ML20049H292
| ML20049H292 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230081 | |
| Download: ML20049H292 (200) | |
Text
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O a
i APPENDIX 15B BWR/6 GENERIC ROD WITIIDRAWAL ERROR ANALYSIS I
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l GESSAR II 22A7007 i
238 NUCLEAR ISLAND Rev. O APPENDIX 15B CONTENTS 4
Section Title Page 1
58.1 INTRODUCTION
15B.1-1 15B.2
SUMMARY
AND CONCLUSIONS 15B.2-1 1
15B.3 RWL SYSTEM OVERVIEW 15B.3-1 15B.3.1 RWL and RBM System Comparison 15B.3-1 15B.3.2 RWL System Operational Description 15B.3-2 l
15B.3.3 Why Replace the RBM System?
15B.3-2 15B.3.4 Generic Versus Plant Specific RWE Analyses 15B.3-3 15B.3.5 Applicability of the Generic Analysis 15B.3.4 15B.4 GENERIC RWE ANALYSIS 15B.4-1 15B.4.1 Introduction 15B.4-1 15B.4.2 RWE Design Criteria 15B.4-1
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15B.4.2.1 RWE Design Criteria Statement 15B.4-1 15B.4.2.2 Procedure for Satisfying RWE I
MCPR Design Criteria:
Overview 15B.4-2 l
15B.4.3 Detailed Procedure Discussion 15B.4-3 15B.4.3.1 Block Diagram 15B.4-3 15B.4.3.2 Projected BWR/6 IMCPR Capability 15B.4-4 i
15B.4.3.2.1 Base Rod Patterns - Step 1 15B.4-4 l
15B.4.3.2.1.1 Rod Pattern Development 15B.4-4 15B.4.3.2.1.2 IMCPR Database 15B.4-5 15B.4.3.2.1.3 IMCPR Database Biases 15B.4-6 l
15B.4.3.2.2 Statistical Evaluation of IMCPR - Step 2 15B.4-7 15B.4.3.2.2.1 Engineering Model 15B.4-7
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15B.4.3.2.2.2 Expected IMCPR Over the l
Operating Map 15B.4-9 15B.4.3.2.3 IMCPR95/50 Capability Curve -
Step 3 15B.4-9 15B.4.3.2.4 Operating Plant Data Comparison to the Projected BWR/6 15B.4-ll IMCPR95/50 l
15B-i l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)
Section Title Page 15B.4.3.2.5 Summary 15B.4-12 15B.4.3.3 Ganged Rod Withdrawal AMCPR
Response
15B.4-12 15B.4.3.3.1 Ganged Rod Withdrawal Simulation - Step 4 15B.4-13 15B.4.3.3.2 Statistical Evaluation of AMCPR/
IMCPR - Step 5 15B.4-14 15B.4.3.3.2.1 AMCPR Database 15B.4-14 15B.4.3.3.2.2 AMCPR Database Biases 15B.4-15 15B.4.3.3.2.3 Engineering Model 15B.4-15 15B.4.3.3.3 Maximum Allowable Control Rod Withdrawal Distance - Step 6 15B.4-17 15B.4.3.3.4 Optimum Withdrawal Increments 15B.4-18 15B.4.3.3.5 IMCPR Technical Specification as a Function of Core Power 15B.4-21 15B.4.3.3.6 Rod Movement Restriction Technical Specification 15B.4-21 15B.4.3.3.7 Comparison of Generic and Deterministic RWE Analyses 15B.4-22 15B.5 U. CERTAINTIES 15B.5-1 15B.S.1 Introduction 15B.5-1 15B.5.2 IMCPR Uncertainties and Biases 15B.5-1 15B.5.3 AMCPR Uncertainties and Biases 15B.5-2 15B.5.4 Statistical Procedure Conservatisms 15B.5-4 15B.6 GENERIC ANALYSIS CONSERVATISMS 15B.6-1 15B.6.1 Introduction 15B.6-1 15B.6.2 IMCPR Conservatisms 15B.6-1 15B.6.3 AMCPR Conservatisms 15B.6-2 15B.6.3.1 High Worth Control Rod Gang Withdrawals 15B.6-2 15B.6.3.2 Single Rod Withdrawals 15B.6-4 15B.6.3.3 AMCPRs at Maximum Core Flow 15B.6-5 15B.6.3.4 Constant Xenon During Rod Withdrawal 15B.6-6 O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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CONTENTS (Continued)
Section Title Page 15B.7 SPECIAL STUDIES 15B.7-1 15B.7.1 Introduction 15B.7-1 15B.7.2 Xenon Distribution Study 15B.7-1 15B.7.3 Rod Sequence Exchange Study 15B.7-2 l
15B.8 MLHGR CONSIDERATIONS 15B.8-1 I
15B.9 REFERENCES 15B.9-1 15BA ATTACHMENT A TO APPENDIX 15B - IMCPR DATABASE 15BA-1 15BB ATTACHMENT B TO APPENDIX 15B - IMCPR DATABASE CROSSPLOTS 15BB-1 15BC ATTACHMENT C TO APPENDIX 15B - OPERATING PLANT IMCPR DATA 15BC-1 15BD ATTACHMENT D TO APPENDIX 15B - AMCPR DATABASE 15BD-1 15BE ATTACHMENT E TO APPENDIX 15B - AMCPR DATABASE CROSSPLOTS 15BE-1 15BF ATTACHMENT F TO APPENDIX 15B - RCIS POWER SIGNAL 15BF-1 15BF.1 Introduction 15BF.2 Effect of Biased Power Signal on RCIS System Function 15BF-3 15BF.3 RCIS Power Signal Source 15BF-5 15BF.3.1 Off-Normal Causes of Power Signal Bias 15BF-8 15BF.3.1.1 Safety Relief Valve Steamflow 15BF-8 15BF.3.1.2 Extraction Upstream of Turbine Stop Valves 15BF-8 15BF.3.1.3 Turbine Extraction Steam 15BF-9 15BF.3.1.4 Bypassed Steamflow 15BF-10 15BF.3.1.4.1 Deliberate Operator Action 15BF-10 15BF.3.1.4.2 Normal Operating Transients 15BF-12 15BF.3.1.4.3 Abnormal Operating Transients 15BF-12 15BF.4 Consequences of Technical Specification Violation 15BF-15 15BF.5 First Stage Turbine Pressure Signal Instrumentation 15BF-17 15BF.6 Control Room Indication of
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Reactor Power / Bypass Valves 15BF-19 15BF.7 RC&IS Setpoint Selection 15BF-21 15BF-23 15BF.8 Summary 15B-iii/15B-iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O APPENDIX 15B C)
TABLES Table Title Page 15B-1 Sequence of Events - RWE in Power Range-for BWR/2-5 15B.10-1 15B-2 BWR/6 RWE Transient Sequence of Events 15B.10-2 15B-3 BWR/6 Standard Plants Incorporated in the IMCPR Database 15B.10-3 15B-4 Typical Results From the Statistical Fit of IMCPR 15B.10-4 15B-5 Typical Values of ( MCPR/IMCPR) 95/95 and MCPR 95/95 15B.10-5 15B-6 IMCPR Technical Specification Requirements for Optimum Withdrawal Distances 15B.10-6 15B-7 Typical Results of BWR/6 Plant-Specific Deterministic RWE Analyses 15B.10-7 15B-8 Typical BWR/6 MCPR Response to the With-drawal of High Worth and Low Worth Control Rod Gangs 15B.10-8 O
15B-9 Comparison of AMCPRs for Single and Ganged Rod Withdrawals 15B.10-9 15B-10 Nominal RWE Response at Selected Power / Flow States 15B.10-10 ISB-ll Impact of Xenon Distribution on IMCPR 15B.10-ll 15B-12 AMCPRs for Gang Rod Withdrawals During Rod Sequence Exchanges 15B.10-12 15B-13 LUGR Performance Limit Margin Reduction Summary 15B.10-13 15BF-1 Consequenceq of RWL Malfunction 15BF-25 L
b ILLUSTRATIONS i
Figure Title Page l
15B-1 Generic Procedure Flow Diagram 15B.10-15 i
15B-2 IMCPR Datapoint Locations 15B.10-16 j
15B-3 Nominal IMCPR as a Function of Database Size 15B.10-17 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued)
O Figure Title Page 15B-4 Probability Plot for the IMCPR Data Base 15B.10-18 15B-5 Nominal IMCPR as a Function of Power and 15B.10-19 Flow 15B.10-20 15B-6 BWR/d IMCPR(P)95/50 Curve 15B-7 Comparison of Operating Plant Data and 15B.10-21 Projected IMCPR Capability 15B-8 Typical MCPR Response to a Ganged Rod 15B.10-22 Withdrawal 15B-9 Location of AMCPR Datapoints 15B.10-23 15B-10 Nominal AMCPR/IMCPR as a Function of Database Size 15B.10-24 15B.10-25 ISB-ll AMCPR Relationship to IMCPR 95/95 as a Function of Withdrawal 15B-12 AMCPR 15B.10-26 Distance at 100% Power 15B-13 Maximum Allowable Withdrawal Distance 15B.10-27 15B-14 Rod Block Setpoints as a Function of Core 15B.10-28 Power 15B-15 Rod Withdrawal Limiter IMCPR Technical Specificatio" as a Function of Core Power 15B.10-29 15B-16 Comparison of AMCPR Frequency Distributions for a Random Sample and Biased Database 15B.10-30 Rod Pattern Prior to Gang Withdrawal During 15B-17 A Rod Sequence Exchange 158.10-31 15B-18 Margin Reduction as a Function of Distance 158.10-32 Withdrawn 15BB-1 Crossplot of Fit Residual Versus Core Power 15BB-1 15BB-2 15BB-2 Crossplot of Fit Residual Versus Core Flow 15BB-3 15BB-3 Crossplot of Fit Residuals Versus Plants 15BB-4 Crossplot of Fit Residuals Versus Cycle 15BB-4 Average Exposure 15BB-5 15BB-5 Crossplot of Fit Residuals Versus Cycle 15BB-6 Crossplot of Fit Residuals Versus Rod 15BB-6 Sequence 15BB-7 Crossplot of Fit Residuals Versus Exposure 15BB-7 Interval 15BB-8 Crossplot of Fit Residuals Versus Core 15BB-7 Average Enrichment ISB-vi
1 GESSAR II 22A7007 j
238 NUCLEAR ISLAND Rev. O d
ILLUSTRATIONS (Continued)
Figure Title Page j
j 15BE-1 Crossplot of Fit Residuals Versus Core Power 15BE-1 15BE-2 Crossplot of Fit Residuals Versus Core Flow 15BE-2 l
15BE-3 Crossplot of Fit Residuals Versus Plant 15BE-3 15BE-4 Crossplot of Fit Residuals Versus Cycle Average Exposure 15BE-4 15BE-5 Crossplot of Fit Residuals Versus Distance j
Withdrawn 15BE-5 j
j 15BE-6 Crossplot of Fit Residuals Versus Exposure l
Interval 15BE-6 j
15BE-7 Crossplot of Residuals Versus Rod Sequence 15BE-6
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15BF-1 RCIS System Function 15BF-26 l
15BF-2 First Stage Turbine Pressure 15BF-27 1
j 15BF-3 Simplified Turbine Steam Design 15BF-29 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O j
ABSTRACT
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This appendix describes a generic method of analyzing the BWR/6 Rod Withdrawal Error (RWE) transient from power levels greater than 20% of rated.
The method employs a statistical evaluation of the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) response to the withdrawal of ganged control rods for both rated and off-rated conditions.
The MCPR response, which is a func-tion of withdrawal distar.:c, is combined with the safety limit MCPR to establish ar. operating limit MCPR technical specification and associated rod withdrawal distances that ensure, with a high degree of confidence, that the design basis AMCPR will not be violated.
The design basis AMCPR is the difference between the operating limit MCPR and the safety limit MCPR (i.e.,
the RWE transient thermal margin loss).
For the above withdrawal distances, it is also veri-fied that the design basis ALHGR is not exceeded.
The design
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basis ALHGR is the difference between the LHGR corresponding to 1% plastic strain and fuel licensing basis LHGR perform-ance limits.
7he withdrawal distances which are specified as design requirements for the Rod Withdrawal Limited (RFL) system as a function of core power are 2 ft between 20% and 70% power and 1 ft above 70% power.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.1 INTRODUCTION O
The consequ'ences of an operator initiated Rod Withdrawal Error (RWE) on BWR/6 are mitigated by the Rod Withdrawal Limiter (RWL)
System, a subsystem of the Rod Control and Information System (RCIS).
The RWL System restricts control rod motion to fixed displacements as a function of core power for each rod or gang selection.
This report deals exclusively with a description of the methodology and generic evaluations that establish the bases for the core physics design requirements for these fixed displacements.
On pre-BWR/6 product offerings, the Rod Block Monitor (RBM) System blocks inadvertent rod withdrawals based on local neutron flux increases measured by Local Power Range Monitors (LPRM).1 Instru-ment response and uncertainty, plant-dependent MCPR operating
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limits, and plant-unique core loadings complicate the application of a generic approach for these plants.
Therefore, a plant and cycle specific worst-case deterministic analysis is currently per-formed for all licensing submittals.
With the introductice of the standard plant concept and the RWL system, a generic BWR/6 RWE analysis is developed.
For this anal-ysis, anticipated operating states (including rod patterns) are examined for several cycles in a BWR/6.
An analysis of the resulting initial minimum critical power ratios (IMCPR) permits the designation of IMCPR values as a function of power that will not be violated with 95% probability /50% confidence.
This represents the projected MCPR capability of BWR/6.
The change in MCPR (AMCPR) is calculated for ganged rod withdrawal events which are initiated from reactor states (rod patterns) among those used to determine the IMCPR distribution.
These AMCPR values at 95% probability /95%
confidence are then subtracted from the 95% probability /50% confi-
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dence MCPR to obtain final MCPR values for comparison to the safety limit.
This allows the maximum withdrawal increments to be 15B.1-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.1 INTRODUCTION (Continued) determined.
A cost / benefit analysis considering both the operational and safety impact is performed to establish the optimum withdrawal increments of 2 ft between 20% and 70% power and 1 ft above 70%
Combining the safety limit and the 95% probability /95% con-power.
fidence MCPR response corresponding to these optimum withdrawal increments results in the final MCPR technical specification to protect against an inadvertent RWE.
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15B.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i
15B.2
SUMMARY
AND CONCLUSIONS The following conclusions are drawn from the generic analysis of the BWR/6 RWC transient above 20% of rated power:
(1)
Between rated (100%) and 70% power and between 70% and l
20% power, the RWL rod block setpoints are 1.0 ft and l
2.0 ft, respectively.
This is based on 95% probability /
95% confidence that the design basis AMCPR (AMCPR DB l
design basis 6LHGR (ALHGR no e violated.
The i
DB Rod Pattern Control System (RPCS) mode of the RCIS j
asrures adequate RWE protection below 20% power, where the RWL mode of the PCIS is not operational.
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(2)
This analysis demonstrated that MCPR was the most limit-t ing safety criterion for this event.
The LHGRs, after a j
ganged control rod withdrawal, were shown not to exceed l
the values corresponding to the 1% plastic strain limit on the cladding.
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(3)
Comparisons between the statistical approach presented in this report and the deterministic approach currently employed for FSAR submittals showed good MCPR response agreement.
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The rod block distances were selected such that there was adequate operating margin between the expected i
operating MCPR performance and the technical specifica-l l
tion operating limit MCPR established for the RWE.
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Figures 15B-14 and 15B-15 were the final outputs of this generic analysis that require implementation into i
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f respectively.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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15B.3 RWL SYSTEM OVERVIEW 15B.3.1 RWL and RBM System Comparison The RWL System proposed for BWR/6 performs the same function as the RBM System on earlier DWR product lines (i.e., blocking an inadver-tent rod withdrawal such that the RWE design criteria are not violated).
However, the hardware and operations of the two systems are significantly different.
A brief discussion of these dif-forences follows.
The sequence of events prior to rod block for the RBM System is described in Table 15B-1.
Basically, the system consists of two redundant RBM channels, each receiving input signals from up to eight LPRMs surrounding the selected control rod.
Each RBM channel signal is the average of the input LPRM signals.
Prior to rod withdrawal, each RBM channel reading is normalized to an assigned b,
Average Power Range Monitor (APRM) channel reading.
After normali-zation and subsequent rod withdrawal, a rod block occurs if either RBM channel signal reaches a preset rod block trip setpoint.
The dual-channel RWL System does not require direct core response feedback from the LPRMs during a rod withdrawal.
Instead, the distance a rod is withdrawn is monitored by position indicator switches situated along the control rod drive (CRD) mechanism.
The relationship between withdrawal distance and the margin to fuel safety limits is analytically determined.
The allowable rod with-drawal distance as a function of core power is set such that there is a high degree of confidence the AMCPR an ALHGR re not DB DB violated.
When an operator attempts to withdraw a rod further than the prespecified withdrawal distance, the rod is blocked.
The time sequence of events for a BWR/6 RWE transient is given in Table 15B-2.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.3.2 RWL System Operational Description The RWL System blocks rod withdrawals at prespecified, power-dependent increments.
This system is operational between the Low 5
(20+0% of rated power) and rated power.
Power Setpoint (LPSP)
[Below the LPSP, rod pattern restrictions are enforced by the Banked Position Withdrawal Sequence (BPWS)].
A High Power Setpoint (HPSP) is established at (70% core power.
Between the LPSP and the IIPSP, rod withdrawals are limited to 2 ft, while between the HPSP and rated power, withdrawals are limited to 1 ft.
These withdrawal restrictions were established by the generic BWR/6 RWE analysis discussed in this appendix.
15B.3.3 Why Replace the RBM System?
The concept of ganged control rods was introduced with BWR/6.
A gang consists of a maximum of four control rods that can be selected and withdrawn simultaneously.
This improves plant startup times and fuel performance, since symmetrical radial power shapes can be maintained during power changes.
The advantages of ganged rod withdrawals complicate a RBM System approach.
Instead of monitoring the neutron flux increase around a single rod, it would be necessary to monitor the response around up to four rods.
This would require the analysis of up to 64 LPRM input signals.
The proper treatment of all combinations of instru-ment failure and response would require a more complex hardware system and supporting analysis than on pre-BWR/6 plants.
Since the RWL System blocks strictly on incremental distance with-drawn, all concerns relative to LPRM instrument response are eliminated.
In addition, the required hardware logic is simplified.
The two systems are equivalent from an analytical standpoint, since setpoints are based on calculated AMCPRs in both cases.
Although 15B.3-2 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I
15B.3.3 Why Replace the RBM System? (Continued)
V) the RBM system makes use of relative neutron flux measurements via the LPRMs, AMCPR is not directly measured.
In addition, the RBM rod block trip setpoint is derived by calculating both AMCPR and the associated instrument response introducing additional potential error sources.
15B.3.4 Generic Versus Plant Specific RWE Analyses On pro-BWR/6 product lines, a single plant and cycle-specific deterministic RWE analysis is performed.
The reactor core bchavior during the RWE transient is modeled by a series of steady-state three-dimensional coupled nuclear thermal-hydraulics calcula-tions using the three-dimensional BWR core simulator code.3 The analysis consists of:
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(1) selecting a control rod pattern at rated conditions which places bundles near the fully inserted high worth control rod on or near thermal limits, MCPR and MLHGR (zero xenon inventory is conservatively assumed, allowing greater rod pattern flexibility in putting the core on limits near the fully inserted error rod; this results in overly conservative LHGR responses when compared to typical operating rod patterns);
(2) incrementally withdrawing the high worth control rod to determine the AMLHGR and AMCPR response; and (3) calculating the final RBM rod block trip setpoint from simulated LPRM response data to assure fuel safety limits are not violated.
The BWR/6 standard plants with standard operating MCPR limits, core and fuel designs and loading patterns ensure similar RWE
%J 15B.3-3 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.3.4 Generic Versus Plant Specific RWE Analyses (Continued) transient responses.
Thus, a generic analysis based on a statistical evaluation of the AMCPR and MLHGR response to a rod withdrawal can be developed.
The primary advantage of the generic approach is that many, if not all, operating conditions as a func-tion of rod pattern, operating power and flow, core average exposure, etc., can be evaluated; whereas, the deterministic approach only evaluates a few selected limiting operating states.
An additional advantage is the decrease in both the licensing sub-mittal preparation and review process on future fuel loadings.
The computer methods used to simulate the RWE transient are identi-cal in both the generic and plant specific analyses.
However, base rod pattern assumptions and rod withdrawal strategies differ.
These differences are discussed in ensuing sections.
15B.3.5 Applicability of the Generic Analysis The RWL generic analysis described in this appendix is applicable to the BWR/6 standard plants (218-624, 238-748, 251-848) and the nonstandard 251-800 plant for the following conditions:
(1) initial, transition and equilibrium cycles; (2) reload batch sizes to 40%;
(3) bundle average enrichments up to 3.25 w/o U-235; (4) up to and including 18-month cycles; (5) the power / flow region of Figure 15B.4-2; and (6) core average discharge expos.res <25800 mwd /t and 40000 mwd /t peak pellet exposure.
15B.3-4
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GESSAR II 22A7007 j
238 NUCLEAR ISLAND Rev. O I
15B.3.5 Applicability of the Generic Analysis (Continued)
)
The generic analysis is not applicable to a control cell core loading strategy or high energy, high discharge exposure cycles (i.e., 33,000 mwd /t core average discharge exposure and 50,000 mwd /t peak pellet exposure).
As new design features are implemented, compliance checks will be performed and documented to demonstrate I
the applicability of the generic analysis.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f))
15B.4 GENERIC RWE ANALYSIS 15B.4.1 Introduction The generic RWE analysis for BWR/6 is significantly different from the plant-specific RWE analysis in that:
(1) many calculational cases at both rated and off-rated conditions are analyzed to estab-lish an adequate data base for the statistical calculations; (2) typical rod patterns, not limiting rod patterns, are assumed prior to rod pulls (a limiting rod pattern is one in which at least one bundle within the 36 bundles around the fully inserted high worth rod is on thermal limits); (3) the AMCPR response to ganged rod withdrawals is statistically determined; and (4) rods are blocked on distance rather than RBM response.
As stated in the previous section, however, both analyser utilize the same three-dimensional BWR core simulator code.
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Although both MCPR and LHGR are of interest with. respect to the RWE transient, it was determined that the MCPR was most limiting, and, hence, the rod withdrawal distances will be established based on MCPR.
