Text

Forwards Info Requested in NUREG-0737 Re Shift Technical Advisor Training Program,Performance Testing of PWR Relief & Safety Valves,Containment Min Pressure Setpoint & Instrumentation for Detection of Inadequate Core Cooling
	ML18085A447
Person / Time
Site:	Salem
Issue date:	12/31/1980
From:	Mittl R; Public Service Enterprise Group
To:	Miraglia F, Varga S; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-2.E.4.2, TASK-2.F.2, TASK-2.K.2.14, TASK-TM NUDOCS 8101070329
	Download: ML18085A447 (141)
	v • d • e

--

e OPS~G-.

Public Service Electric and Gas Company 80 Park Place Newark. N.J. 07101 Phone 201 430-7000 December 3l, l980 Director of Nuclear Reactor Regulation U. s. Nuclear Regulatory Commission Washington, D. c.

20555 Attention:

Gentlemen:

Mr. s. A. Varga, Chief Operating Reactor Branch 1 Division of Licensing Mr. F. J. Miraglia, Chief Licensing Branch 3 Division of Licensing CLARIFICATION OF POST-TMI REQUIREMENTS NUREG-0737 UNITS 1 AND 2 SALEM NUCLEAR GENERATING STATION DOCKETS NOS. 50-272 AND NO. 50-3ll Public Service Electric and Gas hereby submits, in the enclosure to this letter, its response to information requested in NUREG-0737 by January 1, l981.

Should you have any questions.in this regard, please do not hesitate to contact us.

~.:..... ~.'

Enclosure CENTENNIAL OF LIGHT

'Zl 1879 1979 Very t7u1~ yo~s,

/,

I r

.~ /,/-/,!' '~*

~. ;/ /t///,,l1t1 R. L. Mittl General Manager -

Licensing and Environment v

95-2001 i300tvll 1 T*.

NUREG 0737 I.A.1.1. RESPONSE Shift Technical Advisor Training Program NUREG 0737, page I.A.1.1-2, requests information regarding the training program for the January 1, 1981 Shift Technical Advisor (STA) and the status of conformance with the STA segment of the October 30, 1979 letter from H. Denton, addressed to all oper-ating nuclear plants dealing with clarification of short-term STA requirements.

In addition, it requests information on PSE&G's long-range plans for the position, including the eventual phase out of the STA program. is the course outline for our STA program.

As can be seen, it is presented in 10 modules over a scheduled period of 45 weeks.

Presently, modules I-VIII are presented by Westing-house "Nuclear Service Division" under contract from PSE&G.

Modules IX and X are presented by PSE&G personnel.

Modules IX and X were developed by PSE&G after comparing the as-purchasep Westinghouse STA program with the April 30, 1980 INPO guideline for Nuclear Power Plant Shift Technical Advisor.

With respect to the 1981 STA candidates, and after considering their educational background and the overall STA training pro-gram, we will have STAs on-shift by January 1, 1981, who are fully qualified in accordance with the October 30, 1979 letter previously referenced.

As far as the STA requalification program, our STAs will complete all phases of the Salem licensed operator requalification pro-gram, except the annual examination (unless they are appropri-ately licensed for the plant).

Our present requalification program includes the accident/transient analysis and simulator training outlined in the INPO guidelines for the Nuclear Power Plant Shift Technical Advisor {the STAS will function in their advisory capacity at the simulator in order to demonst~ate their capability to analyze the situation and advise the Shift Super-visor).

We believe that participation in the licensed operator requalification program exceeds any STA requalification guide-lines. is a copy of the Salem licensed operator requalification program.

Appendix D of that procedure (page 24, ) addresses the STA retraining.

The requalification procedure has been revised to comply with requirements of the March 28, 1980 letter from H. Denton, titled "Qualifications Of Reactor Operators".

The procedure is presently under review, with respect to the.October 8, 1980 INPO guideline titled "Nuclear Power Plant Requalification Program For Licensed Personnel".

NUREG 0737 I.A.1.1. Response 2 -

Our long-range plans include phase out of the STA position.

The STA function will be performed by an individual with the title of Shift Supervisor-Engineer (SS-E).

This SS-E will be a degreed engineer who holds a senior reactor operator license.

Develop-ment of these individuals will be a process such as outlined in the block diagram on Attachment 3.

Basically, a graduate engi-neer will be selected and enter the STA training program as outlined in Attachment 1.

Upon satisfactory completion of that program, they will be assigned to shift as an assistant to the reactor operator.

After three month on-shift, this person will return for additional training prior to obtaining a Reactor Operator license.

They will then return to shift as an assistant to the senior shift supervisor.

After approximately one year, they will receive additional training prior to obtaining a Senior Reactor Operator license, after which they will be promoted to Shift Supervisor-Engineer.

It is contemplated that we will phase out the existing STA program by July 1982.

As requested in NUREG 0737, Attachment 4 is a comparison of our STA program to the INPO STA guideline previously referenced in this letter.

+

,. \\

'CPJ:hs 12/31/80

Public Service Electric and Gas Shift Technical Advisor Training Program Module I -

STA Fundamentals (B weeks)

a.

Basic PWR Orientation

b.

Nuclear Reactor Theory

c.

Neutron Physics

d.

Reactor Physics

e.

Reactor Kinetics

f.

Subcritical Reactor Theory

g.

Introduction to Reactor Control

h.

Control Rod Reactivity Effects

i.

Soluble Boron Control

j.

Moderator Temperature Effects

k.

Fuel Temperature Effects

1.

Total Power Defect

m.

Poison Effects

n.

Estimated Critical Conditions and Shutdown Margin Calculations

o.

Core Operational Concepts

p.

PWR Thermodynamics ~ Fundamentals (1)

Introduction to Power Plant Function (2)

Definitions (3)

The First Law of Thermodynamics (4)

Heat Engines (5)

PWR The~modynamics - Applied ATTACHMENT 1 Page 1 of 16

(6)

Heat Transfer (7)

Heat Exchangers (8)

Heat Transfer with Change of Phase

1.

The Boiling Curve

2.

DNB

3. - The Condenser

4.

Fouling (9)

Natural Circulation (10)

Plant Transient Response

1.

Steam Generator Shrink and swell

2.

Primary System Temperature Changes (11)

Fluid Flow (12)

Pump and Flow Characteristics ATTACHMENT 1 Page 2 of 16

Public service Electric and Gas Shift Technical Advisor Training Program Module II - Nuclear Training Reactor -

Zion, Illinois (2 weeks)

a.

General Introduction to the WNTR

b.

Review of NTR systems and Reactor Checkout

c.

Rod.Bank l/M Approach to Critical

d.

e.

f.

g.

h.

i.

j.

k.

1.

m.

n.

o.

P*

q.

r.

Reactor Startups Reactor Shutdown and Restart Power Level Changes Control Rod Trading and Relative Reactivity worths Reactivity Measurements Delayed Neutron Effects Compensation Effect on CIC Channels Reactivity Measurements Critical Control Rod Configuration and Changes in Rod worths Use of the Reactivity Computer Control Rod Caiibration Void Reactivity worth Step and Ramp Reactivity Changes Moderator Differential worth Neutron Absorber worths

~

ATTACHMENT 1 Page 3 of 16

Public Service Electric and Gas Shift Technical Advisor Training Program Module III - Health Physics and Radiation Protection (2 weeks)

a.

Introduction to Sample Radiation Safety Calculations

b.

Biological Effects

c.

Principles of Radiation Protection

d.

Principles of Radiation Detection

e.

Radiation Monitoring and Dosimetry

f.

Radiation Detector Usage in Plant

g.

Chemistry - Primary and secondary Systems (1)

Primary Chemistry (2)

Secondary Chemistry O

ATTACHMENT 1 Page 4 of 16

A o *** *.* ****!**. **~. * *..;...... *;* *.* < **,;

P*****-~- * *<

~-*--' **

Public Service Electric and Gas Shift Technical Advisor Training Program Module IV - Plant Systems (15 weeks)

a.

Reactor Coolant System

b.

Reactor Vessel, Internals and Fuel

c.

Reactor Coolant Pumps and Steam Generator

d.

Chemical and Volume Control System

e.

Containment Structure and Systems

f.

Residual Heat Removal System

g.

Safety Injection System

h.

Containment Spray-System

i.

Iodine Removal System

j.

Containment Isolation System

k.

Component Cooling Water System

1.

Auxiliary Feedwater System

m.

Excore Nuclear Instrumentation System

n.

Incore Instrumentation System

o.

Full Length Rod Control System

p.

Rod Position Indicating System

q. *Rod Insertion Limit

r.

Pressurizer Pressure. and Level Control Systems

s.

Steam Dump System

t.

Steam Generator Level Control System

u.

Reactor Protection System

v.

Process Control System Logic Diagrams ATTACHMENT 1 Page 5 of 16

w.

Protection/Safeguards Logic Diagrams

x.

I&C Systems Integration

y.

Main Steam System

z.

Auxiliary Steam System aa.

Condensate System bb.

Feed System cc.

Electrical Distribution dd.

Electrical Controls ee.

Diesel Generator ff.

Safeguards Sequence gg.

Fuel Handling hh.

Spent Fuel Pool ii.

Spent Fuel Pool C)oling System jj. service Water System kk.

Ventilation System

11.

Instrument and service Air System mm.

Fire Protection System nn.

Stearn Generator Blowdown system oo.

Sampling System pp.

Turbine Generator qq.

Turbine Generator Support Systems rr.

Electro-Hydraulic Control System ss.

Voltage Regulator tt.

Liquid Wast~ System uu.

Gaseous waste System vv.

Solid Waste System ATTACHMENT 1 Page 6 of 16

ww.

Plant Computer xx.

Technical Specifications yy.

Seismic Monitoring zz.

Loose Parts Monitoring ATTACHMENT 1 Page 7 of 16

Public Service Electric and Gas Shift Technical Advisor Training Program Module V - Administrative Functions (2 weeks)

a.

Responsibilities for Safe Operation and Shutdown

b.

Equipment outages and Clearance Procedures

c.

Use of Procedures

d.

Plant Modifications

e.

Shift Relief Turnover and Manning

f.

Containment Access

g.

Maintaining Cognizance of Plant Status

h.

Unit Interface controls (multi-unit plants with one or more units still under construction)

i.

Physical Security

j.

Control Room Access

k.

Duties and Responsibilities of the STA

1.

Radiological Emergency Plan

m.

Code of Federal Regulations (appropriate sections)

n.

Plant Technical Specifications (including bases)

o.

Radiological Control Instructions ATTACHMENT 1 Page 8 of 16

\\...._

._j Public Service Electric and Gas Shift Technical Advisor Training' Program Module VI -

General Operating Procedures, Transient and Accident Analysis and Emergency Operating Procedures (3 weeks)

a.

General Operating Procedures (1)

Startup (2)

At Power Operations (3)

Shutdown (4)

Xenon Following While on Standby (5)

Estimated Critical Position and Shutdown Margin Calculation

b.

Plant Systems summary and Integration

c.

(1)

Systems Review and Interface (2)

Control and Protection System Review (a)

Control and Protection Setpoints and Logics (b)

Core Limits (c)

Protection Trips Transient Analysis (1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

Reactor Startup Ramp to 100% Power 5%/minute Up~power Ramp 5%/minute Down-power Ramp 10% Step Load Changes 20%, 50% and 100% Load Rejections Reactor.Shutdown Abnormal Transients ATTACHMENT 1 Page 9 of 16

d.

Instrument Failure Transient Analysis (1)

Temperature sensing Instrument Failures (2)

Flow sensing Instrument Failures (3)

Power Sensing Instrument Failures (4)

Pressure sensing Instrument Failures (5)

Level Sensing Instrument Failures

e.

Technical Specifications Review

f.

Introduction to Accident Analysis (1)

Classes of Accidents (2)

Des{gn Accident Study Assumptions

g.

Accident Analysis (1)

Reactivity Excursion Transients (2)

Increase in secondary Heat Removal (3)

Decrease in secondary Heat Removal (4)

Mass/Energy Release from secondary Break (5)

LOSS of Flow (6)

Locked Rotor (7) over Pressure Protection (8)

Anticipated Transients Without Trip (9)

Loss of coolant Accident (10) small Loss of Coolant Accident (WCAP-9600)

(11)

Steam Generator Tube Rupture (12)

Radiological Assessments ATTACHMENT 1 Page 10 of 16

---~-- -------

Public Service Electric and Gas Shift Technical Advisor Training Program Module VII - Simulator Training -

Zion (2 weeks)

a.

Discuss, analyze and monitor the reactivity manipulations as delineated in the H. Denton letter dated March 28, 1980, and titled Qualifications of Reactor Operators.

b.

Accident/Transient Analysis.

c.

Degraded core cooling.

d.

Mitigating core damage.

e.

Plant operations.

ATTACHMENT l Page 11 of 16

'I*

Public Service Electric and Gas Shift Technical Advisor Training Program Module VIII - Reactor Engineer Training (5 weeks)

a.

Nuclear Fuel Design (1)

Core Design Characteristics (2)

Core Design Report (3)

Plant Application of Core Design Report (4)

Core Control and Load Follow

b.

Fuel Shipping and Accountability

c.

Fuel Handling

d.

Reactivity Computer

e.

Initial Reactor Startup Test Program

f.

Initial Core Loading and Load Power Test

g.

Power Ascension Testing

h.

Incore Nuclear Instrumentation

i. Incore Codes

j.

Incore workshop

k.

Plant Computer

1.

Computer System and Application

m.

NRC Regulations

n.

Licensing Issues - Plant and Reload

o.

