ML17252B047

NEDO-24074 General Electric Boiling Water Reactor Reload-5 Licensing Submittal for Dresden Nuclear Power Station Unit 3.

ML17252B047
Person / Time
Site:	Dresden
Issue date:	11/30/1977
From:	General Electric Co
To:	Office of Nuclear Reactor Regulation
References
77NED353 NEDO-24074
Download: ML17252B047 (43)

Text

NED0-24074 77NED353 CLASS I NOVEMBER 1977 GENERAL ELECTRIC BOILING WATER REACTOR RELOAD-5 LICENSING SUBMITTAL FOR DRESDEN NUCLEAR POWER STATION UNIT 3 NOTICE THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERI.OD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND DOCUMENTS

,- CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) FROM DOCUMENr FOR REPRODUCTION MUST YGM.Athl!VE~~~;;\,, F. COP'/

DEADLINE RETURN DATE RECORDS FACILITY BRANCH GENERAL. ELECTRIC

NED0-24074 77NED353 Class I November 1977 GENERAL ELECTRIC BOILING WATER REACTOR RELOAD-5 LICENSING SUBMITTAL FOR DRESDEN NUCLEAR POWER STATION UNIT 3 BOILING WATER REACTOR PROJECTS DEPARTMENT* GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENERAL. ELECTRIC

\

NE00-24074 IMPOHIANT NOTICE REGARDING CONTENTS OF THIS REPORT.

Please Read Carefully The only undertakings of General Eleatria Company respeating infor'111ation in this doawnent are contained in the contraat betuJeen Corrunonwealth Edison Co.

and General Eleatric Company, and nothing aontained in this doaument shall be aon.strued as ahanging the aontraat. The use of this information by any-one other than CorrmO'YlLJealth Edison Co., for any purpose other> than that for whiah it is intended, is not -authorized, and with respect to any unauthorized use, General Electria Company makes no representation or warranty, and assumes no liability as. to the aompleteness, aaauracy, or usefulness, of the information aontained in this doawnent .

ii

NED0-24074 TABLE OF CONTENTS

1. INTRODUCTION 1-1 2 *.

SUMMARY

2-1

3. :MECHANICAL DESIGN 3-1

4. THERMAL-HYDRAULIC.ANALYSES 4-1 4.1 Statistical Analysis 4-1 4 .1.1 Fuel Cladding Integrity Safety Limit 4-1 4.2 Analysis of Abnormal Operational Transients 4-1 4.2.~ Operating Limit MCPR 4-1

5. NUCLEAR CHARACTERISTICS 5-1 5.1 Nuclear Characteristics of the Core 5-1 5.1.l Core Effective Multiplication, Control System Worth and Reactivity Coefficients 5-1 5.1.2 Reactor Shutdown Margin 5-1 5.1.3 Standby Liquid Control System 5-2

6. SAFETY ANALYSIS 6-1 6.1 Introduction 6-1 6.2 Model Applicability to 8x8 Fuel 6-1

6. 3 Results of Safety Analyses 6-1.

6.3.1 Core Safety Analyses 6-1 6.3.2 Accident Analyses 6-1 6.3.3 Abnormal Operating Transients 6-8 6.3.4 ASME Vessel Pressure Code Compliance 6-18 6.3.5 Thermal-Hydraulic Stability Analysis 6-24

7. REFERENCES 7-1 iii/iv

NED0-24074 LIST OF ILLUSTRATIONS Figure Title 2-1 Dresden 3 NPS Reload-5 Design Reference Core Loading 2...,.2 6-1 Doppler Coefficient Versus Average Fuel Temperature as a 6-3 Function of Moderator Condition 6-2 Accident Reactivity Shape Function Cold Startup, S = 0.005, PL = 1.5 6-4 6-3 Accident Reactivity Shape Function Hot Startup, S = 0.005, PL = 1.3 6-5 6-4 Scram Reactivity Function for 20°C (RDA) 6-6 6-5 Scram Reactivity Function for 286°C (RDA) 6-7 6-6 Dresden 3 Cycle 6 Scram Reactivity and Control Rod Drive Specification (Transients) 6-10 6-7a Generator Load Rejection Without Bypass -2000 MWd/t Before EOC-6 (100% Power) 6-13 6-7b Generator Load Rejection Without Bypass - Cycle 6 Limiting Case (98% Power) 6-14 6-8a Turbine Trip Without Bypass -2000 M~d/t Before EOC-6 (100% Power) 6-16 6-8b Turbine Trip Without Bypass - Cycle 6 Limiting Case (98% Power) 6-17 6-9 Loss of 145°F Feedwater Heating 6-19 6-lOa

