZPZR -ZLowDown + H P"KP 1--144 in2)Where: PMaxUp = peak pressure upstream of LPSI pump [psia]PRCS = peak RCS pressure at SDC initiation
[psia]ZpZR = pressurizer water elevation
[ft]ZLowUp = low point elevation upstream of LPSI pump [ft]HPump =LPSI pump head [ft]p = weight density of liquid [lbf/ft 3]PMaxDown peak pressure downstream of LPSI pump [psia]ZLowDown = low point elevation downstream of LPSI pump [ft]18 LIC-08-0028 Attachment 1 The desired maximum RCS pressure and temperature for SDC initiation are 300 psia and 350'F. Based on procedure OI-RC-8, the pressurizer level is normally controlled between 48% and 60% +/- 4%. Therefore, a pressurizer level of 64% is conservatively assumed. Based on calculation FC07183, a level of 64% corresponds to a water elevation of 1037.15 feet. Based on a review of system isometrics for the piping potentially pressurized during SDC system operation, the low point in the piping upstream of the LPSI pumps is close to valve SI-126 at elevation 972'-11" and the low point in the piping downstream of the LPSI pumps is at the LPSI pumps at elevation 973'-3".Two cases are evaluated.
First, the case where the SDC is initiated at peak initiation pressure and temperature, with the LPSI pumps operating at shutoff head is evaluated.
Based on the certified LPSI pump curve, the LPSI pump shutoff head is 450 ft. The weight density of water for a temperature of 350&deg;F and a pressure of 300 psia is 55.64 lbf/ft 3.Solving Equations 
-1 and -2 for these values, the peak pressures upstream and downstream of the LPSI pumps are: 300 psia, 350 0 F, 0 qpm Flow P M axUp = 324.8 psia (310.6 psig, for a site atmospheric pressure of 14.2 psia)PMaxDown = 498.6 psia (484.4 psig, for a site atmospheric pressure of 14.2 psia)Next, the case where the SDC system is operating at peak SDC initiation pressure but with reduced temperature following RCS cooldown is evaluated.
A temperature of 70&deg;F is conservatively assumed. The weight density of water for a temperature of 70'F and a pressure of 300 psia is 62.36 lbf/ft 3.For cooldown of the reactor, there must be flow through the LPSI pumps. This flow is regulated at 1500 gpm per pump per procedure 01-SC-1. Based on the certified LPSI pump curves, the pump head at 1500 gpm is 423 feet. Solving Equations 
-1 and -2 for these values, the peak pressures upstream and downstream of the LPSI pumps are: 300 psia, 70 OF, 1500 gpm Flow PMaxUp = 327.8 psia (313.6 psig, for a site atmospheric pressure of 14.2 psia)PMaxDown = 510.9 psia (496.7 psig, for a site atmospheric pressure of 14.2 psia)For conservatism, the rerate design conditions for upstream of the LPSI pumps are taken as 350&deg;F and 350 psig and the rerate design conditions for downstream of the pumps are taken as 350'F and 550 psig. This includes major components such as the LPSI pumps and the SDC heat exchangers.
Therefore, margins exist with respect to the maximum nominal pressures.
Impact of Uncertainty on System Component Integrity The pressure temperature design conditions are established per the codes of record USAS 31.7-1968 and USAS B31.1.0-1967.
The instrument uncertainties were not included in the current design specification and as industry practice are not included in establishing the design pressure and temperature.
19 LIC-08-0028 Attachment 1 However, the impact on piping and component pressure integrity of the uncertainty of the instruments was assessed.
The following discussion demonstrates the acceptable functioning of interlocks to prevent piping and component failure due to over-pressurization when the setpoints are chosen without consideration of instrument uncertainty.
PC-1 15A and PC-1 18A provide the pressure interlock signals to HCV-347 and HCV-348. The total uncertainties (TLU) for these loops are calculated in FC06293 and FC06299, respectively.
Since the closure of HCV-347 and HCV-348 is not associated with a design basis accident, the normal environmental condition TLUs are used for this evaluation.
They are 62.0 psi for PC-115A and 69.1 psi for PC-118A.The maximum pressures in the upstream and downstream piping and components considering the interlock uncertainty are tabulated below and compared to the piping design pressure.
The maximum pressures are taken as the pressures determined above plus 69.1 psi for the interlock uncertainty.
Table -3 Maximum Pressure vs. Design Pressure Upstream of LPSI Pumps Parameter Rerate Value Source Interlock Setpoint [psia] 300 LAR Max. Pressure w/o Uncertainty (PN) [psig] 313.6 See above Uncertainty (TLU) [psi] +69.1 FC06293 Max. Pressure w/ Uncertainty (PU) [psig] 382.7 PN + TLU Design Pressure (PD) [psig] 350 LAR 115% of Design Pressure for < 10% of time psig] 402.5 PD x 1.15 Margin exists between the conservatively calculated pressure and the design pressure before uncertainty is applied. The likelihood of both redundant instruments being at the highest uncertainty is unlikely.
Also, the rerate analysis contains the following additional margins: (1) the SDC system piping upstream of the LPSI pumps has been analyzed in calculation FC07234 for a pressure and temperature of 400 psig and 350 0 F, respectively, and (2) the piping system components are Class 300 per ASA B1 6.5-1961.The ASA B16.5-1961 pressure rating for Class 300 components is 675 psig at 350 0 F.Therefore, the piping analysis bounds the maximum system pressures including the interlock uncertainty.
Also for comparison, USAS B31.1.0-1967 Section 102.2.4 provides a 15% allowable allowance for short duration events less the 10% of operating time. The time which the SDC is in cooldown or restart is much less than 10% of the operating time.Table -4 Maximum Pressure vs. Desiqn Pressure Downstream of LPSI Pumps Parameter Rerate Value Source Interlock Setpoint [psia] 300 LAR Max. Pressure w/o Uncertainty (PN) [psig] 496.7 See above Uncertainty (TLU) [psi] +69.1 FC06293 Max. Pressure w/ Uncertainty (PU) [psig] 565.8 PN + TLU Design Pressure (PD) [psig] 550 LAR 115% of Design Pressure for < 10% of time[psig]
632.5 PD x 1.15 20 LIC-08-0028 Attachment 1 Margin exists between the conservatively calculated pressure and the design pressure before uncertainty is applied. The likelihood of both redundant instruments being at the highest uncertainty is unlikely.Valves HCV-347 and HCV-348 will be open with the indicated pressurizer pressure at or near 300 psia for only a short duration.
The condition where SDC would exceed the design pressure of 550 psig is a transient, not a sustained condition.
The code of record, USAS B31.1.0-1967 Section 102.2.4, provides a 15% allowable allowance for transient (short duration) events less the 10% of operating time. Later editions of ASME/ANSI B31.1 define the transient time as not exceeding 10% of any 24-hour operating period (i.e., 2.4 hours). The operating condition with the SDC operating pressure above 550 psig would exist for a short duration not to exceed 2.4 hours because: 1. This condition could only occur during reactor cooldown and then only until the pressurizer pressure is reduced to below 300 psia and reactor startup before the pressurizer reaches 300 psia. A typical cooldown rate is 60-70&deg;F per hour and a heatup rate of approximately 20&deg;F per hour.2. During this time period, the operators are carefully monitoring the reactor pressure and temperature so as to maintain the reactor coolant pressure and temperature with the allowable limits as defined in TDB-lII.7.d.
The repressurization of the RCS after the initiation of the SDC system would require the failure of the auxiliary pressurizer spray system to control pressures and the failure of the operators to take action. Therefore, the operator will be alerted to the failure of the auxiliary pressurizer spray system and will perform compensatory actions. Multiple indications of pressurizer pressure are available to the operators over the entire range of pressurizer pressure.
Figure TDB-III-7.d provides pressure-temperature operating limits which must be monitored and complied with. Instruments P105 or P115 digital display and T113 or T123 are required to be used when transitioning into or out of SDC entry conditions (Ref.TDB-III-7.d).
The uncertainties for these instruments are 5 0 F for temperature and 40 psi for the pressure instruments.
: 3. In addition, an alarm is provided in the control room on CB-1/2/3 to annunciate whenever HCV-347 and HCV-348 are open and pressurizer pressure is greater than 300 psia. The immediate operator actions are to verify that HCV-347 and HCV-348 are closed, and if they are not closed, then reduce pressure to less than 300 psia.Also, the rerate analysis contains the following additional pressure margins: (1) The pressure temperature rating for the pump is at least 600 psig at 350 0 F. The original hydro test pressure for the LPSI pumps was 750 psig.(2) The piping system components are Class 300 or greater per ASA B16.5-1961.
The ASA B16.5-1961 pressure rating for Class 300 components is 675 psig at 350 0 F.21 LIC-08-0028 Attachment 1 (3) The original hydro test pressure for the SDC heat exchangers was 925 psig. Based on ASME Section III, Paragraph N-712, hydro tests shall be performed to a pressure no less than 1.25 times the design pressure.
Hence, the hydro test pressure justifies a design pressure of 740 psig.(4) Per the tables in the Navco Piping Datalog, Rev. 10, June 1, 1974, the maximum working pressures for the piping in this part of the SDC system are: 1.5" OD 1822 psig; 6" OD 983 psig; 8" OD 866 psig; and 12" OD 678 psig at 400 0 F.In Summary: The maximum pressures near to the design limits are sustained only for very brief periods of time (less than 10% of the operating time (i.e., 2.4 hours)). Once the SDC system is initiated, the pressure and temperatures will be reduced rapidly. The following rationale supports this judgment: 1. The operators would not allow the overpressure condition to be sustained during normal cooldown or start-up or during a transient inadvertent repressurization during SDC operation.
: 2. The SDC system was originally designed to reduce the temperatures from 300&deg;F to 130'F in approximately 24 hours. As the temperature decreases, the pressure is also decreased.
: 3. The SDC system is considered to be a moderate energy system by the NRC (i.e., pressure > 275 psig and temperature 
> 200&deg;F for less than 2% of the operating time).4. The system pressure will not exceed the maximum working pressures for the piping or its components during the transient condition.
: 9. The proposed TS 2.1.1(11)(b) specified that the LTOP cannot be placed on SDC until the RCS has cooled to an indicated RCS temperature of less than or equal to 350&deg;F. Table 3 of Attachment stated that the RCS temperature measurement uncertainty is +14'F. With inclusion of the temperature measurement uncertainty, the LTOP may not be put in service until the actual RCS temperature is equal to 364 0 F.a. Please justify that the proposed design temperature of 350&deg;F (Table 1 of Section 3.2.1) is adequate to support the LTOP operation at 364 0 F.OPPD D.9.a. Response: The 14 0 F is the analytical margin added. The actual temperature uncertainty is 5.0&deg;F. A 5.0&deg;F uncertainty has been established for temperature instruments TI 13 and T1 23.When entering or exiting SDC, temperature instruments Ti 13 or T1 23 are required to be used per TDB Figure TDB-III.7.d.
The uncertainty of these instruments is determined in calculation FC06785 and listed in engineering instruction PED-SEI-9.
It is acceptable to apply greater analytical margin than is required to assure that the analysis is conservative.
22 LIC-08-0028 Attachment 1 The pressure-temperature design conditions are established per the codes of record USAS 31.7-1968, USAS B31.1.0-1967, and ASME Section 111-1968 Class C. The instrument uncertainties were not included in the current design specification and as industry practice are not included in establishing the design pressure and temperature.
However, the impact on piping and component pressure integrity of the uncertainty of the instruments was assessed.Please note, the critical 10 CFR 50 and ASME Xl Appendix G P-T curve is developed with the indicated temperature uncertainty factored in, such that it provides the maximum allowed pressure for the indicated RCS temperature.
The bias in this curve is such that the RCS actual temperature may be lower than 350&deg;F and thus the RPV may be slightly more brittle than at an actual temperature of 350'F. The LTOP analysis bias in regards to the indicated 350&deg;F is that the RCS is assumed to be higher than 350'F, so as to leave additional thermal energy in the steam generators when transitioning to SDC. Thus, in a conservative manner, the LTOP system analysis effectively double counts uncertainty.