Therefore, the detailed procedure discussion which follows will primarily concentrate on MCPR.
The LHGR response to a RWE transient will be addressed separately in Section 15B.8 to demon-strate that the selected RWL setpoints based on MCPR also ensure that the LHGR design criterion is not exceeded.
15B.4.2 RWE Design Criteria 15B.4.2.1 RWE Design Criteria Statement The BWR/6 RWE protection criteria can be stated as follows:
The design basis AMCPR for RWEs initiated from the technical specification operating limit and mitigated by the RWL sys-(\\
tem withdrawal restrictions shall be determined such that
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there is 95% probability at the 95% confidence level that 15B.4-1
GESSAR II 22A7007 238 WUCLEAR ISLAND Rev. 0 15B.4.2.1 RWE Design Criteria Statement (Continued) any randomly occurring RWE (initiated on linits) will not result in a larger AMCPR.
In addition, it shall be estab-lished that RWL System withdrawal restrictions set by the above criterion shall not exceed the design basis ALHGR with 95% probibility at the 95% confidence level.
The design basis ALHGR is the difference between LHGRs corre-sponding to 1% plastic strain and fuel licensing basis LHGR performance limits.
15B.4.2.2 Procedure for Satisfying RWE MCPR Design Criteria:
Overview The AMCPR s
1 ence between the RWE required technical DB specification operating limit MCPR and the safety limit MCPR.
The 95% probability /95% confidence values of AMCPR are obtained from a statistical evaluation of the parameter AMCPR/IMCPR as a function of core power, core flow and incremental rod withdrawal distance.
The effects of process instrumentation uncertainties and biases enter twice.
First, the safety limit MCPR analysis includes the effects of the pre-RWE IMCPR monitoring accuracj and licensing basis fuel performance LHGR limits include the effects of monitoring the pre-RWE LHGR distribution.
Second, process instrumentation uncertainties and biases (primarily core power, core flow, and TIP measurements) and calculational uncertainties and biases for IMCPR and AMCPR are considered to establish the "true" AMCPR.
BWR/2-5 plant-specific analyses near rated conditions determine an absolute operating limit MCPR that a plant must achieve to be protected against moderately frequent transient events, as well as the RWE transient with the chosen RBM rod block trip setpoint.
The trip setpcints that result from this single analysis near rated conditions can give unnecessary rod blocks at lower powers.
To 15B.4-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O,'
15B.4.2.2 Procedure for Satisfying RWE MCPR Design Criteria:
Overview (Continued) achieve the minimum withdrawal ~ distance necessary for operational fleribility, the trip setpoint must sometimes be chosen such that the RWE transient is the most limiting.
On BWR/6, the power-dependent allowable withdrawal distances can be varied to reflect an operating limit MCPR that is a function of core power.
To optimize the withdrawal distance at less than rated power, AMCPR f r the RWE is based on a power-dependent DB operating limit MCPR which is easily achievable and not restrictive to lower power operation.
A representative sample of expected BWR/6 operating conditions (i.e.,
rod patterns, core exposure, fuel cycle, enrichment, core size, etc.) was included in the statistical evaluation of the MCPR
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response.
An additional study of the MCPR response to a RWE during a rod sequence exchange was also performed.
Combined with signifi-cant conservatism in the analysis (Section 15B.6), there is a high probability and confidence that any random withdrawal error will not violate the RWE design criteria.
15B.4.3 Detailed Procedure Discussion 15B.4.3.1 Block Diagram A block diagram of the generic analysis is shown in Figure 15B.4-1.
In Step 1, base rod patterns are developed at various power and flow conditions over a reference BWR/6 operating map for various fuel cycles and exposure points within a given fuel cycle.
The resulting (pre-RWE) IMCPRs are fitted versus power and flow with process instrumentation and calculational uncertainties and biases included (Step 2).
Step 3 establishes the 95% probability /50%
~N confidence IMCPR curve (IMCPR95/50)
(O sa function of core power.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.1 Block Diagram (Continued)
This curve represents the projected IMCPR capability for BWR/6.
The 50% confidence level is adequate, since the RWL withdrawal increment setpoints are established by the difference between technical specification MCPR (IMCPRTS) and the MCPR safety limit.
Thus, IMCPR is unrelated to plant safety from a RWE transient standpoint.
The AMCPR response to a RWE transient is evaluated in Steps 4 and 5.
Step 4 simulates the withdrawal of high worth ganged rods from the base rod patterns of Step 1.
The AMCPR/IMCPR response is fitted as a function of power, flow and distance withdrawn in Step 5.
(St p 3) and the AMCPR/IMCPR Next, Step 6 combines the IMCPR95/50 response (Step 5) to set the maximum allowable withdrawal distances as a function of power.
The AMCPR used is the value that will not be exceeded with 95% probability at the 95% confidence level.
In Step 7, the optimum withdrawal distances are established by considering the minimum rod maneuverability requirements, startup time penalties and the impact on rod sequence exchanges.
With the withdrawal distances fixed and the AMCPR response known (Step 5),
the IMCPR as a function of power required to provide the optimum TS withdrawal distances is finalized in Step 8.
The IMCPR is com-TS pared to the IMCPR t
nsure that operating requirements are 95/50 not overly restrictive.
15B.4.3.2 Projected BWR/6 IMCPR Capability 15B.4.3.2.1 Base Rod Patterns - Step 1 15.4.3.2.1.1 Rod Pattern Development The rod pattern search module of the three-dimensional BWR simulator code was used to establish rod patterns over the power and flow map 15B.4-4 l
CESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15D.4.3.2.1.1 Rod Pattern Development (Continued) with ' equilibrium xenon distributions consistent with the power /
1 flow operating state.
The only constraints were: (1) MCPR >
i operating limit MCPR at rated conditions (1.20); (2) MLHGR >
13.4 kW/ft; and (3) the calculated neutron multiplication factor equal to the projected critical k 0.005.
Once these three eff constraints were satisfied, no further attempt was made to flatten the radial power shape to increase margins to thermal limits.
- Thus, at low powers the rod search module is essentially unconstrained and the rod patterns were not optimized to achieve favorable IMCPR and AMCPR performance.
i Projected through-the-cycle rod patterns (1000 mwd /t intervals) at rated conditions were the basis for the history-dependent exposure distributions, as opposed to worst-case distributions.
These exposure distributions had been optimized to meet thermal margins I
and reactivity requirements.
The rod patterns were consistent
(
with current BWR operating philosophy.
i The rod pattern search module was initialized at random exposure points and power / flow conditions.
A nonoptimized initial rod pattern was input to the rod pattern search module, and the margin 4
to constraints was checked to determine if any rod pattern adjust-l ments were required.
The final, nonoptimized rod pattern repre-i sented expected short-term MCPR capability, assuming the core accumulated the greatest portion of its exposure with optimized 4
rod patterns.
This is consistent with normal operations wherein j
many rod pattern adjustments are made at low powers as the core is maneuvered up to rated power, with the major exposure accumula-l tion at steady-state, near rated conditions.
15B.4.3.2.1.2 IMCPR Database The output of Step 1 is a database consisting of sampled values of IMCPR at varicus power and flow conditions (Figure 15B-2 and i"
15B.4-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.2.1.2 IMCPR Database (Continued)
O Attachment A).
A mixture of core sizes, cycles, cycle exposures and rod sequences are included (Table 15B-3).
A total of 71 data-points are included in the database.
Biases introduced by the various parameters are discussed in Subsection 158.4.3.2.1.3.
To establish that the statistically determined value of IMCPR as a function of power and flow is converged (i.e., the database is of sufficient size), the statistical model (Subsection 15B.4.3.2.2) was updated perid'dically to incorporate additional data.
Fig-ure 15B-3 shows the fluctuations in the nominal value of IMCPR at several power and flow conditions ao a function of the number of datapoints.
No core parameter or calculational uncertainties or biases were considered in these sampled IMCPR values.
As the num-ber of datapoints increased, the nominal IMCPR approached constant values.
As more data were added, these values tended to oscillate.
Once this oscillation was obtained at all power and flow conditions
()SS datapoints in Figure 15B-3), the database was considered complete.
15B.4.3.2.1.3 IMCPR Database Biases Biases associated with the variable core parameters (i.e.,
core size, core average exposure, fuel cycle, rod sequence and core average enrichment) were qualitatively evaluated using crossplots of residuals from the statistical fit of the IMCPR database (Subsec-tion 15B.4.3.2.2) versus the subject parameter (Attachment B).
The fit residual is the difference between the observed value and the corresponding estimate of the mean divided by the corresponding estimate of the standard deviation.
Crossplots provide in formation on how well the assumed relationship fits the data and how variables in the model affect a dependent variable.
Lack of fit (i.e.,
random scatter) in a crossplot indi-cates that no biases exist in the database.
Note that all cross-plots of Attachment B, except Figure 15BB-1, are random scatter 15B.4-6
GESSAR II 22A7007
]
238 NUCLEAR ISLAND Rev. O I
15B.4.3.2.1.3 IMCPR Database Biases (Continued) plots and thus do not reveal any obvious biases.
In Figure 15BB-1, I
however, the fit appears skewed between 20% and 50% power.
The fit underpredicts the IMCPR data at 20% power and overpredicts it between 30% and 50% power (bottom end of the rated flow control
~i line).
This is believed to be a result of the rod pattern search module of the three-dimensional BWR simulator.
The skew at 50%
power is %0. 70 in IMCPR, not large enough to present a safety con-
)
corn.
However, this could introduce an operational concern if the maximum allowable distances were used.
As will be seen in Subsec-tion 15B.4.3.3.4, the rod block distances are significantly less J
than the maximum allowable withdrawal distances below 50% power.
i 15B.4.3.2.2 Statistical Evaluation of IMCPR - Step 2 4
15B.4.3.2.2.1 Engineering Model Before the sampling IMCPR database was statistically evaluated, it was renormalized to include known biases and uncertainties between analytical prediction methods and operating plant core monitoring instrumentation.
The details of this database normalization are given in Section 15B.5.
Once the IMCPR data base was normalized, an empirical mathematical fit which correlates the IMCPR as a function of power and flow was developed.
A general program for analyzing data and fitti.ig sta-tistical models was utilized."
The program uses standard regres-sion analysis techniques to determino model coefficients and the standard error and confidence level of the fitted parameter.
The statistical program does not have the capability to automatic-ally select a "best" model.
The user must interpret the results from the statistical tools provided.
These tools include data O
15B.4-7
GESSAR TI 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.2.2.1 Engineering Model (Continued) plotting routines, probability plots, crossplots, correlation coefficients, estimated model parameters, percentile estimates, standard errors and confidence intervals.
Using the above tools, it was found that the IMCPR is accurately represented as a quadratic function of percent core power (P) and percent core flow (F) with a variable standard error, i.e.:
P) + F(C3 + C F) + C PF + C (15B.4-1)
I 1+
2 II 4
5 6
IMCPR F
(15B.4-2)
ER C7+CP+
9
=
8 IMCPR l'IMCPR represents the mean; represents de standad where IMCPR error of the IMCPR distribution; and
-0.765E-01 C
C r
y 6
-0.755 0.383E-03 C
C
=
=
7 7
-0.244E-01 0.377E-01 C
C
=
=
3 8
0.332E-02
-0.193E-03 C
C
=
=
4 g
-0.486E-04 C
=
5 Typical values of the nominal IMCPR and IMCPR are given in 95/50 Table 15B-4 for several power and flow conditions.
Again, the IMCPR values include uncertainties and biases in process 95/50 variables and IMCPR calculations.
The 95 percentile IMCPR is l
based on the assumption that IMCPR is normally distributed about a mean value.
The approximate straight line fit to the datapoints of the probability plot of Figure 15B-4 indicates that this is a valid assumption.
15B.4-8 l
l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
()
15B.4.3.2.2.2 Expected IMCPR Over the Operating Map Figure 15B-5 shows the nominal IMCPR over the reference BWR/6 operating map.
For a constant power level, the expected IMCPR increases with increasing core flow due to the better critical power 5
performance predicted by the GEXL correlation at the higher flow.
For example, at 60% power the expected IMCPR is %1.69 at 45% flow, whereas at 70% flow the expected IMCPR increases to N2.00.
This IMCPR dependence on core flow was one of the primary reasons why the IMCPR curve of Step 3 can conservatively be based on 95/50 reactor power, independent of core flow.
15B.4.3.2.3 IMCPR Capability Curve - Step 3 95/50 Since the RWL System withdrawal increments are only core power dependent, it is necessary to reduce IMCPR to a function of power only.
From the statistical evaluation of IMCPR (Subsec-(
tion 15B.4.3.2.2), it was observed that along a constant power line the minimum IMCPR value occurs at the corresponding minimum flow within the operating domain.
For purposes of the RWL System design, minimum flow was defined to be along the rated flow control line (FCL) (line A of Figure 15B-5) and the minimum flow control valve (FCV) position line at rated pump speed (line B of Figure 15B-5).
Figure 15B-6 plots the IMCPR95/50 value along these trajectories as a function of power.
This curve represents the BWR/6 MCPR per-formance capability for core power levels above 20% of rated.
Achieving this performance is expected to require minimum rod pat-torn optimization consistent with operational startup considera-tions.
For instance, at 50% power there is 95% probability with 50% confidence that a radial power shape can be established with an IMCPR or at least 1.62.
As notted previously, this is conservative at higher flows, but accurate along trajectories A and B of Fig-
/ 'T ure 15B-5.
Conversely, operation to the left of trajectories A Q
1 l
and B with Figure 15B-6 IMCPR values will require more rod pattern optimization.
15B.4-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.2.3 IMCPR Capability Curve - Step 3 (Continued) 95/50 The abrupt change in the slope of Figure 15B-6 at N53% power corresponds to the intersection of the rated FCL and the minimum FCV position line.
Since the limiting MCPR fuel assembly is experiencing a flow reduction as power is lowered along trajectory A of Figure 15B-5, whereas flow is approximately constant along trajectory B, the GEXL critical power for the limiting assembly is expected to fall at a faster rate along trajectory A than trajectory B.
The ratio of critical power to fuel assembly power for the limiting assembly (i.e., MCPR) is expected to rise faster as power is reduced along trajectory B than along trajectory A.
This is evident in Figure 15B-5.
Between 60% and 80% power, the difference in the expected MCPR is 10.18, whereas, between 30% and 50% power, the expected change is %0.95.
Figure 15B-6 should not be interpreted as the IMCPR required to protect a[ainst a RWE transient independent of other RWL system design considerations.
With the BWR/6 RWL System, the withdrawal distances are set to ensure that the design criteria stated in Subsection 15B.4.2.1 are met.
Any value of IMCPR could theoreti-cally be chosen.
All that is required is that the chosen IMCPR technical specification be consistent with the RWL enforced with-drawal limits.
Thus, the RWE event need not necessarily become a limiting transient on BWR/6, since it is dependent on a combination of the selected control rod withdrawal distances and the IMCPR technical specification.
To effectively tradeoff BWR/6's projected IMCPR capability and the maximum allowable withdrawal distances, the 95% probability /50%
confidence IMCPR value was selected.
Choosing higher probability and/or confidence levels would result in a lower IMCPR technical specification and, hence, reduce withdrawal distances.
- However, this would not guarantee a higher degree of plant safety, since RWL enforced withdrawal distances are chosen to be consistent with O
15b.4-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
()
158.4.3.2.3 IMCPR Capability Curve - Step 3 (Continuad) 95/50 whatever value is selected as a technical specification.
It should be noted that this is entirely equivalent to the tradeoffs between RBM trip setpoint selection and technical specification MCPR limits on BWR/2-5.
15B.4.3.2.4 Operating Plant Data Comparison to the Projected BWR/6 IMCPR95/50 Figure 15B-7 compares the actual operating plant process computer data with the projected BWR/6 IMCPR capability.
These data (Attachment C) was only screened to eliminate 7x7 fuel, low power density plants, suspect data and low power operation at high core flows.
All datapoints correspond to high power density (%50.7 kW/t) BWR/4s during steady-state operation.
The BWR/6 IMCPR capability is increased 5.2% to compensate for the average higher power density O'
of BWR/6 (N53.3 kW/t).
Thus, Figure 15B-7 gives some insight to the expected credibility of the analytically determined BWR/6 IMCPR as a function of power.
Above 50% power, some datapoints do not meet the BWR/4 equivalent IMCPR versus power capability.
This implies that additional rod pattern optimization may be required.
(Note:
operating plants are required to operate to a MCPR versus flow technical specifica-tion which bounds the datapoints exceeding the BWR/4 equivalent IMCPR95/50*I The dashed line in Figure 15B-7 represents the nominal value of IMCPR from the fit of aralytically determined IMCPRs corrected for observed biases between analytical prediction methods and plant process computer values.
The good agreement between the equivalent BWR/4 nominal IMCPR and the operating plant data indicates that the normalized analytical methods are accurately predicting
/s expected core performance.
15B.4-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.2.5 Summary Typical rod patterns were developed over the operating map, result-ing in a sampled database of IMCPR values.
Core parameters such as rod sequence, exposure, core size, etc., were randomly varied to develop a database that encompasses all projected BWR/6 steady-state (and quasi-steady-state) operating conditions.
IMCPR was statistically evaluated as a function of power and flow to determine a generic IMCPR capability curve with 95% probability /
50% confidence level as a function of power.
Operating plant data indicate that the projected IMCPR capability provides significant margin at low powers, while some rod pattern optimization may be required above 50% power.
15B.4.3.3 Ganged Rod Withdrawal AMCPR Response With the IMCPR capability established as a function of power, the next step is to calculate the AMCPR response to rod withdrawals.
This response defines the allowable withdrawal distances, which assure with a high degree of confidence that AMCPR will n t be DB exceeded.
Assaring the consistency of the IMCPR technical specification (IMCPRTS) and RWL withdrawal limits is the primary safety-related feature of the entire design procedure.
The most important choice l
remaining in the statistical procedure is the probability and confidence level of AMCPR A
qua e, yet reasonable conserva-DB.
l tism is required.
Consequently, the RWL withdrawal limits that l
correspond to a 95% probability /95% confidence level AMCPR ere DB determined.
l l
O 15B.4-12 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 D)
/
15B.4.3.3.1 Ganged Rod Withdrawal Simulation - Step 4 The RWE transient is assumed to be a slow transient, allowing sufficient time for heat transfer and void redistribution to equilibriate.
The three-dimensional BWR simulator was used to perform a series of steady-state calculations at various ganged rod withdrawa] distances and the MCPR response was tracked.
Throughout the ganged rod pull, xenon was conservatively assumed not to redistribute.
These same approximations are currently employed in BWR/2-6 plant specific deterministic RWE analyses.
Rod withdrawals were initiated from the diverse A-sequence rod patterns developed in Step 1 (Subsection 15B. 4. 3. 2.1).
The use of A-sequence rod patterns is conservative, since they are mirror symmetric with control rod gangs consisting of four rods in the central region of the core.
In a B-sequence, the rod pattern is not mirror symmetric with control rod gangs consisting of three
()
rods.
Thus, B-sequence ganged rod pulls generally result in smaller core average power perturbations and thus smaller MCPR responses than A-sequence pulls.
Theoretically, an operator can inadvertently select and withdraw any rod or gang of rods.
To bias the AMCPR response in the con-servative direction, only the withdrawal of high worth, neutroni-cally coupled, deep rod gangs was simulated.
The smaller core power porturbations from peripheral rod gangs, single rod and shallow rod withdrawals result in a AMCPR response which is signi-ficantly less than the central core ganged withdrawals.
The extent of this conservatism is discussed in Subsection 15B.6.3.1 and 15B.6.3.2.
To encompass the expected allowable withdrawal distance, the with-drawal increments were varied according to the initial power level.
For example, at rated conditions, the AMCPR was typically calcu-
'}
lated for 1, 2,
and 4-ft withdrawals, whereas at 40% power, 2,
6 15B.4-13
GESSAR II_
22A7007 238 NUCLEAR ISLRND Rev. 0 15B.4.3.3.1 Ganged Rod Withdrawal simulation - Step 4 (Continued) and 10-ft withdrawals were analyzed.
In addition, the initial position of the control rod gang varied between full insertion and 4 ft withdrawn.
For fully inserted rod gangs, the MCPR response to the initial portion of the withdrawal (%1 ft) is typically not as large as the response later in a pull.
A typical MCPR response to a high worth ganged rod withdrawal at rated condit'ons is shown in Figure 15B-8.
In this case, the ganged rods were initially positioned at notch 8 (2 ft withdrawn) so that the full withdrawal increment corresponds to 10 ft.
The maximum MCPR response slope occurs between 0 and 4 ft withdrawn.
Beyond 6-8 ft withdrawn, the response to further withdrawals is minor due to the relatively small reactivity insertion and void redistribution.
15b.4.3.3.2 Statistical Evaluation of AMCPR/IMCPR - Step 5 15B.4.3.3.2.1 AMCPR Database The output of Step 4 is a database consisting of AMCPRs at various power acJ flow conditions for several withdrawal increments.
A total of 27 power / flow states were analyzed (Figure 15B-9) with a total of 53 ganged rod withdrawal calculations.
In those instances where more than one calculation was performed at a particular power / flow state, the initial rod patterns were gener-ally different.
However, in a few cases, the initial rod pattern was held constant and different high worth gangs were withdrawn.
The AMCPR database is summarized in Attachment D.
The sufficiency of the AMCPR database is demonstrated in Figure 15B-10.
Periodically, as additional data became available, the nominal value of the fitted parameter (AMCPR/IMCPR) of the statistical model (Subsection 15B.4.3.3.2.3) was evaluated.
Note O
15B.4-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
[m V]
15B.4.3.3.2.1 AMCPR Database (Continued) that process instrumentation and calculational uncertainties and biases were not included in this evaluation.
Beyond approximately 45 datapoints, the mean of the fitted parameter does not change appreciably.
Thus, the AMCPR database is sufficient and the statistical model is converged.
15B.4.3.3.2.2 AMCPR Database Biases The AMCPR database was qualitatively evaluated for biases intro-duced by the variable core parameters (i.e., core average enrich-ment, core size, cycle exposure, etc.), using crossplots as described in Subsection 15B.4.3.2.1.3.
Attachment E contains the crossplots for the AMCPR database.
Since no obvious trends exist in any of the crossplots, it was concluded that the AMCPR database could be considered unbiased with respect to these parameters
-(g) without introducting s2gnificant model inaccuracies.