• MSIV Closure .Flux Scram -2000 MWd/t Before EOC-6 6-22 6-lOb MSIV Closure Flux Scram - Cycle 6 Limiting Case 6-23 6-11 Decay Ratio 6-25 v/vi

NED0-24074 LIST OF TABLES Table Title 2-1 Fuel Type and Number 2-1 4-1 Sunnnary of Results - Limiting Transients 4-2 4-2 GETAB Transient Analysis - Initial Condition Parameters 5-1 Nuclear Characteristics of the Design Reference Core 5-3 6-1 Transient Input Parameters 6..,.9 6-2 Dresden 3 Cycle 6 Transient Data Summary 6-12 6-3 RWE and RBM Analysis 6-20 6-4 Rod Withdrawal Error Summary 6-20 vii/viii

NE00-24074

1. INTRODUCTION This document provides the supplemental information for Reload-5 at the Dresden Nuclear Power Station Unit 3. The technical bases, generic design information, and safety analyses are given in Reference 1.

The design reference core loading is based on the use of 20 8x8 bundles having a bundle average enrichment of 2 .SO wt % U-235 and 156 8x8 bundles having a bundle average enrichment of 2.62 wt% U-235.

The objective of this outage is to load the reactor core to ensure sufficient reactivity to operate the 724-element core at a licensed power level of 2527 MWt for a nominal 5650 MWd/t cycle.

Analyses in this document and its references Justify satisfaction of the outage objectives.

1-1/1-2

y NED0-24074

2.

SUMMARY

The design reference core configuration for this license submittal con-sists of bundles defined in Table 2-1. The relative location of each fuel bundle type is shown in Figure 2-1.

Table 2-1 FUEL TYPE AND NirM:BER Fuel

~ Number Initial 164 Reload-1 (70230) 52 Reload-2 (8D250) 44 Reload-3 (8D250) 108 (8D262) 32 Reload-4 (8D250) 60 (8D262) . 88 Reload-5 (8D250) 20 (8D262) 156 Total 724 2-1

v NED0-24074 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57"59

@LPRM LOCATION (LETTER INDICATES TIP MACHINE) BLANK - INITIAL FUEL

• LPRM LOCATION (COMMON LOCATION FOR ALL TIP MACHINES) A - *RELOAD 1 (7D230 GENERIC Bl

@tRM LOCATIONS B - RELOAD 2 (8D2501 6, SAM LOCATIONS c RELOAD 3 (8D2501 D - RELOAD 3 (8D2621 E - RELOAD 4 (8D262l F - RELOAD 4 (8D2501 G - RELOAD 5 (8D2501

0::.,,

H - RELOAD 5 (8D2621 Figure 2-1. Dresden 3 NPS Reload-5 Design Reference Core Loading 2-2

NED0-24074

3. MECHANICAL DESIGN The two types of Reload-5 fuel which will be employed have the same mechanical configuration and fuel blindle enrichments as the 8D262 an~ the 8D250 fuel assemblies described in Reference 1. Reload 5 incorporates the improved water rod design described in Section 3.1 of Reference 1. The design criteria, models, and results from design evaluation presented in Section 3 of Reference 1 *

. The design criteria, models, and results from design evaluation presented in Section 3 of Reference 1 are applicable to the subject reload.

All Reload-5 fuel incorporates finger springs of the type described in Reference 1.

3-1/3-2

NED0-24074

4. THERMAL-HYDRAULIC ANALYSES 4.1 STATISTICAL ANALYSIS The statistical analyses of the reactor core were performed using the uncer-tainty inputs described in Section 4.5 of Reference 1. The results of the analyses show that at least 99. 9% of the fuel rods in the core are expected to avoid boiling transition if the MCPR is 1.06 or greater.

4.1.l Fuel Cladding Integrity Safety Limit Based on the results of the statistical analysis, the fuel cladding integrity safety limit is a MCPR of 1.06.

