ML023040322

Part 1 of 4 - Course Outline for R-304- and BWR Systems Lesson Plans

ML023040322
Person / Time
Site:	Browns Ferry
Issue date:	09/19/2002
From:	- No Known Affiliation
To:	Office of Nuclear Reactor Regulation
References
FOIA/PA-2002-0343
Download: ML023040322 (112)

Text

Course Outline for R-304-B Day-Title

--Chapter TTC Introduction 1

Course Overview 1.0-1.6 Reactor Physics, 1.7 Review 2

Reactor Vessel System 2.1 Reactor Vessel Instrumentation Systemr 3.1 Review

-3,*

Fuel and Control Rods 2.2 Thermal Limits 1.8 Review

'4 Control Rod Drive 2.3 Reactor Manual Control 5-25 Review Rod Worth Minimizer System 7.5 5

Recirculitiofi System 2.4 "Recirculation Flow Control System 7.2 Review -..

Reactor Water Cleanup System 2.8 6

Reactor Core Isolation Cooling System 2.7 Main Steam System 2.5 Circulating Water System.-11.1 Review 7

Electro Hydraulic Control Systemr_

3.-2 Condensate and Feedwater System 2.6 Review Turbine Building Service Water System 11.4 Turbine Building Closed Loop Cooling Water System 11.5 Feedwater Control System 3.3 8

Introduction to Neutron Monitoring 5.0 Source Range Monitoring System 5.1 Intermediate Range Monitoring System 5.2 Local Power Range Monitoring System 5.3

+ýs

Revision Date BWR SYSTEMS LESSON PLAN A.

REACTOR PRESSURE VESSEL AND INTERNALS B.

REFERENCES

1. BWR Systems Manual, Chapter 2.1

2.

Reactor Assembly GEK 779 Volume I!

3.

Browns Ferry Technical Specifications

4.

Reference' Card File 2.1 C.

OBJECTIVES

1. Description of the Reactor Vessel

2.

Nomenclature, Placement and Purpose of the Reactor Vessel and Related Components

3.

Flow Paths Through the Reactor Vessel

4.

Basic Concepts of Brittle Fracture in Low Alloy Carbon Steels

5.

Operational Limitations GE:IERAL DESCRIPTiOdN

1. Design Basis

a.

To ccntain the reactor core. reactor internals and the reactor ccoianz-,-cderator.

b.

To serve as a high integrity barrier against the le-:kage of radioactive materials to the drywell.

c.

0o provide a floodabie volume in which the reac-:or core can be adeauately cooled in the event of a breach in the primary system external to the reactor vessel.

2.

Description (Figure 1)

a.

Dimensions and Weights

" -veral-e- cht - 7.'

7

-5 E/:'

'ctO Zr tcsK-Zoot Co center neac noz:ie -'ance see drawing,0':,935" Inside diameter - 251" Wall Thickness Cylindrical Sections 7/16"- plus 1/8" stainless steel overlay on the interior Bottom head 7/16"-plus 1/8" stainless steel overlay on the interior (Extra thickness is needed to compensate for the CRD housing penetrations.)

Weights Reactor vessel without head -

624.5 Tons Head Only 126.0 Tons Total -

750.5 Tons Note:

The reactor vessel head is the heaviest lift for the reactor building overhead crane and is the basis for sizing of this crane.

Head weight includes 30 tons which is the weight of the head studs, nuts and washers.

b.

'Materials I)

Base metal - low manganese - molybdenum low carbond steel alloy

2)

Inside overlay - weld overlaid of austenitic stainless steel applied to all interior carbon steel surfaces of both the vessel and head.

Purcose - to cover all carbon steel surfaces to minimize zcrrosisr. which would adversely affect water clarity during refueiinc operations.

c.

'Head Azzachment ii heic on by 92 studs and nuts (more later)

d.

Vessel Nozzles (Figure 1)

1) Level instrumentation and reactor vessel vent nozzle quantity:

I a)

Purpose - provide a tap for wide range level instru

-.entation and a vent for non-condensible cases.

b)

The vent function is used during shutdown and is primarily needed following isolation of the main steam lines.

Gamma radiation from the core dis associates water into H2 and 0 which must be ven-.d to permit vessel floodih and cooldown.

2)

Head Spray Nozzle - Quantity:

1 a)

Purpose - provide for spraying cool water from the head spray mode of the residual heat removal system to collapse steam in that area to permit vessel floodingwhile shutting down the plant.

3).Spare Head Nozzle - Quantity:

1 a)

Purpose - Originally used for vibration measurement instrument leads connected to the jet pumps, shroud head and steam separators, and steam dryer assembly.

(1)

Also used for moisture carry over measurement sensing lines to test performance of the steam separators and steam dryer assembly.

b)

All the above instrumentation is usually removed during the first outage following successful completion of testing.

The nozzle is then blank flanged.

1)

Main Steam Nozzles - Quantity:

4 a)

Purpose - to conduct dry steam out of the reactor.

N) note that nozzles are not at 900 angles fron. each other.

This is due to the conficuration of tne steam dryer (more later).

E) hnstrument Nozzles - Quantity:

6 a)

Puroose - provides for the sensing of reactor water level and pressure.

6)

Feedwater Nozzles - Quantity:

6 a)

Purpose - conducts feedwater into the reactor vessel as required to reolace steam sent to the turbine in order to maintain a constant water level.

7)

Core Spray Nozzles - Quantity:

2 a)

Purpose - provides for the low pressure spraying of the core in the unlikely event of a loss of coolant accident.

8)

Control Rod Drive Hydraulic System Return Nozzle Quantity:

1 a)

Purpose - returns excess water to the reactor vessel that is not used for moving or cooling the control rod drives.

9)

Recirculation Suction Nozzles - Quantity:

2 a)

Purpose - provides water from the reactor to the suction of the recirculation pumps.

10) Recirculation Inlet Nozzles - Quantity:

10 a)

Purpose - routes water from the discharge of the recirculation pumps to the driving nozzles of the jet pumpts to provide the required core flow.

ii)

Jet Pump Instrument Nozzles - Quantity:

2 a)

Purpos-e - routes jet pump differential pressure flow sensing lines out of the reactor vessel.

12)

Core Differential Pressure and Standby Liquid Control Nozzle - Quantity:

1 a)

Puroose - to provide 'or injection of sodium pentaborate to shutdow.n the reac-.or in the event of failure of the control rod drive and/or reactor protection system to scram the plant.

To provide instrumentation taps for measurement of reactor core differential pressure.

Also provides a pressure reference for determining jet pump flow.

13)

Bottom Head Drain - Quantity:

1 a)

Puroose - to provide for draining of the vessel during construction testino and flushing.

b)

Located at the lowest point of the reactor vessel to assure cood crud removal.

c)

Connects to the Reactor Water Cleanup system to pro vide a continual flow of water ou c-f the bottom head region to prevent the accumulation of cold water.

Cold water comes from the cooling water in flow in the control rod drives and can build up in the bottom head regiona" during low core flow conditions.

e.

Refueling Bellows Skirt (Figure 1)

1)

Purpose a)

Provides a weld attachment for the refueling bulkhead bellows.

b)

The bulkhead and bellows combination provide a water tight seal at the flange area of the reactor vessel to permit flooding of the reactor cavity to permit transfer of spent fuel out of the vessel.

2)

Bellows and bulkhead will be covered in detail in the Fuel Pool Cooling System presentation.

f.

Insulation Supports (Figure 1)

I) Welded rings provided at various locations to supoort the vessel metallic insulation.

c.

Thermccouole Mountina Pads tFicure I)

"Lccatec at several locat-cns on the vesse, proper anr supoort skirt to provide for monitoring of temperature of the reactor vessel.

(Will be discussed in detail in zhe Reactor Vessel Instrumentation presentation.)

h.

.'essel Construction (Figures 1 and 2)

1) Shell Sections (Figure 1) -

A total cylindrical shell courses, each shell. course consisting of 2 or more rolled plate sections welded together at vertical seams.

2)

Shell Course No. 3 '(Figure 2) - This particular shell course has 5 Vertical welds, all located away from the nc:: l es.

3)

Top Head (Figure 2)

Consists of:

"Dollar" piece on to4 that accommodates the 3 head nozzles.

Tapered sections form the remainder cf the spherical head.

Fiange section is two semi circular pieces welded together.

4)

Bottom Head (Figure 2)

Approximately the same design as used for the top head.

Note that all welds are outside the area where the CRD and incore housing penetrations are located.

i.

Reactor Vessel Support (Figures 3 and 4)

1) Vertical Support (Figure 3) a)

Purpose - to vertically support the entire weight of the vessel, internals, fuel and moderator.

b) Consists of the following:

Vessel Support Skirt (Welded to Bottom Head)

Ring Girder Sole Plate Concrete Founded on Bedrock

2) Lateral Support (Figures 3 and A) a)

Purpose - to provide lateral support for the vessel.

Desicned to accorm-odate both seisric forces and jet forces resulting fror, the breakage of any Dipe

=.ached to the reactor vessel (e.c., a main stear,

.ine).

b)

Vessel to Biological Shield Stabilizers A

tie down points are provided on the reactor vessel at 90° angles from each other.

B tensioned soring washer stabilizers tie the reactor vessel to the biological shield.

These are basically spring loaded turnbuckles.

The tension is adjusted following stabilizer installation

v rctatinc the a-"usz,-rc nut.

c)

Biological Shield to Containment Stabilizer S*

• 0 pairs of stabilizers located at 45 angles from each other, Consists of 10" diameter pipes welded to top of biological shield and bolted to fittings on the containment wall.

Gibs between the exterior of the containment and the reactor building accommodate vertical differ ential expansion of the drywell and reactor building concrete.

Lateral support is provided in compression only, so that no tension forces are applied that will pull the containment inward.

E.

COMPONENT DESCRIPTION (Figure 5)

1.

Review Briefly go over Figure 5 and review the location and configuration of the major reactor vessel internal components.

2.

Order of Discussion.

The reactor vessel internals discussed below will be covered in the order in which they were installed in the reactor vessel during plant construction.

3.

Baffle Plaze (Figure 5 and Drawing I0l-R935)

a.

-!sc :-nown as diffuser seas rinc and shroud supoort ia-ae.

b.

-*urpose l) To orovide a mounting surface for the jet pump diffuser.

2)

To separate the recirculating suction area (downcomer area) from the core inlet plenum area (below core C,!aze area).

c.

Installation

1)

Installed in the factor, prior to shipment to the field.

-o 'he vessel,-eI

3)

Supoorted by 6 column members welded directly to the vessel bottom head (See l0aR935, Sheet 1)

-*) The support columns support the weight of:

Jet Pumps Shroud and Core Spray Sparger Core Plate Top Guide Peripheral Fuel Bundles

5)

Nots:

There are 2 access holes in the core plate, 180 from each other.

(See Drawing 104R935, Sheet 2)

'These provide access to the below core plate area during construction.

Upon completion of construction,these access holes are welded shut using cover plates.

Jet Dumps (Figures 5, 6 and Drawing 104R935)

a.

Purpose To provide forced flow of coolant-moderator through the reactor to yield higher reactor po~ver output than would be possible with natural circulation.

escripzion (Figure 6) 10 jet pump assemblies each consisting of the following:

niet Riser and Thermal Sleeve

-rasizion PieCe Welded to T7o of inlet Riser S07oz: 1es 2 'fixinc Sections 1 'racket and Restrainer Gate Assembly 2 ;iffuser Sections Having a 19" Inside Diameter at

ne Botzom

c.

installation

1) Riser Assembly a)

Thermal Sleeve On end of riser, is welded into the nozzle at the outer end (pipe attachm.en-end).

(See Drawing 104R935, Sheet 1) b)

Purpose of Thermal Sleeve To prevent overstressing the reactor vessel nozzle due to differences in temperature between inlet water and vessel wall and nozzle temperatures.

c)

Worst Case Conditions

"(I)

Normal Operation Operation of Shutdown Cooling during Plant Cool down.

(2)

Emergency Operation Injection of cold water from the Low Pressure Coolant Injection System.

d)

Riser Brace Arms (Drawing 104R935)

(1) Purpose To support the uoper end of the riser and provide for the vertical differential expansion between the riser and reactor vessel during heatuD and cooldown.

(See &rarwinc i0"-R935, Sheet 1 and 2)

To provide adequate res:raint :o accommoda:e riser vibration which can occur under certain modes of recircuiation system operation.

(2)

Two Sets of Brace ArMs Tuning Fork Type Single Bar Tyoe (3)

Tuning Fork Type Welded to vessel wall and bracket which is welded tc riser.

(L) Single Bar Type Added for additional restraint of riser.

(Newer plants have a different design riser brace arm system. )

Located just below the tuning fork brace arms.

e)

Reason for Risers Used to permit lowering of the Recirculation inlet nozzles to get them'out of the active core region so that they don't receive any significant fast neutron exposure which could change the mechanical properties of the materials.

2)

Diffusers a)

Adaptor (1) 6" high, 19" diameter adaptor is first welded to the baffle plate.

(2)

This is done to make the welding and orecise alignment of the diffusers easier.

(3)

The diffusers are then aligned and welded to the diffuser adaptors.

(L) Upper end of diffuser has 4 guide vanes and a slid fit joint to accommodate the mixer sections.

r-,

Guide vanes -rake it e*asy to chance cut i*.er section from the reactor flange area.

3),ixer Sections a)

Slips into upper end of diffusers.

b)

P.eszrained at midooinz by the bracket and restrainer cate assembly.

(See Drawing 104R935, Sheet 2) c)

Eazh slide of the bracket assembly consists of:

2 restrainer cates that can be unbolted to swina to :er it

,OT.

0

11 -

2 adjustable bolts 1200 from each other that are set to provide the proper lateral restraint of the mixer section.

A wedge assembly that is spring loaded in the down ward direction holds the mixer section against the above 2 adjustment bolts.

(more on purpose of this assembly later.)

d)

A "D" handle is provided on the wedge assembly so that it can be lifted-to permit closing of the restrainer gates with the mixing assembly in place.

4)

Nozzles a)

Nozzles change the direction of the driving flow 1800.

The nozzle end has the flared opening to accommodate the plenum water as an integral part of the nozzle.

b)

Bolted down at upper end to the transition piece.

c)

The lower end is secured to the upper cylindrical shaded mixing section by a clamp.

d.

Purpose of having a slip fit at the diffuser and mixer sections.

1)

The inlet riser enters the reactor vessel at a hicher elevation than the bottom of the diffusers which is permanently-welded to the shroud support (baffle plate).

2)

There will be different vertical crrcwths during reactor heatuc and ccoldown cf:

Stainless Steel Riser Assembly Stainless Steel Diffuser Assembly (Different Length than Riser)

Carbon Steel Reactor Vessel

3)

Therefore an expansion joint is recuired to acco*modate the net differential expansion of these components.

e.

Purpose of having a bracket and restrainer gate assembly:

1) A ohenorenon known as fretzina corrosion has been observed s~ainiess steel in re=c-or en.i-

-:,z 2".re:zn a -corrosion is orimarI ly vibration induced.

12 -

3)

A bracket and restrainer gate assembly is emoloyed to assure there is no vibration between the mixer to diffuser slip fit.

4)

The combination of the bracket, restrainer gates, wedges, etc., assures maximum stiffness of the mixer section and prevents any vibration from occurring and hence fretting corrosion does not occur.

f.

Flow Paths and Basic Operation

1)

Drivina water enters the inlet riser, the transition piece, and its direction is turned downward by the nozzles.

2)

The nozzles increase the velocity of the driving water and at the same time lower its pressure.

3). The lower pressure in the upper end of the mixing section draws the water in the plenum and the two flows are carried downward together and mixed in the mixing section.

4)

The diffuser section slows the flow and thereby increases the pressure of the fluid.

5)

The higher pressure water exits the lower end of the jet pumo and into the lower plenum where it then passes up through the core.

c. Jet Pump Principles of Operation "Will be covered in more detail during the Recirculation System presentazion.

h.

jet Pumrp Heicht Consideration See discussion under G.2. Core Floodability

i. Jet PuTp Flow Sensing Lines (Drawing 104R935, Sheet 2)

1) Purpose To orovide a mechanism for measuring flow in each jet pump.

2)

Every Jet pump has an instrument line taped into the U~rer end of the diffuser.

3) Ten lines (all lines for one recirculation loop) are fed out of the vessel through a single penetration.

Other loop has identical conficuration.

5.

Core Shroud (Figure 7 and Drawing 104R935)

a.

Purpose

1) Divides the downcomer flow from the core flow.

2)

Provides lateral support for the core plate and top guide and hence -lateral support for the fuel bundles.

Note:

Vertical support of the top guide, core plate and all fuel is provided basically by the vessel bottom head and vessel support skirt through the shroud.

b.

Description I)

A 2" thick stainless steel cylindrical assembly welded in tne field to the upper lip of the baffle plate (See Drawing 104R935, Sheet 1) extending up beyond the top of the jet pumps and fuel.