15B.4.3.3.2.3 Engineering Model The statistical fitting program discussed in Subsection 15B.4.3.2.2 was used to fit AMCFR/IMCPR as a linear function of percent power (P) and percent flow (F) and a quadratic function of distance withdrawn (D):
AMC R = Cl * (D + C2
- D ) * ( 1. 0 + C 3
- P ) *
(15B.4-3)
(1.0 + C4
- F)
AMfR = EXP [C5 + C6
- P + C7
- F]*
(15B.4-4)
[1.0 - EXP (-EXP (C8 )
- D)]
15B.4-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.3.2.3 Engineering Model (Continued) where C1 = -0.836E-01 C5 = -2.403 C2 = -0.503E-01 C6 = -0.104E-01 C3 = -0.560E-02 C7 =
0.698E-02 C4 = -0.154E-03 C8 = -0.765 The param>3ter AMCPR/IMCPR is fitted, since studies have shown that AMCPR is proportional and, hence, is approximately a linear func-tion of IMCPR (Figure ISB-ll).
This results in an accurate fit to the data.
Typical 95% probability /95% confidence AMCPR/IMCPR values are given in Table 15B-5.
These values incorporate process instru-mentation, calculational and operating rod pattern selection uncertainties and biases.
Comparison of Cases 2 and 3 or 4 and 5 indicates that, for a constant power level, the maximum value of
( AMCPR/II1CPR) 95/9 5 ccurs at the maximum core flow (100% in these cases).
Since IMCPR increases with increasing core flow (Figure 15B-5), AMCPR is also maximized at high core flows.
Although the smaller core average void fractions at higher flows would be expected to result in lower rod worths on the average, increased power peaking response to rod withdrawal at lower voids more than compensates for the rod worth effect.
The value of AMCPR as a function of core power is calculated 95/95 from the following equation:
(15B.4-5)
(P)
AM PR P) 95/95 "
PR MAX 95/95 95/50 (P)
IMCPR l
l 15B.4-16
GESSAR II
'22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.3.2.3 Engineering Model (Continued) where AMCPR is the 95% probability /95% confidence IMCPR MAX 95/95 value at the maximum core flow, F analyzed at the power P and IMCPR s o Mained hom M gure 15B-6.
95/50 Application of (AMCPR/IMCPR)95/95 at the maximum analyzed core flow is conservative at lower core flows.
Since the RWE transient is assumed to be initiated from IMCPR n
r dit is taken 95/50, for the probability that margin exists to IMCPR For opera-95/50 tion to the right of trajectories A and B of Figure 15B-5, significant margin between the actual IMCPR and the IMCPR95/50 value exists.
ISB.4.3.3.3 Maximum Allowable Control Rod Withdrawal Distance -
Step 6 The maximum allowable centrol rod withdrawal distances are set such that there is 95% probability with 95% confidence that the change in MCPR during the event (AMCPR) will not exceed AMCPR DB during any randomly occurring RWE transient.
To establish the maximum allowable withdrawal distances, it is initially assumed that IMCPR is quiv 1 nt to a preliminary IMCPR technical 95/50 specification (IMCPR
).
Thus, a preliminary design basis p
AMCPR (AMCPR can M M nM as de Merence Mween de PDB preliminary technical specification IMCPR and the safety limit MCPR.
From the MCPR response to a ganged rod withdrawal, the withdrawal distance corresponding to AMCPR ere M erminei PDB For example, Figure 15B-12 gives AMCPR95/95 versus ft withdrawn at 100% power and 100% flow.
From Figure 15B-7, IMCPR equals PTS 15B.4-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.3.3 Maximum Allowable Control Rod Withdrawal Distance -
Step 6 (Continued) 1.17, which corresponds to a AMCPR
~
PDB where the safety limit MCPR = 1.07).
Plotting AMCPR n
gure 15B-13 indicates that die allowable PDB withdrawal dictance is 1.0 ft.
Similar calculations were performed at other core powers.
The resulting maximum allowable withdrawal distances as a continuous function of power are shown in Figure 15B-13.
The distances increase with decreasing core power from 1.0 ft at rated condi-tions to 12 f t at 20% power.
15B.4.3.3.4 Optimum Withdrawal Increments The maximum allowable withdrawal distances of Subsection 15B.4.3.3.3 are consistent with projected BWR/6 MCPR capability and the chosen safety criterion (Subsection 15B.4.2).
To deter-mine the optimum withdrawal increments, it was necessary to factor the operational impact of the IMCPR n
e analysis.
A cost,/
PTS benefit analysis was performed to answer the basic question:
"Is it economically desirable to increase the required IMCPR at off-rated conditions beyond that necessary to protect against non-RWE transients to allow greater RWL setpoints at the risk of having limits so high as to require more frequent and accurate monitoring of IMCPR during startup with the associated capacity factor losses?"
Review of current operating plant practices indicated that a 2-ft withdrawal between 20% and 70% core power and a 1-ft withdrawal between 70% and 100% power allows adequate rod maneuverability during startup and power changes.
Thus, the primary benefit of 15B.4-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.3.4 Optimum Withdrawal Increments (Continued) large withdrawal increments is a reduction in the time required to perform a rod sequence exchange.
Assuming six sequence exchanges per year at 40% power and $300,000 per day power replacement costs, the increased costs associated with a 2-ft withdrawal instead of a 4-ft withdrawal was estimated to be
$22,500 per year:
6 "*
""9 8
hours lost day
$300,000 x 0.5 year exchange 24 hrs full power day (100-40)% full power lost
$22,500/ year
=
100%
Por fixed withdrawal increments, it is a simple calculation to determine the corresponding IMCPR requirements.
Substituting IMCPR and AMCPR f
0" E in Equation TS 95/50 DB 95/95
[G) 15B.4-5 results in the following equation:
AMCPR
}
- IMCPR (15B.4-6)
DB " It MAX 95/95 TS Further, substituting AMCPR M
an carranging DB TS gives:
IMCPR
=
(15B.4-7 TS
-AMCPR(p,pMAX(p)_
95/95 y_
_IMCPR where SLMCPR = safety limit MCPR.
l l
Thus, for any power level, the minimum required IMCPR ra TS fixed withdrawal increment can be calculated from Equation 15B.4-7.
l i
15B.4-19 l
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 15B.4.3.3.4 Optimum Withdrawal Increments (Continued)
At 40% power, the minimum IMCPR requirement for a 4-ft withdrawal was calculated to be 1.87, while the IMCPR requirement for a 2-ft withdrawal was 1.53.
Again, assuming six plant startups per year, it was estimated that each would have to be slowed down <0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> by increased MCPR monitoring to justify a 4-ft versus 2-ft RWL increment at 40% power.
Since a single recalibration of the process computer typically requires 3-4 hours, it was concluded that it is not likely to be cost effective to increase RUL distances to reduce the time required to perform rod sequence exchanges.
In addition to the operational impact, the final withdrawal incre-ments must be consistent with the intent that the RWE not impose IMCPR requirements which are significantly more stringent than those required to protect against other moderate frequency tran-sients.
A review of the MCPR response to the slow flow runout transient and both nonpressurization and pressurization transients established the minimum IMCPR requirements of approximately 1.37 at 70% power and 1.54 at 40% power.
Using the procedure discussed in Subsection 15B.4.3.3.3, the corresponding withdrawal increments are 1.7 ft and 2.0 ft at 70% and 40% power, respectively.
- Thus, the minimum rod maneuverability requirements discussed previously l
are satisfied at 40% power, whereas at 70% power a slightly higher l
operating limit IMCPR is required to ensure a 2-ft withdrawal.
Using Equation 15B-7, the necessary IMCPR is 1.41.
i Since the control rod drive latches only at 6-in. intervals, the l
final optimum RWL withdrawal increments are 2 ft between 20% and 70% power and 1 ft between 70% and 100% power.
A histogram of l
the RWL rod block setpoints is given in Figure 15B-14.
l l
l l
l 15B.4-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/~^
15B.4.3.3.4 Optimum Withdrawal Increments (Continued)
U Below 20% power, the RWL System is not functional.
Instead, rod movements are restricted by the Rod Pattern Control System (RPCS) enforcement of the Banked Position Withdrawal Sequence (BPWS).
The BPWS requires rod groups to be banked at 1, 2,
3 and 12 ft out.
Thus, the maximum withdrawal distance under BPWS constraints is 9 ft.
The IMCPR required to give a 9-ft withdrawal at 20% power is 2.58.
This corresponds to a 99% probability /50% confidence IMCPR.
Thus, the BPWS assures an acceptable RWE response.
15B.4.3.3.5 IMCPR Technical Specification as a Function of Core Power With the RWL withdrawal increments fixed and the AMCPR response known, Equation 15B.4-7 was used to calculate the required IMCPR technical specification to protect against the RWE transient.
The results of this calculation are given in Table 15B-6 and graphi-O cally presented in Figure 15B-15.
Comparison of this curve to Figure 15B-6 indicates that BWR/6 should have minimum difficulty in establishing rod patterns which satisfy the IMCPRTS*
15B.4.3.3.6 Rod Movement Restriction Technical Specification The input power signal to the RWL System originates from the first-stage turbine pressure.
When operating with steam bypass, this signal gives a biased power indication.
This can result in greater withdrawal distances being allowed than the design and licensing basis support.
For example, for a core operating at 50% power with 35% bypass, the input power signal co-responds to
%15% power.
Since this power is below the low power setpoint (LPSP)
I (typically 20% of rated power), the RPCS is functional instead of the RWL System.
The RPCS enforces the BPWS, which allows a maxi-mum 9-ft withdrawal.
Thus, instead of rod withdrawals limited to 2 ft at 50% power, the potential exists for a 9-ft withdrawal.
3 l
15B.4-21
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.4.3.3.6 Rod Movement Restriction Technical Specification (Continued)
If a 9-ft withdrawal did occur, however, the probability of violating the safetly limit MCPR is small.
At 30% power /36% flow, IMCPR is 2.41 (Figure 15B-6).
From Equations 15B.4-3 and 95/50 15B.4-4,
( t:MCPR/IMCPR) 9 5/ 50 is 0.48.
Thus, the final MCPR follow-ing a 9-ft withdrawal at 30% power is 1.25 (2.41 - 0.48 2.41),
which is greater than the safety limit MCPR.
Above 30% power, the operator is typically in a power-shaping mode.
In this case, the rod pattern will most likely not conform to BPWS requirements.
As a result, when operating above 30% power with sufficient bypass to result in a RPCS input power signal less than the LPSP, viola-tion of BPWS constraints will result in both insert and withdrawal blocks on all rods.
To ensure that the above situation does not occur, the following technical specification is required:
"Do not withdraw control rods when operating above the LPSP with steam bypass."
A detailed evaluation of this power signal bias and its impact is provided in Attachment F.
15B.4.3.3.7 Comparison of Generic and Deterministic RWE Analyses Plant-specific deterministic RWE analyses are typically performed only at rated conditions.
Several plant-specific analyses have been completed in support of early BWR/6 FSARs prior to the com-pletion of the generic analysis.
The allowable withdrawal dis-tances from these plant-specific analyses (Table 15B-7) are in general agreement with the 1-ft withdrawal distance from the generic analysis.
Thus, at rated conditions, the overall conser-vatism in the generic analysis is equivalent to the conservatism inherent in the deterministic analysis.
15B.4-22
_. - -.. ~. -. _... _ -...
GESSAR II 22A7007 i
238 NUCLEAR ISLAND Rev. O I
1 O
158.4.3.3.7 Comparison of Generic and Deterministic RWE Analyses (Continued) 1 1
}
Two additional deterministic RWE analyses were performed at 70%
I power on the rated FCL.
The resultant withdrawal distances are in agreement with the generic analysis results.
i j
i l
l I
i l'
l I
I l9 I
i i
e i
l 15B.4-23/15B.4-24 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
{
15B.5 UNCERTAINTIES AND BIASES 15B.
5.1 INTRODUCTION
As part of the statistical evaluation of IMCPR and AMCPR, process instrumentation.
calculational uncertainties and biases must be 4
incorported.
Process uncertainties include measured inputs to the process computer thermal limits calculation.
Calculational uncer-tainties include errors in the three-dimensional BWR simulator model and the process computer calculation of thermal limits rela-tive to the true value.
The safety-related effects of all uncertainties associated with the monitored IMCPR prior to a RWE event are accounted for in the GETAB safety limit MCPR. 5 True variability in pre-RWE IMCPR performance and AMCPR margin loss during a RWE event was included through the diversity of initiat-ing states included in the statistical database.
(
15B.5.2 IMCPR UNCERTAINTIES AND BIASES Compliance with the technical specification operating limit MCPR is based on the process computer calculated IMCPR (IMCPR PC Prior to statistical evaluation, each analytical IMCPR (IMCPRAN calculated by the three-dimensional BWR simulator model was trans-formed into a corresponding process computer IMCPR by the PC general equation:
PC PC' PC AN PC' PC) +B{
IMCPR
+
PC
' AN PC (15B.5-1) where the process measured core power and flow; P
=
PC' PC I
[ )'l B "N+PC the bias betvaen analytical and process computer
=
(,
A calculated IMCPR values; 15B.5-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.5.2 IMCPR UNCERTAINTIES AND BIASES (Continued) a random normal deviation from Monte Carlo c
=
sampling from a frequency distribution with a mean of one and a unit standard deviation; and IMCPR the standard deviation between analytical and o
=
AN+PC process computer calculated IMCPR values.
A different random value of c is used for every IMCPR datapoint.
From comparisons of the three-dimensional BWR simulator and pro-cess computer bundle power peaking factors for 106 states of 10 BWRs, the standard deviation in MCPR values was determined to be 3.01.
The bias between the two methods is 5.2%, with the three-dimensional predictive model bundle power peaking being lower than the corresponding process computer value.
Since these comparisons are performed at the same apparent power and flow (i.e., PAN " EPC
=
es a ues include de uncedainy and Mas and FAN PC between measured power and flow and the true value.
Thus, IMCPR PC in Equation 15B.5-1 is indicated as a function of P and F PC PC*
As discussed in Subsection 15B.4.3.2.2, the IMCPR was then fit PC as a function of power and flow.
The 95% probability /50% confi-dence values of IMCPR are calculated internal to the statistical Pg i
l fitting program.
These IMCPR v lu s are used as indications 95/50 l
of the MCPR performance capability of BWR/6 to assure that tech-l nical specifications IMCPRs are operationally acceptable.
i l
15B.5.3 AMCPR UNCERTAINTIES AND BIASES The withdrawal distance prior to rod block is based upon AMCPRDB*
To ensure that AMCPR n
oa as necessary to de W -
DB I
mine the true MCPR response to a ganged rod withdrawal, SMCPRTRUE.
The incorporation of uncertainties and biases in the analytical l
158.5-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
()
15B.5.3 AMCPR UNCERTAINTIES AND BIASES (Continued) prediction of AMCPR (AMCPRAN) used the same Monte Carlo technique as employed in the IMCPR calculation, i.e.:
AMCPR
)
AMCPR I
+B
=
TRUE PC' PC, AN PC' PC, A
UE AMCPR gAN+TRUE (15B.5-2) where the measured values of power and flow; PPC,FPC
=
withdrawal distance (ft);
D
=
e biases between analytical and true values of B
=
AJ RUE AMCPR; AMCPR the standard deviation between analytical and
=
AN+TRUE true values of AMCPR; and c is defined in the previous section.
Note that the AMCPR error contributed by the mechanical tolerance on control rod drive latching positions is insignificant and was neglected.
Comparison of the standard deviation of planar normalized three-dimensional BWR simulator nodal power distributions relative to gamma scans l
for typical rod patterns before and after rod withdrawals indicated that:
AMCPR A CPR 8.0% and B
+4.8%
=
c
=
l AN+TRUE AN+TRUE
{
\\
O 15B.5-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.5.3 f.MCPR UNCERTAINTIES AND BIASES (Continued) where a positive bias indicates three-dimensional simulator over-prediction of AMCPR.
Since the values of power and flow input to the three-dimensional simulator are subject to independent choice, tMCPR is interpreted as a function of P and F TRUE PC PC*
For each measured power and flow, AMCPR datapoint and corresponding RWE initial state IMCPRAN( PC' PC), Equation 15B.5-1 was used to determine IMCPR IPPC'EPC) using M nte Carlo sampling.
Measured PC powers and flows were used, since the IMCPR f Equation 15B.4-7 TS (Subsection 15B.4.3.3.4) is implemented based on measured power and flow inputs to the process computer.
Also, measured power and flow is mathematically consistent with the methods used to estimate the biases and uncertainties.
The parameter (AMCPRTRUE}!
was dien fit as a function of PC measured power, measured flow and distance withdrawn (Subsec-tion 15B.4.3.3.2).
The desired 951 probability /95% confidence withdrawal distances corresponding to AMCPR were the primary out-DB puts of the fitting program.
15B.5.4 Statistical Procedure Conservatisms The model described in Section 15B.5 establishes a method for modifying calculated va2ues of MCPR and aMCPR responses to attain typical expected true values.
This involves a nominal correction for the mean calculational bias and a random correction for vari-ations from the mean bias.
It will now be shown that this is valid for instances where only one or a few input calculations are available and is conservative when many input cases are available.
The general problem is one of estimating (predicting) the nature of the true distribution of a parameter (characteristic) H of a population of actual RWE everts, from a sample of calculated 15B.5-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 15B.5.4 Statistical Procedure Conservatisms (Continued) values, h, from a set of hypothesized events.
It is assumed i
that the set of chosen hypothesized events can be assumed to be a random sample of (potentially) actual events.
For hypothesized event, i, the relationship between,111, and h1 can be expressed as:
ll)
(15B.5-1) h (1 + pH+C i Il U
=
y i
where it is assumed that the fractional differences between the actual and calculated parameter values can be characterized as a normal distribution with mean, p and standard deviation, o g,
g.
The event specific unknown error factor is c The values of c f.
i form a standard normal distribution.
The expected value for lif i (1+pg).
Assuming p and o are known, confidence is therefore h H
g intervals for Il can be calculated by standard techniques.
i The problem becomes more complicated when it is desired to esti-mate the relationship between several physical parameters -- as in the RWE analysis.
By example, we desire to evaluate an esti-mate of
[p (X)1 (15B.5-2)
H (X)
E i
by a study of calculated values of Amcpr and imcpr.Ili can be expressed as:
Amcpr(X) (1+p +c3,(X)o )
3 g
H (X)
E (15B.5-3) g imcpr (1+py+c; g X)cy) g whereas before the mean biases in IMCPR and AMCPR are given by q
p and p respectively.
Standard deviations are similarly y
g, 7 g(X) expressed.
The event specific unknown error factors are c 15B.5-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.5.4 Statistical Procedure Conservatisms (Continued) and o
- (X).
Assuming p u t., c
""d U
are known, the expected g
y, I
A values and confidence intervals for Ili(X) can be determined by standard techniques.
Up to this point, we have been concerned with estimating a caram-eter or functional relationship for a single hypothetical event.
We now consider the related problem of estimating the distribu-tion of li from many events.
Since estimation of the mean is g
simplest, we consider it first:
11
~ h, 11 hEh (1 + u, + c i 33 )
=
0 1
i g
(1 + p H)
Zh
+
Th c
=
i i
1 (1 + pg) h+0 (15B.5-4)
=
Since in the limit as N++,
by definition for every h the i,
probabilities of positive and negative values of c of the same i
absolute value are equal.
Correcting all values of the database by the mean difference will thus assure estimation of the expected value.
Now, consider the problem of estimating the standard deviation o f 11,
a.
I f !!
is f rom a normal population of mean 11 and stan-1 dard deviation c, we can write (15 B. 5-5 )
II ( 1 +
1.
a- )
II.
=
1 1
where the r form a standard normal distribution.
g O
15B.5-6
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 e
i 15B.S.4 Statistical Procedure Conservatisms (Continued)
We can now write (using 15B. 5-1 and 15B. 5-5) 1 H (1+T c) (1+pH}
y (1 + pH)hi
=
(1+p +'i H}
H 4
i i
ii (1+T a) i (15B.5-6)
=
o o
y g
+ 1+p i
C l
H
/
Using the standard relationship for the variance of a function Y (Xg) where the X are independent Y
=
g l
3Y \\
E.
Var X.
(15B.5-7)
Var Y
=
3X /
1 1
f i
applied to Equation 15B.5-6 gives 2
\\
Ho l
-(1+pH) h H
(H c)2 (15B.5-8)
I
+
l Ver
=
i_
y, H/
4 l
For the case where 1 + p
%1 H
1 2
2)
(15B.5-9) i var (1+pH)h,% if (H
+
I I
l Thus, the estimated variance from a database corrected only by the mean difference between a calculation and actuality will be larger than the variance of the actual values.
This is due to the added calculational uncertainty component of the overall variance.
Usage of these larger values will give wider (con-servative) confidence intervals.
The conclusion can be extended r
j 15B.5-7 i
.,_,,.----e,
-,,--.y
,e.,-
y-,
-,,,,,,,n,
,e
.x-
_,_-.-,,.,c n _,--,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.5.4 Statistical Procedure Conservatisms (Continued) to the evaluation of the relationship between AMCPR/IMCPR and X.
It is required only to correct the calculated AMCPR and IMCPR values by the mean error.
The effects of random calculational error is manifested in conservatively large confidence intervals.
The degree of conservatism depends on the relative magnitude of the calculational uncertainty to the other sources of the total variance.
For the generic RWE analysis, the variance due to reactor state and error rod withdrawn is large compared to cal-culational uncertainty and the degree of residual conservatism in the model that results by modifying each datapoint per Sub-sections 15B.5.2 and 15B.S.3 is exoected to be small.
O l
O 1
15B.5-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
(
15B.6 GENERIC ANALYSIS CONSERVATISMS 15B.
6.1 INTRODUCTION
t l
Conservatisms exist in both the IMCPR and AMCPR calculations.
Whereas IMCPR conservatisms primarily impact plant operability, maneuverability and capacity factor by imposing technical specifi-cation operating IMCPR requirements, the primary impact of AMCPR conservatisms is to increase the overall probability and confidence that AMCPR is n t violated.
It should be noted that choosing a DB low IMCPR also nhances safety considerations indirectly, since, TS in reality, the probability of a RWE transient initiation with more margin than assumed in the design basis is increased.
How-ever, the design basis assumes that the RWE transient is initiated from IMCPR and no credit was taken for this effect.