4.2 ANALYSIS OF ABNORMAL OPERATIONAL TRANSIENTS The results of the limiting abnormal operational transient analyses and the Rod Withdrawal Error (RWE) are summarized in Table 4-1; the specific analyses are described in Section 6. The most severe transient from rated conditions for the 7x7 fuel is a RWE which has a maximum 6CPR of 0.23. The most severe transient from rated conditions for the 8x8 fuel is a generator load rejection without bypass which has a maximum 6CPR of 0.23. Addition of the 6CPR to the Safety Limit MCPR gives the minimum initial MCPR to avoid violating the Safety Limit MCPR during the most severe transient from rated conditions. The GETAB analysis initial conditions for the abnormal operational transients are given in Table 4-2.

4.2.1 Operating Limit MCPR Based on the Fuel Cladding Integrity Safety Limit and the results of the transient analyses, the Operating Limit MCPR is 1.29 for 8x8 fuel and 1.29 for 7x7 fuel.

4-1

NE00-24074 Table 4-1

SUMMARY

OF RESULTS LIMITING TRANSIENTS Maximum 6CPR 7x7 8x8 Turbine Trip w/o Bypass (Rated Conditions) 0.16 0.22 Load Rejection w/o Bypass (Rated Conditions) 0.17 0.23 Loss of 145°F Feedwater Heating 0.16 0 ..18 Rod Withdrawal Error (107% RBM Set Point) 0.23 0.19 Table 4-2 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS*

(Abnormal Operating Transients) 7x7 8x8 Peaking Factors (Local, Radial, Axial) 1. 30' 1. 52' 1. 40 1.22, 1. 6 7, 1. 40 R-Factor 1.100 1.094 Bundle Power (MWt) 5.191 5.699 Nonfuel Power Fraction 0 .035 0.035 Core Flow (Mlb/hr) 98 98 Bundle Flow (103 lb/hr) 113.56 109.43 Reactor Pressure (psia) 1030 . 1030 Inlet Enthalpy (Btu/lb) 522.5 522.5 Initial MCPR 1.25 1. 31

• 100% power/100% flow initial conditions (Do not apply to RWE).

4-2

NED0-24074

5. NUCLEAR CHARACTERISTICS The bundle characteristics, analytical methods, and model descriptions pre-sented in Subsections 5.1 through 5.4 of Reference 1 are applicable to this reload. Results of specific reload core calculations are g.iven below.

5.1 NUCLEAR CHARACTERISTICS OF THE CORE This section presents the results of the calculation on:

1. reactivity control characteristics; and

2. core average reactivity coefficients.

The core characteristics were calculated using the design reference loading pattern shown in Figure 2-1. The loading pattern was designed to accommodate 176 Reload-5 fuel bundles by discharging a like number of fuel bundles from*

the Cycle 5 core.

5.1.1 Core Effective Multiplication, Control System Worth and Reactivity

• Coefficients A tabulation of the typical nuclear characteristics of the reconstituted core is given in Table 5-1. The nuclear characteristics of the Reload-5 fuel bundles are identical to those previously loaded. Therefore, the total con-trol system worth and the temperature and void dependent behavior of the reconstituted core will not differ significantly from those values previously reported.

5.1.2 Reactor Shutdown Margin The reconstituted core fully meets the established technical specification criteria in that it may be maintained sub critical by at least 0. 25% ~k in the most reactive condition throughout the subsequent operating cycle with the strongest control rod fully withdrawn and all other rods fully inserted.

5-1

NED0-24074 A minimum shutdown margin of 0.014 6k calculated for the assumed refueling at a core average exposure of 15,134 MWd/t is the most reactive condition throughout the subsequent operating cycle with the strongest control rod fully withdrawn and all other rods fully inserted. The Beginning of Cycle 6 (BOC-6) shutdown margin is 0.014 b~. Thus, R, the differences between the BOC-6 and the minimum shutdown margin plus the effect of B c settling in the absorber 4

tubes is 0.0004 bk.

5.1.3 Standby Liquid Control System A boron concentration of 600 ppm in the moderator water will bring the reactor subcritical by 0.033 bk at 20°c, xenon free.