E. In the LAR, the licensee states in order to ensure adequate pump seal and bearing cooling, the CCW inlet temperature at the seal cooler must not exceed the design value of 100&deg;F when the RCS temperature is between 300&deg;F and 350&deg;F. The evaluation refers to the CCW inlet temperature to the SDC HX of 110&deg;F would result in an approximately 26 percent increase temperature difference and an increased decay heat load. The evaluation refers to a calculation that verifies the SDC/CCW system has the capability to cool down the RCS from 350 OF to 130 OF at full nominal power of 1500 MWt in the original design basis time of 24 hours.Assuming the CCW supplied to the pump seal and bearing cooling is the same CCW being supplied to the SDC HX, then CCW cannot be supplied to the SDC at 110 &deg;F when the RCS is between 300 OF and 350 OF (i.e., limited to 100 degrees).The application refers to procedure controls to ensure the limit is not exceeded.1. Considering the design basis maximum for river water temperature is 90 OF (with a limit of the outlet of the CCW HX to less than 100 OF, not 110 OF) and future power uprate to 1765 MWt, what would be the effect on the SDC HX's ability at these limits and constraints to remove the required design heat load at 350 OF and achieve a cooldown of the RCS in the time required?OPPD E.1. Response: The purpose of revising Calculation FC05694 "Calculation of Minimum Reactor Coolant Time Using Shutdown Cooling System," was intended to be informative; cooldown of the RCS by the SDC system is not a safety related function.USAR Section 14.15 states that the reactor coolant is reduced from 300&deg;F to 140&deg;F in about 24 hours with the event being Loss-of-Coolant-Accident during shutdown.
The 24-hour time frame is not a limiting DBA analytical limit. The 24-hour limit is a sizing criteria established by CE. [Contract No. 750, CEND23866].
The TS limits to be in cold shutdown are: 6 hours to hot shutdown and cold shutdown within the next 30 hours [TS 2.0.1 (1) & (2)]. There are no safety limits with respect to cool down for DBA [full LOCA 23 LIC-08-0028 Attachment 1 at power]. By engineering judgment and the results of FC05694, the licensing limit of 36 hours to cold shutdown (210'F) can easily be met.A review of calculation FC05694 revealed that the administrative limit placed on CCW temperature between an RCS temperature of 350&deg;F and 300&deg;F was not considered in Revision 2. (Reference Condition Report 2008-1534.)
Therefore, the calculation is being revised to reflect this administrative limitation to show conformance with cooldown of the RCS within the time requirement.
This calculation revision will be completed by March 31, 2008. [AR 42008] This calculation revision will reflect the licensed full power rating of 1500 MWt. Future extended power uprate to 1765 MWt is not being addressed under this submittal.
: 2. The current design temperature limit for the SDC pump discharge piping and SDC HX is 350 OF. The design temperature limit for the SDC pump suction pumping and the SDC pump seals is only 300 OF. The licensee proposes to increase the system operating temperature to 350 OF, which would equal the design temperature for SDC LPSI pumps, suction piping, discharge piping, and the SDC HX, and would exceed the design temperature for the SDC pump seals. This increase would result in no safety margin between the design and operating parameters.
Additionally, in the LTOP analysis, the licensee states there is a 14 OF uncertainty in RCS temperature, which could result in the operating SDC with the RCS higher than operating/design limits. General design criteria under 10 CFR 50, Appendix A, requires the RCS and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the RCP boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
: a. The licensee is requested to explain how operations with no margin and an expected uncertainty that would exceed design limits would be acceptable.
OPPD E.2.a. Response: The pump seals are addressed in Question E.3 below.With respect to the design margins for the pressure boundary components (LPSI pumps, suction piping, discharge piping and the SDC heat exchangers), the component stresses have been demonstrated to be within the design basis limits as defined by appropriate codes. The SDC piping is considered Class 2 piping under the original code of record USAS 31.7-1967.
The SDC components were considered Class C under the original code of record ASME Section 111-1968.
The system is classified Class 2 under ASME Section X1. The uprate meets the requirements identified in these Codes. The use of later design codes is reconciled per ASME Section XI.The 14'F is the analytical margin added. The actual temperature uncertainty is 5&deg;F. A 5F uncertainty has been established for temperature instruments T113 and T123.When entering or exiting shutdown cooling, temperature instruments T113 or T123 are required to be used per TDB Figure TDB-III.7.d.
The uncertainty of these instruments is 24 LIC-08-0028 Attachment 1 determined in calculation FC06785 and listed in engineering instruction PED-SEI-9.
It is acceptable to apply greater analytical margin than is required to assure that the analysis is conservative.
The pressure-temperature design conditions are established per the codes of record USAS 31.7-1968, USAS B31.1.0-1967, and ASME Section 111-1968 Class C. The instrument uncertainties were not included in the current design specification and as industry practice are not included in establishing the design pressure and temperature.
However, the impact on piping and component pressure integrity of the uncertainty of the instruments was assessed.The 5&deg;F change in temperature will have an insignificant change in the stress allowable.
Per USAS B31.1.0-1967, for Type 304 Stainless Steel the allowable at 300'F is 15550 psi and at 400&deg;F is 14950 psi. A 5&deg;F change in temperature results in an allowable change of approximately 30 psi or 0.2%.3. The current design temperature limit for the LPSI pumps seals is 300&deg;F (reference USAR, page 13 of 31 in Section 6.2). The LAR implies that the LPSI pumps are currently rated for 350&deg;F (Table 1 on page 5 and discussion for LPSI pumps on page 7).a. The licensee is requested to explain the difference in ratings when the seals are usually encompassed with the pump ratings. Additionally, the licensee is requested to provide an evaluation of whether the pump seals can safely operate at the new proposed operating limits with an adequate safety margin, considering the higher temperature and pressure, including uncertainties.
OPPD E.3.a. Response: The following evaluations were performed to address the impact of the increase in the process fluid temperature on the pump performance and integrity including the seals. In every case, except the hold-down bolts, the LPSI pump parts are within original design without modification.
Shutdown Cooling Mode of Operation:
LPSI Pump and Cooler Pressure and Temperature Evaluation The LPSI pumps and coolers have been evaluated for the conditions of the 600 psig and 350&deg;F by Flowserve in evaluation TR-2007-09 (FC07096).
The new conditions were found to be within the maximum working pressures and temperature for the pump parts (volute, suction/discharge flanges, main casing bolting and suction bracket).As documented in the evaluation, the rerating of the LPSI pumps to 600 psig and 350&deg;F requires the replacement of the current pump hold-down bolting with bolting composed of a stronger material.
This was the result of increased nozzle loads as calculated in the piping analysis.
The stresses within the nozzles themselves were within allowable values. These bolts are scheduled to be replaced during the 2008 RFO via modification EC 35639.25 LIC-08-0028 Attachment 1 The design pressure and temperature of the LPSI pump coolers of 5140 psig and 800'F bound the rerate conditions of 600 psig and 350 0 F. Note: While the system rerate is to 550 psig and 350 0 F, the LPSI pumps were conservatively evaluated to 600 psig and 350 0 F.LPSI Pump NPSH Evaluation As a result of the SDC system rerate, the static pressure and vapor pressure at the LPSI pump inlets during SDC system operation will change. This will impact the available net positive suction head (NPSHA) of the pumps during initial SDC system operation.
The design flow rate of the SDC system is 3000 gpm, based on 1500 gpm per each LPSI pump. The current LPSI pump NPSH calculation for the SDC operating mode is performed at the design flow rate of 3000 gpm. The required net positive suction head (NPSHR) for the LPSI pumps at a flow rate of 1500 gpm is 14.5 feet.The table below compares the parameters used in the original design NPSHA assessment with the rerate initiation parameters, and summarizes the resulting NPSHA for each case.Elevation Head (ft)36.25 12 33.25 (5)Vapor Pressure (ft) 6.9 (3) 350 (6)Friction Loss (ft) 11.33 7 11.33 "1 NPSHA (ft) 52 365 (= S.H. + E.H. -V.P. -F.L.)NPSHR (ft) 14.5 14.5 NPSHA > NPSHR Yes Yes Notes: (1) Static Head = atmospheric pressure = 0 psig (2) Elevation Head = 1009' -972.75' = 36.25 ft (3) Vapor pressure based on water at a temperature of 140&deg;F (4) Static Head = (300 psia) x ((144 in2/ft 2)/(62.4 lbf/ft 3)) = 693 ft (5) Elevation Head = 1006.5' (01-SC-1, Precaution
: 20) -973.25' (Ref. File 35743 and File 35745) = 33.25 (6) Vapor Pressure = 135 psia (Ref. Crane 410 based on 350 0 F) x ((144 in 2/ft 2)/ (55.6 lbf/ft 3)) (Ref. Crane No. 410 based on 350'F) = 350 ft (7) Friction loss based on flow rate of 3000 gpm (8) Conversion to head conservatively based on the use of a density @ 700F.As shown by the comparisons in the table above, significant margin exists between the NPSHA and NPSHR for both the original design condition and the SDC initiation condition considering the rerate. Even considering the NPSHR at pump runout of 25 feet at 2800 gpm per pump, the NPSHA exceeds the NPSHR by a signification margin for the rerate SDC initiation conditions.
26 LIC-08-0028 Attachment 1 Additionally, to prevent the pump runout (and insufficient NPSHA), the full travel of FCV-326 was reduced from 4 inches to 3 inches by-the installation of a diaphragm stop extension in the actuator housing. For normal SDC system operation, FCV-326 is typically less than 50% open with a SDC system flow adjusted to 1500 gpm. As the SDC flow is not changed by the rerate, and significant valve travel margin exists at the normal SDC system flow rate, the function of the stop extension is not impacted by the rerate conditions.
LPSI Pump Thermal Transient Evaluation Rapid temperature cycling in pumps can cause the casing ring to grow faster than the pump casing, resulting in the loosening of the casing rings. The LPSI pumps were originally designed to withstand a thermal transient of 40&deg;F to 300&deg;F in 5 to 10 seconds.The pumps have been evaluated for a thermal transient of 40&deg;F to 350&deg;F in 5 to 10 seconds in calculation FC07096. The results of the evaluation show that the expected stresses due to the transient are bounded by the material yield strength with significant margin.LPSI Pump Cooler Performance Evaluation (For Seal and Bearing Oil Cooling)The LPSI pump coolers are part of the LPSI pump cooling loop that receives cooling water from the CCW system. After passing through the LPSI pump coolers, the CCW flows through the CCW pump bearing housings and stuffing box jacket prior to being returned to the CCW system. The CCW is provided to the LPSI pump coolers at a design temperature and flow rate of 100'F and 15 gpm.Per the pump manufacturer, the LPSI pump seals are rated for a temperature and pressure of 400&deg;F and 600 psig. This however applies to conditions in the seal chamber and not to the process fluid. When not cooled the seal chamber temperature is normally higher than the process fluid temperature due to heat soak through the pump, friction between the seal faces, and the friction caused by shearing of the liquid pumped.Flowserve was contacted to evaluate the impacts of the rerate on the required CCW flow rate and inlet temperature to the LPSI pump coolers to ensure adequate seal cooling and bearing cooling. The Flowserve evaluation concluded that for process (reactor coolant) temperatures at the pump suction of 300'F to 350 0 F, that current.design cooling water conditions for the pump cooler inlet of 100&deg;F at 15 gpm must be maintained to ensure proper seal cooling and bearing cooling. As discussed in USAR Section 6.2, no CCW is necessary for the LPSI pumps to be considered operable with the suction water temperature 300&deg;F or less.The following summarizes the Flowserve conclusions:
: 1. The normal upper limit for cooling water to the pump bearings is 120'F.2. Bearing cooling water is required for pumping applications above 3000F.3. To ensure adequate bearing cooling, the CCW inlet temperature at the seal cooler must not exceed the design value of 100'F when the reactor coolant temperature is between 300&deg;F and 350&deg;F and the total CCW flow rate is 15 gpm. For the cooler, as 27 LIC-08-0028 Attachment 1 it is arranged on the pump, the inlet CCW water must be 1 00&deg;F for the bearing water to be 120&deg;F. There is a 20'F increase across the cooler.The margins for the seal and bearing water cooling are discussed below. Operator procedures and actions that would be used to maintain the CCW temperatures within the limits are also discussed.
