ML17191A300

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Final Safety Analysis Report
ML17191A300
Person / Time
Site: Dresden  Constellation icon.png
Issue date: 07/10/2017
From:
Commonwealth Edison Co
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8207270119
Download: ML17191A300 (42)
v • d • e


Text

li TABLE OF CONTENTS SECTION 1 -- INTRODUCTION AND

SUMMARY

Page 1.1 PURPOSE AND ORGANIZATION OF REPORT 1.1.1-1 1.1. l PURPOSE OF REPORT 1.1.1-1 1.1.1.1 Introduction 1.1.1-1 1.1.

1.2 Purpose and Scope

of Safety Analysis Report 1.1.1-1 1.1.1.3 Updating of Original FSAR 1.1.1-2 1.1.2 ORGANIZATION OF REPORT 1.1.2-1 1.1.2.1 General Format 1.1.2-1 1.1.2.2 Revisions 1.1.2-1

. 1.2 PLANT DESCRIPTION 1.2.1-1

1. 2 .1 PRINCIPAL DESIGN CRITERIA 1.2.1-1 1.2.1.1 Reactor Core 1.2.1-1 1.2.1.2 Reactor Core Cooling Systems 1.2.1-2 1.2.1.3 Containment 1.2.1-2 1.2.1.4 Control and Instrumentation 1.2.1-3 1.2.1.5 Electrical Power 1.2.1-3 1.2.1.6 Radioactive W~ste Disposal 1.2.1-3 1.2.1.7 Shielding and Access Control 1.2.1-3 1.2.1.8 Fuel Handling and Storage 1.2.1-4 e 1.2 .2 1.2.2.1

SUMMARY

DESIGN DESCRIPTION AND SAFETY ANALYSIS Design Bases Dependent On Site Characteristics 1.2.2-1 1.2.2-1 1.2.2.2 Station Arrangements 1.2.2-3 1.2.2.3 Reactor Sys terns 1.2.2-3 1.2.2.4 Containment Systems 1.2.2-4 1.2.2_.5 Shutdown Cooling System and ECCS 1.2.2-7 1.2.2.6 Unit Control and Instrumentation 1.2.2-8 1.2.2.7 Radiation Monitoring Systems 1.2.2-9 1.2.2.8 Fuel Handling and Storage 1.2.2-9 1.2.2.9 Turbine System 1.2.2-10

1. 2. 2 .10 Electrical System 1.2.2-10 1.2.2.11 Shielding, Access Control, and Radiation Protection Procedtires 1.2.2-10 1.2.2.12 Radioactive Waste Control 1.2.2-11
1. 2. 2 .13 Summary Evaluation of Safety 1.2.2-11
1. 2.3

SUMMARY

OF TECHNICAL DATA 1. 2. 3-1 1.2 .4 INTERACTION OF UNITS 1, 2, & 3 1. 2. 4-1 1.2.4.1 Gaseous Waste Effluents 1. 2. 4-1 1.2.4.2 Liquid Wa~te Effluents 1. 2. 4-1 1.2.4.3 Unit Auxiliary Power Supplies 1.2.4-2 1.2.4.4 Common Auxiliary Systems 1.2.4-2 1.2.4.5 Inter-Plant Effects of Accidents 1.2.4-4 e

8207270119 820720 PDR ADOCK 05000237

' K 'PDR

1.1.1-2 1.1.1.3 Updating of Original FSAR This document is the Updated Final Safety Analysis Report (UFSAR), a report separate and distinct from the original Final Safety Analysis Report.

The original FSAR and the associated docket files (50-237 and 50-249) are the basis for the licensing of the plant. In the event that a discrep-ancy exists between the orginal FSAR and the UFSAR, the orginal FSAR will be the final authority. The Technical Specifications may reference the UFSAR.

The UFSAR is revised annually as required in 10CFR50.71e. The UFSAR is designed to serve as a reference document, reflecting the current con-figuration of the plant, including infonnation and analyses required by and submitted to the NRC since submission of the original FSAR, and con-taining the infonnation in a contiguous fonnat.

2.2.2-5 TABLE 2.2.2:3 RECREATIONAL AND INSTITUTIONAL FACILITIES NEAR STATION Distance from Station (miles) Direction Illinois, Kankakee, and Des Plaines rivers Adjacent Goose Lake State Park 1.0 SW Collins Lake 2.0 w Des Plaines Conservation Area 2.5 SE Illinois Department of Corrections, 3.2 w Morris Juvenile Residential Center There are additional private recreational facilities such as gun clubs and picnic grounds scattered throughout the strip-mined areas south* of the

.station. A small unn.amed public park is also located 1.5 miles east of the station on the Des Plaines River. Public access is available to the Dresden Lock and Dam and a public path parallels the Illinois and Michigan Canal

  • e \

which approaches within 0.7 miles north of the station. The recreational facilities are apparently being actively expanded and improved and data on daily use. indicate a substantial increase in recreationists in recent years.

2.2.2.3 Summary. The EAB of the Dresden Nuclear Power Station, as reported previously, has no pennanent residents. Pennanent population distribution around the station has not changed significantly although total population within the five mile LPZ has increased to an estimated 10,400 residents from 5,090 reported in the FES. The 1980 population was projected to be 8,003 in the LPZ {FES Figure 2.2). Industrial facilities and recreational facil-ities have also expanded although their distribution is largely unchanged.

The daily maximum transient population including visitors to recreational facilities and workers employed by industries within five miles of the station is estimated to be approximately 11,000.

Rev. 1 June 1983 Si .

TABLE OF CONTENTS SECTION 5 -- CONTAINMENT SYSTEMS Page 5.1

SUMMARY

DESCRIPTION 5.l.0-1 5.2 PRIMARY CONTAINMENT SYSTEM 5. 2. l-1 5.2.1 DESIGN BASIS 5.2.1-1 5.2.2 DESC RI PTI ON 5.2.2-1 5.2.2.1 Drywel 1 5.2.2-3 5.2.2.2 Vent Pipes 5.2.2-3

5. 2. 2 .3 Pressure Suppression Chamber (Torus) 5.2.2-4 5.2.2.4 Electrical Penetration Seals 5.2.2-4 5.2.2.5 Fluid Pipe Penetrations 5.2.2-5 5.2.2.6 Isolation Valves 5.2.2~13 5.2.2.7 . Traversing In-Core Probe (TIP) 5.2.2-20 5.2.2.8 Venting and Cooling System 5.2.2-20 5.2.2.9 Automatic Vacuum Relief 5.2.2-20 5.2.3 DESIGN EVALUATION 5.2.3-1 5.2.3 .* 1*
  • Siz:i ng. of Primary Conta.intnent . s*.2.3-I 5.2.* 3.2** . Containment Response to LOCA '5.2.3-3 5.2.3~3 - Characteristics After Reactor Blowdown s~ 2 *.3-8 5.2.3.4. Capability with Respect to Metal-Water 5.2~3-12 Reactions -

5.2.3.5 Seismic Analysis 5.2.3-16 5.2.3.6 Drywell Expansion Gap Design Allowance 5.2.3-26 5.2.3.7 Drywell Missile Protection 5.2.3-30 5.2.4 INSPECTION AND TESTING 5.2.4-1 5.2.4.1 Drywel 1 and Suppression Chamber 5.2.4-1 5.2.4.2. Containment Penetrations 5.2.4-2 5.2.4.3 Containment Isolation Valves 5.2.4-3 5.2.4.4 Containment Ventilation System 5.2.4-4 5.3 SECONDARY CONTAINMENT -- REACTOR BUILDING 5.3.1-1 5 .3 .1 DESIGN BASIS

  • 5.3.1-1.

5.3~2 DESCRIPTION OF SECONDARY CONTAINMENT .BARRI ER . 5.3.2-1 5.3.2.l Reactor Building 5 .3. 2.-1

5. 3. 2*. 2 Airlock Doors 5.3.2'."'2 5.3.2.3' Pipe and Electrical Penetrations 5.3.2-2 5.3.2.4 Reactor Bldg Ventillation Isolation Valves 5.3.2-2 5.3.2.5 Standby Gas Treatment System 5.3.2-3 5.3.3 PERFORMANCE EVALUATION . 5. 3. 3-1 /

/

5.3.3.1 Exfiltration 5.3.3-1

Rev. 1 June 1983

  • 5.2.2:1 5.2.2:2 5.2.2:3 LIST OF FIGURES -- SECTION 5, CONTAINMENT SYSTEMS Pressure Suppression Containment System Typical Electrical Penetration Assembly Canister Center Section -- Low Voltage Power and Control Electrical 5iii Penetration Assembly 5.2.2:4 Center Section -- Shielded Signal Cable Electrical Penetration Assembly 5.2.2:5 Center Section -- High Voltage Power Electrical Penetration Assembly 5.2.2:6 Center Section -- Hot Fluid Piping Penetration Assembly 5.2.2:7 Center Section -- Cold Flu.id Piping Penetration Assembly 5.2.2:8a *Containment Vessel Instrument Line Penetration*

5.2.2:8b Process Stop Valve and. Excess Flow Check Valve Piping*

5.2.2:9 Main Steam Isolation Valve. -- Section 5.2.2:10 Main Steam Isolation. Valve -- Control Diagram 5.2.3:1 Recirculation Line Break -- Illustration 5.2.3:2 Pressure Response to Loss~of-Coolant Accident (LOCA) 5.2.3:3 Temperature Response to LOCA 5.2.3:4 . Pressure Response -- Calculations and Measurements 5.2.3:5a Bodega Bay Tests -- Vessel Pressure and Drywell Pressure 5.2.3:5b Bodega Bay Tests -- Vessel Pressure and Drywell Pressure 5.2.3:6 Comparison of Calculated and Measured Peak Drywell Pressure 5 *. 2. 3: 7 Contaj nment Capability .

