ML20210R312
| ML20210R312 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/01/1986 |
| From: | Cook K LOUISIANA POWER & LIGHT CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20210R318 | List: |
| References | |
| W3P86-3328, NUDOCS 8610070234 | |
| Download: ML20210R312 (301) | |
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LO UISI AN A / 317 BARONNE STREET
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P. O box 60340 POWER & L1GHT NEW ORLEANS, LOUISIANA 70160 + (504) 595-3100 UiluTIES SYSTEM October 1, 1986 W3P86-3328 3-A1.01.04 A4.05 QA Mr. George W. Knighton, Director PWR Project Directorate No. 7 Division of PWR Licensing-B Office of Nuclear Reactor Regulation Washington, D.C. 20555
SUBJECT:
Waterford 3 SES Docket No. 50-382 Reload Cycle 2 Reports
REFERENCES:
1.
LP&L Letter W3P86, August 29, 1986 2.
USNRC Letter, From Robert A. Clark et. al.
to William Cavanaugh, dated August 11, 1982
Dear Mr. Knighton:
Provided herewith are four items to support the Cycle 2 reload: Technical Specification Matrix, Reload Analysis Report (RAR), Middle South Utilities System Reactor Physics Methodology Report, and the Boric Acid Concentration Reduction Effort Report, CEN-341(C).
Attachment One (three copies) provides a Technical Specification Matrix. The Technical Specification Matrix lists the basis documents which are applicable to all of the reload cycle 2 safety analysis based technical specifications submitted by LP&L.
Attachment Two (three copies) provides the second part (Part B) of the Waterford Reload Analysis Report (RAR). As previously discussed with you, the RAR has been submitted to the NRC in two parts (Part A and Part B), Part A submitted via the reference letter one. This Part B of the RAR is the final part of and completes the RAR.
Part B of the RAR contains new sections 7.0, 8.0, 10.0 and supplements to sections 4.0 and 12.0.
Section 4.0 of the RAR Part B incorporates section 4.5, shoulder gap adequacy. Sections 7.0 and 8.0 contain the Non-LOCA Safety Analysis and ECCS Analysis, respectively. Section 10.0 provides a sumary of the proposed cycle 2 Reload Technical Specifications. Section 12.0 is the references section.
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8610070234 861001 NS20579 PDR ADOCK 05000382
\\
P PDR "AN EQUAL OPPORTUNITY EMPLOYER" I
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Page 2 W3P86-3328 October 1, 1986 Attachment Three (three copies) provides the CEN-341(C) Report," Boric Acid Concentration Reduction Effort, Technical Bases and Operational Analysis, Waterford Nuclear Power Plant Unit 3".
This report defines the methodology, and outlines the technical bases, which allows a reduction in the boric acid makeup tank concentration to the point where heat tracing of the boric acid makeup system is no longer required in order to prevent boric acid precipitation.
Attachment Four (three copies) provides the Supplement #1 to MSS-NAl-P. The purpose of this supplement is to demonstrate that the reliability factors presented in MSS-NAl-P are appropriate for application to physics parameters determined by Middle South Services, Inc. (MSS) in support of the Waterford Steam Electric Station Unit No. 3 (WSES-3). This supplement extends the model validation presented in MSS-NAl-P for Arkansas Nuclear One Units 1 and 2 to WSES-3.
In the 1982 safety evaluation report of MSS-NAl-P (Reference 2), the physics methods and reliability factors were not approved for WSES-3 use because comparisons of predictions to WSES-3 measurements were not available when the original report was submitted by LP&L. This supplement provides comparisons between measurements and predictions for WSES-3 and establishes the reliability factors appropriate for WSES-3 applications.
The safety analyses which support the foregoing reports are believed to be technically firm and accurate.
However, a formal QA verification of the analyses has not been completed by CE. The NRC would be expediently notified in the event that the formal QA verification reveals any findings contrary to the conclusions stated in the subject reports.
Please contact me or Robert J. Murillo shoul u have any questions.
/
.W.
Cook Nuclear Support & Licensing Manager KWC/RJM/pim Attachment cc:
B.W. Churchill, W.M. Stevenson, R.D. Martin, J.H. Wilson, NRC Resident Inspector's Office (W3) i
o (ATTACHMENT ONE)
TECHNICAL SPECIFICATION MATRIX T.S.
Section Title Content Description Applicable Basis Document 3.1.1.1 Shutdown Margin-T Change title and content RAR (avg) Greater to provide shutdown 7.1.5 than 200 F margin both above and 7.4.4 below 200'F when any CEA FSAR is at lease partially 15.1.3.1 withdrawn, consistent 15.4.1.4 with program for tempera-ture dependent shutdown margin.
3.1.1.2 Shutdown Margin-T Change title and content Same as T.S. 3.1.1.1 (avg) less than to provide shutdown or equal to margin requirements when 200*F all CEA's are fully inserted. New figure gives temperature dependent shutdown margin in Modes 2 through 5, consistent with 3.1.1.1.
3.1.2.1 Boration Systems Change in BAM Tank CEN-341(C)
Flow Paths -
Concentration Shutdown 3.1.2.2 Boration Systems Change in BAM Tank CEN-341(C)
Flow Paths -
Concentration Shutdown 3.1.2.4 Charging Remove value for shutdown Same as T.S. 3.1.1.1 Pumps -
margin in action statement Operation consistent with 3.1.1.1 and 3.1.1.2 changes for tempera-l ture dependent shutdown margin.
l 3.1.2.6 Boric Acid Same as for 3.1.2.4.
Same as T.S. 3.1.1.1 Makeup Pumps -
Operating 3.1.2.7 Borated Water Change in BAM Tank CEN-341(C) i Sources -
Concentration Shutdown NS30623
- (ATTACHMENT ONE)
TECHNICAL SPECIFICATION MATRIX Section Title Content Description Applicable Basis Document 3.1.2.8 Borated Water Change in BAM Tank CEN-341(C)
Sources -
Concentration Operation 3.1.2.9 Boron Dilution Revise Table 3.1-1 on RAR monitoring frequency with 7.4.4 1 or 2 boron dilution FSAR alarms inoperable, 15.4.1.4 consistent with 3.1.1.1 and 3.1.1.2 changes for temperature dependent shutdown margin.
Also revised charging pump operability require-ments.
2.1.1.1 DNBR Increase DNBR Safety Limit RAR to include allowances for 7.1.5 hot channel factors and 7.3.2 rod bow penalty per method 7.3.3 of Statistical Combination 7.4.1 of Uncertainties.
7.4.3 7.7.1 FSAR 15.1.2.3 15.3.2.1 15.3.3.1 15.4.1.3 15.1.3.1 2.2.1 Reactor Trip Increase DNBR trip Same as T.S. 2.1.1.1 Setpoints setpoint consistent with 2.1.1.1 change. Delete BERRI value from Tech Spec consistent with previous deletion of addressable constants.
3.1.3.6 Regulating Revise PDIL in Figure RAR CEA Insertion 3.1-2 to account for 7.1.5 Limits Cycle 2 core characteris-7.4.1 tics.
7.4.2 7.4.6 FSAR 15.4.1.1 15.4.1.2 15.4.1.3 15.1.3.1
. (ATTACHMENT ONE)
TECHNICAL SPECIFICATION MATRIX Section Title Content Description Applicable Basis Document j
3.2.1 Linear Revise linear heat rate RAR Heat Rate limit Figure 3.2-1 when supports conclu-COLSS is out-of-service sions of 8.0 (limiting event) 3.2.4 DNBR Margin Revise COLSS penalty RAR when neither CEAC is supports conclu-operable. Revise CPC sions of 7.4.3 monitoring limit on DNBR (limitingevent) in Figures 3.2-2 and 3.2-3 when COLSS is out-of-service.
Delete surveillance requirement on COLSS & CPC for rod bow penalty consistent with changes to 2.1.1.1. and 2.2.1 3.1.1.3 Moderator Increase allowable range RAR Temperature for MTC to accommodate 7.1.5 Coefficient higher BOC boron concen-FSAR tration and E0C burnup in 15.1.3.1 in Cycle 2.
3.1.3.1 CEA Position Add new figure to impose RAR power reduction between 7.4.3 15 minutes and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> FSAR after CEA misalignment 15.4.1.3 consistent with the removal of the CPC penalty factor for inward CEA deviation, per CPC Improvement Program.
3.1.3.7 Part Length Add new PDIL with more RAR CEA Insertion restrictive insertion 7.4.3.1 Limits for part length CEA's to FSAR improve rod drop 15.4.1.3 consequences and to clarify rod position vs.
allowable duration.
i
. (ATTACHMENT ONE)
TECHNICAL SPECIFICATION MATRIX Section Title Content Description Applicable Basis Document 3.3.1 Radiation Revise Table 3.3-2 RAR Monitoring Response Times 7.1.5 Instrumentation 7.3.2 7.4.1 7.4.2 7.7.1 FSAR 15.1.3.1 15.3.2.1 15.4.1.1 15.4.1.3 5.3.1 Fuel Assemblies Fuel Enrichment limit MSS Fuel increases from 3.7 w/o Storage Rack to 4.0 w/o Report, 304-37, 5.6.1 Fuel Storage Update to reflect new MSS Fuel Criticality fuel storage criticality Storage Rack analyses Report, 304-37
y
. ATTACHMENT 2 RELOAD ANALYSIS REPORT FOR WSES-3 CYCLE 2 (PART B)
r 4.0 FUEL SYSTEM DESIGN 4.1 MECHANICAL DESIGN 4.1.1 Fuel Design The mechanical design of the Batch D rel oad fuel assemblies is identical to that of the Core 1 fuel assemblies (Reference 4-1) with the exception of the design features listed bel ow.
No changes in mechanical. design bases have occurred since the original fuel design.
However, the ' Batch D design incorporates a number of refinements for the purpose of improving fuel handling, improving the burnup capability of the fuel, and improving the reconstitution feature of the fuel.
The specific changes are discussed in the following paragraphs:
1.
The perimeter strip lead-in tabs of the lowest spacer grid (made of Inconel) have been changed f rom trapezoidal-shaped to curve-shaped to improve fuel handling by reducing the chance of grid hangup. "The remaining spacer grids (nade of Zircaloy) al ready have curve-shaped perimeter strip lead-in tabs.
2.
The shoulder gap in the Batch 0 fuel has been increased to improve the burnup capability of the fuel.
The Batch 0 shoulder gap is 2.382 inches, compared to 1.332 inches for part of Batch B and 2.032 inches (after shimming) for the remainder of Batch B and Batch C.
The Batch 0 shoulder gap was achieved by l
lengthening the guide tubes by 0.9 inches and shortening the fuel rods by 0.15 inches.
The changes do not result in the violation of any design criteria.
l 3.
The designs of the CEA guide tubes and wear sleeves have been modified to facilitate fuel bundle reconstitution.
Reference 4-3 l
l is the NRC's acceptance of the design change.
4-1
e I
4.1.2 Clad Collapse The fuel to be included in the Cycle 2 core satisfies the bases for not requiring cald collapse analyses, as documented in Reference 4-5. '
Therefore, no cycle specific clad collapse analysis was performed for Cycle 2, 4.2 MITIGATION OF GUIDE TUBE WEAR All fuel assemblies which will be placed in CEA locations in Cycle 2 will have stainless steel sleeves installed in the guide tubes to prevent guide tube wear.
4.3 THERMAL DESIGN The thermal performance of composite fuel rods that envelope the rods of fuel batches B, C and D present in Cycle 2 have been evaluated using the FATES 3A version of the C-E fuel evaluation model (References 4-7 and 4-8) as approved by the NRC (Reference 4-9).
The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, f rom beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of Cycle 2.
Results of these burnup dependent fuel performance calculations were used in the Transient Analysis presented in Section 7 and in the ECCS l
Analysis presented in Section 8.
l 4.4 CHEMICAL DESIGN The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch D fuel are identical to those of the fuel batches included in Cycle 1.