5-2

NE00.:.24074 Table 5-1 NUCLEAR CHARACTERISTICS OF THE DESIGN. REFERENCE CORE Core Effective Multiplication and Control System Worth (No Voids. 20°c)

BOC keff Uncontrolled 1.119 Fully Controlled 0.954 Strongest Control Rod Out 0.986 R, Maximum Increase in Core Reactivity With Exposure 0.0004 Into Cycle, 6k (including effects of inverted B4C tubes in control rods)

Reactivity Coefficients, Range During Operating Cycle Steam Void Coefficients at Average Voids; -11. 9 x lo-4 to (6k/k)/6V, 1/% Void -10.4 x *10-4 Power Coefficient at Rated Conditions -0 .057 (6k/k)/(6P/P)

Fuel Temperature Coefficient at 650°c -1.15 x 10-5 to (6k/k) /6T, l/OF -1.24 x io-5 5-3/5-4

y NED0-24074

6. SAFETY ANALYSIS

6.1 INTRODUCTION

The safety analysis for reloads consists of three categories: (a) generic safety analysis, which is applicable to all reloads; (b) bounding analysis; and (c) specific analysis applicable only to the current reload. Wherever a bounding analysis is. applied for an accident or transient, the key parameters need only to. be compared with the worst case and, if they are within "bounds,"

all limits and margins applicable to the accidents or transients will be met.

6.2 MJDEL APPLICABILITY TO 8x8 FUEL Information on the applicability to the 8x8 design of existing models used for safety analyses is given in Reference 1.

6.3 RESULTS OF SAFETY :ANALYSES

6. 3 .1 Core Safety Analyses The General Electric Thermal Analysis Basis (GETAB) (Reference 2) is used to establish thermal margins in reload cores. The operating limits, margins, and fuel damage limits previously used are applicable to this reload. Where neces-sary, further discussions of these and other controlling factors are presented below.

6.3.2 Accident Analyses

.6. 3.2.1 Main Steamline B.reak Accident The consequences of the main steamline break analysis depend on the basic thermal-hydraulic parameters of the overall reactor, as discussed in Reference 1. Because these parameters do not normally change as a result of reload, the referenced analysis applies.

6-1

NED0-24074 6.3.2.2 Refueling Accident The description and analyses of the refueling_accident provided in the FSAR and discussed in Reference 1 apply to this reload. The factors involved are such that the conclusions of these evaluations remain valid.

6.3.2.3 Control Rod Drop Accident The technical bases (bounding analyses) which are presented in Reference 1 were*used to verify that the results of a rod drop excursion in the reloaded core would not exceed the design criteria. For a,pplication to Dresden 3 Reload 5,. the actual Dopp],.er coefficient, accident reactivity shape functions

• and scram reactivity functions are compared with the technical bases in .

Figures 6-1 through 6-5. Since the maximum values of the parameters after this reload will be well below the boundary value, the consequences of a rod drop excursion from any insequence control rod would be below the 280 cal/gm design limit. Further, the radiological consequences will be no greater than those evaluated in Reference 1.

6.3.2.4 Loss-of-Coolant Accident .

The loss-of-coolant accident analysis will be submitted under a separate cover on a schedule consistent with NRC requirements.

6.3.2.5 Loading Error Accident 6.3.2.5.1 Event Description A loading error for the reference core configuration is defined as:

- (1) a Reload.-5 bundle is inserted in an improper location; and (2) the error is not discovered in the subsequent core verification and the reactor is operated.

6-2

NED0-24074 5

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-30

-35 0 400 800 1200 1600 2000* 2400 FUEL TEMPERATURE (°C)

Figure 6~1. Doppler Coefficient Versus Average Fuel Temperature as a Function of Moderator Condition 6-3

NED0-24074 24 20 0 BOUNDING VALUE FOR 280 cal/g 0 CALCULATED VALUE iii

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Figure 6-2. Accident Reactivity Shape Function Cold Startup, S = 0.005, PL = 1.5 6-4

NED0-24074 24 0 BOUNDING VALU.E FOR 280 cal/g 0 CALCULATED VALUE u;

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NED0-24074 BO 0 BOUNDING VALUE FOR 280 cal/g 70 0 CALCULATED VALUE 60 iii

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6-6

NED0-24074 120 0 BOUNDING VALUE FOR 280cal/g Q CALCULATED VALUE 100 u;

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- NED0-24074 Since two independent errors are assumed to occur, the single-error criterion is violated; thus, the event is not classified as an abnormal operational transient. The following are the results and consequences for a worst-case error.

