ML042600388
| ML042600388 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 09/16/2004 |
| From: | Dennig R NRC/NRR/DSSA/SPSB |
| To: | Marshall M NRC/NRR/DLPM/LPD2 |
| HAYES J NRR/DSSA/SPSB 415-3167 | |
| References | |
| TAC MB5105 | |
| Download: ML042600388 (29) | |
Text
September 16, 2004 MEMORANDUM TO: Michael L. Marshall, Jr., Acting Chief Project Directorate II Division of Licensing Project Management FROM:
Robert L. Dennig, Chief /RA/
Containment and Accident Dose Assessment Section Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis
SUBJECT:
SAFETY EVALUATION ON H. B. ROBINSON REQUEST INVOLVING FULL IMPLEMENTATION OF THE ALTERNATIVE SOURCE TERM (TAC NO. MB5105)
Attached is our Safety Evaluation (SE) addressing the proposed changes to the H. B. Robinson Unit 2 Technical Specification (TS) and Robinsons full implementation of the alternative source term (AST). Our assessment of the proposed changes included the verification that the offsite and onsite consequences of various postulated accidents would not result in doses which would exceed the criteria of 10 CFR 50.67. Those TS changes which were reviewed involved the definition of Dose Equivalent 131I, the allowable levels of Dose Equivalent 131I in primary coolant, and the allowable steam generator primary to secondary leak rates.
The results of the staffs review show the acceptability of the proposed implementation of the AST with the exception of the loss-of-coolant accident (LOCA). Appendix A of Regulatory Guide 1.183 describes the manner in which containment spray removal rate constants are to be selected. The manner chosen by the licensee was inconsistent with this guidance. Since the licensee did not provide a sufficient amount of information to allow the staff to reach a conclusion on the acceptability of their proposed method, the use of AST for the LOCA can not be approved. Additional details on this issue and the staffs basis for approving the implementation of AST and the proposed TS changes are presented in the Attachment to this memorandum.
CONTACTS: Leta Brown, NRR/DSSA/SPSB - Atmospheric Dispersion 415-1232 Jack Hayes, NRR/DSSA/SPSB - Radiological Dose Assessment 415-3167
M. Marshall 2
The TS and radiological dose aspects of this review were performed by John Hayes. The atmospheric dispersion aspects of the review were performed by Leta Brown.
Docket No.: 50-261
Attachment:
No. 1 - Safety Evaluation CONTACTS: Leta Brown, NRR/DSSA/SPSB - Atmospheric Dispersion 415-1232 Jack Hayes, NRR/DSSA/SPSB - Radiological Dose Assessment 415-3167
M. Marshall The TS and radiological dose aspects of this review were performed by John Hayes. The atmospheric dispersion aspects of the review were performed by Leta Brown.
Docket No.: 50-261
Attachment:
No. 1 - Safety Evaluation CONTACTS: Leta Brown, NRR/DSSA/SPSB - Atmospheric Dispersion 415-1232 Jack Hayes, NRR/DSSA/SPSB - Radiological Dose Assessment 415-3167 DISTRIBUTION:
Central File SPSB R/F SPSB S/F, H. B. Robinson AST R. Dennig, L. Brown J. Hayes ADAMS Accession #ML042600388 TEMPLATE NO.: NRR-096 PUBLICLY AVAILABLE? (Y or N) Y SENSITIVE?(Y or N) N hayes\\version 09072004.wpd OFFICE SPSB/DSSA SPSB/DSSA SC:SPSB NAME JJHayes LABrown RLDennig DATE 09/ 14 /04 09/ 15 /04 09/ 16 /04 OFFICIAL RECORD COPY
ATTACHMENT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. TO FACILITY OPERATING LICENSE NO. DPR-23 PROGRESS ENERGY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
In a letter to the U.S. Nuclear Regulatory Commission (NRC) dated May 10, 2002, and as supplemented in letters dated March 12, 2003, March 5, 2004, and July 22, 2004, Carolina Power & Light Company, now part of Progress Energy, (the licensee) requested an amendment to Operating License DPR-23 for H. B. Robinson Unit 2. The amendment would make the following modifications to the Robinson Technical Specifications:
The definition of Dose Equivalent 131I in Section 1.1 would be revised to reference as the iodine dose conversion factors those listed under the Effective column of Table 2.1 of Federal Guidance Report 11.
The reactor coolant system (RCS) operational leakage limits, stated in Limiting Condition for Operation (LCO) 3.4.13, RCS Operational Leakage, for total primary to secondary leakage through the steam generators would be reduced from 1 gpm to 0.3 gpm. In addition, the allowable primary to secondary leakage for any one steam generator would be reduced from 500 gpd to 150 gpd.
The maximum allowable activity levels of Dose Equivalent 131I in TS 3.4.16, RCS Specific Activity, would be reduced from 1µCi/g in Condition A and in Surveillance Requirement 3.4.16.2 to 0.25 µCi/g. In addition, Figure 3.4.16-1, Reactor Coolant Dose Equivalent 131I Specific Activity Level Versus Percent of Rated Thermal Power, would be deleted. Required Action A.1 would be revised to replace the reference to the acceptable region of Figure 3.4.16-1 with a limit of 60 µCi/g Dose Equivalent 131I. The second entry condition of Condition C would be revised to replace the reference to the unacceptable region of Figure 3.4.16-1 with a reference to >60 µCi/g Dose Equivalent 131I.
The licensee withdrew a proposal to revise the description of the Explosive Gas and Storage Tank Radioactivity Monitoring Program in Technical Specification 5.5.12. The licensee had also proposed a change to Appendix B, Additional Conditions, related to cycle length restriction. This aspect is no longer a part of this amendment request as this restriction was removed in an amendment issued by the NRC on March 10, 2004.
2 The licensee proposed these revisions to the technical specifications as a result of their application for a full-scale implementation of the alternative source term (AST) under 10 CFR 50.67.