2)

Includes the core spray spargers Insta'lled upon completion of -et pump installation

c.

Closure SurfacE 2pDer surface is machined to orovice a leak tiaht fit a.:iz:

the shroud read.

2' a8 sets of lugs provided for securing shroud head to tne shroud.

2, 2 a!icn.ent brackets, each having 2 holes, serve zhe following purposes.

a) inner Hole Accmmodates short alignment Din on the shroud head thereby assuring proper shroud head azimuth alignment with the shroud prior to boltuD.

b)

Outer Hole Provides a slip fit restraint for the lower end of the two shroud head guide rods (See Drawing IOAR935) c)

The upper end of the shroud head guide rod is secured by a bracket welded to the side of the reactor vessel.

d)

The purpose of the shroud head guide rods is to aid in guiding the shroud head into the proper place on top of the shroud to assure proper boltup.

6.

Core Spray Spargers (Figures 7 & 8)

a.

Purpose To provide 2 redundant spray networks that will yield a spray pattern covering the entire top of the core in the event of a loss of coolant accident.

b.

Description

-1) Consists of spargers with nozzles and the supply piping.

2)

Spargers are permranently mounted inside the upper part of the shroud.

3) 2 elevations of spray headers.

4)

Each of the 2 eveltions will provide 1001 of design flow and spray coverage.

5) Eacn elevation consists of 2 individual l3 0 Csections.

E! Each soray heacer ise:urea zo the shroua inner wall by several brackets which weld only to the shroud.

This leaves the piping free to contract when the Core Spray syszem injects ccid wa:er into the hot reactor vessel during accident conditions.

7) The noz:le angles are alicned ouring orecoe-ational testing to assure proper spray pattern and are then tack welded in place.

3) The supoly piping is arranged as shown on Figure 8 to accommo date siving cortraction during injection of ccld water into

-E W

-a=" Vessel.

7.

Core Plate 1,04R935, Sheet i and Figure 9)

a.

Purpose i) Provides vertical and lateral support for the 24 peripheral fuel bundles.

2)

Provides lateral support for the control rod guides tubes and hence lateral support for the fuel support castings and fuel bundles.

3)

Note:- Vertical support of all fuel exceptthe peripheral fuel bundles is provided-by the fuel support castings, control rod guide tubes and the bottom head of the reactor vessel.

4)

Acts as a partition to force the majority of the moderator coolant up through the fuel bundles rather than outside them.

b.

Description i)

Location Bol:ed to the lower shoulder of the shroud.

2)

Consists of:

Thin Stainless Steel Top Plate Stiffener Plate Members Below the Top Plate Tie Rods to. Cross Brace the Stiffener Mlembers 18E Ccntrol Rod Guide Thbe Holes 185 Z.Iicnment rains for -ssuring Prooer Control

--.od Guide Tuze and Fuel S p:ert Castling rien:at-4on 2, Peripheral Fuel Suocor: Pieces 55 :ncore Guice Tube Holes 12 Guide Sleeves for SRF

& IPR Tncore Guide Tubes

,,oze:" These uide se
eesU are only. provided to provide continuity of the cross-tie rocs which happen to run through the SRI! and IRIM locations.)

I' Neutron Source Location Holes Bolts to Hold the Core Plate to the Shroud

c.

Installation

1) Sequence is installed in the field following shroud installation.

2) Ai cnmert Aligned using 4 alignment pads welded to step in core shroud so that centers of the control rod guide tubes are directly over the center of their respective con trol rod drive housing penetrations in the reactor bottom head.

8. Top Guide (Drawing 104R935 and Figure 10)

a.

Purpose

1) Provides lateral support for the unper end of all fuel bundles.

2) Provides lateral support for upper end of the neutron monitoring instrument assemblies (Source Range, Inter mediate Rance and Local Power Range Monitors).

3) Provides lateral support for upper end of the neutron sources.

b.

Description

1) Box-like structure of stainless steel plate set in a step at the top end of the shroud.

2)

Each central box openino accommodates 4 fuel bundles and one conzrol rod.

(This is the definition of a fue: cE 1.)

3-There are 2L' ocenings near the outside, each of which accC-7o.ate one of the 2a peripheral fuel bundles.

4) Shallow no:ches are cut in the tops of the cross members.

This was to accommodate poison curtains which are no longer used.

5)

Each nuclear instrument assembly and source holder is suoported by a cutout on the bottom side of the core plate at the junction of the cross member (See Ficure 10).

9.

Incore Housings and Guide Tubes (Drawing IlOR935)

a.

Definitions Incore Housing:

That portion starting at the weld (shown on Drawing 104R935,.Sheet 1, coordinates D-27 and ending at the flange surface at coordinates D-32.

Incore Guide That portion starting at the weld mentioned Tube:

above and terminating at the core plate.

b.

Purpose:

"1) Is an extension of the reactor vessel which provides for mounting of the incore nuclear instrumentation (Source Range Monitors, Intermediate Range Monitors and Local Power Range Monitors) and convenient bottom leadout of their electrical cables and mechanical drives.

2)

To prevent jet pump flow (core flow) impingement cn the nuclear instrumentation assemblies in the below core plate area and thereby eliminate possible vibra tion damage to these assemblies.

c.

Description

1) Housing and guide tube are 2" OD stainless steel pipe.

Housing has'bolting flange at bottom end.

2)

Guide tube extends uoward and terminates ir, a slii fit in the core plate 0.5" below the too surface of tne ccre plate.

3)

OuantitV:

43 for Local Power Range Monitors (LPRM) 8 for Intermediate Ranoe Monitors (IRM)

A for Source Range Monitors (SR.)

55 Total

4)

Assemblies are located in the water gap between fuel cells (See Drawing 10!R9311

18 -

5)

SRM and IRM Dry Tubes and LPRM assemblies are loaded into the guide tubes from the top of the reactor.

The nuclear instrumentation assemblies extend upward from the core plate and are secured by the under side of the top guide.

SRM and IRM detector are installed from below the vessel.

d.

Installation

1)

Incore housings and guide tubes are orovided to the field as separate units.

"Reason:

There is inadequate clearance under the reactor vessel to accommodate the overall length if they were fabricated as a one piece unit.

2)

Housings are aligned with their respective opening in the core plate and welded directly to the reactor vessel bottom head (See Drawing 104R935 for vessel'penetration details).

.3) Incore guide tubes are welded to the housings upon comple tion of installation of all incore housings and control rod drive housings.

) 'Cross Bracing 900 box-like cross bracinc is installed thereby tying all the incore guide tubes together (See Drawing I0A,935)

Bracing is to eliminate vibration of the guide tubes and nuclear instrumentation assembiies.

"-) Co^inc Holes

- cooling holes 1/"" diameter are driliec at 900 from each other near the lower end of the guide tubes used by LPRM irnstru-ents.

Purpose:

To provide a ccoling Dath for the nuclear instru mentation assemblies.

iO.

Control Rod Drive Housings (Drawing 104R935)

a.

Purpose r

r (J

19 -

2)

Provides both vertical and lateral support for the drives.

3)

Transmits weight of the fuel, fuel support casting and control rod drive guide tube to the reactor bottom head for support.

b.

Description

1) 14+ feet long, 5" ID stainless steel

2)

Flange at Bottom is for:

a)

Permanent attachment of the CRD hydraulic system insert and withdraw lines.

b)

Bolting of control rod drive mechanism.

3)

Keyway at the flange end is provided for locking the control rod drive thermal sleeve and preventing it from rotating.

c.

Installation

1) Housings are raised into position from below the reactor vessel.

2)

Housings are optically aligned through their resoective core.late openings (hence housing installation cannot begin until core plate is installed).

3) eeloed to an Inconel Stub Tube Stub tube allows easie-weldinc and alicnment of the

-ou3incs tnan *.ould be :Icssible if housincs ;ere zc

e welded to the reactor directly.

Tcp surface of the housincs are all at the samie elevation.

]i.

Control Rod Guide Tube (Drawing 104R935 and Figures 12, 13 and i")

a.

Purpose

1) To provide lateral support for the control rod blade velocity limiter.

2)

To transmit the weight of the fuel and fuel supoort

a:-incs to the reac:cr oot-*- nead via the c:r.:rc,

,'r. -ri;- nousing.

b.

Description (Ficure 13)

1) 1" diameter by 13+ feet long thin wall cylinder

2)

Top has four 3" diameter holes to pass the core flow from the below core plate area into the fuel bundle.

3)

Bottom end is machined to mate with the CRD housing.

c.

Installation (Figure 14)

1) Guide tube is lowered through the core plate from above,

• aligned with the core plate alignment pin, and set on its CRD housing.

2)

CRD thermal sleeve is 'inserted upward into the CRD housing from below the reactor, engaged with the guide tube and rotated to lock the guide tube in place.

A3)

A key is inserted in the lower end of the CRD housing to orevent the thermal sleeve from rotating and releasing the guide tube.

12.

Orificec Fuel Support (Figure 15)

a.

Durpose

1) Provide lateral alignment for bottom end of fuel assemblies.

2)

Transmit weichz of fuel to the control rod guide tube and down :o the reactor botton head.

"o---

T.e,.,ei.,: o-the =-el is not carried ov the core oi ate.

2 Thrcucr, orifices, control the amount oc flow to each fuel

undle.

b.

Description (Figure 15)

A

Lobe Stainless Steel Casting

2)

Supsorts 4 Fue! Lundles

3)

ndivi-.4ual Orifices and Flow Paths for Each Fuel Bundie Slii

'-:s into w.e "ontr:l roc cuide tube.

2)

Alignment dog fits over pin on core plate to assure that orifices in fuel support are in line with openings in the control rod guide tube.

13.

Assembled Drive Line See Figure 16 for overall configuration of the completed installa tion.

W.

Peripheral Fuel Support (Figure 17)

a.

Purpose

.1)

Provide vertical and lateral support to the 24 individual peripheral fuel bundles that are not part of complete 4 bundle fuel cells.

2)

To properly orifice the flow to the peripheral bundles.

b.

Description

1) Body of support piece is permanently fixed into the core plate.

2)

Separate removable orifice can be changed out with special handling tool from the reactor flange area if necessary.

15.

Orificina of Flow Through Fuel Bundles

a.

";hy Orifice

!) Assume a core that has no flow orificinc.

21 In a SW,.,.,

as power is increased, the amount of boiling (two phase flow) within a fuel bundle increases.

"2

• 1ore coclin-is necessary `or a hither power bundle (cne witn more boiling) tnan a lower power bundle.

a)

Peripheral bundle, for example, run at approximately half the power of the central region bundles.

)

.s t;.o phase flow increases within the higher power bundles, increased resistance to flow occurs.

71.1S..'cuId -end -c rer-ute f'c,:

c o :ne I cwer :ower "f-el S.m...-rE--

trereb, tenoiný --c starve -:hE n i*-,er ccwer pur rooes.o ThiS is precisely w-nat 'you don't wanz -.o hiappen.

6)

By providing flow orificing in the fuel support oieces, the majority of the pressure crop across the core is taken at the orifice.

7)

The pressure drop across the orifice is large in compari son to the pressure drop in the fuel bundle itself and consequently, any changes in two phase flow within indi vidual fuel bundles causes insignificant changes in flow patterns between high and lower power fuel bundles.

b.

Flow Orificing (Figure 18)

1)

The core is divided into 2 orificing zones.

Central Region Peripheral Region

2)

The peripheral region consists of bundles near the outside of the core that are supported on regular 4 bundle supports as well as the individual peripheral fuel supports.

3)

Orifice Sizing:

Central Recion 2.3" diameter Peripheral REgion 4 Bundle Support Orifices 1.4" diameter Peripheral Support Orifices 3-3/4" diameter holes

16.

FeedW~ater SDarCers (Drawing 104R935)

a.

Purpose Te eveny ds-:ribuze feedwater to tne fet u-.ps and recircula

',r suctiOns in A manner such tnat:

1) Coic water does not impinge upon the reactor vessel walls and

2)

The core flow up through the core is properly mixed and of consistent temperature.

b.

Description (Drawing 104R935)

1) Six -,o Spargers 2:

Each has a ther:*al slee"ve weke tO the sparcer and Slid

.- 'i

-o vr

."esse!

rc::==e.

Thermal sleeve is needed to -revent excessive thermal stress in the feedwazer nozzles which would occur if cold feedwater-came into contact with the hot nozzle.

3)

Two rows of holes provide for even distribution of the cold feedwater throughout the annulus.

c.

Installation

1) The thermal sleeves are slip fit into their respective feedwater nozzles.

2)

Both ends of each sparger are secured to brackets on the reactor vessel wall.

3)

Spargers are removeable anytime in plant life.

4)

Because of sparger vibratidn problems, the sparger thermal sleeve is often welded or presses fit into the reactor vessel.

i7. Standby Liquid Control/Core Differential Pressure Piping (Figure 19)

a.

Purpose

1) Inlet and distribution line for sodium pentaborate.

2)

Provides for measurerent of above and below core plate pressures and hence core differential pressure.

3)

Provides for measurement of lower plenum (below core plate) pressure for use in jet pump fl-ew measurement.

.rovies input to core sDray 7ine tre-k detection instrumentation.

b.

DescriDtion

1)

A pipe within a pipe.

2) inner pipe connected to the Standby Liouid Control System and senses below core plate pressure.

3)

The use of spray line break detection, core differential pressure and jet pump flow taps will be discussed in detail in later lesson olans.

S.

.sta la t4 -in i

Permanently Installed

18.

Vessel Bottom Head Drain (Drawing INR935)

a.

.rpose

1)

Drain Functions as a low point drain during construction flushing of the reactor vessel.

2)

Crud Removal By being connected to the Cleanup System suction, a con

• tinual outflow of water is maintained to remove crud from the bottom head area and prevent its buildup and resultant radiation exposure problems to personnel in the below vessel area.

3)

Cold Water Stagnation Prevention By being connected to the Cleanup System suction, a con 7inual outflow of water is maintained to assure that cold water (from cooling water inflow from the control rod drives) does not stratify in the bottom head area under low core flow conditions.

.ny cold water accumulation in this area with.the remainder of the reactor vessel hot would cool down the bottom head area.

When core flow is increased, sweeping out the cold water and reolacino it with hotter water, an uncontrolled heatup ozcurs in the bottcm head area ezhich cculd ove' stress the C-Ub tube and reaz:~r vessel skhrt welds.

B 3ottom Head Temperature '.leasuremient A tnermocouple is attached =* the cumside of the 2" drain line.

Wall thickness of the 2" pipe is a small fraction of the vessel wall thickness and a more accurate measure ment of bottom head tempera.ure can be obtained than by placing of thermocouple on the vessel wall.

b.

Description

• 2" nozzle at the low point of the reactor vessel bcttc-head.

a.

ur:)ose 7- :lose o. the ccre cutlet so that a]l moderator and szear

"-.:rced

-toucr

--e s-,ear serators.

b.

Description 1, Consists of Shroud Head Stand Pipes Steam Separators Vibration Crossbracing Tie Down Lugs Tie Down Bolts Bolt Support Rings

• 2) Assembly has to be removed for refueling.

c.

Installation

1) A strong back attaches to the 4 lifting eyes.

2)

Lifted by the reactor building overhead crane.

3) 2 guide rods 1800 from each other secured az the upper end by brackets welded to the vessel wall and at the lower end by brackets on the upoer edge of the shroud provide an easy method of remotely guiding the shroud "head into piace during installation.

-')

2 mating guide brackets on the shroud head (Figure 20) fit over the aoove 2 cuide rods.

This assures that the bol-s and both sets of lugs are Properly aligned.

5 The shrcud head is attached :c the shrouc with Iona s tr,7 sOUeeze:znee zne jcs orn the shrzuc anc shroud head (Figure 21).

2 The bolt consists of The following parts:

nconel Bolt Has a "T" hook at the bottom to fit under the lugs on the shroud periphery.

Through use of the rotating dog at the top end, the bolt can be rctated a -naxin.um of 900.

Stainless Steel Sleeve

'_cs l

~-izs over inc:ne' 5oi-.

-;,: i',:

ber=

e-er..,ear sheulders on tne sleeve ocose fit inzo tne -,;o circular Dolt cuioe rincs on the shrcud head

'See

-i:ure 2).

These cuide rires.rovtde la-rai 60ttom end has flat "T" shaped surface to mate with the top of the lugs on the periphery of the shroud head.

This orovides the vertical suoDort for the bolt and Dre vents the sleeve from rotating.

Bottom end h'as a cut out slot that permits rotation of the inconel bolt 900.

Indent in bottom of cut away accepts pin on inconel bolt and assures bolt is in position to slip between the shroud*

lugs when the shroud head is lowered.

-Tensioning Nut Threaded onto inconel bolt.

Pushes downward on the stainless steel sleeve while at the same time pulling up on the inconel bolt thus squeezing the lucs on the shroud and shroud head together.

Keeper :lut Sprinq loaded in the uoward direction.

Keyed into the stainless steel sleeve to prevent rotation.

PurPose is to prevent loosening of the tensioning nut once it is tightened.