6.3.2.5.2 Results and Consequences Analysis of the loading error accident results in a peak linear heat genera-tion rate (LHGR) of 18.87 kW/ft and a minimum critical power ratio (MCPR) of 0.95* in the misplaced reload (8D262) bundle. The peak LHGR is not large enough to cause fuel damage. Since only a single assembly has an MCPR lower than the safety limit MCPR, the number of rods in the core expected to experience boiling transition is small (<0.01%). Thus, the results of th.is accident are ~ar less severe t.han the major accidents.

Fuel bundles adjacent to the misplaced bundle are insignificantly affected by the presence of the misplaced bundle.

6.3.3 Abnormal Operating Transients 6.3.3.1 Transients and Core Dynamics

• 6.3.3.1.1 Analysis Basis This subsection contains the analyses of the most limiting abnormal operational transients for Dresden 3 Cycle 6. All transients which are the*basis of the existing license were reviewed, and those transients which have been limiting in the past with respe~t to safety margins and are significantly sensitive to the core transient parameter deviations were reanalyzed.

6.3.3.1.2 Input Data and Operating Conditions The input data and operating conditions are shown in T'able 6-1 and represent the nominal basis for these analyses. Each transient is considered at these conditions unless otherwise specified.

• From an initial MCPR of 1.25.

6-8

NED0-24074 Table 6-1 TRANSIENT INPUT PARAMETERS Thermal Power (MWt) 2527 Rated Steam Flow (lb/hr) 9.76 x 10 6 Rated Core Flow (lb/hr) 98.0 x 10 6 Dome Pressure (psig) 1005 Turbine Pressure (psig) 935 RV Setpoint(l) (3)

(psig) 1@1136, 2@1141. 2@1146 RV/Capacity (at Setpoint) (No./%) 5/29.2 RV Time Delay (msec) 650/400(3)

RV Stroke Time (msec) 2001100< 3 )

SV Setpoint(l) (psig) 2@1252, 2@1262, 4@1272 SV Capacity (No./%) 8/52.5 Void Coefficient( 2 ) (S:/%Rg) -9.552, -8.637 Void Fraction( 2 ) (%) 36.04, 33.92 Doppler Coefficient( 2) (C/°F) -0.208, -0.213 Average Fuel Temperature (oF) 1203 Scram Reactivity Curve Figure 6-6 Scram Worth (2) ($) (%) -29.2, -30.1 (l)Includes 1%

(i)2000 MWd/t before EOC-6 and Cycle 6 limiting case respectively C3 )Target Rock combination safety/relief valve 6-9

NE00-24074 100 45 90 40 80 35 70 30 j::

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Figure 6-6. Dresden 3 Cycle 6 Scram Reactivity and Control Rod Drive Specification (Transients) 6-10

NED0-24074

6. 3. 3 .1. 3 Transient Summary A summary of the transients analyzed and their consequences is provided in Table 6-2.

6.3.3.2 Transient Descriptions The abnormal operating transients which are limiting according to saf_ety criteria and which als.o are sensitive to nuclear core parameter changes have been analyzed and are evaluated in the following narrative.

6.3.3.2.1 Generator Load Rejection With Failure of the Bypass Valves The primary characteristic of this transient is a pressure increase due to the obstruction of steam flow by the turbine control valves. The pressure increase causes a significant void reduction, which yields a pronounced positive void reactivity effect. The net reactivity is .sharply positive and causes a rapid increase in neutron flux until the net reactivity is forced negative by scram initiated from pressure switches sensing control valve fast closure and by a void increase after the relief valves have automatically opened on high pressu_re *. Figure 6-7a and b illustrate this transient for the EOC6-2000 MWq/t and the limiting Cycle 6 cases.*

The parameters of concern are the peak steamline pressure margin to. the first spring safety valve ~etpoint and the peak average surface heat flux correlated to MCPR.

For the limiting Cycle 6 case, the neutron flux (the precursor of heat flux)

.. rises to a peak of 259. 2% with a corresponding peak heat flux of 111. 0%. The change in criticai power ratio (ti.CPR) is presented in Table .4-1.