Thus, the chemical or metallurgical performance of the Batch D fuel is expected to remain unchanged from the performance of the Cycle 1 fuel (Reference 4-1).
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l 4-2
e 4.5 SHOULDER GAP ADEOUACY All fuel assemblies in Cycle 2 were reviewed for shoulder gap clearance using the technique described in Reference 4-10 (ifmiting fuel rod growth rate from ANO-2 Batch C fuel and a conservative lower estimate of guide tube growth)and were found to be acceptable for Cycle 2.
In addition, all fuel assemblies were evaluated and found to be acceptable using the technique described in Reference 4-11 (SIGREEP Model). Therefore, it is concluded that no inspections are necessary and the fuel is acceptable for Cycle 2 operation with re-gard to shoulder gap adequacy. A summary of the supporting data. the analytical techniques and their application to WSES-3 Cycle 2 is given in Reference 4-12.
4-3
r 7.0 Non-LOCA Safety Analysis 7.0.1 Introduction This section presents the results of the Waterford Steam Electric Station Unit 3 (WSES-3), Cycle 2 Non-LOCA safety analyses at 3410 MWt.
The Design Basis Events (DBEs) considered in the safety analyses are listed in Table 7.0-1.
These events are categorized into three groups:
Moderate Frequency, Infrequent and Limiting Fault events.
For the purpose of this report, the Moderate Frequency and Infrequent Events will be termed Antici-pated Operational Occurrences.
The DBEs were evaluated with respect to four criteria:
Offsite Dose, Reactor Coolant System Pressure, Fuel Performance (DNBR and Centerline Mitit SAFDLs) and Loss of Shutdown Margin.
Tables 7.0-2 through 7.0-5 present the list of events analyzed for each criterion.
All events were re-evaluated to assure that they meet their respective criteria for Cycle 2.
The DBEs chosen for analysis for each criterion are the limiting events with respect to that criterion.
The write-ups for those events presented consist of a discussion of the reason (s) for the reanalysh a -discussion of the cause(s) of the event, a description of the analyses performed, results and conclusions.
Some events previously analyzed with and without a single failure in the Cycle 1 FSAR (Reference 7-1) have been combined into the same section for presentation.
7.0.2 Methods of Analysis The analytical methodology used for WSES-3 Cycle 2 is the same as Cycle 1 (Reference Cycle) methodology (Reference 7-1) unless otherwise stated in the event presentation.
Only methodology that has previously been reviewed and approved on the WSES-3 docket or on other dockets is used.
i 7-1
7.0.3 Mathematical Models For most events presented herein, the mathematical models and computer codes used in the Cycle 2 Non-LOCA safety analysis are the same as those used in the Reference Cycle analysis (Reference 7-1).
The exceptions to this are the application of the TORC code to the Sheared Shaft / Seized Rotor event and the HERMITE code to the Total Loss of Forced Reactor Coolant Flow, the Asymmetric Steam Generator Transient, the Return-to-Power Steam Line Break and the Sheared Shaft / Seized Rotor events, all of which have been reviewed and approved on other dockets.
Plant response for Non-LOCA Events was simulated using the CESEC III computer code (Reference 7-3).
Simulation of the fluid conditions within the hot channel of the reactor core and calculation of DN8R was performed using the CETOP-D computer code described in Reference 7-7.
The. TORC computer code is used to simulate the fluid conditions within the reactor core and to calculate fuel pin DNBR for the sheared shaft event.
The TORC code is described in Referances 7-10 and 7-11.
Detennination of DNBR for the post trip return to power portion of the steam piping failure events is based on the correlation developed by R. V. Macbeth (Reference 7-5) with corrections to account for non-uniform axial heat flux developed by Lee (Reference 7-6).
This methodology is the same as that employed in the Reference Cycle analysis.
The number of fuel pins predicted to experience clad failure is taken as the number of pins which have a CE-1 DNBR value below 1.26.
The sole exception is the Sheared Shaft event for which the statistical convolution method, described in Reference 7-12, was used.
The HERMITE computer code (Reference 7-9) was used to simulate the reactor core for analyses which required more spatial detail than is provided by a i
point kinetics model.
7-2 l
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7.0.4 Input Parameters and Analysis Assumotions Table 7.0-6 sumarizes the core parameters assumed in the Cycle 2 transient analysis and compares them to the values used in the Reference Cycle.
Specific initial conditions for each event are tabulated in the section of the report sumarizing that event.
For some of the DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculated Cycle 2 values (i.e., CEA worth at trip, moderator temperature coefficient).
7.0.5 Conclusion All DBEs are evaluated for WSES-3 Cycle 2 to determine whether their results are bounded by the Reference Cycle.
Those events whose results were not bounded by the Reference Cycle and those events for which analysis methodology differs from the Reference Cycle methodology are presented herein.
All DBEs have results within NRC acceptance criteria.
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7-3
Table 7.0-1 WSES Unit 3, Design Basis Events Considered in the Cycle 2 Safety Analysis 7.1 Increase in Heat Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature 7.1.2 Increase in Feedwater Flow 7.1.3 Increased Main Steam Flow 7.1.4 Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve 7.1.5*
Steam System Piping Failures 7.2 Decrease in Heat Removal by the Secondary System 7.2.1 Loss of External Load 7.2.2 Turbine Trip 7.2.3 Loss of Condenser Vacuum 7.2.4 Loss of Normal AC Power 7.2.5 Loss of Normal Feedwater 7.2.6*
Feedwater System Pipe Breaks 7.3 Decrease in Reactor Coolant Flowrate 7.3.1 Partial Loss of Forced Reactor Coolant Flow 7.3.2 Total Loss of Forced Reactor Coolant Flow 7.3.3*
Single Reactor Coolant Pump Shaft Seizure / Sheared Shaft 7.4 Reactivity and Power Distribution Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Conditjon 7.4.2 Uncontrolled CEA Withdrawal at Power 7.4.3 CEA Misoperation Events 7.4.4 CVCS Malfunction (Inadvertent Boron Dilution) 7.4.5 Startup of an Inactive Reactor Coolant System Pump 7.4.6*
Control Element Assembly Ejection 7.5 Increase in Reactor Coolant System Inventory 7.5.1 CVCS Malfunction l
7.5.2 Inadvertent Operation of the ECCS During Power Operation l
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l 7-4
r
~ Table 7.0-1 (continued) 7.6 Decrease in Reactor Coolant System Inventory 7.6.1 Pressurizer Pressure Decrease Events 7.6.2*
Small Primary Line Break Outside Containment 7.6.3*
Steam Generator Tube Rupture 7.7 Miscellaneous 7.7.1 Asymmetric Steam Generator Events Categorized as Limiting Fault Events l
l 7-5
c Table 7.0-2 DBEs Evaluated with Respect to Offsite Dose Criterion Section Event Results A) Anticipated Operational Occurrences 7.1.4
- 1) Inadvertent Opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmospheric Oump Valve 7.2.4
- 2) Loss of Normal AC Power Bounded by Reference Cycle B) Limiting Fault Events
- 1). Steam System Piping Failures:
Presented 7.1.5a a) Pre-Trip Power Excursions 7.1.5b b) Post Trip Analysis Bounded by 7.2.6
- 2) Feedwater System Pipe Breaks Reference Cycle 7.3.3
- 3) Single Reactor Coolant Pump Presented Shaft Seizure / Sheared Shaft 7.6.2
- 4) Small Primary Line Break Outside Bounded by Containment Reference Cycle 7.6.3
- 5) Steam Generator Tube Rupture Bounded by Reference Cycle i
7-6
r Table 7.0-3 CBEs Evaluated with Rescect tc RCS Pressure Criterion Section Event Results A) Anticipated Operational Occurrences 7.2.1
- 1) Loss of External Load Bounded by Reference Cycle 7.2.2
- 2) Turbine Trip Bounded by Reference Cycle 7.2.3
- 3) Loss of Condenser Vacuum Bounded by Reference Cycle 7.2.4
- 4) Loss of Nomal AC Power Bou'nded by Reference Cycle 7.2.5
- 5) Loss of Nomal Feedwater Bounded by Reference Cycle 7.4.1
- 6) Uncontrolled CEA Withdrawal from Bounded by Subcritical or Low Power Condition Reference Cycle 7.4.2
- 7) Uncontrolled CEA Withdrawal at Power Bounded by Reference Cycle 7.5.1
- 8) CVCS Malfunction Bounded by Reference Cycle 7.5.2 9)
Inadvertent Operation of the Bounded by ECCS During Power Operation Reference Cycle B) Limiting Fault Events 7.2.6
- 1) Feedwater System Pipe Breaks Bounded by Reference Cycle 7-7
r-
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Table 7.0-4 DBEs Evaluated with Respect to Fuel Performance Section Event Results A) Anticipated Operational Occurrences 7.1.1
- 1) Decrease in Feedwater Temperature Bounded by Reference Cycle 7.1.2 2)
Increase in Feedwater Flow Bounded by Reference Cycle 7.1.3 3)
Increased Main Steam Flow Bounded by Reference CycTe 7.3.1
- 4) Partial Loss of Forced Reactor Bounded by Coolant Flow Reference Cycle 7.3.2
- 5) Total Loss of Forced Reactor Presented
- Coolant Flow 7.4.1
- 6) Uncontrolled CEA Withdrawal from a Presented
- Subcritical or Low Power Condition 7.4.2
- 7) Uncontrolled CEA Withdrawal Bounded by at Power Reference Cycle 7.4.3
- 8) CEA Misoperation Events Presented
- 7.6.1
- 9) Pressurizer Pressure Decrease Bounded by Events Reference Cycle 7.7.1
- 10) Asymmetric Steam Generator Events Presented
- B) Limiting Fault Events
- 1) Steam System Piping Failures:
7.1.5a a) Pre-Trip Power Excursions Presented 7.1.5b b) Post Trip Analysis Presented 7.3.3
- 2) Single Reactor Coolant Pump Presented Shaft Seizure / Sheared Shaft 7.4.6
- 3) Control Element Assembly Ejection Bounded by Reference Cycle
- The results of this event remain bounded by the Reference Cycle.
The event is presented due to a change in analytical methodology.
7-8
Table 7.0-5 DBEs Evaluated with Resoect to Shutdown Margin Criterion Section Event Results A) Anticipated Operational Occurrences 7.1.4 1)
Inadvertent Opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmospheric Oump Valve 7.4.4
B) Limiting Fault Events 7.1.5b
- 1) Steam System Piping Failures, Presented Post Trip Analysis 7-9 1
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Table 7.0-6 WSES Unit 3, Cycle 2 Core Parameters Input to Safety Analyses Reference Cycle 1 Safety Parameters Units Values Cycle 2 Values Total RCS Power MWt 3478 3478 (Core Thermal Power
+ Pump Heat)
Core Inlet Steady State
- F 542 to 560 542 to 560 Temperature (70% power and (70% power and above) above) 520 to 560 520 to 560 (below 70% power)
(below 70% power)
Steady State psia 2000 - 2300 2000 - 2300 RCS Pressure Rated Reactor gpm 396,000 to 396,000 to Coolant Flow 410,000 410,000 Axial Shape Index LCO ASI
.3 to +.5
.3 to +.3 Band Assumed for Units All Powers Maximum CEA Insertion
% Insertion 28 28 at Full Power of Lead Bank
% Insertion 90 25 of Part-Length Maximum Initial Linear KW/ft 13.4 13.4 Heat Rate Steady State Linear KW/ft 21.0 21.0 Heat Rate for Fuel Center Line Melt CEA Drop Time from sec 3.0 3.0 Removal of Power to Holding Coils to 90%
Insertion i
Minimum DNBR CE-1 (SAFDL) 1.19 1.26 Macbeth (Fuel failure 1.30 1.30 limit for post-trip SLB with LOAC) 7-10
)
- Table 7.0-6 (continued)
Reference Cycle Values Safety Parameters Units (Cycle 1)
Cycle 2 Values Moderator Temperature 10-#ao/*F
-2.5 to +0.5
-3.3 to +0.5 Coefficient (below 70%
(below 70% power) power)
-3.3 to 0.0
-2.5 to 0.0 (70% power and (70% power and above) above)
Shutdown Margin (Value
%ao
-5.15
-5.15 Assumed in Limitin EOC Zero Power SLB 7-11
7.1 Increase in Heat-Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature The results are bounded by the Reference Cycle.