Fuel Management and Fuel Cycle ATTACHMENT 1 Page 12 of 16

Public Service Electric and Gas Shift Technical Advisor Training Program Module IX -

Supplement to Westinghouse STA Program (3 weeks)

a.

Mitigating Core Damage (2 days)

(1)

Natural Circulation Following a Small Break LOCA (2)

Excore Source Range Detector Response (3)

Incore Thermocouples (4)

Incore Movable Detector Response (5)

Post-Accident Primary Chemistry (6)

Radiological Hazards of Sampling (7)

Methods of Determining Dose Rate Inside Containment (8)

Hydrogen Hazards During Severe Accidents (9)

Post-Accident Containment Environmental Effects (10)

Radiation Monitoring System Response

. (11)

Alternative Methods of Determining Vital Plant Parameters

b.

Electronics - Electrical Theory (2.5 weeks)

(1)

Basic DC Circuits and Laws (2)

Semi-Conductor Devices and Circuits

{a)

Physics and Conduction Properties

{b)

Devices

{c)

Circuits

{d)

Operational Amplifiers

{e)

Digital Electronics ATTACHMENT 1 Page 13 of 16

(3)

Instrumentation and Control Theory (a)

Control Loops (b)

Primary Elements and Detectors (c)

Secondary Elements (d)

Transducers and Transmitters (e)

Controllers (f)

Final Control Elements (g)

Modes and Methods (h)

Circuit Protection and Control Logic (4)

Electrical Sciences (a)

Motors

1)

Introduction

2)

DC Motors

3)

AC Motors (b)

Generators and Alternators

1)

Introduction

2)

DC Generators

3)

Alternators (c)

Transformers (d)

Switchgear

c.

STA Chemistry Training (1 week)

(1)

Introduction (2)

Corrosion Chemistry (3)

Reactor Coolant Chemistry (4)

Stearn Generator Chemistry Control ATTACHMENT 1 Page 14 of 16

circulating water chemistrY control 161 component cooling water chemistrY control 111 water Treatment Methods 101 chemical and volUll\\e control system l9)

Rad-~aste systems tlO)

RaaiochemistrY 1111 sources of RadioactivitY in a pWR 1121 Accident conditions 1131 Technical specifications ATTACHMENT l page 15 of 16

Public Service Electric and Gas Shift Technical Advisor Training Program Module X -

PSE&G Supervisory Training Program (2 weeks)

a.

Company Organization

b.

Planning Organizing and Controlling

c.

Communications

d.

Tagging Rules

e.

OSHA

f.

Motivation

g.

Equal Opportunity and Affirmative Action Activities

h.

Industrial Relations

i.

Leadership

.j.

Counseling and Appraising

k.

Administrative Reports

1.

Plant Security and Safety

m.

Time Management ATTACHMENT 1 Page 16 of 16

ATTACHMENT 2 TRAINING PROCEDURES SALEM GENERATING STATION N3C LICENSED OPEF.ATOR R!::G,UALIFICP.~'ION PROG::\\A~

ATTACHMENT 2 Page 1 of 24

I..

Formal Lectures/Self Study - the following subjects must be covered during each two year period.

Unit differences will be stressed,

l.

Principles of Reactor Theory and Operation.

2.

Plant Design and Operating Characteristics.

3.

Plant Instrumentation and Controls.

4.

Plant Safety, Emergency, and Protection Systems*

5.

Operating, Emergency, Abnormal and Administrative.

Procedures.

6.

Technical Specifications.

7.

Radiation Protection and Controls.

8.

Fuel Handling System and Core Parameters.

9.

A?plicable Portions of Title 10, Chapter I, Code of Federrl Regulations.

10. Current Topics/Applicaole LE~'s.

11. Eeat Transfer-Fluid Flow.

12. Mitigating Core Damage.

All N3.C licensed personnel must participate in tee ~e-qualification Progra=.

Most material will be ?rese~te~

via lectures on a schedule consistent with t~e O?erati~~

Departsent sc=edule.

Non-operators may attend a~y session, on differe~t deys i~

differe~t sessions, but ~ust atte~d or make U? all re~~irea classes wit~i~ the period cf ?rese~tatio~ ~or t~at lect~=e series.

It is ?Ossi~le t~at u~der cer~ei~ ci~cu~sta~ces e~ o;er~:or will ~iss ~!s shi~t's tra!~i~g c7:le.

!ver7 e~~ort s~o~ld ATTACHMENT 2 Page 2 of 24 Date 12/20/80 Re". __

4 __ _

be made to send him through with another group.

It is also possible that a non-operator could miss portions of a trai~-

ing segment.

That individual will be assigned to a prepared self study program culminating in the administration of a rela:ed lecture series test.

Every lecture series will be concluded with a test related to the subject matter covered.

Whenever practical it will i~-

elude questions concerning applicable Emergency Instructions.

A score of 80% or greater will constitute satisfactory per-formance.

Anyone scoring less than 80% will be assigned remedial training consisting of_ a prepared self-study progra~.

Co~pletion will be verified by the training staff.

A score of greater than 80% on the applicable section o~ t~e a~n~al examination allows that individual to skip the lect~res associated with the tested material.

He still m~s~ ?ass the End of Lectures Series Test and is subject to reoe~ial trai~i~g.

There may be times where heavy vacation sc~edules or o~e~ati~g conditions ~arrant the postponeoent of the Requalificatio~

?rogram.

Those cancelled lectures must be resched~led e.t t~e earliest opportunity.

Every a£te~?t should be ~ade to a!~e~e to the Annual Exa~ination.schedule.

to Manage~-Sale~ for rev!ew.

... -~ -'*: *-..

ATTACHMENT 2 Page 3 of 24 De.te

~e*r.

12/:0/80

Lecture Series Breakdown l,

Principles of Reactor Theory and Operations.

a.

Reactor Theory - in general, as covered in W On-Site Training Manual, Volume I,Section I.

b.

Reactor Characteristics

c.

Reactor Response

d.

ECP's, SDM's, I/M's

2.

Plant Design and Operating Characteristics

a.

Systems Review*

(1)

Reactor, Reactor Coolant (2)

Che~ical and Volume Control (3)

Residual Heat Removal (4)

Com?onent cooling Waste (Li~uid, Gaseous, Solid)

(5)

(6)

Spent ?~el Cooling

( 7)

Ventilation-F!B, Aux. 3ld.g., Co:ltrol Rao::.

(8)

Steam an~ Turbine (9)

Turbine Auxiliaries (10)

Cooling - Service Water, TAC (11)

Feed - Condensate - Aux. Feei (12)

Steam Ge~erator (13)

?ire Protection

b.

T:-e.::i.s ient and Accident Analysis /Integrated Plant Response

All do not.ha7e to ~e co7ered ~ith a ~or~al lecture.

ATTACHMENT 2 Page 4 of 24 Date pt?Q/80

?. e *1 *

3.

Plant Instrumentation and control

a.

Systems Review*

( 1)

N!S (2)

In-Core (3)

SGWLC-Feed Pump Speed control (4)

Pressurizer Pressure and Level, POPS (5)

Reactor Control (a)

FLRC (b)

Steam Dum-p (6)

Reactor Protection (7)

Radiation Monitoring (8) computer

~Al~ do not have to covered ~itb formal lecture.

4

?l*ot Safety, Emergency, aod Protection Syste~s.

a.

Systems Revie~

(1)

Safety Injection (2)

Containment-Containment Syste~s (3)

ECCS Controls (4)

Emergency power (5)

Accident A~a~7sis (6) zc:s Act~a':.io~

ATTACHMENT 2 Page 5 of 24

!)a,':.,:

l'.?./'.?.O/SO

i.e*r. __

4 __

. 5.

r 0.

Opeiating, Emergency, Abnormal, and Administrative Procedures.

a.

Review (1)

General OI's

( 2)

Emergency Instructions

( 3)

Abno!"mal Occurrences (4)

Operating Incidents

( 5 )

Administrative Procedures (6)

Emergency Plan Technical Specifications

a.

General Familiarity

b.

Bases

c.

De:!'initio:is

d.

License Limits

7.

Radiation Protection and Controls

a.

AP-24

b.

10CF~20

c.

!~strumentat!on ~a~iliarity

d.

Control Point Procedures

e.

Drills

8.

Fuel Ha:idling S:r_ste_!!l. and Co!"e ?ara:nete:::-s

a.

~ef~eli:g ?:::-oced~res

b.

Ref~el~~g Eazaris

c.

Core

?ara~eters-~!, ?uel Eur~u~, Effects of !~e!

3u: :::u.p.

ATTACHMENT 2 Page 6 of 24

~e*r.

12/20/80 4

I I

9.

Applicable Portions of Title 10, Chapter l

a.

QA Progra~ -

10CFR50

b.

10CFR19

c.

10CFR20

d.

10CFR55

10. Current Topics

a.

Facility Design Changes

b.

Revisions to Instructions and Procedures

c.

Operational Philosophy Changes

d.

Revie~ wea~ areas as detected

e.

LER's -

Sale~ and others

f.

Significa~t !vents at other Facilities

12. *~itigating Core Da~age.

Su~ject matter is presently being developed.

II.

Cn-the Job Training

1.

Beactivit7 Manipulations - during the two year period of the BC/SRO license, each indlvi~ual is re;~~re~ to perfor~ the fcllo~ing approved ~anipulations.

The starred ite~s ~ust oe ~erfor~ed annually.

i)a't.e 12/20/80 ATTACHMENT 2 Page 7 of 24 4

- ~-....

(l)

Plant or reactor startups to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established.

(2)

Plant shutdown.

(3)

Manual control of steam generators and/or feedwater during startup and shutdown.

(4)

Boration and/or dilution during power operation.

(5)

Any significant ( > 10%) power changes in man~al

rod control.

(6)

Any reactor power change of 10% or greater w~ere load change is perfor:ned with EHC in manual.

(7)

Loss of coolant including:

l.

significant steam generator leaks.

2.

inside and outside pri~ary contain:nent.

3.

large and small, including leak-rate determination.

4.

saturated reactor coolant system res:ponse (?~).

(8)

Loss of instr~:nent air (if simulated ?la=t specific).

( 9)

Loss of electrical power (and/ or de grad.e.::. ;io-.;e.. :-

sources).

(10)

Loss of core coolant flow/nat~ral circ~lat~o~.

(11)

Loss of condenser vac~u:i.

(12)

Loss of safety related service water.

(13)

Loss of *shutdown cooling.

( 14)

Loss of co:::i.pone:it cooli:ig sys't '*

~.,!..: ;ling to a:i i:idivid~al coo?one!l.t.

(15)

Loss of !l.C!':ial f'eed.-.n3."':er or =.o.:*::. __ :'ee-:.*.;e."':e::-

syste~ fai~~re.

ATTACHMENT 2 Page 8 of 24 12;20/SC I I

I I

(16)

(17)

(18)

(19)

(20)

(21)

(22)

(23)

(24)

(25)

(26)

Loss of all feedwater (normal and emergency).

Loss of a protective system channel.

Mispositioned control rod or rods (or rod drops)

Inability to drive control rods.

Conditions requiring use of emergency bore.tion or standby liquid control systen.

Fuel cladding failure or high activity i~

reactor coolant or offgas.

Turbine or generator trip.

Malfunction of automatic control syste~(s) which affect reactivity.

Malfunction of reactor coolant pressure/volu~e control system.

Reactor trip.

Main stea~ line brea~ (inside or outside :~=ta~~~e:

(27)

Nuclear instru~entation failure(s).

Reactor Operators must perfor~ the manipulation to ta~e cred.it.

Senior Reactor Operators and Shift Technical Advisors

may take credit for the manipulation if they direct or evaluate the control manipulations as they are performed.

The use of Tech~~cal Specifications should be =ex!-

~!zed during t~e nanipulations.

Mani~ulat!o~s s~all ~e doc~~ente1 in the Manipulations Log and be~o=e a ;a~t o~

t~e ~ndi7iduals training ~eco~d ~hen t~e fo~~ is ~~11.

ATTACHMENT 2 Page 9 of 24 Je.te

?.e*r.

12/20/8*

4

structions (EI).

An annual revie~ of each OI and EI

2.

Review of Applicable Operating (OI) and Emergency In-listed in Appendix A to this procedure is required.

There shall be 4 means of review for these procedures.

All 4 means need not be completed for each procedure.

a.

In di vid ual ReYiew - licensed ind iv id ua.l rev ie*,.; s -

documents by initialing A?pendix A.

b.

Lecture/Seminar - there shall be a lecture/se~ir.ar presented on all procedures listed in Appendix A on an annual basis.

c.

Simulator - all conditions and syste~s capable of simulating sy~ptoms/respcnse leading to the use o~

~alem E~ergency Instructions shall be utilize~ at the si~~lator currently under contract by ?S3~~

for operator training.

The Si~ula~or Instr~ctor The evaluation process will ~e de7e~opea ~Y t~e Senior Nuclear Training Super7isor a:d Ver.dor Re;re-senta.ti'7e.

d.

Sale~ Cont<ol Boa<d - eacb Licensee stall de~o~st<ate console.

o"'*~ 1... ~ * ~~

ATTACHMENT 2 Page 10 of 24

a.~e _ 12/"0/RQ

"?.e-;.

~--------------- ---===-:-::::~~~--

Training shall be documented on Appendix A o procedure for each individual licensee.

When the annual review is complete, Appendix A shall become part of his individual training folder.

For tracking purposes, a master OI-EI Chart showing all licensed individuals current OI-EI review status will be posted in the training area.

This chart shall be maintained by tbe training staff in a timely manner following the completion of required training.

3.

Facility Design Changes - the training staff is re-sponsible for scheduling of special training lectures and demonstration to ensure all operators are cogni-ze.nt of facility design and license changes.