Stco u

Is tack welded in place once it is ad.usted to vrovi_':

-or zne prcper arount of zac<inc :ff of tne..,s-.oni**. nu:

necessary to olace the inconel bolt in its down and 90' rota:ed position.

P-revents accidental removal and croDopace of the tensionir.c nut into the reactor vessel.

7)

Once zhe shroud head is set in place on the shroud, Zhe shroud head bolts are rotated 900 so that the inconel bolt head will engage the shroud lugs.

8)

Tensioning nuts are torcued to 50 ft. lbs.

This promotes cn:-' partiai tensionino of the bolts.

9)

The balance of the tensioning is provided when the moder ator heats up the bolt and sleeve during plant heat up.

10) The stainless steel expands more than the inconel bolt thus squeezing the shroud and shroud head lugs together harder as the reactor temperature increases.

11) Partial tensioning through use of dis-similar metals is used to make it easier and quicker to bolt down and remove the shroud head.

d.

Precautions "1) Since the shroud head bolts are tensioned only partially when the vessel is cold, the recirculation pump cannot be run at full speed during preoperational testing until the moderator temperature is >1400 F.

Otherwise the shroud head will partially lift off of the shroud.

20.

Steam Separators (Figure 22)

a.

Purpose To increase steam quality from 10 -

13c at core exit to at least 90".

b.

Description j) StandpiDe T-sr~iz: steam -

liquiý -ixzure zo :he moiszure sea=ratcrs.

2)

Separators a)

Centrifugal Type Separator b)

Permanently Welded to Standpipes c)

Cross bracing promotes a ricid structure and prevents vibration (See Figure 20).

d)

Turning vane at inlet imparts a rotation to the in cc-.inc two chase fluid.

e, "re dense !tung

• -s znrcn :-:o the cu-:s-de by cen-ri fuCa'i force foriing a conzinuous wall of water acainst the inside of the inner ri:e.

f)

Three stages of separation return the.liouid to the downroner area and on to the recirculation oumo and jet pump suctions.

g)

Steam qualities at rated power Core Exit 101 - 131Z Separator Exit 95%

c.

Moisture Carryover I) -Definition That moisture exiting the steam separator at the top.

2)

Problems with Moisture Carryover Too much carrybver to the steam dryers will overload them with a resultant decrease in steam quality exiting the reactor vessel.

Moisture carryover is minimized in order to:

a) increase Turbine Efficiency b)

Decrease Turbine Wlear c)

Minimize Radioactivity Carried Over to the Balance of Plant.

3) d.ar Level Effect on Carryover a)

If the water level surrcunding the separators is too high, the water in the seoarator tends to backuo with resultant m~oisture carryover out the zop of the separator.

b)

Note that on Figure 23, a very wide deviation in water level is Dossible before any significant carryover results.

d.

Steam Carryunder (Ficure 23)

S) 2eflnition Tnma: s-ea-,'h-* e;..ts z*e se:arz:crs thaz is er,-a rec in the licuid seDarated from, the separators.

2)

Where M1easured At the outlet of the separators before the colder feedwater is mixed into the separated liquid.

3) Origin of Steam Carryunder Some steam carryunder is always present (See Figure 23).

Can become excessive due to running with the reactor water level too low.

"4) Problems with Excessive Steam Carryunder a)

Recirculation Pumps (1)

Steam bubbles in liquid yield a lower density fluid than would be the case if there were no steam bubbles.

(2)

A 0.2' weight percent steam carryunder means

,Q2S by volume.

(3)

If the lower density fluid reaches the recircula tion pumps, there is increased chance of oump cavitation.

(,)

M6re pumping power wiil be required to pump the leSs dense fluid due to the necessarily increased fluid velocities and resultant frictional line losses.

b)

Plant Efficiency The plant operates less efFiciently since steam that is carrieG under is not sent to the Lurbine.

c)

Core Considerations (I)

Core average void content will increase slightly.

(2)

Core pressure drop will be slightly increased due to higher two phase flow.

"i

>'nr, i

cr

• cal B I

.E r r;-

ic;..:

wi~ e s i

,hti recucec.

"Note: These considerations will be discussed in detail under Ther-mal qydraulics.

The intent here is tc ac-cualnz the stu-tent ";ith the concep. effec-s and contrcl co s~ e, c~rrvuncer.

21.

Steam, Dryer (Fiaure 24 and Drawing I0AR935)

a.

Purpose

1) To dry the flu'id coming out of the steam separators to

>99.9-quality.

2)

To provide a seal between the wet steam area (steam exiting the moisture separators) and the dry steam flowing to the turbine.

b.

Description (Figure 24)

1)

One piece assembly with no moving parts except the hold down mechanisms.

2)

Upper section consists of Peerless type steam dryers and moisture collection troughts and drain lines.

3)

Upper section has sides "cut away" to permit steam flow to zhe main steam lines.

S:,acint main steam nozzles as shown rather than at 000 ancles from each other, more steam dryer panels car. e I.uilz into the dryer with a resultant lower overa~i dryer pressure drc.

4) Lower.section consists of a seal skirt.

c.

Stear. Drver Panels (Figure 25" y-,

ee rlies ::rand Dryers - s,-!re as u iei the t.,.bir.e

,noiszure separators.

2)

Principle of C.oeration a)

Ocerates on centrifugal force.and gravity principles.

b)

'e: steam.-. ixture rises frcm the s.ear seDarazwrs throuch baffling, and is forced horizontally through the dryer panels.

c)

The wet szeam is forced to make a series of raoid

-E r-ges in direction ;hi e traversin c -the drver r: :nese traverses. noi sture is thrown tC -nE

sice wnere it is cauzch' by tne many -oisture

,,ae:tior, hooks.

e)

Dry steam of 99.9" quality exits the top of the steam dryer unit.

f)

Removed moisture drops down into collecting troughs and is routed to the outside of the dryer assembly and into the downcomer annulus area by means of drain pipes.

d.

Dryer Installation (Figure 24 and Drawing 104R935)

1) Lifted by 4 eye bolts by bjse of a strong back and the Reactor Building overhead crane.

2)

Strong back is the same one that is used to install and remove the shroud head.

3)

Proper azimuth alignment is provided by 2 dryer guide rods 1000 from each other that are attached to brackets welded to the reactor vessel wall.

L) Slots in the sides of the dryer mate with the guide rods.

5)

Dryer is supported vertically in the reactor vessel on brackets welded to the reactor vessel wall.

, Four hold down assemblies on the inside cf the vessel head prevent the dryer from lifting during high steam flow transients.

22.

-ateriais Surveillance Sample Program (Figure 2E)

a.

-urncse o -.cnitor the effect of fast neutron exposure (fluence) on the,-ech-anical properties of the reactor vessel steel.

b.

Description of Program and Hardware

1) Samples of the vessel carbon steel Parent metal, weld netal and weld heat affected zone metal are used.

2)

Sufficient samples are placed in the reactor vessel to permit measurement of changes to the vessel mech anical Droperties throuchout zhe 4O vear vessel life.

Ty,:es :f.-est s eci.r USEC:

Charny V-:;otch Test Bars - for checking shifts in ni, atures (Nil-ductility transition temperature is tnat :emoer ature below which low alloy carbon steel fractures in a brittle rather than in a ductile manner.)

Tensile Specimens

- for monitoring shifts in tensile properties (yield*

strength and ductility)

"* Flux Monitoring

- To determine the fast flux fluence received by the test specimens and vessel metal.

Consists of iron, nickel and copper dosimeter wires.

-a) Specimens and "flux monitors are sealed in leak tight metal capsules having a dry helium atmosphere.

5)

The individual encapsulated specimens and dosimeters are then placed in baskets.

Six baskets are hung on the reactor vessel wall opposite the core mid-plane where they will receive the highest fasZ neutron exposure (See Figure 26).

5) in addition, there are three sets of specimens and dosi meters that are hung from the top guide in order to get acceleratec fasz neutron exocsure daza.

7

.he s:eci.ens are,:,ec and sent tc a ho: ce*i i*i,'azcry and tested accordino to the following schedule.

Vessel Wall Soeci-ens Too Guide Soecimens 1, 2. -, 8, 16 and 32 years 1, 2 and 4 years

c.

.nstalIaticn W

,all Mounted Assemblies a)

Basket assemblies are suspended from brackets on t'e re.ctor vessel wail in 5 different Iccations.

.*r oa c- ýf naske-is spring "caec -o hcd the asse:.bly in place.

S"screw aDDroxim-ztly 1/2 way down the basket ssem~bi-is aoauszed :o bear against the vessel

-,.-Zr_- is no ':ibra-ion in" :,e

,;s, : asse~"e '..c

,r

33 -

2)

Top Guide Mounted Asse~mblies Are hung from the top guide.

F.

OTHER REACTOR VESSEL COMPONENTS AND RELATED COMPONENTS

1. Vessel Head and Closure (Figure 27)

a.

Vessel Flange

1)

Heavy flange is welded to the cylindrical portion of the reactor vessel.

2) 92 studs screw into the vessel flange

b.

Vessel Head and Flange i)

HemisDherical head fabricated in the same manner as the bottom head.

2)

Head flange has bored holes to permit slipping the head over the studs in the flange.

3)

Inside surface of the head does not have a stainless steel weld overlay (Earlier plants did have overlay on the heads).

c.

Studs, -Nuts and Washers I)

Studs screw'into a bushing in the vessel flange.

(Bushings not used on later Dlants.)

21 Zusninc is provile-' as a safa.-ez easure.

In case threads are damacec, no repair work is recuired on the reactor vessel parent material.

3)

Castellated nuts are used to secure the head in place.

4)

SDherical washers are used zo assure that the force of the nuts is evenly distributed on the top surface of the head flance.

d.

Flance Seal "0'r-imszsrc*

r 2

-c*cr, riZ

.C,

.I...s.

' 0'

,-inc mazer'.Al is siP'"r&azdIcoil

3)

Keepers that screw into the head flange hold the "0" rings cnto the head for ease of installation.

4)

Note:

Ruber "0" rings are generally used for cold vessel hydrostatic te~ts during construction since their cost is only a fraction of the metallic "0" ring cost.

e.

Stud Protectors'and Guide Caps

1) Aluminum "cans" are fitted over each stud prior to installa tion or removal of the head to protect the stud threads.

2)

Three guide caps are bolted to the tops of 3 stud protectors at 1206 intervals to help guide the head into place during installation.

f.

Head Installation

1) This is a complex process and only the basics will be dis cussed here.

2)

Following setting of the head, the stud protectors are removed and the spherical washers and nuts are installed.

)

-'easurinc rods are installed -in each stud.

--)

The temperature of the entire flange area is brouchz to greater than 100OF and stabilized.

5)

Stud tensioning devices clamo cnto the threaded stud and bear uoon the tcD surface o- -:he head flance.

E) The studs are stretches a prescribed amount as -measured by a dial nicrcreter cn top of the stud.

(Stud stretches, measurinc rod doesn't.)

7)

The nut is run down to take up the.slack and the zensioner is then backed off.

8)

This tensioning operation has to be done twice for each stud.

Once for the initial pass to seat the "0" rings Once for the 'inal tensioninc Z:

e" CreS ua:-c tiaco sUr z's

'/

cecrease the i.-;'-_a.iaz4icn ti.e.

10) Measuring rods are removed and top insert plugs are 9

ins:alled to keep out water and crud,hich coulo affect the accuracy of subsequent bolt elongation measurments.

11)

Effects of Over and Undertensioning Undertensioning - will result in inadequately seating the "0" rings and will permit leakage.

Overtensionino

- will actually "rotate" the flange sur faces somewhat so that the outside surfaces pull together and the inner surfaces ("0" ring seal area) pull apart resulting in leakage.

2.

Vessel Insulation (Figures 28 and 29)

a.

Purpose To minimize the heat loss in the primary system.

b.

Description

1) Type of Insulation (Figure 28) a)

PRefiective insulation of layered stainless steel.

b)

Fabricated in panels that are held together with snao buckets.

Ove,,:-all Conficuration Fire29';

a)

Tco nead insulation is fabricated as a single unit.

-as to be removed in order to remove the reactor vessel head.

Note: The cylindrical insulation around the top head nozzles can be removed separately to permit disconnecting the piping attached to these nozzles.

b)

Insulation on the reactor vessel nozzles is removable zc -r,* c r ccc in-_serv.'ce Jlt s=,c,.-

na M

A :1S, of -. en oze we"'ds.

c)

Cn nlants uo througn Cocper, zne insulation on tne YCyindricai oart of the vessel is not removable for sn service nr-sDeczzon as tnere is too iitzie rocm between the biological shield and the vessel to permit rermovai of the insulation, insoection and re-installa tion of the insulation.

Plants built after Cooper have the capability for in service inspection of all vessel welds.

3.

Biological Shield (Figures 3 and 30)

a.

Purpose

1) *-Reduce neutron and gamma radiation from the reactor to a)

Permit drywell access and maintenance with minimum radiation exposure to personnel.

b)

Extend the life time of drywell components such as cable insulation to the design life of the plant (prevents gamma radiation degredation of organic commounds).

c)

Prevent neutron activation of components within the dr.';ell and the resultant radiation exposure to per sonnel..

b.

Description (Figure 3)

1)

Basic Structure a)

Cylindrical structire of high censity concrete (magnetite) havinc vertical ' be=-:- supzort columns and steel outer s,,",*is 'inside and out; b)

Supported off of the reactor pedistal.

2)

• .ozzie ýccess Openings krFcdre 30) a)

Access openings are provided around the nozzles to oermit removal of the insulation for in service inspection during outages.

b)

Access Opening Components (1 Steel Gates -

.rov;de oarma shieldi*,c ni'ces CeI-it o.e.ino cares fcr (2)

Permali Shielding - poly impregnated plywood provides neutron shielding Several layers, each layer

'-l" thick Total depth -l ft.

g..

Control Rod Drive Housing Supoort Network (Figure 31)

a.

Purpose

1)

To prevent the rapid ejection of a control rod in the unlikely event of a control rod drive housing failure with the reactor at pressure.

2)

Some typical Engineered Safeguards are:

Emergency Core Cooling System Szandby Coolant Supply System Steam Flow Restrictions Control Rod Velocity Limiters Control Rod Drive Housing Supports Standby Liquid Control System

b.

escripzion:

1) Support bea[ns are founded on the inside of the concrete reactor cedestal.

2

• anzer roGs with sprinc,iashers are sus-enced frcr the Dea7~s.

3ri* cia.os, crid plates and support bars are bolted tC -he hancer rods to vertical supoort the tottom end of each housing and drive.

L) Installed to provide approximrately a 1" gap between

,oz:om, o' the drive and the support steel when the reactor is cold.

Gap reduces to -0/4" when the reactor is hot (CRD housings exoanded downward while vessel skirt exoands uoward.)

I fa

.ails. the "_- cc rz s-s e zrave=

i-n the out;.arc c.re-:icn -* 3".

G.

REACTOR 'JESSEL FLOWS AND CORE FLOODABILITY

1. Flow Paths and Flow Rates (Figure 32)

a.

Core Flow Driving flow from jet pump risers Driven flow from annular region into jet pumps Total Core Flow

1) Flows Out:

Steam flow 13.38 Flow to Cleandup System 0.13 Total Flow Out 13.51

2)

Flows in:

Feedwater Flow 13 Control Rod Drive System 0

combined cooling and return water flows 34.2 x 106 lbs./hr.

68.3 x 106 lbs./hr.

102.5 x 106 lbs./hr.

x 106 lbs./hr.

x 1O6 lbs./hr.

x16 lbs./hr.

.33 x 106 ibs./hr.

).05 x 106 lbs./hr.

T:z-zl Flow ir.

3.5i x,

l.s./hr.

note: Recirculation driving flow not included in above ficures.

2.

Core Floodability (Figure 33)

a.

Aociicability

1) Apolicable to a loss of coolant accident.

2)

The worst case loss of coolant accident is a 28" recir culation sucticn line break with the reactor at full In znis case -ne core will Decoe co.Dlezel-unccvered.

.) This will be discussed in detail during the Emergency

'cre Coolinc System presentations.

b.

Design Features

1) The Emergency Core Cooling Systems and the reactor vessel design must be compatible so that following a loss of coolant accident, the core can be adequately cooled.

-2)

There are several systems. that will provide water to the reactor following a loss of coolant accident.

"3) One of these systems is the Low Pressure Coolant Injection System (LPCI) mode of RHR.

-)

For simplification, only the LPCI system will be discussed here.

5)

The LPCI system injects water into the reactor vessel using the PRHR oumps via both recirculation inlet lines and down the 20 jet pumps.

6)

This flooding water then increases the wa-er level in the reactor starting at the bottom of the vessel and working its way up into the core.

7)

When the water level reaches the top of the jet pump mixing sections, water will begin spilling out into the downcomer area and out of the vessel throuch the broken recirculation line.

E; This elevation where wa:e:r tbecins to soill cu-.

of the jet pumps is 2/3 of the heicht of the active fuel.