Also, the TS degree of conservatism chosen in design basis AMCPR values affects operability, maneuverability and capacity factor to a secondary
'T degree, since more conservatism reduces allowable rod withdrawal
[O distances.
15B.6.2 IMCPR CONSERVATISMS Nonoptimized rod patterns are the primary source of conservatism in the IMCPR calculation.
In this context, conservatism means that lower MCPR performance is calculated than expected in actual plant operation.
MCPR, MLHGR and criticality error band con-straints are used in the analytical searches to generate the IMCPR critical rod patterns.
These constraints were held constant over the power / flow operating map at values corresponding to rated conditions.
The result is that it is very easy to satisfy the rated thermal limits constrainte at low core powers.
Thus, some analytical rod pattern optimization is required near rated condi-tions; whereas, at low powers, significant margin to thermal limits exists and no optimization is necessary.
ew fO 15B.6-1
GESSAP II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.6.2 IMCPR CONSERVATISMS (Continued)
It is operationally undesirable to require a plant to optimize the margin to thermal limits at low powers, as this would slow down plant startup, maneuverability and reduce capacity factors.
This is consistent with current BWR thermal limits technical specifica-tions, and, as a result, the conservatism in the nominal IMCPR from the statistical fit to the data is expected to be small, assuming BWR/6s follow current operating practices.
This is sub-stantiated by the excellent agreement of the nominal IMCPR with operating plant data in Figure 15B-7.
Significant conservatism in the nominal predicted IMCPR does exist relative to the expected IMCPR capability as a function of power if rod patterns are optimized during startup.
This margin gives added assurance that BWR/6 will be able to operate within the proposed MCPR technical specification.
15B.6.3 AMCPR CONSERVATISMS ISB.6.3.1 High Worth Control Rod Gang Withdrawals Central core high worth control rod gangs were withdrawn to deter-mine a conservative AMCPR response.
Typical values of AMCPR as a function of distance withdrawn for both high worth and low worth gangs withdrawn from the same base rod pattern are given in Table 15B-16.
The withdrawal of the high worth gangs results in a greater change in MCPR.
Table 15B-16 also shows typical results for peripheral rod gang withdrawal.
Peripheral rods are located in the second row in from the core boundary as opposed to edge rods, which have at least one bundle with a water face in the four-bundle cell containing the rod.
Peripheral rods are restricted by the RWL system enforced withdrawal limits, whereas adge rods are not (i.e., edge rods can 9
15B.6-2
l GESSAR II 22A7007 238 N'JCLEAR ISLAND Rev. 0 N
15D.6.3.1 High Worth Control Rod Gang Withdrawals (Continued) sv be continuously withdrawn from full-in to full-out).
This represents no safety concern for the following reasons:
(1) edge rods typically are fully withdrawn prior to the LPSP; (2) the IMCPR's for edge bundles are significantly greater than the core limiting MCPR (thus, much larger AMCPR's can be tolerated); and (3) edge rod gangs are neutronically uncoupled and the rod worths are extremely small.
The resultant perturbation to core power during edge rod withdrawal is small and the MCPR response is typically less than that for peri-pheral rod gangs.
The statistical evaluation of AMCPRs based entirely on high worth gangs is biased in the conservative direction.
For a truly random sampling of AMCPRs for a specific withdrawal increment at a speci-fic operating condition, a distribution similar to frequency dis-tribution A of Figure 15B-16 would result.
The 95% probability /
50% confidence AMCPR value is designated as Point 1.
In the generic analysis, a censored sample of AMCPRs representing the AMCPRs of Curve A are evaluated resulting in frequency distribu-
[
tion B.
The 95% probability /50% confidence value of distribution B is indicated as Point 2.
Only high worth control rod gangs were included in the database for distribution B.
U: is estimated that'the resulting AMCPRs p
all fall within the upper 30% of distribution A.
This is not an area of safety concern, since the stated statistical design basis would allow a completely random gang selection.
Also, if an (n) error in engineering judgment exists which significantly under-estimates the AMCPR ';esponse for a few selected ganged rods, these 15B.6-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.6.3.1 High Worth Control Rod Gang Withdrawals (Continued) few outlier points would increase the curve B standard deviation to compensate for any slight reduction in the AMCPR.
In the limit, the inability to select high worth ganged rods would result in distribution B being identical to distribution A, which again is still acceptable.
The degree of conservatism in the AMCPR calculation due to high worth control rod gang selection was not evaluated since a truly random sample was not generated for comparison.
At this time, no formal credit is taken for this conservatism.
However, based on AMCPR comparisons like those of Table 15B-8, the conservatism would not be expected to result in changes in plant operability or safety margins afforded by this current procedure.
15B.6.3.2 Single Rod Withdrawals Single rod withdrawals izere not included in the AMCPR database because of the smaller MCPR response relative to ganged rod with-drawals (Table 15B-9).
In the context of this report, the cen-tral control rod is designated as a single rod and not as a gang.
The MCPR response for both the highest worth single rod (usually the central control rod) and a rod contained in the high worth gang is included for comparison.
In both instances, AMCPR s
GANG a minimum of 1.8 times AMCPR with the MCPR response for the SINGLE central high worth single rod being slightly smaller than that for off-center single rods.
As discussed in Subsection 15B.6.3.1, it is difficult to quantify the absolute conservatism in this design basis when single rod withdrawals are not considered in the random RWE database.
No credit is taken for this conservatism.
However, assuming a random i
distribution of single and ganged rod withdrawals, the conserva-tism in the average AMCPR is estimated to range from 0.01 at 100%
power to 0.15 at 40% power.
15B.6-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.6.3.2 Single Rod Withdrawals (Continued)
The currently defined RWL hardware system does not distinguish between single and ganged rod withdrawals and applies the same withdrawal blocks to each.
A system update that applies separate blocks for single and ganged rod withdrawals would be able to take advantage of this conservatism.
However, support for such a sys-tem is not the intent of this report.
15B.6.3.3 AMCPRs at Maximum Core Flow To determine the allowable withdrawal distance for a particular power level, the AMCPR at maximum core flow was used.
As dis-cussed in Subsection 15B.4.3.3.2, the MCPR response is largest at the higher core flow.
Since (AMCPR/IMCPR)95/95 n reases for increasing core flow (Table 15B-5), the maximum conservatism in this parameter occurs at the minimum allowed core flow at any O
selected power level.
Comparison of AMCPR values for 2-ft 95/95 withdrawals in Table 15B-5 along constant power lines indicates that the conservatism is approximately 0.02 at 80% power /70% flow and 0.13 at 60% power /44% flow.
The AMCPR values at any power and flow point are obtained by multiplying the (AMCPR/IMCPR)95/95 by the appropriate IMCPRTS (Figure 15B-15).
This also results in conservatism through the following process.
Figure 15B-5 shows that the nominal IMCPR performance is a significant function of flow as well as power.
Even though the parameter ( AMCPR/IMCPR) (P,F) is slightly larger at maximum flow than along a typical startup path, IMCPR is much larger.
Thus, while the product:
^"fp AMCPR(P,F)
(15B.6-1)
IMCPR(P,F)
(P,F) *
=
is larger at maximum flow, in reality, the limiting condition at
()
any power levels occurs at the lower flow because at higher flows 15B.6-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.6.3.3 AMCPRs at Maximum Cora Flow (Continued) the plant is likely to initiate the RWE with more than the required excess MCPR margin (to the safety limit) to compensate for the slightly larger AMCPR.
Table 15B-10 illustrates this.
In the current procedure, the MCPR prior to a RWE is assumed to equal the low IMCPRTS(P) value at all core flows for a selected power.
The high ( AMCPR/IMCPR) (P, Fgg) value is also used at all flows corresponding to power P.
The product of these two gives the conservative design basis value of AMCPR used to set RWL dis-tance limits at all flows for power P.
This conservative procedure was chosen to eliminate the need for complex power and flow-dependent MCPR technical specifications and RWL withdrawal limits.
The degree of conservatism is judged operationally acceptable.
15B.6.3.4 Constant Xenon During Rod Withdrawal During the rod withdrawal simulation, it was conservatively assumed that the xenon distribution remained constant.
In the actual case, xenon does slightly redistribute; the degree is dependent upon how long it takes for an operator to recognize a RWE and take correc-Live action.
As a rod is withdrawn and the power increases within a localized region, the xenon concentration might initially decrease due to a faster burnout rate but it will ultimately increase, suppressing bundle powers and increasing MCPR.
Iloweve r,
since the half-life of I-135 (the precursor to Xe-135) is 6.58 hr and the control rod is withdrawn from the full-in to full-out posi-tion in less than one minute (12-ft withdrawal at the nominal con-l Linuous drive speed of 3 ips i 20%), xenon redistribution and its impact on MCPR is negligible.
O 15B.6-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.7 SPECIAL STUDIES 15B.
7.1 INTRODUCTION
Two studies were performed in conjunction with the generic RWE analysis to ensure that the IMCPR and aMCPR databases were conser-vative.
The purpose of the first study was to determine if any significant bias exists due to th7 assumption of an equilibrium xenon concentration prior to the RWE as opposed to rated xenon or xenon-free conditions.
The second study hypothesizes that a RWE transient occurs during a rod sequence exchange when extreme asymmetry exists in both the rod pattern and xenon distribution.
15B.7.2 XENON DISTRIBUTION STUDY All rod patterns developed for the RWE generic analysis assume an equilibrium xenon distribution consistent with the power / flow
[}
state.
In some instances, this assumption may not be explicitly correct.
For example, during load-following along the rated flow control line, rated xenon conditions could exist down to approx-imately 60% power.
Also, essentially xenon-free conditions could occur at low powers during plant startup.
Control blade maneuvers and combinations of the above can result in other transient xenon conditions.
However, these were considered to be covered by the above conditions and were not analyzed.
Several cases were run to approximate these conditions.
Starting with the rod patterns determined with equilibrium xenon, new rod patterns were developed, if necessary, to satisfy the rod pattern search module constraints.
Comparisons were then made to the IMCPR value from the design basis, generic RWE analysis.
Table ISB-ll lists the power and flow conditions at which these comparisons were made for IMCPR.
i l
l [}
Since the choice of the IMCPR is n t directly safety related, TS
\\'
the primary concern is that the plant be reasonably capable of l
i l
15B.7-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15B.7.2 XENON DISTRIBUTION STUDY (Continued) g satisfying the MCPR technical specification when other than equilibrium xenon conditions exist.
All values given in Table ISB-ll (except for one case each at 20% power /38% flow and 40% power /40% flow) fall within the 95% probability to 5% proba-bility range.
Since the IMCPR is significantly less than the TS nominal IMCPR capability at low powers (comparison of Fig-urcs 158-6 and 15B-15), no difficulty in meeting technical specification limits is expected.
In fact, at low powers the no-xenon condition predicts higher IMCPRs than the equilibrium xenon condition, but this is considered a statistical fluctuation of two datapoints and not a trend.
At higher powers (40-60% power), the difference between IMCPR values are on the order of 0.10, with the equilibrium xenon IMCPRs generally higher.
However, some xenon would be present, suppressing bundle powers somewhat in the high power regions and increasing the core limiting IMCPR.
The inverse of the above trend in IMCPR versus power level occurs when a rated xenon distribution is assumed down the rated flow control line.
At the low end of the flow control line, the rated xenon distribution overpredicts IMCPR relative to the equilibrium xenon condition.
As the power level increases, the equilibrium xenon distribution approaches the rated xenon distribution and the IMCPRs converge.
15B.7.3 ROD SEQUENCE EXCHANGE STUDY l
rod sequence exchange, highly asymmetric rod patterns Throughout a can exist.
Two typical rod patterns during a rod sequence l
l exchange were analyzed to demonstrate that the MCPR response to a ganged rod withdrawal is conservative relative to the generic analysis MCPR response.
l The two rod sequence exchange cases were developed in accordance with current operating strategies.
Figure 15B-17 shows the base 15B.7-2
_~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ll 15B.7.3 ROD SEQUENCE EXCHANGE STUDY (Continued) rod pattern assumed for Case 1 prior to an inadvertent gang with-drawal for an exchange from Sequence A2 to B2 at 40% power and 40% flow.
This is a typical rod swap operating condition under current guidelines.
The indicated deep gang was fully withdrawn and the resulting AMCPR calculated.
A similar rod pattern was developed for a rod swap from Sequence B1 and A2 (Case 2).
Table 15B-12 compares the AMCPRs for the two rod sequence exchange cases and the statistical fit of the AMCPR database.
In both cases, the AMCPR values from the generic analysis are 95/95 greater than AMCPR H wever, in Case 2, the nominal ROD SWAP.
AMCPR s greater than the predicted nominal AMCPR.
Since ROD SWAP the nominal oMCPR from the generic analysis is an average value, it is expected that some AMCPR will be larger.
This is ROD SWAP considered an expected statistical fluctuation.
O l
l l
l t O 15B.7-3/15B.7-4 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
[J 15B.8 MLHGR CONSIDERATIONS In addition to the MCPR response to a RWE event, the linear heat generation rate (LHGR) response must be evaluated to determine the most limiting of the two parameters.
The current applicable licensing basis for LHGR is the 1% plastic strain limit on the fuel cladding.
All GE fuel is designed to performance limits which have >60% margin to the 1% plastic strain limit.
For pur-poses of the BWE generic analysis, the differences between the 1%
plastic strain limit and the performance limits are defined as the dosi i basis ALHGR (ALHGR DB Performance limits are a complex function of fuel enrichment, residence time, gadolinia concentration and nodal exposure.
- Thus, the nodal ALHGR during ganged rod withdrawals must be evaluated to ensure with a high degree of confidence that the ALHGR is not DB exceeded.
bd The three-dimensional BWR simulator code calculates the margin to performance limits after each withdrawal increment.
The reduction in the margin over each withdrawal increment was then evaluated.
As long as the 95% probability at the 95% confidence level value of the maximum margin reduction during an unrestricted withdrawal is <60%, the ALHGR s not exceeded and the LHGR is not the DB limiting parameter.
Note that analyzing margin reduction conser-vatively assumes no credit for initial (pre-RWE) operation below performance limits.
Ten of the most limiting cases (based upon core MLHGR response) observed in the development of the AMCPR l
database (Subsection 15B.4.3.3.2.1) were re-evaluated for margin reductions to performance limits.
Figure 15B-18 shows a repre-sentative margin reduction as a function of feet withdrawn.
l Typically, the maximum margin reduction occurs between 0 and 6-8 f t withdrawn.
l 15B.8-1 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
158.8 MLl!GR CONSIDERATIONS (Continued)
The results of the LIIGR analysis are given in Table 15B-13.
The 95% probability /95% confidence level maximum margin reduction over any 2-ft withdrawal increment is 34.2%.
Since this is less than the 60% margin to 1% plastic strain, it was concluded that the LilG R is not the limiting parameter.
O l
i l
l I
O 15B.8-2
3 I
GESSAR II 22A7007 l
238 NUCLEAR ISLAND Rev. O l
r 1
s i
h 15B.9 REFERENCES i
i 1.
W.
R. Morgan, "In-Core Neutron Monitoring System for General i
Electric Boiling Water Reactors", April 1969 (APED-5706).
f l
2.
C. J.
Paone, " Banked Position Withdrawal Sequence," General Electric Company, January 1977 (NEDO-21231).
l 3.
J.
A.
Woolley, "Three-Dimensional BWR Core Simulator",
f General Electric Company, May 1976 (NEDO-20953).
l 1
i 4.
W.
B. Nelson, et al, "STATPAC - A General-Purpose Program for i
Data Analysis and for Fitting Statistical Models to Data",
General Electric Company, May 1972 (75 GEN 012).
5.
" General Electric BWR Thermal Analysis Basis (GETAB):
- Data, I
Correlation and Design Application", January 1977 (NEDO-10958-A).
I l
l O
15B.9-1/15B.9-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/%
(_)
Table 15B-1 SEQUENCE OF EVENTS - RWE IN POWER RANGE FOR BWR/2-5 Elapsed Time (sec)
Event 40 Core is assumed to be operating at rated conditions.
NO Operator selects and withdraws maximum worth rod.
s1 Initial RBM signal is normalized to reference APRM and rod withdrawal starts.
%1 Total core power and local power in vicinity of rod increases.
%5 LPRM system indicates excessive localized peaking.
%5 Operator ignores warning and continues withdrawal.
/s N15 RBM System indicates excessive localized peaking.
O,
%15 Operator ignores warning and continues withdrawal.
%20 RBM System initiates a rod block, inhibiting further withdrawal.
440 Reactor core stabilizes at higher core power level.
gs
' L]
15B.10-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O j
Table 15B-2 BWR/6 RWE TRANSIENT SEQUENCE OF EVENTS Elapsed Time (sec)
Event 0
Core is operated in a typical control rod pattern on limits.
0 Operator withdraws a single rod or gang of rods continuously.
n1 The local power in the vicinity of the withdrawn rod (or gang) increases.
Gross core power increases.
%8*
RWL blocks further withdrawal.
%25 Core stabilizes at slightly higher core power level.
- For a 2-ft RWL incremental withdrawal block.
Time would be longer for a larger block, since rods are withdrawn at approximately 3 in./sec.
O O
15B.10-2 i
}
i GESSAR II 22A7007
)
238 NUCLEAR ISLAND Rev. O l
1 Table 15B-3 BWR/6 STANDARD PLANTS INCORPORATED IN THE IMCPR DATABASE i
Plant Core S{ze Cycle Enrichment l
l f
A 624 Initial Low I
B 624 Initial Med l
l C
624 Transition Med i
D 748 Initial Med l
l E
748 Equilibrium Med F
748 Equilibrium High G
800 Initial Low I
II 848 Initial Med O
k i
e t
l 1
I t
O 15B.10-3 1
4
- ww
.----ew--
_------,----------m----.---r-.--
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 15B-4 TYPICAL RESULTS FROM THE STATISTICAL FIT OF IMCPR Power Flow Nominal
(%)
(%)
IMCPR 95/50 100 100 1.27 1.17 80 71 1.49 1.35 60 44 1.67 1.46 40 36 2.31 1.99 20 38 3.46 2.93 O
O 15B.10-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
[)
Table 15B-5 v
(AMCPR/IMCPR)95/95 AND AMCPR TYPICAL VALUES OP 95/95 2 ft 6 ft Case Power Flow
-AMCPR-
-AMCPR-AMCPR AMCPR 95/95 No.
(%)
(%)
_ IMCPR_95/95
_IMCPR_ 95/95 95/95 1
100 100
-0.14
-0.17 2
80 71
-0.16
-0.22 3
80 100
-0.18
-0.24
-0.34
-0.46 4
60 44
-0.18
-0.27
-0.36
-0.53 5
60 100
-0.22
-0.40
-0.42
-0.76 6
20 32
-0.25
-0.74
-0.49
-1.45
- Uncertainties and biases incorporated.
- AMCPR at the specific power and flow state, as opposed to 95/95 AMCPR95/95(P) of Equation 15B.4-5, which is calculated fcir maximum core flow at power P.
s O
15B.10-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 15B-6 IMCPR TECHNICAL SPECIFICATION REQUIREMENTS FOR OPTIMUM WITHDRAWAL DISTANCES Power ra a ncrement IMCP
(%)
TS (ft) 100 1.17 1
80 1.25 1
70 1.41 2
60 1.46 2
40 1.53 2
20 1.51 2
O l
15B.10-6 I
GESSAR II 22A7007 i
238 NUCLEAR ISLAND Rev. 0
(
Table 15B-7 TYPICAL RESULTS OF BWR/6 PLANT-SPECIFIC DETERMINISTIC RWE ANALYSES Power Allowable Withdrawal
(%)
Plant Size Enrichment Distance (ft) 100 800 Low 1.5 100 624 Low 1.0 100 624 Med 1.0 100 748 Med 1.5 70 800 Low 2.0 70 748 Med 2.5 i
O O
15B.10-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 15B-8 TYPICAL BWR/6 MCPR RESPONSE TO THE WITHDRAWAL OF IIIGli WORT!! AND LOW WORTl! CONTROL ROD GANGS Distance Power Flow Withdrawn Iligh Worth Low Worth Peripheral
(%)
(%)
(ft)
Gang Gang Gang 40 40 2
-0.20
-0.16 (second row of rods 6
-0.65
-0.53 boundary) 60 100 2
-0.06
-0.06 6
-0.48
-0.31 12
-0.65
-0.51 55 110 2
-0.18
-0.12
-0.03 4
-0.48
-0.33
-0.10 6
-0.65
-0.53
-0.20 10
-0.67
-0.58 g
80 110 1
-0.05
-0.01 2
-0.13
-0.02 4
-0.26
-0.06 l
l I
I I
l l
l l
l l
l O
15B.10-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/~'s l
j Table 15B-9
,v COMPARISON OF AMCPRs FOR SINGLE AND GANGED ROD WITHDRAWALS
- AMCPR Distance "9
- Power Flow Withdrawn 4-Rod Highest Worth One Rod of
(%)
(%)
(ft)
Gang (Usually Central) 4-Rod Gang 40 40 2
-0.204
-0.032
-0.045 6
-0.648
-0.331
-0.360 80 100 2
-0.220
-0.097 4
-0.412
-0.136
-0.185 100 100 1
-0.029
-0.005 2
-0.055
-0.024
/N 60 100 6
-0.483
-0.056 i'-
12
-0.653
-0.243 80 110 1
-0.010
-0.005 2
-0.035
-0.012 4
-0.127
-0.030 80 70 1.5
-0.06
-0.034 4
-0.223
-0.094 40 40 2
-0.117
-0.032 6
-0.752
-0.331 12
-0.739
-0.350
- Uncertainties and biases not incorporated.
O 15B.10-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 15B-10 NOMINAL RWE RESPONSE AT SELECTED POWER / FLOW STATES Distance AMCPR Power Flow Nominal Withdrawn FINAL
(%)
(%)
IMCPR (ft)
NOMINAL AMCPR MCPR 80 71 1.48 1
0.04 0.06 1.42 80 100 1.51 1
0.04 0.07 1.45 60 44 1.67 2
0.10 0.17 1.50 60 100 2.06 2
0.10 0.20 1.85 40 36 2.31 2
0.12 0.28 2.03 40 100 2.92 2
0.12 0.35 2.57 9
i f
9 15B.10-10 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/~ N
(
Table 15B-11
\\_/
IMPACT OF XENON DISTRIBUTION ON IMCPR Power Flow Nominal IMCPR
(%)
(%)
Equil. Xenon Xenon-Free Rated Xenon 20 38 3.46 3.83,4.66 (2.93-4.00)*
30 36 2.83 2.90,2.84 (2.41-3.25)*
2.85 40 40 2.40 1.97,2.29 2.56,2.58 (2.07-2.73)*
60 40 1.60 1.59,1.52 1.78,1.67 (1.39-1.80)*
1.65
/"'\\
t, i
80 70 1.49 1.63 m/
(1.35-1.63)*
- 95% PROB to 5% PROB values determined from a statistical analysis of the entire equilibrium xenon IMCPR database without uncertain-ties and biases incorporated.