4.

Simulator Training - the goal of si~~lator use is to maintain and/or upgrade operator profic:e~cy d~ring normal transient and accident situations Opera to!"

must attain a minimum grade of SAT in eac~ area lis~ed on t~e NUS Simulate~ Evaluatio~ Report (Appe~dix C) for those events si~ulated.

Re~edial tre.i~ing to upgrade UNSAT grades is required.

This me.y be a.c-complished at t~e simulator or by the Sale~ tra.i~i~g sta.!~,

a~ the ea~liest o~~ortunity.

'.,., -r s i..,.. ' "'... o.......,.. a ;,... ~,.. g..,.. ~...

°'"' e io....,.....,.... -r"" ~

~ -r

~T-:;. ~

.LJ

-~

.J-

...... "-~""

.,;J

~---

... ~-

opera~or Licensi:g Era~c~ prior to o~~ic!a~ inc:~s!c~

ATTACHMENT 2 Page 11 of 24 De.~e

?.e*r.

12/20/80

I~I.

Evaluation -

4 methods

1.

Job Performance - all personnel are evaluated at least annually in the performance of tbe duties for their assigned position.

It is the responsibility of the Chief and Operating Engineers to ensure that only those personnel who demonstrate capability to bandle abnormal and emergency situations, are safety-minded, and have the general ability to disch~~£e licensed duties are allowed to perform licensed duties.

T!:ey may, at their discretion, remove an individ~al from licensed duties until the deficiency is corrected.

2.

Observation, by supervisors and instructors, o~ t~eir action in abnor~al and e~ergency situations at the facility or si~ulator.

.. 3.

Lecture Series Examinations - at the end cf eac~ le~-

ture series an exa~ination will be ad=inistere~ related to the material presented and applica~le emerge~cy a~t ab.:J.or:na.l situations.

A graC.e of > 80% "Will constit-.ite sa~isfactory perfor:nance for that lecture series.

Anyone <80% will ~ave to perfor.?J. re:nedial tra~:i~g.;...

const!tute satis~actcry ;er~or=a.:ce for that lecture series.

All licensee. ?erson~el ~ust ;ass the Le:t~:e ATTACHMENT 2 Page 12 of 24

_Ll.C0/80

\\e *r. -----

4 I

/

I

Series tests whetb~r they attended t~e lectu~e or not.

A grade of< 803 on this test will not disqualify a man from performing his licensed duties.

Assignme:::t to Remedial Training will be reported to the Assistant-to-Manager.

4.

Annual Written Examination -

An annual exa~ination will be administered.

The content and scope of this examination will parallel and utilize the sa~e ti~e requirement as that of an ~C licensing exa~.

Licensei individuals must take all sectio~s t~at are a.pplica~le to their license.

The approved passing grade for the A~nual 3xa~ is Be~

o v e r a 11, wit h no s e ct i on < 7 0%.

Ar.. y i n d !. -; i :1. 'J. a l :'a :!. l i ::

below this mini~u~ grade shall be reportei to tbe Assista.nt-to-Ma~ager and be removed fro~ lice:lsei duties.

They will then be assigned to a specifically des:gne~

Accelerated Re~~alification Progra~.

.Ar..yor.e scoring <

80% on any section o:' t!;.e.:..nr..::.al Examination will u~dergo re~edial trai~!~g i~ *ea~ areas as s~ecified by an SRO licensed trai~i~g staf~ ce:~er.

Satisfactory co~pletion of this re~edial trai~!~g cay ATTACHMENT 2 Page 13 of 24

ate

~e"T.

12120/~Q

_J_

I I

Accelerated Requalification Program - Individuals falling below the minimum passing grade on the Annual Examination will be immediately assigned to an Accelerated Requalifi-cation Program.

The training staff will develop a

formal Accelerated Training Course, both lecture and self study, that is specifically designed to correct the indi-vidual's deficiencies.

This Accelerated Training Course will be documentec by the training staff and approved by the Se~ior Nuclear Training Supervisor.

At the conclusion of the course, the indi~idual will be ad=inistered a vritten examination.

He/She must attain greater than the minimu~ ap~licable score in order to be reassigned to normal operator duties.

At t~e difcretion of the Senior Nuclear Training Supervisor re -e:<a:iina. t ion may only consist of a~ individual section(s), de?e~dent on his/her proximity to the overall 80% passing re~uire~e~t.

P~rticipation in the Acce~erated Requalification ?rogra=

does not ~xempt the individual from the normal re~~ali~!c~~~o~

~rogra~ lecture series.

~e~res~er Traini~g - If a licensee has ~ot bee~ actively per~or~ing the ~~nctio~s of an RO or SRO

~or a period c~

four ~onths or longer, he shall be enrollei in a

~cdi~ie~

ATTACHMENT 2 Page 14 of 24

s.-: e 12/20180

~.:..

while not performing licensed duties.

He will then, prior to resuming activities licensed by t~e NRC, demonstrate that his knowledge and understanding of facility operatio~s and administration are satisfactory.

This de~onstration shall include an NRC type RO or SRO written examination and walk-through.

The results of this demonstration are su~-

mitted to the NRC -with a request to allow the indi*ridual to resume licensed duties.

IV.

Requalification Prog=am Administration

1.

Overall responsibility for the Requalificatioc ?rogra~

rests with the Senior Nuclear Training Supervisor in co-ordination with the Sale~ Assistant-to-Ma~a.ger.

T::iese include:

a.

Co-or~ination o~ prcgra~ scheduling

b.

Co~pliance with regulations

c.

Instructor scheduli~g

d.

?rogra~ docume~taticn

e.

A~n~al exa~i~ation pre~aration

-..... ~-4i""l-c--*.... __ ~*

The indivi~ual (s) responsible for makin5 ~;

""".,..'°~:-

~

.... !"'\\,...

>J _....,_

a..:i a *1erage g!'ade of 853.

A =i.ax:..:::iu.::i. c :' 3 ;e:-sc=.s do se g:ie n. t.

Ja.-:e 12.'~0/SC ATTACHMENT 2 Page 15 of 24 3.e-r.

g.

Insure that license holders are meeting requirements for license renewal.

h.

Periodic review of the reactivity manipulations log.

3.

Records

a.

General Training File -

The Gen~ral Training File contains all infor~ation pertaining to the adminis-tration and execution of the requalification progra=.

In particular, it contains:

(1)

A chronological ~og of major requalification events.

Examples of major events are:

-Start of t~e requalification progra~.

-Administration of the Ann~al ~xa~i~ation.

-Sta=~ of a trai~ing seg~ent lect~re series.

-A~~inistatio~ of a lecture series test.

-Enrollme~t and final disposition of an individual in an accelerated req~~lificatio~

progra:::i.

-Execution of a drill.

-Simulator training.

-Start of a special training lect~re de~onstration series.

-Start o~ a ne~ requalificatio~ ;rcgra~ cy~le.

Sale:n.

ATTACHMENT 2 Page 16 of 24

,.,,._ Q i r -

--- _..,=

_,::i

a.-:e 12/20/SO 4

(2)

A copy of the current annual examination with approved answers.

(3)

A list of all individuals (and their grades) to whom the annual exa~ination was ad~inistere~.

(4)

A chronological list of all lectures and instructors.

This includes regular, re~edial, accelerated progra~s, facility design changes, and operating and emergency instruction lectures (5)

A copy of any post-lecture series tes: and answers.

(6)

Attendance for all si~ulator training.

(7)

Overall attendance record in the Requalificatio~

Program.

(8)

An overview of the sc~edule marked up to s~ow a participants status in the program.

b.

An individual training folder will be maintai~ed by t~~

training staff for each licensed operator for the life of his license.

The folder will contain:

(1)

An initial traini~g ~refile.

(2)

The graded copy of his annual exa=inatio~.

(3)

T~e graded co~y of ~is post-lecture se~:es tes (4)

The results of a~y re~edial traini~g.

(5)

Documentation of assig:~e~t to a~e outli~e c!

any Accelerated Requali!icetio~ ?rogram (A~?).

(6)

The graded copy o~ the AR? re-exa=i~atio~ e~d fi~al dis?osition.

( 7)

( 3)

~ua:i!icatio~ Cari !or ov~ra:l a~d ~~e~ge~~7 :~s~~~c::o~s.

ATTACHMENT 2 Page 17 of 24

':>1.. -*

Ja'";e 12/20/80

?.e*:.

4

Note:

(9)

Documentation of any simulator training.

Assignment to and completion of annual examination re~edial training.

(10)

General and Individual Files may be microfilmed and destroyed folloving each 2 year cycle/or e.t the ter~inatio~

of a license.

No change may be me.de to the e.pprovea reque.lificatio~

PROGRAM CH..A..NGES training progra~ that decreases the sco?e, ti~* allotted for the prograo, or frequency in conducting di~ferent parts of tbe progr*=, mnless that chanse is re7ieve! e~d ATTACHMENT 2 Page 18 of 24

!)e.te i:/:?.0/80

\\e '!T

4

3 3

) a

~

'1

~

-3 v

tj

.II 1

!)

)

-ti 0

°'

rt rD I-'

N N

0 co 0

l~nm t:<JtHlCV E. I.

J, s

CH E.I.

J, s

CR 4.0

4. l 4.2 4.3 4.4 4.5 4.6 4.1 4.0 4. 'J

4. l 0 l.t:*Jt!lul -*

I. -

1.. :1*1111e,

~; ** Sl11111l<11.. 1*, Cll -

l"u11t.rol l{onm 11......,,..11 )( l\\

J.i 1:en:->ed Operdtor Procedure Rrview

2.

3.

4.

Appendix B to the Licensed Operator Requalification Program:

'Requalification Program for Licensed Operations Instructors.

Licensed operations instructors shall attend or administer all aspects of the Licensed Operator Requalification Program.

Those individuals who make up and/or grade the quizzes and major examinations are exempt from this ~ortion of the requalification program.

A maximum of three instructors may be exempted from testing.

Licensed operations instructors shall participate in si~ula~or training to the same level as any other licensed perso~~el.

The assignments of licensed operations instructors will be rotated by the Senior Nuclear Training Supervisor such that they maintain proficiency in all areas of licensed operator trai~i~~.

Each licensed operations instructor should stand at leas~ 1 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months watch each month.

This shall be as a reactor operator or control room senior reactor operator in order to maintair. pro-ficiency and contact with plant operations.

Watchstanding may be waived when neither Salem unit is in Mode 1 or 2, as defined in Technical Specifications.

Each watch will be docu~e~ted in the instructors individual folder by use of a summary report.

This summary should include a description of major events, re-activity manipulatio~s, logkeeping, etc.

5.

Licensed operations instructors will attend professionel meet-ings, seminars, etc., as determined necessary by t~e ~rainir-g Engineer.

6.

The Senior Nuclear Trai~ing Supervisor will scted~!e instr~ctor seminars, on a periodic basis, to introduce new or revise~ =a-terial, design changes, significant L~~*s or re-i~for:e ol~

material.

ATTACHMENT 2 Page 20 of 24 Date 12 /20/80 Rev. --~~-4~~~

SURRY TR.~"rNING CENT.t:.K........ * *-

APPENDIX C SIMULATOR EVALUATION REPORT ENT:

MAL-FUNC-POSITION EVALUATION TIONS u-r EVOLUTION USED SUPV RO BOP OBS E

VG SAT Si

1. Plant/Reactor Startup

2. Plant Shutdown

3. Manual S/G level control 4. Boration/Dilution during power operation

5. Powe::- changes

(~10%) in

,r manual rod control I

6. Power changes

(~ 10%) in load limit control

(

~

~*..

I

7. Loss of coolant (l) SG tube leak (2) inside/outside containment I'

I I

( 3) leak rate cete::-mination ii I

I 11

( 4) saturated RCS response

'I 1 I I

8

Loss of inst:::-Ur.le:i.t.

l air

9. Loss o= elect=:.cal 1*

I I

powe.:-

I I,

I I

0. Loss of coola:it f lcw/nat'.lral I

ci.=culation I

I JI '!'TAC HMEN'I 2 Pa je 21 of,>4 I

EVOLUTION 1.,.1,,,.,J........, ---

ll. Loss of condenser vacuum l

oss of service water

13. Loss of shutdown cooling

14. Loss of component cooling

15. Loss of normal f eedwater

16. Loss of all f eedwater

17. Loss of protective system channel

18. Mispositioned rod/rod drop

19. Inability to drive c:mtrol rods

~o. Emergency boration

. Fuel clad failure/High RCS activity

22. Turbine/Generator trip

23. Auto control syste.-n malfunction which affects reactivity I

I

24. Malfunction of RCS pressure/

I I

volume control syste.rt1 I

25. Reactor trip I

r

26. Main stea.-n line break I

l I

I I

I

27. NIS failure I

I I

(continued on Page three)

ATTACHMEN 2

Page r2 2 o::E 24 I

... ~,

'NALUATION KEY eel lent Very Good Satisfactory Unsatisfactory

- Performance and knowledge far above the average.

No weakness noted in performance or knowledge.

Only minor weakness noted in performance or knowledge.

Serious weakness noted in performance or knowledge.

OVER~LL EVALUATION Should assess the student 1 s :

(1) Level 6f understanding of overall plant response (2).Knowledge of individual systems involved (3) Ability to identify accidents or malfunctions and take proper action in a reasonable time (4) Adherence to procedures RE~.ARKS:

Instructor Signature Page 23 of 24 NL.JS ccr:;pc;=:ATl(

..... -..... -......... -......... ~

-***~.

Appendix D to the Licensed Operator Requalification Program:

Requalif ication for Shift Technical Advisors

1.