93 Calculations show. that if floodin-of the reactor vessel is accomplished within a sDecifiec time frame and the level maintained at the 2/3 point, the core will be adequately cooled indefinitely and the integrity of the fuel cladding mai ntai ned.

a)

Lower 2/3 of the Core Cooled because it is flooded with water.

• ,1 t*

e i/,.',

-ne C..cre

'!icorous boiiinc in the lower 2/1 of -he core provides a mixture 3f steam. and water..hich, uDon flowing ucward cools zhe uc-er,!3 of the ccre.

P0 -

Long term, (after fuel decay heat has decreased), there wiEl be less boilino in the lower 2^3 of the core to provide the flow of steam and water to cool the upper 1/3 of the core.

Fuel clad temperature would increase with time.

How ever, it would still remain acceptable under these conditions.

10)

Note:

Under the above assumed conditions, water would have_

to be continually made up to the vessel to accommodate for

-the following cooling losses:

Boil off and, Leakage through the jet pump diffuser to mixing section slip joints in the amount of -.10O gpnm.

il N

• nte:

The above discussion serves to illustrate the princi Ple of 2/3 core coverage.

Under all loss of coolant acci dent conoitions there will be more than one Emergency Core Cooling Syster operating to reflood and cool the core.

The use of any of the Emergency Core Cooling Systems that provides water to the reactor vessel will not only provide the mirniur, required 2/3 core coverage but will provide comolete core coverage.

R "IT.-: OTHE: SYSTEM¶S (Figures 1 and 5)

'.essei.:nst*ue-~za'.cr.

Level, Dressure, temperature and flow sensed from Cer.:er Too -ead Nozzie

.r"nszrueret,Xo::~es Szandby Licuid Control/Core Differential Pressure Lines Jet Pu-o 7low Lines

-,er7-ccouoie Pacs and Bottom Head Drain Line

2.

Control Rod Drive System

a.

Control rod drives are mounted on CRD housings.

Z : *

-*.*e

-C

"=

. e S z =

--r c- -

r *. :

tCafne,,C

.,osfee

-'eturn thrruCn...... e on side of vessel.

c Fead Spray :,ozzle on Head Used to soray head area steam sDace (not,;aV~s of vessel) and vessel flooding.

3.

Main Steam System Provides steam to drive the main turbine.

Also,rovides over pressure protection for the reactor vessel.

4.

Auto Depressurization System An emergency core cooling system which functions through the pressure relief valves on the main steam lines to blow down the vessel pressure under certain emergency conditions.

Feedwater Systen Provices hich purity makeup water :-o he reactcr to replace stea.-'

sent to -ne turoine.

5.

.ic

ressure Scslant Injecticn System, (H'....

a.

-roides water to the reactor vessel to cool the core in the in"ikeiv event of a loss of coolanz acdef-.

b.

-:ive zow;er is stean from one of the -,ai.n steam lines which

r.'-eS a z-r:,ine..

c.

,.a-e e-ze's t eactor via -ne Feec sa-er

,ines.

~r:, r:

a r sDray-in

-f the core in t.e.jrlikeiy

,.de

--or I c:w ore __ure h

r - - i SCss S-" :CC:anz acczzC':..

.Recirculation Sy.szem.

,forceo ciculation of -ne reacm--r c

'an to yeied rc=r-r reactor Dower outouz tnan would be possible uncer natural circ

-*cn cc-diti'-ns.

e.

Peactor,:ater eanup Syste.

,~~-- --

---

" Cv :

] -,e

~ '

ss.r..

b.

Suction tzaken from recirculation !Ooo suction line and tcttom head draainr.

r..:h tie

.ee..ater lines.

11.

Shutdown Cooling Mode.of RHR

a.

Provides for the removal of decay heat from the fuel during normal plant shutdown.

b.

Suction taken from the "A" recirculation loop suction oiping.

c.

Returns through both recirculation loop discharge piping.

12'.,

Low Pressure Coolant Injection System (LPCI)

Mode of RHR

a.

Provides flooding water to the reactor via the recirculation discharge piping to cool the core in the event of a loss of coolant accident.

13.

Standby Licuic Control System (SBLC)

Used to inject neutron absoroing sodium pentaborate solution into the reactor tc shut it down in the unlikely event of failure of the Control Rod Drive and/or Reactor Protection Systems.

i"

-Neutron -oni:*rinc Systems

a.

Consists of:

Scurce..arae M,'onitoring System (SPM) 1nzer-.ec':*.e ?.ance :onizcrinc Sseis--er m'l, L~cac-a

=,r ar.ace ;cr, ir.E System ' L.P....

b.

Provides :or ronitoring of reactor cower under all rcdes of operazicn.

A s'v'ste, of sensors and relays which scrars the reactor uccn

osi, s whi:n re:;resen: pozentially unsafe cc-ci--ions to the reactor.

TECHNICAL SPECI-:

17:ONS

a.

Linitation The reactor coolant system pressure shall not exceed 1325 psig at any time when irradiated fuel is present in the reactor.

Note:

The reactor coolant system refers to everything attached to the reactor vessel and extending out to the isolation valves.

b.

Where M'easured In the vessel steam space (dome)

c.

Basis for Limitations

1) Based upon the more restrictive allowable pressures for Reactor Vessel and Piping and Related Equipment

2) Piping Limits a) Design Suction - 11i8 psig Discharge - 1326 psig b) Piping code allows a 20-.' overpressure during transients or:

120% x II4' = 137' psic 120' x 1326 = 1591 pslc S.

Reactor Vessel Limits a)

Designed to 1250 psig b) Boiler code allows a 10:1 overpressure during transients or, 110Z x 1250 =.1375 psig d) 1375 psig vs. 1325 psig S -ust be *su.e thet t÷*e enzt.re rec-.or coolant s

cf.a-er.. _r. -ne C.e.'--e sJoe -."n S" e.r.f OCC;*r5.

b)

Pressure is highest in the reactor coolant system at the lowest point in the system.

c)

Even though this low point in the system is the Recirculation System suction piping, it is assumed for conservation that the more stringent reactor vessel limitations apply.

d)

Since pressure is measured in the steam spare, coj nection for this static head of water (-.50 psig) must be applied.

-This is the reason for not exceedi'ng 1325 psig as indicated on Control Room instrumentation.

d.

How Reactor Vessel is Protected from Overpressure

1) Through proper operation of the reactor

2)

Through relief and safety valves on main steam lines during abnormal conditions.

3*

Through the Reactor Protection System which will scram "the plant upon receipt of signals which indicate poten tially unsafe conditions.

2.

Thermai Limitations

a.

Limitations i)

The average raze of reactor coolant tempera:ure chances dur-o normal heatuo. cr coc-,cwn shall no-exceec KC*°F.

when averagea over a one hour pericd.

b.

ehere Ieasured The following will be permanently recorded:

Steam Dome Pressure (Converted to Upper Vessel Temoerature)

Reactor Bottom Drain Tenperature Reactor Vessel Shell Adjacent to Shell Flange

• .ecirculation Loops A & B Reactor Vessel Bottom Head Temperature

• sis

=c*n Lii~

s

",F"-'./hr. Hea:up Pate a)

Provides for efficient but safe plant operation.

b)

When maintained, it assures that stresses caused by thermal transients are within those analyzed.

3.

Pressurization Temperature (Figures 34 and 35)

a.

Purpose of Limitations "To specify the minimum temperature for a given reactor pres sure for the following mcdes of operation:

In-Service Pressure Testing Subcritical Heatup and Cooldown Operations with the Core Critical

b.

Limitations

1)

During all operations with a critical core, other than for low level physics tests, except when the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve -3 of figure 34 (Figure 3.6-1 in Tech. Specs.).

2)

Durinc hec-:up by non-nuclear means, except when the vessel is vented, cooldown following nuclear shutdown on low-level physics tests, the reactor vessel temnoer atures shall be at or above the temperatures of curve

-'2 of ficure 3.

7 The reactor vessel shell temoeratures during inservice

/ycýs-a:'c or leak es.:Inc shall be at or aocve -*,e temoeratures sr.ow-n on curve =1 of figure 34.

-his curve :o -e modified by the increase in temoerature recurrerd by neutron exposure as shcwn in Ficure 3E (Figure 3.6-2 in Tech. Specs.).

) The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head is greater than lOOF.

c.

'asis for Limiations 12

-Eriztie fra.ure in rea::cor vessel r.ateria*S a;

-i ferriC-i steels.. cluding olain caron stees anc low al.cv steels can break in a bri:-ie r.anner

=- icw te,-:eratures, typically below -1 00F.

- 46 b)

The low alloy steel used in the construction of the reactor vessel fal, s into tnis caZegory.

c)

Under unusual circumstances, the temperature at which brittle fracture occurs for the reactor vessel steel can be within the temperature range over which the reactor can be expected to operate.

d)

Austenitic stainless steel alloys used in reactor components, such as the shroud, shroud head, etc.,

can break in a brittle manner at temperatures in

.the -200 0 F. to - 3 0 0 0F. range.

This temperature range is well below any that the vessel will see.

Revision Date BWR SYSTEMS LESSON P'L"N A. REACTOR VESSEL PROCESS INSTRUMENTATION B.

REFERENCES

1. BWR Systems Manual Chapters 2.2 and 3.2

2.

GEK 32555 - Nuclear Boiler System (Brown's Ferry)

3.

GEK 32556 - Reactor Protection System (Brown's Ferry)

4.

GEK 32550 - Feedwater Control System (Brown's Ferry)

5.

GEK 779 - Volume V - Instruction Manuals for Vendor Supplied Equipment

6.

Final Safety Analysis Report - Brown's Ferry Nuclear Plant 7.. Card File Index - Chapter 2.2 C.

OBJECTI6VES

1. Fully understand the purpose of the system and its design objectives.

2.

Know the types and ranges of instruments comprising the Reactor Vessel Instrument System.

3.

Learn all setDoints and trip functions associated with the various i nstruments.

Vnow the urn, erivinc reasons oehind these setpoinzs and t-,:

funct:4ons.

Learn the Technical Specifications pertaining to the system.

Become familiar with control room readcuts.

1. Design -asis

a.

Provide the operator with sufficient information in the con trol room to protect the vessel from undue stresses.

b.

Provide information which can be used to assure that the reactor core remains covered with water and that the separators are not flooded.

c'.

Provide automatic, redundant, reliable inputs to the Reactor Protection System (RPS) to shut the reactor down when fuel damage limits are approached.

d.

Provide automatic initiation of the Emergency Core Cooling System (ECCS) and the Primary Containment Isolation System when safe operational parameters are grossly exceeded.

e.

.Provide.a method of detecting leakage from the reactor vessel head flange seal.

2.

Processes measured include:

a.

Vessel Level

b.

Vessel Pressure

c.

Vessel Temperature

d.

Jet-Pump and Core Flow

e.

Vessel Flange Seal Leakage E. CO.MPONET DESCRIPTION

i.

Vessel Level Instrumentation

a.

enitions ýFýcure I:

1) Reactor Vessel Zero a)

Reactor pressure vessel bottom head invert (the top of the bottom head) b)

Provides reference for all incore comDcnents and vessel nozzle zaos 2" :nszrument Zero a) 528" above vessel zero n k..

n7 ure o l) Normal Control Range a) 0 - 60" span covering normal operating range.

b)

Referenced to instrument zero.

c)

GEMAC Instruments (3) used by Feedwater Control System (1) Temperature Compensated by a Pressure Signal (2)

Most accurate level indication available to the operator d)

Yarway Instruments used for Trip Functions (later)

(1) Not Temperature Compensated (2)

Calibrated for Normal Operating Reactor Pressure (3)

Used for trip functions because a Yan-iay instrument does not require electrical power for indication.

e) 3 control room indicators and one recorder monitor this range of level indication - all Feedwater Control System, instruments

2)

Emergency System Range a)

-155" to 60" span covering normal operating range and dow;n to the lower instrument nozzle.

b

= --=-rncs o instrtren-zer c)

Temperature compensatec via heat clamps between reference and varianle lec.

d)

Provides trip functions associated with the level i nstrur.enta-i on.

e)

Caliirated for normai operating reactor pressure f)

Two control room inaicators monitor this range of level indication a) 0" to -0.O sDan cce....ng upper portion of reactor vessel bi Referenced to inszrument zero.

C).,c te-erature comoensation (1) Calibrated for cold (<212OF.,

0 psic) moderator temperature.

A calibration curve must be used at other than zero psig to find actual level.

d)

Provides level indication during vessel flooding on cooldown.

(1)

No trips or alarms associated with this range.

e) *One control room indicator monitors this range of level indication

4)

Post Accident Flooding Range a)

-100" to +200" span covering active core area and overiapping the lower portion of the Emergency Systems Range b)

Instrument zero for this instrument is 360" above vessel zero, which is the top of the active fuel.

c)

No Temperature Compensation (i) intended for use only under accident condizions with reactor at 0 psia and re~irc :umos tripped.

(2)

Variable tap is from di ffuser c' t :JUe S and 6 (or 1 r

ard 16).

Flcw thc:h tne ":-e pu-p interferes ;,,4n zne variaoie e:- 4ina renderino instrument indication inaccurate.

d)

Prcvides indication during and after a -oss of cool.anz accicent.

t also provides a si!naI to in-ericck the Ccntaimr..ent Coolina ;oce of zhe Residual heat Removal System (RHR).

'I) Prevents using the -RHR System for :ontainmenz deoressurization -.,hen it is needed to flood tne core region.

toe -

t 0

i00": portion cf this,a n cez z s ra'_

-r,

r teA

.. -r

-".ne cE :,tesE ir-srue r-s.

c.

Level Instrumentation and Piping Layout (Figure 3)*

1) Condensate Chambers a)

Piping run is uphill from vessel to chamber.

(1) Allows excess chamber condensation to overflow back to reactor vessel.

b)

Normal Control and Post-Accident Flooding Ranges

-share a common condensing chamber.

c)

Piping runs are insulated and as short as practical.

d) Condensate chambers are unlagged.

2)

Yanray Column.(Figure 4) a)

Provides physical temperature compensation." Compensa tion required since external reference leg is colder (denser) than internal variable leg (at operating pressure).

(1)

Steam condensing in the condensate chamber keeps the reference leg full.

(2)

Piping is run downhill into condensate chamber to ensure adeouate condensate in the chamiber.

(3)

Excess condensation overflows into the variable leg (4) A natural circulation effect keeps variable leg condensate recirculating back into the vessel.

15)

As a result. the variable leg is kept not k..-,*cC.

)

(6)

Heat transfer clam.os maintain reference leg temperature at %290oF.

b) Auxiliary Chamber Operation (1

ormaily full o-F cool (<10 0°F.% water.

'2-z7n

-vent z, ran*id cecrease -n" reac.cr pressure, condensate charmber may vegin to flash, decreasing reference lec height.

s-s n,/.Y.al: t-e

'._ _ i s-r'r-hs 'wi-. "e the exce:zi c. n o,

-. e snj:cJs.z..n.

':e-s floccInc rance of.;nr,.ch trere io o'iv one i nsZruefa...

(3)

As the condensate level drops uncovering zhe auxiliary chamber part, cool water flows into the condensate chamber:

(a)

Quenching the Flashing Action (b)

Restoring the Reference Leg Level (4)

The equalizing tube allows steam to flow back into the auxiliary chamber where it condenses.

(5)

Without this action, the reference leg height would decrease, giving the appearance of an increase in actual vessel level.

Note:

The other condensate chambers do not need the auxiliary chamber since they operate at a much lower temperature

(<150 0 F.).

d.

Level instrument Functions Indicated Setpoint(s)

Type Range (Indicated)

Function Normal Control 0': to 60" NA Provides level Range inouts to te Feedwater Level Control System 51

1) Tri.s "z.in Turzý

.e

2)

Trips Reactor Feed Pu,.ps

3)

Trips Hich Pressu'e Coolanz Injec tion Turbine z' '

Tri's :e=czor Core Isolation Coolinc System rt, n

r'*,' r-r Aw

-- r.

,1 Czcn tr'

¢S...

Sezpoint(s)

(indicated)

+3911

+27"

+18 Emergency Systems Range

-155" to

-60"1

-51.5" High Level Alarm*

• Normal Operating Level

1)

Low Level Alarm*

2)

In-conjunction with RFP Trip, initiates Recir culation Pump Runback.*

1)

Reactor Scram

2)

Containment isolations

3) Permissive to Automatic Depres surizations System

4)

Start Standby Gas Treatment System

1)

Initiates the High Pressure Coolant ineczicr, SWsien (HPC)

2)

Initiates the Reaczor Core isolation Coolinc Syste7 -. C- *

3)

Main Steam Line isolation

4)

Recirculation

-urm.