7.1.2 Increase in Feedwater Flow The results are bounded by the Reference Cycle.
7.1.3 Increased Main Steam Flow The results are bounded by the Reference Cycle.
7.1.4 Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Oump Valve The results are bounded by the Reference Cycle.
7.1.5 Steam System Piping Failures Failures in the main steam system piping were analyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.
7.1.5a Steam System Piping Failures:
Inside and Outside Containment Pre-Trip Power Excursions 4
The Steam Line Break with Loss of Offsite AC Power (LOAC) was analyzed to evaluate the maximum number of calculated fuel pin failures for the site boundary dose calculation.
Changes in the pin census for Cycle 2 and the availability of the CPC Variable Overpower Trip (V0PT) make this analysis necessary.
7-12
7.1.5a.1 Identification.of Causes A rupture in the main steam system piping increases steam flow from the steam generators.
This increase in steam flow increases the rate of RCS heat removal by the steam generators and causes a decrease in core coolant inlet temperature.
In the presence of a negative moderator temperature coefficient of reactivity (MTC), this decrease in temperature causes core power to
~
increase.
A Loss of Offsite AC Power (LOAC) during the transient can contribute to an additional reduction of themal margin due to the associated loss of power to the reactor coolant pumps.
The excursion in core power is terminated by the action of one of the following Reactor Protection System (RPS) trips:
Core Protection Calculators (CPCs). Low Steam Generator Pressure (LSGP), High Linear Power Level, or High Containment Pressure.
Further reduction of thermal margin due to LOAC is terminated by the Delta Pressure Low Flow trip.
7.1.5a.2 Analysis of Effects and Consequences Steam Line Breaks (SLBs) inside containment may be postulated to have break 2
areas up to the cross section of the largest main steam pipe (7.88 ft ),
Those SLBs occurring outside the containment building have break areas limited 2
by the areas of the main steam flow venturies (3.14 ft ) which are located upstream of the containment penetrations.
Only those trips which are not subject to environmental degradation are credited for the inside containment SLB cases.
The required input into the V0PT includes output from the Resistance Temperature Detectors (RTDs) and ex-core power detectors.
In addition to the V0PT, the LSGP trip is credited and the environmentally degraded value of the Delta Pressure Low Flow trip is used to determine the most adverse timing of a Loss of Offsite AC Power.
The outside containment SLB is not subject to the same adverse environmental effects as the inside containment SLB, Therefore, the full array of RPS trips are credited.
)
7-13
A parametric analysis in both MTC and break area was performed on the SLB events.
This parametric analysis identified the limiting inside and outside containment SLB events in terms of fuel pin failure caused by the pre-trip power excursion.
Table 7.1.5a-1 lists the most limiting values of key parameters for the inside containment steam line breaks.
The limiting values of key parameters, Table 7.1.Sa-2, for the outside containment SLB are the same as those for the inside containment SLB except the break area.
Using the criteria that all fuel rods for which CE-1 DNBR fall-s below the 1.26 SAFDL are assumed to experience cladding failure, the results of the inside and outside containment cases indicate that fewer than 4 percent of the fuel pins fail for the inside containment breaks and that fewer than 1.5 percent of the fuel cins fail for the outside containment steam line breaks.
Although the fuel failure is less, the outside containment SLB is more Ifmiting in terns of radiological consequences.
This is because a steam generator iodine decontamination factor (DF) of 100, as specified in Reference 7-2, was used in the calculation of radiological consequences for the inside containment SLB.
The calculation for the outside containment SLB used a DF of 1.
7.1.5a.3 Results Radiological consequences were considered for three cases.
Each case assumed that all iodine transported to the secondary side was released to atmosphere during periods of steam generator dryout.
Each case used an assumed maximum steam generator leakage of 1 gpm to the affected steam generator, The first case was for an outside containment SLB with the equilibrium iodine concentration for continued full power operation in combination with an assumed accident initiated iodine spike.
The Reference Cycle analysis of this case remains valid.
The resultant doses were a small fraction of 10CFR100 limits.
7-14 1
The second case was for pn outside containment SLB with a pre-existing iodine spike.
The Reference Cycle analysis of this case remains valid.
The resultant doses were within 10CFR100 limits.
The third case was for an outside containment SLB with a 1.5% predicted fuel failure. The resultant calculated two hour site boundary doses were less than 250 REM thyroid and less than 1. 2 REM whole body.
This is within 10CFR100 limits.
Table 7.1.5a-3 contains the sequence of events for the outside containment pre-trip SLB with LOAC. The key plant parameters (core power, core heat flux, RCS pressure, RCS temperatures, steam generator pressure and reactivities) are shown in Figures 7.1.5a-1 through 7.1.Sa-6.
7.1.Sa.4 Conclusions The results of the pre-trip analysis of SLB with LOAC demonstrate that a coolable geometry is maintained during this event as the number of fuel pins calculated to fail is small.
For an accident initiated iodine spike, the resultant doses are a small fraction of the 10CFR100 limits.
For a pre-existing iodine spike or for the predicted fuel failure, the resultant doses are within the 10CFR100 limits.
l 7-15
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Table 7.1.5a-1 Key Parameters Assumed for the Steam Piping Failures Event Inside Containment Pre-Trip Power Excursions, Reference Cycle Cycle 2 Parameter Units Value Value Total RCS Power MWt 3478 3478 (Core Thermal Power
+ Pump Heat)
Initial Core Coolant
- F 560 560 Inlet Temperature Initial Reactor Coolant psia 2000 2000 System Pressure Initial Reactor Coolant gpm 356,400 356,400 System Flow Initial Steam Generator Pressure psia 963 976 Effective Moderator Temperature 10~4ap/*F
-3.3
-3.3 Coefficient 2
Break Size ft 7.88 5.0 l
/-16 l.
Table 7.1.5a-2 Key Parameters Assumed for the i
Steam Piping Failures Event Outside Containment Pre-Trip Power Excursions Cycle 2 Parameter Units Value Total RCS Power MWt 3478 (Core Thermal Power
+ Pump Heat)
Initial Core Coolant
- F 560 Inlet Temperature Initial Reactor Coolant psia 2000-System Pressure Initial Reactor Coolant gpm 356,400 System Flow Initial Steam Generator Pressure psia 976 Effective Moderator Temperature 10-4ao/*F
-3.3 Coefficient 2
Break Size ft 3.14 7-17
Table 7.1.5a-3 Sequence of Events for the Steam System Piping Failure Event Outside Containment Pre-Trio Power Excursions with LOAC Time (sec)
Event Setooin,t or Value 2
0.0 Failure in the Main Steam System 3.14 ft Piping 6.15 CPC V0PT RPS Trip Condition 115% of 3410 MWt 6.70 Trip Breakers Open 7.0 CEAs Begin to Drop 7.1 LOAC Occurs, Reactor Coolant Pumps Begin to Coastdown 7.4 Maximum Core Power 126% of 3410 MWt 7.8 Maximum Core Heat Flux 115% of 3410 MWt 8.8 Main Steam Isolation Signal 675 psia 9.7 Main Steam Isolation Valves Begin to Close 19.7' Main Steam Isolation Valves Closed 20.3 Safety Injection Signal 1560 psia 27.3 Pressurizer Empties 38.8 Safety Injection Pumps Reach Full Speed 7-18
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STEAM LINE BREAK, PRE-TRIP POWER EXCURSIONS Wotorford Steam CORE POWER vs TIME Electric Station 7-19
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7.1.5b Steam System Piping Failure, Post-Trip Return to Power The Steam t.irf6 Break (SLB) post-trip return to power was reanalyzed due to a more adverse moderator cooldown curve and an increase in maximum inverse boron worth.
This analysis was done to demonstrate that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.
7.1.5b.1 Identification of Causes A break in the main steam system piping will cause an increase in steam flow.
This increase in flow results in increased heat removal frcm the Reactor Coolant System (RCS).
In the presence of a negative moderator temperature coefficient of reactivity (MTC) the cooldown will cause positive reactivity to be added to the core. Highly negative MTCs and large break sizes can combine to cause an approach to criticality.
This approach to criticality is teminated by the addition of safety injection boron and the increase in temperature following either 1.
Tennination of steam flow and heat removal by the action of the MSIVs in both steam lines.
or 2.
Termination of steam flow from the unaffected steam generator by the MSIV action and dryout of the affected steam generator.
The Hot Full Power. (HFP) and Hot Zero Power (HZP) Steam Line Break (SLB)
Events with and without loss of Offsite AC Power (LOAC) were analyzed to determine that critical heat fluxes are not exceeded during this event and site boundary doses do not exceed 10CFR100 guidelines.
7.1.5b.2 Analysis of Effects and Consequences It was found that the HFP SLB with LOAC was the most limiting of the four SLB cases described above.
The analytical basis for the HFP simulation is discussed below.
7-25
2 A.
A double-ended guiljotine break (7.88 ft ) causes the greatest cooldown of the RCS and the most severe degradation of shutdown margin.
)
8.
A break inside the containment building, upstream of the MSIVs causes a non-isolatable condition in the affected steam generator.
This results in an. approach to criticality which is terminated by the dryout of the affected. steam generator.
C.
A reactor. trip is initiated by either Low Steam Generator Pressure, Low Steam Generator Water Level, High Linear Power Level, Low ONBR, or Delta-
~ Pressure Low Flow Trip (Loss of AC Power).
. D.
A safety injection actuation signal (SIAS) is actuated when the pressur-izer pressure drops below the setpoint.
Time delays associated with the safety inj'ection pump acceleration, valve opening, and flushing of the
)
unborated safety injection lines are taken into account.
Additionally,
,the event was -initiated from the highest pressure allowed by the technical specifications to delay the effect of safety injection baron.
E.
The cooldown of the RCS is terminated when the affected steam generator blows dry.
As the coolant temperatures begin increasing, positive reactivity insertion from moderator reactivity feedback decreases.
The decrease in moderator reactivity combined with the negative reactivity inserted via boron injection cause the total reactivity to become more
~
negative.
j The' conservative assumptions included in the HFP simulation are discussed below.
I The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to the most negative value allcwed by the Technical Specifications.
This negative MTC results in the greatest positive reactivity addition during the RCS cooldown caused by the steam line break.
Since the coefficient of reactivity associated with moderator feedback varies significantly over the range of mcderator density covered in the analysis, a 7-26
.1-----
curve of reactivity insgrtion versus moderator density rather than a single value of MTC is assumed in the analysis. The moderator cooldown curve used in the analysis was conservatively calculated assuming that on reactor trip, the highest worth control element assembly is stuck in the fully withdrawn position.
The reactivity defect associated with fuel temperature decrease is also based on a most negative Fuel Temperature Coefficient (FTC). This FTC, in conjunction with the decreasing fuel temperatures, causes the greatest positive reactivity insertion during the steam line break event.
The delayed neutron fraction assumed is the maximum value including uncertainties for end-of-life conditions.
This too maximizes subcritical multiplication and thus increases the potential for return-to-power (R-T-P).
The minimum CEA worth assumed to be available for shutdown at the time of' reactor trip at the maximum allowed power level is -7.80%Ao.
This available scram worth corresponds to the moderator cooldown curve and stuck rod worth used in the analysis.
~
During the return-to-power, negative reactivity credit was assumed in the analysis.