2.0 BACKGROUND
A licensees adoption of AST requires analyses of those accidents appropriate for the type of reactor facility. NRC guidance on the performance of such analyses is presented in Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000. An acceptable demonstration involves showing that both the exclusion area boundary (EAB) and low population zone (LPZ) doses are less than 25 rem total effective dose equivalent (TEDE) or some fraction thereof, depending upon the accident. In addition, licensees must demonstrate that the control room operator dose meets 10 CFR 50.67. Accidents which are typically analyzed are based upon reactor type. For a pressurized water reactor, the typical accidents analyzed include the main steamline break (MSLB), the locked rotor, the rod ejection, the steam generator tube rupture (SGTR), the fuel handling accident and the large break loss-of-coolant accident (LOCA). For the application of the AST to Robinson, the licensee calculated the dose to all of these accidents except the rod ejection and the fuel handling accident. Previously, the licensee had submitted the fuel handling accident utilizing AST (March 13, 2002). Approval of the use of AST for the fuel handling accident was issued for the Robinson Station in Amendment 195, dated October 4, 2002. The licensee did not calculate doses for a rod ejection accident for the reasons discussed in Section 3.3.5 of this Safety Evaluation. The licensee did calculate the consequences of a single rod control cluster assembly (RCCA) withdrawal.
The licensee calculated doses for individuals located at the EAB, the LPZ and for individuals located in the control room and the Technical Support Center (TSC)/ the Emergency Operating Facility (EOF). As part of the licensees implementation of the AST, the licensee also calculated new onsite and offsite atmospheric dispersion values.
A licensees implementation of AST may necessitate changes to a facilitys technical specifications. For Robinson, those technical specifications which were proposed for change included the definition for Dose Equivalent 131I, the operational leakage limits for the RCS primary to secondary and the maximum allowable activity levels of Dose Equivalent 131I in the RCS.
3.0 EVALUATION 3.1 Definition of Dose Equivalent 131I The licensee proposed a change in the definition of dose equivalent 131I. They proposed to define dose equivalent 131I using the Effective Column from Table 2.1 of Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion.
3 The staff has reviewed the proposed revision to the definition and finds the licensees proposal acceptable.
3.2 Explosive Gas and Storage Tank Radioactivity Monitoring Program The licensee originally proposed to change the criterion from 500 rem whole body to 500 rem TEDE. The staff indicated that the appropriate TEDE criterion for the release of the contents of a waste gas decay tank is 100 mrem TEDE. The 100 mrem value is consistent with limits of 10 CFR 20.1301 and the guidance in Branch Technical Position (BTP) ETSB 11-5.
The licensee is not required to change the criterion to a TEDE as a part of the full implementation of AST. They may maintain their existing criterion at 500 mrem whole body.
But if they wish to change to a TEDE criterion, the criterion must be 100 mrem TEDE.
The licensee chose to withdraw their request rather than to change the dose criterion to 100 mrem TEDE. They did that in their March 5, 2004, letter. Consequently, the acceptable dose criterion will remain at 500 mrem whole body for the release of the contents of the waste gas decay tank.
3.3 Assessment of Radiological Consequences 3.3.1 Large Break LOCA The staff performed confirmatory calculations of the potential consequences of a LOCA based upon information provided in the licensees May 10, 2002, and March 12, 2003, submittals and in response to staff questions. The staff calculated EAB, LPZ and control room operator doses.
The staffs calculations could not confirm that Robinsons implementation of AST for the LOCA resulted in doses which met the 10 CFR 50.67 dose acceptance criteria for the EAB. This occurred because only a fraction of the total forced flow of 130,000 cfm could be justified as a mixing rate between the sprayed and unsprayed regions. The licensee confirmed the accuracy of the staffs conclusion. During the time that the licensee was assessing the staffs conclusion, the licensee also identified a non-conservative error in the modeling of containment leak rate.
In order to compensate for the impact of these two items on the calculated dose consequences, the licensee re-evaluated and revised certain LOCA dose inputs. The changes were presented in the licensees March 5, 2004, submittal. The changes related to the modeling of the containment and to the removal mechanisms in containment. There were no changes to the modeling of the dose contribution from engineered safety features (ESF) leakage outside containment. There were no changes to the control room and Technical Support Center (TSC) design inputs and there were no changes to the assumptions involving atmospheric dispersion factors or core inventory source terms. Because of the change in containment modeling, the amount of activity in containment was altered which affected the quantity of activity released from containment and the direct dose the control room and TSC as a result of direct shine and plume shine. The licensee re-evaluated these direct doses. The specific changes to containment modeling were the following:
4 a) the sprayed and unsprayed volumes in containment; b) the origin of the source term associated with containment leakage; c) the aerosol spray removal model; d) the natural deposition removal cut-off time; e) the mixing rate between the sprayed and unsprayed containment regions; and f) spray removal cut-off times.
Additional details follow.
The volume of the sprayed area of the containment was originally 52%. It was originally postulated that only gravity affected the spray coverage patterns. In the May 10th submittal, the licensee had neglected any spray coverage occurring due to air movement patterns. In their March 5, 2004, letter, the licensee revised their spray volume calculation based upon air movement patterns described in NUREG/CR-4102, Air Currents Driven by Sprays in Reactor Containment Buildings. As a result of this re-calculation, the spray volume of the containment was determined to be 82.9% with spray train A operating and 81.5% with spray train B operating.
In the licensees May 10, 2002, calculation of the LOCA consequences, they had assumed that all of the containment leakage originated from the unsprayed region of the containment. During their re-assessment of the consequences of a LOCA they determined that this assumption had been implemented incorrectly into the calculation model. The licensees March 5, 2004, analysis assumed that containment leakage occurred from both the sprayed and unsprayed regions in proportion to the volumes of each region.