TriD

-:unc:cn V Feecwazer Level Control System TzEL Indicated

_Rance Function

Indicated Ranae Setpoint(s)

(Indicated)

Emergency Systems Range

-143.5"

1)

Initiate Core Spray

Shutdown, Vessel Flooding Range Post Accident Flooding Range 0" to +400" None

-100" to

+200"

2)

Initiate Residual Heat Removal Low Pressure Coolant Injection Mode

3)

Permissive to Automatic Depres surization System

4)

Start Diesel Generators Level indication during vessel flooding Pevents inadvertent operation of contain ment coolino during acci dent condi ti ons.

I N

oze that the level setoi nzs at 4-i0", -51.5" arnd -I13,-..5.

are co.rmcnly referred to as Ic.,

low-low and !zw-Tcw-Iow vazer level respeczively.

e.

Level trip settings were chosen for the following reasons:

"Note:

All levels shown are incicated levels.

1) Main Turbine, Reactcr Feed Pump Turbine, High Pressure Coolant injection Turbine and Reactor Core isolation Cooling Turbine Trip at +54".

a)

+54" level is point above which moisture carryover becores much more pronounced.

Turbine blading may te da nace,: i f

.eI oc

'orao*

on ccn-z nued

-"ove

h-;r. s eve,. Ty2e Function

-. 9" b)

In addition, tripping all but the main turbine prevents feedwater from overflowing into main steam lines during sressurized operation.

Water "slugs" in steam line may cause hydraulic pressure surges which may lift safety/

relief and/or safety valves.

2)

High Level Alarm at +39" a)

Defines upper end of normal operating region.

b)

If level is <_+39", transients such as a recirculation pump trip will not cause level to increase to turbine trip point (+54").

3)

.Low Level Alarm at +27" a)

Defines lower end of normal operating region.

b)

If level is z.+27", transients such as a reactor feed pump trip from full power will not cause level to decrease to reactor scram point (+10") with a runback of the recirculation pumps.

-)

Reactor Scram at +10" a)

Assures the reactor will not be operated without suf ficient water above the reactor core.

b) in conjunction with the containment isolation signal of +10", prevents fuel cladding perforation.

c)

Prevents operation wi-h separator skirts uncovered.

Carryunder could resulz with:

Reduces core inlet subcooling

"(2)

Reduces jet pump and recirculation pump net positive suction head.

d)

Set low enouch to prevent spurious operation for normal operating transients 5"

Conzain-nent Isolations at i-0" a)

In conjunction wih the reactor scram settino (-a -

),

initiates closure of certain primary system isolation valves to limit inventory loss and orevent fuel cad

-"c--_i" ua 5-

-I, Se 0 *rc-,,--T to

'.ae C-.r.'.i r.IS erh S

e7.r c-tr-ani.nt.

esm....._=,lns--"

iish -

seconcary containment.

6)

HPCI and RCIC System initiation at -_3-l-5' a)

HPCI initiates at -31.5" is set to allow adeouate core cooling for small line breaks.

b)

?CIC initiation at.5 is set to Drovide adequate cooling under loss of feedwater flow and/or main steam isolation conditions.

- S7, S"

7)

Main Steamline Isolation at 6" a)

Initiates closure of Group I isolation valves to prevent excessive release of radioactive products to the environment and loss of vessel inventory.

8)

Recirculation Pump Trip at a)

Prevents operation of recirculation pumps without adequate net positive suction head.

9)

Diesel Generator Start Signal at -143.5" a)

Starts Diesels so that they are already up to proper speed and voltage in the event of a subsequent loss of normal power.

(1) If. normal power was already lost, Diesels will start and automatically close onto the emergency buses when they reach prcper speed and voltage.

10)

Emergency Core Cooling System (-Cr-)

Initiation Signal at 1-,.

") Set low enough to orevent sDurious ooeration b)

Sez hich enough to aV1zw tire

o ac-:e the I,.,

pressure ECCS so that no fuel melzing will occur.

Long term cooling will be cossible wizn no fuel eli tdown.

"?Ij 2/3 Core Covered Permissive Interlock at -30" a)

Allows Residual Heat Removal

(,HR) System to be used for conzainment spray.

b)

If core has been reflood tc at least 2/3 core heignPt, sufficient water is available to keeD core cool.

The u:cer 1/3 oF the ccren.*

l :e)s cooicC 5'as':nc

11 -

f.

Steam Flow Effect on Reactor Water Level (Figure 5)

1) Steam flowing through the dryers is forced to change direction several times, resulting in a pressure drop across the dryers.

2)

At 100% steam flow, the pressure drop is V'7" of water.

3)

Therefore, at 100% steam flow, P1 is 7" of water less than P2.

4)

The level outside the dryer skirt (downcomer region) is 7" higher than inside the skirt.

5)

Since the vessel level instruments compare the reference column height to the downcomer (variable column) height, setpoints are adjusted to compensate for this error.

6)

The water level inside the dryer skirt is slightly dome shaped.

a)

Moisture separator drains must flow' to the outside (downcomer) recion to return to the core.

b)

In order for drains from the interior separators to flow outward, a hydraulic gradient is recuired.

c)

At 100% power, the "top" of the dome is

\\4" higher than the "outside", providing the hydraulic gradient.

Note:

The degree of hydraulic gradient is variable, ranging from 0" at 0% power to 4" at 100:' power.

c. ;single overall reactor differential pressure inszru.ent

,. so utilizes the level pivina.

1)

Range 0 - 50=

2)

Used to evaluate jet pump performance during initial szayrtuo testing.

3)

No operational significance.

2.

Reactor Pressure instrumentation and Piping Arrangement (Figure 6)

a.

Ultilizes same piping as vessel level instrumentation

b.

Control Room Indications (Panel 9-5)

nstrument Narrow Range Reaczor Pressure Reactor Pressure Wide Range Reactor Pressure*

R~ed Recorder 3 Indicators Recorder Indi cation 950 - 105-psi9 0 - 1200 psig 0 - 1500 psig

c.

Pressure Instrument Functions Type SetDoint(s)

Pressure Switches (4)

Pressure Switches (2) tressure Switcnes (2; P-essure Pecorcer

?:-essure Switches ()

230 psig 450 psig 600 Osig 10,0 psig 1055 psi 9 Function Signal Reci rculation Discharge Valves to Close for Residual Heat Removal Logic

1) Permissive for opening core spray and Residual Heat Removal admission valves.

2)

In conjunction with high dryvelell pressure

(+2 psig) initiates core spray and Low Pressure Coolart Injection Inzerlocks -he !-,echanica&.,

Vacuum Pur.os if

?.eatcor Pressure >600 psig and condenser vacuum >22" ng.

,eaczor Hizh Pressure A-.arr.

reactor Scram Sicnal

-F'unctions from FWCS Pressure Recorder Type Pressure Switches (")

Pressure Switches (4)

Pressure Indicating Transmitters Setpoint(s) 1055 psig 1120 psig 0 - 1500 psig Function Automatically byDasses main condenser Io4 vacuum or main steam isolation valve closure (only If zhe M*ode selector switch is not in Run).

Initiates Recirculation Pump Trip for Anticipated Tran-sient Without Scram Provide temperature compensa tion for vessel level signal (Feedwater Control System)

d.

Pressure switch settings were chosen for the following reasons:

1) 230 psig a)

Recirculation Discharge Valves close on Loss of Coolant Accident for RHR Logic, ensures RHR Injection into Reactor Vessel

2) 450 Psig a)

Permissive ooeration.

above this pumps were for RHR and Core Spray ECCS Logic and Low pressure systens will not inject pressure even if valves were open and running.

3) 600 psig a) 7ces no! aliow opera=ion of when not needed to maintain pressures where sicnificant released.

Xechanical Vacuum Pump condenser vacuum at radioactivity could be I0c-0 psic a), Alert operator of impending high pressure condition.

5) 1055 psig a)

Reactor scram at this pressure prevents possible violation of the Nuclear Steam Supply System pres slret

14 b)

Reactor scram, in conjunction with relief and/or safetv valve operation, limits pressure transient experienced during a turbine trip at high power conditions.

6) 1055 psig a)

Startup testing on KRB revealed possible core insta bilities during operation at >600 psig and main steam isolation valves closed.

b)

'Recent tests indicate this interlock may be unnecessary, setpoint has been raised to allow Hot Standby Operation (subcritical, MSIV's closed) at normal operating pressure.

7) 1120 psig a)

Trips recirculation pumps, in the event the RPS scram function fails, tripping these pumps will reduce core power to within the safety and safety relief valve capacity.

3.

Vessel Temperature Instruments

a.

Provides indication of vessel shell, flange and support skirt temperatures.

b.

Allows operator determination of temperature gradients and the resulting stresses.

c.

Such monitoring is especially imDortant during heatup, cooldown and transient conditions.

. co*-:r-ccns'Pn'an zhermocouoles with b.raided -lass insula-:ion and stainless steel cladding are positioned on vessel.

'K Thermocouole Pad Locations (Figure 7)

Vessel Suppor:- Skirt b)

essel Head Stud Bolt c)

Vessel Flange d)

Feedwater Nozzle (3)

15 -

e)

Vessel Shell (Above and Below Normal Water Level) f)

Above Vessel-to-Skirt Junction g) Vessel-to-Skirt Joint h)

Bottom Head i)

Vessel Drain Line to'Cleanup System j)

Vessel Flange-to-Shell AT

2) 32 additional thermocouples are installed on the vessel and hooked up to the junction box but do not readout in the control room.

3)

Thermocouple Installation a)

Most consist of two adjacent pads welded onto the vessel during fabrication (Figure 8).

(1)

One Dad has a thermo~ouple sheath clamping screw.

(2)

The other holds the hot junction of the thermocouple against the vessel surface.

b)

Thermocouples measuring the vessel head and head flance areas are maonetically clamoed.

S1)

Easily removed durina vessel head removal.

(2)

Can be positioned for optimum temperature measurement.

c)

Another thermocouDle is inserted into a drilled cut vessel head bolt.

e.

The vessel shell-to-flancelT is rreasured to limit thermal stress on the shell-to-flange weld.

I)

The flange is very massive and changes temperature slower than the shell.

, -.-, s e neatu, and coc' oc :r'n raze ar.a

- r. s

.re,:t q er:"u,-' and Ccre Flow Instruments

. it oower'outDuz is oroportional to the abilizy to. recve

.ea::er er

c:re c-:.w jrat' are,-ecuied zo evaluate reactor,cw.;er ie've.

16 -

b.

Since the total flow of coolant must pass through the jet pumps to reach the core inlet plenum, flow is measured in each jet Dump and summed to yield total flow.

c.

Each jet pump has a pressure tap on the pump throat.

This pressure is compared to the core inlet plenum pressure to nroddce a differential pressure signal proportional to f1! I.

i) Core inlet plenum-pressure is essentially jet pump discharge pressure.

2)

Individual jet pump differential pressures are indicated in the control room.

d.

The square root of the differential pressure is a signal representing flow.

e.

Core Flow Measurement (Figure 9) l) Flow for jet pumps l through 5, 6 through 10, 11 through 15 and 16 through 20 are summed.

2)

Sumned flows for 1 through 5 and 6 through 10 are again summed to give "A" recirculation system jet pump flows; and similarly 11 through 15 and 16 through 20 to give "B" 1o op flows.

3 These loop.flows are summed to give total core flow.

"4 In the event one recirculation oumo is secured, that loocs jet pumos will have reverse flow, and the inactive ioco's,lowto give a -., re accuraz-e core flow indication.

5)

Total core flcw is recorded in the control room.

f.

Fully Instrumented Jet Pumps

1)

Four jet pumps were calibrated at a test facility and then installed, one per quadrant.

2)

The other 16 jet pumps were then calibrated against them.

3)

The fully instrumented jet pumps have additional pressure taps.

) re on -:ne our zn

.a:

b) 1 ne on the pump dif-user F-,c',Y nc-.cation is iaye.

a n :ontrzi

.cc....

17 -

d' Jet pumps 1, 6, 11 and 16 are fully instrumented.

5)

The lower taos on the instrumented jet Dunps also orovide

,he variable leg signal :o the post accident flooding level indication.

g.

ADnormal differences in flow indicat-ons between jet pumps

.Pay be indicative of an inoperative jet pump.

Refer tc the Recirculation System Lesson Plan for more information.

h.

Core Plate Differential Pressure

1) Compares standby liquid control injection line pressure (below core plate) to above core plate pressure.

2)

-May be used to determine long term trends in the indicated core flow and core plate differential pressure relationship.

5.

Vessel Head Flange Leak Detection (Figure 10)

a.

Detects leakage from the inside of the reactor vessel past the inner seal ring.

b.

The detection line is connected to a drilled passage in the vessel flange.

c.

Any inner seal leakage is piped to a collection chamber installed between 2 air operated valves.

d.

A level switch detects accumulation of water in the chamber.

e.

A pressure switch detects abnormally high pressure in the collection piping.

-. clenoid ocerated control valves are ce-enercized in the monitor

=csizicn.

o. Placing the switch in the drain position energizes both control valves, reversing the position of the isolation valves.

I This blocks the leakace and drains the detection circuilt to the Drywell Equipment Drain Sump.

h.

eturn the swizch to Normal.

i.

The leakage rate may be determined by timing the perioc required to activate the level alarm again.

c.:e:...iv

._Z*.rCPC,

-hat c-er2-4 in C.a.- D sc.is.na va,'e s b, a"OiGec 3nze eo:-cee th.... n the firs-seAl is detected.

Opera-ing exoerience has shown that the amount of steam leakace Dast the first seal increases after each coeration of the collecticn cnarter fill anc drain valves..reich will wire cr-a flance seaiilc surface.

j.

Failure of both flange seals is detected by the primary containment leak detection system.

k.

Note that it is possible to fill the sensing line with water turing refueling operations.

If the line is not drained prior to heatup, an erroneous inner seal leak will be annunciated.

F.

OPERATIONAL

SUMMARY

-1. Protective System Philosophy

a.

Redundant sensors employed to. reduce probability of system failure.

.b.

Trips or protective actions are initiated from level or pressure switches only.

c.

Switches serve but one function and that is to transmit trip

-functions.

d.

All signals required for remote (control room) level and pressure indication/recorders are completely independent of the protective circuitry.

e.

Failure of:remote recorders or indicators will not prevent proper Drotective system response.

EM 4;1 ZNTERRELATTMONSHI PS

1.

Level

a.

NJormal Control Rance Inputs to:

1) Feeowater Control System

2)

Main Turbine Trip Control Circuit

3)

Peactor Feedwazer Pump Control Circuit

4)

Hich Pressure Coolant Injection Turbine Trio Control Circuit

5)

Reactor Core Isolation Cooling Turbine Trio Control Circuit

'J -eactor Prctection Systei 7' Oontainnent isclation Sys:e:

8)

Recirculation Flow Control System

b.

Emergency Systems Range Inputs to:

1) Emergency Core Cooling System Controls

2)

Reactor Core Isolation Cooling System Controls

3)

Containment Isolation System

4)

Recirculation MG Set Control Circuit

5)

Diesel Generator Control Circuit

c.

Post Accident Flooding Range Inputs to:

1), Residual Heat Removal System Control Circuit

2.

Pressure

a.

Reactor Pressure Inputs to:

1)

Emergency Core Cooling System Control Circuitry

2)

Reactor Protection System

3)

Mechanical Vacuum Pump Control Circuit

4)

Recirculation.G Set Control Circuits

5)

Feedwater Control System H.

TECHNICAL SPECIFICATIONS

1.

Tc icai S.cecificazions associa:ed with vesse, instrumerts are z3o numerous t3 enumerate here.

Refer to the references for the various sections in Tech.

Specs. and in Lesson Plans for sys:ems ising this instrumentation.

Revision Date W.%o,*R SYSTE7,S LESSON PLA-:

A.

FUEL B.

REFERENCES

1.

BWR Systems Manual, Chapter 2.3

2.

Desification Considerations in BWR Fuel; NEDM-10735, Supplement 5

3.

GE Nuclear Engineer's Manual

GEI-92823B

4.

Pre-Conditioning Interim Operating Management Recommendation (PCIOR)

5.

BWR/4 and BWR/5 Fuel Design, NEDE-20944-1P, September, 1974

6.

Licensing Topical Report, GE BWR Generic Reload Application for 2 x W Fuel, NEDE-20360-P, April, 1974, Revisions 1, 2, 3 & 4

7.

GE BWR Reload Licensing Amendment for Browns Ferry Nuclear Plant Unit 1, NEDO-24020,

May, 1977

3. Final Safety Analysis Report Browns Ferry Nuclear Plant 3.2, 3.6, 3.7

9.

Reference Card File 2.3 C.

OBE CTIVES I.

To underszanc the mechanical ccns-ruction and nuclear dezicr 0 tZe deeI.

2. 'Te understand the use cf burnable ooisons in the fuel anc treir effeczs on core reactivity.

-. To,xicersand -ne basis anc zrne method of"Fuel Pre-Ccnoi-i,'"ng.

To uncerszand the past and oresenz fuel problems and the c-rrective actions T.aken.

2.

GENEPAL DESCRIPTION ic',e-Oer~er-t'cn C 'ecti'e

5. uei,ec."ani*"cal*_.*, le*c S1)

To Drovide a hich inte~ri-y assembly cf fissionable Sazeriai

.,hi

n c--n e,-e

-,rran-ed in a critical array.