5.2.3. :8 Mathema*trcal Model' :..;.._Earthquake ir:i N-S Direction*

5.2.3:9 Mathematical Model. -- Earthquake in E.. W Direction 5.2.3:10* Mathematical Model --Drywel 1 5.2.3:11 Drywell Displacement Di*agram ;.._ N-S Direction 5.2.3:12 Drywell Shear Di'agram -- N-S Direction 5.2.3:13 Drywell Moment Diagram -- N-S Directi.on 5.2~3:14 Drywell Displacement Diagram -- E-W Direction 5.2.3:15 Drywell Shear Diagram -- E-W Direction

  • 5.2.3:16 Drywel 1 Moment Diagram -- E-W Direction 5.2.3:17 Torus Se.ismic Model 5.2.3:18 Ring Header --*General Plan &. Location of Snubbers 5.2.3:19 Ring Header -- Attachment of Hydraulic Snubbers 5.2.3:20 Ring Header -- Location of Masses, Segment 1 5.2.3:21 Ring Header -- Plan of 24" Header 5.2.3:22 Ring .Header -- Section A-A 5.2.3:23 Ring Header -- Typical Header Support Assembly 5.2.3:24 CB&I Drawing 224, Rev. 8, 24 11 Header for Suppression Chamber 5.2.3.:25 CB&I Drawing 228, Rev *. 1, Support' Assembly .for 24 11 Header
5. 2 .. 3: 26 Drywel 1 Thenna l Expansion *

..5.2.3:27 Typical Penetration Joint*

5.2 *. 3:28 Resilient Characteristics of Polyurethane 5.2.3:29 Energy Necessary to Penetrate Drywell Containment

. 5.3.2:1 Reactor Building Superstructrue -- Blow-off Details 5.3.2:2 Diagram of Standby Gas Treatment System 5.3.3:1 Secondary Containment Pressure & Exfiltration After LOCA 5.3.3:2 Secondary Containment Pressure & Exfiltration After Refueling Accident 5.3.3:3 Perfonnance Curve*, Standby Gas Treatment System Exhaust Fan

DRYWELL SHELL MULTIPLE FLU ED HEAD FITIINGS PENETRATING PIPE INNER SLEEVE 2'6" MAX.

TYPE "1" FIGURE 5.2.2: 6 CENTER SECTION - HOT FLUID PIPING PENETRATION ASSEMBLY

Rev. 2 June 1984 5.2.2-7

  • Shielded signal cables are provided to interconnect low noise circuits between the reactor and the control ro001; in particular, the reactor neutron monitoring channels. Figure 5.2.2:4 shows a cutaway view of the containment penetration assembly for shielded signal cables. One type of circuit uses coax connectors mounted directly on the headerplates and isolated from ground. Another type of circuit uses connectors mounted on the penetration assembly auxiliary structure. The cable density is restricted to one circ-uit per three square inches of headerplate surface for the first type, and approximately 80 circuits of the latter type for each 12 inch penetration nozzle. '

A sectional view of the high voltage power cable penetration assembly is shown on Figure 5.2.2:5. The penetration assembly accomodates voltages up to 5 KV and cables as large as 1000 MCM and is designed to maintain low gas leakage rates and high insulation resistance. The high voltage cables are passed through ,openings in the headerplates and potting canpound applied to both sides of the headerplates to effect a pressure seal. The header-plates are constructed of stainless steel, a nonmagnetic material', in order to eliminate the possibility of eddy current heating.

5. 2. 2. 5 Fluid i e enetrations are of two general types; i.e., those which accommodate t erma movement hot), and those which experience relatively little thennal stress (cold). Fluid piping penetrations for which movement provisions are made are high temperature lines such as the main steam line and certain other reactor auxiliary and cooling system lines. A typiaal penetration of this type is shown in Figure 5.2.2:6. These penetrations have a guard pipe between the hot line and the penetration nozzle in addition to a double-seal arrangement. This permits the penetration to be vented to the drywell should a rupture of the hot line occur within the penetration:

The guard pipes are designed to the same pressure and temperature as the fluid line and are attached to a multiple flued head fitting, a one piece forging with integral flues or nozzles. This fitting was designed to the ASME Pressure Vessel Code,Section VIII. The penetration sleeve is welded to the drywell and extends through the biological shield where it is welded to ,a bellows which in turn is welded to the guard pipe. The bellows acccxnmodat'es the thermal expansion of the steam pipe and drywell relative to the steam pipe. A double bellows arrangement permits remote leak testing of the penetration seal. The 1ines have been constrained at each end of the penetration assembly to limit the movement of the line relative to the containment, yet permit pipe movement parallel to the penetration.

The only lines which connect to a high-pressure system which do not have a double-seal penetration sleeve are the* hydraulic 1ines to the control rod drives. These involve 354 small, stainless steel lines, shop-welded to three sections of the drywell plate. The mechanical problems involved with this number of small penetrations in a relatively small area make it impractical to provide individual penetration sleeves. The pipes are designed to deflect with the drywell shell. They a*re not individually testable, but will be tested as part of the overall containment leak rate test .

BIOL. SHIELD DRYWELL PENETRATION SLEEVE DRYWELL SHELL MULTIPLE FLUED

  • PENETRATING PIPE HEAD FITIINGS IFORGEDI TUBE TURN OR GUIDES AS APPROVED EOUA REOD.

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SLEEVE llN PLACE)

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4; FIGURE 5.2.2:7 CENTER SECTION - COLD FLUID PIPING PENETRATION ASSEMBLY

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1000

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~

800

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  • 0.0147

~ H *6 FT

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CJ TEST DATA

"'> 600

- CALCULATED 50 40

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  • -~ * * * ><

o *o CJ Toi

  • 184° F, CARRYOVER :::::95%

30

  • Toi* 70° F, CARRYOVER :::77%

)( Toi* 64° F,CARRYOVER:::::30%

0 TQi

  • 103° F, CARRYOVER ::::: 8%

20 - CALCULATED HOMOGENEOUS CARRYOVER

---CALCULATED ZERO CARRYOVER

- * -CALCULATED ZERO 10 CARRYOVER AND CONDENSATION 0

0 2 3 4 5 6 7 8 TIME !SECONOSl FIGURE 5.2.3:4 PRESURE RESPONE - CALCULATION a MEASUREMENTS

5.2.3-29

  • 7. Using a special T-shaped tool, a 3 strip of epoxy impregnated 11 fiberglass tape was placed behind the panel joint.
8. The joint was filled with epoxy and a second 3 strip of epoxy 11 impregnated fiberglass tape was placed over the joint to complete the closure.
9. After the tie plates in the fiberglass were rigidly attahed to the outside plywood forms, the fiberglass shell became the inner form for the pouring of the concrete structure .

In addition to the steps followed above, the following special precau-tions were taken at the junction of the expansion gap filler and pipe pene-trations:

1. The polyurethane foam sheets were applied on the drywell shell tight against the penetration.
2. The penetration pipe sleeve was placed on the penetration stopping at the polyurethane foam sheet (i.e. 2-1/4 from drywell shell).

11

3. The cover panels were placed to within approximately 1/4 of the 11 sleeve and the joint between the cover panels and the sleeve was caulked with epoxy caulking.
4. Epoxy and fiberglass tape were applied to join the sleeve with the cover panels.

A diagram of the joint at pipe penetrations is given in Figure 5.2.3:27b.

TABLE 5.2.3:4 MATERIALS USED TO FILL DRYWELL EXPANSION GAP

1. Polyurethane Foam: This-material is a polyester base flexible polyurethane foam manufactured to exacting controls from refined raw materials to produce a quality foam suitable for use in areas of high radiation. Sheets used conform to the following require-ments:
a. Base Specification: MIL-PPE~200F.
b. Chemistry: Isocyanate foam formed by reaction of poly-isocyanates with polyester polyols.
c. Density: 2 pcf +/-0.10 pcf.
d. Thermal Value: .26 K factor.
e. Service Temperature: 285°F.
f. Physical Properties:

(1) Tensile - 12 psi minimum.

(2) Elongation - 100%

(3) Canpressibility - 35% at 1.0 psi maximum.

.e g.

(4) Compression Set - 10% at 50% compressibility .

Sheet Size: 2-1/4 x 2' x 8 with tolerances as specified by MIL-C-26861.

11 1

6iv LIST OF FIGURES -- CHAPTER 6, ENGINEERED SAFEGUARDS 6.2.2:1 Emergency Core Cooling System Versus Break Spectrum 6.2.3:1 Core Spray Cooling System 6.2.3:2 Core Spray Cooling System Pump Characteristics 6.2.3:3 Core Spray Pipe Protection 6.2.3:4 Core Spray System Functional Control Diagram 6.2.3:5 Unassisted Core Spray Performance 6.2.3:6 Core Spray Distribution Effect of Total Flow rate (Lm'ler Header) 6.2.3:7 Core Spray Distribution Effect of Updraft (Lower Header) 6.2.3:8 Core Spray Distribution Effect of Open Elbow Inclination (Lower Header) 6.2.3:9 Core Spray Distribution Effect of Open Elbow Azimuth (Lower Header) 6.2.4:1 LPCl/Containment Cooling System 6.2.4:2 LPCI/Containment Cooling System Pump Characteristics 6.2.4:3 LPCI Break Detection System Logic Arrangement 6.2.4:4 LPCI/Containment Cooling System Functional Block Diagram 6.2.4:5 LPCI Logic Control System Arrangement 6.2.4:6 Unassisted LPCI Performance 6.2.5:1 HPCI Sys tern 6.2.5:2 HPCI System Functional Block Diagram 6.2.5:3 Peak Clad Temperature vs. Break Size, HPCI and ADS with Core Spray or LPCI 6.2.5:4 Depressurization Rates, HPCI and ADS with Core Spray 6.2.5:5 Unassisted HPCI Performance 2 6.2.5:6 Level Transient Following a 0.2 ft Steam Break (Both HPCI and ADS Initiated) 2 6.2.5:7 Level Transient and Flow to HPCI Nozzle Following a 0.2 ft Steam Break 6.2.5:8 HPCI Pump Characteristics 6.2.5:9 Example Turbine Capacity Curves 6.2.6:1 Automatic Depressurization System -- Functional Block Diagram 6.2.6:2 Automatic Depressurization Relief System -- Functional Control Diagram 6.2.6:3 Jet Pump Model for Transient Analysis 6.2.6:4 Reverse Flow Resistances of Jet Pumps 6.2.6:5 Slowdown from a Bottom Location 6.2.6:6 Vessel Pressure and Level Traces -- Bodega 30 6.2.6:7 Vessel Pressure and Level Traces -- Humboldt 17 6.2.6:8 Vessel Pressure and Level Traces -- CSE Data, Run ~-15 6.2.7:1 Required HPCI Mixing Efficiency vs. Break Area (ft ) To Prevent Clad Melt 6.2.7:2 Break Analysis Model -- Flow In and Out of Reactor 6.2.7:3a Main Steam Flow Simulation Model 6.2.7:3b Core Spray Decay Heat Removal Model 6.2.7:4 Reactor Coolant Level Swell 6.2.7:5 Short Term Core Inlet Flow and Pressure Transient

6.2.3-9 Periodic system tests using test lines.