(
'w) 15B.10-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 15B-12 AMCPRs FOR GANG ROD WITHDRAWALS DURING ROD SEQUENCE EXCHANGES AMCPR*
Normal Operating Rod Pattern Case Power Flow Rod Sequence Design Basis No.
(%)
(%)
Exchange Nominal 95/95 1
40 40
-0.619
-0.78
-1.06 2
50 40
-0.664
-0.57
-0.78
- AMCPR for 12-ft gant; rod withdrawal without uncertainties and biases incorporated.
O O
15B.10-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I
't i
i Table 15B-13 LIIGR PERFORMANCE LIMIT MARGIN REDUCTION
SUMMARY
1 l
Case Power Flow 3
No.
(%)
(%)
2 ft 1
30 36 8.26 2
20 32 15.04 3
100 100 18.62 4
40 40 10.68 5
60 115 30.02 1
6 100 80 12.04 j
7 60 70 12.62 8
100 100 8.91 l
9 80 70 7.97 10 97.3 100 19.45 Mean 14.36 1
h Standard Deviation 6.83 0
34.16 95/95 I
- A2 ft maximum decrease in margin for the limiting 2-ft
=
l withdrawal increment.
l l
9 O
15B.10-13/15B.10-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 BASE ROD (1)
PATTERNS l f 1 r i f GANGED ROD IMCPR FIT AS FUNCTION OF (2)
WITHDRAWAL (4)
SIMULATION POW E R, F LOW 1 r 1 r AMCPR FIT AS PROJECTED IMCPR FUNCTION OF CAPABILITY g3)
POWER, F LOW, (5)
(IMCPR95/50)
FEET WITHDR AWN, IMCPR
%j 1 P MAXIMUM ALLOWABLE WITH DR AWAL DISTANCES AS (6)
A FUNCTION OF POWER 1 r OPTIMUM WITH DR AWAL III DISTANCES AS A FUNCTION OF POWER 1 f R EQUIR ED IMCPR TECHNICAL (8)
SPECIFICATIONS Figure 15B-1.
Generic Procedure Flow Diagram g(
15B.10-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0
120 AEGION OF APPLICABILITY 110 REFERENCE BWR/6 OF GENERIC ANALYSIS OPER ATING M AP 100 pO-O. -
90 00 '
\\
7
/
O
\\
80 O
/
6 00 Q 5
/ O
\\
- 70
/O O
O O
8*
I 8
7O O
p 0 50 e
5 O
/
/
40 O
Oh O
30 O
O 20 0
10 l
l l
l l
l l
l l
l g
l 0
10 20 30 40 50 60 70 80 90 100 110 120 PERCENT CORE FLOW Figure 15B-2.
IMCPR Datapoint Locations l
l i
O 1
l ISB.10-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I9 2.7 1.8 2.6 60% POWER /44% FLOW 1.7 2.5 1.6 0% POWER /71% F LOW i
2 N
~
E 40% POWER /36% FLOW 2
l 8
~
l z j,5 2.3 z 5
Oz
{
i 1.4 2.2 100% POWER /100% FLOW 1.3 -
yg 1.2 l
l l
l l
2.0 0
20 30 40 50 60 70 NUMBER OF IMCPR DATAPOINTS Figure 15B-3.
Nominal IMCPR as a Function of Database Size 15B.10-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0
3.39 y
3.28 3.00 2.73 2.51 2.29 2 00 e
1.85 e
1.63 e
e 1.41 1.19 e
a i.000 a
e2 g 0860 e.
e r- 0 640 2
0.420 3
3 0.200 2
2 0
2.
e
-0.130 2
ee
- 0.350
- 32 22
-0 570 2e 23
- 0.790 2
-1.00 2
e
-1.12 ee e
-1.34 e2 e e
-1.56 e
-1.78 e
I l !
t I
f l
f i
f f
i i
-2 00 0.1
- 0. 5 1 2 5
10 20 30 50 70 80 90 95 98990.5 0.9 CUMULATIVE PROBABILITY Figure 15B-4.
Probability Plot for the IMCPR Data Base 15B.10-18
GESSAR II 22A7007
)
238 NUCLEAR ISLAND Rev. O i
120 110 1.27 100 1.30 90 E 80 h
s'
""~_________
o s'-
70 A
s',s' Eo s
u s
/
,/
-~~----~
~
y GO s'
{
1.50 /
u
,M
$ 50 1.70 e'
""~
s'
,s**,-
e a
2 00 '
,s' 40 B
"~
2.40 '
30 s"
2.80 ' a 20 3.30 " #
10 I
I I
I 0
O 10 20 30 40 50 60 70 80 90 100 110 PERCENT CORE FLOW Figure 15B-5.
Nominal IMCPR as a Function of Power and Flow t
t i
1 15B.10-19 i
1
~-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 30 28 26 24 2.2 l
l g
2.0 l
8
\\
~
O 1.8 l
10 14 1.2 I
I f
I 10 O
20 40 60 80 100 120 CORE POWER (%)
Figure 15B-6.
BVR/6 IMCPR(P)95/50 Curve ISB.10-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O
30 11 1
\\
28
\\
\\
EQUlVALENT BWH/4 NOMIN AL iMCPR 2.6
{
(1.052
- BWR/8 NOMINAL IMCPR)
\\
\\
2.4 NOMIN AL IMCPR
\\
\\
2.2 EQUlVALENT DWR/4 k
IMCPR CAPABILITY ti.052
- eWR/s IMCPRggg)
E 20 g
\\
1.8 k
OPERATING PLANT DATA
\\ k 16 OO O
PROJECTEo /
\\
00 B W R/G 1.4 IMCPR 95/50 1.2 I
1.0 O
20 40 60 80 100 120 140 CORE POWER (%)
Figure 15B-7.
Comparison of Operating Plant Data and Projected IMCPR Capability 15B.10-21
...,,.e
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1 40 1.35 1.30 1.25
[
N 9
E 120 2;
O 8
1,15 1.10 I
135 l
I I
I I
- oo o
2 4
6 8
to 12 INCHEMENTAL HOD WITHDH AWAL DISTANCE (f t) j vioure 15B-8.
Typical MCPR Response to a Ganged Itod Withdrawal 15B.10-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
120 BOUNDARY OF APPLICABILITY 110 OF GENERIC ANALYSIS 100 O
A - O - --
/
\\
~
/
\\
/
O
@k (o
80 o
f w
/
\\
h 70 f(f O
\\
E f
O
\\
h" l
cn)
$ s0 s'
r O
o 4
O 30 O
s 10 I
I I
I I
I I
I I
I o
O 10 20 30 40 50 60 70 80 90 100 110 120 PERCENT CORE FLOW Figure 15B-9.
Location of AMCPR Datapoints O
15B.10-23
GESSAR II 22A7007 238 IJUCLEAR I SLAtJD Rev. 0 0
-0.10
-04
-0.14 (20,32 5)
+
-0.12
-0.3
-0 10 N
J<
z
\\
(70.113.2.2')
_2 2
E b
-008
-02 g b
0 8
n 2
- 0.06 x
( _.,, 0..,
-O 02 i
[
0 O
0 20 30 40 50 60 70 NUMBER OF DATAPOINTS Figure 15B-10.
Nominal liMCPR/IMCPR as a Function of Database Size 158.10-24
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0
\\
\\
i
\\
g
\\
\\
\\
E
\\<o
\\
o g
W s
g
[
\\
\\ O 0
\\
\\
v
\\
\\
k
.S
\\
\\
\\
~
~
e"
\\
g o\\
o o) 5 b
O\\o
\\
O N
0<
9
_ 2 8
3 O
s N
\\
~
e m
<\\
\\
< D
\\
M N
\\
g
\\
\\S 9
oI m
W
\\
\\
Z odS
\\
\\
e,,,
\\
\\
o
\\
\\
\\
5a
= "
jo0 4
\\
\\
\\
o g
\\
g E
g i
m N
\\
\\\\s\\
I
\\\\
\\- d
\\\\
l
\\\\g
\\
l r$ o l
l I
I I
n o
m e
v n
o Ud3WT' 15B.10-25
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0
0.3 0.2 k
E M
1
.1MCPHgg 01 l
I i
l I
l I
O O
1 2
3 4
5 6
7 DISTANCE WITHDRAWN (f t) l Figure 15B-12.
lMCPR s a Funcdon of MMrawal 95/95 Distance at 100% Power l
15B.10-26 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
14 12 10
.i g
U i
8 t;
5
_J<m
$o 6 t-3:
4 2
O f
f I
t O
20 40 60 80 100 120 140 CORE POWER (%)
Figure 15B-13.
Maximum Allowable Withdrawal Distance ISB.10-27
Omet>W H NNyWoo4 nn a3 NLC 2Coe1>W H h C
%O<.
a t
e ym yC I
)
ft
. S
(
E E ST I
G R N
N U I
AS E
I S.
G O
RS P
A T
S H WE S
T P S K
N B N C
I F PO O
L LOS B
A E
T D
o N N R O
O I EE R
TMW CER NC E
UR L
FOB TF A ON T E P N SE L CC WPC oe i
RRA O
C 8
nO 8
._8
" d.
c
,Mam2* 8mm O2m D
n 0
n t-tHOOF iOemOPDn[
@(
- nOcPOD Om oOMO TOEOn L
Mr-QCnO
[i e mOG HUw HO'MO
-=
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
1.6 1.5 1,4 O
N f
1.3 y
t.2 1.1 I
I I
1.0 O
20 40 60 80 100 120 140 CORE POWER (%)
O Figure 15B-15.
Rod Withdrawal Limiter IMCPR Technical Specification as a Function of Core Power 15B.10-29
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
f ( AMCPR)
DISTRIBUTION B DISTR!8UTION A I
G i
I 1
l I
I l
.:J I
I I
I I
2 AMCPR (X) l Figure 15B-16.
Comparison of t2MCPR Frequency Distributions for a Random Sample and Biased Database 15B.10-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/m v
59 55 14 10 14
]
51 36 36 36 36 14 10 2
10 14 47 43 36 36 32 32 36 36 39 14 10 2*
O 2*
10 14 35 36 32 30 30 32 36 31 10 2
0 2
0 2
10
/
- \\
/
\\
27 36 32 30 30 32 36 23 12 0
2*
O O
23 10 14 19 38 36 34 36 38 12 4
4 12 15 38 36 38 11 07 8
8 I
03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58
- G ANG WITHDR AWN O.
I Figure 15B-17.
Rod Pattern Prior to Gang Withdrawal During A Rod Sequence Exchange 15B.10-31
GESSAR II 22A7007 l
238 NUCLEAR ISLAND Rev. 0 O
35 30 25 7
z9 20 U
a E
z 15
we r Flow Exposure No.
(%)
(%)
IMCPR (mwd /t)
Plant
- l 67 60.4 70 1.9800 0
F i
68 80 112.5 1.5300 0
F 69 20.3 32 3.9500 0
F l
l 70 100 100 1.3900 0
F i
71 60 115 1.9700 0
F i
1 4
i l
- Plant designations correspond to Table 15B-3.
l I
i l
i t
e i
i l
O 15BA-3/15BA-4
~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f
O h
l I
1 I
l ATTACIIMENT B To i
t l
APPENDIX 15B l
IMCPR DATABASE CROSSPLOTS O
i i
h l
I
]
l I
O l
1 r'
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
('N
'NUM BE RS IN DICAT E NUMBER OF DATA POINTS AT THAT LOCATION CORE POWER (%)
NO.
CELL IN LOWER ROW ENDPT 20 30 40 50 60 70 80 90 100 ABOVE 3.2 1
3.0 1
2.8 2.6 2.4 1
2.2 1
2.0 2
18 1
1 2
1.6 1
1 4
2 1.4 1
1 1
1.2 1
1 1.0 1
g 5
0.8 1
1 1
1 1
0 3
06 1
1 1
[
5 0.4 1 1 1
1 1
E 5
02 1
1 1
1 1
5
- 0.0 2
1 1
1 3
- 0.2 1
1 1
8
-04 3
3 1 1
6
-06 3
1 1
1 8
-08 3
1 1
1 1
1 2
- 1.0 1
1 3
- 1.2 s
1 1
4
- 1.4 2
1 1
3
-16 1
1 1
1
-18 1
BELOW NO. IN COL.
5 2
12 1 4
13 1 2 2 1 13 1 2
29 71 TOTAL 3
V Figure 15BB-1.
Crossplot of Fit Residual Versus Core Power 15BB-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 e
'NUMBE RS IN DICATE NUMBER OF DATA POINTS AT THAT LOCATION CORE FLOW (%)
NO.
CELL IN LOWER ROW E NDPT 40 50 60 70 80 90 100 110 ABOVE 32 1
30 1
2.8 26 2.4 1
2.2 1
2.0 2
1 1
2 1.6 1
1 2
1.4 1
1 1
1.2 1
3 1
10 1
5 08 1 1 1
1 1
y 3
06 1
1 1
y 5
0.4 1
3 1
t 5
0.2 1
2 1
1 5
-00 1
2 2
3
- 0.2 1
1 1
8
-04 11 1
2 3
6
-06 1
1 2
2 8
-08 2 1 1
1 1
1 1
1 2
- 1.0 1
3
- 1.2 1
1 1
4
-14 3
1 3
- 1.6 1
1 1
1
-18 1
411132 1 2 1 3
14 2 1 2
17 4 1 1 1 71 TOTAL l
Figure 15BB-2.
Crossplot of Fit Residual Versus Core Flow 15BB-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1
l l
'NUMBE RS IN DIC AT E NUMBER OF DATA POINTS AT THAT LOCATION PLANT NO.
IN HOW A
F E
G 8
O H
C ABOVE 3.2 1
30 1
28 26 2.4 1
2.2 1
20 2
1.8 1
1 2
16 1
1 3
2 1.4 2
1
- 1. 2 1
1
,9 1
1.0 y
5 08 1
1 1
1 1
b 3
06 1
1 1
5 04 1
1 2
1 5
02 2
1 1
1 5
-00 1
3 1
3
- 0.2 1
1 1
8
-04 1
1 1
1 3
1 6
-06 2
1 1
1 1
l 8
- 0.8 1
1 1
1 1
1 2
2
- 1.0 1
1 1
3
- 1.2 1
1 4
-14 1
2 1
3
- 1.6 1
2 1
-18 1
BELOW NO IN COL.
11 6
10 7
12 10 5
10 71 TOTAL Figure 15BB-3.
Crossplot of Fit Residuals Versus Plants 15BB-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
'N UM BE RS IN DIC AT E NUMBER OF DATA POINTS AT TH AT LOCATION CYCLE AVERAGE EXPOSURE (MWD /T)
N O.
CELL IN LOWER ROW ENDPT 0
1.0 2.0 30 4.0 50 60 7.0 ABOVE 3.2 1
30 1
2.8 2.6 2.4 1
2.2 1
2.0 2
18 2
2 16 1
1 2
1.4 1
1 1
12 1
y 1
1.0 1
5 08 1
1 1
1 1
3 06 2
1 E
5 04 1
1 1
2 h
5 02 1
1 2
1 5
- 0.0 2
1 1
1 3
- 0.2 2
1 8
-04 2
2 2
1 1
0
-06 1
1 3
1 8
-08 2
1 2
1 2
2
- 1.0 1 1 3
-12 1
1 1
4
- 1.4 1
1 1
1 3
-16 2
1 1
- 1.8 1
15 1 10 8
12 9
8 6
2 71 TOTAL Figure 15BB-4.
Crossplot of Fit Residuals Versus Cycle Average Exposure 15BB-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
'NUMBE RS INDICATE NUMBER OF DATA POINTS AT THAT LOCATION CYCLE NO.
CELL IN LOWER ROW E NDPT 3.0 55 8.0 10.5 13.0 15.5 ABOVE 3.2 1
30 1
2.8 2.6 2.4 1
2.2 1
2.0 2
1.8 1
1 2
1.6 2
2 1.4 2
3 1
1.2 1
1 1.0 1
h 5
08 1 2 2
3 V6 2
1 y
5 04 4
1 5
02 4
1 5
- 0.0 2
3 3
-02 2 1 8
-0.4 6 1 1
6
-0.6 3 1 2
8
-08 4 2 2
2
- 1.0 2
3
- 1.2 1 1 1
4
- 1.4 3
1 3
- 1.6 1 2 1
- 1.8 1
44 11 16 71 TOTAL Figure 15BB-5.
Crossplot of Fit Residuals Versus Cycle 15BB-5
GESSAR II 22A7007 238 tJUCLEAR ISLAIJD Rev. O O
' NUMBERS IN DICATE NUMBE R OF DATA POINTS AT THAT LOCATION HOO SE QUENCE N O.
CELL IN LOWER HOW ENDPT A1 A2 01 02 ABOVE 32 1
30 1
2.8 2G 24 1
22 1
20 2
18 1
1 2
16 1
1 1
2 14 1
1 g
1 12 1
f 1
10 1
9 5
08 1
3 1
h 3
06 3
b 5
04 1
3 1
5 02 1
2 1
1 S
-- 0 0 1
1 1
2 3
-02 1
1 1
8
-04 2
2 3
1 6
-06 1
2 1
2 8
-08 1
3 2
2 2
-10 2
3
- 1.2 3
4
-14 1
2 1
3
-16 2
1 1
-18 1
BELOW NO IN COL.
12 33 14 12 11 TOTAL Figure 15BB-6.
Crossplot of Fit Pesiduals Versus Rod Sequence 15BB-6
GESSAR II 22A7007
{
238 NUCLEAR ISLAND Rev. 0
' NUMBERS INDICATE NUMBER OF DATA POINTS d
AT THAT LOCATlON f
ENRICHMENT E XPOSURE INTE RVAL NO.
CELL NO CELL IN LOWER IN LOWEH HOW E N DPT BOC EOC MOC ROW ENDPT HI LOW MED ABOVE ABOVE 3.2 32 1
30 1
1 3.0 1
2.8 28 2.6 l
26 2.4 2.4 1
2.2 1
1 22 1
20 f
20 2
1.8 1
1 2
1.8 1
1 f
2 1.6 1
1 2
16 2
2 1.4 1
1 2
1.4 2
1 1.2 1
1 1.2 1
g 1
1.0 1
y 1
1.0 1
5 08 3
1 1
3 5
08 1
4 g
3 06 3
{
3 06 1
2 y
E 5
04 2
2 1
E 5
04 1
4 5
02 1
2 2
h 5
02 4
1 5
-00 3
1 1
5
- 0.0 1
4 3
- 0.2 2
1 3
-0.2 1
2 8
-04 4
1 3
8
-0.4 1
2 5
6
-06 2
4 6
- 0.6 1
5 8
-08 5
3 8
- 0.8 1
2 5
j 2
-10 2
2
- 1.0 1
1 3
-12 2
1 3
- 1.2 1
1 1
4
-14 2
1 1
4
- 1.4 4
3
- 1.6 1
2 3
-16 1
2 1
8 1
1
-18 1
38 9
24 NO. IN COL.
6 18 47 71 TOTAL 71 TOTAL Figure 15BB-7.
Crossplot of Fit Figure 15BB-8.
Crossplot of Fit Residuals Versus Residuals Versus Exposure Interval Core Average Enrichment 15BB-7/15BB-8
GESSAR II 22A7007 j
238 NUCLEAR ISLAND Rev. O i
1 1
' O l
i i
1 i
I h
l
}
1 i
1, 1
1 i
1 i
i 1
l i
ATTACilMENT C i
l TO APPENDIX 15B 4
OPERATING PLANT IMCPR DATA I
i 1
l ll l
1, 4
I h
6 L
l i
I e
h i
t i
P h
{
O E
e B
-w - m-g e
,ww ew~~r_
~,e-m-
mm
- mm_,
n-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Power Flow Plant
(%)
(%)
IMCPR*
1 97.4 99.3 1.35 94.7 95.0 1.34 71.7 59.1 1.43 73.6 61.1 1.44, 47.1 42.0 1.98 48.6 42.0 1.85 52.8 42.0 1.85 58.7 46.1 1.71 57.8 45.9 1.74 59.1 46.3 1.72 97.8 99.7 1.33 96.8 99.6 1.34 81.2 71.4 1.42 88.7 81.5 1.44 24.5 34.3 2.99 2
95.0 99.0 1.44 96.0 99.0 1.41 3
98.7 100.0 1.42 97.1 95.9 1.45 96.0 90.4 1.38 100.0 99.3 1.32 99.3 99.9 1.30 99.8 94.6 1.30 98.7 99.7 1.35 76.9 61.1 1.58 4
89.5 77.4 1.39 100.0 99.4 1.30 h(a
- Process Computer measured 15BC-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
Power Flow Plant (t)
(%)
IMCPR*
4 96.9 94.7 1.32 94.6 90.0 1.39 98.9 97.3 1.30 89.0 63.0 1.42 5
83.0 73.0 1.54 83.5 78.7 1.60 82.7 73.7 1.57 82.4 76.4 1.50 61.9 54.4 1.89 82.9 76.1 1.60 59.3 46.1 1.84 65.4 52.7 1.73 61.3 48.2 1.86 88.0 86.0 1.59 84.0 77.0 1.60 6
77.6 73.5 1.67 78.7 74.5 1.57 67.9 70.9 1.65 73.1 60.0 1.58 60.8 47.5 1.54 59.5 46.4 1.58 55.4 48.1 1.73 70.4 62.0 1.53 I
69.9 60.6 1.60 79.8 75.0 1.52 60.9 47.0 1.59 7
94.1 95.8 1.37 92.0 99.9 1.42
- Process Computer measured 15DC-2
GESSAR II 22A7007 238 NUCLEAR I.cLAND Rev. 0 Power Flow Plant
(%)
(%)
IMCPR*
7 88.8 78.3 1.40 79.2 74.4 1.50 89.7 78.4 1.40 8
26.6 40.0 3.49 i
64.1 54.1 1.84 43.2 31.2 2.22 61.1 53.7 1.82 j
9 96.4 100.3 1.38 s
10 78.0 72.0 1.53 87.0 75.8 1.49 73.5 62.9 1.47 i
11 77.0 66.0 1.51
)
i 83.0 79.0 -
1.50 99.0 100.0 1.37 88.0 85.0 1.59 i
97.0 100.0 1.39
- Process Computer mea'sured 1
e i
4..
f 0
15BC-3/15BC-4 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i >
a 1
f l
P ATTACIIMENT D 1
l TO l
i j
APPENDIX 15B l
l AMCPR DATABASE l
t
!O 4
)
'I j
l I
i l
i
[
t t
I l
l 1
/
t r
i T
f l
.=---r..-.