Shift Technical Advisors (STA) shall participate in all phases of the Licensed Operator Requalif ication Program

.except for the annual examination requirement.

The waiver of the annual examination requirement is not applicable for STA's holding a current reactor or senior operator license.

2.

Quiz and lecture series grade criteria requirements are applicable.

Attendance requirements are the same as for licensed personnel.

3.

The Refresher Training requirements on pages 13 and 14 are not applicable unless the STA is currently licensed for the facility.

Individuals who have not performed the STA function for less than 30 days shall recieve training sufficient to ensure cognizance of procedure and plant design changes.

Individuals who have not performed the STA function for less than 6 months shall, prior to assuming the duties of the position, complete a special training pro-gram that is tailored to the individual's needs.

This program must pe approved by the Chief Engineer and Senior Nuclear Training Supervisor.

As a minimum, it shall cover all emergency instructions, design changes within the last 6 months and any necessary simulator training.

It is the responsibility of the Chief or Operating Engineer to inform the Senior Nuclear Training Supervisor when the time requirements, as outlined here, are exceeded.

ATTACHMENT 2 Page 24 of 24

Date:

Rev.:

12/20/80 4

DEVELOPMENT OF SHIFT SUPERVIsm-; ~~4'~

- ~

- - ~~... *I.*

graduate engineer hired as new employee t

undergo STA training (approx. 45 weeks)

J assigned to shift for 3 mos. as Assistant Reactor Operator

.l further training in preparation for NRC reactor operator examination J

acquire reactor operator license l

shift support as Assistant-to-Senior Shift Supervisor (approx. 1 year)

I further training in preparation for NRC Senior Operator Examination l

acquire senior reactor operator license 1

promoted to Shift Supervisor - Engineer ATTACHMENT 3 Page 1 of 1

Educational &

Comparison of PSE&G STA Program to the April 1980 INPO Guideline Training Requirements Guideline 6.1.1 Education

a. Mathematics Total of 270

b. Inorganic Chemistry contact hours

c. Physics 6.1.2 College-level Fundamental Education

a. Mathematics 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months

b. Reactor Theory 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months

c. Reactor Chemistry 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months

d. Nuclear Materials 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months

e. Thermal Sciences 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months

f. Electrical Sciences 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months

g. Nuclear Instrumentation and 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months Control

h. Nuclear Radiation Protection 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months and Health Physics 6.2 Applied Fundamentals: Plant Specific

a. Plant specific Reactor Technology

b. Plant Chemistry & Control

c. Reactor Instrumentation & Control Total of 120

d. Reactor Plant Materials hours

e. Reactor Plant Thermal Cycle

.3 Management/Supervisory Skills

a. Leadership

b. Interpersonal Communications

c. Motivation of Personnel

d. Problems and Decision Making 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months

e. Command Responsibility & Limits

f. Stress

g. Human Behavior 6.4 Plant Systems 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months 6.5 Administrative Controls 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months 6.6 General Operation Procedures 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months 6.7 Transient & Accident Analysis 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months 6.8 Simulator Training 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 6.9 Annual Requalification.

80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months Page 1 of 1 PSE&G Program Covered by Engineering degree Covered by degree 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months 40 hours 0 hours0 days 0 hours 0 weeks 0 months Covered by degree and by 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in program 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months 11 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 88 hours 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months 360 hours 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months Addressed at simulator 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 80 hours 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months

1 NUREG-0578 Section 2.1.7.A

1.

The Salem SER (Supplem~nt No. 4 - April, 1980, Pages II.E-2 and II.E-3) states that the Salem Unit No. 2 AFW initiation circuitry meets the NUREG-0578 short-term re-quirements.

Therefore, PSE&G has already provided suf-ficient information to demonstrate that the Salem AFW System design meets the intent of paragraphs 4.1,.4.2, 4.9, 4.10, and 4.17 of IEEE Standard 279-1971.

Informa-tion previously submitted to the Staff is contained in References A and B.

Logic Drawings 221054-B-9545-2, 221056-B-9545-3, and 221064-B-9545-3 of the Salem Protection System depict the bypasses, indication of bypasses and channel inde-pendence of the AFW automatic initiation ci~uitry.

These drawings were submitted to the Staff at the January 3, 1980 meeting in Bethesda, Maryland.

In addition, "the February 4, 1980 submittal to the Staff (Reference A), contains a description of the system level annunciation available in the control room and the administrative controls used to alert the operators to equipment unavailability.

The above mentioned ref-erences include the necessary information to substan-tiate that the Salem AFW System design meets the intent of Paragraphs 4.6, 4.11, 4.12, and 4.13 of IEEE Standard 279-1971.

The functional diagrams submitted to the Staff on February 4, 1980 (see Reference A) and the logic dia-grams submitted at the January 3, 1980 meeting in Bethesda, Maryland, reveal the interactions between con-trol and protection functions for the Salem AFW System.

It can be seen from the drawings and the Reference A re-sponse that the Salem AFW System design meets the intent of paragraph 4.7 of IEEE Standard 279-1971.

The Salem SER (Supplement No. 4 - April, 1980) Page II-E-2 states:

"The auxiliary feedwater (AFW) system for Salem Unit No. 2 was designed as a safety-related system, aside and apart from any TMI-related require-ments imposed subsequently by the NRC."

The components used in the AFW System were purchased as safety-related equipment and subject to the Quality Assurance program.

The intent of paragraphs 4.3 and 4.4 of IEEE Standard 279-1971 has been met in the original design (and any subsequent design modifications) of the system

GN 08/2

2.

It is our position that no deviations from the above in-dicated paragraphs of IEEE Standard 279-1971 exist in the Salem AFW System design.

3.

A description of the AFW System is contained in Section 10.2.l.2 of the Salem FSAR.

A brief description of pro-tection circuit isolation is provided in the Salem FSAR on pages 7.2-6 and 7.2-16.

The Salem FSAR references two Westinghouse documents for a detailed discussion of isolation circuits.

These two reports are: WCAP-7824, "Isolation Test-Process Instrumentation Isolation Amp-lifier, Westinghouse Computer and Instrumentation Divi-sion, Model 131-110" and WCAP-7672, "Solid State Logic Protection System Description."

Logic diagrams, electrical schematics, piping diagrams, and instrument schematics for the Salem AFW System were submitted to the Staff at the January 3, 1980 meeting in Bethesda, Maryland.

Functional diagrams were submitted to the Staff on February 4, 1980 (see Reference A).

Al-though the drawings were submitted for the Unit No. 1 review, they are applicable to Unit No. 2.

The Salem AFW System surveillance procedures which dem-onstrate the testability of automatic initiation for Unit No. 1 were submitted to the Staff on December 27, 1979 (Reference B).

Although written for Unit No. 1, the procedures are applicable to Unit No. 2.

Technical specifications for the Unit 2 Auxiliary Feed-water System have been issued

GN 08/3

NUREG-0578 Section 2.1.7.B

1.

The Salem SER (Supplement No. 4 - April, 1980, Page II.E-4) states that the Salem AFW instrumentation design satisfies the control grade requirements specified in NUREG-0578.

Therefore, PSE&G has already provided suf-ficient information to demonstrate that the Salem AFW instrumentation design meets the intent of paragraphs 4.1, 4.2, 4.9, and 4.10 of IEEE Standard 279-1971.

In-formation previously submitted to the Staff is contained in References A and B.

The Salem AFW System is equipped with one AFW flow indi-cator, one wide-range level indicator, and three narrow-range level indicators for each steam generator.

The Salem SER (Page II.E-4) states:

"Taken together, the combined (direct and indirect) AFW flow indication cap-abilities satisfies the single failure criteria."

This evaluation by the Staff is due in part to the fact that the Salem plant is equipped with four vit~l instrument buses and, that the vital instrument buses' normal power source is backed-up by batteries.

The Salem AFW system instrumentation arrangement satisfies the NRC criteria for four loop plants with four vital buses.

The instrumentation provided for the AFW System can be seen on the instrument and functional diagrams which were submitted to the Staff on January 3, 1980 and February 4, 1980 (see Reference A), respectively.

As a result of these considerations, the Salem AFW System de-sign meets the intent of paragraph 4.6 of IEEE Standard 279-1971.

Steam generator narrow range level indicators have a demonstrated accuracy of + 20% for post-accident condi-tions and + 10% for auto-1nitiation (short-term, first hour).

Westinghouse documents are available to substan-tiate the accuracies stated.

The remaining instrumenta-tion is presently being reviewed by the Staff for ac-ceptable environmental qualification.

The AFW flow in-dication channels possess an accuracy of + 2% as stated in the SER (Supplement No. 4 - April, 1980).

In addition, the Salem AFW system was designed as a safety-related system.

The components used were pur-chased as safety-related and were subject to the GN 08/4

Quality Assurance program.

Consequently, the intent of paragraphs 4.3 and 4.4 of IEEE Standard 279-1971 have been conformed with in the original design (and any sub-sequent design modifications) of the system.

The functional diagrams submitted to the Staff on February 4, 1980 (see Reference A) and the logic dia-grams submitted at the January 3, 1980 meeting in Bethesda, Maryland reveal the interactions between con-trol and protection functions for the Salem AFW System.

It can be seen from the drawings and the Reference A re-sponse that the Salem AFW System design meets the intent of paragraph 4.7 of IEEE Standard 279-1971.

2.

It is our position that no deviations from the above in-dicated paragraphs of IEEE Standard 279-1971 exist in the Salem AFW System design.

3.

A description of the AFW system is contained in Section 10.2.1.2 of the Salem FSAR.

A brief description of pro-tection circuit isolation is provided in the Salem FSAR on pages 7.2-6 and 7.2-16.

The Salem FSAR references two Westinghouse documents for a detailed discussion of isolation circuits.

These two reports are: WCAP-7824, "Isolation Test-Process Instrumentation Isolation Amp-lifier, Westinghouse Computer and Instrumentation Divi-sion, Model 131-110" and WCAP-7672, "Solid State Logic Protection System Description."

Logic diagrams, electrical schematics, piping diagrams, and instrument schematics for the Salem AFW System were submitted to the Staff at the January 3, 1980 meeting in Bethesda, Maryland.

Functional diagrams were submitted to the Staff on February 4, 1980 (see Reference A).

Al-though the drawings were submitted for the Unit No. 1 review, they are applicable to Unit No. 2.

The Salem AFW System surveillance procedures which dem-onstrate the testability of automatic initiation through signals generated by AFW related instrumentation for Unit No. l was submitted to the Staff on December 27, 1979 (Reference B).

Although written for Unit No. 1, the procedures are applicable to Unit No. 2.

Technical Specifications for the Unit 2 Auxiliary Feed-water System have been issued

RB:gs GN 08 1/5

Automatic Initiation of the Auxiliary Feedwater System for PWRs (Section 2.1.7.a)

NRC Position Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwa~er system, the following requirements shall be implemented in the short term:

1.

The design shall provide for the automatic initiation of the auxiliary feedwater system.

2.

The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of *auxiliary feedwater system function.

3.

Testability of the initiating signals. and circuits shall be a feature of the design.

4.

The initiating signals.and circuits shall be powered from the emergency buses.

5.

Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.

6.

The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.

7.

The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFHs from the control room.

In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements.

Responses The Auxiliary Feedwater ~ystem is described in Section 10.2.1.2 of the FSAR.

The systen is designed to Class IE criteria and is powered from the emergency power source.

M P79 54 01/26 Salem 1 & 2

.JAN i

1980

Automatic initiation of the Auxiliary Feedwater System is provided by the following signals.

Motor Driven Pumps

a.

Loss of Offsite Power

b.

Loss of Main Feed

c.

Low-Low Level in One Steam Generator

d.

Safeguards Sequence Signal Turbine Driven Pump

a.

Loss of Offsite Power

b.

Low-Low Level in Two Steam Generators

c.

4kV Bus Undervoltage Manual initiation of the systems may be accomplished from either the Control Room, or locally at the pumps.

The system and its components are designed for single failure*

considerations and are testable.

M P79 54 01/27 Salem l & 2 JAN 1 1980

ITEM II.E.1.2.:

AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION

RESPONSE

The attached information addresses the requirements of Item II.E.1.2., for Salem Unit No. 2.

It has been previously submitted during the review of the Salem Unit No. 1 AFW System.

The Salem SER, Supplement 4 contains the results of the NRC review of this information.

Auxiliary Feedwatcr Flow Indication to Steam Generators for PWHs (Section 2.1.7.b)

NRC Position Consistent with satisfying the reguirem~nts set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perf*)rm its intended function, the following requirements shall be implemented:

1.

Safety-3rade indication of auxiliary feedwater flow to each steam generator shall be provided in the contrql room.

2.

The auxiliary feedwater flow ins~rument channels shall b~ powered from the emergEncy buses consistent with satisfying the emergency power diversity requirements of the auxiliary f~~dwater system set forth iri Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

Response

Safety-grade indication of auxiliary feedwater flow to each 1

steam generator is provided in the Control Room.

These I

indicating cha~nels are designed to the same criteria as the protection system indicators.

One testable flow instrument with an accuracy on the order of +/- 2%, is provided for ~ach steam generator.

In addition, three level instruments are provided for each steam generator.

The instruments are all powered from the vital buses, seismically qualified with environmental qtlalif ication for the level instruments which are located inside the containment *

. M P79 54 01/28 Salem 1 & 2 MAR ~ o 1980

Assurance of sufficient water being provided to the steam generators is of primary concern.

This is accomplished by control of valve demand with steam generator level indica-tion.