The seal vendor specifies that 12.75 gpm are required for the seal with the process water temperature greater than 300 0 F. The 15 gpm specified above provides margin.The CCW temperatures are administratively monitored and controlled in operating procedures 01-CC-1 and 01-SC-I. The existing temperature limitations are discussed in operating instruction 01-CC-1, Precautions, paragraph number 5: In modes 1 or 2 nominal CCW temperature is 55 0 F to 110 &deg;F. CCW temperature must remain below 120 0 F. CCW temperature may fall below 55 0 F or exceed 110&deg;F during testing periods, but the following parameters must be closely monitored to ensure the sudden temperature change does not induce an undesirable transient or violate a design limit:* Letdown -Maintain -12 0&deg;F" Spent Fuel Pool -Normally maintain greater than or equal to 45&deg;F and less than or equal to 100 0 F. The design operating temperature for the Spent Fuel Pool and Storage Racks is 40&deg;F to 140 0 F." Reactor Coolant Pumps -Maintain parameter within the limits specified in operating instruction OI-RC-9, Tables 1-4. Monitor pump parameters closely to ensure the temperature change caused by testing does not adversely affect [sic] pump performance.
* Control Room Air Conditioner Waterside Economizer 
-Maintain Control Room ambient temperature less than 105&deg;F.Consistent with the limitations on CCW temperature in procedure 01-CC-1 discussed above, the 100&deg;F temperature limitation to the LPSI pumps when the reactor coolant temperature is greater than 300'F is addressed administratively.
OPPD determined that the appropriate procedure for this limitation is O-SC-I. Pending NRC approval of the LAR, as part of implementation of the amendment, the following precaution should be added to O-SC-I: "When the LPSI pumps are in operation in SDC mode with the RCS temperatures greater than 300 0 F, the CCW heat exchanger outlet temperature shall be limited to 98&deg;F." Note: The 98 0 F limitation accounts for a 2 0 F water temperature increase across the CCW pumps, therefore limits the temperature to 100&deg;F at the CCW loads.The 100&deg;F limitation applies only to the normal operating conditions for the LPSI/SDC seal and bearing cooling. When the reactor coolant temperature is below 300 0 F, the pump vendor has indicated that no cooling from the CCW system is necessary for the 28 LIC-08-0028 Attachment 1 seals and bearings.
The peak reactor coolant (sump water) temperature post-LOCA has been calculated to be 196.6 0 F.The margins that exist in CCW system capacity and operational flexibility are discussed next.Margins exist in the CCW system with respect to capacity and operational flexibility.
Based on historical operating data, CCW temperature is maintained between -70'F and-90&deg;F during normal operation.
Operating data from June 8, 2005 to September 6, 2005 indicates a maximum temperature of 87&deg;F. Operating data from June 8, 2006 to September 6, 2006 shows a maximum temperature of 85 0 F. This data shows that margin typically exists between the actual CCW heat exchanger outlet temperature and the new 100&deg;F limitation for SDC system operation with RCS temperatures above 3000F.Significant margins exist in the CCW system. The design SDC mode heat load on the CCW system (61.1 MBtu/hr) is greater than the design normal operation mode heat load on the CCW system (23.75 MBtu/hr).
However, it is significantly lower than the post Recirculation Actuation Signal (RAS) design load of 117.8 MBtu/hr and the pre-RAS design load of 352.2 MBtu/hr. All non-essential loads on the CCW system can be removed during the initial shutdown period when the SDC system heat load is greatest.Furthermore, CCW flow through the SDC heat exchangers can be throttled to limit the heat load of the SDC system on the CCW system if the CCW heat exchanger outlet temperature limit is challenged.
The actual CCW return temperature can be regulated by controlling the SDC cooldown rate, using operator actions. A discussion of these actions, along with any impact as a result of the increased in SDC initiation conditions is described below (Ref. procedures 01-SC-1 and OP-ST-RC-0008):
: 1. SDC flow is initiated after warming up and sampling the system by opening the SDC suction valves HCV-347 and HCV-348 on Loop 2 and shutting the LPSI pump suction valve from the Safety Injection and Refueling Water Tank (SIRWT). Two LPSI valves are opened and FCV-326 is throttled to maintain 1500 gpm flow through the SDC system. A second LPSI pump may be started to meet cooling requirements or change system alignment.
During initial cooldown, the temperature difference for heat transfer is large, thus only a small portion of the total SDC flow is diverted through the SDC heat exchangers.
As cooldown proceeds, the temperature difference becomes smaller, and thus the flow rate through the heat exchangers must be increased to maintain the desired cooldown rate. As a result of the increase in SDC initiation temperature, the SDC flow through the SDC heat exchangers may be reduced even further during the initial cooldown.2. SDC flow through the SDC heat exchangers and SDC heat exchangers bypass line, along with CCW flow, are periodically adjusted to maintain the desired cooldown rate until the RCS temperature is reduced to the desired temperature.
System cooling flow is then adjusted to remove decay heat while maintaining the desired temperature.
The SDC heat exchanger bypass flow is controlled by valve FCV-326.The CCW flow rate through the SDC heat exchangers is controlled using CCW line control valves HCV-484 and HCV-485. While the SDC heat exchanger bypass and 29 LIC-08-0028 Attachment 1 CCW flow rates during the initial cooldown may change as a result of the increase in SDC initiation temperature, the overall flow balancing process to obtain the desired temperatures will not be impacted.3. With the SDC system in operation, two CCW pumps and three CCW heat exchangers are in operation.
CCW is supplied to the SDC heat exchangers and to all or some of the normal operating components.
The CCW flow rate and temperature will be verified in the post modification testing as follows: The CCW water temperature at both the inlet and outlet of the seal water cooler SI-I-IA & B shall be recorded starting with the reactor coolant temperature as near as possible to the initiation of 350'F and then recorded until the reactor coolant temperature is reduced to less than 300 0 F. The results shall be provided to design engineering for evaluation.
The CCW inlet temperature to the seal water cooler shall not exceed 100'F. The outlet temperature of the seal water cooler shall not exceed 120'F. Should these temperatures be exceeded, additional CCW flow capacity shall be provided to lower them to within the accepted criteria.The CCW flow rate shall be measured.
The flow rate of at least 15 gpm is expected.
This test must be run at the same time as the temperature test described above. Record the results and send them to design engineering for future evaluation.
If the 15 gpm criterion is not met, the CCW configuration shall be revised to increase the flow as much as possible.ECCS Mode of Operation:
The accident process parameters are based on the accident conditions which conservatively assume the reactor at power. The full reactor power conditions are not impacted.
The LPSI pumps are auto initiated via a Safety Injection (SI) initiation signal.The pump design is conservatively evaluated in the USAR based on the bounding accident parameters.
Therefore, this rerate of the shutdown cooling system to initiate reactor cooling at an increased pressure and temperature during normal shutdown has no impact on the LPSI pump performance parameters or integrity for the safety related ECCS response as discussed in USAR Sections 6-2, Safety Injection System, and 14-15, Loss-of-Coolant Accident.APPENDICES:
Appendix A -OPPD SDC RAI Calculation Excerpts Appendix B -Calculation FC07096, Evaluation of Temperature and Pressure Increase for[Pump Type] 6 UCL 30 LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. IA SHAW GROUP COMPANY CALCULATION TITLE PAGE CLIENT & PROJECT: OMAHA PUBLIC POWER DISTRICT PAGE 1 of 37 Ft. Calhoun Station -Unit .1 TOTAL PAGES: 100 CALCULATION TITLE (Indicative of Objective):
QA CATEGORY i- I Nuclear Safety Related Evaluation of Shutdown Cooling Mode Temperature and Pressure (CQE)Increase on SI System Piping and Pipe Supports El -II El-Il[] -Non-Safety Related LE-CALCULATION IDENTIFICATION NUMBER _ _ -Fossil/Industrial Plant J.O. orW.O. NO. DIVISION & GROUP CURRENT CALC NO. OPTIONAL OPTIONAL TASK WORK PACKAGE NO.CODE N/A N/A FC07234 N/A N/A APPROVALS 
-SIGNATURE 
& DATE , d 7 CONFIRMATION
__________REQUIRED INDEPENDENT REV. NO. SUPERSEDES PREPARER(S)IDATE(S)
REVIEWER(S)/DATE(S)
REVIEWER(S)/
OR NEW CALC. NO. OR REV. YES NO ,_ DATES(S) CALC. NO. NO.See Ramasastry Sal a&#xfd; John K. Manning John K Manning 0 N/APage EEl ElE DISTRIBUTION GROUP NAME & LOCATION COPY GROUP T NAME & LOCATION COPY SENT SENT OPPD Fort Calhoun Station El_ (original)
S&W Darlene McDonald I [Document Control Denver (electronic copy) I LI ____f______
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. / A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICA-TFION-NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A NIA FC07234 N/A PAGE NO. 5 of 37 1. BACKGROUND 
& OBJECTIVE Background EC-35639 (Reference
: 27) changes the Shutdown Cooling (SDC) Entry Conditions from 250 psia / 300 0 F, to 300 psig / 350 0 F.This calculation evaluates the increase in temperature from 300&deg;F to 350&deg;F on all piping and pipe supports in the SDC flow path of the Safety Injection (SI) System. It also evaluates the increase in the maximum operating pressure during SDC Mode from 250 psia to 400 psig upstream of the SI pumps and from 500 psi to 550 psig downstream of the pumps.Obiective The objective of this calculation is to evaluate the piping and supports included in the analytical boundaries of the Seismic Subsystems described in Section 2 of this calculation.
Specifically, this calculation will:* Demonstrate that the piping/support configuration is adequate and meets the OPPD Analysis and Modeling Criteria (References 4, 5, 6 & 8), the ASME Section III Code requirements (Reference
: 2) and Ft. Calhoun USAR (Reference
: 1) commitments.
* Determine and evaluate support loads in accordance with criteria from References 3, 7 & 8. Initiate any required pipe support modifications necessary to bring components within Design Basis limits." Determine and evaluate valve accelerations for cases where the ASCM ARS are substituted for the original ARS." Determine nozzle loads on Pump SI-1A, Pump SI-1B, Heat Exchanger AC-4A and Heat Exchanger AC-4B for evaluation by others." Determine movements at Wall Penetration Bellows shown on Piping Fabrication Isometric IC-189, which is included in Seismic Subsystem S1-191A, for evaluation by others.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A N/A FC07234 N/A PAGE NO. 22 of 37 4. CONCLUSIONS The following specific objectives of this calculation have been satisfied:
* After implementing EC-35639 (Reference 27), the piping /support configuration is adequate and meets the OPPD Analysis and Modeling Criteria (References 4, 5 & 6), the ASME Section III Code re ' waets (Reference_
1.85. --A_- -S 'dun all pief sen detemoinged an are acceptable.
Supports where the load has exceeded the previous design or capacity loads by 10% should be revised to incorporate the evaluations performed in the attachments to this calculation.
Support SIH-287 shall be removed.* Valve accelerations for cases where the ASCM ARS have been substituted for th ori inal ARS have been determined and evaluated as acceptable.
T-he nozzlean -4A--a'n-d Heat xchI ovmet a heWalPeeraio elos hononPpig ariaio Isomtri /C19 hc sicue nSimcSbytmS-9A aebe fowre fo vlain 
..