The peak steamline pressure for the worst Cycle 6 case is limited to 1215 psig as a result of the high-pressure actuation of the four electromatic valves and one combination of safety/relief valve,* which provides a 25 psi margin to the 1240 psig setpoint of the first spring safety valve.

6-11

NED0-24074 Table 6-2 DRESDEN 3 CYCLE 6 TRANSIENT DATA

SUMMARY

Core Power Flow

=::s RISE 1rsn 300 1 1------.---,----i-----+:73~Si:;"ET! >U: F'..:'rl 17.l(TR)

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Figure 6-7a. Generator Load Rejection Without Bypass -2000 MWd/t Before EOC-6 (100% Power) 1 NE'JTF.:N ~:.. '.r.< 1 VESS~L F~::s Rj:E rPSll f 2 FEq~( FL*:'..! '.:E'I":'::~ iEMf> 2 STH liP;E *?RES ~I~E !?SU 150. 13 Av: ~~=~~::: ~::~T ~LUX 300. 3 s::;"ETT * .:i_.,*:. F:_~..; ~::1(TR) lj t:.t*_*.. ;;1:.i ":..:~ 0

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o. I .. "* 6. TIHE ISECI 12 *. 16. Figure 6-7b. Generator Load Rejection Without Bypass - Cycle 6 Limiting Case (98% Power) NED0-24074 6.3.3.2.2 Turbine Trip With Failure of the Bypass Valves The primary characteristic of this transient is a pressure increase due to the obstruction of steam flow by the turbine stop valves. The pressure increase causes a significant void reduction, which yields a pronounced positive void reactivity effect. The net reactivity is sharply positive and causes a rapid increase in neutron flux until the.net reactivity is forced negative by scram initiated from 90% open switches on the turbine stop valves and by a void increase after the relief valves have automatically. opened on high pressure. Figure 6-8a and 8b illustrate this transient for the EOC 6 -2000 MWd/t and the limiting Cycle 6 cases. The parameters of concern are the peak steamline pressure margin to the first spring safety valve setpoint and the peak average surface heat flux correlated to MCPR. For the limiting Cycle 6 case, the neutron flux (the precursor of heat flux) rises to a peak of 243.3% with a corresponding peak heat flux*of 109.3%. The change in critical power ratio (llCPR) is presented in Table 4-1. The peak steamline pressure for the worst Cycle 6 case is limited to 1214 psig as a result of the high~pressure actuation of the four electromatic valves and one combination of safety/relief valve, which provides a 26 psi margin to the 1240 psig setpo~nt of the first spring safety valve. 6.3.3.2.3 Loss of Feedwater Heating The loss of feedwater heating is analyzed in FSAR's and other submittals because it constitutes the most limiting cool water transient. Feedwater heating can be lost if the steam extraction line to the heater is shut and the heat supply to the heater is removed, producing a gradual cooling of the tubes. The reactor will receive cooler feedwater flow which will pro-duce an increase in core inlet subcooling and, due to the negative void reac-tivity coefficient, an increase in core power. The delay in the flow from the tripped feedwater heater to the feedwater sparger is ignored, thereby adding conservatism to the analysis. 6-15

8. 12. 16.

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TIME ISECI r1gure 6-8b. Turbine Trip Without Bypass - Cycle 6 Limiting Case (98% Power) NED0-24074 Figure 6-9 shows the response of the plant to the loss of 145°F of the feed-water heating capability of the plant. This represents the maximum expected single heater (or group of heaters) to be tripped or bypassed by a single event. The reactor is assumed to be.at maximum.power conditions* on manual flow control when the heating capability was lost. Note that in manual flow control mode the core flow remains essentially constant through.out the transient. Neutron flux increases above the initial value, however, in order to produce the same steam flow with the higher inlet subcooling . . The reactor settles out with a neutron flux 120. 8% of initi.al power and fuel average surface heat flux peaks at 119.5% of its initial value. The change in critical power ratio (~CPR) is presented in Table 4-1. 6.3.3.2.4 Plant Operation The operating plan for Dresden 3, Cycle 6 is to start up and operate out to 2000 MWd/t before EOC-6 at 100% power, then reduce power to 98% by coasting down and operating at this power level out to EOC-6. Previous analyses of all rods out coastdown at EOC have consistently shown that operating limits as determined*by pressure transients are not exceeded. These.analyses are applicable to Dresden 3,for Cycle 6 operation. 6.3.3.2.5 Rod Withdrawal Error

• Assumptions and descriptions of rod withdrawal error are given in Reference 1.