This negative reactivity credit is due to the local heatup of the inlet fluid in the hot channel, which occurs near the location of the stuck CEA.
This credit is based on three-dimensional coupled neutronic-thermal-hydraulic calculations performed with the HERMITE/ TORC code (References 7-9
~
and 7-10) for Calvert Cliffs Uriit 1 Cycle' 7 (Reference 7-4).
Only a fraction of the negative reactivity credit justified for Calvert Cliffs Unit 1 Cycle 7 was used.
The analysis assumed that, on a safety injection actuation signal, one high pressure safety injection pump fails to start. A maximum inverse baron worth of 110 ppin/?.ao was conservatively assumed for safety injection.
A conserva-tive MSIV closure time of 10.0 seconds was assumed in this analysis.
The key parameters used for the post-trip R-T-P SLB are listed in Table 7.1.5b-1.
t 7-27
7.1.5b.3 Results The results for the Hot Zero Power SLB events and the Hot Full Power SLB witn no Loss of AC are bounded by those for the Hot Full Power SLB with concurrent LOAC presented herein.
Table 7.1.5b-2 presents the sequence of events for the HFP SLB with concurrent LOAC. The key plant parameters (core pcwer, core heat flux, RCS pressure, RCS temperatures, steam generator pressure and reactivities) are shown in Figure's 7.1.5b-1 through 7.1.5b-6.
The minimum post-trip DNBR experienced during the transient was 2.12 using the Macbeth low flow DNBR correlation.
This value results in no calculated fuel failure during the course of this transient.
7.1.5b.4 Conclusions The results of the post-trip analysis of SLB demonstrate that there is no calculated fuel failure, a coolable geometry is maintained, and the Cycle 2 radiological release is bounded by the Reference Cycle.
l l
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7-28
Table 7.1.5b-1 Key Parameters Assumed for the Post-Trip Return to Power Steam Line Break Event Reference Cycle Cycle 2 Parameter Units Value Value Total RCS Power MWt 3478 3478 (Core Thermal Power +
Pump Heat)
Initial Core Coolant Inlet "F
560 560 Temperature Initial Reactor Coolant Flow gpm 356,400 356,400 Initial Reactor Coolant psia 2000 2300 System Pressure CEA Worth at Trip
%Ao
-8.55
-7.80 Inverse Boron Worth ppm /%Ao 95
-110 Initial Steam Generator psia 963 976 Pressure High Pressure Safety Injection One Pump One Pump Pumps Inoperable Inoperable 2
Break Area ft 7.88 7.88 Blowdown Fluid 100% Steam 100% Steam 7-29
Table 7.1.5b-2 2
Sequence of Events for the Hot Full Power. 7.88 ft,
Inside Containment Steam Line Break with Loss of Offsite Power Time (sec)
Event Setpoint or Value 2
0.0 Double-Ended Guillotine Break 7.88 ft in a Main Steam Line with Concurrent LOAC, Reactor Coolant Pumps Begin to Coast Down 1.7 Reactor Trip Signal Generated on 675 psia Low Steam Generator Pressure, Main Steam Isolation Signal 2.6 Trip Breakers Open 2.7 MSIVs Begin to Close 2.9 CEAs Begin to Drop 12.7 MSIVs are Completely Closed 17.1 Pressurizer Empties 22.0 Safety Injection Actuation Signal 1560 psia Generated on Low Pressurizer Pressure 52.0 Safety Injection Pumps Reach Full Speed 100.9 Affected Steam Generator Empties 136.1 Maximum Post-Trip Power 6.8% of 3410 MWt 154.2 Ma'ximum Post-Trip Reactivity
.012%ao 165.5 Minimum Post-Trip MacBeth DNBR
>1.30 1
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Weterford Steen Electne 5,,,,oa REACTIVITIES VS TIME 7.1.Sb-6 7-36
7.2-Decrease in Heat Removal by the Secondary System 7.2.1 Loss of External Load The results are bounded by the Reference Cycle.
7.2.2 Turbine Trip 1
The results are bounded by the Reference Cycle.
7.2.3 Loss of Condenser Vacuum The results are bounded by the Reference Cycle.
7.2.4 Loss of Normal AC Power The results are bounded by the Reference Cycle.
7.2.5 Loss of Normal Feedwater The results are bounded by the Reference Cycle.
7,2,6 feedwatgp $ygtgg pjpg @pgaks 1
The results are bounded by the Reference Cycle.
7-37
7.3 Decrease in React'or Coolant Flowrate 7.3.1 Partial Loss of Forced Reactor Coolant Flow The results are bounded by the Reference Cycle.
7.3.2 Total Loss of Forced Reactor Coolant Flow The 4-pump Loss of Coolant Flow (LOF) Event is analyzed to determine the' minimum initial margin that must be maintained by the Limiting Conditions for Operation (LCOs) such that in conjunction with the Reactor Protection System (RPS) the DNBR SAFDL will not be exceeded. This event was reanalyzed due to a change in the CPCs.
The CPCs were modified to include a trip on low reactor coolant pump (RCP) speed which replaced the previous trip on low projected DN8R.
7.3.2.1 Identification of Causes A loss of coolant flow may result from a loss of electrical power to one or more of the four reactor coolant pumps.
Reactor trip on loss of coolant flow is initiated by the CPCs on low RCP speed.
For a loss of flow at any power operating cond.ition, a trip will be initiated when the RCP shaft speed drops to 96.5 percent of its initial speed.
The reduction in core flow lags the decrease in RCP shaft speed.
7.3.2.2 Analysis of Effects and Consequences The 4-pump Loss of Flow transient is characterized by the flow.coastdown curve given in Figure 7.3.2-1.
Table 7.3.2-1 presents the initial conditions assumed for this event.
i 7-38
.~
7.3.2.3 Results Table 7.3.2-2 presents the sequence of events for the 4-pump loss of Flow Event. This is a representative case and is initiated at an axial shape index of zero.
The CPC low RCP speed trip setpoint is reached at 0.55 seconds and the CEAs start to drop into the core 0.60 seconds later. A minimum CE-1 ONBR of 1.26 is reached at 2.5 seconds.
Figures 7.3.2-2 to 7.3.2-5 present the core power, heat flux, RCS pressure, and RCS temoeratures as a function of time.
7.3.2.4 Conclusions The 4-pump Loss of Flow event initiated from the Technical Specification LCOs in conjunction with the CPC trip on low RCP shaft speed does not exceed the DN8R SAFDL.
4.
7-39
Table 7.3.2-1 Key Parameters Assumed for the Total loss of Forced Reactor Coolant Flow Event Reference Cycle Cycle 2 Parameter Units Value Value Total RCS Power MWt 3478 3478 (Core Thermal Power
+ Pump Heat)
Initial Core Coolant
'F 560 560 Inlet Temperature Initial RCS Vessel Flow Rate gpm 396,000 396,000 Initial Reactor Coolant psia 2400 2350 System Pressure 10-# o/*F
+0.5
+0.5 Moderator Temperature Coefficient A
Doppler Coefficient Multiplier 0.85 0.85 Projected 96.5% Initial RPS Trip Setpoint DNBR of 1.20 shaft speed CEA Worth at Trip
%Ao
-7.95
-6.0 4-Pump RCS Flow Coastdown Figure 7.3.2-1 Figure 7.3.2-1 D
7-40
Table 7.3.2-2 Sequence of Events for Total Lo:: of Forced Reactor Coolant Flow Event Time (sec)
Event Setpoint or Value 0.0 Loss of Power to all Four Reactor Coolant Pumps 0.55 Low Reactor Coolant Pump Shaft Speed 96.5% of Initial Trip Condition shaft speed 0.85 Trip Breakers Open 1.15 CEAs Begin to Drop 2.50 Minimum CE-1 DNBR 21.26 4.5 Maximum RCS Pressure 2518 psia 7-41
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7.3.2-3 Electne Station REACTOR COOLANT SYSTEM TEMPERATURES VS TIM 7-46
7.3.3 Single Reactor Coolant Pump Shaft Seizure / Sheared Shaft The single reactor coolant pump shaft seizure / sheared shaft with loss of Offsite AC power was reanalyzed due to a change in the fuel failure pin census and CEA worth at trip.
The event was reanalyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed 10CFR100 guidelines.
Both the RCP shaft seizure and sheared shaft were evaluated for the current cycle.
Loss of Offsite AC power was evaluated for each and the sheared shaft was determined to be more limiting.
7.3.3.1 Identification of Causes l
l A single reactor coolant pump sheared shaft is caused by mechanical failure of the pump shaft.
Following the shearing of a reactor coolant pump shaft, the core flowrate rapidly decreases to the value that wot.ld occur with only three i
reactor coolant pumps operating.
The reduction in coolant flowrate causes an l
increase in the average coolant temperature in the core and may produce a i
departure from nucleate boiling (DNB) condition in some portions of the core.
A reactor trip is generated when the rapid flow reduction in the affected loop decreases steam generator delta-pressure below the delta-pressure low flow I
trip setpoint.
The reactor trip produces an automatic turbine trip.
Following turbine trip, AC power is assumed to be unavailable due to a failure to fast transfer to offsite power.
The loss of Offsite AC power (LOAC) results in the coastdown of the remaining three reactor coolant pumps, which further decreases the reactor coolant system flow.
The operator can initiate a controlled system cooldown using the atmospheric dump valves any time after reactor trip after restoring power using the diesels.
7.3.3.2 Analysis of Effects and Consecuences i
i The sheared shaft plus LOAC was assumed to occur at hot full power and at core thermal hydraulic conditions such that the minimum initial margin allowed by the Limiting Conditions for Operation (LCOs) is available.
Table 7.3.3-1 l
l l
7-47
contains the initial conditions for Cycle 2 and for the Reference Analysis (Reference 7-1). The minimum DNBR for this event was calculated with the TORC computer code and the calculation of the amount of fuel failure employed the statistical convolution method described in Reference 7-12.
7.3.3.3 Results The sheared shaft with loss of Offsite AC power results in a minimum calculated CE-1 DN8R of 0.86.
This results in a predicted ' fuel failure of less than 8.5%.
The resultant offsite doses are less than 30 REM thyroid and less than 2.5 REM whole body.
Additionally, the peak RCS pressure is less than 2750 psia.
7.3.3.4 Conclusions For the sheared shaft with the loss of Offsite AC power the radiological doses are less than 10% of the 10CFR100 limits of 300 REM thyroid dose and 25 REM whole body dose.
The consequences of the sheared shaft with LOAC are more limiting than the seized rotor with LOAC.
r 7-48
Table 7.3.3-1 4
Key Parameters Assumed for the Single Reactor Coolant Pump Sheared Shaft Event with Loss of AC Power i
Reference Cycle 2 Parameter Units Cycle Value Value Initial Core Power Level MWt 3478 3478 Core Inlet Coolant Temperature
- F 560 560 0
Core Mass Flowrate 10 lbm/hr 137.2 142.5 Reactor Coolant System Pressure psia 2115 2100 i
Maximum Radial Power Peaking Factor 1.59 1.7 I
Moderator Temperature Coefficient 10-4 ap/*F 0.5 0.5 Doppler Coefficient Multiplier 0.85 0.85
]
CEA Worth for Trip
%ao
-8.55
-6.00 I
i 7-49
Table 7.3.3-2 Sequence of Events for Single Reactor Coolant Pump Sheared Shaft Event With Loss of AC Power Time (sec)
Event Setpoint or Value 0.0 Sheared Shaft Occurs on One Reactor Coolant Pump 0.5 Low Reactor Coolant System Flow 0.60 of Initial RCS Trip Condition Flow through the Affected SG 1.20 Trip Breakers Open 1.50 CEAs Begin to Drop into Core Turbine Generator Trip; Loss of AC Power Occurs 2.90 Minimum CE-1 DNBR 20.86 6.4 LH (Unaffected Steam Generator) Safety Valves Open 1096 psia 6.7 RH (Affected Steam Generator) Safety Valves Open 1096 psia 6.9 Maximum RCS Pressure 2490 psia 11.3 Maximum SG Pressure 1127 psia 7-50
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7.4 Reactivity and Pcwer Distribution Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or low Power Condition The uncontrolled CEA withdrawal (CEAW) from suberitical or low power condi-tions is analyzed to ensure that the departure from nucleate boiling ratio (DNBR) and the fuel centerline melt (CTM) specified acceptable fuel design limits (SAFDLs) are not violated. Additionally, the CEAW from subcritical and low powers is analyzed to verify that the peak RCS pressure is less than the design limit of 2750 psia. The CEAW from low power is included to demonstrate the benefit gained from the new Core Protection Calculator (CPC) Variable Overpower Trip (V0PT).