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET.CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUPI CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A N/A FC07234 N/A PAGE NO. 2 7 of 37 7. RESULTS 

==SUMMARY==
Stress Summlna Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.MODEL SI-082C From SUPERPIPE file: si082c350.s pie.out, dated 07/10/07, 13:42:55 7I 1wa 'e- AIi~ I Cmponp~ent Stes ,-P s $e (psi) ss Type Equation 8 16700 5266 0.315 A31 AWBW, i = 2.0 Equation 9 (Upset) 20049 5519 0.275 A31 AWBW, i =2.0 Equation 9 (Faulted) 23109 5698 0.247 37A AWBW, i =2.0 Equation .10 27675 18033 0.652 37A AWBW, i =2.0 Equation 11 N/A N/A N/A N/A N/A MODEL S.-201A j -2 From SUPERPI E file: si20 2rs.spi .ou dated 06/2 07,14:21:42 1 ' V -' 'nin.AllbIabl L&#xfd;M x 14m qvr~ :;Node' Compongnt E ation 8 1E 600 45 /0.356 -16 STRP/,i1.0 quationP9 pset) / 6WC)80 122018>' 0.620 20 Fl i=2'.0 Equat' 9 (Faulted) 23109 3 [ 'l J 20 LWI 0 alqition 10 /A 2760d 354
* I 21/, BTEE, ' =,2.Equation 11 i 50449 / 1.147 BTE , i =2.057//(1/* Reuii Deaile AnI'Ws of Te (ANSY'S)./Detailed of TeeFrom SUPERPIPE file: si201b0rs.spipe.out, dated 06/21/07, 08:41:12 Equation 8 15900 4030 0.253 48 STRP, i =1.0 Equation 9 (Upset) 19080 6003 0.315 102 AWBW, i =2.384 Equation 9 (Faulted) 23109 7040 0.305 102 AWBW, i =2.384 Equation 10 27475 12571 0.458 62 AWBW, i =2.0 Equation 11 N/A N/A N/A N/A N/A LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. IA SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUPI CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: NIA N/A FC07234 N/A PAGE NO. 2 8 of 37 Stress Summary (cont'd).Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.MODEL SI-203A From SUPERPIPE file: si203 t.spipe.out, dated 03/22/07, 08:14:57 g;r~ (P O Q ." .7 ' '7&#xfd;,"a T~ype.Equation 8 16400 5453 0.333 219 BELB, i = 3.702 Equation 9 (Upset) N/A N/A N/A N/A N/A Equation 9 (Faulted)
N/A N/A N/A N/A N/A Equation 10 27600 19604 0.710 221 BTEE, i =2.906 Equation 11 N/A N/A N/A N/A N/A MODEL SI-205A From SUPERPIPE file: si205a350rs.spI e.out, dated 06/27/07, 12:13:24 ifo -,7 a f Maxiiium..,S~e~.
Yde -o Cmpne~4~& ~ K s~re~s(psi (psi) Name~St~rRais aio Tpe, Equation 8 18600 7366 0.396 375 AWBW, I =3.580 Equation 9 (Upset) 22320 10763 0.482 375 AWBW, i =3.580 Equation 9 (Faulted) 23109 10858 0.470 375 AWBW, i =3.580 Equation 10 28015 28218 1.007 20 AWBW, i =4.050 Equation 11 46075 31798 0.690 20 AWBW, i =4.050 MODEL SI-191A From SUPERPIPE file: si191a350rs.spipe ut, dated 06/22/07, 13:09:28 lo hIoab~e.. 
--Maximumn Stres&#xfd; J'1Nt!6&#xfd; ~;: qomponeht Strss (Psi),' Srs....aio Nmc "K y~Equation 8 16400 6100 0.372 Al15 AWTT, i =1.90 Equation 9 (Upset) 19680 8041 0.409 127 AWBW, i =2.0 Equation 9 (Faulted) 23109 9023 0.390 127 AWBW, i =2.0 Equation 10 27600 19751 0.716 127 AWBW, i =2.0 Equation 11 N/A N/A N/A N/A N/A LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER .J.O. OR W.O NUMBER DIVISION AND GROUPI CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A NIA FC07234 N/A PAGE No. 2 9 of 37 Stress Summary (cont'd)Maximum Pipe Stress Levels and Allowable Stresses -ASME IIl Class 2 and 3 piping.MODEL SI-187A From SUPERPIPE file: sil87ar2x350rs.spi e.out, dated 06/22/07, 13:08:15 Condition Alowabl6:!.:&#xfd; iti trs -den Equation 8 16400 8154 0.497 440 BELB, i= 3.751 Equation 9 (Upset) 19680 10678 0.543 370 BELB, i = 3.751 Equation 9 (Faulted) 23109 11895 0.515 370 BELB, i = 3.751 Equation 10 27600 10489 0.380 370 BELB, i =3.751 Equation 11 N/A N/A N/A N/A N/A MODEL SI-192A, SI-195A, SI-341A (combined)
From SUPERPIPE file: si192asm350.=
dated 03/28/07, 15:22:05 Condti~ri>
Al6~i~~ ~ ~ 4d~. Cmponent Equation 8 16400 9171 0.559 115 AWBW, i =2.100 Equation 9 (Upset) 19680 13439 0.683 115 AWBW, i =2.100 Equation 9 (Faulted) 23109 14545 0.629 A501 AWBW, i =2.100 Equation 10 27600 28793 1.043 97 BTEE, i =2.172 Equation 11 44000 33787 0.768 97 BTEE, i =2.172 MODEL SI-080C From SUPERPIPE file: si08Oc6scd350.spipe.out, dated 03/29/07, 11:53:29 Colfff5Vn- (pi)F'SrsstreSS -Component Equation 8 18800 9930 0.528 D173 FILW, i =2.1 Equation 9 (Upset) 21900 18639 0.851 D35 BRED, i =2.0 Equation 9 (Faulted) 23109 19293 0.835 D35 BRED, i =2.0 Equation 10 28200 39085 1.386 W83 AWBW, i =1.08 Equation 11 47000 46072 0.980 W83 AWBW, i =1.08 LIC-08-0028 Appendix A Soc Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. /IA SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A N/A FC07234 NIA PAGE NO. 30 of 37 Stress Summary (cont'd)Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.MODEL SI-079C From SUPERPIPE file: Si079c3z350.pire.out.
dated 03/30/07.
12:49:00 Condftion 
",:IAllowable&#xfd; 
<Mqximumt Stress' Node, j~omppnent~~4;'
K St~ress, (pD ~~ fe(i>~~ a~~Typpe-~ ~~ ~~ 4'. tib~I ..Equation 8 18800 7395 0.393 C15A FILW i = 2.1 Equation 9 (Upset) 22560 11399 0.505 B35C FILW i = 2.1 Equation 9 (Faulted) 23109 10984 0.475 C15A FILW i = 2.1 Equation 10 27600 37800 1.370 11 BTEE, i = 2.172 Equation 11 44000 41392 0.941 11 BTEE, i = 2.172 MODEL SI-073C From SUPERPIPE file: si073c3c350.s I e.out, dated 03/30/07,12:46:47 Conditiony jVq' Altoat&s No~d, Corn&#xfd; rent Equation 8 18800 6865 0.365 170 FILW i = 2.1 Equation 9 (Upset) 22560 12057 0.534 170 FILW i = 2.1 Equation 9 (Faulted) 23109 10757 0.465 170 FILW i = 2.1 Equation 10 28200 37047 1.314 J73 STRP, i =1.0 Equation 11 47000 38097 0.811 J73 STRP, i =1.0 MODEL SI-074C From SUPERPIPE file: si074c5a350x.out, dated 06/26/07 07:17:25 o~n tj r > IlionAll6 wabIe6 in u -&#xfd;; comp < Stes N e m F nt Stes s. p i4 Nme Type Equation 8 18250 9905 0.543 D35 BRED, i = 2.0 Equation 9 (Upset) 22560 16847 0.747 215A FILW, i =2.1 Equation 9 (Faulted) 23109 17693 0.776 215A FILW, i =2.1 Equation 10 28200 25107 0.890 C04A FILW, i =2.1 Equation 11 N/A N/A N/A N/A N/A Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A NIA FC07234 N/A PAGE NO. 31 of 37 Stress Summary (cont'd)Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.Load Conditions for ASME III Class 2/3 piping: Equation 8 Pd + W 1.0 Sh Ref. 2 Equation 9U Pd + W + E _ 1.2 0 Sh Ref. 2 Equation 9F Pd + W + E' K Sh = PB Ref. 2, 1 Equation 10 T + A S SA Ref. 2 Equation 11 Pd+W+T+A Sh + SA Ref. 2 Defini Sc Sh PB SA Pd W E A E'tion of Terms:= Allowable stress at minimum (cold) temperature.
= Allowable stress at the maximum analyzed piping temperature due to either operating conditions or maximum ambient temperature for Class 2 piping. (Used Design Temperature)
= Faulted allowable stress from Ref. 1, see next page.= Allowable stress for expansion 
= 1.25 Sc + 0.25 Sh.= Stress due to internal pressure loads at design pressure.Stress due to sustained mechanical loads including deadweight of piping, components, contents, insulation and lead shielding.
= Stress due to inertia effects of the Operational Basis Earthquake (OBEI).= Stress due to displacement effects of the Operational Basis Earthquake (OBEA).= Stress due to inertia effects of the Safe Shutdown Earthquake (SSEI).
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 NIA N/A F007234 N/A PAGE NO. 32 of 37 Stress Summary (cont'd)Maximum Pipe Stress Levels and Allowable Stresses -ASME III Class 2 and 3 piping.Eq. 9F allowable for all nodes, Conservative Temperature, Pressure & Diameter Used Dnom OD t D, Temp Sm SE f PD PM PB In In in in OF psi Psi I psig psi .. psi 12 12.750 0.375 12.000 350 16700 20040 600 5585 23109 12 12.750 1.312 10.126 650 16700 20040 2485 5094 23509 From Ref I (USAR, Appendix F, Table F -1)PB = K Sh P 8 = 4 Cos(-X ,)j" 2 SD , ="1.2 2 SM = Stress Intensity Allowable, Ref 2 So = 1.2SM LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. / A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 NIA N/A FC07234 NA PAGE NO. 3 3 of 37 Support Evaluation A review of the loads on all supports in Attachments A -N shows: " Support SIH-287 shall be removed.* All other supports are acceptable without modification.
* Many Pipe Support Calculations require revision.
These are the calculations associated with those supports with a "Change Factor" >= 1.1 as shown in Attachments A -N.Containment Penetration Evaluation Seismic Subsystems S1-082C, Sl-201A, S1-192A, SI-195A, SI-341A and SI-080C have boundary anchors modeled at Containment Penetrations M-16, M-17, M-86 and M-89. A review of the calculated stresses in the piping at the boundary nodes simulating Containment Penetrations shows that the calculated piping stresses are below the ASME Code allowable stress limits for all loading conditions.
As stated in Section 6.7.2 of PED-MEI-8 (Ref. 6), since the piping satisfies criteria consistent with the classification of the piping/penetration assembly, the loads upon the penetration are acceptable.
Embedded Pipe Evaluation Seismic Subsystems SI-201A, SI-201B, SI-203A, SI-205A, SI-191A, Sl-185A, SI-192A, S1-195A and SI-341A have anchors and guides modeled at piping that is embedded in the walls and floors of the Auxiliary Building.
A review of the calculated stresses in the piping at the nodes simulating embedded piping shows that the calculated piping stresses are below the ASME Code allowable stress limits for all loading conditions.
As stated in Section 6.7.2 of PED-MEI-8 (Ref. 6), since the piping satisfies criteria consistent with the classification of the embedment, the loads upon the embedment are acceptable.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. IA SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.: 0 N/A N/A FC07234 N/A PAGE NO. 34 of 37 Nozzle Evaluation Nozzles Loads have increased due to the increased temperature during SDC Mode.a The Loads on Pumps SI-1A & SI-1B have been submitted to the Flowserve for evaluation of their acceptability.
* The Loads on Heat Exchangers AC-4A & AC-4B are evaluated in Calculation FC07235.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. I A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE IREV. NO.: 0 NIA NIA FC07234 N/A PAGE NO. 35 of 37 Valve Acceleration Evaluation A review of the accelerations from the SUPERPIPE analysis for both the OBE and SSE cases shows that the maximum acceleration in the vertical direction is less than 2 g's at all nodes and that the maximum acceleration in the horizontal direction is less than 3 g's at all nodes.These accelerations satisfy the limits from Section 5.9 of Reference
: 4.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. /A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE R REV. NO.: 0 NIA NIA FC07234 N/A PAGE NO. 3 6 of 37 Wall Sleeve Bellows Evaluation Movements at the Wall Sleeve Bellows have increased due to the increased temperature during SDC Mode.* The Movements at the Wall Sleeve Bellows in Seismic Subsystem SI-191A for the Bellows shown on Drawings A-4436, D-4264 and IC-189 have been forwarded to S&L for evaluation.