Tne ass -zl,1 Tust Oecaoabie of ef-fcienTcly transerrinc the gener=a:-

-r-,a=-".

coclr.-n water wh',e ai Zainir.: szruczural -n-ecri-y and con-airninG the fission DrZ.U.

S...

b.

Nuclear Design

1) To attain rated power generation' from the nuclear fuel for a given period of time.

2)

To attain reactor nuclear stability throughout core life.

3)

To allow normal power operation of the nuclear fuel with out sustaining fuel damange.

2.

Safety Design Basis

a.

Fuel Mechanical Design

1) The nuclear fuel shall be utilized as the initial barrier to the release of fission products.

The fission product retention capability of the nuclear fuel shall be substan tial during normal modes of reactor operation so that significant amounts of radioactivity are not released from the reactor fuel barrier.

b.

N*uciear Design

1) Fuel nuclear design shall provide negative reactivity feed back that is sufficient, in combination with other Dlant systems, to prevent fuel damage as a result of any abnormal operational. transient.

2)

Fuel nuclear design shall exhibit such nuclear characteristics as recuired to assure that the nuclear system has no inherent tendency toward divergent or limn: cycle operation.

3)

Fuel nuclear design shall limit the excess reactivizy of the core sufficiently to assure that reactivity control syste-ns are capable of makinc the core subcritical at any 4-e

.ni

.the contro rod cf h-ihest worth fully w;i-hdrawn.

1.

Fuel Ass":-lv X.echanical Cons-ruction

a.

Fuel Bundle i) 76a Bundles 2, Len:zr. Co Active Fuel

- l-inches 0

Total.eight -.6E0 lbs.

b.

Lower Tie Plate (Fabricated from Type 304 Stainless Steel Casting)

1) Directs coolant to fuel rods.

2)

Provides grid -for fuel rod positioning.

3)

Contains 8 threaded holes for tie rods.

4)

Positions the channel relative to the fuel rods and relative' to the other channels in the four bundle cell.

S5)

"Has a nose piece which supports the fuel assembly and guides it into the fuel support piece during fuel loading.

6)

All 1 inch bypass flow holes in the core support plate are plugged and two 9/32 inch holes are drilled in the lower tie plate of each reload 8 x 8 assembly.

c.

Upper Tie Plate (Fabricated from Type 304 Stainless Steel Casting)

1) Positions the upper end of the fuel rods.

2)

Positions.the fuel channel - the outside mating configuration of machined bosses maintains the channel-to-upper-tie plate alignment.

3)

Provides fuel handling capability - lifting bail.

A)

Provides fuel bundle orientation.

a)

Serial numzer stamped on handle -

can be read from center of cell.

b)

Eundle installed with the lug on the bail pointing zoward the center of the four bundle cell.

5) Eignt fuel tie-rods pass through the upper tie plate.

a)

On top of the upper tie plate, the tie rods are secured with stainless steel nuts.

b)

A locking tab washer is installed over each pair of

=c`acent t'e rods f-icure 2).

'I) -he tazs are benz up agcainsz the nits z tc.revez.Z rotation and loosening c-the nuts.

(2)

Another tab fits into a notch in the tie rod to prevent rotation and loosening of the tie rod.

d.

Fuel Channel

1) Constructed from Zircaloy-4 a)

Alloy of 98% Zirconium, 1-1/2% tin, and small amounts of iron and chromium.

b)

Lower rate of hydride formation than Zr-2 because nickel not used in alloy.

c)

Better neutron economy, lower neutron absorption cross section than stainless steel.

d)

Channels are to be used repeatedly in successive cycles.

2)

Fastened to the fuel bundle with a spring clip channel fastener and one cap screw (Figure 3).

3)

Purpose a)

Channels the coolant flow upward through the fuel bundle (approximately 90% of the flow is through the bundle 10'- is bypassed around the fuel bundle).

b)

Provides a bearing surface for the control rod blades.

c)

Provides protection for fuel rods during fuel handling.

d)

Provides the primary resistance to lazeral acceierazior loadings on the fuel assembly (seismic loadings).

e)

Correct control rod passage clearance ensured by stain less steel buttons at the too of the channel - the but:on roazes with the button on the opposing channel.

e.

Fuel Rod Spacers ( Figures 4 and 5)

1) Fabricated from Zircaloy-- sheet metal for neutron economy.

2)

Located at ý,20" intervals along the fuel rod (total of seven spacers).

,-r-Ji:e zssizi-e c3rtact*`srZor fuel rocs.

• )

Radially positions the rods relative to one another and relative to the fuel channel.

")

A solid support is formed for each rod by springs which push the rods against their support.

a)

Prevent fuel rod vibration b)

Prevent fretting wear associated with fuel rod vibration c)

On 7 x 7 and 8 x 8 bundles, Inconel-X is used for

-springs for reliable spring properties and character istics.

(Inconel-X is an alloy of chromium, nickel and iron.)

6)

Functional requirements placed on the fuel spacer are that it must transmit the acceleration loadings imposed by the fuel rods to the channel without loss of its capability to properly position the fuel rods.

7)

Spacer-Capture Rod - The spacers are held in their axial position by a sDacer-capture rod.

a)

Fuel Spacer-Capture Rod (7 x 7 Bundles)

(1)

Center rod of fuel assembly - contains fuel (2)

Segmented with square end plugs to mate with the spacer and prevent the spacer from sliding axially.

(3j Segmented sections of the center rod are drilled to allow transport of the fissicn cases to the upper plenum.

(4)

Center opening of the spacer has a square shape to accommodate the fuel spacer capture rod.

(5)

Fuel spacer-capture rod has a square lower end plug shank to nate with a square hole in the lower tie plate -

rod. cannot rotate and release the spacers.

b)

Spacer-Capture Water Rod (8 x 8 Bundles) (Figure 4)

(I Loca:ec off-cer,:e-o' fuel asse-+/-%.'

- ccntains no rue!

(2)

Hollov,,Zircalloy - 2 rod (3)

Several holes are drilled around the circumference of the rod at each end to allow coolant water to flow through the rod.

(4)

Welded tabs relative to the axial spacer locations.

During assembly, the water rod is passed through the spacer and then rotated to lock the spacer in its axial position.

(a) 14 tab water rod, design improvement to provide additional. design margin.

i.

Spacei grid (14 tab water rod) is posi tioned by a tab above and below a struc tural member (divider) of the spacer grid.

ii.

14 tab water rod has the following advantages over the 7 tab water rod:

- Provides a stronger positioning device

- Reduces loads induced in the water rod

- Minimizes spacer cocking

- Provides improved tolerance control iii.

Strength of spacer arid positioning device for the 14 tab water rod is inproved by approximately 33-over the 7-tab design.

(b)

Old design 7 tab water rod positioned spacer by being located between twzo structural

-.emrbers (bars) of the spacer grid.

(5)

Bottom end Dlug of the water rod has a square shank which:

(a)

M'ates with a square hole in the lower tie clate and prevents rotation and release of the soacers.

(b)

Prevents bundle assembly with the water rod in the wrong position.

!6)

Effect of the "'.ater P~d

%a) increases 41 erz:ion in t re ir..erior cf t-,e bundle.

(b) improves neutron econony.

(c)

Flattens bundle power distribution - reduces tne rod-to-rod power peaking.

(d)

Reduces the void coefficient of reactivity.

(e)

Flattens axial power shapes by reducing the yoid coefficient and increasing the ratio of, non-boiling-to-boiling water in the top of the core.

f.

Finger Springs (Reload 8 x 8 Bundles Only) (Figure 6)

1)

Sheet metal stampings

2) *Positioned by the ends of the fuel rods

3)

Fill the space between the channel and the lower tie plate.

4)

Maintain a constant pressure against the lower tie plate and maintain a constant bypass flow.

5)

Reload 8 x 8 fuel incorporates finger springs for controlling moderator/coolout bypass flow at the interface of the channel and fuel bundle lower tie plate.

g.

Expansion Spring

1)

Located over the top end plug pin of each fuel rod.

2)

Keeps the fuel rods seated in the lower tie plate.

3)

Takes up differential expansion during operation and compen sates for minor differences in fuel rod lengths by sliding within the holes of the upper tie plate.

1) Protects the fuel rods against any impacts which might occur against the upoer tie plate.

5)

Spring fabricated from Inconel-X on 7 x 7 and reload 8 x 0 fuel.

E.

C3:.**ONIENT DESCRiPTION

1.

Fuel Rod (Figure 7)

a.

Types of Fuel Rods

8-I)

Standard Fuel Rods

2)

Tie Rods a) Active fuel length same as stanaard rod b)

End plugs threaded

3)

Fuel Spacer-Capture Rod a)

Active fuel length same as standard rod

b.

Cladding

• )

Fuel pellets contained in Zircaloy-2 tubing.

(Zircaloy-2 is

,92% Zirconium alloyed with small amounts of tin, iron, chromium and nickel.)

2)

Zircaloy-2 has less resistance to hydriding than Zircaloy-4 but a better heat transfer coefficient, better neutron economy than stainless steel.

3)

Cladding thickness is adequate to be "free standing" (i.e.,

caoable of withstanding external reactor pressure without collapsing onto the pellets within).

L)

Cladding surface pre-oxidized by an autoclave process making it more resistant to contamination and easier to keep clean.

c.

Plenum Volume

1) Free volume for the accumulation of fission gases.

9' Su-'icien+ volume provid-e to oýreven: excessive inte-nal pressure from fission cases liberated over the design life of the fuel.

3)

Fission roCduc-s for-ed during.oer-z~on are mcSt:v ccna=_`rec I.,.hin zne fueI :e e-is -

a re atively s,-il a.*:rt is releasec and must *e acc..odated in.sice -he rod.

d.

?lenum Spring

1) Prevents movement of the fuel column inside the fuel rod during fuel shipping and handling.

2)

Allows for axial expansion of the fuel pellets.

3)

Keeps the fuel pellets down in the active fuel length of the fuel rod.

• ")

Spring is fabricated from stainless steel.

e.

Hydrogen Getter (Reload 8 x 8 Only)

1) Hydrogen will combine to form a hydride of low strength which tends to migrate to local areas in the cladding.

Blisters are formed which can result in fuel failure.

2)

The getter is added a's a precaution during fuel manufacture "to avoid hydrogenous contamination (internal hydriding) in addition to hot vacuum outgassing of the fuel rod.

3)

The Zirconium-based getter is 100 times more reactive with water than is the Zircalloy-2 cladding and absorbs any moisture left in the fuel rod.

_)

Getter Construction a)

Zirconium alloy in the form of small chips is loosely packed in a stainless steel tube.

b)

One end of tube capped.

c)

Other end of tube is covered by wire screening.

f.

End Plugs i)

Unper and lower end plugs are fabricated from Zircaloy-2 (Zr-2).

2)

End plugs are seal welded to the fuel tube to provide a sealed tube after hot vacuum outgassina and backfilling of the fuel rod with helium gas to one atmosphere Dressure.

t -eiium is an inert gas witn A-cood heat transfer coefficienz.)

2)

Shanks on the end plugs encage the uoper and lower tie plates and maintain rod position in the bundle.

Shanks of upoer end plugs are sized according to the enrich ment of the fuel rod (7 x 7 and Reload 8 x 8).

1i-nacer :he i

t u:-er end : 'h "

z'er

-,:=. Enri;`,,-,-.=,n: Of -.:'E

.,r.

10 -

b)

The holes in the upper tie plate are drilled to mate the end plug.

c)

This prevents incorrect enrichment location in the bundle.

d)

It is not.mechanically possible to completely put together a fuel assembly.with any high enrichment rods in positions specified to receive a lower enrichment.

e)

Provides visual check of fuel assembly orientation.

6)

Type II and Type III fuel assemblies have a thermal barrier "interposed between the bottom fuel pellets and the lower end plug of the basic fuel rods and between the connect6rs and their adjacent fuel pellets in the spacer capture rods.

a)

The purpose of the thermal barrier is to reduce the operating temperature difference between the lower end plug (or connector) and the cladding in the weld regions.

b)

The barrier consists of four close winds of 35 mil-dia meter.stainless steel wire with 30 nil thick stainless steel waters spot welded to each end.

c)

Recent calculations and tests indicate that all stress limits are satisfied at the welds without the thermal barriers - thermal barriers will not be used in Unit 3

,uel.

.- E! Pellets

1) Sinzered ryiinder of UOL.

or UO2 - Gd20 a)

Hich density ceramic uranium dioxide or uranium dioxide - Gadolinium cxide.

b)

Dowder cold pressed at high pressure to form pellets and then sintered in a reducing atmosphere at 1650 to 17500C.

c) Average pellet immersicn density is about 95 of

-.:hcr--'c& censi-; -

-. :csitv of 3 *,,o,.,s. r

-11

2)

Pellet Shape a)

Shorter Fuel Pellets with Chamfered Corners (Reload 8 x 8)

(1)

Length-to-diameter ratio is approximately 1.0 (2)

Corners of pellets chamfered (3)

Less mechanical interaction between the fuel and the cladding.

(4)

Reduced possibil.ity of pellet-clad interaction fuel failure.

(5)

Pellets are.416 inches in diameter,.420 inches in length.

b) 7 x 7 fuel pellet edges are chamfered to reduce mech anical interaction between the fuel and cladding.

"3) Radial gap between fuel oellet and cladding provided to allowi for pellet growth from thermal expansion and irradi ation swelling.

a)

Gap.0)2 inches, Type I, II, III b)

Gpa

.009 inches, Reload 8 x 8

4)

Multiple Enrichments "a) Power density orcportional to:

Dower Density

-:z- =.TA where

= fission cross-section of the fuel thermal neutron flux N

number density of fuel atoms b)

Desirable to maintain a uniform Dower density across the core.

c)

Greater thermal flux exists in the water gaps due to Aetter ioderation (Ficure 10,.

Revision Date BWR SYSTEMS LESSON PLAN A. CONTROL ROD DRIVE HYDRAULICS B.

REFERENCES I.

Boiling Water Reactor Systems Manual, Chapter 2, 4

2.

Browns Ferry FSAR Chapter 3, 4

3.

Browns Ferry Instrumentation and Control Manual Volume IX Part I

4.

Browns Ferry Operating Instruction #85 5.Air Operated Control Valve Instruction Manual

6.

Control Rod Drive Pump Instruction.Manual

7.

Reference card for CRD Chapter 2, 4 C.

OBJECTIVES

i.

Describe operation of hydraulic control unit

2.

Describe components and operation of CRD Hydraulic Systeni

3.

Describe air systems needed for control rod scrams

,a. Discuss the associated control room and local instrumentation Review applicable Tich Specs

1. The CRD System controls changes of reactivity by incrementally positioning control rods within the core in response to Re.:-er Vanuai Control sionals.

2.

r.e sys:e is also required tic uickly shut down -he reactor by ra:i clv inserting con-rol rods into the core in response to a manual or automatic signal.

• . Cor.r.i Rod Drive Hydraul ic Control Unit

a.

One hydraulic control unit is orovided for each control rod.

-s -

b.

Combines all operating valves and components needed for the normal positioning or scran of a single control rod.

c.

Unit functions on pressures supplied by CRD Pydraulic System to insert or withdraw its asscciated drive and provide cooling water to the drive mechanism.

d.

Provides stored ehergy to give initial scram energy to rod.

4.

Control Rod Drive Hydraulic System

a.

Purpose CRD Hydraulic System provides the necessary pumps and valves to provide water at the proper pressure to the HCUs for all operations of control rod motion.

Piping Arrangements

b.

Provides piping and necessary volume to ensure that all rods can scram yet prevent loss of an excessive amount of vessel inventory.

E.

CO!IPOQNE'JT DESCRIPTIONS

1. Hydraulic Control Units (Fig.

1)

a. Purzose Combines all ooera.inc valves and components recuired for the normal or scram operation of a single control roc.

r-znen~s "Lydraulic risers

'!;nifoid and directional control valves Scram inkE: and ou.iet valves c

.,u c

Ia: cr inszru-ment assembly

c.

nsailatior rcnfi.u tirn 1i Quantity 185 (one for each WRD)

2)

Arrangement Roughly equally divided into d banks of HCUs on each side of the reactor building at ground floor.

d.

Piping Assembly (Fig. 2)

1) Seven hydraulic risers Insert line - to CRD underpiston area Cooling water - from cooling water header

.Exhaust line - to CRD Hydraulic System return line Scram discharge - to scram discharge volume Drive water - from drive water header Charging water -

from charging header to accumulator Withdraw line - to CRD overniston area

2)

Manifold Sa)

Purpose Directs water between the seven risers and valves on the HCU b)

CoMoonents Ccrzains cooiinc and drive water check valves, fil:er elements to Drotect the direcoicnal control valves and the CRD ana pressure test plugs.

3) inlet and Outlet Scram Valves a)

Durpose Control water flow to and frcm the CRD for scram insertion.

e.