Leak-off lines between isolation valves.

Drainline on pump side of outboard isolation valves.

Motor-valves can be exercised independently.

3. Pumps Preoperational test of entire system.

Periodic system tests using test lines.

Pump seal leakage is monitored.

4. Spray Sparger Preoperational test of entire system.
5. Spray Nozzles Pre-operational test of entire system.
6. Relief Valves Can be removed and tested for tesi point.
7. Screens Preoperational test of entire system.

Periodic system tests using test lines.

Pressure indicator on pump suction during above tests.

Each core spray subsystem may be tested individually during reactor operation as follows:

1. The pu~p of the loop under test may be started by its manual control switch. The test bypass valve is opened to allow the pump to be tested at full flow. Flow and pressure instrumen-tation is observed for correct response and the system outside the drywell may be checked for 1ea ks.
2. The admission valves and the testable check-isolation valves may be tested independently of the pump and flow test as follows:
a. The nonnally open maintenance valve upstream of the normally closed admission valve is closed by the control switch.

Limit switches on the maintenance valve act as a permissive to open the admission valve which may then be exercised (opened and closed) by manual actuation of the control switch.

Rev. 1 June 1983 6.2.3-10

b. At th~ end of the test, with the startup valve fully closed,.

the maintenance valve must be opened.

In the event that a reactor low,.. low water level and reactor low pres-sure actuation signal, or a high drywell pressure actuation signal occurs during a loop test, the loop not under test will start automatically. The loop being tested wil 1 return automatically to the operational mode and will then* restart aLitomatical ly.

The pressure differential between :the system piping inside the ves.sel

  • and an internal reference pressure is monitored dudng power opera ti on.

Changes in these pressure.readings will* provide indicatiori of* loss of integ-rity of piping within the containment vessel. In addition, pipes,-pumps, valves, and other working components outside of the primary containment ca~ be visually inspected ~t any time.

  • All n'ecessary experimental programs to confinn .the perfonnance of the core spray system were perfonned.
  • These are descri~ed below:

,.'* The.purpose* of these tests was to. detennine a core spray no*zzle

~ ,_*. arrangemen*t . and *aiming angles .such that every bund*le *in the* core:

  • .would' receive an adequate* ambun.t of* water* during* core spray ope.ration.
  • . :e**

Adequate flow is defined as 2.45 gpm per* bundle.

  • The *test section used

. to conduct these tests employed a full. scale mocku~ of the: upper secti6n_

  • ~* ....  :

of the c.ore and. al sq emp 1oyed a ~pray .spa rger ri rig fabricated to: t_he

  • exact dimens i o'ns. of the spa rger* ring. Tes ts were conducted for both the
  • upper and lOwer spray* rfog* Sparger. ,The effects of Sparger fl OW, a'imi ng angle tolerances, and updraft were investigated thoroughly~

The core spray nozzle* arrangement adopted for the units consists.of 65 1HH12 full jet nozzles and 65 1-inch open elbows. The. spray outlets

  • are equally spaced with the full jets and open el bows alternating around the sparger. The core spray sparger ring is actually two 130° sections with flow entering each section 15° from the midpoint.
  • Unifonn distribution of flow around the sparger was obtained by the use of orifices placed in the entrance of the spray branches around the sparger.
  • Aiming angles for the full jet and open el bows were* detennined for both the upper and lower sparger.
  • Variation in nozzle aim-

. ~ . ing angle, updraft velocity,. core spray* flow rates. were, tested to detennine the sensitivity to the variables. These effects are.discussed in the following paragraphs .. - However, only the results of the .lower sparger are reported here since both upper and lower spargers gave the same result. *

1. Effect of Flow Variation Tests- were run with flows as 1ow as 4100 gpm. and* as high. as .,'

5850 gpm. The flow distributions achieved are shown ii1 . f Figure 6.2.3:6. Design flow rate for the core spray is 4500 gpm .

. ()

i f.

6. 2 .3-11
2. Effect of Updraft The results of these tests are the same as reported in Topical Report APE0-5453 and Oyster Creek, Docket 50-219, Amendment #30. Updrafts expected in the core are in the range v1here they 1>/i 11 have little or no effect on the over-all spray distribution. Figure 6.2.3:7 illustrates this effect.
3. Effect of Nozzle Aiming Angle The effect of inclination angle is shown in Figure 6.2.3:8 for the open elbows of the low header. Variation of the full jet inclination angle has even less effect and is not shown. Effect of azimuth setting is shown in Figure 6.2.3:9.

The design value for azimuth for all nozzles is 0°. As a result of these sensitivity tests, an allowable tolerance of

+/-1° has been set for all angle settings.

B. Core Spray Heat Transfer Tests The core spray heat transfer tests were completed and are reported in Topical Report APED-5459. These tests covered fuel bundles having an initial power level of 6.5 MWt, well in excess of the peak bundle power for the units. Peak initial temperatures of 1600°F were also tested well in the range of those predicted for the units, which are all. under 2000°F. Actual core spray operating temperatures of 2350°F were tested,*

well in excess of the maximum 2000°F predicted for the units. Bundle flow rates from 0.5 gpm to over 3.25 gpm were tested including the 2.45 gpm minimum measured flow rate for the spray distribution. As shown in Topical Report APED 5458, flows into the bundle as low as 0.5 gpm effectively cooled the rods. A more detailed discussion is given in the above report. Considerable margin exists beb1een the minimum flm'ls required for cooling the fuel rods and that actually being injected.

6.2.6-4 (See discussion of channel burnout later in this section). Since the ind~ced equivalent shut off head of the jet pump is well in excess of this amount, it is clear that plenum flow will be prevented from bypassing the core through the broken loop jet pumps. Thus all the flow from the unbroken loop must go through the core until such time that the top of the jet pumps uncover.

When the mixture in the downcrnner region of the reactor is completely discharged through the broken recirculation line, the depressurization rate in the vessel will increase markedly since steam will be leaving through the break rather than liquid. This will produce vigorous flashing of the lower vessel plenum inventory, resulting in high flow rates into the core and backwards through both sets of jet pump diffusers. The split of the flow leaving the lower plenum depends upon the resistances of three flow paths, i.e., through the unbroken side jet pumps, through the core, and through the broken side jet pumps. The reverse flow resistances of the two sets of jet pumps are calculated using the models shown in Figure 6.2.6:4.

Since it is conservative for the reverse flow in the jet pumps to have a low resistance, the minimum expected loss coefficients are used.

If the area of the jet pump throat is AT and the specific volume of the flow leaving the lower plenum is v, the tota pressure loss between 1 and 4 is 2

Kl-4 Wjl v + _z_

2g A t

2 v for case (a), and

+

for case (b).

WR is calculated by assuming critical 1flow at the minimum nozzle area and applying the Moody blowdown model .

For flow through the core, the total pressure loss between core inlet and outlet is given by an expression of the form 1Moody, F.J., "Maximum Two Phase Vessel Blowdown from Pipes" ASME Paper No. 6 5-WA/HT -1.

6.2.6-8 For the reasons above, the assumption that the flow distribution through the core during a loss-of-coolant transient is unchanged from the steady-state distribution is justifiable.

6.2.6.3.4 Justification for using the Hench-Levy CHF correlation for a system undergoing a rapid depressurization can be shown by examining the fuel rod time constant, the vessel depressurization rate, and the core inlet flow rate. Fundamentally, the CHFR is evaluated using a model which simulates the BWR 49 fuel rod bundle. The basic input consists of time varying pressure, core inlet flow, and core power which have been determined by other calculations. The model includes axial, radial, and local power distribution as well as distributed loss coefficients. The Bl~R fuel rod time constant is approximately 10 seconds. The maximum vessel depressurization rate would be approximately 50 to 100 psi/sec depending on the postulated break of either one recirculation line or one steamline at the vessel respectively. At rated core flow, it would take approximately 0.3 seconds for the coolant to travel through the core. Even under degraded flow conditions, the flow transit time in the core is not drastically altered primarily due to the voiding effect and its associated increase in velocity.

The average depressurization of the coolant per axial node (20 nodes/

core) would be about 1 psi/sec. Furthennore, since the time constant of the fuel is approximately 10 seconds the node surface heat flux variation during the transit time of the coolant would be negligible. Therefore, for all practical purposes, the coolant is subjected to quasi-steady conditions because the changes in the key parameters are not rapid and the Hench-Levy steady state CHF data can be applied.

Another basis for applying the CHF correlation is that all evidence to date indicates that the use of steady-state CHF data for predicting transie~t CHF conditions is conservative even during truly rapid tran-sients. Therefore, by applying steady-state multi rod CHF data to this transient analysis, there is an inherent factor of conservatism. Further margin exists in this analysis, especially at the end of the transient because the actual improvement in CHF which occurs at pressure down to 600 psi over that at 1000 was neglected. Thus, there are also only two heat transfer correlations used for large break anal,yses during blowdown.