Nodes Y45, 45 & Z45 apply.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts CALCULATION COVER SHEET CLIENT & PROJECT: OMAHA PUBLIC POWER DISTRICT PAGE 1 OF 23 Ft. Calhoun Station -Unit I TOTAL PAGES INCL. ATTACH: 290 CALCULATION TITLE (Indicative of Objective):
QA CATEGORY.; -1 Nuclear Safety Related Structural Evaluation of Shutdown Cooling Heat Exchangers for a (CQE)Shutdown Mode Pressure Increase El-Il El-,I LI -Non-Safety Related LI-CALCULATION IDENTIFICATION NUMBER _] -Fossil/Industrial Plant J.O. or W.O. NO. DIVISION & CURRENT CALC NO. OPTIONAL OPTIONAL GROUP TASK WORK PACKAGE NO.CODE N/A N/A FC07235 APPROVALS 
-SIGNATURE 
& DATE 7 CONFIRMATION
&916 R1150 7REQUIRED INDEPENDENT REV. NO. SUPERSEDES PREPARER(S)/DATE(S)
REVIEWER(S))
REVIEWER(S))
OR NEW CALC. NO. OR REV. YES NO DATE(S) DATES(S) CALC. NO. NO. ..John Spizuoco ohn Manning "r Ga r 0 N/A El El El1 El DISTRIBUTION GROUP NAME.& COPY GROUP [ NAME & LOCATION COPY LOCATION SENT SENT OPPD Fort ..El CalhounI Station I (original)
S&W Darlene j []Document Control McDonald Denver (electronic copy)__ElE EI El...._______ _________
__________
__________________I________
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. A SHAW GROUP COMPANY CALCULATION SHEET J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 0 N/A NIA FC07235 NIA PAGE NO. 5 1.0 CALCULATION OBJECTIVE This calculation analyzes the OPPD Shutdown Cooling Heat Exchangers, AC-4A & 4B, for structural loads including a higher than original design pressure (Ref. 1), new seismic loads from the Alternate Seismic Criteria and Methodologies (ASCM) program (Ref. 1), and new pipe connection (nozzle) loads (Refs. 17, 20 & 23). The heat exchangers were originally designed and fabricated in accordance with the ASME Code, Section III, Class A for the tube side and Class C for the shell side (Ref. 10). The tube side was reclassified as Section III Class C per the Fort Calhoun Station USAR (Ref. 11).This analysis of the Shutdown Heat Exchangers is for 550 psig (original design is for 500 psig) at 350'F at the tube side in order to accommodate an increase in the Shutdown Cooling System (SDC) entry conditions (Ref. 1). The shell side remains as originally designed.

==2.0 REFERENCES==

I1. Email from Douglas Molzer (OPPD) to John Manning (Shaw) dated May 17, 2007 transmitting new SDC Ht Ex design information.
: 2. Heat Exchanger Drawings: 2.1 Whitlock Mfg. Co. Drawing No. L-26133-Rev.
6 (OPPD No. 18676), Size 35-B-300 Type 1-R-2 Shutdown Heat Exchangers 2.2 Whitlock Mfg. Co. Drawing No. A-25888 (OPPD No. 10408), Size 35-B-300 Type l-R-2 Shutdown Heat Exchangers 2.3 Whitlock Mfg. Co. Drawing No. B-26133-S (OPPD No. 18618), Cradles- Part S 2.4 OPPD Drawing No. D-4749, Sh. 3, Rev. 0, Anchorage Deficiencies for Various Electrical 
& Mechanical Equipment 2.5 Whitlock Mfg. Co. Drawing No. B-26133-F Rev. 1 (OPPD No. 18619), Shell -Part F 2.6 Whitlock Mfg. Co. Drawing No. B-26133-A Rev. 2 (OPPD No. 18620), Channel -Part A 2.7 Whitlock Mfg. Co. Drawing No. B-26133-AF Rev. 2 (OPPD No. 18625), 12 to 36 Sweepolet 
-Part AF 2.8 Whitlock Mfg. Co. Drawing No. B-26133-L Rev. 1 (OPPD No. 18621), Baffles -Part L 2.9 Whitlock Mfg. Co. Drawing No. B-26133-Y (OPPD No. 18622), Gaskets -Part Y 2.10 Whitlock Mfg. Co. Drawing No. B-26133-FA (OPPD No. 18623), Welding Neck Flange for Channel -Part A-3 2.11 Whitlock Mfg. Co. Drawing No. B-26133-FS (OPPD No. 18624),,Welding Neck Flange for Shell -Part F 2.12 Whitlock Mfg. Co. Drawing No. H-26133-D Rev. 2, Tube Sheet Part D for Shutdown Cooling Exchangers Item # AC-4A, 4B 2.13 OPPD Drawing No. D-4077 Sh. 1 of 3, Seismic Restraints for 5 Block Shield Walls in the Auxiliary Building in Response to NRC Masonry Wall Bulletin IE-80-11 LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. A SHAW GROUP COMPANY CALCULATION SHEET J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 0 NIA NIA FC07235 N/A PAGE NO. 21 152 7948 2758 154 -5535 -9536 162 0 5261 164 0 -8411 Comparing the AC-4B support saddle loads due to the nozzle loads to those for AC-4A clearly show that the AC-4A heat exchanger supports are more highly loaded. Therefore, the AC-4B Hx supports need not be evaluated.
The anchor bolt stresses are acceptable., 9.0 RESULTS A summary of the results of the structural analysis for the OPPD SDC Heat Exchangers is presented below.Component Actual Thickness/
Required Thickness/
Ref. Page Stress Allowable Stress Tubeside Head 0.625 inches 0.5897 inches A-21 Tubeside Shell 0.625 inches 0.5995 inches A-16 Tuesd FaneSH = 18252 psi SH = 22875 psi A3 Tubeside Flange H 2PH PA-30 ST = 12247 psi ST = 15250 psi tubeside'Flag 
.** 6421 &#xfd;tincheA31 ae062. 2 A-hes Area Avail. = 7.2548 Tubeside Nozzles in 2  Area Req'd = 7.2547 in 2  A-41, 46 (Worst of two) 17156 psi max due to 24525 psi Allowable pipe loads Shell Side Flange SH = 21387 psi 1.5Sfo = 30000 psi A-36 ST = 16658 psi Sfo = 20000 psi A-36 Shell Side Flange 0.4375 inches 0.1387 inches A-38 Hub Shell Side Shell 0.4375 inches 0.1584 inches A-18 Shell Side Head 0.4375 inches 0.1578 inches A-27 Shell Side Nozzles Area Avail. = 3.33 in 2  Area Req'd = 2.49 in 2 (Worst of three) 26248 psi max due to 26250 psi Allowable LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC.I A SHAW GROUP COMPANY CALCULATION SHEET J.O. OR W.0 NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 0 N/A N/A FC07235 N/A PAGE No. 22 Component Actual Thickness/
Required Thickness/
Ref. Page Stress Allowable Stress pipe loads Tube Sheet 26695 psi 32700 psi B-4 Support Saddles (Bending at .21895 psi 26250 psi A-83 Saddle)Anchors 0.751 IF 1.0 20 Building Concrete 128 psi 750 psi A87 

==10.0 CONCLUSION==
S All of the analysis results, with the exception of the flange hub on the tube side, are acceptable and meet the minimum design requirements of the ASME Code Section III Division 1 2004 Edition w/2006 Addenda.The flange hub on the tube side flange is shown in Attachment A, page A-3 1, as requiring a thickness of 0.642 1 inches. The actual thickness of the flange hub is 0.625 inches. This represents a difference of less than i3%. In actuality, the flange hub has a 0.625 inch thickness.
at only the weld centerline.
The flange hub is tapered and increases in thickness through a 14-degree angle up to the tube sheet main cross section. The thickness at a fraction of an inch away from the weld centerline is equal or greater than the required thickness of 0.6421 inches.Since the flange forging has actual physical properties of 347 10 psi yield strength (Ref. 7)and the allowable stress is based on the minimum yield strength of 30000 psi, it can be concluded if the ratio of the actual yield to the minimum yield strength is equal or greater than the approximately 3% thickness difference, then the structure is acceptable.
The ratio of the yield strengths is 34710/30000 
= 1.157 or the actual yield strength is nearly 16%greater than the minimum required.
The flange hub is therefore acceptable based on the above reasoning.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. / A SHAW GROUP COMPANY CALCULATION TITLE PAGE CLIENT & PROJECT: OMAHA PUBLIC POWER DISTRICT PAGE 1 OF 25 Ft. Calhoun Station -Unit 1 TOTAL PAGES: 403 CALCULATION TITLE (Indicative of Objective):
QA CATEGORY;- I Nuclear Safety Related Calc- Stress Analysis for Subsystem SI-201A (CQE)i-I!LI-Ill EL -Non-Safety Related r-I-CALCULATION IDENTIFICATION NUMBER _ _ -Fossil/industrial Plant J.O. or W.O. NO. DIVISION & GROUP CURRENT CALC NO, OPTIONAL OPTIONAL I TASK WORK PACKAGE NO.CODE N- A N-A FC01029 NIA N/A APPROVALS 
-SIGNATURE 
& DATE CONFIRMATION
______)6_1_0__
REQUIRED INDEPENDENT REV. NO. SUPERSEDES PREPARER(S)/DATE(S)
REVIEWER(S)/DATE(S)
REVIEWER(S)/
OR NEW CALC. NO. OR REV. YES NO'DATES(S) CALC. NO. NO.FC01029, Rev. I rI 2 &John K. Manning R~bert McAuiiffe,'/" R rtcAuf' FC00914. Rev. I FC00916, Rev. I/--9-67 -,-0,,/ FC06524, Rev. 1 DISTRIBUTION GROUP NAME,& LOCATION COPY GROUP NAME &LOCATION COPY 1SENT !SENT OPPD Fort Calhoun Station SN & SENT (original)
Shaw Document Control Stone &Webster Denver (electronic copy)F]!D LI!I -i 1.LIC-08-0028"Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER.0 NUMBER DIVISION AND GROUP CALCULATION NUMBER IOOPTIONAL TASK CODE REV. NO.I/A NA FC01029 NIA PAGE NO.BACKGROUND 
& OBJECTIVE Background Revision 2 of this calculation for seismic subsystem SI-201A consolidates four calculations for this seismic subsystem into one calculation.
It incorporates the analysis performed in Reference 18 which addressed revised operating conditions for Shutdown Cooling per EC-35639 (Reference 19). A supplemental ASME III NB-3600 (Class 1) analysis is performed to show that the stresses in the Tee at Node 21 are acceptable per Reference 30.Objective The objective of this calculation is to evaluate the piping and supports included in the boundary of Seismic Subsystem SI-201A, as shown on the piping isometric drawing (Reference 10).Specifically, this calculation will:* Demonstrate that the piping / support configuration is adequate and meets the OPPD Analysis and Modeling Criteria (References 4, 5 & 6) and the ASME Section III Code requirements (References 2 & 30).* Determine and evaluate pipe support loads per References 3 & 7.* Verify that accelerations at valves remain within acceptance limits." Verify that containment, floor and wall penetration loads remain within acceptance limits.* Qualify Flanges.Summary of Changes made in Rev. 2 Revised component weight and flange weight.Revised run pipe stress intensification factor for Tee at Node 21.Added ASME Class 1 analysis of Tee at Node 21.Added documentation.
Incorporated input data and historical information from the following calculations which are being and superseded by this calculation:
FC01029, Rev. I Reference 8, See Attachment F.FC00914, Rev. I Reference 15, See Attachment G.FC00916. Rev. 1 Reference 16, See Attachment H.FC06524, Rev. 1 Reference 17, See Attachment J.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER ,J.O. OR W.0 NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 2 NIA NIA FC01 029 NIA PAGE NO. 8 4. CONCLUSIONS The following specific objectives of this calculation have been satisfied due to the valve weight, flow element weight, thermal condition and pipe support function changes: " After implementing the change in the operating conditions at entry to Shutdown Cooling Mode described in EC-35639 (Reference
: 19) and removing Support SIH-287 (also required by EC-35639), the piping / support configuration is adequate and meets the OPPD Analysis and Modeling Criteria (References 4, 5& 6) and the ASME Section III Code NC-3600 requirements (Reference 2). A supplemental ASME Section III Code NB-3600 Class I analysis is used to show acceptance of stresses in the 12 X 8 Reducing Tee at Node 21 per Reference 30.* Loads upon all pipe supports have been determined.
All revised loads are no greater than 110% of the loads used in previous support qualification calculations and are therefore acceptable without additional analysis.* The loads applied to flanges have been evaluated and are acceptable." The loads on Containment Penetration M-16 and the embedded wall and floor sleeves that form boundary anchors for this analysis are acceptable since the stresses in the connected piping are acceptable.
* The accelerations at valves remain within acceptance limits.Record of Actions Required to Support the Conclusion of this Calculation.