Driving Water Section (Fig.

2)

1) CoT0:cnents FNr Formally ueenegi:Ke soen~c o::eraze di-s=Tneq con:rcl -VaEves are 7ounted on the :ping assenmly -anifolc.

SSolenoids powered by 120 vac Nus.

e.

2) Directional Control Valves Energized by the action of the Reactor Manual Control System to control pressure on -he under and overpiston area of the associatea CRD mechanism.

3)

Operation By energizing two of the four valves simultaneously, the drive water header is connected to either the under or overpiston area while the exhaust header is simultaneously connected to the opposite side of

,the drive piston.

a)

Insert operation Valves85-40A and 85-40D open Valve 85-4OB opens to settle the drive Valves 85-AOA and 85-40D close Valve 85-40B closes b)

Withdraw Valves85-40A and 85-L0D open to lift drive off of the collet fingers Valves85-10A and 85-aOD close Valve 85--I0D and 85-*CC cpen Valve 85-L0C closes allo'.*.,ic drive :o settle Valve 85-40B closes

".l-w Control Valves Two of the directional control valves are ecuipoed with inzegral flow control valves.

These valves are arranged so that they pass flow to or from the underpiston area.

I Tnsert throttle valve

'l.jow wo directicna vor flow to underaiston voiume Flow to underpiston volume is greater than exhaust from overpiston (because of larger underpiston area.)

Thus, better speed control obtained by throttling larger flow, i.e., unGerpiston flow.

Throttle drive water b)

Withdraw throttle valve Located with directional control valve porting flow from underpiston line.

During withdraw, rod is falling; therefore, it is necessary to brake the rod by throttling flow from the underpiston area.

Flow from underpiston area is greater than to above piston (because of larger underpiston area.)

Therefore, speed control is only feasible on underpiston port.

In addition, full drive water pressure is needed to hold collet fingers extended.

Jf drive water were throttled, collet fingers could not be extended and rod movement would not be consistent.

Throttles exhaust water.

F slow 1Rate Flow is approximately 4 GP* when the drive is being inserted at 3 inches per second and 2 GPMI during drive wiVhdrawjai ooeration.

Dene~oer that

-under/P-,ver areas are 4.0 in -/1.2 in Therefore it wo.Id see-ttE 2 G'. is an excessive flow rate.

This hiýn raze, however, is recuired to accommodate the higher P-over collet seal leakage.

(k-under seal leakage is nuch smaller.)

6) Drive water and cooling water check valves Curing a scram, the drive insert header is pressurized to accumulator Dressure.

Back flow is prevented through the cooling water line by the coolihg water check valve Back flow through the drive header line is prevented by the drive water check valve since accumulator pressure could lift the drive water supply valve 85-40A off its seat.

NOTE:

A check valve cannot be placed in exhaust header for the same purpose because it would prevent normal flow operation of the exhaust header.

Valve 85-40B is oriented so that increased pressure tends to force the valve closed rather than than force it open.

7)

Filters Filters are provided in the drive water line to the directional control valves and in the drive insert and drive withdraw header to the CRD mechanism to prevent damage to the directional control valves and the drive mechanisms from rust or scale in the CRD Hydraulic System water (rhost of the supoly viping

'is carbon steel.)

Cooling Water Section (Fig. 2)

1) Components Consists of coolinr water riser and cooling check valves

2)

Flow Path Cocolin-water f-::s frcr. :he coolin. water header

hrough the cneck valve to tne drive insert line at ai, times wnen the rod is stationary.

3) Fl ow Rate Flow.25-.33 GPM az about 20 psi above reactor pressure.

.:ner a drive is i' motion, oressure in -,he insert l4ne is at drive pressure (260 osi; reactor pressure).

Therefore, the check valve will be closed to close off the flow of cooling water and prevent drive water from recirculating back to the CRD Hydraulic System.

g.

Scram Section (Fig.

2)

1) Operation a)

Scram action Two scram pilot valves are directly connected to the Reactor Protection System so that the inlet and outlet scram valves open in response to scram signals.

b)

Inlet valve When open, the inlet scram valve permits the scram accumulator to supply the initial energy to raoidly insert the control rod.

c)

Outlet valve The opening of the outlet scram valve permits water.vented from the overpiston area of the CRD to exhaust to the scram discharge volume.

d)

Scram pilot air valves (Fig.

2)

(1). Quantity Two per HCU (2)

Type 3-way solenoid operated valves

'3)

Power supply Energized normally by 120 VAC Reactor Protection System power A)

~Normal position When one or both are enerGized, they port ir scý-r valves Szr~n ac:itr1 Both deenergize on scram to vent aIr from scraz' valve diaunracm.

e) Scram valves (1) Type Both glove valves with Teflon seats to minimize leakage (2)

Normal lineup Both air operated valves normally held closed by air pressure from instrument air header (3)

Scram action Open on internal spring pressure upon removal of control air pressure (scram (4)

Timing Start to open within

.15 seconds after pilot valves lose voltage (5)

Sequencing Outlet valves open slightly faster to prevent buildup of high pressures in CRD (has stronger opening spring)

(6)

Position indication Bc:h provided with s~ring mounted acsition switch.

Vnen both valves ooen, pos~ion s,:i:ches cause olue roo scram signal :r full core display on panel five (7)

Air pressure in the valves is maintained at 75 psig.

"iyher pressures would (a)

Cause deformation of the Teflon seats with resultant valve leakage (b) Slightly increase scram times, Darticularly for the initial phase of rod movement.

()

Scram inlet valve leakage or canse prEsswrib:ly in ot wo e

d

!-'jY:!S area and possibly slow rod inser:ion.

(9)

Scram outlet valve leakage Causes depressurization of the over piston area.

Cooling water pressure on the under piston area can cause the drive to drift in.

f)

Scram accumulator and instrument block (Fig. 2)

(1)

Purpose Serves as independent source of stored energy to initiate scram insertion of the associated CRD.

(2)

Description Piston type accumulator connected to N2 cylinder (3)

Piston Piston serves as barrier between high pressure N. (source of stored energy) and the water used to initiate a control rod scram.

Under normal conditions, the piston is in the full down location.

4A) Piston seal Piston sealed by two-Teflon seals and a rubber O-ring.

(5) Iormal lineup Accumulator continuously charged by CRD Hydraulic System charging water header.

(6)

Check valve Check valve prevents flow back to charging header from accurulatzor.

Accumulators will retain charce for some time in the event of loss of pressure (pumo trip) in charging water header.

(7) 1,, Source 2Rternai source.

(8) overpressure protection Rupture disc on inszrument block bursts at 2000 psi to protect accumulator from overpressure.

(9)

Monitoring Continuously monitored for gas and water leakage (a)

Water leakage past accumulator piston seals Float type magnetic reed level switch senses water in H* side.

Alarms when 37 cc of watir accumulates.

(b)

H 2 pressure Low pressure switch senses low gas pressure (940 to 970 psi)

(c)

Alarms Either will cause "Accumulator Trouble" alarm in control room (audible and visual)

(amber light on full core display) on panel five.

(d)

Operation Action Operator must zhen go to local acc.u.,a:cr :rouble panei to se if low pressure or water level switch tripped.

Locai cane! will have alarm button lit for EQC.

Operator,us: push button.

If lignt goes OUT, alarm is water leakage.

if light stays on, alarm is low gas pressure.

(10) Nitrogen charging (Figs.

3 & 4)

(a*-

itrogen fc-accumulator cnarging is sucDlie&

":ne on -ac, s-'e of :ne reacz:r :uicin near the

-an's of HUs.

(b)

Components (Fig.

3) in each charging station normally consists of Nitrogen bottle Bottle pressure regulator and pressure monitoring instrumentation 2 isolation valves Safety valve Vent valve Flexible hose (c)

Gas pressure Since gas expansion during initial charging results in low nitrogen temperature, the final accumulator precharge pressure must be set only when the temperature of the nitrogen in the accumulator has reached equalibrium (room temperature)

Fig. A is used to determine the proper accumulator precharge gas pressure.

Reason for requiring a variable charging pressure is to assure that the piston in the accumulator is always full down during operation, thereby guaranteeing the required amount of water is available to scram the drive.

Too high a pressure would prevent this.

If too loow a

^ pressure is used, :he low Dressure ata-r will te -ripped even wnen :ne cyiiincer pistor. is fu7" town.

Too low a pressure will affect scram zimes.

If accumulators are charged to a low pressure in hich room temoerazure and the room :emperature sucsequenziy decreases, it isprobable that the low pressure alarm will be tripped.

(d)

Accumulator charging procedures (Fic.

2)

(See Browns Ferry Operating instructions 85Step !17' and Tech Manual GEl 92807A for details.)

(i)

Close all HCU riser manual isolazion valves, including charqing water valve $5-5E (ii)

Connect a flexible hose from valve 85-590 to the floor drain and slowly open valve 85-590 and drain the accumulator.

NOTE:

All manual isolation valves should be closed rather than only valve 85-588 (in step 1) above because if the reactor scrams while valve is open, 85 590 is open.

(aa)

Cooling water header pressure will be directly connected to valve 85-590 (bb)

Reactor water at temperature and pressure would be ported to the drive P-under area throuch the ball check valve and could, if the bail check valve failed, blow

'ack through the insert line and valve 85-590.

(iiil Verify that valve 85-229A is closed.

(iv)

Slowly remove cap on P6 to bleed off the instrument block gas pressure.

v Connect cas charcinc line to PS.

(vi)

Open valve 85-229A.

(vii)

Fill accumuiaton to specified pressure, close valve 25-220A and recap P5.

(viii)

Ooen valve 85-229A and reverify correct pressure.

kix)

Close valve 85-590 (x)

Slowly open charging water valve 35-588 to push the rechanicai piszon to the "xi,-Reocen a!'

rise- :7anial isolation valves.

(11)

Procedures for draining water from instrument block (See Fig. 2)

(a)

Close valve (35-229A) on accumulator instrumentation block (b)

Slowly remove cap on plug P-6.

(c)

Briefly.crack open valve 85-229A to blow out moisture.

(d)

Return valve 85-229A and plug P-6 to normal condition.

(e)

Check N for proper pressure and recharge if neceisary.

(12)

HCU isolation (See GEi -92807A for details.)

(a)

Precautions (i)

Control rod whose HCU is to be isolated should be latched in the full-in position.

(ii)

The HCU should not be isolated for extended periods when tne reactor is in cold shutdown in order to prevent sea!

damage due to loss of cooling water.

(b)

Procedure (i)

Fully close tne isolation valve (65-612) in the insert riser (ii)

Fully close tne isolation valves :n the withdraw and charcing wazer risers (85-615 5-58 )

(iii)

Open the accumuiazor drain valve -c discharge the water from zhe accumulator 85-590 to zhe floor drain fiv)

Fully close tne isolation valve in

ne scra7 :.ilo: air i-e.

--he sc_':

tes: scare in ahe toners1 'mc.

ea.evie zse scran miot air valves.

-14 (v)

Fully close the isolation valves in the scram discharge riser 85-617 the cooling water riser 85-599, drive waZer riser £5-593, and zne exnaust water riser 85-600.

(vi)

Electrically isolate the HCU from the Reactor Manual Control System.

(vii)

Discharge the gas side of the accumul ator.

(13) Electrical isolation of the HCU (a)

Prevents operation of directional solenoid valves (b)

Accomplished by removing the color coded amphenal connectors.

2.

C3htroi rod drive hydraulic system

a.

Purpose Provides water at. the. proper pressure to the HCUs for all zypes of control rod motion

b.

Basic Flow Path (Fig.

5)

1) CST

_ Suc-ion strainers

3) " CRD hydraulic pumps

-)

Drive water filters a

Prcrculazion Pump Seal fiushing -water Charcina water header 7:

Flow control stazion S)

Drive water pressure control station "CJ P.eturn :o -eactor I

"rr~al syszem flow is 72 G:"'

c.

H1ydraulic Supply

1) System suction - condensate storace tank

2)

Pumps a)

Type Two 100% multistage centrifugal electric motor driven pumps (1) Pump 1A is specific to Unit 1 while pump 18 is a swing pump that can be used for Unit 1 or Unit 2.

(a) 1A pump normally valved in lB valved out.

b)

Rating 76 GPM at 1500 psig (plus 20 GPH to minimum flow

-V4#-e)

During a reactor scram, however, the pump will be required to deliver a total flow of 200 GPM at 1350 psig (maximum rating) in order to recharge the accumulator (179 GPH to system and 20 GPM to minimum flow line.)

c)

Speed increaser Constant sDeed mechanical speed increases provided between motor and oumo.

d)

Suction strainers Provided to protect-punp seals, etc.

e)

Vayving Each oump is provided with a manual suction and discharCe isolation valve.

Discharge valve is a stoo check valve to prevent reverse flow in the idle Dump.

E~

ri-'e ;ic* """e-: --..

ei-:.erc uni":

or

,ni:

-16 f)

Minimum flow (1) Purpose A minimum flow line is inciuaed upstream of the discharge valve to prevent over

• heating of the pump in case the discharge valve is inadvertently shut.

(2)

Flow path Minimum flow line is locked open and recirculates 20 GPM to CST through orifice when the pump is running (3)

Valve Minimum flow valve is stop check to prevent backflow through the idle pump g)

Power source 4160 Vac motors, 1A - Unit Board IC 1B Shutdown Board A h)

Cooling Cooling water is provided to the oil cooler and thrust bearing by the Raw Cooling Water System i)

Recirculation Pu7:o Se-al Purge CRD flow ("G?/.e1irc Dump orificed flow,) is directed to each recirculation pump seal zo flush tho during startup conditions.

This flush keeps dirg of seal piping to increase seal life and to limit radioac:ive waste discharge j)

Pump test - a pumo test line is provided (1) Flow patn - frc-pump discharge to return line (2)

Purpose - allows pump tests when balance of system is shuz cown k.,

• eais Operating pump maintains seal pressure on standy pump to Drevent air =-om entering the seais o0 the standby pump if suc:icn pressure to the running pump sHould drc; zec: a:7.ospneric.

3)

Drive water filter a'

Piping arrangement Two filters installed in parallel; one is usually on line, the other in standby.

b)

Filter mesh 50 micron absolute (25 mincons nominal) filter prevents rust., scale, etc. from entering HCUs.

c)

Alarm Condition of filters monitored by dp indicating switch.

Alarm in control room at 20 psid.

d)

Type Cartridge type filters can be vented, drained and cleaned and reused.

e)

"Y" Strainer

."Y" strainer is provided downstream of filters to protect system against large particles if a filter cartidge should fail.

d. Svstem Pressure Control

1)

Rc~ur.azcr cnar:in-line a)

Purpose Suoplies charging -. azer to the HCU accumulators b)

Normal lineup Accumulators "Ifloaz" at purmný discharge pressure c)

Check valve If pumps fail, accumulators are held charged C-./

C,C K Y-a :n,B r, g r.c

. i-I ntS 7 f GW C f waIE' 7:

1-

ie "S

S 0%

e m d)

Runout protection During a scram, the HCU accumulators will be fully discharged The CRD pumps will try to recharge all the

-accumulators at once.

To prevent pump runout and probable tripping of the pump motor on over-current (1)

A restricting orifice is provided to limit the maximum rate of recharging to 179 GPM (maximum flow is with reactor at 0 psig).

(2)

A throttle valve doWnstream of the restricting orifice is provided to provide additional throttling if required NOTE:

The accumulators cannot be recharged until the scram is reset (scram inlet and outlet valves closed) due to drive seal leakage being

.greater than pumD capacity.

e)

Charging water pressure - is independent of reactor vessel pressure f)

Caution Hich accutulacr Pressu,-eS can cause very racid acceleration cf control roc and possible damace by deforming tubes or causing severe shock when drive hits stop piston at end of travel (can irvert Seiieviile wasners.)

Therefore, do not exceed 1510 psig charging water~pressure (GE Design Engineering value).

2; Flow control station (Figs.

5, 6) a)

Purpose T7 -r1vi'e a -ehcd or c-at:ca*!!

b)

Valve arrangement Two pneumatically operated valves in parallel c)

Valve lineup One in'operation, the other in standby d)

Control Flow controlled by flow indicating controller in control room with the fl/A Station in Automatic (1) Manual Mode Operator sets controller output.

Hence, valve remains at a constant open value (2)

Automatic Mode Operator sets flow setpoint.

Flow feedback signal from venturi type flow element in line.

Controller compares setpoint and flow signal and adjusts valve position to hold flow constant.

e)

Detailed valve control (Fia.

6)

(1) Air supplies Manual loading control air through pressure regulator Auto loadina control air tnrcuch E/P unit "otive air fro c-,.osicion.Vr (2)

Local manual operation of flow control valve (a)

Position 3-way valve tn provide manual control air to desired POV positionor from the pressure reculating valve (b)

After verifyino FCV oDerability, open manual isolation valves and position Flow Control valve as required by adjusting Pressure Regulatino valve flow t

a o a c -:, e n

-z'

-cr.--ro', va Iv e (a)

Locally posizicn 3-;.,,av valves to orovide auto loading ccnro-l air to desired flow conzrol valve -.-rs

-1?