When the MCHFR is greater than unity, based on the Hench-Levy CHF corre-lation, the Jens-Lottes heat transfer correlation is used to obtain the nucleate boiling heat transfer coefficient. The nucle~te boiling heat transfer coefficient is of the order of 10,000 BTU/(ft -°F-hr) or hig2er.

If the surface heat transfer coefficient is above few hundred BTU/(ft -°F-hr),

the peak clad temperature is unaffected by higher valves of heat transfer coefficient. This is because the heat transfer from the fuel is then limited by conduction through the fuel pin itself rather than the surface heat transfer coefficient.

1Tong, L. S. "Transient CHF Prediction" Presented at ASME-AICHE Heat Transfer Conference, Philadelphia, Aug 1968.

6.2.6-9 Hhen the MCHFR is less than unity the surface heat transfer coefficient is assumed to be zero. This is conservative because film boiling heat transfe2 coefficients would exist which are of the order of 100 to 1000 BTU/(ft -°F-hr) and would actually provide additional cooling during the blowdown phase.

6.2.6.3.5 Level swell during blowdown. Mixture level and local properties in a vessel during loss-of-coolant detennine core heat transfer environment and nature of 1blowdown. Mixture level predictions were made with an ana-lytical model which has been compared with experiments, and reasonabl2 level swell compariso~s have been made with EVESR steam blowdown tests and CSE blowdown data .

It consistently has been stated and shown that calculated ~lowdown rates used in loss-of-coolant analyses are faster than expected . There-fore an acceptable evaluation of the mixture level model should be uncoupled from the blowdown model. Ideal data to use in an evaluation of the mixture level r:iodel is a time-dependent plot of (1) actual level, (2) mass remaining in the vessel, and (3) vessel pressure. Data usually published do not include all three measurements. However, pressure-time traces for bottom blowdowns exhibit a 11 knee 11 , which is characteristic of a sudden change from liquid (or mixture) to vapor blowdown. It is, therefore, reasonabJe to use the mixture level model with a blowdown which closely predicts the given pressure trace. Time at which the measured knee appears can be compared with the time required for mixture level to reach zero elevation.

It is the purpose of this work to compare predicted mixture level with selected measurements showing either a pressure-time knee, or actual level measurement.

Flowing Quality and Bottom Slowdown

  • Slowdown rate depeijds on break geometry, vessel stagnation pr,essure, and stagnation enthalpy . The actual value of stagnation enthalpy is determined partly by mixture properties in the vessel, and partly by the fact that vapor bubbles rise through liquid. Figure 6.2.6:5 indicates blowdown frarn a bottom location. Symbols and subscripts used in the*

analysis are shown in Table 6.2.6:2.

1Moody, F.J., "Liquid-Vapor Action in a Vessel During Blowdown 11 , APED-5177, 1956.

2ranni, F.W.: Fritz, J.R.: and Law, D.D., "Design and Operating Experiences of the ESADA Vallecitos Experimental Superheat Reactor (EVESR) 11 , APED-4784, 1965.

311 Nuclear Safety Quarterly Report, Feb., March, April, 1968, for Nuclear Safety Branch of USAEC Division of Reactor Development and Technology",

BNWL-885, October, 1968.

4Moody, F.J., "Maximum Two-Phase Vessel Bl owdown from Pi pes 11 , APED-4827, 1965.

6.2.6-11 Total blowdown rate 1~

8 is composed of vapor and liquid flows WgB and WfB:

wgB = XF WB = XF Ge AB (1) wfB = (1 - Xf) WB = (1 - Xf) Ge AB (2)

The tenn X is flowing quality, or vapor mass flow fraction leaving the bre~k. Critical flow rate per unit break area G (P 0 ,h 0 )

can be considered a function of P0 and Xf if stagnation enthglpy h0 is expressed by ho = hf(Po) + XF hfg(Po) ( 3)

Vapor and liquid mass conservation equations are written as follows for the dotted control volume where elevation y is very small:

wg 8 + wg = o (4) wfB + wf = o ( 5)

The tenns W and I~ are upward flm<1s of vapor and liquid 1 at elevation y, exp~essed by:

(6)

VJ = A u [-v- ( 1 - X) - _u_J ( 7) f v v vf vJhere V is bubble rise velocity relative to the vessel, u is bubble rise velocity in stationary liquid, and v is local mixture specific volume based on X, which is local instantaneous vapor mass fraction. Equations 1 through 7 can be combined to give:

( 8)

=

Equation 8 relates P , X, and XF to the property A /(A u). If flm'ling quality X is zer8, then stagnation enthalpy is 8 h (~),and bottom blowdown of saturated liquid occurs; i.e., the con~it9on for A8/(Avu) to be satisfied for bottom liquid blowdown is that 1Moody, F.J., "Liquid - Vapor Action in a Vessel During Slowdown", APED-5177, 1956.

6.2.6-12 A8/(Avu) for liquid is less than or equal to Gc(Po,hf(PJ) vf(p ) (9) 0 However, if the above inequality is not satisfied, vapor bubbles will be entrained in the blowdown flow, and Equation (8) must be used to detennine X

  • Bottom liquid or mixture blowdowns from 1000 psig characterist,cally produce X in the range O to 0.10.

The corresponding effect on blowdown rate is small. It follows from Equation 8 therefore that 1

(10)

Equation (10) can be used to help select test data for comparisons which can be closely approximately by liquid blowdown until the mixture level reaches the break.

Slowdown Rates and Pressure Traces Calculated graphs already are available for liquid blo~down from 1000, 1250, and 2000 psia initial stagnation pressures .

The graphs include a variable time scale with break area A8 as a*

parameter. Proper selection of an equivalent A to bring theore-1 tical and measured blowdown pressure traces into 8 agreement will enable a better evaluation of the mixture level model.

Mixture Level Prediction.

Mixture level has been calculated for vessel bottom blowdown from 1000 psia initial pressure with initial liquid level equal to 75 percent .}-

of the vessel overall height.

Where necessary, mixture level calculations can be made from the following equations whenever XF is greater than o from Equation 10:

1Moocty, F.J., "Liquid-Vapor Action in a Vessel During Blowdown 11

, APED-5177, 1956.

2Moody, F.J., "Perfect Nozzle Slowdown Study", APED-4398, 1963; Moody, F.J., "An Analytical Model for Pressure Suppression 11 , APED-4734, 1964.

6.2.6-13 YL (t')

H

( 11)

I 1(t') = Integral from zero tot' of

-[sf(o) - sf(t')J I [Sg(t') - sf(o)J dt' (12)

The time t' is given by AB t' = -t (13)

Mo Selected Tests for Comparison Three t3sts were selected for this comparison from CSE 1 , Bodega 2 ,

and Humboldt , blowdowns. Pertinent experimental quantities are listed in Table 6.2.6:3.

1Nuclear Safety Quarterly Report, Feb., March, April, 1968, for Nuclear Safety Branch of USAEC Division of Reactor Development and Technology, BNWL-885, October, 1968.

211 Preliminary Hazards Summary Report, Bodega Bay Atomic Park Unit No. 111 ,

PG&E, Dec. 28, 1962.

3Robbins, C.H., "Tests of a Full Scale 1/48 Segment of the Humboldt Bay Pressure Suppression Containment", GEAP-3596, 1960.

6.2.6-14 TABLE 6.2.6:3 HENCH-LEVY CORRELATION -- IMPORTANT EXPERIMENTAL QUANTITIES Bodega 30 Humboldt 17 CSE, B-15 Vessel Volume, ft 3 80 55 ' 150 Initial Pressure, psia 1250 1250 2000 Vessel Height H, ft 20 31.4 15 Vessel Area Av, ft 2 3.97 1.75 9.36 2 .0573 .044 Break Area AB' ft .0643 Initial Fluid Mass, lbs 2420 1795 6700 YLo Initial Level Fraction -H- .685 .735 1.0 Equivalent Break Area A B1

.053 .0388 .0446 A' B sec Value for ft .0134 .0222 .00477 Av The tenn GU 1 ' sec .00513 .00513 .00366 0 f tt Bodega and Humboldt vessels ~~ere cylindrical without internal mechanical components to obstruct internal flows. The CSE vessel contained a dummy core plate.

The equivalent break areas were detennined so that theoretical and measured pressure-time curves were closely aligned up to the pressure trace knee. The quantity A' /A u was based on u = 1.0 foot per second bubble rise velocity, which ~ee~s to be more characteristic of small vessels. All three cases show that A' /Au is larger than the tem1 l/(Gf Uf) initially so that mixture bl~wd~wn is assured. (Gf decreases as pr~ssure drops so that mixture blowdown is assured througnbut the blowdown.) It follows that Equations 11 and 12 can be applied for necessary calculations.

6.2.6-15 Rather than calculate level curves for Bodega 30 and Humboldt 17, it was decided to make an approximate comparison with already available calculations from initial vessel pressures of 1000 psia. The tests begin at 1250 psia. Figures 6.2.6:6 and 6.2.6:7 show true vessel pressure traces with attention called to the knee. The lower half of each graph gives mixture level calculations based on 1000 psia, and 75 percent initial water level. A dotted curve also is shown, based on the appropriate A1 R/ u Hhen mixture level reaches zero, the pressure 11 knee 11 should occur.

Evenvthough the calculation is based on 1000 psia and the tests were run from 1250 psia, mixture level disappearance was predicted within 12.5 per cent using the model described and this is a direct index of its accuracy.

The CSE Data in Fig. 6.2.6:8 includes both pressure, mass remaining, and level for comparison with the mixture level model. Comparing only the level data in Figure 6.2.6:8 it is seen that a bubble rise velocity between 0.5 fps and 1.0 fps brackets the CSE measured level. Note that the level swell model is not applied to the test data until the system is ~aturated. Note also that the 0.5 fps curve shows liquid vanishing at about the right time as indicated by the knee in the experimental curve. Thus reasonable agreement is also shown with the CSE data.