The licensees May 10, 2002, analysis included a particulate spray removal coefficient based upon a model specified in Standard Review Plan Section 6.5.2. The licensees March 5, 2004, letter incorporated an analysis which utilized the 50 percentile Powers spray removal model as incorporated into the RADTRAD code. This model represented a best estimate or mean removal rate values expected from mechanistic models for any given set of containment spray parameter inputs.
In the licensees original analysis, they had assumed that natural deposition removal would cease when a decontamination factor (DF) of 1000 was achieved. The revised calculation assumed that natural deposition occurred throughout the duration of the accident.
The assumption of mechanical mixing in the containment was modified in the March 5, 2004, analysis. Originally, it was assumed that 130,000 cfm was removed from the sprayed region and transferred to the unsprayed region and 130,000 cfm was removed from the unsprayed region and transferred to the sprayed region. The revised calculation assumes that 65,000 cfm is taken from the sprayed region and transferred to the unsprayed region and visa versa.
The original calculation assumed that elemental spray removal was cutoff at t = 2.46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> when a DF of 200 was achieved. With the changes in the sprayed and unsprayed volumes and
5 mixing rates, the time at which an elemental iodine DF of 200 was achieved became 2.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for spray train A and 2.11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> for spray train B.
The May 10, 2002, analysis established a time (20.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) for particulate iodine at which a DF of 50 would be achieved. At that time, the spray removal coefficient for iodine was to be reduced by a factor of 10. With the use of the time dependent Powers spray removal mode in the March 5, 2004, analysis, it was not necessary to apply a factor of 10 reduction when the DF = 50. In both the previous and the revised licensees calculations, the sprays were secured before a DF = 50 was ever achieved.
The licensee calculated the potential consequences of a postulated large break LOCA to the control room operators and to individuals located offsite at the EAB and at the LPZ. It was postulated that the occurrence of a LOCA would result in releases to the environment from containment leakage and emergency core cooling system (ECCS) recirculation loop leakage.
The release of alkali metals and the elemental and particulate forms of iodine to containment would be reduced by containment sprays and by natural deposition when the sprays were not in operation. Containment fans would mix the radioactivity between the sprayed and unsprayed regions.
The licensees calculations assumed containment leakage occurred at the maximum allowable leakage value in the TS containment leakage program. The containment leakage value was reduced to 50% of the TS value at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident.
The licensee assumed ECCS leakage was twice the limit (2 gph) in the Robinson Technical Requirements Manual Specification 3.23, Post Accident Recirculation Heat Removal System Leakage. This leakage was presumed to begin at the earliest time that recirculation flow starts and end at the time recirculation flow ceases. The licensee stated that the results of their analyses demonstrated that in all cases 10 CFR Part 50.67 doses were met.
Appendix A of Regulatory Guide 1.183, contains a footnote which indicates the manner in which the spray removal rate constants developed by the use of the Powers model, (i.e., the model described in NUREG/CR - 5966, A Simplified Model of Aerosol Removal by Containment Sprays) are to be selected. For design basis calculations, these constants are to be selected in a manner which maximized the dose consequences. The licensees selection of the 50th percentile value of the Powers model did not maximize the dose consequences.
In addition, the licensee did not provide sufficient information for the staff to conclude that the use of the 50th percentile value was an acceptable alternative. Therefore, the staff could not approve the licensees proposed use of the AST for the LOCA.
3.3.2 Main Steamline Break The licensee evaluated the consequences of a MSLB. Three cases were assessed. The first case assumed the accident occurred following an iodine spike, referred to as the pre-existing spike case. The second case assumed the MSLB initiated an iodine spike, referred to as the accident-initiated spike. The third case assumed that the MSLB induced fuel failures. In all
6 cases, a 150 gpd primary to secondary leak rate was assumed to the steam generator with the steamline break (referred to as the faulted steam generator). It was also assumed that the primary to secondary leak to the unaffected steam generators (referred to as the intact steam generators) was the remaining primary to secondary leakage allowed by technical specifications, i.e., 0.19 gpm (0.3 gpm - 150 gpd).
The pre-existing iodine spike case assumed reactor coolant activity level was at the TS limit of 60 µCi/gm of dose equivalent 131I. The accident-initiated spike case assumed that the reactor coolant activity level was at twice the proposed TS 3.4.16 limit of 0.25 µCi/gm dose equivalent 131I. For this RCS activity level it was assumed that, concurrent with the MSLB, an iodine spike occurs which results in the release of iodine from the fuel to the reactor coolant.
Iodine was assumed to be released at a rate which is 500 times the normal iodine release rate associated with a reactor coolant activity level of 0.5 µCi/gm dose equivalent 131I. The spike was assumed to occur for eight hours. For the fuel failure case, it was assumed that two fuel assemblies were breeched and that these assemblies had a maximum radial peaking factor of 1.8. For all cases, the secondary system activity was assumed to be at the Technical Specification limit of 0.1 µCi/gm dose equivalent 131I.
In all cases, steam generator dryout was assumed for the faulted steam generator. Primary to secondary activity released to the faulted steam generator was presumed to be released directly to the environment. Primary to secondary leakage to the intact steam generators was presumed to be mixed with the bulk liquid in the steam generators. The activity in the intact steam generators was assumed to be released to the environment as vapor based upon the steaming rate and the partition coefficient of the particular nuclide group. Primary to secondary leakage was assumed to continue until the primary side pressure leakage was reduced below that of the secondary side or until the temperature of the leakage was reduced below 212oF.
For Robinson, the analysis incorporated 53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to initiate residual heat removal (RHR) operation and 98.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to terminate primary to secondary leakage.
The staff performed confirmatory calculations of the potential consequences of a MSLB accident based upon information provided in the licensees submittals and in response to staff questions. The assumptions which form the basis for the staff calculations are presented in Table 3.3-1. The staff calculated EAB, LPZ and control room operator doses. The results of these calculations are presented in Table 3.3-5. The staffs calculations confirmed that the Robinsons implementation of AST for the MSLB does not result in doses which exceed the acceptance criteria in Regulatory Guide 1.183.