* Pipe support SIH-287 shall be permanently removed from the plant. This activity is included as part of EC-35639.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts.STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONALTASK CODE* REV. NO. 2 N/A NIA FC01 029 N/A PAGE NO. 17 7. RESULTS 

==SUMMARY==
ASME III CLASS 2 PIPE STRESS 

==SUMMARY==
(from SUPERPIPE file: si201a-rs2.spipe.out, dated 10/03/07, 10:55:33, See Attachments C & L)Condition Allowable Maximum Stress Node Component Stress Stress Ratio Name Type (psi) (psi)Equation 8 16400 6665 0.406 20 FILW SOP43W SIF =2.000 Equation 9 19680 16030 0.815 20 FILW (Upset) SOP43W SIF = 2.000 Equation 9 23344 15948 0.683 20 FILW (Faulted)
SOP 43W SIF = 2.000 Equation 10 27600 41940 1.520 21 TEE Highest
* SOP 18BL SIF = 1.830 Equation 10 27600 27030 0.979 20 TEE Second highest SOP 16BL SIF = 2.172 Equation 11 44000 45850 1.042 21 TEE Highest
* SOP 18BL SIF = 1.830 Equation 11 N/A N/A N/A N/A Except for Node 21, all nodes Second highest passed Eq 10.Therefore, Eq 11 is N/A* See the results from the ASME III NB-3600 Class 1 for the evaluation of the Tee at Node 20.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 2 NIA NIA FC01029 N/A PAGE NO. 18 Load Conditions for ASME III Class 2/3 piping: Equation 8 Pd + W < 1.0 Sh Refs. 2 & 4 Equation 9U Pd + W + E +A < 1.20 Sh Ref's. 2 & 4 Equation 9F Pd + W + E' K Sh = PB Refs. 1,2 & 4.Equation 10 T-+ 2A <SA Refs. 2 & 4 Equation 11 Pd + W + T + 2A Sh + SA Refs. 2 & 4 Definition of Terms: Sc = Allowable stress at minimum (cold) temperature.
S = Allowable stress at the maximum analyzed piping temperature due to either operating conditions or maximum ambient temperature for Class 2 piping. (Used Design Temperature)
PB = Faulted allowable stress from Reference 1, see next page.SA = Allowable stress for expansion 
= 1.25 S, + 0.25 Sh.Pd = Stress due to internal pressure loads at design pressure.W = Stress due to sustained mechanical loads including deadweight of piping, components, contents, insulation and lead shielding.
E = Stress due to inertia effects of the Operational Basis Earthquake (OBEI).A Stress due to displacement effects of the Operational Basis Earthquake (OBEA).E' = Stress due to inertia effects of the Safe Shutdown Earthquake (SSEI).T = Stress due to thermal expansion of the system in response to average fluid temperature.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC./A SHAW GROUP COMPANY CALCULATION SHEET LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. A SHAW GROUP COMPANY' CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.0 NUMBER DIVISION AND GROUP CLCULATION NUMBER IOPTIONAL TASKOCODE RV O NIA NIA I Fc0 0129 NIA PAGE NO. 20 ASME III CLASS 1 PIPE STRESS 

==SUMMARY==
(TEE @ NODE 21)(from NUPIPE-SWPC file: output.sum, dated 10/09/07, 13:20:51, See Attachments E & L)Material: Design Temperature:
Maximum Temperature:
Sm at Design Temperature:
SA403-WP304 350&deg;F 350&deg;F 19350 psi Spec., Reference 36 References 19 & 26 References 19 & 26 ASME Code, Ref. 2 NORMAL AND UPSET CONDITION ASME III NODE MEMBER MAX. CALC. ALLOWABLE NB 3600- POINT TYPE STRESS (psi) STRESS (psi)Equation 9 (Design) 21 TEE 8225 29025 Equation 10 21 TEE 86987 58050 Equation 12 ** 21 TEE 51968 58050 Equation 13 ** 21 TEE 28889 58050 ASME III NODE MEMBER CUMULATIVE ALLOWABLE NB 3600 POINT TYPE USAGE FACTOR***Equation 14 21 TEE 0.8601 1.0 Equation 9 (Design) Pd + D + E + H _ 1.5 Sm Equation 10 Pmax+ T + R + H + E +A+ L < 3.0 Sm Equation 11 Pmax + T + R + H + E + A + L Equation 12 T+R _< 3.0Sm Equation 13 Pmax + E + H + L + D _ 3.OSm Equation 14 Pax + T + R + H + E + A + L CUF< 1.0 FAULTED CONDITION ASME III NODE MEMBER MAX. CALC. ALLOWABLE NB 3600 POINT TYPE STRESS (psi) STRESS (psi)Equation 9 (Faulted) 21 TEE 8601 58050 Equation 9 (F)Pf+D+H + E'+Y'_< 3.0 Sm The load combinations are defined on the next page.** Either the requirements of Eq 10 or Eq 12 and 13 must be satisfied.
Usage Factor is based on Eq. 11 (peak stress range), Ke and Sm.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 2 NIA NIA FC01029 NIA PAGE NO. 21 Definition Of Terms S, = Stress intensity value at the higher of design or maximum operating temperature for Equation 9. Sm for the remaining equations is based on the operating temperature.
D = Stress due to sustained mechanical loads including deadweight of piping, components, contents and insulation.
E = Stresses due to inertia effects of the 0BE.A = Stresses induced in the piping due to response of the connected equipment and/or structures to the OBE (commonly referred to as OBE anchor movements).
E'= Stresses due to inertia effects of the SSE.H = Stresses due to occasional loads other than seismic. Examples of these loads would be: water hammer, steam hammer, opening of safety relief valves, etc.T = Stress due to thermal expansion of the system in response to average fluid temperature.
R = Stresses induced in the piping due to thermal growth of equipment and/or structures to which the piping is connected as a result of plant normal or upset conditions.
L = Local stresses in piping and/or piping components due to sudden changes in fluid temperature (commonly referred to as thermal transient effects).Y' = Stresses in piping and/or piping components due to pipe striking pipe (pipe whip) or blowdown of adjacent system (jet impingement).
Pd = Internal pressure loads due to design pressure.Pf = Internal pressure loads due to faulted plant operation.
Pna = Internal pressure loads due to range of operating pressure.
LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 2 N/A NA FC01 029 N/A PAGE NO. 22 SUPPORT EVALUATION A review of the loads on all supports in Attachment B shows:* Support SIH-287 shall be removed." Small bore supports are assumed acceptable for 600 pounds in each restrained direction. (See Section 3.) Calculated loads are less.* Since all "Change Factors" < 1.1, further evaluation of supports is not required.CONTAINMENT PENETRATION EVALUATION A boundary anchor at Node M-16 is used to model Containment Penetrations M-16.A review of the calculated pipe stresses at node M-16 shows that they are below the ASME Code allowable stress limits for all loading conditions.
As stated in Section 6.7.2 of PED-MEI-8 (Ref. 6), since the piping satisfies criteria consistent with the classification of the piping/penetration assembly, the loads upon the penetration are acceptable.
EMBEDDED PIPE EVALUATION Anchors and guides are modeled at piping embedded in the walls and floors of the Auxiliary Building.
A review of the calculated pipe stresses at nodes simulating embedded piping shows that they are below the ASME Code allowable stress limits for all loading conditions.
As stated in Section 6.7.2 of PED-MEI-8 (Ref. 6), since the piping satisfies criteria consistent with the classification of the embedment, the loads upon the embedment are acceptable.
Sleeve Material is assumed to be the same as attached pipe material. (See Section 3.)VALVE ACCELERATIONS Seismic accelerations are reported in Attachments C of this calculation.
A review of the SSE accelerations from these analyses shows that the maximum acceleration in the vertical direction is less than 2 g's at all nodes and that the maximum acceleration in the horizontal direction is less than 3 g's at all nodes. These accelerations satisfy the limits from Section 5.9 of Reference 4.US SUPPORT UPLIFT Uplift occurs on the two clevis hangers located at nodes 73 and 77 on the 2"-601R line. These supports were considered inactive in all cases except for the Weight analysis.
A review of the support summary shows that the movements at these supports are always upward except for the case when the lines are cold. Since they are unnecessary and only support weight when the piping is cold, they are acceptable as is.FLANGE CHECK The moments for flanges at Nodes X21, A21, 29 and 58 are enveloped and compared to the allowable for Class 151 and 301 flanges using the methodology from NC-3658.3 of the 1980 ASME III Code, Reference
: 2. Flange data is from Reference
: 5. Bolt data is from Reference
: 27. Temperature and Pressure data is LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INCJ A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO.NIA NIA F0F01029 NIA PAGE NO.from Section 6 of this calculation.