..mm -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
)
Distance Case Power Flow Withdrawn No.
(%)
(%)
AMCPR (ft)
Plan t*
1.1 60 100
-0.590E-01 2.0 D
1.2 60 100
-0.483 6.0 1.3 60 100
-0.653 12.0 2.1 60 100
-0.550E-01 2.0 D
2.2 60 100
-0.316 6.0 2.3 60 100
-0.513 12.0 3.1 61 40
-0.244 2.0 B
3.2 61 40
-0.524 6.0 3.3 61 40
-0.587 9.0 4.1 100 100
-0.860E-01 1.0 B
4.2 100
'.00
-0.159 2.0 4.3 100 100
-0.281 4.0 5.1 45 70
-0.122 2.0 D
5.2 45 70
-0.600 6.0
()
5.3 45 70
-0.934 12.0 6.1 60 70
-0.223 2.0 B
6.2 60 70
-0.530 6.0 6.3 60 70
-0.627 10.0 7.1 47.5 40
-0.132 2.0 G
7.2 47.5 40
-0.557 6.0 7.3 47.5 40
-0.784 12.0 8.1 97.3 100
-0.170E-01 1.0 E
8.2 97.3 100
-0.380E-01 1.5 8.3 97.3 100
-0.580E-01 2.0 9.1 80 70
-0.300E-01 1.0 E
9.2 80 70
-0.930E-01 2.0 9.3 80 70
-0.222 4.0 9.4 80 70
-0.318 6.0 10.1 80 100
-0.330E-01 1.0 E
10.2 80 100
-0.970E-01 2.0 10.3 80 100
-0.234 4.0
()
10.4 80 100
-0.332 6.0
- Plant designations correspond to Table 15B-3.
15BD-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Distance Case Power Flow Withdrawn No.
(%)
(1) t.M C P R (ft)
Plant
- 11.1 100 70
-0.540E-01 1.0 E
11.2 100 70
-0.102 2.0 11.3 100 70
-0.177 4.0 12.1 80 40
-0.540E-01 1.0 E
12.2 80 40
-0.112 2.0 12.3 80 40
-0.200 4.0 13.1 100 100
-0.770E-01 1.0 A
13.2 100 100
-0.104 1.5 13.3 100 100
-0.126 2.0 14.1 99 86
-0.300E-01 1.0 C
14.2 99 86
-0.540E-01 2.0 14.3 99 86
-0.920E-01 5.0 14.4 99 86
-0.100 7.0 15.1 100 100
-0.280E-01 1.0 C
15.2 100 100
-0.430E-01 1.5 15.3 100 100
-0.550E-01 2.0 16.1 80 70
-0.924E-01 2.0 C
16.2 80 70
-0.170 4.0 16.3 80 70
-0.220 6.0 16.4 80 70
-0.254 10.5 17.1 40 40
-0.220 2.0 C
17.2 40 40
-0.780 9.0 17.3 40 40
-0.780 12.0 18.1 40 70
-0.353 2.0 C
18.2 40 70
-0.835 6.0 18.3 40 70
-0.870 9.0 l
18.4 40 70
-0.860 11.0 19.1 60 70
-0.155 2.0 G
19.2 60 70
-0.470 6.0 19.3 60 70
-0.436 11.5 20.1 60 100
-0.273 2.0 C
20.2 60 100
-0.680 11.5
- Plant designations correspond to Table 15B-3.
15BD-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 I
Distance
( j Case Power Flow Withdrawn No.
(t)
(%)
AMCPR (ft)
Plant
- 21.1 40 70
-0.328 6.0 A
21.2 40 70
-0.244 8.0 22.1 70 90
-0.640 11.5 A
23.1 85 90
-0.175 2.0 G
23.2 85 90
-0.330 4.0 24.1 20 32
-1.276 4.0 A
24.2 20 32
-1.231 8.0 24.3 20 32
-0.704 12.0 25.1 55 100
-0.609 3.5 B
25.2 55 100
-0.981 7.5 25.3 55 100
-1.042 11.5 26.1 40 50
-0.421 4.0 A
26.2 40 50
-0.665 8.0 26.3 40 50
-0.501 12.0
[T 27.1 100 80
-0.920E-01 1.0 B
27.2 100 80
-0.166 2.0 27.3 100 80
-0.262 4.0 28.1 80 100
-0.220 2.0 B
28.2 80 100
-0.412 4.0 28.3 80 100
-0.517 6.0 29.1 60 70
-0.192 2.0 B
30.1 80 110
-0.550E-01 1.0 D
30.2 80 110
-0.132 2.0 30.3 80 110
-0.262 4.0 31.1 40 40
-0.117 2.0 D
31.2 40 40
-0.739 12.0 32.1 40 40
-0.160 2.0 D
32.2 40 40
-0.529 6.0 32.3 40 40
-0.531 11.5 33.1 70 45
-0.106 2.0 D
33.2 70 45
-0.225 6.0
- Plant designations correspond to Table 15B-3.
15BD-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Distance Case Power Flow Withdrawn No.
(%)
(%)
AMCPR (ft)
Plant
- 34.1 65.3 60
- 0. 6 6 0E-01 2.0 G
34.2 65.3 60
- 0.163 4.0 34.3 65.3 60
- 0.211 6.0 35.1 100 110
- 0. 6 00E-0 2 1.0 G
35.2 100 110
-0.250E-01 2.0 35.3 100 110
-0.730E-01 4.0 36.1 40 70
- 0.514 2.0 G
36.2 40 70
-1.078 6.0 36.3 40 70
-1.191 10.5 37.1 97.5 100
-0.320E-01 1.5 E
37.2 97.5 100
-0.500E-02 2.0 38.1 20 32
-0.502 2.0 F
38.2 20 32
-1.439 6.0 38.3 20 32
-1.156 12.0 39.1 80 70
-0.690E-01 2.0 F
39.2 80 70
-0.265 6.0 40.1 80 110
-0.100E-01 1.0 F
40.2 80 110
-0.350E-01 2.0 40.3 80 110
-0.127 4.0 F
41.1 80 110
-0.200E-01 1.0 i
41.2 80 110
-0.630E-01 2.0 41.3 80 110
-0.156 4.0 l
42.1 55 110
-0.187 2.0 F
42.2 55 110
-0.485 4.0 42.3 55 110
-0.674 10.0 43.1 55 110
-0.121 2.0 F
43.2 55 110
-0.337 4.0 43.3 55 110
-0.539 6.0 1
43.4 55 110
-0.586 10.0 44.1 100 100
-0. 6 3 0E-01 1.0 D
44.2 100 100
-0.125 2.0 l
44.3 100 100
-0.260 4.0 44.4 100 100
-t.301 8.0
- Plant designations correspond to Table 15B-3.
15BD-4
l l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
j l
Distance Case Power Flow Withdrawn No.
(%)
(%)
AMCPR (ft)
Plant
- i 44.5 100 100
-0.288 10.0 45.1 40 40
-0.204 2.0 D
i 45.2 40 40
-0.648 6.0 1
6 45.3 40 40
-0.614 11.5 46.1 100 100
-0.900E-02 1.0 F
l 46.2 100 100
-0.340E-01 2.0 46.3 100 100
-0.235 10.0 l
46.4 100 100
-0.223 12.0 l
l 47.1 60.4 70
-0.710E-01 2.0 H
47.2 60.4 70
-0.247 4.0 47.3 60.4 70
-0.476 6.0 48.1 80 112.5
-0.200E-01 1.0 H
48.2 80 112.5
-0.500E-01 2.0 I
48.3 80 112.5
-0.239 4.0 49.1 20.3 32
-0.831 2.0 H
49.2 20.3 32
-1.730 6.0 1
49.3 20.3 32
-1.551 9.0 49.4 20.3 32
-1.385 12.0 50.1 100 100
-0.290E-01 1.0 H
50.2 100 100
-0.550E-01 2.0 50.3 100 100
-0.920E-01 4.0 51.1 60 115
-0.352 2.0 H
51.2 60 115
-0.623 4.0 51.3 60 115
-0.705 6.0 52.1 20 32
-1.214 2.0 D
I i
52.2 20 32
-1.657 6.0 I
52.3 20 32
-1.438 9.0 52.4 20 32
-1.358 11.5 53.1 30 36
-0.526 2.0 A
53.2 30 36
-1.091 6.0 53.3 30 36
-0.946 9.0 53.4 30 36
-0.8680 11.5 i
- Plant designations correspond to Table 15B-3.
l l
l i
15BD-5/15BD-6 4
i
I l
GESSAR II 22A7007 l
238 NUCLEAR ISLAND Rev. 0 l
r I
f Ih l
,i l
}
I i
i 1
l' l
ATTACHMENT E TO APPENDIX 15B j
AMCPR DATABASE CROSSPLOTS l
i 4
I i
I I
l k
l E
I f
l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
- NUMBE RS IN DICATE NUMBER OF DATA POINTS AT THAT LOCATION CORE POWER (%)
NO.
CELL IN LOWE R ROW ENDPT 20 30 40 50 60 70 80 90 100 ABOVE 2.4 1
- 2. 2 1
1 2.0 1
1 1.8 1
2 1G 1
1 3
1.4 1
1 1
6 1.2 1 1 1
1 1
1 14 1.0 1
1 1
2 1
4 1 3 15
- 0. 8 2
1 7
2 1 2 N
15 0.6 1
4 1 2 1
2 1
3 11 04 2
1 1 1 2
1 1 2 6
0.2 1
1 1
1 1 1
13
-00 1
2 5 3
2
[
7 0.2 2
2 2
1 C
12
- 0.4 1
1 2 1 1 6
5
-06 2
2 1
10 0.8 3
2 1
4 8
-10 2
1 1
1 1
2 6
- 1. 2 1
1 1
3 7
- 1.4 2
1 2 1 1
5 16 1
1 1
2 8
- 1.8 1 4 1
1 1
2 20 2
1
- 2. 2 1
1
- 2.4 1
BELOW NO IN COL 14 4
23 3 31024 3
3 32 2 5
4 30 160 TOTAL Figure 15BE-1.
Crossplot of Fit Residuals Versus Core Power 15BE-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
- NUMBERS INDICATE NUMBER OF DATA POINTS AT TH AT LOCATION CORE FLOW (%)
NO.
CELL IN LOWER ROW ENDPT 40 50 60 70 80 90 100 110 ABOVE 2.4 1
2.2 1
1 2.0 1 1
1.8 1
2 1.6 1
1 3
1.4 1
1 1
6 1.2 1
1 2
1 1
14 1.0 1 1 1 1
2 1
2 5
15 0.8 2
3 1
5 2 2 g
15 0.6 1 1 1 1
5 5
1 11 0.4 2 1 2
1 3
2 j
9 6
- 0. 2 11 2
1 1
h 13
-00 2
5 3
3 t
7
- 0.2 2
3 1
1 122
-04 11 4 3
2 1
5
-06 1
1 3
10
- 0.8 1
4 1
4 8
- 1.0 2 2 4
6
- 1.2 1 1
4 7
- 1.4 21 1 3
5
- 1.6 11 3
8
- 1.8 1
1 2
1 3
2
- 2.0 2
1 2.2 1
1
- 2.4 1
BELOW NO. IN COL.
14420 2 3 3
35 3 4 3 44 19 3 3 160 TOTAL l
1 l
Figure 15BE-2.
Crossplot of Fit Residuals Versus Core Flow I
15BE-2 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
'NUMBE RS IN DICAT E NUMBER OF DATA POINTS AT THAT LOCATION PLANT NO.
IN ROW A
F E
G B
C H
C ABOVE 2.4 1
2.2 1
1 2.0 1
1 1.8 1
2 16 1
1 3
1.4 1
1 1
6 1.2 2
1 2
1 14 1.0 1
3 4
3 3
15 0.8 4
3 3
3 2
g 15 06 1
1 3
1 3
2 4
y 11 03 2
1 1
1 2
4 9
6 02 2
1 1
1 1
13
- 0.0 4
2 1
1 4
1 g
7 0.2 1
1 2
1 1
1 12 0.4 1
1 5
1 2
2 5
-06 2
2 1
10 0.8 1
4 1
3 1
8 1.0 1
1 1
1 2
1 1
6
- 1.2 2
1 1
1 1
7 1.4 2
4 1
5
- 1.6 1
2 2
8
- 1.8 1
1 3
3 2
- 2.0 2
1
- 2.2 1
1
-2.4 1
16 22 19 17 19 31 16 20 Figure 15BE-3.
Crossplot of Fit Residuals Versus Plant 15BE-3
GESSAR II 22A7007 238 NUCLEAR ISlu.ND Rev. 0 O
l
- NUMBE RS INDICATE NUMBER OF DATA POINTS AT THAT LOCATION CYCLE AVER AGE EXPOSURE (MWD /T)
NO.
CELL IN LOWER ROW ENDPT 0
1.0 2.0 3.0 4.0 50 60 ABOVE 2.4 1
2.2 1
1 2.0 1
1 1.8 1
2 1.6 1
1 3
1.4 1 1 1
6 1.2 1
1 1
1 1
1 14 10 1
2 4
3 2
1 1
15 0.8 3 3 3
1 2
1 2
15 06 3 3 1
3 1
3 1
g y
11 04 3 1 2
2 3
0 6
0.2 1 1 1
1 2
U 13
-00 2 2 3
1 1
3 1
a:
7
- 0. 2 2 1 1
1 2
g 12
- 0.4 2 5 1
3 1
5 0.6 1
1 1
2 to
-08 4 2 1
3 8
- 1.0 2
2 2
1 1
6
- 1.2 1 1 1
2 1
7
- 1.4 6
1 5
1.6 2
1 2
8
- 1.8 5 1 1
1 2
- 2.0 2
1
- 2. 2 1
1 24 1
B E LOW 1
NO IN COL.
42 21 17 13 20 17 18 12 160 TOTA L i
l Figure 15BE-4.
Crossplot of Fit Residuals Versus Cycle Average Exposure 15BE-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
'NUMBE RS INDICATE NUMBER OF DATA POINTS AT THAT LOCATION DISTANCE WITHDRAWN (FEET NO CELL IN LOWER HOW E N DPT 3.0 5.5 8.0 10.5 ABOVE 2.4 1
2.2 1
1 2.0 1
1 1.8 1
2 16 1
1 3
1.4 2
1 6
1.2 2
3 1
14 10 1
5 2
1 2
1 1
1 0
15 08 4
1 7
1 1
1 15 0.6 1
2 6 2
2 1
1 11 0.4 3
4 2
1 1
g 6
0.2 2
1 1
1 1
g 13
-00 2
2 5
2 1
1 m
7
-02 2
2 1
2 h
12 04 3
3 3
2 1
5 06 1
1 2
1 10 08 1
5 1
1 1
1 8
1.0 1
1 1
1 1
1 2
6
- 1.2 1
2 1
1 1
l 7
- 1.4 3
1 1
1 1
5 1.6 1
1 3
8
- 1.8 1
1 2
1 1
2 2
- 2.0 1
1 1
- 2. 2 1
1 2.4 1
l BE LOW NO. IN CO L.
18 4 48 1 23 1 26 1
1 4
6 5 2 1
8 11 160 TOTAL i
N Figure 15BE-5.
Crossplot of Fit Residuals Versus Distance Withdrawn 15BE-5
'NUMBE RS INDICATE NUMBER OF DATA POINTS AT THAT LOCATION EXPOSURE INTERVAL ROD SEQUENCE NO CELL NO.
CELL IN LOW ER IN LOWER ROW ENDPT BOC EOC MOC ROW ENDPT A1 A2 ABOVE ABOVE 24 24 1
2.2 1
1 22 1
1 2.0 1
1 24 1
1 1.8 1
1 1.8 1
2 16 1
1 2
16 1
1 3
1.4 2
1 3
1.4 2
1 6
1.2 3
1 2
6 1.2 2
4 b
14 1.0 10 2
2 14 1.0 5
9 3
15 08 12 1
2 3
15 08 1
14 15 06 5
10 2
4 15 06 7
1 7
nh 9
11 04 5
1 5
9 11 04 2
9
{]
6 0.2 3
2 1
6 02 2
4 t
13 00 8
1 4
b 13
-00 2
11 I
7
- 0. 2 4
2 1
7 0.2 2
5 12
- 0.4 10 1
1 12
-04 3
9 HH 5
06 2
3 5
-06 1
4 10 0.8 7
3 10
- 0.8 10 g
8
- 1.0 7
1 8
- 1.0 2
6 O
6
- 1.2 5
1 6
- 1.2 2
4 7
- 1.4 7
7
- 1.4 7
5
- 1.6 3
2 5
1.6 5
8
- 1. 8 7
1 8
1.8 8
2
- 2.0 2
2
- 2.0 2
1
- 2. 2 1
1
- 2.2 1
1 2.4 1
1
- 2.4 1
106 16 38 NO. IN COL.
34 126 160 IN TOTAL 160 TOTAL o>
Figure 15BE-6.
Crossplot of Fit Figure 15BE-7.
Crossplot of
< a Residuals Versus Residuals Versus 0"
Exposure Interval Rod Sequence O
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j GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O
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4 ATTACHMENT F
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APPENDIX 15B j
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RCIS POWER SIGNAL
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ATTACHMENT F V
TO APPENDIX 15B RCIS POWER SIGNAL 15BF.1 INTRODUCTION This section is a technical expansion of the information given in Subsection 15B.4.3.3.6.
It describes the power signal source input to the Rod Control and Information System (RCIS), those plant conditions under which that signal is biased, and the impact on the Rod Withdrawal Limiter (RWL) system.
In particular, the technical basis for the proposed plant technical specification (15B.4.3.3.b) and the conclusion that the current signal source is adequate is presented.
Related issues on the acceptability of core thermal power level instrumentation, bypass valve position indication, and the need for additional alarms are also discussed.
On U
15BF-1/15BF-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15BF.2 EFFECT OF BIASED POWER SIGNAL ON RCIS SYSTEM FUNCTION Figure 15BF-1 is a schematic diagram of relevant RCIS functions.
RCIS has two modes of operation.
At operating core thermal powers below the Low Power Set Point (LPSP), the Rod Pattern Control Sys-tem (RPCS) becomes functional.
RPCS enforces Banked Position f
Withdrawal Sequence (BPWS) constraints on control rod positions.
BPWS constraints reduce control rod incremental reactivity worths to values much less than those required to satisfy control rod drop accident design criteria.
This is accomplished by assigning all control rods to groups such that rods in any group are uni-formly dispersed over the core.
Further, the order of group selection, relative positions of rods within a group, and the relative positions of (banked) groups is tightly controlled by hardwired logic (Reference 2 of Section 15B.9).
Control rods can be moved singly or simultaneously by gangs (sub-groupings of one to a maximum of four control rods).
The longest withdrawal of a single or gang of control rods under BPWS, during the same control rod selection, is a 9-ft withdrawal from notch position 12 to 48 (from 3 ft withdrawn to the fully withdrawn position).
This is allowed only when the control rod pattern prior to the error meets all BPWS constraints.
If BPWS is violated, no rod movements (insertions or withdrawals) are per-mitted, eliminating the possibility of any RWE.
When core power input to RCIS exceeds LPSP, an automatic switchover to the RWL system is made (and vice versa).
In the RWL range, control rod patterns are modified from strict compliance to BPWS at powers near LPSP, to the rated patterns as power is increased.
It is estimated that only 25% of rated rod patterns comply with BPWS.
This means that, if when operating at a power significantly above LPSP, the RCIS input power signal is biased low such that RPCS enforces BPWS, it is likely that no rod movements will be possible, eliminating the possibility of a RWE.
However, in 15BF-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O those infrequent instances when BPWS is met, the worst RWE is limited to 9 ft.
When the RWL is functional, incremental rod withdrawals are limited to mitigate the potential consequences of a RWE.
Preset withdrawal increments are built into the RWL logic (Figure 15B-14) as a step function of core power level.
Between LPSP and HPSP (High Power Set Point;
%70% rated) a 2-ft withdrawal increment is enforced.
At powers greater than HPSP a 1-ft withdrawal increment is enforced.
Thus, if the core is operating above HPSP, but a low biased power signal indicated a value between LPSP and HPSP, RWL would allow a 2-ft withdravral instead of a 1-ft withdrawal.
Similarly if the low biased power signal is indicated below LPSP and BPWS is satisfied, a 9-ft withdrawal would be possible instead of 1-ft.
Therefore, for operation at core powers aaove LPSP (%20% rated),
the general effect of a low biased RCIS input power signal on RWL function is to potentially allow incremental rod withdrawals greater than intended in the design basis; and, when operating at any power level, high biased signals potentially restrict withdrawals to be shorter than intended.
From RWE mitigation considerations then, high biasing results in either the intended or conservative rod blocks and are not considered further.
Potential causes for low biasing are now examined in detail.
O 15BF-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
'~'
(s-)h 15BF.3 RCIS POWER SIGNAL SOURCE RCIS gets its core power input signal from a measurement of first-stage turbine pressure.
Figure 15BF-2 is a schematic showing the location of this measurement in the reactor vessel steam distribu-tion nystem.
The approximate relationship between first-stage turbine steamflow and core thermal power (CTP) is as follows:
CTP = C
- SF p
1 or equivalently, CTP % C lSF2 + Al + SF3 + SF4}
(15BF-1) where SF is the reactor total steamflow.
y C
is a proportionality constant as follows:
p
)
~~/
Cp % (h
-hgy)
(15BF-1.1) st where enthalpy of reactor steam, h
st h
E enthalpy of reactor feedwater.
l fy l
SP is steamflow through safety relief valves (SRV),
2 Al is the total extraction steam upstream of the turbine stop valves (TSV),
l l
SF is the steamflow bypassed to the main condenser, and 3
SF is the total steamflow delivered to the turbine first 4
stage.