Table 6-3 shows the results of.the worst case condition for Dresden 3 ( Reload 5. The rod block monitor (RBM) setpoint of 107% is selected to allow for failed instruments for the worst allowable situation. This case demon-strates that even if the operator ignores all alarms during the course of this transient, the critical power ratio (CPR) does not go below the 1.06 MCPR safety limit. 6.3.4 ASME Vessel Pressure Code Compliance All Main Steam Line Isolation Valve Closure-Flux Scram (Safety Valve Adequacy) The pressure relief_ system must prevent excessive overpressurization of the primary system process barrier and t~e pressure vessel to preclude an uncon-trolled release of fission products. 6-18 1 tflJTRCN fftux 1 VESSa. ~S RISE lPSI J 2 PE~ FL!"~ cmrrn ~ 2 STH Lit~ ~S RISE !PSI> 3 A'.'E S!.:~ta: ~RT FLUX 3 ~ *,*:t_',~ fL'.:;1 Cl.) ISO *-----+-----+------t-r.-.ac~r:iiEflCC~.-- 350.i-----...---------rsyr~~Ett:::;rtkJ s VESSEL SlERH FLCW S ffl.IEF VR..',E FLCi-4 CU 6 TLmII£ P,RE5 FHSE lPSH

100. 21D.i-----...-----------.----+----~

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20. ~. 60. tn.

TIHE lSECJ Figure 6~9. Loss of 145°F Feedwater Heating NED0-24074 Table 6-3 RWE AND RBM ANALYSIS (WITH INSTRUMENT FAILURE) Rod Block -~CPR/ ~CPR* Rod Position Seq~oint 7x7 8x8 Ft Withdrawn 1.04 1.2829/0.1553 1.1690 Io .1110 4.* 00 1.05 1. 2610/0.1772 1.1550/0.1250 4.50 1.06 1. 2410/0.1972 1.1420/0.1380 5.00 1.07 1. 2098/ 0. 2.284 1.1042/0.1758 6.00 1.08 1.1480/0.2902 1. 0130/0. 26 70 9.00 1.09 1.1329/0. 3053 0.9764/0.3036 10.00 1.10 1.1240/0.3142 0.9520/0.3280 11.00

• Based on an estimated MCPR of: 1.4382 (7x7)

1. 2800 (8x8)

Table 6-4 ROD WITHDRAWAL ERROR

SUMMARY

Rod.Position MLHGR. Kw/ft MCPR Ft Withdrawn 8x8 7x7 8x8 7x7 0 -13. 40 14.54 1.280 1.438 2 13.43 14.58 1.236 1.376 4 14.95 15.00 1.169 1.283 6 14.50 14.85 1.104 1.210 8 13.68 14.61 1.060 1.167 10 15.78 16.65 0.976 1.133 12 17 .31. 18.46 0.930 1.120 6-20

NED0-24074 The Dresden 3 pressure relief system includes 4 electromatic relief valves, one Targe~CRock dual~purpose safety/relief valve and 8 spring safety valves located on the main steamlines within the drywell between the reactor vessel and the first isolation valve. These valves provide the capacity to limit nuclear system overpressurization.

The ASME Boiler and Pr.essure Vessel Code requires that each vessel designed to meet Section III be protected from the consequences of pressure in excess of the vessel design pressure:

(a) A peak allowable pressure of 110% of the vessel design pressure is allowed (1375 psig for a vessel with a design pressure of 1250 psig).

(b) The lowest qualified safety/relief valve setpoint must be at or below vessel design pressure.

(c) The highest safety valve setpoint must not be greater then 105% of vessel design pressure (1313 psig for a 1250 psig vessel).

Dresden 3 safety/relief and spring safety valves are set to self-actuate at the pressures shown in Table 6-1, thereby satisfying (b) and (c), above."

Requirement (a) is evaluated by considering the most severe isolation event with indirect scram. The relief valves are assumed to be inactive.