Loads are from the SUPERPIPE output in Attachment C.FLANGE QUALIFICATION OF 8" 150# RF FLANGED JOINTS Node Pt: 58,29 Note: This procedure is only applicable for bolting materials having an allowable stress of 20000 psi or greater at 100 0 F.Class Special 301/I 1R Type Weld Neck Flange Mat'l SA- 182 Gr. 304 Rating 150 Lb.Bore 8 NPS Sch 40 BOLTING Type Threaded Stud Bolt with 2 Heavy Hex Nuts Stud Mat'l SA 193 Gr. B7 Stud Nominal Diameter 0.750 in Stud Allowable 25000 psi Nut Mat'l SA 194 Gr. 2H FLANGE DATA C Diameter of Bolt Circle 11.750 in DF Diameter of Raised Face 10.625 in N Number of Bolts 8 Dr Root Diameter of bolts (UNC Series) 0.620 in A I Cross-Sectional Area of Each Bolt = Tr/4 DrI 0.302 in'AB Total Bolt Cross-Sectional Area = n
* Al 2.415 in 2 TD Design Temperature 500 'F PD Design Pressure 150 psi PO Max. Operating Pressure (N&U) .150 psi PFD Max. Operating Pressure (Faulted) 150 psi SY Yield strength of flange material at TD 19.40 ksi LOADING MFS Bending or torsional moment applied to the joint due to the combination of deadweight, thermal expansion, thermal anchor movements, relief valve steady state thrust and other sustained loads. (in-lbs)MFD Bending or torsional moment as defined for MFS but including the combined effect of all concurrent loadings. (in-lbs)EVALUATION Allowable Moments MFS 3125 (SY/36) CAB 47792 in-lbs MFD(N&U) 6250 (SY/36) CAB 95583 in-lbs MFD(Fltd)
[ 11250 AB -(Tr/16) DFa PFD ] C (SY/36) 150997 in-lbs Allowable Pressure Is PFD <= 2 PD ? Yes LOADINGS (from Attachment C)Mx My Mz Mb NODE Case (in-ibs) (in-lbs) (in-lbs) (in-lbs) Moments are in ALL DEAD 9423 551 10861 The Membcr local THRI 14766 4606 13054 system.THR2 18180 6442 32949 OBEI 5946 4126 2986 Moments are SSEI .8405 5778 3762 Acceptable "Yes" or "No" MFS 27603 6993 43810 44365 Yes MFD(N&U) 33549 11119 46796 48099 Yes MFD(Fltd) 36008 12771 47572 49256 Yes LIC-08-0028 Appendix A SDC Entry Conditions LAR RAI -Calculation Excerpts STONE & WEBSTER, INC. A SHAW GROUP COMPANY CALCULATION SHEET CALCULATION IDENTIFICATION NUMBER J.O. OR W.O NUMBER DIVISION AND GROUP CALCULATION NUMBER OPTIONAL TASK CODE REV. NO. 2 N/A NIA FC01 029 NIA PAGE NO. 24 FLANGE QUALIFICATION OF 8" 300# RF FLANGED JOINTS Node Pt: "X21, A21 Note: This procedure is only applicable for bolting materials having an allowable stress of 20000 psi or greatcr at 100'F.Class 301 R Type Weld Neck Flange Mat'[ SA-182 Gr. 304 Rating 300 Lb.Bore 8 NPS Sch 40S BOLTING Type Threaded Stud Bolt with 2 Heavy Hex Nuts Stud Mat'l SA 193 Gr. 97 Stud Nominal Diameter 0.875 in Stud Allowable 25000 psi Nut Mat'l SA 194 Gr. 2H FLANGE DATA C Diameter of Bolt Circle 13.000 in DF Diameter of Raised Face 10.625 in n Number of Bolts 12 Dr Root Diameter of bolts (UNC Series) 0.731 ih .A I Cross-Sectional Area of Each Bolt =Ti /4 Dr 2  0.420 in'AB Total Bolt Cross-Sectional Area = n
* Al 5.036 in'TD Design Temperature 350 "F PD Design Pressure 350 psi PO Max. Operating Pressure (N&U) 350 psi PFD Max. Operating Pressure (Faulted) 350 psi SY Yield strength of flange material at TD 21.60 ksi LOADING MFS Bending or torsional moment applied to the joint due to the combination of deadweight, thermal expansion, thermal anchor movements, relief valve steady state thrust and other sustained loads. (in-lbs)MFD Bending or torsional moment as defined for MFS but including the combined effect of all concurrent loadings. (in-lbs)EVALUATION Allowable Moments MFS 3125 (SY/36) CAB 122758 in-lbs MFD(N&U) 6250 (SY/36) CAB 245517 .in-lbs MFD(Fltd)
[11250 AB -(Tr /16) DF" PFD] C (SY/36) 381417 in-lbs Allowable Pressure Is PFD <= 2 PD ? Yes LOADINGS (from Attachment C)Mx My Mz Mb NODE Case (in-lbs) (in-lbs) (in-lbs) (in-lbs) Moments are in ALL DEAD 9423 1165 29388 the Member local THR1 1821 984 9333 system.THR2 1961 5251 36065 OBEI 5945 4111 7575 Moments arc SSEI 8405 6047 8930 Acceptable "Yes" or "No" MFS 11384 6416 65453 65767 Yes MFD(N&U) .17330 10527 73028 73783 Yes MFD(Fltd) 19789 12463 74383 75420 Yes LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 PRODUCTION ENGINEERING DIVISION QUALITY' PROCEDURE FORM PED-QP-3.1 R8 PAGE 1 OF 2 CALCULATION COVER SHEET Calculation Number: FC07096 Page No.: I QA Category:
[X] CQE I Non-CQE [ ] LCQE Total Pages: 17 Calculation Title: Short Term Calc: [ I Yes [X] No Evaluation of Temperature and Pressure Increase Vendor CaIc. No. TR-2007-09 for 6 UCL Associated Project::
EC35639 Software Tracking No.: Responsible NED Dept No.: 356 (from PED-MEI-23, if applicable)
Owner Assignment (by Dept Head): (Required only if there are affected documents to be changed)OPPD Engineer Assignment (by Dept Head): D. Molzer (Required only for verification of vendor/contractor calculations)
Verification of Vendor/Contractor CaIc. assumptions, inputs and conclusions complete: OPPD Engineer:
Doug Molzer _Date:  APPROVALS 
-SIGNATURE AND-DATt" Confirmation (Multiple preparers shall identify section prepared per PED-QP-3, Section 4.3.) Required?Supersedes Rev. Preparer(s)
Reviewer(s)
Required for CQE Calc No. Yes No No. Independent
_ _ _Reviewer(s) 0 See Attached NA X Cover Page LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 PRODUCTION ENGINEERING DIVISION QUALITY PROCEDURE FORM PED-QP-3.1 R8 PAGE 2 OF 2 CALCULATION COVER SHEET Calculation Number: FC07096 Page No.: 2 Applicable System(s) 
/ Tag Number(s)SI, AC /SI-1A, SI-1B, SI-1A-1 ,SI-1 B-1 EA's and/or Calculations Used as input in this Calculation FC07234 External Organization Distribution (Groups affected by this calculation)
Name and Location Copy Sent (./) Name and Location Copy Sent (/)
LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 PRODUCTION ENGINEERING DIVISION QUALITY PROCEDURE FORM PED-QP-3.2 R6 CALCULATION REVISION SHEET Calculation No.:FC07096 Page No.: Rev. # Description/Reason for Change 0 Initial Issue LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 1rn o'FLOWSERVE Pump Division Evaluation of Temperature and Pressure Increase for 6 UCL TR-2007-09 CUSTOMER:
Omaha Public Power District END USER: Fort Calhoun Nuclear Plant CUSTOMER ORDER No: 00106796 FLOWSERVE ORDER No: SERVICE: Low Pressure Safety Injection PUMP TYPE: 6 UCL PUMP S/N: 0669-58 0669-59' o Colnt.$Prepared by: Timothy Nish __._,____"_____"_
Design Engineer teviewed by: Paul Kasztejna 7A Supervising Design Engine I Approved by: John Lawler P.E.Product Engineer LL L'A'V TR-21)(1-09 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 -C C) o q , 4,. C Revision C: 09-July-2007 Remarks Date Prepared Reviewed !Approved A Initial Issue- 08-June-2007 Tim Nish PJK Preliminary
_B Final Report 26-June-2007 Tim Nish PJK JFL .-PE-4 ___-______
I JFL-P C Revisions per Customer Comments 09-i uly-2007 Tim Nish PJK fFI.- -PE;i LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 cc 0 -&#xfd;0)q 6 Pa4,krC2 TR-2D07-09 Revision C: O9-Juy-W j2007e CONTENTS:* R evisions .............................................. 
.............
i* A bstract ...............................................................
iI" Table of Contents .... ......... 
...................................
iv* Introduction 
.Results ..............................................................
2* Method of Analysis ......................................................
4* Discussion 
...........................................
6* C onclusion 
...............................................................
8* R eferences: 
............................................................
9 iv LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 A-C 0  TR-2007-09 Revision C: 09.July-2007 Replacing the existing hold down bolts, as shown on DWG No. 6UCL86X 1 9A Rev B, with ASTM At93 Gr. B7 would provide bolting material that has a maximum allowable stress that exceeds the maximum stiresses in the hold-down bolting. The 2001 ASME Boiler & Pressure Vessel Code, Section II, Part D lists the allowable stress limit as 25,000 psi at the desired operating temperature of 350 degrees Fahrenheit.
Normal practice for pressure certification requires a hydrostatic test of the pressure boundary components of the pump to 1.5 times the Maximum Allowable Working Pressure.
These pumps were originally designed with a Maximum Allowable Working Pressure of 500 psig and \Nvre hydrostatically tested at 750 psig. The 2001 ASME Boiler& Pressure Vessel Code, Section XI. Article IWA-4334.
states that only a leakage test is required.
Flowserve acknowledges this, however, it is considered 
-'good practice" to perforn the hydrostatic test.'therefore, it is the recommendation of Flowserve that these pumps be re-hydrostatically tested at 900 psig to qualify the new Maximun Allowable Working Pressure of 600 psig.iii LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 rC- O'70cf (6 k-?e, TRo2007-09 Revision C: 09-July.2007 ABSTRACT: Omaha Public Power District (OPPD) currently has (2) Flowserve 6 UCL pumps for low pressure safety injection at their Fort Calhoun Nuclear Plant. OPPD wants to change operating conditions of the pump without modifying the original equipment.
Specifically.
the pump's Maximum Allowable Working Pressure (MAWP) will be increased to 600 psig from the original MAWP of 500 psig and the suction pressure will increase from 300 psig to 350 psig. In addition, the maximum pumping temperature increased to 350 degrees Fahrenheit from 300 degrees.Flowserve has been contracted to analyze the operational changes requested by OPPD, and conclude whether or not the current equipment, as is, is capable of running safety and efficiently.
Specification F-6701, Project No. 07751-405 outlines the technical requirements of the evaluation.
Included in the evaluation will be the effects of nozzle loads applied to the suction and discharge nozz.les of the pump. OPPD has supplied a set of revised nozzle loads to be used in the evaluation.
OPPD also supplied a thermal shock scenario to be evaluated.
The pumps will be subjected to a large temperature change in the pumping fluid over a short duration, and the effects of this thermal shock on the pump are to be determined.
Considering the requested changes, Flowserve identified components that would be affected by theses changes and require analysis to be performed.
These components and the results of the analysis are summarized below in Table 1. The calculations were performed in accordance with 2001 ASME Boiler & Pressure Vessel Code. Section XI.Division 1, Article IWA-4000.
and under the QA provisions of lAW I OCFR 50.Appendix B.Table 1: Affected Components and Analysis Results Component Analysis Result Volute Wall Thickness Acceptable Suction/Discharge Flanges, Nozzles Acceptable Main Casing Bolting __ Acceptable Suction Bracket Thickness Acceptable Heat Exchangers Acceptable Anchor Bolting Acceptable Pump Hold-Down Bolting Unacceptable Casing Ring-Thermal Cycling Acceptable Based upon the results of the analysis, Flowserve Engineering concluded the operational changes to the low pressure safety injection pumps-desired by OPPD would require upgrading the pump hold-down bolts because of high stress levels. These stress levels are due to the Nozzle Loads that are applied. These bolts are currently ASTM A307 Gr. B.Flowserve design standards use an allowable stress limit of 15,000 psi for this material.ii LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 IOC ? 0 c 6 Q TR-2007-09 Revision C: 09-July-2007 INTRODUCTION:
In 1969, (2) Flowserve model 6 UCL pumps were supplied to Omaha Public Power District for use in their Fort Calhoun Nuclear Power Plant in a low pressure safety injection (LPSI) application.
The 6 UCL is a single stage, double suction. overhung pump design. Both, the suction and discharge nozzles, 10 and 6 inches respectively, are aligned in the vertical axis and connect to the volute on opposite sides of the shaft centerline.
This is typically referred to as top-top nozzle orientation.
The pump produces 1700 GPM and 400 feet of head at 3560 RPM. and is powered by a 300 hp electric motor.The LPSI pump is the entry position for the Shutdown Cooling System (SDC) at Fort Calhoun. The original, and current. conditions for this process require a reactor coolant temperature and pressure less than 300 degrees Fahrenheit and 250 psia. OPPD infonred Flowserve that if they could raise these limitations, and transition earlier to the SDC from the steam generator heat removal, the overall plant operation could be improved by reducing outage duration because the SDC offers a much higher cool down rate.OPPD advised Flowserve they would like to increase the Maximum Allowable Working Pressure of the LPS1 pumps and increase the operating temperature.
They requested an inlet pressure of 350 psig, compared to an original 300 psig, and a Maximum Allowable Working Pressure of 600 psig, compared to the original 500 psig. An increase to an operating temperature of 350 degrees Fahrenheit from 300 degrees was requested.
In addition, OPPD has supplied a set of nozzle loads and a thermal shock scenario to be evaluated.
Flowserve was contracted by OPPD to analyze the effect of this re-rate on the pump and integral heat exchangers, and determine if any modifications to the existing equipment were required.Specification F-6701, Project No. 07751-405.
details the scope of supply of the evaluation and technical requirements of the analysis.
In accordance with this specification, all results and recommendations of the re-rate were based upon calculations performed in accordance with the 2001 ASME Boiler & Pressure Vessel Code, Section X1, Division 1, Article IWA-4000, and under the provisions set forth by lAW IOCFR 50, Appendix B.The pumps in question were designed and constructed to acceptable industry standards and practices at the time. Given that the current study is being performed to a different set of standards than that of the original design and construction, this hereby serves as reconciliation that the analysis techniques and acceptance criteria, employed in this report, are equivalent or more stringent than those of the original design and construction.
I LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 cJc- Q ) 96 ' (5 TR.2007.09 Revision C: 09-July-2007 RESULTS:* Increase in MAWP and Temperature Flowserve Engineering identified critical components that would be affected by the pressure increase, and determined if the stress levels in these components would exceed design values. Table 2 summarizes the results of the analysis in which maximum allowable working pressures for each component was calculated and compared to the proposed pressure of 600 psig. The calculations were based on the current dimensions and material of the component and a temperature of 350 degrees Fahrenheit.
Table 2: Masimum Allowable Working Pressure by Component Description Result of Analysis Volute Satisfactory 
.Suction/Discharge Flanges Satisfactory)
Main Casing Bolting Satisfactory Suction Bracket L __ Satisfactorv The integral Hleliflow seal cooling heat exchangers.
model 8X4C- 10, were evaluated solely on the OEM vendor data. Flowserve Engineering has reviewed this information and concluded the heat exchangers are acceptable for the aforementioned temperature and pressure increases.
This specific model has a pressure rating of 2500 psig for the tubes at 350 degrees Fahrenheit, which is well above the desired 600 psig requested.