Additional comparison with the only other av~ilable data is discussed by Moody in which the level rise data from EVESR is discussed.

It is concluded from the comparisons v1ith this data that the model has sufficient accuracy for predicting level rise.

6.2.6.4 Inspection and Testing 6.2.6.4.1 Environmental testing was perfonned to verify the operability of the electromatic main steam relief valves actuators under accident condi-tions. The actuator was placed in an autoclave on a special base which ,,

supported it in an upright position. A spring load was applied to.. the actuator, representative of the relief valve operating load on the solenoid.

Instrumentation and electrical leads to register the solenoid action and to energize solenoid coils were carried through the autoclave wall by means of insulated plugs. The electrical leads were Vulkene cable which is the same as used in operation. A pressure gage, thermometer, and a safety valve were installed in the autoclave. The limit switches on the actuator were also connected to signal lights to prove their integrity. The test was accomplished by energizing the solenoid at five minute intervals for one minute duration after the autoclave had been pressurized with saturated steam at 62 psig (308°F). A record of the total test time to pressurize, time at pressure and number of successful operation cycles \'las recorded.

The autoclave atmosphere was maintained at 62 psig with saturated steam for the duration of the test.

1Moody, F.J., 11 Liquid - Vapor Action in a Vessel During Blowdown 11 , APED-5177' 1956.

2ranni, F.W.; Fritz, J.R.; and Law, D.0., "Design and Operating Experiences of the ESADA Vallecitos Experimental Superheat Reactor (EVESR)", APED-4784, 1965.

Rev. 1 June 1983 8.2.1-2 The auxiliary power supply for Unit 3 is the unit auxiliary power transfonner which is connected to the generator leads. *The re.serve auxiliary power supply is from the reserve *au*xil iary power transformer*

which is connected to the 345 kv bus at Dresden. *

  • The system auxiliary transfonners step the transmission voltage down to the station 4160 volt system. Each reserve auxiliary transfonner is size~ to provide the total auxiliary load of one unit plus one division of engineered safeguards auxiliary power for the other unit. There are two breakers to allow Unit 2 4160 volt Bus 24-1 and Untt 3*4160 volt Bus 34-1 to be>tied in an emergency .. This configuration provides availability of *redundant sources of offsite power. *
  • The auxiliary power supplies. from the 138 and 345 kv transmission systems are protected against the effect of unplanned outages by the diversity of six separate 345 kv circuits and five 138 kv circuits and three major generatin~ units feeding into the two swi~chyards at the Dresden site.
  • Each unit has adequate auxiliary power supply from either the 138 or 345 kv swi.tch-yard or from diesel generators.

It. i.s imposs i b.l e for the failure of any orie component* of ei the*r the "138 O:r 345 kv transmission sys terns to caus*e *a s;imultaneoµs pu;t~ge '.9:f:'

  • both bt:.ls~s at>or~sde~'*' *::. '. >**>'. *:.* ',. '*, ' ' ..:.' " .,. '*;t'_: ... ... *~<- *.: ':~**' :: .
8. 2 . 1. 4 An aTys i s .l The probability of losing the offsite el-ctric power supply has been
  • minimized by the design of the Commonwealth Edison generation and trans-

. mission system.* In.creased reHability is provided through interconnections to neighboring systems. In 1977, the Commonwealth Edison transmission sys-tem consists, in part, of seventy-four 345-kV lines totaling 2244:miles, and three 765-kV 1 ines totaling 152 miles. The transmission* system is intercon-nected with neighboring electric utilities at 28 points, 12 at 138;..kV, 15 at 345-kV and 1 at 765-kV.

Commonwealth Edison is a member crf Mid--American Interpool Network (MAIN).

In general, all electric utilities in Illinois, Missouri, Upper Michigan, and the eastern half of Wisconsin are members of MAI.N. At the beginning of 1977, .

the transmission within MAIN consisted of 120 345-kV lines totaling 4714 mile.s and 3 765--kv lines totaling 152 miles. One of the functions of .MAIN is to en-

. sure that the transmission system is reliable and adequate. Power flow.and*

  • transient stability studies are con.ducted on a regular basis using the criteria*

stated in MAIN Guide No. 2 (Reference .1), a portion of which is as follows:

"The: generation and transmission system shall be adequate to withstand the most severe of the following set of contingenties without fesulting in an uncontrolled widespre.ad tripping of lines and/or generators with resulting loss. of load over a large area: *

1. Sudde.n/~utage of any tower 1 ine at the. time when any other one circuit is out of servfc~ *.

1.

Rev. 3 June 1985 8.2.2-6

  • All protective circuit breakers are sized according to standard electrical industry practice where maximum current interrupting capabilities of the circuit breakers exceed the available line-to-line or 3 phase short circuit current taking into account the impedences of the generator, transformers and other electrical system components.

On loss of auxiliary power the reactor will scram (if above 453 Turbine first stage pressure), and if auxiliary power is not restored immediately, the diesel generators are designed to automatically start and carry the vital loads for an indefinite period. The buses are so arranged that the vital shutdown loads are automatically transferred to the diesel generators.

Auxiliary power is normally supplied by the unit auxiliary power transformer and the reserve auxiliary transformer with the load divided between them. These transformers supply power to the equipment used to maintain a safe and operable plant. It is very improbable that both electrical power sources would be lost simultaneously\

because each is supplied from a different source. Nevertheless the loss of all.auxiliary power is assumed for design purposes.

Unit 2 does not depend on Unit 3 to achieve. redundancy in offsite power. Units 2 and 3 export power from the station on the 345 kv system. Off-site power for Dresden Unit 3 is supplied by the 345 kv system. The switchyard is designed so that no single bus fault or errant opening of a breaker will cause a loss of power to Unit 3. Similarly,

  • Unit 2 is supplied by the 138 kv system. Again, no single bus fault or accidental opening of one breaker will cause a* loss of power or subsequent. scram. Transformers 81 and 83 also .connect the 345 kv and 138 kv switchyards together. Thus, it is possible to supply Units 2 and 3 from alternate high voltage sources. This power is supplied through the reserve auxiliary transformer, which is capable of carrying the .. entire load, and the 138 kv switchyard. Redundancy in offsite pow~r is achieved by multiple 138 kv lines to the 138 kv switch-yard.

The following systems from Unit 3 are capable of feeding the Unit 2 systems:

(a) 125v battery (b) 125v de main bus and reserve bus (c) 125v battery chargers Numbers 3 or 3A (d) 125v de cable feeders to switchgear (e) 250v de system The following systems or equipment are required for operation for both Units 2 and 3, and so have redundant power supplies.

(a) Cardox C0 2 System Controls (b) Standby Gas Treatment System (2 trains, each of which can be used on either unit; *Each unit supplies power to one of the trains.)

0034f OOlOf

Rev. 2 June 1984

8. 2. 3-11 The Unit 2 Reactor Building Bus No. 2 is nonnally supplied from the
  • Unit 2 Turbine Building Main Bus No. 2 (as described above) and is the nonnal source of control power for the Unit 2 4160 volt switchgear 23-1 and 480 volt switchgear No. 28, Reactor Building escape lighting, etc.

It is the reserve source of control power to Unit 2 4160 volt switch-gear 24-1 and 480 volt switchgear 29.

The Unit 2 Turbine Building Reserve Bus No. 2 is nonnally supplied from the Unit 3 Turbine Building Main Bus No. 3 and is the nonnal source of control power for 4 kv switchgear 24-1 and 480 volt Switchgear 29 and all of the Turbine Building control buses for Unit 2.

Note that the control power for one set of reactor building switchgear is canpletely independent (including the battery) from the control power to the other set of switchgear in the reactor building.

The 125 volt battery discharge rating is 62.3 amperes for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

{498 ampere hours). The battery is sized to carry the following loads for the time* periods as indicated:

TABLE 8.2.3:4 125 VOLT DC BATTERY LOADINGS

.1.

125 Volt DC Battery Loads ,limp &

Direct Connected Loads Estimated Power Required Dura ti on Escape lighting 8.16 Kw 65 amp/4hr

2. Annunciator relay cabinet & visual annunciator 1.0 Kw 8 amp/3hr
3. Indicating lamps and auxiliary relays 2.0 Kw 16 amp/3hr
4. Plant sirens 2.0 Kw 15 amp/lOmi n
5. Electranatic relief valves 30 amp/15min
6. Trip (4) 345 Kv O.C.B. 120 amp/lmin 7, Trip (!}-Generator field breaker 10 amp/lmin
8. Trip {20} 480/4160 Kv A.C.B. 120 amp/lmi n
9. Turbine main trip solenoid 10 amp/lmin
10. Close (4) 4160 A.C.B. 40 amp/lmin
11. Standby diesel generator field flashing 140 amp/lmin
12. HPCI turbine controls 5 amp/4hrs
13. TIP system shear valves 50 amp/lmin
14. HPCI Turbine drain.valves 5 amp/4hrs 8.2.3.2.3 Nuclear lnstument Supply Systems 48/24 Volts DC The electrical supply for the source range monitor and intennediate range monitor systems, and the stack gas, radwaste discharge, and off-gas radiation monitors, consists of two duplicate 48/24 volt 3 wire, grounded neutral systems. See Figure 8.2.8:3. Each system consists of two 24 volt, 80 ampere hour (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rate) batteries in series and connected to a de distribution panel. There are two silicon rectifler type 25 ampere battery chargers on each system, one of which is connected to each of the 24 volt batteries. The source of power for the battery chargers is the 120 volt ac instrument bus. Each 48/24 volt system is equipped with under-voltage and over-voltage alanns. The battery chargers are capable of completely recharging the battery in approximately 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> while simultaneously supplying the nonnal continuous load, estimated at 15 amperes.