In reviewing the licensees analysis, the staff determined that the licensee had utilized ICRP 30 dose conversion factors for determining the dose equivalent 131I in primary coolant. With the proposed change in the definition of dose equivalent 131I utilization of ICRP dose conversion factors, use of ICRP 30 dose factors is no longer appropriate. Thus, this aspect of the licensees methodology is no longer acceptable. The methodology utilized by the licensee to determine the dose equivalent 131I in primary coolant needs to utilize the Effective dose conversion factors in Table 2.1 of Federal Guidance Report 11. A necessary and sufficient
7 condition for granting this amendment is that the licensee change this methodology in future assessments of the consequences of this accident.
3.3.3 Steam Generator Tube Rupture The licensee evaluated the consequences of a SGTR. Two cases were assessed. The first case assumed the accident occurred following an iodine spike, referred to as the pre-existing spike case. The second case assumed the SGTR initiated an iodine spike, referred to as the accident-initiated spike. For the steam generators without the tube rupture, referred to as the intact steam generators, primary to secondary leakage was assumed to be 150 gpd per steam generator for both the pre-existing and the accident-initiated spike cases. The remaining primary to secondary leakage allowed by technical specifications 0.08 gpm (0.3 gpm - [2 steam generators x 150 gpd/sg]) was to the steam generator with the tube rupture (referred to as the faulted steam generator).
For the pre-existing iodine spike case, it was assumed that the reactor coolant activity level was at the TS value of 60 µCi/gm dose equivalent 131I. For the accident-initiated spike case, it was assumed that the reactor coolant activity level was at the proposed TS 3.4.16 limit of 0.25
µCi/gm dose equivalent 131I. In the licensees March 10, 2002, letter they had indicated, for the accident-induced iodine spike case, a similar assumption is made with one exception.
The primary coolant iodine activity increases during the first eight hours of the transient as a result of the release from the defective fuel at rate 335 times the iodine equilibrium appearance rates consistent with an initial dose equivalent (DE) I-131 concentration twice the value of the proposed TS 3.4.16 limits. Based upon this statement, it would be expected that the licensee incorporated into Robinsons SGTR analysis a release rate from the fuel which equated to an initial primary coolant activity level of 0.50 µCi/g of dose equivalent 131I. This release rate is a factor of two greater than the release rate which would occur with primary coolant at 0.25 µCi/gm dose equivalent 131I.
The licensees March 5, 2004, letter indicated that the statement from the March 10, 2002, letter was incorrect and that the release rate was based upon primary coolant activity being at 0.25 µCi/g of dose equivalent 131I. The licensee indicated that the SGTR would not result in any fuel failures.
For both the pre-existing and accident-initiated spike cases, secondary system activity was assumed to be at the TS limit of 0.1 µCi/gm dose equivalent 131I. In addition, for both cases, the licensee assumed the faulted steam generator would be isolated within 30 minutes and that offsite power was lost. Thus, the main condenser was unavailable for steam dump. The licensees justification for the 30 minute assumption for isolation was that assumption was part of their current licensing basis. The licensee indicated that the original plant licensing basis established the 30 minute steam generator isolation time for the SGTR event and that this basis was re-affirmed through the NRCs review of the steam generator replacement and power uprate licensing changes in the early 1980s. Neither break flow nor primary to secondary leakage is assumed to be terminated within those 30 minutes.
8 The licensee indicated in their March 5, 2004, submittal that simulator experience has shown that the continuing break flow will not result in the opening of any main steam safety valves nor does it result in a steam generator overfill condition. The March 5, 2004, submittal also indicated that isolation of the affected steam generator was consistent with Robinsons current operating procedures. In a July 22, 2004, letter which supplemented the March 5th response, the licensee indicated that they had run a SGTR scenario in June 2004 on the Robinson simulator. This scenario included a 695 gpm primary to secondary leak rate due to a tube rupture with an open power-operated relief valve (PORV) on the steam generator with the tube rupture and a loss of offsite power. When the open PORV was manually closed, the steam generator with the tube rupture was isolated from the environment in 20 minutes compared to the 30 minutes assumed in the AST scenario. During this scenario, the appropriate response and mitigation procedures were followed during the simulator scenario.
For both the pre-existing and the accident-initiated spike cases, it was assumed that a portion of the ruptured tube flow would flash to steam. The portion which flashes to steam was assumed to rise through the bulk water in the steam generator and to enter the steam space where it would be immediately released to the environment without mitigation, i.e., no credit for scrubbing within the bulk water. Primary to secondary leakage to the faulted steam generator and that portion of the tube rupture flow which does not flash was assumed to mix with the bulk water and to be released to the environment based upon the steaming rate and the partition coefficient for the particular nuclide group. For intact steam generators, the activity released to the environment was also based upon the steaming rate and the partition coefficient for a particular nuclide group. Primary to secondary leakage would continue until either the primary side pressure was reduced below that of the secondary side or until the temperature of the leakage was reduced below 212oF. The release of radioactivity was assumed to continue until the RHR system was place in operation. The licensee assumed that it took 53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to cool the reactor down to a point where no further release of steam and radioactivity would occur to the environment. At that time releases from the steam generators would terminate.
Item 14 of Attachment II of the March 10, 2002, submittal stated, The ratio of radioiodines to other radionuclides provide in the UFSAR, Table 11.1.1-2, is assumed to be a constant.
Based upon this statement, it could be assumed that for the pre-accident spike case, where the dose equivalent 131I activity level is assumed to have spiked to 60 µCi/g, a comparable spike has occurred for the other radionuclides such as cesium. This would also apply to the accident-initiated spike case. The licensees consequences analysis for the SGTR did not appear to maintain these ratios when considering spikes. The licensee was requested to clarify their assumptions for the SGTR. The licensee provided such clarification in the March 5, 2004, letter. They indicated that the statement on ratios only applies with regard to the reactor coolant activity level at the technical specification values (60 µCi/g or 0.25 µCi/g). In regard to spiking, only the iodine isotopes were considered to spike consistent with Appendix F of Regulatory Guide 1.183. The licensee assessed the potential for the spiking of other isotopes in addition to iodine. They concluded that some cesium spiking would occur and it would be in the form of CsI and that it would be in the particulate form. They concluded that very little of this activity would be released from the secondary side of the steam generators especially if the
9 steam generator tubes remain covered with water. The Robinson SGTR analysis indicates that the tubes will remain covered in the event of a SGTR.