The results of the analysis with this trip are improved over the Reference Cycle results.
The CEAW from subcritical conditions is included due to an assumed increase in the subcritical reactivity addition rate.
7.4.1.1 Identification of Causes An uncontrolled withdrawal of CEAs is assumed to occur as a result of a single failure in the control element drive mechanism (CEDM), control element drive mechanism control system (CEDMCS), reactor regulating system, or as a result of operator error.
7.4.1.2 Analysis of Effects and Consecuences The withdrawal of CEAs from subcritical or low power conditions adds reac-tivity to the reactor core, causing both the core power level and the core heat flux to increase together with corresponding increases in reactor coolant temperatures and reactor coolant system (RCS) pressure.
The withdrawal motion of CEAs also produces a time dependent redistribution of core power.
These transient variations in core thermal parameters result in the system's approach to the specified fuel design limits and RCS and secondary system pressure limits, thereby requiring the protective action of the Reactor Protection System (RPS).
7-57
The reactivity insertion, rate accompanying the uncontrolled CEA withdrawal is dependent primarily upon the CEA withdrawal rate and the CEA worth since, at subcritical and lower power conditions, the normal reactor feedback mechanisms do not occur until power generation in the core is large enough to cause changes in the fuel and moderator temperatures.
The reactivity insertion rate determines the rate of approach to the fuel design limits.
Depending on the system initial conditions and reactivity insertion rate, the uncontrolled CEA withdrawal transient is terminated by either a high logarithmic power trip, high power level trip, high pressurizer pressure trip, CPC lo'w departure from nucleate boiling ratio (DNBR) trip, CPC high local power density trip, or CPC variable overpower trip (V0PT).
A CEA withdrawal from subcritical conditions was initiated from the conditions
~
in Table 7.4.1-1.
A moderator temperature coefficient (MTC) of +0.5x10-4e/*F was used in this analysis.
This MTC, in conjunction with the increasing core coolant temperatures, yields an increase in core heat flux.
The least negative fuel temperature coefficient (FTC) is used in this analysis.
The minimum CEA worth assumed for shutdown at time of reactor trip is 5.0% for
- ero power critical (Mode 2) operation and 4.0% for subcritical (Mode 2) operation.
7.4.1.3 Results The uncontrolled CEA withdrawal from subcritical conditions resulted in a reactor trip on.high logarithmic power at 73.8 seconds.
The minimum CE-1 CNBR calculated for this event initiated from the conditions of Table 7.4.1-1 was greater than the design limit of 1.26.
The peak linear heat generation rate (PLHGR) was calculated to be in excess of the steady state ac<:eptable fuel to centerline melt (CTM) limit of 21 kw/ft.
Fuel centerline temperature does not exceed 4900*F and the fuel is not predicted to melt.
Additionally, the peak RCS pressure is less than the design limit of 2750 psia.
Table 7.4.1-2 presents the sequence of events for this event.
Figures 7.4.1-1 through 7.4.1-6 present the NSSS response for core power, core heat flux, RCS temperatures, RCS pressure, reactivities and steam generator pressure.
7-58
The uncontrolled CEA withdrawal from low power is presented because the V0PT added to the CPCs is credited to mitigate the consequences of this event.
The low power CEAWs were analyzed to determine the maximum RCS pressure increase and to evaluate the potential for fuel performance degradation.
The initial conditions for the CEAW that maximize fuel performance degradation are listed in Table 7.4.1-3.
V0PT is generated at 32.35 seconds and the CEAs begin to drop at 33.20 seconds.
The sequence of events is presented in Table 7.4.1-4.
Figures 7.4.1-7 through 7.4.1-12 present the NSSS~ response for this event.
The results of the CEAW event to maximize the RCS pressure demonstrated that the peak pressure expected for the Cycle 2 transient is less than that presented in the FSAR.
7.4.1.4 Conclusions An uncontrolled CEA withdrawal from either subcritical or low power conditions does not exceed the DN8R or CTM limits The RCS pressure limit of 2750 psia is not exceeded during this event.
7-59
Table 7.4.1-1 Key Parameters Assumed in the CEA Withdrawal From Subcritical Concitions Event Parameter Units Cycle 2 Initial Core Power Level MWt 3478 x 10-10 Initial Inlet Coolant
- F 545 Temperature 6
Initial Core Mass Flow Rate 10 lbm/hr 150.2 Initial RCS Pressure psia 2000 Moderator Temperature Coefficient 10-4 ao/*F 0.5 Fuel Temperature Coefficient 0.85 Multiplier Minimum CEA Worth at Trip
%ap
-4.0 Maximum Reactivity Addition 10~4ao/sec 1.9 Rate l
7-60
Table 7.4.1-2 Sequence of Events for the CEA Withdrawal from Subcritical Time (sec)
Event Setpoint or Value 0.0 CEAW Initiated 52.3 Reactor Reaches Criticality 73.8 High Logarithmic Power Trip Condition 0.6% of 3410 MWt 74.35 Trip Breakers Open 74.6 Peak Core Power 115.3% of 3410 MWt 74.65 CEAs Begin to Drop 74.9 Peak Core Heat Flux 28.5% of 3410 MWt 74.9 Minimum CE-1 DNBR 21.40 f
i 7-61
Table 7.4.1-3 Key Parameters Assumed for the CEA Withdrawal From Low Power Event Reference Parameter Units Cycle Cycle 2 Initial Core Power Level MWt 3410x10-6 3410x10-6 Initial Inlet Coolant
- F 530.5' 545 Temperature 6
Initial Core Mass Flow Rate 10 lbm/hr 175.4 145.5 Initial RCS Pressure psia 2000 2000,
Moderator Temperature Coefficient 10-4 ao/'F 0.5 0.5 Fuel Temperature Coefficient 0.85 0.85 Multiplier Minimum CEA Worth at Trip
%ao
-4.45
-5.00 Maximum Reactivity Addition 10-4ao/sec 1.1 1.1 Rate Pressurizer Pressure Control System Manual Mode (heaters off)
(sprays fully on) 7-62
Table 7.4.1-4 Sequence of Events for the CEA Withdrawal from Low Powers Event Time (sec)
Event Setpoint or Value 0.0 CEAW Initiated 32.35 V0PT Trip Condition 40% of 3410 MWt 32.90 Trip Breakers Open 33.20 CEAs Begin to Drop 33.3 Peak Core Power 65% of 3410 MWt 33.6 Peak Core Heat Flux 31% of 3410 MWt 33.4 Minimum CE-1 DNBR 2 2.0 9
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7.4.2 Uncontrolled CEA Withdrawal at Power The results are bounded by the Reference Cycle.
7.4.3 CEA Misoperation Event The single full length and part-length CEA drop events are analyzed to detennine the initial thermal margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNBR and Fuel' Centerline Melt (CTM) specified acceptable fuel design limits (SAFDLs) will not be violated.
The CEA position-related penalty factors for downward single CEA deviations in the Control Element Assembly Calculators (CEACs) have been set equal to one (no penalty).
This applies to both full-length and part-length CEA downward deviations.
Sufficient thermal margin will be maintained by the LCOs to compensate for' the removal of CEA position-related penalty factors for downward ' single CEA deviations.
A part-length PDIL has been added which restricts the part-length CEA insertion to less than 25% for power levels greater than 50%.
From these initial conditions the part-lendh single or subgroup drop inserts only negative reactivity similar to full-length single or subgroup drop. The method used to analyze the single CEA Drop event is the same as the method described in Reference 7-13.
For CEA subgroup drops, the CEA position-related penalty factors for downward deviations are still used by the Core Protection Calculators as in the Reference Cycle to provide a trip when necessary.
7.4.3.1 Identification of Causes The CEA Misoperation Events are defined as the inadvertent release of a single CEA or CEA subgroup causing it to drop into the core.
The occurrence of an electrical or mechanical failure in a CEA drive mechanism could result in a CEA drop.
7-76
7.4.3.2 Analysis of Effects and Consequences The single full length CEA Drop is analyzed since this event requires the maximum initial margin to be maintained by LCOs.
Table 7.4.3-1 presents the initial conditions assumed in the analysis.
Additional conservative assumptions include:
a)
The turbine load is not reduced, but is assumed to remain the same as prior to the CEA drop.
This results in a power mismatch between the primary and secondary systems, which leads to a cooldown of the RCS.
b)
The most negative moderator and fuel temperature coefficients of reactivity are used because these coefficients produce the minimum RCS coolant temperature decrease upon return to 100 percent power and thus minimize DNBR.
c)
Charging pumps and pressurizer heaters are assumed to be inoperable during the transient.
This maximizes the pressure drop during the event and minimizes DNBR.
d)
All other systems are assumed to be in the manual mode of operation and have no impact on this event.
The event is in.itiated by dropping a full length CEA over a period of 1.0 second.
A value of 9.0 percent is used for the radial peaking factor increase.
The axial power shape in the hot channel remains unchanged and, hence, the increase in the 3-D peak for the maximum power is directly proportional to the maximum increase in radial peaking factor of 9.0 percent.
Since there is no trip assumed, the peaks will stabilize at these asymptotic values after a few minutes as the secondary side continues to demand 100 percent power.
7-77
7.4.3.3 Results Table 7.4.3-2 presents an illustrative sequence of events for the Full Length CEA Drop Event initiated at the conditions described in Table 7.4.3-1.
A minimum CE-1 DNBR of greater than 1.26 is obtained at 900 seconds as determined from the 9.0 percent radial power peaking increase following CEA drop plus 15 minutes of xencn redistribution at the final coolant conditions.
At this time the operator will take action to reduce power in accordance with Figure 3.1-1A of the Technical Specifications, if the misaligned CEA has not been realigned.
A maximum allowable initial linear heat generation rate of 17.0 kW/ft could exist as an initial condition without exceeding the Acc'eptable Fuel Centerline Melt Limit of 21.0 kW/ft during this transient.
This amount of margin is assured because the linear heat rate LCO is based on the more limiting allowable linear heat rate for LOCA (13.4 kW/ft, see Table 7.0-6).
The results for the CEA Subgroup Drops are no worse than the Reference Cycle.
7.4.3.4 Conclusions The full length CEA drop event initiated from the Technical Specification LCOs does not violate the DNBR and CTM SAFDLs.
7-78
Table 7.4.3-1 Key Parameters Assumed for the Full Length CEA Drop Event Reference Cycle Cycle 2 Parameter Units Value Value Total RCS Power MWt 3478 3410**
(Core Power +
Pump Heat)
Initial Core Coolant "F
560 553**
Inlet Temperature Initial Reactor Coolant psia 2000 2250**
System Pressure Initial RCS Vessel gpm 396.000 396,000 Flow Rate Moderator Temperature 10-4ao/*F
-3.3
-3.3 Coefficient Doppler Coefficient 1.15 1.15 Multiplier CEA Insertion at Maximum
% Insertion of 28 28 Allowed Power Bank 6
% Insertion of 90 25 Part-Length Dropped CEA Worth
%ao
.059
.05 CEA Drop Radial 1.090 Peaking Distortion Factor Xenon Redistribution 1.043 Radial Peaking Distortion Factor for 15 Minutes
.i
- CPC applied the penalty directly to DNBR and LPD calculations.