Rev. 3 June 1985 8.2.3-12

  • Modifications have been made to Unit-3's 48/24 volt DC system. The existing cells (96 cells_of Gould type DPR-19 Plante Battery) have been replaced with 96 cells of the Gould type 2-KCX-190 lead calcium battery.

The existing racks were modified (seismically qualified) to accept the larger physical size since the batteries have a 190 amp-hour capacity compared to the old 80 amp-hour, subsequently the neutron monitoring reliability is expected to be improved. The rated capacity of this new battery is 2.38 times greater than the battery it replaces. To accomodate the new batteries,*the float and equalization voltages were adjusted upward on the battery chargers.

8.2.3.2.4 D.C. System Alarms The reliability of the D.C. Systems has been improved by the installation of D.C. Voltage Indicator and Under Voltage Alarms on the 12SV; 250V and 48/24V Batteries. This will assure on adequate state of charge exists on the station batteries at ali times. ,

The'HPCI 125VDC annunciation alarm circuit has been modified to provide a means to monitor the status of both the main and reserve bus feeds. Should the o.c. feed auto-transfer, it would alert the operator if the reserve Bus is de-energized .

  • '0034f OOlOf

12.1.1-3 Standby Liquid Control System Emergency Core Cooling Systems High Pressure Coolant Injection System Automatic Depressurization System Core Spray Systems Low Pressure Coolant Injection/Containment Cooling System Containment Cooling Water System Standby Gas Treatment System Standby Coolant Supply System Fuel Storage facilities, to include spent fuel, handling equip. and new fuel storage equipment.

Standby Electrical Power System Station Batteries Emergency Diesel Generators Essential Buses and other electrical gear for power to critical equipment.

Control Room - Instrumentation & Control Reactor Level Instrumentation Feedwater Control Instrumentation Standby Liquid Control System Instrumentation Manual Reactor Control System Control Rod Instrumentation Control Rod Position Indicating System Reactor Protection System Neutron Monitor System In-Core Neutron Monitor System Area Radiation Monitors Process Radiation Monitors

TABLE 12.1.1:1 ALLOWABLE STRESSES FOR CLASS 1 STRUCTURES Structural Steel Reinfrcg. Steel Concrete Maximum Tension Shear Compr.

Maximum Allowable Stress: on Net on Gross on Gross Loading Conditions Allowable Stress Compr. Shear Bearing Section Section Section Bending I

1. Dead, Live, Operating, 0.5Fy 0.45f c 1.l(fc)'2 0.25f c 0.60Fy 0.40Fy Varies 0.66F and Seismic (O.lg) with Slen- taY derness 0.60Fy Ratio I
2. Dead, Live, Operating, 0.667Fy 0.60f c 1.467(. f c )'2 0.333f c 0.80Fy 0 .53Fy Vari es 0.88F and Hind with Sl en- toy derness 0.80Fy Ratio
3. Dead, Live, Operating, [Safe shut~own of the plant can be and Seismic (0.2g) achieved. ]

FY= minimum yield point of material f c = compressive strength of concrete Note 1: The structure was analyzed to assure that a proper shutdown can be made during ground motion having twice the intensity of the spectra shown in Figure 12.1.1:3 even though stresses in some of the materials may exceed the yield poi~t.

Rev. 3 June 1985 12 .1.2-29

  • 12.1.2.4.4.1 Applicable Codes and FSAR
1. Piping The analysis of the piping bas been performed in accordance with USAS ANSI 831.1 1967. In addition to this Code requirement, the Safety Analysis Report also requires a verification that safety- related systems remain functional in the event of a Design Basis Earthquake (DBE).
2. Pipe Supports The qualification of existing pipe supports has been per.formed in accordance with original design criteria documented in the FSAR, including AISC Manual of Steel Construction (Sixth Edition) and MSS-SP58 (1967). Additional design bases such as vendor data and other limitations not covered in the FSAR or the above-mentioned
  • Codes were reconstructed based on the existing design drawings and the understanding of the industry standards at the time. New supports were designed in accordance with MSS-SP58 (1975), ANSI B31.l (1977 Addenda through Summer 1979), and AISC Manual of Steel Construction (Seventh Edition). Standard.support components were selected to conform with MSS-SP69 (.1976). Existing elements of '

.modified supports comply with the criteria of existing supports. New

  • elements of modified supports comply with criteria of new supports .

Attachments to supports utilizing integral pipe attachments.are of material compatible with the pipe and conform to the piping jurisdictional Code.

12;1.2.4.4.2 Special Analysis Limits

1. Piping Analysls Initial Acceptance Cri teria 2 *.

IE Bulletin 79-14 requires that the as-built stress level of p1p1ng systems be determined when additional supports are required to meet the FSAR design loading requirements for the acceptance review of a given piping system. While the appropriate FSAR requirements are intended to assure safe operation after an Operating Basis Earthquake (OBE), the Initial Acceptance Criteria is based on the capability of a system to function during and immediately after a Safe Shutdown Earthquake (SSE). Initial acceptance was evaluated when FSAR allowables were exceeded, and was based on tw.ice yield for SSE. The basic criterion for initial acceptance for carbon steel and stainless 1

  • steel is:
    • 0040f 0012f

Rev. 1 June 1983 13i TABLE OF CONTENTS SECTION 13 -- CONDUCT OF OPERATIONS 13.1 ORGANIZATION AND RESPONSIBILITY 13.1.1-1 13.1.1. ORGANIZATION 13.1.1-1 13.1.2' CECO AND GE STARTUP ORGANIZATION . 13.1.2-1 13.L2.l General Electric Organization and Responsibilities 13 .1. 2-1'

'13.1.2.? Commonv1.ealth Edison. Organization and Responsi,.. 13 .1.2-5 bilities 13.1.3 *STATION ORGANIZATION/MANAGEMENT* 13 .1.3-1 '

13.1.3.1

  • Station Superintendent '13.1.3-l 13.1.3.2 Operating Assistant Superint~ndent 13 .1.3-1 13.1.3.3 Administrative and Support Service Assistar:it Superintendent *
  • 13.1.3-1 13.1.3.4 Maint~nance Assistant Sup~rintendent 13 .. L3-1 13.1.3.5
  • Operations. Department Organizatin 13.1.3-2 13.1.3.6 Rad-Chem Department Organization 13 .1.3-4 13 .1.3. 7 Technical Staff Organization. ' 13.1.3-4' 13.1.3.8 Maintenance* Depa rtme.nt
  • '1.3.1.3-8,

, .. r:L.r.4 ' ' PERSONNEL :QUAUFICAnONS 13. r.4: .*'

  • e

. . .. 13 .1.4. L . . Initi a.l Opera ting Pe*rsonhel .i3. l. 4.:.1

' ' ' . *13.1.4.2. *MininiumShift Mannf11g Requirements

  • 13 ..IA-l l'

r

' ' I 13 .2* .OPERATIONAL TRAINING* 13 .2 .1-1 13.2.1 PRE-OPERATIONAL TRAINING 13.2.1-1 13.2.2 OPERATIONAL TRAINING 13.2.2-1 13.3 STATION GENERATED PROCEDURES 13.3.0-1 13.4 PRECAUT.IONARY PLANNING 13.4.1-1 13.4.1 GENERATING STATION EMERGENCY PLAN (GSEP) 13.4.1-1 13.4.2 RAD I AT ION CONTROL STANDARDS 13.4.2-1 13.4.2.1 General 13.4.2-1 13.4.2.2 . Personnel Monitoririg 13.4.2-1 13.4.2~3 Personnel Protective Equipment. 13.4.2,..1 13.4.3 STATION. ACCESS

  • 13.4.3-1' 13.5 RECORDS 13.5.0-1
  • n.6 ADMINISTRATIVE CONTROLS . 13.6.1-1 13.6.1 AUTHORITY TO TERMINATE POWER PRODUCTION 13.6.1-1.

13.6.2 REVIEW AND INVESTIGATIVE FUNCTION 13.6.2-1.

lJ.6.2.1 Action to be Taken in the Event that a Safety 13 .6 .2-l

. Limit is Exce.eded -

13.6.2.2 Acti oh to be Taken in the Event of a Reportab 1e 13.6.2-1 Occurence r

Rev. 1 June 1983 13iv LIST OF FIGURES ~- SECTION 13, CONDUCT OF OPERATIONS 13.1.2:1 Unit 2 Startup Organization 13.1.3:1 Dresden Organizational Chart 13.1.3:3 Dresden Administrative and Support Departmental Chart 13.1.3:4 Dresden M~intenance Departmental Chart

.,a

Rev. 1 June 1983 13.1.3-1

-- 13 .1.3 STATION ORGANIZATION/MANAGEMENT Thts section outlines the .overall organization of the station ..

Stat.ion Staff is organized in acco*rdance with the Quality Assurance See figures 13.1.3:1 through 4.

The Manual~

13.1.3.1 Station Superintendent The Station Superintendent is responsible for direct management. of the station including indu-strial relations-, plann-ing, coordfoatfori, -

_*di'rectfon of the operation, maintenance, refueling and techni.cal activities.* The Station Superintendent is respons.ible for compliance with the station's NRC Operatin~ ~icense, government regul~tions, ASME Code requirements and the Qu.al i ty Assurance. Program. All reports issued by the station: to the Nuclear Regulatory Commission (NRC) *are_ the Station Superintendent's responsibi:lity. He also* authorizes the use of approved procedures* at the station, and is responsible for final approval a.nd , *

...;. distribution of station reports. -

The order to place a unit in operation or shutdown for maintenance-or refueling is issued by the Superintendent or Operating Assistant Superintendent.*

      • The Station Superintendent authorizes. all approved r.iodiffca-ti-ons -

- _ . to the* station after the issuance of an Operating_ Li.c~nse and completion,. "

  • -*of preoperationar*testing.. He forwards-'requests for.niod*ifi.cat:i.ons to the.

Station Nuclear Engtneering- D~partment. - * *-

He: provides di rec ti on for the station's on-s.ite review and' audit function as provided in the Administrative Section 6.0 of- the Technical Speci fica ti ons.