The staff has performed their assessment of the potential consequences of a SGTR event.
The staffs assessment assumed that the reactor coolant activity level for dose equivalent 131I was at 60 µCi/g for the pre-existing spike case and at 0.25 µCi/g for the accident-initiated spike case. Table 3.3-2 presents the assumptions utilized by the staff in their assessment.
The potential consequences of a SGTR accident are presented in Table 3.3-5. The staffs calculations confirm that the consequences of a steam generator tube rupture accident met the dose criteria (25 rem TEDE pre-existing spike case and 2.5 rem TEDE accident-initiated spike case) of Regulatory Guide 1.183 at the EAB.
In reviewing the licensees analysis, the staff determined that the licensees calculations had utilized ICRP 30 dose conversion factors for determining the dose equivalent 131I in primary coolant. With the proposed change in the definition of dose equivalent 131I, utilization of ICRP 30 dose conversion factors is no longer appropriate. Thus, this aspect of the licensees methodology is no longer acceptable. The methodology utilized by the licensee to determine the dose equivalent 131I in primary coolant needs to utilize the Effective dose conversion factors in Table 2.1 of Federal Guidance Report 11. A necessary and sufficient condition for granting this amendment is that the licensee change this methodology in future assessments of the consequences of this accident.
3.3.4 Reactor Coolant Pump Shaft Seizure (Locked Rotor)
The licensee assessed the consequences of a postulated reactor coolant pump shaft seizure (locked rotor) event. The occurrence of such an event could result in fuel failures. In the event of fuel failures, the radioactivity from the fuel would be dispersed to reactor coolant. As a result of primary to secondary leakage, radioactivity would be transferred to the secondary side of the steam generator. Since it is presumed that the event will occur with a subsequent loss of offsite power, activity from the secondary side of the steam generators will be released to the environment via the process of removing the reactors decay heat using the steam generator PORVs.
The licensee assumed that the sources of radioactivity in reactor coolant would be the operating reactor coolant activity levels assumed to be at the technical specification values for dose equivalent 131I, the iodine spike contribution and gap activity due to fuel failures.
The primary to secondary leakage was assumed to mix instantaneously and homogeneously within the secondary side without flashing. The licensees analysis consisted of a determination of (1) that fraction of fuel that reaches or exceeds the initiation temperature of fuel melt, and (2) that fraction of fuel elements for which the fuel clad is breeched. The licensee relied upon departure from nucleate boiling ratio (DNBR) as the fuel damage criterion for estimating fuel damage for the purpose of establishing releases. The licensees analysis assumed 17 breeched assemblies. The licensees submittal indicated that fuel melting would not occur during a locked rotor event.
10 Releases to the environment were assumed to occur until shutdown cooling was initiated thereby terminating releases from the steam generators. Activity released from the steam generators would be a function of the steaming rate and the partition coefficient for the particular nuclide group.
The staff performed confirmatory calculations of the potential consequences of a locked rotor accident based upon information provided in the licensees submittals and in response to staff questions. The assumptions which form the basis for the staff calculations are presented in Table 3.3-3. The staff calculated EAB, LPZ and control room operator doses. The results of these calculations are presented in Table 3.3-5. The staffs calculations confirmed that the Robinsons implementation of AST for the locked rotor accident does not result in doses which exceed the acceptance criteria in Regulatory Guide 1.183.
3.3.5 Single Rod Control Cluster Assembly (RCCA) Withdrawal Appendix H of Regulatory Guide 1.183 provides guidance on the assessment of rod ejection type of accidents. The licensee did not perform an assessment of the consequences of a rod ejection accident. They indicated that such an accident at Robinson does not result in fuel damage and the consequences are bounded by the consequences of other accidents.
The licensee did perform an assessment of the consequences of a postulated single rod control cluster assembly (RCCA) withdrawal accident. The amount of radioactivity released as a result of this accident was based upon the number of fuel rods breeched, the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and the radial peaking factor.
The licensees analysis assumed one fuel assembly had its rods breeched. Three other assemblies were assumed to reach or exceed the initiation temperature for fuel melt.
The degraded fuel assemblies were assumed to release various quantities of isotopes to reactor coolant based upon the extent of the fuel damage. Radioactivity from reactor coolant would enter secondary coolant as a result of primary to secondary leakage. Radioactivity in the secondary side coolant would be released to the environment based upon the steaming rate and the partition coefficient for the particular nuclide group. Primary to secondary leakage was assumed to continue until the primary side pressure was reduced below that of the secondary side. The release of radioactivity from the steam generators was assumed to be terminated when the RHR system was placed in operation.
In Appendix H of Regulatory Guide 1.183, two release paths to the environment were considered. One pathway is via the secondary side through the steam generator power-operated relief valves (PORVs). This pathway is analyzed because all activity is assumed to be released and remain in the reactor coolant system (RCS). Therefore, the only release path is via the secondary system. For this case, no releases from the containment are assumed.
The second pathway is via containment leakage. This pathway occurs because the rod ejection is postulated to result in a loss of the RCS barrier. For this situation, 100% of the
11 activity released from the fuel is assumed to be released into the containment. This would result in the pressurization of the containment to well above normal operating conditions due to the small to medium size LOCA. Therefore, Appendix H of Regulatory Guide 1.183 specifies that releases due to containment leakage are also to be modeled for a rod ejection accident.
These two pathways are to be analyzed separately and there is no summation of the dose results.