4
- Nominal parameters are used to illustrate the transient responses for this event.
7-79
Table 7.4.3-2 Sequence of Events for Full Length CEA Drop Time (sec)
Event Setpoint or Value 0.0 CEA Begins to Drop into Core 1.0 CEA Reaches Full Inserted Position 100% Inserted 1.1 Core Power Level Reaches Minimum and 92.5% of Initial Begins to Increase due to Reactivity Feedbacks 27.9 Core Power Returns to its Maximum Value 100% of Initial 250.0 Core Inlet Temperature Reaches a New 552*F Steady State Value 250.0 Reactor Coolant System Pressure Reaches 2235 psia New Steady State Value 900.0 Minimum DNBR is Reached
> 1.26 901.0 Operator Action - Core Power is Reduced if Dropped CEA has not been Realigned l
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7-80
7.4.4 CVCS Malfunction (Inadvertent Baron Dilution)
The Inadvertent Baron Dilution event is analyzed for Cycle 2 to demonstrate that sufficient time is available for an operator to identify the event and to terminate an approach to criticality for all subcritical modes of operation.
The results of the analyses establish shutdown margin requirements for Modes 3 through 6.
This event was reanalyzed due to the increase in critical boron concentrations as shown in Table 7.4.4-1.
The results of an inadvertent boron dilution at power (i.e., Modes 1 & 2) are no more severe than for Cycle 1.
7.4.4.1 Identification of Causes A CVCS malfunction which results in unborated water being pumped at the maximum possible rate into the RCS by the demineralized water supply system is assumed to occur. For this to occur, one or more charging pumps must be on, the primary makeup water pumps must be on, and the demineralized water supply system must be aligned to supply water to the charging pump suction via the volume control tank. Since at least three simultaneous equipment malfunctions would be required to produce the above conditions, the incident could only be the result of improper operator action accompanied by a single equipment malfunction.
l In Mode 5 with the RCS partially drained for system maintenance, operation of l
only one charging pump at a maximum rate of 44 gpm is allowed when Keff is between.96 and.98.
Power to the other two charging pumps is required to be l
removed with their breakers locked out.
In Mode 5 with the RCS partially l
drained, only one charging pump is required to be locked out when Keff is l
between.96 and.94.
In Mode 5 with the RCS partially drained', no charging l
l pumps are required to be locked out when Keff is less than.94.
l l
In Mode 6, operation of only one charging pump at a maximum rate of 44 gpm is l
allowed. Power to the other two charging pumps is required to be removed with their breakers locked out.
I l
7-81
y 7.4.4.2 Analysis of Effects and Consequences For the subcritical modes (i.e., Modes 3 through 6), the time required to achieve criticality due to boron dilution is dependent on the initial and critical boron concentrations, the inverse boron worth, and the rate of dilution.
Table 7.4.4-1 compares the values of the key transient parameters assumed in each mode of operation for Cycle 2 and the Reference Cycle. The analysis conservatively assumed higher critical boron concentrations and lower inverse boron worths than.are expected for Cycle 2.
These choices decrease the calculated times to criticality in initially subcritical modes. The time to criticality was determined by using the same methodology as in the FSAR.
-(Reference.7-1,Section 15.4.1.4.3).
s 7.4.4.3 Results,
Table 7.4.4-2 compares the results of the analysis for Cycle 2 with those for the Reference Cycle. The key results are the minimum times required to lose the prescribed negative _ reactivity in each operational mode.
As for Cycle 1, the startup flux channel alarm will provide indi. cation of any baron dilution event before the reactor reaches critical. For some conditions this alarm may be the first positive indication that a dilution is in progress. For these cases, the alarm will assure that the minimum specified intervals of 30. minutes for Mode 6 and 15 minutes for Modes 3-5 will exist
' after the alarm announces to terminate the event.
7.4.4.4 Cenclusion
' The results of this analysis demonstrate that sufficient time exists for the
}
operator to take appropriate action to identify and mitigate the consequences of the Inadvertent Boron Dilution Event.
7-82
Table 7.4.4-1 Key Parameters Assumed in the Inadvertent Boron Dilution Event Reference Cycle Cycle 2 Parameter Value Value Critical Baron Concentration, PPM (All Rods Out, Zero Xenon)
Startup (Mode 2) 882 1670 Hot Standby (Mode 3) 882 1670 Hot Shutdown (Mode 4) 882 1670 Cold Shutdown (Mode 5) 940 1670 Refueling (Mode 6) 940 1600 Inverse Boron Worth, PPM /%ao Startup 68 80 Hot Standby 68 75 Hot Shutdown 61 75 Cold Shutdown 56 75 Refueling 56 75 Minimum Initial Subcriticality Assumed, %ao Startup 5.15 2.0 Hot Standby 5.15 2.0 Hot Shutdown
!. '5 2.0 Cold Shutdown 2.0 2.0 Refueling N/A*
5.0
- Technical Specification minimum refueling boron concentration with uncertainty is assumed.
7-83
Table 7.4.4-2 Results of the Inadvertent Boron Dilution Event Time to Criticality Acceptance Criterion Mode (Minutes)
(Minutes)
Reference Cycle Cycle 2 Startup (Mode 2)
>72
>40 15 Hot Standby (Mode 3)
>72
>40 15 Hot Shutdown (Mode 4)
>72
>45 15 Cold Shutdown'(Mode 5)
RCS Full
>72
>45 15 RCS Partially Drained *
>72*
2>40*
15
>40**'
15
>40***
15 Refueling (Mode 6)
>72
>105****
30
I
7-84
7.4.5 Startup of an (nactive Reactor Coolant Pump Event The results are bounded by the Reference Cycle.
7.4.6 Control Element Assembly Ejection The results are bounded by the Reference Cycle.
7.5 Increase in Reactor Coolant System Inventory 7.5.1 CVCS Malfunction The results are bounded by the Reference Cycl'e.
7.5.2 Inadvertent Operation of the ECCS During Power Operation The results are bounded by the Reference Cycle.
7.6 Decrease in Reactor Coolant System Inventory 7.6.1 Pressurizer Pressure Decrease Events The results are bounded by the Reference Cycle.
7.6.2 Small Primary Line Pipe Break Outside Containment The results are bounded by the Reference Cycle.
7.6.3 Steam Generator Tube Rupture The results are bounded by the Reference Cycle.
7-85
7.7 Miscellaneous l
7.7.1 Asymmetric Steam Generator Events The transients resulting from the malfunction of one steam generator are analyzed to determine the initial margins that must be maintained by the LCO's such that in conjunction with the RPS (CPC high differential cold leg temperature trip) the DNBR and Fuel Centerline Melt (CTM) SAFDLs are not violated. This event is presented due to a change in analytical methodology.
7.7.1.1 Identification of Causes The four events which affect a single steam generator are identified below:
a) Loss of Load to One Steam Generator (LL/ISG) b) Excess Load to One Steam Gerierator (EL/ISG) c) Loss of Feedwater to One Steam Generator (LF/ISG) d) Excess Feedwater to One Steam Generator (EF/1SG)
Of the four events described above, it has been determined that the Loss of Load to One Steam Generator (LL/ISG) Event is the limiting asymmetric event.
Hence, only the results of this transient are reported.
The event is initiated by the inadvertent closure of a single Main Steam Isolation Valve (MSIV), which results in a loss of load to the affected steam l
t generator. Upon the loss of load to the single steam generator, its pressure and temperature increase to the opening pressure of the secondary safety valves and its water level decreases. The core inlet temperature of the loop with the affected steam generator increases resulting in a temperature tilt across the core.
In the presence of a negative moderator temperature coefficient the radial peaking increases in the cold side of the core, 7-86
resulting in a condition,which potentially could cause an approach to DNBR and CTM SAFDLs. The CPC high differential cold leg temperature trip serves as the primary means of mitigating this transient. Additional protection is provided by the steam generator low level trip.
7.7.1.2 Analysis of Effects and Consecuences The most negative value of the moderator temperature coefficient is assumed to maximize the calculated severity of the associated power peaking.
The LL/ISG is initiated at the initial conditions presented in Table 7.7.1-1 and is analyzed parametrically on axial shape index to determine the maximum initial margin needed to ensure the SAFDLs are not violated.
The NSSS response is calculated using the 'CESEC III code. The resulting core-parameters (core flow, RCS inlet temperature, RCS pressure, and reactor trip time) are the input into a 2-D simulation of the core using the HERMITE code.
HERMITE is used to model both the effects of the temperature tilt on radial power distribution and the space-time impact of the scram. The thernalmargin changes are evaluated with the CETOP code.
7.7.1.3 Results l
A reactor trip is generated by the CPC's at 6.0 seconds based on high differential cold leg temperature between the cold legs associated with the steam generators.
Table 7.7.1-2 presents the sequence of events for the loss of load to one steam generator.
Figures 7.7.1-1 to 7.7.1-5 show the NSSS response for core power, core heat flux, RCS temperatures, RCS pressure, and steam generator pressure.
The minimum CE-1 DNBR calculated for the LL/ISG Event is greater than 1.26.
7-87
A maximum allowable initial linear heat generation rate of 17.0 kw/ft could exist as an initial condition without exceeding the Acceptable Fuel to 1
i Centerline Melt Limit of 21.0 kW/ft during this transient.
j This amount of I
margin is assured because the linear heat rate LCO is based on the more limiting allowable linear heat rate for LOCA (13.4 kW/ft, see Table 7.0-6) 7.7.1.4 Conclusions The loss of load to one steam generator event, initiated from the Technical Specification LCO's does not violate the DNBR and CTM SAFDLs.
7-88
Table 7.7.1-1 Key Parameters Assumed for the Loss of Load to One Steam Generator Event Cycle 2 Parameter Units value Total RCS Power MWt 3478 (Core Thermal Power
+ Pump Heat)
Initial Core Inlet Temperature
'F 553 Initial Reactor Coolant System psia 2250 Pressure Moderator Temperature Coefficient 10-4 ao/*F
-3.3 Doppler Coefficient Multiplier 0.85 Radial Distortion Factor for a 1.13 18'F Core Inlet Temperature Asymmetry 6
7-89
~
Table 7.7.1-2 Sequence of Events for the Loss of Load to One Steam Generator Event Time (sec)
Event Setpoint or Value 0.0 Initiate Closure of a Single Main Steam Isolation Valve (MSIV) 0.1 MSIV on Affected Steam Generator is Fully Closed 0.1 Steam Flow from Unaffected Steam Generator Increases to Maintain Turbine Power 6.0 CPC Delta-T Setpoint Reached 18'F (Differential Cold Leg Temperature) 6.1 Safety Valves Open on Isolated Steam 1096 psia Generator 6.40 Trip Breakers Open 6.70 CEAs Begin to Drop 7.30 Minimum CE-1 DNBR
>1.26 10.9 Maximum Steam Generator Pressure 1135 psia 7-90
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1100 5
1000 w
E i
G w
E "O
3 "J
5 800 CTED 3.G.
v 700 I
I I
I 600 0
4 8
12 16 20 TIME, SECONOS l
l LOUISlANA POWER & LIGHT CO.
ASYMMETRIC STEAM GENERATOR Waterford Steam Electrie Station STEAM GENERATOR PRESSURE VS TIME 7.7.1-5 7-95
8.0 ECCS ANALYSIS
8.1 INTRODUCTION
AND
SUMMARY
An ECCS performance analysis of the limiting break size (0.8 DEG/PD*) was performed for WSES-3 Cycle 2 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1). The analysis justifies an allowable peak linear heat generation rate (PLHGR) of 13.4 kw/ft. This PLHGR is equal to the existing limit for WSES-3. The method of analysis and detailed results which support this value are presented herein.