The event which satisfies this specification is the closure of all main steam-line isolation valves with indirect (flux) scram *. The initial conditicins assumed are those specified in Table 6-1. Figures 6-lOa and b graphically illustrate the event for the two cases. An abrupt pressure and power rise o.ccurs as soon as the reactor is isolated. For the worst case, the safety valves open to limit the pressure rise in the steamline at the valves to 1277 psig and at the bottom of the vessel to 1311 psig. This response provides a 64 psi margin to the vessel code limit to 1375 psig. Thus, requirement (a) is satisfied and adequate overpressure protection is provided by the pressure relief system.

6-21

I~ ffLUX 2 PEA!< f'~L WlTFR TEW ISO. 3 RVE s;.::i::- :a t'EAT FLUX tr"C~~:E~11T~--

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Figure 6-lOa. MSIV Closure Flux Scram -2000 MWd/t Before EOC-6

1 VESSEL f'~:s A!:~ !PS! I 2 srn ui.:: ::=:::s =:sf *!"Sil 300:...._----+------'li------+=3-i~:::::E~ T *;:i_ *;::: ;::_::,; tZI ll fl:::llff ~i .* t:: ;:_~,, !'i:l(TR) 5 BTP~SS V~-~::: FL:~ IZI 6 UfiS i hE Pi'.~5 Ai X: IPSI J I

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TIHE lSECI Figure 6-lOb. MSIV Closure :Flux Scram - Cycle 6 Limiting Case

NED0-24074 6.3.5 Thermal-Hydraulic Stability Analysis Descriptions of the types of thermal-hydraulic stability considered and the analytical method used for evaluation are given in Reference 1. The results for Dresden 3 Reload-5 are given below.

6.3.5.1 Channel Hydrodynamic Conformance to the Ultimate Performance Criteria The channel performance calculation yields decay ratios as presented below:

100% Rod Line - (

Channel Hydrodynamic Performance Natural Circulation Power \

• Decay Ratio, x2;x0 8x8 Channel o. 09*

7x7 Channel 0.01 At this most responsive condition, the most responsive channels are clearly within the bounds of the ultimate performance criteria of < 1.0 decay ratio.

6.3.5.2 Reactor Conformance to Ultimate Performance Criteria The decay ratios determined from the limiting reactor core stability conditions are presented in Figure 6-11. The most responsive case is again 100% rod line - nattiral circulation condition.

100% Rod Line -

Reactor Core Stability Natural Circulation Power Decay Ratio, x2;x0 0.48 These calculations show the reactor to be in compliance with the ultimate per-formance criteria, including the most responsive condition at 100% rod line - natural circulation power.

6-24

NED0-24074 1.4 . - - - - - - - - - - - - - - - - - - - - - - . . ; . __ _ _ _ _ _ _ _ _ _ _ _ _ __

1.2 1.0 0 0.8 x

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NATURAL CIRCULATION LINE 0.4 0.2 00"'-----------"-----------"'---------....1----------"""'"i....----------'

20 40 60 80 100 PERCENT POWER Figure 6-11. Decay Ratio 6-25

NED0-24074 6.3.5.3 Channel Hydrodynamic Conformance to the Operational Design Guide Channel Hydrodynamic Rated Low End of Performance Conditions Flow Control Range Decay Ratio, .xz1x0 8x8 <0. en 0.02 7x7 ~0.01 0.01 The most responsive channel is in.conformance with the operational design guide of < 0.5 decay ratio.

6.3.5.4 Reactor Core Conformance to Operational Design Guide The calculated value of the decay ratio of the reactor power dynamic response for rated operating conditions and for the low end of the flow control range (55% power, 39% flow) are presented below.

Reactor Core Rated Low End of Performance Conditions Flow Control Range Decay Ratio <0,02 0.25 As noted earlier, Figure 6-11 describes the.variation of decay ratio over the entire power flow range.

6-26

NED0-24074

7. REFERENCES

1. GE/BWR Generic Reload Licensing Application for 8x8 Fuel, Rev 1 Supplement 4, April 1976 (NED0-20360).

2. General Electric Thermal Analysis Basis (GETAB): Data Correlation and Design Application, General Electric Company BWR Systems Department, November 1973 (NEDE-10958, Class III).

7-1/7-2

\r - '~

?

GENERAL fj ELECTRIC