The licensees analysis did not incorporate a containment release pathway nor did it use the guidance of Appendix H. The licensees analysis utilized the model in Appendix G (Locked Rotor Accident) of Regulatory Guide 1.183.
The licensee concluded that analysis of the containment leak path for the RCCA withdrawal was unnecessary because primary coolant boundary is expected to remain intact and the only leakage from the reactor coolant system to containment would be minor and would not result in a pressurization of the containment. Therefore, any release from containment would be insignificant. Consequently, the licensee assumed that the only pathway necessary for consideration for release to the environment would be the primary to secondary leak path and the resultant release via steaming through the steam generator power-operated relief valves.
The staff performed confirmatory calculations of the potential consequences of a RCCA withdrawal accident based upon information provided in the licensees submittals and in response to staff questions. The staff also assessed the licensees position that the containment pathway did not need to be evaluated for the RCCA withdrawal accident.
The staffs assessment concluded that the licensee did not need to analyze a RCCA withdrawal accident assuming a containment leak pathway. The assumptions which form the basis for the staff calculations are presented in Table 3.3-4. The staff calculated EAB, LPZ and control room operator doses. The results of these calculations are presented in Table 3.3-5. The staffs calculations confirmed that the Robinsons implementation of AST for the locked rotor accident does not result in doses which exceed the acceptance criteria in Appendix H of Regulatory Guide 1.183.
3.4 Assessment of Control Room Habitability The licensee calculated the control room operator dose for the accidents evaluated in Section 3.3 of this Safety Evaluation. When operated to mitigate the consequences of an accident, the control room emergency ventilation system brings 400 cfm of outside air, passes it through a filter and a charcoal absorber and distributes the air to the control room envelope.
Air from the control room envelope is recirculated back through the filter and the charcoal absorber at a rate of 2600 cfm.
The control room emergency ventilation system is actuated either by a safety injection signal or a signal from a radiation monitor. The licensees analyses for a locked rotor event and for a single RCCA withdrawal assumed that the control room normal ventilation system operated for
12 one hour prior to switching to the control rooms emergency filtration system. For the SGTR accident, it was assumed that the control room emergency ventilation system was initiated 310 seconds following the start of the accident. During the period when the normal control room ventilation system is operating, it is assumed that the normal makeup air to the control room envelope is 400 cfm, which is unfiltered, and that the unfiltered inleakage into the envelope is 300 cfm.
During the periods when the control rooms emergency filtration system is operating, the licensees analyses assumed that unfiltered inleakage into the control room envelope is initially 300 cfm for all accidents except the LOCA. For the LOCA, it is assumed that the unfiltered inleakage is initially at 170 cfm. After one hour, the unfiltered inleakage is assumed to be reduced by 70 cfm when the inleakage from the Hagan Room is reduced. The Hagan Room is a source of unfiltered inleakage into the control room envelope when there is a loss of the Auxiliary Building exhaust fan HVE-7. Loss of this fan results in the Hagan Room, which is adjacent to the control room envelope, being at a higher pressure than the control room envelope. The Hagan Room will remain in this condition until operators can take certain actions which will result in the reduction of the pressure in the Hagan Room to below that of the control room envelope. The licensees analysis assumed that it would take approximately one hour to implement the actions and reduce the pressure in the Hagan Room.
The licensee has performed testing of their control room envelope to establish its inleakage characteristics. A summary of these testing results were presented in an April 10, 2003, letter to the staff. In this letter, the licensee indicated that the inleakage test acceptance criteria are less than 560 cfm (400 cfm unfiltered makeup plus 160 cfm of unfiltered inleakage) when in the normal operating mode; 160 cfm when the control room emergency filtration system is operating in the emergency pressurization mode and the Hagan Room is at a greater pressure than the control room envelope; and less than 90 cfm when the Hagan Room is at a lesser pressure than the control room envelope. The licensee did test the control room envelope to determine its inleakage characteristics when the normal control room ventilation system was operating. The staff will review the licensees response to Generic Letter 2003-01, Control Room Habitability, to determine the manner in which the licensee confirmed the inleakage.
The staff performed calculations to determine whether the licensees implementation of the AST would result in postulated doses which would meet 10 CFR 50.67. The staff calculations confirmed that control room operators doses do meet 50.67 for the accidents analysed in Sections 3.3.2 - 3.3.5.
3.5 Atmospheric Relative Concentration Estimates 3.5.1 Meteorological Data Progress Energy calculated new relative concentration (X/Q) values for the design-basis accident dose (DBA) assessments described above using onsite meteorological data collected during calendar years 1988 through 1996. These data were previously evaluated and are
13 discussed in the safety evaluation report associated with Amendment 195 dated October 4, 2002.
3.5.2 EAB and LPZ Relative Concentration Estimates The licensee calculated X/Q values for the EAB and LPZ using site-specific inputs and the PAVAN computer code. The PAVAN code, documented in NUREG/CR-2858, PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Plants, uses the methodology described in Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. The licensee made calculations for an EAB distance of 425 meters and LPZ distance of 7242 meters. Releases were assumed to be ground level.
The licensee provided EAB X/Q estimates for time periods of longer than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration, but such estimates are not appropriate for use in this EAB dose assessment and they are not approved as part of this license amendment.
3.5.3 Control Room Relative Concentration Estimates Progress Energy used the ARCON96 methodology (NUREG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wake) for calculation of control room X/Q values with a modification to the surface roughness length and averaging sector width constant.
These two modifications are acceptable to the NRC staff. Calculations were made for postulated DBA releases to the control room from the plant stack, closest main steam safety valve/relief valve, closest main steam line, nearest point of the containment building, and residual heat removal heat exchanger room, and to the technical support center/emergency offsite facility from the nearest point of the containment building and residual heat remover heat exchanger room. All releases were assumed to be ground level point releases.