Additionally, the current bearing cooling water supply is sufficient for the desired increase in operating temperature and pressure for the lPSI pumps A and B.* Effects of Thermal Shock The thermal cycling criteria detailed in Specification F-6701 identified a temperature change of the operating fluid of 290 degrees Fahrenheit.
from 40-350 degrees Fahrenheit, in 5-10 seconds. This thermal shock would cause stress values in the casing ring to increase to 3970 psi, which is less than the 35,000 psi yield strength of the material.* Effects of Nozzle Loads There are (6) .75" anchor bolts that secure the base plate to the foundation, and (4) 1.0" bolts that secure the pump casing to the pedestal.
Tables 3 and 4 summarize the stress levels that will be seen in the bolting that secure the pump structurally.
The load descriptions in Tables 3 and 4 refer to the (3) operating classifications."Thermal" refers to a Normal scenario where only thermal and deadweight loadings are considered. "Upset" refers to a scenario where Normal + OBE loadings are used, and"Faulted" represents the final scenario of Normal + SSE. The value following the description refers to the iteration number of the nozzle loading results found in DIT-35639-01, provided to Flowserve by the Stone & Webster. All possible combinations for suction and discharge nozzle loadings were evaluated.
2 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 fco-w% je 4--E)TR-2007-09 Revisioni C: 09-"uy-2007 Table 3: Stress in Anchor Bolts by Load Case (LPS8I B)]uf&#xfd;I Thermal 1 Thermal I2 Thermal 2 Thermal 2 1 n1'rma 1 Thermal 2 Thermal 3 Thermal I Thermal 2 Thermal 3 4079 3742 3695 4029 3696 f1u(9205 7141 6804 8663 6838 upset 1 Upset 1 5629 15847 Upset I Upset 2 5424 18026 Upset 1 Upset 3 5617 15866 Upset 2 Upset 1 I 5815 15124 Upset2 Upset 2 5389 17305 Upset 2 Upset 3 5604 15146 Faulted 1 Faulted 1 5798 16193 Faulted 1 Faulted 2 5374 16487 Faulted 1 Faulted 3 5656 15957 Faulted 2 Faulted 1 5637 15131 Faulted 2 Faulted 2 5336 15683 Faulted 2 Faulted 3 5625 15153 Table 4: Stress in Pump Hold-Down Bolts by Load Case (LPSI IB)Thrml Thrml1 93 35 IThermall1IThermal 2 81822 3W6700 IThermal IlThermnal 3 5918 898I IThermal 2 Thermal 1 5446 35541 3 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 F6 0 -2 0 96 A.-C y ') 0 G'W_ 11, TR.2007-9 Revision C: 09-July-2007 The suction and discharge nozzles are cast into the casing and have a minimum thickness of 0.56 inches. Tables 5 and 6 summarize the stress levels in the nozzles due to the nozzle loadings.
The data and nomenclature are the same as those used in Tables 3 and 4.Table 5: Stresses In Suction Nozak Suction Nozzle1 20531 5435*Upsdt2 19281 5070 Faulted 1 20531 5481 Faulted 2 , 19221 5128 Table 6: Stresses in Discharge Nozzle Dischar Nozzle Thermal 1 1875 5437 Thermal 2 2896 8365 Thermal 3 1891 5441 Upset 1 4282 8119 Upset 2 7714 11530 Upset 3 4303 8128 Faulted 1 4326 8168 Faulted 2 5318 8796 Faulted 3 4347 8175 METHOD OF ANALYSIS:* Increase in MAWP and Temperature Flowserve Engineering analyzed the proposed increase of the Maximum Allowable Working Pressure using currently accepted dynamic and static methods, and compared these results to allowable stress levels at the revised temperature (350 degrees Fahrenheit).
The pressure limitations for this style pump are determined by (4) design criteria:
volute wall thickness, suction/discharge flange ratings, main bolting stress, and suction bracket thickness.
4 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 F-C. 0 -)v 76 TR-2007-09 Revision C; 09.July.20G7 Pressure limitations based on volute wall thickness are determined using the 1992 ASME Boiler & Pressure Vessel Code, Section VIII. Division I. Article UG-27...PR t =SE -0.6P Solving for P...SE: R + 0.6t Prcssure limitations based on suction and discharge flanges are detenrnined using the 1996 ANSI Standard 1316.5-Pipe Flanges and Flanged Fittings:
Pressure-Temperature Ratings. Both flanges are cast from ASTM A351. a group 2.2 material; therelore Table 2-2.2 is used.The main casing bolting is analyzed using the 1992 ASME Boiler & Pressure Vessel, Section ViIl. Division 1. Mandatory Appendix 2: Rules for Bolted Flange Connections with Ring Type Gaskets. to determine the maximum working pressure.
The hydrostatic force ofthe pressure and the compression load of the gasket seating must be resisted by the bolting load...AhS, =Ir G2P + 2Pb;7Gm 4 Solving for P...-G + 2bthn 4 The 1992 ASME Boiler & Pressure Vessel. Section VIII, Division 1, Article UG-34 is used to determine pressure limits based on the thickness of the suction head bracket...(SE)Solving for P...SEt Cd 2 5 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 ." C-6 4 -.10)p2&#xfd;ayd /3 TR.200709 Revision C: 0W-July-2007
* Effects of Thermal Shock Stress values in the casing rings were calculated using fundamental theories of stress and strain. In a free state, the increase in temperature would cause the casing ring to strain in order to alleviate induced thermal stresses.
However, the casing acts as a constraint not allowing the expansion, and thermal stresses build in the ring.The change in temperature is used to calculate the strain in the ring via the coefficient of thermal expansion.
A final change in diameter can be determined and used to calculate the area loading on the ring. Finally, the hoop stress of the ring can be determined.
.* Effects of Nozzle Loads The base plate anchor bolts and pump hold-down bolts were analyzed considering noz7lc loads, deadweight of the equipment, motor torque, and seismic loading. Applied forces and moments are translated to an origin, located at the geometric centroid of the bolting.and divided equally among the bolts. Vertical forces (Ry) are reacted by tensile loads.and horizontal forces (R\ & Rz) are reacted by shear loads. Moments about horizontal axes are resolved into vertical tbrees. and moments about the vertical axis are resolved into horizontal forces. From these reactions, the bolt stresses were calculated and compared to allowable stresses for the material provided The structural integrity of the suction and discharge nozzles was analyzed using fundamental theories of stress. Nozzle loading, deadweight of the equipment, working pressures.
and the geometry of the pump casing were all considered in determining maximum principal and shear stress values.DISCUSSION:
The pump's pressure vessel components must be able to meet the Maximum Allowable Working Pressure without exceeding allowable stress values. Given the dimensions and material properties a maximum allowable working pressure can be calculated for each critical component.
The governing component is that with the lowest maximum working pressure.
In this case, it is the volute wall thickness.
The maximum working pressure calculated for the volute is still greater than the Maximum Allowable Working Pressure of 600 psig. Therefore, the structural integrity of these pumps will not be affected by the increases in operating pressure and temperature.
However, it is important to note, qualifying the pump to this new Maximum Allowable Working Pressure requires a hydrostatic test at 1.5 times the operating pressure.
The pumps in question were originally hydrostatically-tested at 750 psig. Flowserve recognizes the 2001 ASME Boiler & Pressure Vessel Code, Section XI, Article IWA-4334, states only a leakage test is necessary.
However, it is considered "good practice" to re-hydrostatically test the casings whenever a change in MAWP is made. Therefore, it is the recommendation of Flowserve to re-hydrostatically test these pumps at 900 psig to qualify the pressure re-rate to 600 psig.6 LIC-08-0028 Appendix B Cc )O 0 j J&#xfd;C )SDC Entry Conditions LAR RAI -Calculation FC07096 TR-2007-09 Revision C: 09Juty-2007 The small size of the casing ring responds quickly to the temperature gradient, whereas the casing does not. Thermal expansion of the casing ring is prohibited by the casing, and thus thermal stresses rise in the casing ring. The thermal shock scenario will cause the hoop stress in the casing ring to increase, but will remain less than the yield strength of the casing ring material.As a result of the proposed increases in pressure and temperature revised thermal loadings were required.
The calculations for the anchor bolting and pump hold-down bolting were completed using the revised nozzle loadings provided by the Shaw Nuclear Group of Stone & Webster in document DT-35639-01.
The seismic accelerations were provided by OPPD in document DT-35639-02.
and the weight of the equipment was provided by FLS. Each LPSI pump (IA & 1 B) had unique load cases for each of the three nuclear load combinations:
Normal. Upset (OBE), and Faulted (SSE). LPSI B is the governing cawe due to the significantly higher loadings observed at the nozzles.The maximum tensile stress in the base plate anchor bolts is 18.000 psi. The anchor bolts were originally provided by the customer and therefore the material is unknown. The 18.000 psi tensile stress is slightly higher than the maximum allowable for ASTM A307 Gr. B, (or,,,, = 15,000psi).
a common carbon steel material.
The calculations used to determine the stresses in the anchor bolts. however, did not account for the manner in which the bedplate is installed.
Once bolted down, the bedplate is also filled with grout to further secure it. Considering this fact, Flowserve Engineering has determined the anchor bolts to be acceptable for the re-rated conditions The pump hold-down, bolts were originally provided by Flowserve:
they are a carbon steel ASTM A307 Gr. B. (om*, =15,000psi).
However, as shown in Table 4, the maximum stress that may occur in the pump hold-down bolts is 19,400 psi. Unlike the anchor bolts which are cemented in, the pump hold-down bolts are the only means by which the pump is secured to the pedestal.
These bolts will experience the entire loading.It is for this reason the pump hold-down bolts are unacceptable for the proposed re-rate conditions of the LPSI pumps at OPPD's Fort Calhoun. Flowserve recommends upgrading the material of the pump hold-down bolls, as shown on DWG No.6UCL86XI9a Rev B, to ASTM A193 Gr. B7 bolts (a.' = 25,000psi), which is above the maximum stress in the hold-down bolts, and would be acceptable for the new operating conditions.
Stress levels in the suction and discharge nozzles were evaluated using the same information referenced in the bolting analysis.
The maximum stress calculated for the discharge nozzle was 11,500 psi, while the maximum calculated stress for the suction nozzle was 5,500 psi, Both of these values are less than the yield stress of 35,000 psi for the material.7 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 LZ-t5 2 "'?C A0 TR.2007.09 Revision C: 09-July-2007 CONCLUSION:
At the request of OPPD, Flowserve analytically analyzed several critical components to determine if the Flowserve model 6 UCL LPSI pumps at the Fort Calhoun Nuclear Station were capable of an operational increase in conditions.
The proposed conditions change include increasing the suction pressure from 300 psig to 350 psig, increasing the Maximum Allowable Working Pressure from 500 psig to 600 psig, and increasing the operating temperature from 300 degrees Fahrenheit to 350 degrees.Flowserve concluded that the pump set. as is. is not acceptable for the re-rated conditions.
However. the only critical components are the pump hold-dowii bolts due to stresses exceeding the maximum allowable working stresses.
An upgrade of the pump hold down bolts, as shown on DWG No. 6UCL86XI9A Rev B, from ASTM A307 Gr. B to ASTM A193 Gr. B7 would sufficiently raise the maximum allowable working stress of the bolts beyond the calculated working stress requirements.
In addition.
Flowserve recommends the casings be re-hydrotested at 900 psig to satisfy a pressure certification requirement of a hydro test at 1.5 times the operating pressure.8 LIC-08-0028 Appendix B SDC Entry Conditions LAR RAI -Calculation FC07096 C 9 6 -TR.207.09 Reylsion C: 09uJ*-200 

==REFERENCES:==
* Specification F-6701, Project No. 07751405 0 ASME Boiler & Pressure Vessel Code, 200 1 Edition 0 ASME Boiler & Pressure Vessel Code, 1992d Edition 0 lAW IOCFR 50* OPPD document DT-35639-01:
Revised Nozzle Loadings* OPPD document DT-35639-02:
Seismic Accelerations
* l-etter from FLS Engineering (Paul Kasztejna) to OPPD. dated 16-May-2007
* 1999 Annual Book of ASTM Standards, Volume 15.08: Fasteners* 1996 ANSI Standard B16.5-Pipe Flanges and Flanged Fittings:
Pressure-Temperature Ratings* Roark's Formula for Stress and Strain, 6'h Edition 9}}