Present indication of pump operation, valve demand/

position, auxiliary feedwater flow (one/steam generator),

auxiliary feedwater discharge pressure and steam generator level (three/steam generator) is adequate to meet the information requirements necessary to assure appropriate operator action.

All of the above equipment is powered from vital buses, and is considered adquate to meet short and long term requirements.

I -

M P79 54 01/29

-2la-Salem l & 2

...L l :. -.

-~-- -*- *--

Mr. A. Schwencer, Chief Operating Reactor Branch tl Diviaion of Operatinq Reactor*

v. s. Nuclear Regulatory Comm11*ion Wa1hinqton, D. c.

20555 Gentle.mans BAL.EM GENERATING STATION WIT MO. l RESPONSES TO WRC INQUIRIES AT AUXILIARY PEEDWATER SYSTEM MEETING February 4, 1110 Jtepre1entativea from yonr *taff and PSE'G met in Bethe*da on

- January J, 1980 to review the Salem response to NUREG 0578, Item 2.1.7& and 2.l.7b on the auxiliary feedvater *Y*tem.

At thi1 meeting, a number of item* vere requested to be *ubmitted by PSE*G to a1aiat in your revie1!' of Salem.

Attached are the respon*e* to the inquiries.

'1Yo*eopie1 of the d.ravinq* re-que1ted are alao attached.

Or..,.....

d 1gm~.:... s~="

F. P. Llbri=zi.

General Mdllaqer -

Electric Production

(

I SALEM GENERATING STATION UNIT NO. l INFORMATION SUBMITTAL TO RESPOND TO NRC INQUIRIES FROM AUXILIARY FEEDWATER MEETING JANUARY 3, 1980

1.

Submit functional drawings showing the isolation devices for protection and control circuits.

RESPONSE

Two copies of the following PSE&G drawings for Salem are attached:

220431-B-9524-3 220433-B-9524-3 220435-B-9524-3 220437-B-9524-3 Functional Diagram -

No. 11 & 21 Stea~ Generator Control and Protec-tion Levels.

Functional Diagram -

No. 12 and 22 Stearn Generator Control and Protec-tion Levels.

Functional Diagram -

No. 13 and 23 Stearn Generator Control and Protec-tion Levels.

Functional Diagram -

No. 14 and 24 Steam Generator Control and Protec-t ion Level?,

These drawings show the steam generator narro~ range level initiating signals circuitry for reactor trip and auxiliary feedwater actuation and their use in feedwater control.

Signal isolators are shown in the circuits for separation between the protection and control portions of the circuits.

2.

Provid~ documentation references describing the initiat-ing circuit isolation devices and fault protectior..

RESPO~SE:

A brief description of protection circuit isolation is provided in the Salem FSAR on Page 7.2-6 and 7.2-16.

The Salem FSAR references two Westinghouse documents for a detailed discussion of isolation cir-cuits.

These two reports are WCAP-7824, *isolation Test Process Instrumentation Isolation Amplifier, Westing-house Computer and Instrumentation Division, Model 131-llo* and WCAP-7672, *solid State Logic Protection System Description*.

(..-

l.

Submit the 125 volt DC control schematics for the motor operated pumps.

RESPONSE

Two copies of the following PSE~G drawings for Salem are attached.:

203312-B-9768-13 203315-B-9769-13 Control Schematic -

No. 11 and 21 Auxiliary Feedwater Pumps Control Schematic -

No. 12 and 22 Auxiliary Feedwater Pumps 4a Provide information on minimum requirements for auxil-iary feedwater system operation as defined in safety analyses.

RESPONSE

The Salem FSAR establishes the safety re-quirements for the auxiliary feedwater system in the

Accident Analysis*, Section 14.

In the loss of normal feedwater transient, the safety analysis of the incident was performed assuming only one motor driven auxiliary feedwat~r pump is operable deliv-ering flow to two stearr. generators (refer to FSAR Page 14.1-38).

In the f eedwa ter 1"i'rie rupture ace ident, the safety analysis of the incident was performed assuming auxiliary feedwater initiation with a flow of 426 gpr..

{refer to FSAR Page 14.3-22).

The results of both analyees demonstrated that the plar.t is maintained in a safe condition.

As.stated at the meeting and presented in the Salem FSAR Section 10.2, the two motor driven pumps are each rated at 440 gprr. and the one turbine driven pump is rated at 880 gprn.

The design of the automatic, as well as the manual initia-tion circuits of the auxiliary feedwater system at Salem, is adequate to assure that the minimum require-ments are met.

In addition, the Salem Unit 11 design was reviewed by the NRC staff and found acceptable on the basis of mini-mum requirements and single failure criterion to assure that a minimum auxiliary feedwater flow of 440 gpm is delivered to two steam generators.

The results of that review were documented on Pages 10-S to 10-7 of the AEC (NRC) Safety Evaluation of the Salem Nuclear Generating Station, Units l and 2, dated October 11, 1974.

(

- ~ --- - -- -

-~..

5.

Provide a description of the system level annunciation when portions of the auxiliary feedwater system are out of service.

RESPONSE

Administrative controls are utilized for alerting operators of the inoperability of the auxiliary feedwater system due to maintenance/testing of equip-ment.

The unavailable equipment log and tagging request procedure assure that the operators are aware of the out-of-service status for the system.

Technical speci-fications have been established which limit the out-of-service time for the system.

In addition, individual alarms are printed out on the Salem auxiliary annunciator system in the Control Room alerting the operator to conditions/potential conditions in which equipment of the auxiliary feedwater system may not operate properly when called upon.

Such alarms would result in investigation or confirmation of the status of the auxiliary feedwater system.

The alarms provided are indicated on Table 1.

The combination of administrative controls and the above noted alarms provide the opet"ator with sufficient infor-mation to determine the operability status of the auxil-iary feedwater system.

6.

Provide qualification documentation references for the following specified auxiliary feedwater system equipment:

Transmitter (low level)

Logic Equipment Instrument RacY.s Actuation Device

RESPONSE

Table 2 lists an example of each item re-quired with data concerning equipment qualification.

Similar information was previously provided in the re-sponse to FSAR questions 7.18 and 7.30.

Only the steare generator narrow rang~ level transmitters have undergone environmental qualification testing.

The remaining items are not located in areas subject to an adverse en-vironment (provided with Class lE ventiliation system) or are not required to operate due to redundant equip-ment in other areas if an adverse environment results because of high energy line breaks.

~

.. *--... -~--.:.. -....

7.

Revise Salem response to NUREG 0578, Item 2.l.7b to in-dicate the use of the wide range steam generator levels as a backup indication for auxiliary feedwater flow.

RESPONSE

The response to Item 2.l.7b will be revised to reflect the operator's use of the following parame-ters for the determination of proper auxiliary feedwater system operation:

auxiliary feedwater flow stea~ generator narrow range levels steam generator wide range levels auxiliary feedwater valve positions auxiliary feedwater pump discharge pressure auxiliary feedwater pump status JPG: Jb H?l2 1/4 r

(

SALEM GENERATING STATION UNIT NO. 1 RESPONSE TO NRC/AFW MEETING 1/3/M Tab1e 1 Aux111ary Annunciator A1anns for Auxiliary Feedwater Systerr Alanns due to testing of protection system 1ogic Train A - No. 11 Aux. FW Pump Start Blocked Train A - No. 13 Aux. FW P"1lp Start Blocked Train B - No. 12 Aux. FW P"1lp Start Blocked Train B - No. 13 Aux. FW Pump Start Blocked Alarms for PllTlP 1noperability No. 11 Aux. FW Pump

~ Loss of 125V DC No. 11 Aux. FW P"11p - Loss of 28V DC No. 12 Aux. FW Pump - Loss of 125V DC No. 12 Aux. FW Pump - Loss of 28V DC No. 13 Aux. FW Pump & Turbine - Loss of 125V DC No. 13 Aux. FW Pump & Turbine - Loss of 2BV DC No. 13 Aux. FW Pump Speed Control - Loss of 115V AC A1anns for valve inoperability Aux. FW Valve llAFll - Out of positioniLoss of llSV AC Aux. FW Valve 12AF11 - Out of position/Loss of llSV AC Aux. FW Valve 13AF11 - Out of position/Loss of llSV AC Aux. FW Valve 14AF11 - Out of po~ition/Loss of llSV AC Aux. FW Valves 11-14AF11 - Loss of 28V DC Aux. FW Va1ves 11-12AF21 - Out of position/Loss of 115V AC Aux. FW Valves 11-12AF21 - Loss of 28V DC AJ.Jx. FW Valves 13*14AF21 - Out of position/loss of 115V AC Aux. FW Valves 13-14AF21 - Loss of 28V DC Alanns for misce11aneous functi~ns Aux. FW Temp & Suction Alarms - Loss of 125V DC Aux. FW Temp & Suction A1arms - Loss of 28V DC

I**

(

Alanns for miscellaneous functions (Cont'd)

Aux. FW Tank Level Alanns - Loss of 12SV DC Aux. FW Tank Level Alanns - Loss of 2BV DC NOTE:

The initiating signals (low steam generator level) are not bypassed during testing but are placed in the tripped mode - refer to Technical Specifications.

A description of the auxi1 iary annunciator system 1s presented in the Salem FSAR Response to Question 7.7.

-****--*------=--...........-~*----

SAi.Eli! G£M£AATING STATTOtt 1'tll..:>. l RESPONSE TU ltRC/AFW ME£T1ftG 1/3/f!IJ T1blt 2 Egu1~nt T r1n9111 tter (low level 1nput) logic [qu1~ent Actuatfon Oevfce

~vfct Ste* ~.

Narnnr Range level So11d State Protec-tion Syst!.'"1 Process Control &

Protection Cabinets Au11. FW Pump Motor

!!anuf. /~_de_l Barton 764 (Lot 2)

Westinghouse West1nqhouse/

Hagan Allh Chalmers 5BRllS-Rr.- 36'JORPM 6flf111P qu~lification Oocurne!!._t Westinghouse Letter Report NS-TMA-2184 WCAP 7817, Supp. 3 and Westinghouse letter NS-Cl-692 WCAP 7817, Supp. l and WCAP-Rl\\J2 Allfs Chalmers Rotor Shaft Oeflections, Stress Calcu-lations, Seismic Shock fl/ 717n fSAR Ref.

Q7.JO,

()7. <11 07.18,

()7. ?CJ Q7.18,

{)7.79 117. l R, Q7. 7° Enviro~ntal A Seh*1c Seh1111c Mo environntnt.11 requfred-loc1t.t in control roo~ 1rea which is provided with redundant Class J[

Vent Systems Se151111C No enviro~tal requfred-located in relay rOOlll 1re1 whfch fs provided with rerlundant Class 1[ Vpnt 5yst~s Sef s111fc Ho envfro"""'tal requfred-redundant Aux. Fll PIJlllP (turbfnt) located in other arpa wtiich ls not subjPCt to thp s.- postu-1 atPd ndvprse envlro~nt.

NOTE:

The se1s111fc doc""4!ntlt1on p1ckaqes for the SS~. reactor protection racks and au11111ary feed'jjfttpr ptll"IP motor~~ ~vftwf'd by ttt. M!C Sehmlc QU111flcat1on Rev1tw Tram for Salem durinq lhr early part of 1979.

~

The Wpstinqhouse rpports ire generic to a number of plants and have already bePn submfttPd to the NQC Staff.

I.

~ *:

I l

I

"'" I I

I

REFERENCE B

December 27, 1979 Director of Nuclear Reactor Requl.ation o.s. *uclu.r Requlatory CollD.iaaion Waahi.nqton, D.c.

20555 Attention1 Mr. A. Schvancer, Chi*!

Operating Reactor* Branch tl Division o! Opera~in; Reactor*

Gentl9men1 ADDITIONAL INFORMATION AUXILIARY FEEDWATER SYSTEM LESSONS LEARNED RECOMMENDATIONS 2.1.7.a and 2.1.7.b NO. l UNIT SALEM NUCLEAR GENERATING STATION DOCKET HO. 50-272 Ill our letter ~ you of December 14, 1979, we conmitted to *upply you vit.h the procedures which demonstrate t.he testability of the automatic initiation of Ule Auxiliary Feedwate.r Syatam.

Attached you vill find Surveillance Procedures SP(0)4.7.l.2(a) and SP(0)4.7.l.2(b) which fulfill this commitment.

Should you have any further queationa, please do not hesitate to contact us.

Very truly yours, 0.., C" d

ng::: ::~.,;l ;'TIC' F. P. Librizzi Frank P. Libriz:i General Manager -

Electric Production Jp/ /t.t tachm1J:t1 t r

SP(0)4.7.l.2(a)

SURVEILI.k,..CE PilDCEDURE PLk"T SYSTEMS -

AU>:lLlT,P.Y i"EI:D;.;;,T:t:F:

UNIT ND. l l.O PURPOSE l.1 Demonstrate the No. 13 Auxiliary Feedwatcr P~~? O?era~le: during Modes (l) thr~ (31.

l.2 Demonstrate the Auxiliary Feedwater flow p~ths opc:rat.lc during Mo:les (l) tr.:-*~ (2).

2. 0 INITIAL CON:JITIC>~~S

2. l Stea..'"f, pressure is > 750 psig (Tavg > 512"F) I \\/hen O?erating r:o. 13 1'.t:.xil ia:-y fe.,:*.:a~e:::

purr.p.

2.2 The Auxiliary Feedwater Storage 'l'anY. > 2DC,OOO g:.}.s.

2.3 Che~k Off Sheet 4.2 of OJ III-10.3.l, "Auxiliary Fee~~a~f:::-

Sv.stc~ Op~=~:io~*, i~

satisfa=to:rily complete=.