3.5.4 Summary - Atmospheric Relative Concentration Estimate Analysis The NRC staff has reviewed the inputs to the PAVAN and ARCON96 codes and found them to be generally consistent with NRC staff practice, site configuration drawings, and other information provided by Progress Energy. Although the NRC staff is of an opinion that trees may have had an influence on meteorological measurements at the Robinson site in the 1988 through 1996 time period, the NRC staff does not have sufficient basis for concluding that the impact is significant enough to reject the dose assessment for this amendment given the assumptions used in the calculations. Based on this review, the NRC staff finds the X/Q values listed in Table 3.5-1 acceptable for use in this dose assessment.
4.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that approval of the AST amendment request can be approved for all of the above noted accidents except the LOCA and that the results of the above noted accident analyses confirm that the proposed TS changes to re-define Dose Equivalent 131I, to change the allowable levels of Dose Equivalent 131I
14 in primary coolant, and to change the allowable steam generator primary to secondary leak rates are acceptable.
Principal Contributors: L. Brown J. Hayes Dated: September 7, 2004
15 Table 3.3-1 Assumptions for MSLB Accident Parameter Value Iodine & Alkali Metals Partition Factor Faulted Steam Generator Intact Steam Generators 1
0.1 Steam Release from Faulted Steam Generator (lbs)
At time of break 0-2 hours 2-8 hours 8-24 hours 24-53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 53.2-98.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 137,294 SG plus 23,900 Feedwater 161,304 [Includes the above]
330 881 1608 2512 Steam Release from Intact SGs (lbs) 0-2 hours 2-8 hours 8-24 hours 24-53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 53.2-98.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 300,116 561,235 1,110326 1,611,092 0
Primary to Secondary Leak Rate (gpm)
Intact SGs Faulted SG 0.19 0.11 Primary Coolant Activity Level -
Dose Equivalent 131I (µCi/g)
Pre-existing Spike Accident Initiated Spike 60 0.50 Secondary Side Activity (µCi/g)
Dose Equivalent 131I Alkali Metals 0.1 0.1 of Primary Coolant Value Number of Failed Assemblies 2
Total Number of Fuel Assemblies 157 Time before RHR operation (hours) 53.2
16 Time before Primary to Secondary Leak Terminated (hours) 98.8 Steam Generator Mass - Minimum (lbs) 88,641 Primary Coolant System Mass -
Minimum (lbs) 372,137 Maximum Nominal Letdown Flow (gpm) @130oF, 2235 psig 120 Uncertainty Applied to Letdown Flow 10%
Maximum Identified Primary Coolant Leakage (gpm) 10 Maximum Unidentified Primary Coolant Leakage (gpm) 1 Maximum Primary Coolant Mass (lbs) 433,859 Isotopic Equilibrium Appearance Rate (Ci/hr) @Spiking Factor = 500 131I 132I 133I 134I 135I 134Cs 137Cs 138Cs 4162 3914 7766 4742 3805 440 64 2377 Atmospheric Dispersion Factors (sec/m3)
EAB 1.77E-3 LPZ 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 8.92E-5 3.50E-5 2.19E-5 7.95E-6 1.85E-6
17 Control Room (Faulted SG) 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 2.48E-3 1.57E-3 7.05E-4 4.74E-4 3.93E-4 Control Room (Intact SG) 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 2.60E-3 1.65E-3 7.22E-4 4.97E-4 4.01E-4 Breathing Rates (m3/sec)
Offsite 0-8 hours 8-24 hours 1-30 days 3.47E-4 1.75E-4 2.32E-4 Control Room 3.47E-4 Chemical Form of Release to Environment Elemental Organic 0.97 0.03
18 Table 3.3-2 Assumptions for SGTR Accident Parameter Value Partition Factor for Iodine & Alkali Metals Flashed Steam (Faulted Steam Generator)
Non-flashed (Faulted)
Intact Steam Generators 1.0 0.1 0.1 Steam Release from Faulted SG (lbs) 0-0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 0.5-2 hours 95,500 0
Steam Release from Intact SGs (lbs) 0-0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 0-2 hours 0-8 hours 0-24 hours 0-53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 104,641 302,696 871,641 2,002,409 3,650,872 Break Flow to Faulted SG(lbs) 0-0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 0.5-2 hours 131,000 0
Primary to Secondary Leak Rate (gpm)
Intact SGs Faulted SG 0.22 0.08 Faulted SG Isolated (min) 30 Primary Coolant Activity Level -
Dose Equivalent 131I (µCi/g)
Pre-existing Spike Accident Initiated Spike 60 0.25 Flashing Fraction 0-30 minutes 0.3027 Duration of Plant Cooldown by Secondary System (hr) 53.2
19 Steam Generator Mass - Minimum (lbs) 88,641 Primary Coolant System Mass -
Minimum (lbs) 372,137 Maximum Nominal Letdown Flow (gpm) @130oF, 2235 psig 120 Uncertainty Applied to Letdown Flow 10%
Maximum Identified Primary Coolant Leakage (gpm) 10 Maximum Unidentified Primary Coolant Leakage (gpm) 1 Isotopic Equilibrium Appearance Rate @ Spiking Factor = 335 (Ci/hr) 131I 132I 133I 134I 135I 2081 1958 3886 2371 1907 Atmospheric Dispersion Factors (sec/m3)