13.1.3.2 Operating Assistant Superintendent Res.ponsibility for the day-to-day operating and refueling activities*

- for the station is delegated to the Operating Assistant Superintendent.

Reporting to him are the Station Operating Engineers. * -

~.: . 13.1.3.3 Administrative and- Support Services Assistant Superintendent

. The Administrative and Support Services Assistant Superintendent*

reports to the Superintenderit and perforins various administrative duties.

and support serviCes as assigned. Reporting to him are the (1) Technical Staff Supervisor,. (2) Office Supervisor, (3)-station Security Administrator, (4) Quality Control Supervisor and (5) Radiation Chemistry Supervisor.

1.3.1.3.4 Maintenance Assistant Superintendent -

The Mai~tenance Assistant Superintendent is ~esponsible for directing

.the maintenance, including repair, of all mechanical and electrical equip-ment, including instrumentation. His responsibility includes planning work, providing on-the-jo~ training of maintenance* personnel, maintaining calibra-

,tton listings_ for majntel}ance, arranging for maint_enanc_e w_ork and_ inspection to be perfonned and initiating requisitions for the procurement of tools, materials, equipment and parts from vendors and services from Contractors.

Rev. 1 June 1983 13.1.3-2 13.1.3.5 ,operations Department Organization Operating Engineers The Operating Engineers report to the Operating Assistant Super-intendent and are responsible for the operation of the mechanical and electrical equipr.ient and certain common plant systems, such as fuel handling and radioactive waste processing, assigned to them by the Operating Assistant Superintende.nt. They are responsible for recommending maintenance for such equipment and for authorizing functional acceptance tests to be conducted by the Operating and Technical Staff personnel.

The Operating Engineers are responsible for compliance with thi 11 Limiting Conditions for Operation 11

  • Waste Systems Engine~r The Waste Systems Engjneer reports to the. Operating Assistant Super-intendent and has the responsibility for operations of an waste systems.

These responsibilities include planning, development 6f procedures and coordination of maintenance for liquid and sol id radwaste, waste water*

treatment and domestic sewage treatment.

Shift: Engineer .

The Shift Eng.in~er reports to* the Operating Assistant Superinte.ndent arid superv*ises the Shift Control. Room Engineer, Shift. Foreman~ Nuclear

  • Station Operat6rs, Equipment Operators and Equipment Attendants. ,He has the basic responsibility of.insuring the safe and efficient operation of the station equipment under his direction~

Shift Control Room Engineer The Shift Control Room Engineer reports to the Shift Engineer and directs the Shift Foreman, Nuclear Station Operators, Equipment Operators and Equipment Attendants in the perfonnance of control room operations.

The Shift Control Room Engineer will be dedicated to the concern for the safety of the plant. During nonnal operations he has overal 1 responsibility for control room activities. During transient and accident conditions he will serve as the Shift Technical Advisor to the operating crew.

Shift Foreman (Licensed)

The Shift Foreman (Licensed) reports to the Shift Control Room*

Engineer and has the basic function of supervising operations in the plant and assuring proper coordination between the control room activities and plant activities. He also ensures that the Shift Engineer and Shift Control Room Engineer are properly infonned of all conditions that could adversely affect plant- operations. He supervises the Nuclear Station Operators, Equipment Operators and Equipment Attendants.

Rev. 1 June 1983 13.1.3-3 Refueling Foreman The Refueling Foreman, under the direction of the Shift Engineer or Operating Engineer, plans and supervises the work of the Fuel Handlers.

He conducts training of the Fuel Handlers as directed by the Operating Engineer and Training Department.

The Refueling Foreman performs other duties as assigned by the

  • Operating Enginee~ ..

c Shift Foreman (Non-Licensed)

The Shift For.eman (Non-Licensed) wil.1 supervise .daily Radwaste Operations under the direct. control of the Shift Engineer. He will submit Work Requests as required a~d coordfnate Operations and Maintenan~e activities .i~ Radwaste. He will plan Radwaste Operations follpwing the direction of the Waste Systems Engineer and conduct trai-ning and. super.-

~' . vision of Equipment Attendants as directed by the requirements of the training program. *

  • The. Shift Foreman (Non-Licensed) and the Equipment Attendants for

\. which he is responsible will maintain the cleanliness of Radwaste*..

. NUclea~ Statton Operator. *,,

9

. The NSO reports to the Shift Supervisor .. His bas.ic fUnction is_ to operate the* plant according to approved procedures as directed by .the Shift Supervisor.

Station policy on RO and SRO manning levels and overtime It shall be Dresden Station policy to maintain an adequate number of p~rsonnel o~ the ~tation payroll in the Shtft Engineer, Shift Control Room Engineer, Shift Foreman and Nuclear Station Operator job classific~tions such that t.he use of overtime is not routinely required to compensate for

  • inadequate staffing. Responsibility for executing this policy is assigned to the Operating Assistant Superintendent.

It is* understood that vacancies due to promotion, resignation, extended illness, unit outages or other factor.s may create situations where overtime is required to compensate for manning level deficiencies.

The. Operating Assistant Superintendent shal 1 document the reasons necessary for such overtime, including the.corrective action being taken to restore desired manning levels and receive written *approval of the Station Super-intendent for instances w~ere overtime is necessary for extended periods.

If the period exceeds three months in length, approval of the reason and corrective action by the Division Vice-President for Nuclear Stations shall also be required .

.9 i,

I

Rev. 1 June 1983

-- 13 .1. 3 .6 Rad-Chem Department Or.gani zati on Rad-Chem Supervisor 13.1.3-4 The Rad-Chem Supervisor has overall res pons ibil ity for the department and reports to the Administrative Services and Support Assistant Super-intendent. Under the direct supervision of- the Rad-Chem Supervisor are the Lead Health Physicist, Rad-Chem Foreman and Engineering Assistant.

The Rad-Chem Supervisor evaluates radiological conditions, maintains Rad ... Chem Procedures, ensures compliance with applicable regulations, directs station chemical controls, and maintains records- of department functions such as.radiatiOn surveys, chemica*l and radiochemical analysis, and training. - -

Lead Chemist The Lead Chemist supervises the chemist, an engineering assistant; and Radiation Pro'tectionmen assigned to the laboratory. He is responsible for chemical and radiochemical analysis perfonned.

Lead Health Physicist the Lead He.al th: Physicist supervises the health- physicists and two engineering assistants. He is responsible for:- revie1'ling the hea:lth physics activities of the* station. -

        • Rad~Chem Foremen*

i

_The Rad-Chem foremen supervise.the Radiation Protectionmen. The Foremen are responsible/ for assigning work and evaluating results.

Radiation Protectionme~

Radiation Protectionmen's duties include direct radiation surveys, contamination surveys, chemical and_ radiochemical analysis and training in the use-of the respiratory protection equipment.

13.1!3.7 Technical Staff Organization Technical Staff Supervisor The Te<:hnical Staff Supervisor provides technical support for plant operations, refueling, maintenance, modifications and in-service inspection*

and evaluates process. data ar:id equipment perfonnance and adequacy of -

station procedures. He is. also responsible for numerous surveillances, such as local and integrated leak rate tests, standby gas treatment system filter perfonnance and emergency core cooling system undervoltage tests and other engineering matters. -

The Technical Staff Supervisor reports to the Admi ni.strative and Support Services Assista_nt Superintendent. He makes recommendations and advises the Administrative _and Support Services Assistant Superintendent

  • with respect to quality-assurance. He has the responsibilities. and authority as described in Section 6.0 of the Technical Specifications for -

implementation of the on-site review and audit function.

Rev. 1 June 1983 13 .1. 3-5 Lead Engineers The Lead Engineer of each Unit Group, Special Projects Group and Nucl~ar Group reports t6 the Te~hnical Staff Supervisor .. He is responsible for coordinatin~ the activites of his respective group to ensure technical

.support for plant operations including, but not limited to, refueling, maintenance, surveillances and in-service i~spections.

Cognizant Engineers A designated Cognizant Engineer has the following duties and.

responsibilities:: *

a. Reviews proposed modifiC:ation, test and referenced design
  • docu~ents .and becomes fa~iliar with the objectiv~s of the project. Ensures that the modification satisfies all the requirements stated in any special reports or Nuclear Regulatory C001mi ss ion *cooimi ~en ts.

b; Reviews documents associated w*ith the project for complete_.*

. ness and authorization; If documents are not complete, -

. obtains necessary documents to complete the package. Complet~s .

modification checklist. - * *

- -**.~ - . '. . ' :

, -.. e;. ** c~ordi'nat~s-. pto}eci* between': va rrous>depart~ents-.*: or between~ .. ,.

contractor and.: CEC.o. * * * *

d. , Follows the projectclos~ly and .is prepared to.give status.

of work. * * * *

e.
  • Ensures work was done in atcordance with the approved do~uments. b
f. At the completion of the project, coordinates, conducts; and. .\

documents the required testing. -

g. Witnesses and reports* the results of any testing to *the Technical Staff Supervisor/Operating Engineer.

13.1.3.8 Maintenance Department Maintenance responsibilities are .under the direction of the Maintenance Assistant Superintendent a.nd the Master Mechanic. This depa.rtrnent is responsible for all maintenance, includ.ing instrument, mechanical, and electrical maintenance. During major maintenance su~h as that perfonned during. refueling periods.,-the maintenance staff may be supplemented by other employees.of Commonwealth Edison and/or contractor organiza ti ans as required. Where contractor organi za ti ans

  • .are used, there will be close surveillance, by Commonwealth Edison personnel, of the work
  • e

. .~ ..... .  : .' ~. . .. ' .. '* '*

  • 9* Rev. 1 June 1983 DRESDEN ORGANIZAtIONAL CHART SUPERINTENPENT SENIOR FINANCIAL CO-ORD INA TOR -----------------------~---1 I I I I I I

I CONTRACT PERSONNEL OPER. ASS IT. . ADM IN. & SUP. MAINT. ASS'T.