2.-'

Stea.-:-, Generator lcvf::l is > lOt < 67l an:'! ll & 12 J.;I F'-.i.-:-;:s a:-e cpe:ratJ.e.

3.0 F?.Z'.:1*.".'710::.s 3.2 If withir. 30 r..inutes after a shutdo;..-:i cf ~o. 13 J.Y\\-; Pur..?, it i.s cesi::c..= t.o re-":::.:::

the p;.:r.-.;:>, e):e:rcise the rnan:.ic.l spee::i cha:;ger te> T.'.ir.ir.,;.:r.. an= ttie::-. rcst:t to desi:-c-:.:

specc.

3.3 Flow to a Stea~ Generator must be linited to < 1 x lD! L~~/~::. (200 g?~l...-~e~ t~~

following conditions exist in that Steri~ Generate=:

1)

Level has been <lOi for more than 5 m!n~tcs, an~.

2)

?lo feed flow (H::iin or Aux) to that Stearn G~ne=ato::- for rr..:ire t!'J.::?n S r..i:-."..:~..:?.::.

~.O CHECY. OFF SHEETS 4.l

.Ar;; Valve Lineup 5.0 PROCED:YRE S.1 Once per ~-1 01,ys, verify No. 13 AuY.iliary FC'edwater Pu.-r.? opPrable, in=lu!i:-i~ t.he...-ate::

flow pat.h, by performing Section 5.2 and 5.3.

S.2 1'uxiliary Fel!d-.;ater Pun:p Test

~

1~:01 Ii II :. '

\\;-

.* 1 5.2.l Close (fully) each of the four A!'ll Valvci:;.

-=:io.1.em :Unit l

-l-

,~n "*n

. : I*

Re\\". 5

S:f' ( D l.;. 7. l. 2 (")

5.2.2 Decrease Speed Derr.and to rni1iirr.u:-..

5.2.3 St.art No. 13 AF~ PU."n!J, end increar.e speed tt> obtc:in > l50:i pdg dis:)-;::.:-:::~ i--:--:::-

6urc.

5.2.4 Record d~ta required, except operating time.

TH'.!: S'IhRTED STEA.'"'. PRESS!.iRE (l)

Dl s~p_;..~::;r, PRLSSURE (2)

'l'lY.:.. s:-:..:;:.:.;

(1)

~ini~u.~ acce?table > 750 psig.

(2)

Minir::u.-:: acce?table > 150CJ psig io-: ?:a. 13 J.* :*** P:.::".";..

5.2.7 Ve::-ify that 1~552 Turbine Trip valve is n~t t::-i~?e~ by che:J:i~~ l~:e:~y t~at the v~lve is reset.

5. 2. r 5.:2.~

to desin::'!

sett.in-=

~.. ~

~ **-.......

I 5.3 Auxiliary Feed~ater Valves 5.3.l

~e:ord the position of ea:h valve liste~ on Check Off Sheet ~.1.

~h!.:SE p:$i-tior.s r.lt:St rr.atch those 1.mde:- the N*);:_v_:..:, PDSITlO~' col\\!!:".:-..

5.~

Triis su::-veillance shall be acce:*tablt.: if No. 13 Ml*: f'-.1.":lp op::?ration i~ v,;.; li;.it~ c:

6tep 5.2.3 and all valves listed on Cht:k O~f SheEt 4.l are pos~tio~~~ IA~ st~? 5.3.l.

Manager -

Salern Generating Statio~

S'.:>:R::: Y.'i'.'eting tio. ____________ _

r~la'(-.' :

"I

I*.

1:

I I.'.

M

'i,1J 0.fl ;'/

l '.. :r \\/

Sale.":1 Unit l '** i..J..S lfi.e\\*. 5

AF AF A:

l3 13 13 13 SG SG SG SG SG SG SG SG SG SG SG SG SG SG SG SG SG SG SG

... *-'--------~-... -. __ _

STOR 'II< ov PU~P SUCT v PU~P su:T v PU!'!r SUCT v AF PU~? D!S2H A: PU!-'.? ogc;;

AF PU~P D!SCP.

'Af' PU!'!? D!SC"H 1'.F LEV!:I..

co~~'!'

'J,.'F LE\\'EL co:~?

AF L!:\\"!:L co:~:-

AF LE\\'£1.. co::T l-S J\\'

AF,,.

'J,.'F lV l-.:F l\\'

1-.: L:S\\"!:L co:::-

1-.Y AF J -

l.S

]..:!'

J..F 1J" AF M

TO TO TO

~o v

\\"

\\'

v

\\'

CHECK OFF SHEET 4.)

AF'lo> VALVl::

Llt~E1JI' Nl~"'.E

~

ll 5~

12 SG 13 SG H

SG

---*-t---"l:..c3:__AF..c-_P;_UM P D l SCll V TO 11 SG 121..FeG 13 M 0

SP{0)4.7.l.2(a)

J..::J::-----

Pu.::: Jc.::

l_3_AF~6. ___

l.!_AF p~p D~SC!!_~TO 13 __ ~~-------------

---* -- I

....___!~1i.F8_&_.___l3. 1'_'f_ Pi:J~P_P.1S_~H _'{ _ _'!_9_1~-S~---*-- -------*- *- *-

-* ___ l.O_.

___ ----- I I

~--. *.-.----* _

SP(0)4.7.l.2(a)

VALVE ll>..Fl 01 Ar Plll'!P REC!RC \\'

121\\FlOl AF PU~:r REC!P.C' v l.Ml03 AF PUl-'.P REC I RC O\\'

(l) Deffiand 9hall be et 95~ open.

CHEC~ orr SHEE1 4.1 AFi,.; VIJ..VE Lll;Ei.Jr N1'-"'\\L l'CiPY.J*.1.

POS l 'I'l c.:;

LO LO l."J Re*:ie;.;e:: by ______________________________ _

Se~ior Shift S~?~rvisor/Sti~t S~perv~s~~

r ** i

8' **..., ':1*i r\\:'. * ~I

\\ *'

~

1 fl\\ 'I i

\\** :

I.

I ~

I I I,

I I n, 1, \\I *J'l*li'.i' * *. 1 U.\\.l1.;\\I 1.:,:l.l_.i

...:,I

.u.

--- ~

'\\"[' \\'!

1

/

.* *I

'-' ~. ~~

,,,.;:-77;"-;--~

~--**--

Pu:.1 :-J :;:.

I I

I

1 Rev. 5

~*------*

SURVElLLANCE PRD:EDU~

PLJ~;T SYSTE~S -

~UY.lLlAF.Y :n:r:D't;,?.,'f!:P UN:!T t:o. l s J* ( u) 4. 7

l. 2 ( t,)

1.0 PURT'DS!:

l.l 'J'o verify No. ll and No. 12 Au>:ilia:::-y Feed.,.atcr (1'.F\\\\) pu-:-?s ~tart upon rc=e:.iJ.-"- c.:

.either e loss of bot.h Main Fe:ej*~*ater Pwnps o:- a St.ea-Ge:r.e:::-ator Low-Loi.: wati:::

l~--.-c.~

signal.

l. 2

'I'o verify No. 13 Auxiliar:i* Fced...,ater PU.""."?' s th:::-ottle valvr* o;:>er:5 l.l;:>:i:-. rcc(:ip-:. c~

either e loss of offsite (4kV Group Bus) pow~r c: t..::; Stea:-. G::~.::::L1t:.:-I.1:>...--:.,:,1,.* '"':!':.,_:-

level sign<ils.

l.3

'Io verify proper, operatio:-i of J,uto::-.atic Valve" ir. ll c::-.:: 12 J..-.:.;:.

Fe:<:::-~-"'-'-=- :r*..:..--

paths at pW'.1? dis=harge press'.lre: of i:.oo p.::ic_;.

2.1 The u~it is in operational ~oje 3,

~. 5 or £.

2. 2 The: Au>:iliary Feej;.;ater Sto:::-a:;::-: 'Ianl; is ope:::-c.:.le.

2.3 Functional tests are not bein; conducted on either Train *h* o: *e**

Out~.:.: 7e~: ~=-=

l~terface: Cabinet.

ii. 4 J..11 stc?S of this p::-ocedure are to b::: co::-:;:*lct;:-:i in S~:;:-JE.r.::£:.

"H-,,.,e-.*,;,:::-, i:-.iti:::

~- - -

ditions ~ay b~ co::-?leted i~ a~y order.

2. 5 None of the three: AF\\oi purr.ps are: co:-.=urrently i:; use.

2. C Tbe valves liste:: o:-i Ched: o:: Sheet.C. l a::-e: in their 'l'O?_".J..L pcsi tio:-:s.

3.1 Do not operate any AFW pur..;::; with:>ut e suct.io:: o:r cis::h<!:'."££: flo;.;

p~:.:-:.

-'. 0 Cl!ECi: O!"F SHEETS 4.1 AF\\ir1et 'jest Section.

S.3.l Open bot.h front ir.terface cabinet doors ana verify the following:

l)

All of the test s1o1itches arc in the m:BL'J:r. OU7?DT positio:-i.

2) control Roo:r. receipt of tht: sei::...1 D STJ-.7::. PP.o:::::7J er: :,y ST!:" 7?J-. ::~:
J-.' c,::
~ TEST anr.unciat.or.
5. 3. 2 Test the internal test assembl~* as follo*... *s:
l)
Depress and then release the TL/TX: in::ic~tcr.
Th~ re= li;~t ~:.~: i~:~:.
nate end then cxtinguis~.
2)
Insert the r:iete:r test lea= into the TEST J;..cr. t-:=7L:F: "IL.'.:::' j::.:::l:s.
3)
Place the range switch in the lV position.
Depress the ?-'2TEP. TEST push!.::.ittcr;.
The peter sh::.>..:1:: in:::.c::te c~-;:,::-::.:::.:-.:.:.;,:;*
full scale aeflectio:-:.
5)
Remove the meter jack lead fro::-. the TES'T Jl-.Ct'. Y.LTE.?. TES: jc.::i::;.
All subsegue!'lt Section 5. 2 st.e;,:>s are \\!sec..,;. ti-: rcfE::re:-.cc*
to Check Off Sheet ~. 2 (Train 1*." Sectio:.).
5.3.3 Utilizir.g TS632, follow Steps of Table C7.l.2(b)-l.
5.3.4 Utili:ina TS633, follow Ste?s of Talb~ 4.7.l.2(b)-l 5.3.5 Utilizing TS634, follow Steps of Table 4.7.l.2(b)-l S.4 Train.. B" - Output Test and Interface Cabinet Test Se:tio!'l S.4.l Open both front interface c~~~net do~rs and veri!y the follo~in;:
l)
All of the test switcl1es are in the UNBLOCK OU?Pi.JT posit.io:-..
2)
Control Roo:n receipt of the SDLlD STATE PROTECTION SYSTE!"! TRAl!;.. ;..* (.*::
TEST annunciator *
!>.4.2 Test the internal asser:ibly a!: follow!>:
Th~ rco li~~t ~ill ill~.i-Depress a:1::3 U1en releas£* the> TL/T>:C indir.ato:::.
l) nate an~ lhcn e~tin;uish.
r~n-1!:*!:
U lL.1.. '
r*~*: *r *.........
.. I
,.. '. "I '"'.. '
[,.*,.: "/.,.';*,I. I~.,
.. 11.J \\:../' *.,I,.
\\ I I*
I...
11
.IL:
LI
.*I.,
SI' (0) C. 7. 1. 2 (!J)
2)
Inse:rt the meter test le;ac'l int.o the T:C::S'I JACl< METL:i* TLS'J jllcr.:-..
3)
~lace the range s~itch in tn1 lV position.
4)
Depress the METL:R TEST pushb'.ltton.
The rneter sh::i;;ld indicate a;-,;.::-c..xir..::*_.:_c:i*
full s::ale dcflc::t.ioll.
5)
Re::icive the rneter tl'st lead fro::-. the-7:C::S7 JACY
~.:TEP TEST jad-.!:.
All subsequent Se.::tior. S.3 steps are usec with re!e:-ence to Check Off Sheet 4.2 (Train MB* Sect.ion).
5.4.3 Utilizing TS6~2. follo~ Ste?S of T~Ll~ C.7.l.2(b)-J.
5.4.4 Utilizir.9 TS633, follow Ste?S of Ta~le C.7.l.2(t)-1 5.C.5 Utilizir.; TS634, follo~ St~?S of Table 4.7.l.2(b)-l 5.5
.,

:-,,.* P;.:.::-.? Op"':-atior.<:!l Se.::tio::-:
l)
Verify illu.-::inatio:: of tt.e rea s:;..P.:- bezel.
2)
Test res~~ts 5.5.2 Depress the 570? L,,.,,_., fo= No. 13 J..'.l~:iliary Fee:: P:.:..-.;::.
5.5.3 Verify 11, 12, 13, 14t.F21 closed.
5.5.4 Adjust a~~and for 11, 12, 13, 14hF2l V~lves to lOC~.
The sta!'ting of N::1. 11 L 12 J..FI'.' Pu::-:;-*"' with a lOD~
De.:r.an:! c.n 1.!"21 valves will adj water to Stea."'.'
Generators.
M::1nitor Stea~ Gen~:-ator level ~~=
AF21 Valve position to rnini:r.-.ize -..<".tcr ac=it.io:-,
to Ste~~ Generator~.