EAB 1.77E-3 LPZ 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 8.92E-5 3.50E-5 2.19E-5 7.95E-6 1.85E-6 Control Room 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 2.60E-3 1.65E-3 7.22E-4 4.97E-4 4.01E-4
20 Breathing Rates (m3/sec)
Offsite 0-8 hours 8-24 hours 1-30 days 3.47E-4 1.75E-4 2.32E-4 Control Room 3.47E-4 Chemical Form of Release to Environment Elemental Organic 0.97 0.03
21 Table 3.3-3 Assumptions for Locked Rotor Accident Parameter Value Core Thermal Power Level (MWt) 2346 Duration of Plant Cooldown by Secondary System (hr) 53.2 Gap Fraction:
131I 85Kr Other Noble Gases &
Halogens Alkali Metals 0.08 0.10 0.05 0.12 Failed Fuel Assemblies 17 Primary to Secondary Leak Rate (gpm) 0.3 Iodine & Alkali Metals Partition Factor in Steam Generators 0.01 Steam Released from 3 SGs (lbs) 0-2 hours 2-8 hours 8-24 hours 24-53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 301,967 566,768 1,124,996 1,637,910 Maximum Radial Peaking Factor 1.8 Steam Generator Mass - Minimum (lbs) 88,641 Primary Coolant System Mass -
Minimum (lbs) 372,137
22 Atmospheric Dispersion Factors (sec/m3)
EAB 1.77E-3 LPZ 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 8.92E-5 3.50E-5 2.19E-5 7.95E-6 1.85E-6 Control Room 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 2.60E-3 1.65E-3 7.22E-4 4.97E-4 4.01E-4 Breathing Rates (m3/sec)
Offsite 0-8 hours 8-24 hours 1-30 days 3.47E-4 1.75E-4 2.32E-4 Control Room 3.47E-4 Chemical Form of Release to Environment Elemental Organic 0.97 0.03
23 Table 3.3-4 Assumptions for RCCA Withdrawal Accident Parameter Value Core Thermal Power (MWt) 2346 Fuel Defects (No. of Assemblies)
Clad Failure Fuel Melting 1
3 Number of Fuel Assemblies in Core 157 Primary to Secondary Leak Rate (gpm) 0.3 Per cent of Fuel which melts and releases activity to reactor coolant Noble Gases (%)
Iodines (%)
100 50 Per cent of Fuel which melts and releases activity to containment Noble Gases (%)
Iodines (%)
100 25 Iodine & Alkali Metal Partition Factor in the SGs before and after the accident 0.01 Containment Volume (ft3) 1,958,526 Containment Leak Rate (weight
%/day) t = 0-1 day t > 1 day 0.10 0.05 Gap Fraction:
All Iodines All Noble gases Other Halogens Alkali Metals 0.10 0.10 0.10 0.12 Time to Establish Shutdown Cooling (hours) 53.2
24 Steam Released from 3 SGs (lbs) 0-2 hours 2-8 hours 8-24 hours 24-53.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 301,967 566,768 1,124,996 1,637,910 Maximum Radial Peaking Factor 1.8 Steam Generator Mass - Minimum (lbs) 88,641 Primary Coolant System Mass -
Minimum (lbs) 372,137 Atmospheric Dispersion Factors (sec/m3)
EAB 1.77E-3 LPZ 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 8.92E-5 3.50E-5 2.19E-5 7.95E-6 1.85E-6 Control Room (PORVs) 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 2.60E-3 1.65E-3 7.22E-4 4.97E-4 4.01E-4 Control Room (Containment) 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days 4.15E-3 2.74E-3 1.17E-3 8.18E-4 6.74E-4
25 Breathing Rates (m3/sec)
Offsite 0-8 hours 8-24 hours 1-30 days 3.47E-4 1.75E-4 2.32E-4 Control Room 3.47E-4 Chemical Form of Release to Environment Elemental Organic 0.97 0.03
26 Table 3.3-5 Robinson Dose Consequences (TEDE) in Rem Accident Exclusion Area Boundary Low-Population Zone Control Room Operators Main Steamline Break Pre-existing Spike (Acceptance Criteria)
Accident-initiated Spike (Acceptance Criteria)
With Fuel Damage (Acceptance Criteria) 0.10 25
.63 2.5 1.3 25 0.025 25 0.12 2.5 0.29 25 0.085 5
.45 5
0.98 5
Steam generator Tube Rupture Pre-existing Spike (Acceptance Criteria)
Accident-initiated Spike (Acceptance Criteria) 19 25 2.4 2.5 0.97 25 0.12 2.5 2.72 5
.24 5
Locked Rotor (Acceptance Criteria) 0.42 2.5 0.035 2.5 1.6 5
Rod Control Cluster Assembly Withdrawal Secondary Side Release Path Containment Release Pathway (Acceptance Criteria) 2.41 6.3 0.184 6.3 0.902 5
27 Table 3.5-1 Robinson Relative Concentration (X/Q) Values Offsite X/Q Values (s/m3)
EAB Limiting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> interval 1.77 E-3 LPZ 0 - 2 hrs 8.92 E-5 2 - 8 hrs 3.50 E-5 8 - 24 hrs 2.19 E-5 1 - 4 days 7.95 E-6 4 - 30 days 1.85 E-6 Control Room and TSC/EOF X/Q Values (s/m3)
Release -
Receptor Pair 0-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days Plant Stack - CR 1.24E-03 8.97E-04 3.62E-04 2.58E-04 2.14E-04 Closest MSSV/RV - CR 2.60E-03 1.65E-03 7.22E-04 4.97E-04 4.01E-04 Closest Main Steam Line - CR 2.48E-03 1.57E-03 7.05E-04 4.74E-04 3.93E-04 Containment Nearest Point - CR 4.15E-03 2.74E-03 1.17E-03 8.18E-04 6.74E-04 Containment Nearest Point -
TSC/EOF 1.64E-04 1.43E-04 6.49E-05 4.41E-05 3.50E-05 RHR Heat Exchanger Room -
CR 7.13E-03 5.49E-03 2.29E-03 1.71E-03 1.37E-03 RHR Heat Exchanger Room -
TSC/EOF 1.38E-04 1.23E-04 5.52E-05 3.78E-05 3.01E-05 FHB Wall - CR 1.34E-03 1.02E-03 4.31E-04 3.21E-04 2.56E-04 CR - Control Room FHB - Fuel Handling Building MSSV/PORV - Main Steam Safety Valve/Relief Valve RHR - Residual Heat Removal TSC/EOF - Technical Support Center/Emergency Offsite Facility