8.2 METHOD OF ANALYSIS This analysis of WSES-3 Cycle 2 utilized the blowdown calculation for Cycle 1 (Reference 8-2). As in the first cycle, the calculations perfomed for this evaluation used the NRC approved C-E large break ECCS performance evaluation model which is described in i
Reference 8-3 through 8-8.
The refill /reflood hydraulic calculations were performed with the COMPERC-II code (Reference 8-6).
The hot rod clad temperature and clad oxidation were calculated with the STRIKIN-II (Reference 8-7) and PARCH (Reference 8-8) codes.
The COMPERC-II run was made to implement a new axial power shape determined by the procedure described in Reference 8-11.
- Further, l
the containment heat sink and containment free volume data were updated. No single failure was assumed. Significant core and system parameters for Cycles 1 and 2 are shown in Table 8-1.
- DEG/PD = Double-Ended Guillotine at pump discharge.
8-1
Also employed in this analysis is the C-E flow blockage model as described in Reference 8-9.
This blockage model addresses NRC's NUREG-0630 (Reference 8-10) cladding deformation and flow blockage analysis guidelines.
In this C-E model, new rupture temperature, rupture strain and flow blockage models, adopted from NUREG-0630 are used in the STRIKIN-II and PARCH codes. Also the steam cooling heat transfer coefficients calculated by the PARCH code, for use during the less than 1.0 inch /second reflood rate time interval, are calculated using an explicit method for redistribution of steam flow around the blockage region. The core wide clad oxidation percentage was estimated based on the pin census of Cycle 2.
8.3 RESULTS Table 8-2 presents the analysis results for the 0.8 DEG/PD break which produces the highest peak clad temperature. For comparison the results for Cycle 1 are also presented. The results of the evaluation confirm that 13.4 kw/ft is an acceptable value for the PLHGR in Cycle 2.
The peak clad temperature and maximum local and core wide clad oxidation values, as shown in Table 8-2, are well below the 10CFR50.46 acceptance limits of 2200'F,17% and 1%,
respectively. Table 8-3 presents a list of the significant parameters displayed graphically for the limiting 0.8 DEG/PD break.
Burnup dependent hot rod calculations were performed with STRIKIN-II to determine the initial fuel conditions which result in the highest peak clad temperature (PCT). This study demonstrated that the burnup with the highest initial fuel stored energy results in the highest PCT. This occurred at a hot rod burnup of 1000 MWD /MTV.
The 0.8 DEG/PD break produced the peak clad temperature of 2150*F.
For the 0.8 DEG/PD break the peak local oxidation (PLO) was calculated to be 7.8%.
8-2
8.4 CONCLUSION
The ECCS performance evaluation for WSES-3 Cycle 2 results in a peak clad temperature of 2150*F, a peak local clad oxidation percentage of 7.8% and a peak core wide clad oxidation percentage of less than 0.805% compared to the acceptance criteria of 2200*F,17% and 1%,
respectively. Therefore, operation of WSES-3 Cycle 2 at a core power level of 3458 Mwt (102% of 3390 Mwt) and a PLHGR of 13,4 kw/ft is in conformance with 10CFR50.46.
j i
)
8-3
TABLE 8-1 WSES-3 Cycle 2 Core and System Parameters Reference Parameter (Units)
Cycle Cycle 2 Reactor Power Level 9102% of Nominal 3458 3458 Average Linear Heat Rate 9 102% of 5.6 5.6 Nominal (Kw/Ft)
Peak Linear Heat Generation Rate (Kw/Ft) 13.4 13.4 Core Inlet Temperature (*F) 557.5 557.5 Core Outlet Temperature (*F) 618.6 618.6 0
6 System Flow Rate (1bm/hr) 148x10 148x10 0
0 Core Flow Rate (lbm/hr) 144x10 144x10 Gap Conductance (1) (Btw/hr ft.F) 1406 1534 2
Fuel Centerline Temperature (1)(.F) 3319.7 3321.6 Fuel Average TemperatureII)('F) 2131.4 2111.3 Hot Rod Gas Pressure (I)(psia) 1113.3 1113.3 Hot Rod Burnup (Mwd /Mtu) 676 1000 Number of Steam Generator Tubes Plugged per Steam Generator 0
250 Augmentation Factor Function of 1.0 Elevation Minimum Initial Containment 14.4 14.4 Pressure (psia) 3 6
6 Containment Free Volume (ft )
2.677(10 )
2.684(10 )
Axial Peaking Factor 1.52 1.53 (1) Initial values at the limiting hot rod burnup as calculated by STRIKIN-II at 13.4 kw/ft.
8-4
TABLE 8-2 WSES-3 Cycle 2 Limiting Break Size (0.8 DEG/PD)
Cycle 1 Cycle 2 Peak Linear Heat Generation 13.4 13.4 Rate (kw/ft)
Peak Clad Temperature (*F) 2170 2150 Time of Peak Clad Temperature 244.7 259.1 (Seconds)
Time of Clad Rupture (Seconds) 40.9 41.9 Peak Local Clad 0xidation (%)
8.35 7.80 Total Core-Wide Clad 0xidation (%)
(0.805
( 0.805 5
6 l
l l
8-5
,r-.
-n, ~,. - -, -, - - - - - - - - - - - - - - -. -. - - - - -. - ~.. -
TABLE 8-3 WSES-3 Cycle 2 Variables Plotted as a Function of Time for the Limiting Large Break Figure Variable Designation Containment Pressure 8-1 Mass Added to Core During Reflood 8-2 Peak Clad Temperature 8-3 Hot Spot Gap Conductance 8-4 Peak Local Clad 0xidation 8-5 Clad Temperature Centerline Fuel Temperature.
8-6 Average Fuel Temperature and Coolant Temperature for Hottest Node Hot Spot Heat Transfer Coefficient 8-7 Hot Rod Internal Gas Pressure 8-8 e *
- em>w
.p M,$
8-6 j
.,n._.._.--..,,,._.,,-,,,_.._.w._,
~_.___,,._._,...n
_..__.__n
._-y_.,,,,..,. _. _ _
I 60 000 SC.000 40 000 ;^
s 12a*
30 000 E
A 20.000 10.000 000 a
a a
a a
o o
9 9
aa o
a o
o a
a m
e,.
m a
m e-en TIME RFTER BRERK.
SEC LOUISLANA WATERFORD 3 CYCLE 2 Figure POWER & LIGHT CO.
0.8 DEG/PD 8-1
",'g,{*'d f,"*
CONTAINMENT PRESSURE 9
8-7 I
150000 TIME, SEC.
REFLOOD RATE, IN./SEC.
0 14.4 1.80 125000-14.4 62.4 1.22 62.4 600 0.68 100000 b
/
/
75000
-g<2 50000 25000 C
C C
C C
C C'
C O.
O o
l O
O O
O O
O N
v C
C C
j m
m e
I TIME AFTER CONTACT, SEC LOUISIANA WATERFORD 3 CYCLE 2 Figure POWER & LIGHT CO, 0.8 DEG/PD 82 W*'gjd S"m MASS ADDED TO CORE DURING REFLOOD g,
8-8
2200 C
2000 K
1800 1600
.j 1400 j
4 5
f k
1200 i
\\
d 1000 800 600 400 100 200 300 400 500 600 TIME.
SECOND3 LOutSLANA WATERFORD 3 CYCLE 2 Rgure i
POWER & LIGHT CO.
0.8 DEG/PO 83 Waterford Steam PEAK CLAD TEMPERATURE Electrie Sterion 89
180dr 1600 1400 1200
=?
a-+c5 E
1000 m
ui 800 g
8 L
e 600 400 200
~
V N
N l
l l
100 200 300 400 500 6CC TIME SECONOS LOUISIANA POWER & LIGHT CO.
WATERFORD 3 CYCLE 2 Figure Waterford Steam 0.8 DEG/PD 84 Electrie Storion HOT SPOT GAP CONDUCTANCE 8-10
18 16 14 12 m
j 10
-l 5
8 u
6
/
2 1
100 200 300 400 S00 600 TIME, SECONOS fGHT CO.
0 8 DEGjPD
^
PO R
f,*((,(*,'d Ng*
g, PEAK LOCAL CLAD OX1DATION 8-11
~ ' ~ '
2700 i
2400 FUEL
[
/ CENTERLINE 2100 CLAD AVERAGE 1800 s
h
\\
1500 f
\\
I l
\\
3<
5 k 1200 y
'l 900 600
\\
300 i
/
COOLANT 100 200 300 400 500 60C TIME, SECONOS LCU 65(AN A WATERFORD 3 CYCLE 2 - 0.8 DEG/PD Figure POWER & UGHT CO.
CLAD TEMPERATURE, CENTERLINE FUEL TEMPERATURE, 86 Waterford 5 ream AVERAGE FUEL TEMPERATURE AND COOLANT E!eer i: Srcrie" TEMPERATURE FOR HOTTEST NODE 8-12
180j 160 140 u.
A 120 t
x2 Ea p
100 m
M 8
80 x
m E
E 3
60 x
40 7
s 100 200 300 400 500 600 TIME, SECONOS WATERF PO R
GMT CO.
o 8 DEG PD j,',*,',[,'8 5',g'"
3,,
HOT SPOT HEAT TR ANSFER COEFFICIENT 8-13
9 1.4 I.3-1.2-P INITIAL = 1113.3 PSIA 1.1-1-
?
0.9 -
W Q,8-RUPTURE = 41.9 SECONDS C
WO es 0.7-3 3
$f0.6-Wv
[
0.5 -
0.4-0.3-0.2-0.1-0 i
i i
i i
0 20 40 60 80 100 TIME (SEC)
OU 5 NA WATERFORD 3 CYCLE 2 Figure p
g g*
0.8 mM Waterford Steam 88 Electric Station HOT ROD INTERNAL GAS PRESSURE 8-14
!.'W.
2 10.0' TECHNICAL SPECIFICATIONS This section provides a sumary of the proposed changes to the
_N; Technical Specifications for WSES-3 for the Cycle 2 Reload. The changes are arranged in numerical order.
Detailed change pages
v for the Technical Specifications are provided in separate submittals.
. i, i
r s
i
)
4 f
10-1
Tech Spec Title Nature of Change Section 2.1.1.1 DNBR Increase DNBR Safety Limit to include allowances for hot channel factors and rod bow penalty per method of Statistical Combination of Uncertainties.
2.2.1 Reactor Trip Increase DNBR trip setpoint consistent with Setpoints 2.1.1.1 change. Delete BERR1 value from Tech Spec consistent with previous deletion of addressable constants.
3.1.1.1 Shutdown Change title and content to provide Margin-T(avg) shutdown margin both above and below Greater 200*F when any CEA is at least partially than 200'F withdrawn, consistent with program for temperature dependent shutdown margin.
3.1.1.2 Shutdown Margin-Change title and content to provide T (avg) less than shutdown margin requirements when all or equal to 200'F CEA's are fully inserted. New figure gives temperature dependent shutdown margin in Modes 2 through 5, consistent with 3.1.1.1.
3.1.1.3 Moderator Increase allowable range for MTC to Temperature accommodate higher BOC horon concen-Coefficient tration and E0C burnup in Cycle 2.
3.1.2.1 Flow Paths-Delete requirement for heat tracing and Shutdown surveillance on boron flow paths, consistent with 3.1.2.7 changes.
10-2
Tech Spec Ti tle Nature of Change Section
(
3.1.2.2 Flow Paths-Revise boric acid flow paths and Figure 4x Operating 3.1-1 for required BAMT volume vs concentration and delete heat tracing surveillance consistent with 3.1.2.7 and 3.1.2.1 changes. Remove value for shutdown margin in action statement consistent with 3.1.1.1 and 3.1.1.2 changes for temperature dependent shutdown margin.