S.5.5 SiMultaneously, de?ress No. ll an:: No. 12 Main Feedwater Pump £~.ERG!::->CY TRIP pushbuttons (located on test panel on side of ~ain Feedwater Pur:ips).

l)
Vt!:-i!y the starting of No. ll and }b. 12 AF\\\\ Pumps.
f: !."; i Ji i. ~. *'. j:'.p. 1;1 *1-r. : *.'. *.';:
,1: \\'t; JI ',' *'.'
I
. '. : "I I *.) : : \\' IJ.JI'. : \\ ~I L~
Sale::i Unit l ( ;' "',,: \\','/
\\
j;
'I
\\:
-*.I\\,~:: LJ
~ev. 3
~
5.6 Reviewe=
_ ___.___ ___ -----~-------*- -. -*
s" ((Ji 4. 7
1. :> ( ):_,)
2)
Verify level control valves are opening "1hu1 pu:-:.;* c.i.s:har9t* i:;:(::.::.'..:::--t. rt.:<,::'.:;;
ap?roxim~tely llDO psig.
3)
Test results D
D 5.5.6 Depress the STOP bezels for both No. ll anc N~. 12 1..F'~ Pu..~?s.
5.S.7 Verify ll, 12, 13, l4AJ21 Shut.
'This sun*eillance shall be acce?table if all fee:!*.;l!te::.-
p:.ll"";.-
co:-.~::.-c,1 circ-..it.::.-y I.. -
~,..
IA~ T~ble 4.7.l.2(b)-l fo::.- both Train A, Train B, and th~t ~ott Auxil1bry F~e~~~:~:
p\\!."':'.?S start, attain a dis:::harge pressure > 1100 psig a:-i:'l lcv~l c:;:-,trc~ vc.lve.s c.;:.~:- *
µ)~~~~~~~~~~~~~~~~~~~~~~~~~*~~~~~
Date Senior Shi~t Sup~::.-viso::.-/S~ift Supe::.-~isor Manager -
Saler. GPnt.:~a~in; S:a:i.:.:.
f1 '\\' \\ r *I 1'
. 1 l
. ' I* L
"-*.1L*~:.' I Rev. 5
---~~,; *-- -- __,.._
SP(O)C.7.l.2(:.,)
ThllLE ~.7.l.2(b)-l TEST srou:::~::E "/\\./~"
This test sequences a series cf relays, such that the slave rel~~ is energized, ye: th~ c,~:~~:
functions of the slave relay remain unchange:::.
Perform the following steps, in the ordf'r *givt:n, utilizing the designated test s.... ~t:::..
Al)
Turn TS to the BLOCK ODTF'i.J7 position.
Verify an::
re:::>r~ o;. Ch-:-::r. Off She~:.;.2 U:'~
!c,:::.,-~*
ing alarms:
a)
TL/TXC Alarm bl Aux Alarm Safety Equiprae:;t Train BLO:r:::::::>.
h2)
Perform the follo,,..in~ initial checks, rer.o:-cin;i the rest:J.ts i:-i th.-.. F,;,c-:r. u::?*:-:-
!.e:::i:~.
asspciated with the test switch bein3 operated:
R2cord o~ Chee~ Oi! s~~~: C.2 a)
Depress ~he test lights associated with the test s~itch or~r~t~::: i:: Step ~lJ ~~-~.
The test lights should not energize.
Record res~lts c~ Ch~cr C!! Sheet 4.~.
If the test lig'ht illw..ina+:.es d:::. n:::.t F=-:: ::;.,t::::: **i t.:-1 tes:.
Contact the Perforr.-.:=**,:::e Depa:::-tt,e::t a::id re:;'.1est the;* i::-
vestigate the rnalfunctic::.
bl Insert the ~eter test lead into the test ja:ks associated wit~ test ~~itc~ O?~=~:~~
in Step lh) above.
The test ~?ter s~~~ld in~icat~ app=o~i~ately 1/2 f~ll s::3l~ d~
flection.
Record test meter in~icotion on Che~k Off Sheet C.2.
~3)
Turn TS to the OPER.;7~ O~TPUT position.
Re=ord the test light illi.:.::1i~atio:: i~ the O?:~~~:
OUTPU'T sectio::-i.
7.4)
r.:*::ord the test rneter' indication fo:r each o: the.- teJt jo~Y.s in the; C?:::P.;;7:<;
O"..':'?~:t !"'?::ti::::.
AS)
T~?;; TS to th~ P.ES~T OUTPDT position.
t.f __ :... '
'-=- -
test meter readings return to thos~ values recorce= in Ste? h2).
A5)
Turn TS to the UNBLC>CY. OllTPU'.l' po?:ition.
The follo.,.ing alarr.-.s will reset:
P.c::orc o:: C!-."'::::
Off Sheet 4.2.
a)
TL/TXC Alann b)
AUY. Alarm Safety Equipr'lent Train BLOCfJ:D.
~~ -,
1* :'
I
1
'[i,,1,...,.,,,'7",,. -,r-11 ','."'\\., **
11, I'" *.
I **' "** ol * \\
jl: I '..*
I...
~ ! I \\ * '. I *. :' : I : I *.
! I
l ---~-----**---~-~~---.-----------------------~
r Vhl., Vf.
NM::::
NO?.~'..hl, P0~}.'!_20~* j_ ll:J'Il.?.LS I I
I lM'l AF STOR TF OV LO llAF3 M" Pll'-l"" SUCT v LO I~
l2AF3 AF PUMP SUCT v LO l3AF3 AF PUM? SUCT v LO llAFlO 13 1>.F PUMP DISCll TD 1l I
SG LO I
12Af'l0 l3AF PU!'!::' DISCH TD l:? SG L0 I
lJAFlO 13 1>.F PU~'iP DISCH TD l3 SG LO I
I l4AFl0 13 AF PU"I? DlSCH TO 14 SG LO llAF20 SG AF IV LO I
l2Ar20 SG Ar i
IV LO 13AF20 SG 'AF !V L0 HA:'20 SG AF :rv LO 11AY22 SG A"F IV LO l2h.F22 SG AF !\\"
L0 I
13.?-..?22 SG "J..F IV LO I
l4h!22 SG J!.F :rv LC!
I ll1.F23 SG 1>.F I)..,. 'r"... C\\'
I LC:. (l J I 12h.F23 SG 'JU INL!:7 CV L0 (l) I 13.r.:: 2 3 s~ Ar I?;L!:'I' C\\'
I LO (1) I HA.::23 SG AT l?\\LET CV I
LO (1)
V...?~2 1-..? PUM? R!:CEC CLG !SC:.
I 0
H.?51 I N' BACK!J? TIE FP.O~ D"=""-*~,..
--* -.1.1* \\oiJ..7I:R C&T x
I llAF52 AF PU~P Bl.:Y.UP su:T C&T l21.F52 AF PUP.!' BJ\\Cr:UP SUCT CE.T x
13;...?52
'J..F PU!-'!P B1'.':Y.UP
~~!~T C&'T Y.
ll1.Fe6 13 AF PUM? DISCH V TO 11 SG 1.0 12AFB6 13 1-J' PU'.'1? Drs:H v TO 12 SG U>
13..<-.FB6 13 M PUMP DlSCH v TO 13 SG Ul 14AF86 13 w PU~.? DISCH v TO 14 SG I.O llAFlOl
'AF PUMP REClRC v I.O 1"AFlOl AF PUMP REC!RC v 10 1AF103
'AF PUMP REC I RC OV I.O lAFl04 AF PUMP GOV COOL OUTLET 0
lAFlOS
'AF PUMP JACKET OUTLET 0
11i.Fl06 AF PUMP JACKET OUTLET
~
0 Performed by ------
--Reviewed by Date
f,
_,. 0
~
I f

-----...... *---*-*--.,... ---~-

CHECY. orr SHL!.T 4.2 l\\UX J"F.EDn;..Tr..R T:t:ST
~-:-;,
51-."J'i'CH PD.SlTlON FlTNCTlO'-'
TL TXL Al.AP.I:
BLOCK AUX AU<.P..M l>UTF'UT TEST L!G:lT/
,1;,C}:
T!:~T OPERATE
'IESJ LIGE:-/
Ol'-;1'!'
J;,:r: TE~T RESET TEST L!G:!T/
nnT?~"T iJ..,...;.*
't';' C'7' UNBLOCK TL TXL PZSET Ol.'TPUT J..t.:X l,L'*'. P..E:SF.T (1) Jack No. JC632 (2) Jack No.
JL..~633 (3) Jack No. JC634 SV."ITCI!
TAAl :J..,\\" TES'l T5632 (l l s..~\\T
.r.:-.:-.;:-
\\/
I/\\
\\/
l/\\
TJ>J\\Il; B" r---c::,....
J ~e-J.
TS632
( 1)
T5633 (2)
.T