In order to include the first-stage turbine pressure (P4) into the equation and allow investigation of the effect of a change in
(~N l
\\
l extraction steam downstream of the first-stage pressure measure-
%J ment (A
), the simplified design of Figure 15BF-3 is considered.
2 15BF-5 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O There is a single extraction between the highest pressure (HP) stages and lower pressure (LP) turbine stages.
The circuit diagram for the analogous electrical situation is presented, and analysis of this circuit gives the relationship between first-stage turbine pressure and first-stage turbine steamflow:
P 4 SF
(
}
4 C
R(A2) 2 y + C +RIO2 C
2 where C and C re constants (positive) 7 2
and R(A2) represents the functional dependence of extraction steamflow resistance to the extraction flow rate.
The relationship between first-state turbine pressure and core thermal power for this example is therefore:
O P
CTP % (h
-hgy)
SF2+dl + SF3+
4 g
C R(A2) l 2
E1 + C +R(A2)
(15BF-3) 2 (It should be noted that this equation and development is very simplistic and is intended only to indicate the qualitative effects of various system changes, and not to predict quantitative effects.)
For normal operations above %20% rated thermal power, the rela-tionship between first-stage turbine pressure and core thermal power is analyzed for the purposes of choosing appropriate pres-sure setpoints corresponding to LPSP and HPSP.
During the analysis to determine the pressure setpoints, all normal expected modes of plant operation near LPSP and flPSP that could impact the turbine O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 "N
pressure versus core power relationship are considered.
Normal conditions are considered to be as follows:
All safety relief valves fully closed (SF%0)
Normal extraction steam mode (s);
All bypass valves fully closed (SF 0).
3 Under these conditions, CTP % Ky+K2 4
- P (15BF-4) where K K
are constants for a particular normal extraction steam y,
2 mode [feedwater temperature].
When using Equation 15BF-4 to deter-mine the appropriate first-stage turbine pressure corresponding to RCIS setpoints, all normal feedwater system lineups that could
( ')
exist at the power range of interest (i.e.,
different values of K y,
'~
K) re considered.
For LPSP, an appropriate pressure setpoint is 2
one that assures that true core thermal power is greater than 20%
when RCIS switches f rom the RPCS to the RWL f.itetion.
This assures BPWS restrictions are not prematurely relaxed.
BPWS restricts the consequences of most RWEs to be insignificant, and there is a very high probability that the worst RWE allowed by BPWS even at 30%
power [ Case 1, Table 15BF-1] will be acceptable.
For HPSP, an appropriate pressure setpoint is one that assures that the true core thermal power is less than 70% when the RWL switches from allowing 2-ft incremental withdrawals to 1-ft withdrawals.
Therefore, all normal modes of plant operation (and power signal bias) are considered when choosing appropriate RCIS pressure signal i
setpoints.
Off-normal plant operating modes that can result in additional power signal biases are not considered during the selection of LPSP and HPSP pressure setpoints.
The effects of
[O h
off-normal power signal biases will now be addressed along with l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O the reasons for not considering them during the setpoint selection process.
15BF.3.1 Off-Normal Causes of Power Signal Bias 15BF.3.1.1 Safety Relief Valve Steamflow Safety relief valves would be open only under abnormal or transient conditions.
Safety relief valve steamflow would cause first-stage pressure to underpredict core power - low bias (Equation 15BF-3, 3P /DSF2< 0).
4 Consideration of such an occurrence with a simultaneous RWE is not a part of the RWL system design basis.
The probability of such an event would be much lower than that of a moderately frequent event (IC 1 occurrence / year) since a RWE and either a significant tran-sient, multiple operator errors, or a safety relief valve equipment malfunction have to occur simultaneously.
O Leakage of steam past safety relief valves would result in the same low power bias.
Ilowever, the amount of leakage that would be operationally acceptable would be very small and would result in insignificant pressure signal bias.
(SRV leakage would be much less than that of a single fully open bypass valve -
Subsection 15BF.4.)
For the above reasons safety relief valve steamflow is not considered to have a significant potential for compromising the RWL design basis.
15BF.3.1.2 Extraction Upstream of Turbine Stop Valves Above 20% rated power, the most likely occurrence would be a reduction in upstream extraction due to a loss of a piece of equipment requiring steam (e.g.,
steam-driven turbine).
This O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
{'
would result in a high biased power signal (Equation 15BF-3, 3P /3Ay < 0) and conservative RWL function as previously discussed.
4 However, justified operator actions in response to infrequent plant conditions, operator errors, or equipment malfunctions could result in greater upstream extraction than the first-stage power signal calibration state.
Operator errors and equipment malfunc-tions simultaneous with a RWE are not considered within the RWL design basis for previously discussed reasons (Subsection 15BF.3.1).
Justified operator actions to increase upstream steam extraction are not considered to significantly reduce RWL effectiveness for the following reasons:
1.
The frequency of such situations would eliminate such a RWE from the moderate frequency transient classification category; f
2.
Deliberate, conscious operator action is involved, in which case the impact on the plant heat balance and pressure signal calibration would be apparent.
- Thus, I
if the condition persists for a significant duration, the need for recalibration or setpoint adjustment should be evident.
15BF.3.1.3 Turbine Extraction Steam l
An increase in turbine extraction (A2) results in an overall reduction in resistance to steamflow through the turbine (Fig-
< 0).
Therefore, a low biased power ure 15BF-3, 3R(A2)/302 l
measurement will result (Figure 15BF-3, 3P /3A2< 0; Equa-4 tion 15BF-3).
Turbine extraction steam effects on RWL function, then are identical to those of upstream extraction, and the same reasons why these effects are judged acceptable apply (Subsection 15BF.3.1.2).
N 15BF-9 l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15BF.3.1.4 Bypassed Steamflow The effect of bypass steamflow on first-stage turbine signal bias (and on RWL) is identical to that for safety relief valve steam-flow (Subsection 15BF.3.1.1).
Bypass steamflow results in a low biased power signal (DP /DSF 0).
However, there are more plant 4
3 conditions to consider which open bypass valves than safety relief valves.
15BF.3.1.4.1 Deliberate Operator Action That category of situations, which results in the opening of bypass valves, is defined to consist of those normal, planned plant operations, in which the operator either manually opens bypass valves, or takes other action that is expected to open bypass valves by action of the plant control systems.
In these situations the operator is alerted that bypass valves could open and should be giving attention to control room instrumentation of bypass valve status (Section 15BF.6).
This is required in order to comply with the proposed plant technical specification (Sub-section 15B.4.3.3.6) against control rod withdrawals with bypass valves open above (LPSP) s20% reactor power.
Under such circum-stances there is no need to alarm bypass opening, since it is the expected condition.
Control room instrumentation available to the operator that alerts him when core power exceeds the LPSP level is as follows (Section 15BF.6):
(a)
Average Power Range Monitor (APRM)
(b)
Plant heat balance (manual or process computer)
(c)
Feedwater flow (d)
Reactor vessel steamflow 15BF-10
_~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (c)
First-stage turbine pressure
~_/
(f)
Plant generator output At very low core powers in the startup range, plant conditions can exist which introduce biases and/or large uncertainties in all the above measurements.
However, as the power level approaches the 20% rated level, all the above indicators should approach consis-tency.
If bypass valves are open, a consistent first-stage turbine pressure signal should imply lower-than-actual reactor power.
In any case, deliberate operation with bypass valves open at powers greater than 20% rated would be an infrequent occurrence that the operator would be aware of; if uncertainty in operating power level prevented assurance of whether power exceeds LPSP, rod withdrawals could be suspended until the infrequent condition cleared or until the uncertainty in core power measurement was acceptable.
The consequences of even a 50% error in power near 20% (30% instead of 20%) have an insignificant impact on RWE protection given by the s/
RWL (Section 15BF.4) and are considered acceptable and obtainable.
Examples of deliberate actions which could open bypass valves are as follows:
1.
Increasing reactor power by control rod withdrawals, or core-flow increases above the turbine load limit setpoint.
The load limit setpoint basically restricts the degree which the turbine control valve (TCV) will open.
As power is increased, the pressure regulator will bypass steamflow to the main condenser to maintain reactor pressure within normal operating limits.
An increase in the load limit setpoint will admit the bypassed steam to the turbine and close the bypass valves.
2.
A manual reduction in the load limit setpoint below the current operating power level will also open bypass x--
valves as above.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3.
Turbine stop valve (TSV), Main Steam Isolation Valve (MSIV), Turbine Control Valve (TCV), or Bypass Valve (BV) surveillance testing can all result in short duration or sustained bypass valve opening due to the operation of the pressure regulator.
4.
Manual bypass valve opening by operator.
15BF.3.1.4.2 Normal Operating Transients This class includes plant responses to normal operating transients which would open the bypass valves automatically without deliberate operator action.
The only event expected to open bypass valves in this category is a reduction in the load demand signal.
The pres-sure regulator will open bypass valves to maintain normal reactor pressure as the control valves close in response to the load demand reduction.
If operating in the Automatic Load Following (ALF) mode with the load demand signal controlled by the load dispatcher, the control system will run back core-flow automatically to reduce reactor power, and the transient pressure rise might not open bypass valves unless the load demand reduction is rapid.
Even for a rapid reduction, the bypass valves would be open only for a short duration (seconds).
If operating in the ALF with the load demand signal controlled by the operator, this becomes a deliberate operator reduction of load action (Subsection 15BF. 3. 4.1).
Since the only instances in this category which open bypass valves are of very short duration, RWL function is rot significantly compromised.
15BF.3.1.4.3 Abnormal Operating Transients Abnormal operating transients refers to the Chapter 15 moderately frequent events analyzed, and the less severe events that they bound.
This is considered to include single equipment failures of a bypass valve or the control system which would result in a bypass valve opening.
These events themselves represent single failures 15BF-12
/
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GESSAR II 22A7007 l
238 NUCLEAR ISLAND Rev.-0
.i f
or operator errors; and the probability of any occurring i
simultaneously with a RWE is remote, therefore not requiring I
i consideration within the RWL basis for a moderately frequent RWE.
l The significant perturbations in plant operations expected for most of these events (e.g.,
partial load rejection, equipment failures, alarms, etc.) would alert the operator of the abnormal condition.
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13BF-13/15BF-14 d
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15BF.4 CONSEQUENCES OF TECHNICAL SPECIFICATION VIOLATION (O
It has been explained why the RWL system reliability is not signifienntly degraded by potential biases in the input power s ig na l',
if the proposed technical specification is adhered to.
The reasons for believing that the plant instrumentation is ade-quate to all6w effective implementation of that technical speci-fication has*also been discussed.
This section establishes that in almost'all cases the expected consequences of a RWE even with a biased power signal (in violation of the proposed technical specification) are still inconsequential.
The final MCPR, F,
after a RWE can be expressed as:
F= I (1 -
)
(15BF-5) where I E expected (nomin
_e-RWE initial MCPR O(,)
(Equation 15B.4-1)
A AMCPR yE expected (nominal) or considered IMCPR (Equation 15B.4-3) 2 of the final value can be expressed as:
The variance,.a 7, (1-f) 2 2
,,7 (g )2 (15BF-6) o
=
0 where 0 E standard deviation of I (Equation 15B.4-2) 7 A
A and ay standard. deviation of 7 (Equation 15B.4 -4)
E Thus the expected'value of MCPR after a RWE that would be The allowed d;te to a biased power signal can be calculated.
probability that the final value exceeds any chosen value can
~
also be calculated.
Table 15BF-1 indicates that the probability
(_)
of boiling transition for'even the most inaprobable signal biases is very small.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Most domestic BWR/6 plants have a minimum specified 35% sheam bypass capacity.
Assuming that actual performance result.: in an extremely optimistic 401 bypass capacity, the most severe ceral-l 1er.ge s to the RWL system would be Cases 3 and 6 (Tabic 15BP-1).
i These cases require all bypass valves to open fully.itypically 4, j
5, or 6) and is considered unrealistic.
Cases 1 and 5 are con-sidered the more realistic instances of a single valve actuation (conservative 10's bypass capacity-per-valve assumed).
The results show that the probability of boiling transition occurring even without, or in violation of, the proposed technical e.pecification is remote.
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d GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4
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15BP. 5' FIRST-STAGE TURBINE PRESSURE SIG*iAL INSTRUMENTATION Sections 15BP.3 and 15BP.4 have established that the first-stage turbine pressure signal is functionally acceptable for application within the RC&IS.
This section describes the first-stage turbine pressure' signal:
1 1.
Equipment Classification The instruments are classified essential and consist of two channels of transmitters, trip units and alarms which are redundant and divisionally separated.
Both channels must trip to effect RC&IS function changes at 1
LPSP and HPSP.
2.
RC&IS Usage The use of the signal within the RC&IS is described in O
I Section 15BP.2.
L 3.
Qualificatior.
The instruments are qualified to meet the requirements of Table 7.1-6.
4.
Control Room Instrumentation The signal is continuously monitored and indicated on the front control room panel.
Any instrument out-of-service or any gross failure of the instrument is alarmed and indicated in the control room.
O 15BF-17/15BF-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 15BF.6 CONTROL ROOM INDICATION OF REACTOR POWER / BYPASS VALVES The proposed technical specification requires that the operator have sufficient knowledge of reactor power and bypass valve status to effectively implement the requirements.
This has been briefly addressed (Subsection 15BP. 3.1. 4.1) and is now considered in detail.
Control room instrumentation available to the operator and the use of these instruments to determine bypass val've status and core power are as follows:
1.
Status of each turbine bypass valve position is indicated on a control room panel.
The display, often in the form j
of status lights, current to the bypass servoes, and valve positions, indicates whether the valves are partially or fully open, or closed.
The operator has N
ready access to this information should the need arise.
2.
Average Power Range Monitor (APRM) is indicated and recorded in percent-of-core-power on the front control room panel.
The operator can read the power level directly from the indicator.
3.
Process computer code OD3 will provide the operator with core power status.
Upon demand, it takes the process computer approximately 1-2 min to calculate this infor-mation.
If the process computer is not available, the i
operator could perform a manual reactor heat balance to determine core power.
He would need to know feedwater flow rate, temperature, and reactor pressure, which are l
all available in the control room.
The feedwr.ter flow l
j rate and reactor pressure are indicated on the front panel.
The feedwater temperature is located on a panel
)
off to the side of the front panel.
15BF-19 j
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.
Feedwater and mainsteam flow rate indications located on the front control room panel will provide the operator with core power level.
lie can predetermine by extrap-clation (knowing rated conditions) the feedwater and mainsteam flow rates corresponding to the LPSP.
In the event the turbine bypass valves open, he can check these indicators to see whether the flow rates exceed the pre-determined LPSP flow rates and then act in accordance with the technical specifications.
5.
First-stage turbine pressure and plant generator output indicators are located on the front control room panel.
The operator can use these indicators to help determine turbine bypass valve status and reactor power level.
For example, should these indicators read decreasing or less than expected conditions, but the other reactor power indicating devices remain unchanged, he can con-clude the turbine bypass valve / valves have opened.
IIaving to assess reactor power level when the bypass valves open, the operator can use these same indicators to see if the power at the first stage of the turbine and out of the generator exceeds the LPSP.
If it does, he can conclude the reactor power is even greater (bypass capacity) and respond to the technical specification requirements.
O 15BF-20
GESSAR II 22A7007 i
238 NUCLEAR ISLAND Rev. 0 l
15BF.7 RC&IS SETPOINT SELECTION i
t The major biases in the first-stage turbine pressure signal due to atypical plant operating modes has been discussed at length.
Inac-curacies in the power signal under normal, typical expected operating conditions due to normal (instrumentation) sources are included in the implemented RC&IS LPSP and llPSP setpoints.
LPSP will be chosen
>20% and llPSP 170% in recognition of such effects.
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15BF-21/15BF-22 i
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{
15BP.8
SUMMARY
l 1
All known issues associated with the use of first-stage turbine j
pressure as the RC&IS input power signal have been addressed.
It
)
has been shown that the expecP.ed consequences of a Rod Withdrawal
)
I Error even with the most improbable signal biases are acceptable.
1
(
A technical specification is proposed to remove even the most l
remote chance of violating fuel integrity limits during a RWE concurrent with a biased input signal to the RWL system.
The usage of the first-stage turbine pressure signal within the RC&IS is totally adequate.
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13BF-23/15BF-24 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1
Table 15BF-1 CONSEQUENCES OF RWL MALFUNCTION l
k j
Case 1_
2 3_
4 1
6, I
Actual Thermal Power
(% Rated) 30 40 60 60 80 100 i
i Core Flow i
j
(% Rated) 36 36 45 45 70 100 l
Steamflow Bypassed
(% Rated) 110 120
%40
%40 110 R30 i
f RC&IS Sensed Power j
(% Rated) 120 120
%20 20 (70 (70 RPCS Functional Yes Yes Yes Yes No No l
BPWS Rod Pattern Yes Yes Yes No Yes or No Yes or No (Probability, %)
(%100)
(N100)
(N25)
(%75)
(100)
(100)
RWL Functional No No No No Yes Yes p
i Max. RWE Distance 1
(ft) 9 9
9 0
2 2
l RWL Design Basis (Distance, ft) 2 2
2 2
1 1
i k
Nominal Pre-PWE MCPR 2.83 2.31 1.69 1.69 1.48 1.26 Nominal Post-RWE MCPR 1.87 1.58 1.23 1.69 1.35 1.18 j
Probability Post-RWE MCPR 11.07 99.7 98.9 86.5 100 99.8 92.7 i
Probability Post-RWE i
F1CPR 11.00 99.9 99.5 94.4 100 99.9 99.2 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
ItC&ls NO POWEH
1 P 2
l1
?
A l
P A
I f itP HOD POSITIOrJ ep.HP+a O
l f HP I
NO itP ItP - IlP YES
'UP - BLOCK POSITION 1f f tOD B LOCK I
Figure 15BF-1.
RCIS System Function 15BF-26
O O
O d L a2 SF2 l
l t k a
/
S
S
V U
f, U U
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b CTP Pi P4 SRV MSIVt MS:V2 TSV TCV f
/
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U y
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1 r Eo 9
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om 1
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/
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FWP Nm
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O CONT RV RE ACTOR VESSEL CONT CONTAINMENT WHIMARY)
SF3 STE AMF LOW BYPASSED TO MAIN CONDENSER CTP CORE THERMAL TSV TURBINE STOP VALVES TUD TUR BIN E SF1 STEAMFLOW FROM VESSEL TCV TURBINE CONTROL VALVES GEN GENERATOR Pt VESSEL DOME PRESSURE BV BYP ASS V ALVES OOND CONDENSER SF2 STE AMF LOW THROUGH SRV ai SUM OF ALL EXTRACTION STEAM UPSTRE AM OF TSV FWP FEEDWATER PUMPS SRV SAFETY RELIEF VALVES MSIV l INBOARD / OUTBOARD MAIN SF4 STEAMFLOW TO TURSINE a2 EXTR ACTION STEAM DOWNSTREAM y
t OFP4 MEASUREMENT
%N MStV2 J STEAMLINE ISOLATION V ALVES P4 FIRST-STAGE TURBINE k>q J
PRESSURE
- O Figure 15BF-2.
First-Stage Turbine Pressure o
L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
/
/
HP Lp SF4 P4 TO COND 1I i
2 1
-v ;
- w, w,
y N
2 1
q r 0
R 3,
3 G
V P4 4 '
R2R3 C2 R (A2)
Hj
+
C 1+C2 + R (a2I R2+R3 2
2 8P4 C2 3R (a2) 3v R2
+
2
[C2 + R (ag)]2 ba2 1
R1
[3
=
R2+R SINCE.
<0 8 32 WHERE C, C2 ARE CONSTANTS 1
Figure 15BF-3.
Simplified Turbine Extraction Steam Design 15BF-29/15BF-30
1 I
j GESSAR II 22A7007 i
l 238 NUCLEAR ISLAND Rev. O I
l l
SUMMARY
TABLE OF CONTENTS 1
I Chapter /
j Section Title Volume L
1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT l.1 INTRODUCTION 1
}
1.1.1 Type of License Required l
1.1.2 Identification of Applicant i
1.1.3 Number of Plant Units f
1.1.4 Description of Location l
1.1.5 Type of Nuclear Steam Supply System l
i 1.1.6 Type of Containment l
i l
1.1.7 Core Thermal Power Levels
}
i 1.1.8 Scheduled Completion and Operation j
Dates l
1.2 GENERAL PLANT DESCRIPTION 1
l 1.2.1 Principal Design Criteria 1.2.2 Plant Description j
1.3 COMPARISON TABLES 1
1.3.1 Comparisons with Similar Facility i
Designs l.3.2 Comparisons of Final and Preliminary Information 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1
1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design 1.4.3 Nuclear Steam Supply Sy Supplier I
1.4.4 Turbine-Generator Vendor 1.4.5 Consultants y,
1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1
1.5.1 Current Development Programs 1.5.2 PSAR Commitment Items 1.6 MATERIAL INCORPORATED BY REFERENCE 1
r j
iii ww--
-__%,m..
_--,--.-,-.-e.,---------
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 1.7 DRAWINGS AND OTIIER DETAILED INFORMATION 1
1.7.1 Electrical, Instrumentation, and Control Drawings 1.7.2 Piping and Instrumentation Diagrams 1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1
1.8.1 Compliance Assessment Method 1.9 STANDARD DESIGNS 1
1.9.1 Interfaces 1.9.2 Exceptions O
O iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTENTS (Continued)
,.,(J Chapter /
Section Title Volume 2
SITE CHARACTERISTICS 2.0
SUMMARY
l 2.1 GEOGRAPHY AND DEMOGRAPHY l
2.1.1 Site Location and Description 2.1.2 Exclusion Area Authority and Control 2.1.3 Population Distribution 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 1
2.2.1 Location and Routes 2.2.2 Descriptions 2.2.3 Evaluation of Potential Accidents 2.3 METEOROLOGY l
2.3.1 Regional Climatology 2.3.2 Local Meteorology (m}
2.3.3 Onsite Meteorological Measurements Program 2.3.4 Short-Term Atmospheric Diffusion Estimates 2.3.5 Long-Term Atmospheric Diffusion Estimates 2.4 HYDROLOGIC ENGINEERING 1
2.4.1 Hydrologic Description 2.4.2 Floods 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers 2.4.4 Potential Dam Failures, Seismically Induced 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.6 Probable Maximum Tsunami Flooding 2.4.7 Ice Effects 2.4.8 Cooling Water Canals and Reservoirs 2.4.9 Channel Diversions
()
2.4.10 Flooding Protection Requirements v
v
GI:SSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTEf1TS (Continued)
Chapter /
Section Title Volume 2.4.11 Low Water Considerations i
l 1
2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4.13 Ground Water 2.4.14 Technical Specifications and Emergency Operation Requirements 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECl!NICAL ENGIf1EERING 1
2.5.1 Basic Geologic and Seismic Information 2.5.2 Vibratory Ground Motion 2.5.3 Surface Faulting 2.5.4 Stability of Subsurface Materials and Foundationn 2.5.5 Stability of Slopes 2.5.6 Embankments and Dams O
vi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 3
DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2
3.1.1 Summary Description 3.1.2 Criterion Conformance 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 2
3.2.1 Seismic Classification 3.2.2 System Quality Group Classifications 3.2.3 System Safety Classifications 3.2.4 Quality Assurance 3.2.5 Correlation of Safety Classes with Industry Codes 3.3 WIND AND TORNADO LOADINGS 2
3.3.1 Wind Loadings s
s.