ADMINISTRATION ADMINISTRATOR SUPT. SERV. ASS'T. SUPT. SUPT.

PRO I SAFETY I CO-ORDINATOR I I

I I

TRAINING SUPV. I

  • r---~T---~~---~r--~- 7 ~-~r~-------1---
  • - _I* I I .1 I I . I I I I

~

~

SITE SUB-STA. CONST. J I TOUR DIR. I I

Q.A. STA. CONST.

ENGINEERS ENGINEERS INSTRUCTORS __I---.:

CLERKS CUR KS I

Figure 13. L3: 1

Rev. 3 June 1985 13.2.2-1

  • 13.2.2 OPERATIONAL TRAINING Commonwealth has the continuing responsibility to provide sufficiently qualifed personnel to assure safe and efficient operation of the separate units and the entire station. Normally, a vacancy in any grade will be filled by the candidate next in line who qualifies technically and demon-strates satisfactory performance of the job. This philosophy of personnel repl~cement creates incentive for personnel in lower job classifications to expand their ability to perform work in higher classifications. Personnel in higher job c.lassifications have a continuing responsibility to train men in the lower classifications.
  • The Operating Engineers, Shift Engineers, Shift Foremen, Technical Supervisors, Training Supervisor, and Maintenance Supervisors have overall xesponsibility for the trainirig and maintenance of proficiency of the personnel in their respective groups. Ibis training and maintenance of proficiency will be accomplished by lectures, demonstrations, drills, and on-the-job training by supervisors and foremen of the various shifts or other working groups. A rotating work schedule has been extablished under which each operating shift will be relieved of operating responsibilities on specified days and will devote such time to Unit 2 and 3 training.

The Commonwealth Edison Company Topical Report entitled "Requalification Program for Licensed Operators, Senior Operators, and Senior Operators

  • (Limited)" (revised December 1983) is a qualification program designed to assure station management that all licensed operators maintain a high level of competency. The program consists of lecturers, simulator training, and on-the-job trai~ing. Records are maintained and evaluations are made of each individuals performance. (Keets IKI Action Plan Item I.A.3.1. and*

Appendix *A, 10CFRSS, Section 2).

The Station Superintendent has delegated to the Training Supervisor the responsibility for implementing and coordinating the requalification program.

I 0042f I 0013f

14.2.2-7 14.2.2.5 Accident Analysis for GE 8x8 Fuel 14.2.2.5.1 Fuel Damage Dropping a fuel assembly onto the reactor core from the maximum height allowed by the refueling equipment, less than 30 ft, results in an impact velocity of 40 ft/sec. The kinetic energy acquired by the falling fuel assembly is less than 17,000 ft-lb and is dissipated in one or more impacts.

The first impact is expected to dissipate most of the energy and cause the largest number of cladding failures. To estimate the expected number of failed fuel rods in each impact, an energy approach is used.

The fuel assembly is expected to impact on the reactor core at a small angle from the veritcal, possible inducing a bending mode of failure on the fuel rods of the dropped assembly. It is assumed that each fuel rod resists the imposed bending load by a couple consisting of two equal, opposite con-centrated forces. Therefore, fuel rods are expected to absorb little energy prior to failure as a result of bending. Actual bending tests with concen-trated point-loads show that each fuel rod absorbs approximately 1 ft-lb prior to cladding failure. Each rod that fails as a result of gross compres-sion distortion is expected to absorb approximately 250 ft-lb before cladding failure (based on 1 percent uniform plastic deformation of the rods). The energy of the dropped assembly is conservatively assumed to be absorbed by only the cladding and other core structures.

The first impact dissipates 0.80 x 17,000 or 13,600 ft-lb of energy.

It is assumed that 50 percent of this energy is absorbed by the dropped fuel assembly and that the remaining 50 percent is absorbed by the struck fuel assemblies in the core. Because the fuel rods of the dropped fuel assembly are susceptible to the bending mode of failure and because 1 ft-lb of energy is sufficient to cause cladding failure as a result of bending, all 63 rods of the dropped 8x8 fuel assembly and all 62 rods of the 8x8R and P8x8R ...,

assemblies are assumed to fail. Since the tie rods of the struck fuel assem-blies are more susceptible to* bending failure than the other 55 or 54 fuel rods, it is assumed that they fail on the first impact. Thus, 4 x 8 = 32 tie rods (total in 4 assemblies) are assumed to fail.

Because the remaining fuel rods of the struck assemblies are held rigidly in place in the core, they are susceptible only to the compression mode of failure. To cause cladding failure of one fuel rod as a result of compression, 250 ft-lb of energy is required. To cause failure of all the remaining rods of the 4 struck assemblies, 250 x 56 x 4 or 56,000 ft-lb of energy would have to be absorbed in cladding alone. Thus, it is clear that not all the remaining fuel rods of the struck assemblies can fail on the first impact. The number of fuel rod failures caused by compression is computed as follows:

14.2.2-8 (0.5) (13,600) [ 11 11 + 17 ]

250 = 11 (8x8, 8x8R, P8x8R)

Thus, during the first impact, fuel rod failures are as follows:

8x8 8x8R/P8x8R Dropped assembly 63 rods (bending) 62 rods (bending)

Struck assemblies 32 tie rods (bending) 32 tie rods (bending)

Struck assemblies _JJ rods (compression) 11 rods (compression) 106 failed rods 105 failed rods Because of the less severe nature of the second impact and the distorted shape of the dropped fuel assembly, it is assumed that in only 2 of the 24 struck assemblies are the tie rods subjected to bending failure. Thus 2 x 8

= 16 tie rods are assumed to fail. The number of fuel rod failures caused by compression on the second impact is computed as follows:

0.19 11

[ ~2 ] (17,000) [ 11 + 17 ]

250 = 3 (8x8, 8x8R, P8x8R)

Thus, during the second impact the fuel rod failures are as follows:

Struck assemblies 16 tie rods (bending)

Struck assemblies 3 rods (compression)

T9 fa i 1ed rods The total number of failed rods resulting from the accident is as follows:

8x8 8x8R/P8x8R First impact 106 rods 105 rods Second impact 19 rods 19 rods Third impact 0 rods 0 rods 125 total failed rods 124 total failed rods 14.2.2.5.2 Radiological Consequences Based on the linear heat generation rate applicable to 8x8, 8x8R and P8x8R fuel, it can be theoretically predicted that the fractional plenum activity will be approximately one-tenth of that activity contained in the plenum of a 7x7 fuel rod. For the purpose of this evaluation, it is con-servatively assumed that the fraction plenum activity for any 8x8 fuel rod is the same as the 7x7 rod. Since each 8x8 fuel bundle produced the same power as a 7x7 bundle, the average activity per rod for the standard 8x8 bundle will be 49/63, or 0.78, times the activity in a 7x7 rod.

Rev. 1 June 1983 14.2.2-9 Based on the assumption that 125 8x8 rods fail compared to 111 for a 7x7 core, the relative amount of activity released for the 8x8 fuel is (125/111) (0.78) = 0.88 times the activity released for a 7x7 core.

Similarly, for 8x8R ~nd P8x8R fuel is (124/111) (49/62) = 0.88 ti~es the activity released for a 7x7 core. The activity released to the environment and the radiological exposures for the 8x8, 8x8R or P8x8R fuel will, there-fore, be less than 96% of those values presented for the 7x7 fuel are well below those guidelines set forth in 10CFRlOO; therefore, it can be concluded that the consequences of this accident for all 8x8 fuel will also be well below these guidelines.

14;2.2.6 Airborne Effects Over the Refueling Pool The ventilation ducts which will have the greatest influence on any*

activity released to the refueling pool as a consequence of the refueling accident will be those 16 openings located on the periphery of the refueling pool (see Figure 14.2.2:2). Those openings located around the fuel pool and dryer-separator storage pool will be of negligible value until the time that fission products have diffused from the reactor cavity to these locations, -

the time required for this diffusional process possibly being on the order of hours. Since approximately a 62-foot head of water exists between the surface of th~refueling pool and the. top of the reactor core the only activity of importance which will escape initially to the surface of the refueling poolwi.ll be noble gases. If the noble gases ~hould be _released within a-couple of feet of*theperipherar exhaust ducts this aCtivity will

_be removed within a short period of time to the reactor building exhaust. .. -

i plenum header. The radiation level in the exhaust duct will be sufficient*

. *to tr1iP the reactor building ~xhaust plenum monitor thereby causing isolation-.

of the nonnal exhaust path, initiating startup of the standby gas treatment.

system and isolating all inlet ventilation ducts to the reactor building.

As a consequence of reducing the nonnal exhaust flow rate from 1 air change per hour to 1_ air change per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the effectiv~ness of the 16 exhaust ducts around. the refueling pool is al so reduced by a factor of 1/24 thereby resulting in negligible air flow over the refueling -pool. As a consequence of this reduced flow thennal convection currents will be the controlling method for mixing the activity released -from the refueling pool to the_

secondary containment.

GOOOINGS PONTIAC ELECTRIC POWERTON GOOOINGS ELECTRIC GROVE MIDPOIN1 JUNCTION GROVE JUNCTION I

I ,. , r, t- - L. .r - -.---1 i - - - - - - - - - 1 . - - - r " l r - - . . t---t-------4.---:~:-+--..__-t r---4..---E-+* 138 KV BUS I TR. 81 I

I I

RED BLUE I

L--~}-- I I

TR.83 I

AUX.PWR. TR. 21 TR.32 345KV SWYD TR.10 345 KV TR. 81 WILMINGTON KANKAKEE BRADLEY <

RES.

AUX. PWR.~ INDUSTRIAL TR.22 CUSTOMERS JOLIET TR.12 345 KV TR 83. WILL COUNTY MAZON JOLIET I 38KV SWYD 1

FIGURE 8 .2 .I :I SINGLE LINE DIAGRAM 345KV, I 38KV SWYD S

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