1

I

=

-20 (b)

Locally position electrical selection switch to direct signal from flow indicatino controller tc Cesired El? unit.

(c). After verifying operability, open FCV manual isolation valves and position valve as required.

f)

Flow with flow control valve closed (1)

Flow rates 9 GPM with reactor at zero pressure 3 GPM with reactor at rated pressure (2)

Purpose Following a scram, the flow through the charcino water header will be 170 GPM.

The flow control valve will therefore be fully closed trying to maintain 76 GPM,.

Flow of 3-9 GPM is permitted to pass

.through the system and back to the reactor vessel to assure that the reactor vessel return nozzle does not heat up (up on a stoppage of flow) and subsequently rapidly cool down upon re-establishing normal system flow.

This prevents thermal overstressing

-:nE ;'V nozzle.

3)

Drive water pressure control station (Fig.

5) a)

Purpcse To provice a methoýd for adjusting :he preSssure of the water supply used to insert and withdraw control rods.

b)

Pressure control valves (1)

One motor operated pressure control valve c.----

-- a--r r-es---.;---=,nr----

e--,---------.

c :ne motor oper -teo valve (2) Operation Valve is oositioned to maintain drive water pressure at reactor pressure plus 250 psig (PRx + 250)

(4)

Drive water header Provides driving forces to each of the HCUs.

4)

Cooling water pressure control station a)

Purpose Provides water at an adjustable pressure above reactor pressure to maintain CRD temperature much less *than rated reactor temperature to protect CRB mechanism seals.

Normal temperature is 200 F.

b)

Valve arrangement Motor operated pressure control valve bypassed by manually operated pressure control

valve, c)

CRD Instrumentation Temoerature of the CRD mechanism it monitored by a tbermocouple in the Rod Position Information System probe.

7e-oerature is printed out on a multi-poinc recorder in tne conrrol room. A high temmerature alarm alarms at 350 F on a control rod drive mechanism.

P Purpose of temperature monitlring (1) Operation at hich temperatures for extended periods will reduce the life of the drive graphitor seals.

(2)

Temperature instrumentation also provides a means for detection of leaking scram outlet val'ves.

e)

Operation Valve is positioned to maintain cooling water pressure at reactor pressure + 18 - 37 psig (PRx + 18-37) f)

Flow path Cooling water from the.cooling water header flows to the HCUs where it is ported to the drive insert line.

5)

Automatic drive and cooling water header pressure control a)

Pressure in the drive and cooling water headers is dependent upon reactor pressure b)

The flow control valve maintains a constant flow of 78 GPM through the system.

(That is, the flow control valve will have to open further as reactor pressure increases in order to maintain the required system flow.)

c)

If flow stays constant in the system, the pressure drop across the drive and cooling water pressure control valves will stay constant regardless of reactor pressure.

d)

"As a resulz, the drive and cooling water pressure cor. ol vailes wv~ill reouire ad.usztno only once

"*:on sysze-szarz~p' and wiln noz require cons:ant a:,usz.ent during a szartup or shuzdown.

E) Syszem return line

)-"

ow path Returns water from the CRD Hydraulic System to reactor.

1)

Exhaust from cooling water PCV (2)

Exhaust from stabilizing valves n

-j s

')-

Exhaus: from tes**

ioass line b)

Primary containment isolation Primary containment isolation is provided by two check valves backup by a motor operated valve which has no automatic isolation signals c)

RPV nozzle The flow returns to the reactor vessel through the CRD return nozzle.

d)

Purpose of returning to reactor vessel The CRD Hydraulic System must be operated on a constant differential pressure above reactor pressure.

Without returning to the RPV, the drive and cooling water pressure control valves would require continuous operator adjustment or automatic control to yield constant differential pressure.

7)

Stabilizing valves a)

Purpose To provide system flow stability when inserting or.withdrawing a control rod drive.

b)

Confiauration Two identical sets of valves In each set, one valve is for insert o:eration. the czhe- 'cr wi:hdraw oera=:icn.

c)

Normal lineup One set-of valves in operation with the other valved out Solenoid valves are open unless rod is driving.

z'Selec-:~r i:

-Control room swi:ch de-er:Ines which set of valves is in service.

e)

Flow adjustment Needle valves downstream of each solenoid valve proyided for setting the required flow..

f)

Flow path Path is provided from the drive water header to the return line.

,g)

Normal operation Both solenoid valves in one set are normally open with flow set by throttlinq valvps Insert stabilizing valve - passes A GPM (flow required by a CRD while inserting.)

e.

Scram Discharge Volume and Instrument Volume (Figs.

5 & 7)

1) Purpose To receive and contain the water exhausted from all CRDs.during a scram, thereby limiting the ioss of water from the reactor vessel.

"2) Definition a)

Scram discharge volume The hepder pipinc that runs over the tce o0 the hC"s.

It is the piping that connects the HCUs to the i struimenz volume.

b)

Instrument volume The "U" shaDed piping provided for the measurement of water released frcm the drives to zhe scram discharge volume during a scram.

3)

Scram discharge volume valves

-2 T*..:nc t.ýe cZe

)

Drain valve -

1 valve air zo coen, spring -o clese valve

1) Scram discharge volume sizing Indeoendent o" the instrument volume, it is sized to contain the water volume dischaiged from all the CRDs (3.3 gals/drive).

Thus, during a scram, the scram discharge volume partially fills with water.

5)

Scram action On a scram signal, the vAnt and drain valves will shut.

6)

Initial volume pressure during a scram The back pressure in the volume will not exceed 65 psig initially so as not to affect drive scram timing.

7) Pressure rating after scram With all CRDs fully inserted, leakage past the CRD seals from the CRD pump and from the reactor through the scram valves fills the discharge volume and pressurizes the volume to reactor pressure.

• )

Overpressure protection Relief valve set at 1250 psig.

) Vent and drain path

-ke =*. vent and drain valves are Aiped to she Reactor Building Equipment Drain Tank (RBEDT),

to assure no airborne activity is released to the reactor buildino atmoschere when a scram is reset.

"r' r
-n.s:ru*wetaion volume a)

Provides means for measuring amount of water in M.. '.Toime in instrument volume is not needed for scrae.

b)

Normal conditions 7 6Xcessive leakage or inadvertent venz or drain valve closure causes zhe scram discharge

'Yciume :c :ill, conzrol red insertion Tight be

reven**c during a scram.

c)

Level switches Six float type level switches are ccr,rected to the instrument volume to monitor the volume

• for abnormal water level.

d)

Volume not drained switch When 3 gallons accumulate in the instrument volume, one level switch trips and causes an alarm in the control room.

e)

High level When the instrument volume is half full (25 gallons) a second level switch trips and causes a rod block.

f)

High-High level reactor scram The other four level switches trip when the instrument volume is full (50 gallons).

'1hen two switches in a one-of-two-twice logic trio, the reactor will scram since operation without scram capability is forbidden and scram capability is lost if the discharge volume fills further.

a) High-High level byoass "ischarge volume high water level scram can be typasse isnc" a keylock s.izch in :he Shutdown and Refuel nodes of ooeration.

a, This enables the Reac-or Protection System

=z be reseZ SO that the discharce voIjme vent and drain valves can be opened to drain the discharge volume after a scram without initiating a subsequent scram due tn hicn water level.

b)

When the SDV high water level scram is bypassed, control rod withdrawal is blocked c in ala,. a.n n-ciates in.he ccntr.,i

11)

Scram dump valves (pilots and solenoids) (Fig.

7) a)

Purpose These valves apply 75 psig control air to the diaphram operators of the SDV vent and drain valves to hold the valves open b)

Quantity There are two total valves.

c)

Power supply Each solenoid is powered from a separate Reactor Protection System bus.

d)

Scram action When both solenoid operated scram dumD valves are deenergized by the Reactor Protection System,.

air is vented to atmosphere and the SDV vent and drain valves shut.

f)

Conditions for closing

-If only one scram dump valve is deenergized, the SDV vent and drain valves will. remain open.

12)

Discharge volume isolation test valve a)

Puroose Can be manually closed from the control rocm to check for leaking scram outlet valves.

S' Pcver supply Normal power supply instrument and control BusA back uo Dower supply is the uninterruptable 120 Vacrus

13)

Scram pilot solenoid valves a)

PurDose The ciahratr, cce:atcrs of ane sc.n ie: a'c outlet valves are hela closeld

' ins-rurenz air at 75 psig.

b) Quantity There is a pair of valves for each of the 105 HCUs.

c)

Power supply Each solenoid is powered from a separate Reactor Protection System bus.

d)

Scram action When both solenoid operated scram pilot solenoids are deenergized.by the Reactor Protection System, air is vented to atmosphere and the scram inlet and discharge valves are opened by internal spring pressure resulting in the drive scramming.

e)

Conditions for closing If only one scram pilot valve is deenergized, the scram valves will remain closed.

14) Backup scram valves a)

Purpose Provide redundant means of venting air from scram pilot valves and scram dump valves.

Primary purpose is to provide another mechanism for scrammina the 1 or 2 drives which, due to unknown factors, fail to scram.

The valves are not intended to funczion as ar aiernaze -,eznoc of rapidly scraMming zne reac:or.

b) Valves Two solenoid cPerazea three-viav pilot valves normaily ceenergized.

c)

Power suo:iy

ere,.,:a from 125 vdc bus.

d)

Scram operation zriD 6n crcer to energize e,:her z zne valves.

Energizing either valve will vent air off all scram valves and the two scram dump valves.

Scram operation will be slow ( 15-20 secs) because of the large volume of air which must be vented through the small valve ports.

e)

Backup scram valve action The SDV vent and drain valves will also close if either one of the 125 vdc scram backup valves becomes energized.

These valves are normally deenergized.

If either energizes, air is vented off the entire scram air header, opening HCU scram valves and shutting the shutting the SDV vent and drain valves.

f)

Check valve around downstream B/U scram valve is to allow faster venting of air if only upstream valve energizes.

F.

I NSTRUMENTAT I ON

1.

Control Room Instrumentation System flow indicating Controller C!har'ng water pressure IDr",'..

a-t=_

- ` ow Dr*,ve ;...azer.ifferenzial Cc: c i -;

-, Ea-er fl ow Cczlinc wazer differential press-ire Rezurn water flow ere+ re

-e-.

Device Indicator 7ndicatcr Indicator

.ncicazor indicator IndicaTor

'qui tý')i oi t Rance 0-100 GPM 0-2300 -sic 0-8 G?:,

0-305 psid 0-E8 S?FI 0-60 psid 0-50 GPM 0-*0OF

2.

Significant Alarms, Interlocks & Trips

. Item Set Point Scram valve air supply low pressure Scram discharge volume not drained Scram discharge volume hi level rod block Scram discharge volume Channel.9/B high level Rod drive oump low suction pressure Accumulator high level/

O-w oressure 70 psig 3 gals.

25 gals.

50 gals.

18" H3.

absolute 37 cc of water or 970 psig Rod drive hich 3^0°F temperature Charcinc "vater low

• oc Drive wazer filzer

'1::h DP 1!10 Osic 22 psic Alarm only.

indicates abnormally low control air pressure Alarm function only.

Indicates possible leaking scram outlet valve.

Probable significant leakage.

Provides.a rod withdrawal block Reactor will scram at next higher level if not corrected.

Above this level, adequate space for water coming from the drives upon a scram cannot be cuaranteed Therefore, the plant must be scrammed.

Indicate of faulty suction path to pumps - probably a dirty suction strainer.

Will trip the pump.

Indicative of leakina nitrocen fitzin*,

or leaking piston seals in accumulator cylinder or a reactor scram (causing low accumulator pressure;.

Two trips cause a rod out block.

Alarm only. Check cooling water flow to affected drive.

Alarm only.

Low accLu-ua-:r :ressure

/ resu2: in ionrer scram tines.

Aiarm only. indicative of a dirty filter. Valve in standby filter.

Remarks

3.

Significant Local Instrumentation CRD pumD suction pressure Filter differential pressure System flow Charging water pressure Drive water flow Drive water differential pressure Cooling water flow

-Cooling water differential pressure Stabilizing valve flow Return water flow Scram valve instrument air pressure Accumulator nitrogen pressure (on HCU)

G.

OPERATIONAL SUINARY

1. Keep running" at all times:

a.

Prevents damage to seals by providing cooling water for the CRD mechanism.

b.

Provides flcw throuch CRD mechanisr. to vessel, Dreventing

'crud' from failing from vessel into CRD 7echanism, possibly fouling filters or damaging seals.

c.

Keeps system filled and vented.

c.

Keeps recirculation system pump seals pu-ged.

2.

Ven-ino

a.

The CRD Hydraulic System should be well vented Drior to starting CRD pumps to prevent pressure surges (water hammer) in the CRD System lines.

-V,

, ri

.n and :rs jre cin

-_; o ventea onsjre

ro:er c roc z2lv coerati on.

c.

Causes of air in system

1) Occurs primarily when reactor is at zero pressure and CRC Du-Ds are off.

2)

Following removal or reinstallation of a CRD mechanism.

3)

Following hydraulic control unit (HCU) servicing.

d.

Vent locations

1) CRD Hydraulic System Pump vents Filter vents Piping high point vent valvws

2)

CRD pressure overpiston piping CRD pressure underpiston piping Pluc type vents located in high point on piping.

Nio hard picinc installed, hoses must be run from venrt pIucs to floor drains.

e.

• Problems of air in CRD pressure overpiston lines

1) Erratic rod resoonse to control occurs a.i Rc =a no setzle fur'" - an "accumulator discha-:e cc:.rs.

f.

Problems wi-h air in CRD pressure underpiston line 2 The CP2 will fail to notch out because of a loss of unMatcihinc

/

2/ The C:; wjil fail to rctch, in because insufficien:

c/p will be produced to 7ove tne rod to the next la:chec pcsi-icr..

It -

nave a iong seztlino time as it settles back :o its original position.

3.

Scram Valve Leakage

a.

If scram valves leak, a d/p wiill be caused across the Diston resulting in slow rod insertion

b.

This will be seen as "rod drift".

c.

lo clear the condition, insert the rod fully and scram the control rod to try to get scram valves to reseat.

4.

Cooling Water Pressure High

a.

If cooling water pressure is increased to a point where the d/p is enough to lift the control rod, slow inward movement of the control rods.will occur.

5.

CRD.Drive Pressure Variations

a.

The drive water pressure of PRx x +260 psig is a nominal value.

b.

Due to variations of the CRD mechanism seals, i.e.,

different wear rates, etc., more than 260 psid might be required to move a rod.

c.

If a rod does not move with drive water pressure set at reactor pressure +260 psi, increase the drive water pressure until. the rod begins to more.

d.

Then decrease pressure to nominal to.keep rod moving.

"e.

This often occurs after the rod is withdrawn for the first time after a long outage.

6.

Scra='

a.

-ollowinq a scram but before the scram discharce volume is fuli, 'he control rod will be in an overtravel in

osition since there is still a large d/p across the piston.

b.

Therefore, the green full in ligh: on panel five will be on but there will be no rod Dosition readout displayed.

c.

After tne scram discnarge volume is full, there will be no d/p across the piston, and the rod wiil settle into the 00 piston.

7.

Failure of CRD pumps

a.

Operator will be unable to move control rods using normal procedures

b.

Failure of both CRD pumps can be tolerfated for a short time.

c.

Accumulators will start discharging because of back leakage through check valves.

d.

If there is low pressure-on an accumulator, plant should be scrammed if reactor pressure is <550 psig.

H.

RELATIONSHIPS WITH OTHER SYSTEMS

1. Raw Cooling Water "CRD Hydraulic System pumps cooled by Raw Cooling Water System.

Therefore, Raw Cooling mater must be operating in order to operate CRD System.

2.

Power Supply IA pump is powered from 1 C 4160 Unit Board; IB is powered from Shuco';,n Board 1A which can be fed fron a diesel - needed to preven: damage to CRD mechanism seals by providing cooling

.during loss of normal power conditions.

"Deac:or Prot-cz:.,n Syszem Prov-ices si*rals :o ene-cize or denercize scram oilct and dur,.

valves and backup scram valves.

q Reac0or

-a1i_

Control System,

?r;ides sicnaz s to hydraulic c,.,%-r:l unit. -o con-:rol directional cont-o" valve positions to control normal rod motion.

".Con'roi

"-r Syszem Services the flow control valves, scran valves and scram discharge volu-e vent and drain valves.

C,r

.o; is g..Iracte' to booh sEaBs zo.. ur:e the sea's of conza.*.-r7S-t wnic. woulo :u

--he sea's ana poss,10lv increase Rad waste ac:ivity

7. Cn-a-atecfondensate Storage TlanK Pro%' ceS :,aEer -or the S'/sze7

-35

• .I' SPECIFICATI0;IS

1. See Control Rod Drive Lesson Plan for all applicable technical specifications associated with the control rod drive hydraulic system.