3.1.2.4 Charging Pumps-Remove value for shutdown margin in action Operation statement consistent with 3.1.1.1 and 3.1.1.2 changes for temperature dependent shutdown margin.
3.1.2.6 Boric Acid -
Same as for 3.1.2.4.
Makeup Pumps-Operating 3.1.2.7 Borated Water Reduce required boron concentration and Sources-delete surveillance of heat' tracing on Shutdown Boric Acid Makeup Tanks (BAMT) consistent with program for reduction below the boron solubility limit.
3.1.2.8 Borated Water Revise requirements for boric acid makeup Sources-Oper-tanks, revise concentration vs volume in ating Figure 3.1-1 and delete surveillance requirement on heat tracing consistent with 3.1.2.7, 3.1.2.2, and program for reduction of boron concentration below the solubility limit.
10-3
Tech Spec Title Nature of Change Section 3.1.2.9 Boron Dilution Revise Table 3.1-1 on monitoring frequency with I or 2 boron dilution alarms in-operable, consistent with 3.1.1.1 and 3.1.1.2 changes for temperature dependent shutdown margin.
3.1.3.1 CEA Position Add new figure to impose power reduction between 15 minutes and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after CEA misalignment consistent with the removal of the CPC penalty factor for inward CEA deviation, per CPC Improvement Program.
3.1.3.4 CEA Drop Time Redefine the surveillance interval for CEA rod drop timing tests from 18 months to every refueling outage.
3.1.3.6 Regulating Revise PDIL in Figure 3.1-2 to account for CEA Insertion Cycle 2 core characteristics.
Limits 3.1.3.7 Part Length Add new PDIL with more restrictive CEA Insertion insertion for part length CEA's to improve Limits rod drop consequences and to clarify rod position vs allowable duration.
3.2.1 Linear Heat Revise operating limit on PLHR in Figure Rate 3.2-la when COLSS is out of service and make it applicable when 0,1 or 2 CEACs are operable. Delete part 3.2.1.c for auto-matic monitoring by CPC when no CEACs are operable.
10-4
Tech Spec Title Nature of Change Section 3.2.4 DNBR Margin Revise COLSS penalty when neither CEAC is operable. Revise CPC monitoring limit on DNBR in Figures 3.2-2 and 3.2-3 when COLSS is out-of-service. Delete surveillance requirement on COLSS & CPC for rod bow penalty consistent with changes to 2.1.1.1 and 2.2.1.
3.2.7 Axial Shape Reduce the positive limit on ASI to reduce Index associated power distribution uncertainty in COLSS and CPC thereby increasing available thermal margin.
3.3.1 RPS/ESFAS Revise to allow bypassing of the non-safety Instrumentation related high steam generator level trip.
3.3.3.6 Accident Add reactor vessel level monitoring Monitoring instrument and revise format for associated Instrumentation action in Table 3.3-10, 3.3.3.8 Fire Detection Change Table 3.3-11 to list smoke detectors Instrumentation in the Control Room.
3.10.1 Shutdown Increase time period available for testing Margin while shutdown margin is less than require-ment in 3.1.1.1.
3.10.2 Moderator Add reference to 3.1.3.7 to allow Temperature suspension of limits on part length Coefficient, CEA insertion during physics testing.
Group Height, Insertion and Power Distribu-tion Limits 10-5
Tech Spec Title Nature of Change Section 3.10.3 Special Test CPC permissive setpoint change Exception, Reactor Coolant Loops 5.3.1 Fuel Increase maximum limit on fuel enrichment Assemblies to acconnodate longer cycles.
5.6.1 Fuel Storage Update to reflect the new fuel storage Criticality criticality analysis performed for Cycle 2.
10-6
12.0 References 12.1 Section 1.0 References (1-1)
"Waterford Steam Electric Station, Unit No. 3 Final Safety Analysis Report," Docket No. 50-382.
12.2 Section 2.0 References None 12.3 Section 3.0 References None 12.4 Section 4.0 References (4-1)
"Waterford Steam Electric Station, Unit No. 3, Final Safety Analysis Report," Docket No. 50-382.
(4-2)
(Deleted)
(4-3)
C. O. Thomas (NRC), to A. E. Scherer (C-E), " Acceptance for Referencing of Licensing Special Report LD-84-0a3, CEA Guide Tibe Wear Sleeve Modification," September 7, 1984.
(4-4)
(Deleted)
(4-5)
EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding Volume 5:
Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," April, 1985.
12-1
(4-6)
(Deleted)
(4-7)
CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report,"
July, 1974.
(4-8)
CEN-161(B)-P, " Improvements to Fuel Evaluation Model," July, 1981.
(4-9)
R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), " Safety Evalation of CEN-161 (FATES 3)," March 31,1983.
(4-10)
CEN-309(A)-P " Arkansas Nuclear One Unit 2 Cycle 5 Shoulder Gap Evaluation," July, 1985.
(4-11)
CENPD-269-P, Revision 1-P, " Extended Burnup Operation of Combustion Engineering PWR Fuel," July,1984.
(4-12)
CEN-335(C)-P, "Waterford Unit 3 Cycle 2 Shoulder Gap Evalua-tion," July, 1986.
12.5 Section 5.0 References (5-1)
EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Volume 5:
Evaluation of Interpellet Gap Formation and Clad Collapse in ModernPWR Fuel Rods," April, 1985.
(5-2)
(Deleted)
(5-3)
(Deleted)
(5-4)
CENPD-153-P, Rev.1-P-A, " INCA /CECOR Power Pesking Uncert-ainty," May, 1980.
(5-5)
CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," April,1983.
12-2
12.6 Section 6.0 References (6-1)
CENPD-161-P-A, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April,1986.
(6-2)
CENPD-162-P-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," September,1976.
(6-3)
CENDP-206-P-A,
" TORC
- Code, Verification and Simplified Modeling Methods," June,1981.
(6-4)
CEN-160(S)-P, Rev. 1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3," September, 1981.
(6-5)
CEN-283(S)-P,
" Statistical Combination of Uncertainties, Part 1 Combination of System Parameter Uncertainties in Thermal Margin Analyses for San Onof re Nuclear Generating Station Units 2 and 3," June, 1984 (6-6)
CEN-155-(S)-P, "CE-1 Applicability to San Onofre Units 2 and 3 HID-2 Grids, Response to NRC Questions," March, 1981.
(6-7)
CEN-165(S)-P, " Responses to NRC Concerns on Applicability of the CE-1 Correlation to the SONGS Fuel Design," May,1981.
(6-8)
NUREG-0787, Supplement 1, " Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3," Docket No. 50-382, October, 1981.
(6-9)
CENPD-225-P-A, " Fuel and Poison Rod Bowing," June 1983.
(6-10)
Robert A.
Clark (NRC) to William Cavanaugh III, (AP&L),
" Operation of ANO-2 During Cycle 2," July 21, 1981 (Safety Evaluation Report and License Amendment No. 26 for ANO-2).
12-3
12.7 Section 7.0 References (7-1)
"Waterford Steam Electric Station Unit No. 3, Final Safety Analysis Report," Louisiana Power and Light Co., Docket No. 50-382.
(7-2)
" Standard Review Plan," NUREG-0800, Rev. 2, 1981.
(7-3)
"CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981, Enclosure 1-P to LO-82-001, January 6, 1982.
(7-4)
Calvert Cliffs Nuclear Power Plant Unit 1 Docket No.
~
50-317. " Amendment to Operating License DPR 53 Supplement 1 to Seventh Cycle License Application," September 1,
1983.
(7-5)
R. V. Macbeth, "An Appraisal of Forced Convection Burnout Data," Proc. Instr. Mech. Engrs., Vol. 180, Pt. 3C, PP 1
37-50, 1965-1966.
(7-6)
D.
H.
Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Water - Part IV, large Diameter Tubes at about 1600 psia," A.E.E.W. Report R479, 19866.
(7-7)
CEN-191(B)-P "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs 1 and 2," December 1981.
(7-8)
CEN-308-P Revision 00-P, "CPC/CEAC Software Modifications for the CPC Improvement Program," July 1985.
(7-9)
CENPD-188-A, "HERMITE Space-Time Kinetics," July 1975.
(7-10)
CENPD-161-P, " TORC Code - A Computer Code for Determining the Thennal Margin of a Reactor Core," July 1975.
12-4
(7-11)
CENPD-206-P,
" TORC Code Verification and Simplified Mr deling Methods," January 1977.
(7-12)
CENPD-183, " Loss of Flow - C-E Methods for Loss of Flow Analysis," July 1975.
(7-13)
CENPD-199-P, Rev.
1-P, "CE Setpoint Methodology," March 1982.
12.8 Section 8.0 References (8-1)
Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3, Friday, January 4,1974.
(8-2)
K. W. Cook, LP&L, to G. W. Knighton, NRC, " Cycle 2 LBLOCA Analysis Methods", W3P86-1578, July 1, 1986.
(8-3)
CENPD-132-P, " Calculative Methods for the C-E Large Break LOCA Evaluation Model", August 1974.
CENPD-132, Supplement 1 " Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model", December 1974 (Proprietary).
(8-4)
CENPD-132-P, Supplement 2P, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", July 1975.
(8-5)
CENPD-133, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis", April 1974 (Proprietary).
CENPD-133, Supplement 2, "CEFLASH-4A, a FORTRAN IV Digital i
Computer Program for Reactor Blowdown Analysis (Modification)", December 1974 (Proprietary).
i 12-5
(8-6)
CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", April 1974 (Proprietary).
CENPD-134, Supplement 1,
"COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)",
December 1974 (Proprietary).
(8-7)
CENPD-135-P, "STRIKIN, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1974.
CENPD-135, Supplement 2 P,
"STRIKIN-II. A Cylindrica'l Geometry Fuel Rod Heat Transfer Program (modification)",
February 1975.
CENPD-135-P, Supplement 4P. "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1976.
(8-8)
CENPD-138, and Supplement 1, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", February 1975.
CENPD-138 Supplement 2 (P), " PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", January 1977 (Proprietary).
(8-9)
Topical Report, Enclosure 1-P to LD-81-095, CE ECCS Evaluation Model Flow Blockage Analysis, December 1981.
(8-10)
D. A. Powers and R. D. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis", NRC Report, NUREG-0630, April 1980.
(8-11)
LD-85-050, Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC), Enclosure, " Supplemental Material for Inclusion in CENPD-132, Supplement 3-P", November 5, 1985.
l 12-6
,. - -. -, -., -. _,. _., - - -,., -,.. - - - -. _. _. - - - - -, - - -. ~,.
12.9.
Section 9.0 References (9-1)
CEN-304-P, Rev. 01-P, " Functional Requirements for a Control Element Assembly Calculator," May,1986.
(9-2)
CEN-305-P, Rev. 01-P, "Fuentional Requirements for a Core Protection Calculator," May,1986.
(9-3)
CEN-308-P-A, "CPC/CEAC Software Modifications for the CPC Improvement Program " April,1986.
(9-4)
CEN-310-P-A, "CPC and Methodology Changes for the CPC Improvement Program," April,1986.
(9-5)
CEN-3330-P, Rev. 00-P, "CPC/CEAC Software Modifications for the CPC Improv. ment Program Relaod Data Block," May, 1986.
(9-6)
CEN-39(A)-P, Rev. 03, "CPC Protection Algorithm Software Change Procedure," January, 1986.
(9-7)
CEN-39( A)-P, Supplement 1-P, Rev. 03-P, "CPC Protection Algorithm Software Change Procedure Supplement 1,"
- April, 1986.
(9-8)
CEN-323-P, " Reload Data Block Constant Installation Gui,deline," February, 1986.
(9-9)
CEN-281(S)-P, "CPC/CEAC Software Modifications for San Onofre Nuclear Steam Generating Station Units No. 2 and 3,"
July, 1984.
i i
12-7
12.10 Section 16.0 References None 12.11 Section 11.0 References None 1
4 12-8
/
.