3.3.2 Tornado Loadings 3.3.3 BOP Interface 3.3.4 References 3.4 WATER LEVEL (FLOOD) DESIGN 2
3.4.1 Flood Protection 3.4.2 Analytical and Test Procedures 3.4.3 BOP Interface 3.4.4 References 3.5 MISSILE PROTECTION 2
3.5.1 Missile Selection and Description 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles 3.5.3 Barrier Design Procedures 3.5.4 BOP Interface 3.5.5 References bG vii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 2
3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.3 References 3.7 SEISMIC DESIGN 3
3.7.1 Seismic Input 3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis 3.7.4 Seismic Instrumentation 3.7.5 BOP Interface 3.7.6 References 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 3
3.8.1 Concrete Containment 3.8.2 Steel Containment 3.8.3 Concrete and Steel Internal Structures of Steel Containment 3.8.4 Other Seismic Category I Structures 3.8.5 Foundations 3.8.6 BOP Interface 3.9 MECHANICAL SYSTEMS AND COMPONENTS 4
3.5.1 Special Topics for Mechanical Components 3.9.2 Dynamic Testing and Analfsis 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.4 Control Rod Drive System 3.9.5 Reactor Pressure Vessels Internals 3.9.6 Inservice Testing of Pumps and Valves 3.9.7 BOP Interface 3.9.8 References viii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
()
SUMMARY
TABLE OF CONTENTS (Continued)
V Chapter /
Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING llYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualification Criteria (Including flydrodynamic Loads) 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including flydrodynamic Loads) 3.10.4 Seismic Qualification Tests and Analyses (Including flydrodynamic Loads; 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED t
/\\
MECl!ANICAL AND ELECTRICAL EQUIPMENT 5
V 3.11.1 Equipment Identification and Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Loss of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF TIIE NUCLEAR ISLAND 5
APPENDIX 3B CONTAINMENT LOADS 6,7 APPENDIX 3C COMPUTER PROGRAMS USED IN T!!E DESIGN OF SEISMIC CATEGORY I STRUCTURES 8
APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR AND i
PUMP UNDER ACCIDENT CONDITIONS 8
l APPENDIX 3E DESCRIPTTON OF SAFETY RELATED MECilANICAL AND ELECTRICAL EQUIPMENT 8
l l
APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR l
CONTAINMENT VESSEL 8
t O
ix
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume APPENDIX 3G PIPE FAILURE ANALYSIS 8
APPENDIX 31I EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT.,
3 IN. ON SEISMIC DESIGN LOADS AND BUILDING RESPONSES 8
O
\\
l l
O I
l X
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 4
REACTOR 4.1
SUMMARY
DESCRIPTION 9
4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.3 Reactivity Control Systems 4.1.4 Analysis Techniques 4.1.5 References 4.2 FUEL SYSTEM DESIGN 9
4.2.1 General and Detailed Design Base 4.2.2 General Design Description 4.2.3 Design Evaluations 4.2.4 Testing and Inspection 4.2.5 Operating and Developmental Experience
-s
(_)
4.2.6 References 4.3 NUCLEAR DESIGN 9
4.3.1 Design Bases 4.3.2 Description 4.3.3 Analytical Methods 4.3.4 Changes 4.3.5 References 4.4 THERMAL - HYDRAULIC DESIGN 9
4.4.1 Design Basis 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System 4.4.4 Evaluation l
4.4.5 Testing and Verification l
4.4.6 Instrumentation Requirements 4.4.7 References
' (s) l xi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 4.5 REACTOR MATERIALS 9
4.5.1 Control Rod System Structural Materials 4.5.2 Reactor Internal Materials 4.5.3 Control Rod Drive Housing Supports 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 9
4.6.1 Information for Control Rod Drive System (CRDs) 4.6.2 Evaluations of the CRDs 4.6.3 Testing and Verification of the CRDs 4.6.4 Information for Combined Performance of Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6.6 References APPENDIX 4A CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR TYPICAL BWR 9
4A.1 Introduction 4A.2 Power Distribution Strategy 4A.3 Results of Core Simulation Studies 4A.4 Glossary of Terms 4A.5 References O
xii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
SUMMARY
TABLE OF CONTENTS (Continued) b Chapter /
Section Title Volume 5
REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1
SUMMARY
DESCRIPTION 10 5.1.1 Schematic Flow Diagram 5.1.2 Piping and Instrumentation Diagram 5.1.3 Elevation Drawing 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 10 5.2.1 Compliance with Codes and Code Cases 5.2.2 Overpressure Protection 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2.5 Reactor Coolant Pressure Boundary and ECCS System Leakage Detection s,
System s) 5.2.6 References 5.3 REACTOR VESSEL 10 5.3.1 Reactor Vessel Materials 5.3.2 Pressure / Temperature Limits 5.3.3 Reactor Vessel Integrity 5.3.4 References 5.4 COMPONENT AND SUBSYSTEM DESIGN 10 5.4.1 Reactor Recirculation Pumps 5.4.2 Steam Generators (PWR) 5.4.3 Reactor Coolant Piping 5.4.4 Main Steam Line Flow Restrictors 5.4.5 Main Steam Line Isolation System 5.4.6 Reactor Core Isolation Cooling System (RCIC) 5.4.7 Residual Heat Removal System 5.4.8 Reactor Water Cleanup System 5.4.9 Main Steam Lines and Feedwater Piping
()
5.4.10 Pressurizer xiii
l t
GESSAR II 22A7007 i
238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 5.4.11 Pressurizer Relief Valve Discharge System 5.4.12 Valves 5.4.13 Safety and Relief Valves 5.4.14 Component Supports l
5.4.15 References I
l l
9 xiv
\\
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTENTS (Continued)
('%g i
\\v Chapter /
Section Title Volume 6
ENGINEERED SAFETY FEATURES 6.0 GENERAL 11 6.1 ENGINEERED SAFETY FEATURE MATERIALS 11 6.1.1 Metallic Materials 6.1.2 Organic Materials 6.2 CONTAINMENT SYSTEMS 11 6.2.1 Containment Functional Design 6.2.2 Containment Ileat Removal System 6.2.3 Secondary Containment Functional Design 6.2.4 Containment Isolation System 6.2.5 Combustible Gas Control in Containment 6.2.6 Containment Leakage Testing (s) 6.2.7 Suppression Pool Makeup System 6.2.8 References 6.3 EMERGENCY CORE COOLING SYSTEMS 11 6.3.1 Design Bases and Summary Description 6.3.2
System Design
6.3.3 ECCS Performance Evaluation 6.3.4 Tests and Inspections 6.3.5 Instrumentation Requirements 6.3.6 References 6.4 IIABITABILITY SYSTEMS 11 6.4.1 Design Basis 6.4.2
System Design
6.4.3 Systems Operational Procedures 6.4.4 Design Evaluations 6.4.5 Testing and Inspection 6.4.6 Instrumentation Requirements 6.4.7 Nuclear Island / BOP Interface xv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 11 6.5.1 Standby Gas Treatment System (SGTS) 6.5.2 Containment Spray Systems 6.5.3 Fission Product Control Systems 6.5.4 Ice Condensers as a Fission Product Control System 6.5.5 References 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 11 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination Techniques and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) 11 6.7.1 Design Bases 6.7.2
System Description
6.7.3 System Evaluation 6.7.4 Inspection and Testing 6.7.5 Instrumentation Requirements 6.8 PNEUMATIC SUPPLY SYSTEM ll 6.8.1 Design Bases 6.8.2
System Description
6.8.3 System Evaluation 6.8.4 Inspection and Testing Requirements 6.8.5 Instrumentation Requirements APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS ll xvi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
O-Chapter /
Section Title Volume 7
INSTRUMENTATION AND CONTROL SYSTEMS
7.1 INTRODUCTION
(All Systems) 12 7.1.1 Identification of Safety-Related Systems 7.1.2 Identification of Safety and Power Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis 7.3 ENGINEERED SAFETY FEATUuES SYSTEM, INSTRUMENTATION AND CONTROL 13 7.3.1 Description 7.3.2 Analysis
-IIPCS
-Shield Building Annulus Mixing
-ADS (3) n ary gntain-
-LPCS ment Isolation
~ " !b
-Primary Containment
-CRVICS Isolation LCS
-MSPLCS
-Standby Power
-RilR/ Containment
-D-G Support Systems E##Y
-Essential Service
-RIIR/ Suppression Pool Water Cooling
-ESF Area Cooling
-Suppression Pool
-Pneumatic Supply Makeup
-CB Atmospheric
-Combustible Gas Control
-SGTS ed Water 7.4 SYSTEMS REQUIRED FOR SAFE SIlUTDOWN 14 7.4.1 Description 7.4.2 Analysis
-RCIC
-RilR/ Shutdown Cooling
-SLC
-Remote Shutdown xvii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5.1 Description 7.5.2 Analysis
-Nuclenet Control
-BOP Benchboard nsole
-Supervisory Moni-
-Standby Information toring Console
""UI
-Display Control
-Rx Core Cooling BB System 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 14 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References
-Neutron Monitoring
-FPCCS
-Process Radiation
-DW/ Containment Monitoring Vacuum Relief
-Refueling Interlocks
-Vent & Pressure Control
-Leak Detection
-Rod Pattern Control
~ ^
-Suppression Pool
-HP/LP System Interlock emperature Monitoring
-Recirculation Pump Trip 9
xviii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O gN
SUMMARY
TABLE OF CONTENTS (Continued)
U Chapter /
Section Title Volume 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 7.7.1 Description 7.7.2 Analysis 7.7.3 References
-RPV Instrumentation
-Leak Detection
-Rod Control &
-Rod Block Trip Information
-Fire Protection
-Recirculation Flow
-Drywell Chiller &
Control Cooling
-Feedwater Control
-Plant Instrument Air
-Performance Moni-
-Neutron Monitoring toring System
-Radwaste 7.8 NI/ BOP INTERFACES 14 7.8.1 Essential Service Water (Supply)
((j)
System Instrumentation and Controls 7.8.2 Diesel Generator Fuel Oil Transfer System APPENDIX 7A I&C ELEMENTARY DIAGRAMS 15 i
OT V
XiX
i GESSAR II 22A7007 238 NUCLEAR ISI AND Rev. O 5
SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 8
ELECTRIC POWER 8.1 INTRODUCTICN 16 8.1.1 Utility Grid Description 8.1.2 Onsite Electric Power System l
8.1.3 Design Bases 8.2 OFFSITE POWER SYSTEM 16 i
8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 Ot.'",ITE POWER SYSTEMS 16 8.3.1 AC Power Systen s i
8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems e
i 9
XX
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4
SUMMARY
TABLE OF CONTENTS (Continued)
-~
V Chapter /
Section Title Volume 9
AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 17 9.1.1 New Fuel Storage (High Density) 9.1.2 Spent Fuel Storage (High Density) 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.4 Fuel-Handling System 9.2 WATER SYSTEMS 17 9.2.1 Essential Service Water (ESW) System 9.2.2 Closed Cooling Water System 9.2.3 Demineralized Water Makeup System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Conocuoate Storage Facilities and Distribution System
)
9.2.7 Plant Chilled Water Systems 9.2.8 Heated Water Systems 9.2.9 Nuclear Island / BOP Interfaces 9.3 PROCESS AUXILIARIES 17 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Floor and Equipment Drainage Systems 9.3.4 Chemical and Volume Control System (PWR) 9.3.5 Standby Liquid Control System 9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 17 9.4.1 Control Room HVAC System 9.4.2 Fuel Building HVAC System 9.4.3 Auxiliary Building HVAC Systems 9.4.4 Turbine Building Area Ventilation System 9.4.5 Reactor Building HVAC System a
xxi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 9.4.6 Radwaste Building HVAC System 9.4.7 Diesel-Generator Buildings HVAC Systems 9.5 OTHER AUXILIARY SYSTEMS 17 9.5.1 Fire Protection System 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.5 Diesel-Generator Cooling Water System 9.5.6 Diesel-Generator Starting Air System 9.5.7 Diesel Engine Lubrication System 9.5.8 Diesel Generator Combustion Air Intake and Exhaust System 9.5.9 Suppression Pool Cleanup System 9.5.10 Nuclear Island - BOP Interface APPENDIX 9A FIRE HAZARD ANALYSIS 18 O
Xxii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 10 STEAM AND POWER CCNVERSION SYSTEM 10.1
SUMMARY
DESCRIPTION 19 10.2 TURBINE GENERATOR 19 10.2.1 Design Bases Functional Limitations by Design or Operational Charac-teristics of the Reactor Coolant System 10.2.2
System Description
10.2.3 Turbine Disk Integrity 10.2.4 Evaluation 10.3 MAIN STEAM SUPPLY 19 10.4 OTifER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 19 10.4.1 Main Condensers 10.4.2 Condenser Air Removal System t_
10.4.3 Main Condenser Evacuation System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR)
I 10.4.9 Auxiliary Feedwater System (PWR)
O xxiii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 11 RADIOACTIVE WASTE MANAGEMENT t
11.1 SOURCE TERMS 19 11.1.1 Fission Products 11.1.2 Activation Products 11.1.3 Tritium 11.1.4 Fuel Fission Production Inventory and Fuel Experience 11.1.5 Process Leakage Sources 11.1.6 Radwaste System 11.1.7 Radioactive Sources in the Gas
)
Treatment System 11.1.8 Source Terms for Component Failures 11.1.9 Other Releases i
11.1.10 References 11.2 LIQUID WASTE MANAGEMENT SYSTEM 19 11.2.1 Design Basis 11.2.2 System Descriptions 11.2.3 Estimated Releases 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 19 I
11.3.1 Design Bases 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System Description 11.3.3 RECHAR System Operating Procedure 11.3.4 Radioactive Releases 11.3.5 References 11.4 SOLID RADWASTE SYSTEM 19 11.4.1 Design Bases 11.4.2
System Description
O XXIV
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
O' Chapter /
Section Title Volume 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING MJD SAMPLING SYSTEMS 19 11.5.1 Design Bases 11.5.2
System Description
11.5.3 Effluent Monitoring and Sampling 11.5.4 Process Monitoring and Sampling 11.5.5 Calibration and Maintenance 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 19 k]
XXV
GESSAR II 22A7007 238 KUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 12 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 19 12.1.1 Policy Considerations 12.1.2 Design Considerations 12.1.3 Operational Considerations 12.2 RADIATION SOURCES 19 12.2.1 Contained Sources 12.2.2 Airborne Radioactive Material Sources 12.2.3 References 12.3 RADIATION PROTECTION DESIGN FEATURES 19 12.3.1 Facility Design Features 12.3.2 Shielding 12.3.3 Ventilation 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.5 References 12.4 DOSE ASSESSMENT 19 12.5 HEALTH PHYSICS PROGRAM 19 O
XXVi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
i
SUMMARY
TABLE OF CONTENTS (Continued) i Chapter /
f Section Title Volume i
]
13 CONDUCT OF OPERATIONS 19 l
i I
i
{
XXVii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 14 INITIAL TEST PROGRAM 14.1 TEST PROGRAM 20 14.1.1 Administrative Procedures (Testing) 14.1.2 Administrative Procedures (Modifications) 14.1.3 Test Objectives and Procedures 14.1.4 Fuel Loading and Initial Operation 14.1.5 Administrative Procedures (System Operation) 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN SAFETY ANALYSIS REPORTS 20 14.2.1 Summary of Test Program and Objectives 14.2.2 Organization and Staffing 14.2.3 Test Procedures 14.2.4 conduct of Test Program 14.2.5 Review, Evaluation and Approval of Test Results 14.2.6 Test Records 14.2.7 Conformance of Test Programs with Regulatory Guides 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Develop-ment of Test Program 14.2.9 Trial Use of Plant Operating and Emergency Procedures 14.2.10 Initial Fuel Loading and Initial Criticality 14.2.11 Test Program Schedule 14.2.12 Individual Test Descriptions O
xxviii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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TABLE OF CONTENTS (Continued) v Chapter /
Section Title Volume 15 ACCIDENT ANALYSES 15.0 GENERAL 21 15.0.1 Analytical Objective 15.0.2 Analytical Categories 15.0.3 Event Evaluation 15.0.4 Nuclear Safety Operational Analysis (NSOA) Relationship 15.0.5 References 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 21 15.1.1 Loss of Feedwater Heating 15.1.2 Feedwater Controller Failure -
Maximum Demand 15.1.3 Pressure Regulator Failure - Open 15.1.4 Inadvertent Safety / Relief Valve
}
Opening
~/
15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.7 References j
15.2 INCREASE IN REACTOR PRESSURE 21 l
15.2.1 Pressure Regulator Failure - Closed i
15.2.2 Generator Load Rejection 15.2.3 Turbine Trip 15.2.4 MSLIV Closures 15.2.5 Loss of Condenser Vacuum l
15.2.6 Loss of Offsite AC Power 15.2.7 Loss of Feedwater Flow l
15.2.8 Feedwater Line Break 15.2.9 Failure of RHR Shutdown Cooling o
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TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 21 15.3.1 Recirc21ation Pump Trip 15.3.2 Recirculation Flow Control Failure -
Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.3.4 Recirculation Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 21 15.4.1 Rod Withdrawal Error - Low Power 15.4.2 Rod Withdrawal Error at Power 15.4.3 Control Rod Maloperation (System Malfunction or Operator Error) 15.4.4 Abnormal Startup of Idle Recirculation Pump 15.4.5 Recirculation Flow Control with Increasing Flow 15.4.6 Chemical and Volume Control System Malfunctions 15.4.7 Misplaced Bundles Accident 15.4.8 Spectrum of Rod Ejection Assemblics 15.4.9 Control Rod Drop Accident (CRDA) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 21 15.5.1 Inadvertent HPCS Startup 15.5.2 Chemical Volume Control System Malfunction (or Operator Error) 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 15.6 DECREASE IN REACTOR COOLANT INVENTORY 21 15.6.1 Inadvertent Safety / Relief Valve Opening 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Steam Generator Tube Failure 15.6.4 Steam System Piping Break Outside Containment 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Inside Pressure Boundary)
Containment 15.6.6 Feedwater Line Break - Outside Containment 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 21 O)
(_
15.7.1 Radioactive Waste System Leak or Failure 15.7.2 Liquid Radioactive System Failure 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure 15.7.4 Fuel-Handling Accident 15.7.5 Spent Fuel Cask Drop Accidents APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS 21 APPENDIX 15B BWR/6 GENERIC ROD WITHDRAWAL ERROR ANALYSIS 21 OV XXXi
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Section Title Volume 16 STANDARD TECHNICAL SPECIFICATIONS FOR GENERAL ELECTRIC BOILING WATER REACTORS D_FINITIONS 22 16.1 E
16.1.1 Action 16.1.2 Average Planar Exposure 16.1.3 Average Planar Linear Heat Generation Rate 16.1.4 Channel Calibration 16.1.5 Channel Check 16.1.6 Channel Functional Test 16.1.7 Core Alteration 16.1.8 Critica) Power Ratio 16.1.9 Dose Equivalent I-131 16.1.10 E-Average Disintegration Energy 16.1.11 Emergency Core Cooling System (ECCS)
Response Time 16.1.12 Frequency Notation 16.1.13 Identified Leakage j
16.1.14 Isolatior. System Response Time 16.1.15 Limiting Control Rod Pattern 16.1.16 Linear Heat Generation Rate 16.1.17 Logic System Functional Test 16.1.18 Maximum Total Peaking Factor 16.1.19 Minimum Critical Power Ratio 16.1.20 Operable - Operability l
16.1.21 Operational Condition (Condition) 16.1.22 Physics Test 16.1.23 Pressure Boundary Leakage 16.1.24 Primary Containment Integrity 16.1.25 Rated Thermal Power 16.1.26 Reactor Protection System Response Time l
16.1.27 Recirculation Pump Trip System Response Time xxxii
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TABLE OF CONTENTS (Continued)
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V Chapter /
Section Title Volume 16.1.28 Reportable Occurrence 16.1.29 Rod Density 16.1.30 Secondary Containment Integrity 16.1.31 Shutdown Margin 16.1.32 Staggered Test Basis 16.1.33 Thermal Power 16.1.34 Total Peaking Factor 16.1.35 Unidentified Leakage 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 22 16.2.1 Safety Limits 16.2.2 Limiting Safety System Settings 16.B2 SAFETY LIMITS 22 16.B2.1 Bases
()
16.B2.2 Limiting Safety System Settings 16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 22 16.3/4.0 Applicability 16.3/4.1 Surveillance Requirements 16.3/4.2 Power Distribution Limits 16.3/4.3 Instrumentation I
16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systems l
16.3/4.6 Containment Systems 16.3/4.7 Plant Systems 16.3/4.8 Electrical Power Systems i
16.3/4.9 Refueling Operations l
l 16.3/4.10 Special Test Exceptions i
q
.)
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TABLE OF CONTENTS (Continued)
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Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.B3/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.B3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.B3/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 22 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Core 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 Fuel Ste" age 16.5.7 Component Cyclic or Tra1sient Limit l
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l XXXIV k
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TABLE OF CONTENTS (Continued)
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I Chapter /
Section Title Volume 17 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 22 j
j 17.2 QUALITY ASSURANCE DURING THE OPERATING PHASE 22 1
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