s:..: V)::rs I
j
\\
\\ I
\\I
~
I\\
I \\
I I
TS'S~3
( 4 l POSI'.i'J:O~;
FlJl':CTIO~;
~....7
~-~~:;~.~ -r... r:::...~ c::*-
T,..,.. -
'.'C':~
\\. *'.. -._
TL TXL AL.?..?..~'.
BL0Ci:
"t: Y.
1'.LhP-~
OUTPl,""T LIG!-:7/
'TES'I
~
J;... :v '!TST O?E?J,T£
'JES':" L!Gr~: /
0:.TTI'!;:
Jl\\Ct'. Tr-c::*-
~~j RESE:l
'TES'!' LIG:!T/
OUT?UT JJ.CV
'i'IC:":'
UNBLOCK TL TXL P.I:SET OUTPUT*
At:Y. ALM p_zsET (l) Ja=k ~o. JC£32 (4) Test Light ~o. TLB633 (3) Jack No.
JC63~
Senior Shift Sup~rvisoriShift Supervisor
\\/
I\\
I I/\\
\\./
I ' J
~
I \\
I\\/
I \\
l/\\
I I
SP(0)4.7.l.2(bl TL5E34
( ~}
S.~T
J ;:.. r:;
.,..... ~**
)~
I
\\/
/\\
T7 S ( :~:,
(., i
. _-. - ~..
'I" -
\\I 11*\\
,--1 J
\\/
/\\
ri r.n *:"\\!-I',.,""'\\" "1 r:-* r:ljTD :*~~~\\
I\\ t, I ~*, I l : I.. !.. I!
\\
I I 11
' J
'U\\i*I,.:...'
'*11,,*
t ~"
I
'- *.. ~ *.
Q\\ *~::11L]"
I, 'il
'.J'~ Rev. 3
\\\\
POSITION AUTOllA.TlC I!HTIATIOI~ _OF THE AUXILIARY FEED~ATER SYSTEM FOR PWRs (11.E.l.2)
~ RFFF.RENCF.C As part of the Lessons Learned recommendation 2.1.7.a. the Automatic Initiation of the ~uxiliary Feedh*ater System must be "upgraded in accordance \\'/ith safety-grade requirements."
t CLP.RH I CATION The intent of this recor.imendation is to ensure a reliable automatic initiation system. This obiective can be met by providing a system which meets the requirements of 1EEE~279~1971.
The staff has determined that the following salient paragraphs of IEEE-279 should be met or sufficient justification provided for any deviations.
JEEE Paraqraph 4.1* General Functional Require~ents 4.2* Single Failure 1*1<JJr.Ci ucWSING Mfl.Nfl.GER SALEM E. A. LIDEN.
4.3, 4.4 Qualification
~
4.6 Channel Independe~ce 4.7 Control and Protection System Interaction P.UG 19-~980
~~.
NOTED
.:C.1-*" * *:
REFFR 10 0 0 '*' ***:*:**o:o:*
D****** D******.-.*:*:**
r.<~'
~_qp1E.S *.****.*********.*.*.* _..*. *.*.*...
' c:iW........ FILE ***************
4.9*, 4.°JO*
Cc.pability for Testing 4.11 Channel Bypass 4.12 Operating Bypass
4. 13 Indication of Bypass 4.17*
~anual Initiation T
\\.
/
I
j(.,) (~~.--: f (
...:::.---- *. f.
,_)"')..<.-rt:.:-.
f'("bv -~
f\\P-5 r notµ. ;r ~ ft1?
£wS 1)JJ
"'ihese re~uire;;-,ents \\*.*ere part of the short term control grade require;nents.
A??U U-.B IU TY Ail Pi-.'?.s.
J, JMPLEMENTATION DATE January 1981.
TYPE OF REVIEW NRR Pre-Implementation Review.
DOCUl-'1EtHATlON REQUIRED Each licensee/applicant sh~ll provide documentation sufficient to support a re2sonable assurance finding by the NRC that the above requirements are met.
The documentation should include as a minimum:
t.
1.
A discussion of the design with respect to the above paragraphs of IEEE-279.
2.
Discussion of deviations, if any, from the above stated requirements with adequate justification for such deviations.
3. Support information intluding system design description, logic diagrams, electrical schematics, piping and instrument diagrams, test procedures, and technical specifications.
TECH~ICAL SPECIFICATIONS REQUIRED Yes.
REFERENCES NUREG-0578 Section 2.l.7.a.
. c I\\
AUXJLlARY FEED~AlER FL TO STE~'i GrnERAlORS II.E.1.2 r'rtuJtCf Li~*-L:Si 'G MANl\\GER
~;'\\~;.: _,j POSITION E. /;;,. LIDEN As part of ~h~ Lessons Learned recommendations 2.1. 7 -~ the Flo' IndiA~CJr9 7
J,980 for the Aux1l1 ary Feedwater System must be "upgraded in accord nee \\told~&
.t!d:
safety grade requirements."
REFFR 10 gj~~ b"-:.........
CLARI FI. CAT I ON COPIES.,.l_../J... <!.' l'.~Jp -~.J.*.. :....
CUE
'F. '. *r 'I-<
J. * * *
ILE..***.*
The intent of this recorrrnendation is to ensure a reliable indication of auxiliary feed1vater. system performance.
This objective can be r.iet by provi-ding an overall fndication system which meets appropriate design principles outlined in IEEE 279-197~.
The staff has determined that the overall flow indication system should meet the follm'l'ing criteria:
1.
For 8&\\\\ plants to satisfy these requirements they must provide as a minimum:
a.
Two AFW flow indicators for each steam generator.
2.
For \\..1estinghouse.:nd Combustion Engineering p1ants, to satisfy these requireri2nts they must provide as a minimum:
a.
One AFW flow indicat6r for each steam generator.and, two Wide Range level indicators for each steam generator.
3.
For {4 loop) plants with 4 vital buses the following option is acceptable:
a.
One AFW flow indicator for each steam generator.
b.
One Wide Range (lo lo Level) indicator for each steam generator.
c.
T1*:0 !\\a1-roh1 Range (Hi level) indicators for ecch steam generator.
4.
For c.11 plants, specified accuracies must be justif1ed based on overall syst-e::-. requi1-er.,2nts including post-accicent 1mnitoring.
Docwnentation must ce1i,e:nstrate that tr.ese accuracies c.re met by the s_ystem proposed.
In addition, the Starr has deter~ined that the following salient paragraphs of lEEE 279 should be r.r-t in the flo\\ indication S.Ystem design or
sufficient justification provided for any deviations.
JEEE PJ..RJ:.GRAPH
!..
l *
l,.(T:
L
....... )
!... f.
L -
.. I General Functional Requirements S i n g 1 e. F c i l u re 4.4 Qualific2tion
C:-;cr.r:el I nce;:..c:r.dence
C : :-. : r o i e: ;, d r n:- t e ct i on Sys t e ;--:-: I n ~ e 1-2 c ti o ri
t..
l..... *
(:.:-.-~,---,*.;-.v :or lec:.~.iro
-V
-r-C-'
l~J I
-;.. 111_,
i
-*~rt of tf-,e" Si":c*rt
-~,.AP:'L1tABJL11l All P\\*lRs LEMHnAT10N DATl January 1981 E Of REVlE~ NRR Pre-Implementation Review DOCUl"i~N1AT10N REQUIRED.
Ea ch 1i cens ee/ app l i cant sh a 11 provide doc umen tat ion s uffi cl ent to sup po rt a reasonable assurance finding by the NRC that the above requirements are..,t.
The coc.Uiir2ntation should include as a minimum:
1.
A discussion of the de Sign with respect to the above paragraphs of IEEE 2 7g.
2.
Discussion of deviations, 1f any, from the above stated requirements with cdequate justification for such deviations.
Support information including system design description, logic diagrams, electrical schematics, piping and instrument diagrams, test procedures, end technical spec1fications.
TECHlHCAL SPECIFICATION CHANGES RE U!RED (Yes)
REFERENCES NUREG-0578 Se~tion 2.1.7.b
ITEM II.D.l:
PERFORMANCE TESTING OF PWR RELIEF AND SAFETY VALVES By letter dated October 31, 1980, the NRC transmitted NUREG 0737, "Clari:fication o:f TMI Action Plan Requirements", which provides under Item II.D.1 additional guidance as to what the NRC expects to obtain from Industry safety and relief' valve testing.
Each Utility is required to respond b~ December l5, l980 as to their commitment to the NRC requirements.
The following is PSE&G's response to NUREG 0737, Item II.D.l:
"As a sponsor of the EPRI PWR Safety and Relief.Valve Test Program, PSE&G intends to comply with the require-ments of NUREG 0578, Item 2.1.2. By letter dated-Dec-ember 15, 1980, R. C. Youngdahl of Consumers Power *com-pany has p~ovided the current PWR Utilities' positions on NUREG 0737, Item II.D.l clarifications.
Briefly those positions are:
A.
Safety and Relief Valves and Piping - the EPRI "Program Plan for Performance Testing of PWR Safety and Relief Valves", Revision 1, dated July 1, 1980, does provide a program that sat-isfies the NRC* requirements.
Discussion with the NRC staff and their con~sultants are resol-ving specific detailed issues.
B~.Block Valves -
The EPRI Program has not formally included the testing of block valves.
However, a small* number of block valves have been tested at the Marshall Steam Station Test Facility.
The PWR Utilities and EPRI can not provide a detailed block valve test program until results of the Wyle and CE relief valve test a~e avail~
able.
Therefore, a block valve test program will not be provided before July, 1981.
The PWR Utilities and EPRI believe that the proper oper-ation of the TMI-2, and Crystal River block valves and other operational experience, plus knowledge of the Marshall tests, support a less hurried and more rational approach to block valve testini.
c.
ATWS Testing -
PWR Utilities will not ~upport additional eff~rts for ATWS valve testing until regulatory issues are resolved.
The major safety and relief valve test facility (CE) is nearing com-*
pletion*and some measures were taken to provide.
additional test capability beyond the current pro-gram requirements.
The NRC should recognize that results from the current program are likely to provide most of the information necessary to add-ress ATWS evenis (i.e., relief capability at high pressures)."
j II.B.2. -
DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR POST-ACCIDENT OPERATIONS
RESPONSE
PSE&G addressed the requirements of NUREG-0660 concerning plant shielding in our letter to the Director of Nuclear Reactor Regulation, dated July 8, 1980.
The requirements concerning environmental qualifications of equipment for post-accident operations were addressed in PSE&G's letter to the Director of NRR, dated November 26, 1980.
Concerning the requirements of Item II.B.2., there are no deviations from the above mentioned positions.
I ITEM II.B.3. -
POST-ACCIDENT SAMPLING CAPABILITY
RESPONSE
PSE&G has reviewed the design and procedures of the Salem Nuclear Generating Station and we have concluded that there are no deviations from the requirements of Item II.B.3
ITEM II.E.4.2.
CONTAINMENT MINIMUM PRESSURE SETPOINT
RESPONSE
The recommended containment minimum pressure setpoint for initiating containrnerit isolation is < 3.2 psig.
The basis for this setpoint is as follows:*
Cl) The maximum operating pressure based on containment pressure operating history h~s
. been 1.0 to 1.5 psig; (2) The Westinghouse Setpoint Study indicates an expected instrument accuracy of 2.3% of the in-strument span (60 psig) which is equivalent to approximately 1.4 psig; (3) The present technical specification for the containment pressure at which to initiate containment pres-sure relief is 0. 3 psig.
The summation of i terns ( 1), * ( 2)
and (3) above (1.5 psig + 1.4 psig + 0.3 psig) gives the recommended setpoint.
The present containment minimum pressure setpoint for initi-ating containment isolation is.,~ 4. 7 psig.
ITEM SALEM NUCLEAR GENERATING STATION NUREG-0737 -
RESPONSES II.F.1-1 Noble Gas Monitori document and justify deviations (if any) to position and/or clarification.
RESPONSE
Public Service Electric and Gas Company has been working closely with Vendors in order to develop a reliable high range noble gas monitoring system.
Preliminary design details on the* system as currently proposed will be available on January 1, l98l for review.
Final design details are not available at this time since PSE&G has felt compelled to reject earlier designs proposed by several vendor as not being acceptable in satisfying the USNRC guidance.in that several manufacturers suggested use of excessive amounts of lead attenuation in order to achieve the extended ranges required.
II.F.1-2 Effluent Sampling and Analysis:
document and justify deviations (if anyl. to position and/or* clarification.
"RESPONSE As a,subsystem of the high range noble gas monitoring system, PSE&G is evaluating the purchase of an effluent sampling system for particulates and iodines.
Final design details are not available at this time since.the proposed manufacturer of the effluent sampling system**
did not have the design of their system finalized.
With the issuance of NUREG-0737, the specifications of the system can be better defined.
Public Service Electric and Gas has the preliminary design of the proposed system available for USNRC.inspection and review.
II.F.1-3 Containment Radiation Monitor:
document and justify deviations (if any) to position and/or clarification.
NCA:kd
RESPONSE
Installation of the high range containment monitoring will be achieved by January 1, l982.
Final design details will be available fo'r review on January 1, 1981 *.
ITEM II.F.2.
INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING
RESPONSE
Attached is PSE&G's response to the requirements of this item.
The response is a partial one because the information received from Westinghouse on December 16, 1980 pertaining to the Reactor Vessel Level System was generic in nature and must be revised for Salem Plant specifics.
In addition, some of the information supplied is proprietary, and the Westing-house proprietary version will not be available until the week of December 29, 1980.
The final proprietary and non-proprietary submittal for the Reactor Vessel Level System will be available for submittal to the NRC the. week of February 2, 1981 *
',, ""'....., *' 'r,.... - -* *
--**~. ";......., *.. J
i.'
ITEM
. II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING
RESPONSE
1. *
1. a.
A final design description of the reactor vessel level instrumentation system will be submitted at a later date.
l.b.
For a description of the subcooling meter, see the attached response to Section 2.1.j.b. of NUREG 0578.
For a description of the Incore Thermocouple System, see Attachment 1 of II.F.2.
l.c.
No modifications are planned for the instru-mentation systems described in l.b.
2.
The design analysis, including evaluation of various instruments to monitor reactor vessel water level, will be submitted with Item l.a. above.
3.
A description of additional test programs to be con-ducted for evaluation and qualification of the reactor vessel water.level system will be submitted with Item
.1. a.. above.
4.
Information on the Rea~tor Vessel Level System confor-mance to NUREG 737,Section II.F.2, will be submitted with l~a. above.
For information on existing instru-mentation see l.b. above.
The proposed ICC monitoring system conforms to Appendix A references, "NUREG 0578, Recommendation 2.1.3.b", and the let~er from H. ~. Denton, NRC,
~o All Operating Power Plants, dated October 30, 1979, subject:
Discus-sion of Lessons Learned -
Short.Term Requirements" ex-cept that the in-core thermocouple temperature measure-ment system is not environmentally qualified and does not meet all the requirements of II.F.2 Attachment No. 1 (see 1.1.b above).
No ~odif ications are presently planned as indicated in
.1.1.c above.
The justification for continued use of the in-core thermocouples is as follows:
Current procedures instruct the operator to monitor the in-core thermocouples as a guide to determining the existence of adequate core cooling.
Despite a lack of available documentation in our possession, Westinghouse information and operating.history of these de.vices in-dicate that they would function.
If these devices were M P80 132/2 1
ITEM II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING
.RESPONSE
1
l.a.
A final design description of the reactor vessel level instrumentation system will be submitted at a later date.
l.b.
For a description of the subcooling meter, see the attached response to Section 2.1.3.b. of NUREG 0578.
For a description of the Incore Thermocouple System, see Attachment 1 of II*.F.2.
l.c.
No modifications are planned for the instru-mentation systems described in l.b.
2.
The design analysis, including evaluation of various instruments to monitor reactor vessel water level, will be submitted with Item l.a. above.
. *- *~
3.
A description of additional test programs to be con-ducted for evaluation and qualification of the reactor vessel water level system will be submitted with Item l.a. above.
4.
Ihformation on the Reactor Vessel Level System confor-mance to NUREG 737,Section II.F.2, will be submitted with l.a. above.
For information on existing instru-mentation see l.b. above.
The proposed ICC monitoring system conforms to Appendix A references, "NUREG 0578, Recommendation 2.1.* 3.b", and the letter from H. R. Denton, NRC, to All Operating Power Plants, dat~d October 30, 1979, subject:
Discus~
sion of Lesson~ Learned - Short Term Requirem~nts" ex-cept that the in-core thermocouple temperature measure-ment system is not environmentally qualified and does not meet all the requirements of II.F.2 Attachment No. 1 (see 1.1.b above).
No modifications are presently planned as indicated in 1.1.c above.
The justification for continued use of the in-core thermocouples is as follows:
Current procedures instruct the operator to monitor the in-core thermocouples as a guide to determining the
.existence of adequate core cooling.
Despite a lack of available documentation in our possession, Westinghouse information and operating history of these devices in-dicate that they would function.
If these devices were M P80 i32/2 1
~ '.
'.. :-* to fail, backup parameters, such as, RCS pressure and*
.tempera tu re, steam generator level, and auxiliary feed-water flow are available to enable determination of adequate core cooling.
These backup devices are quali-fied and are described in the proc*edure which also alerts the operator to potential errors in the incore temperature readings.
The existing procedure and availability of alternate indications is sufficient to ensure proper determination of corecooling adequacy.
5.
The description of the computer functions and functional specifications for relevant software in the mini~com puters for the reactor vessel level system will be. sub-mitted with l.a. above.
A description of the computer functions for in-core thermocouples and subcooling meters are described in II.F.2., and the attached response to Section 2.1.3.b. of NUREG 0578, respectively.
6.
A current schedule, including contingencies for instal-lation, testing, calibration and implementation of any proposed new instrumentation or information displays, will be submitted with 1. a. above *.
7.
-/ -
~*.> *-:, '
Guidelines for use Of additional instrumentation and analysis used to develop these procedures will be sub-mitted with l.a above.
Emergency Instruction 1-4. 4, "Loss.of Coolant" (Leakage Greater *Than Maximum Charging Flow), has been revised to include operator instructions regarding the use of the subcoolirtg meter and the in-core thermocouples as.indi-cations of the approach to inadequate core cooling.
Reference Sections 4.6, 4.7, 4.84, 4.10.4, and 4.10.5, of the attached Emergency Instruction I-4.4.
The procedures will be modified when the final monitor-ing system is implemented based upon input from the Westinghcius~ Owners Group.
~9.
At presen*t*;.no additional submittals are planned *. *How-ever, if additional information is developed.or request-ed, it will be submitted promptly.
M P80 132/2 2
IIF.2, Attachment 1 This is an evaluation of *conformance of the ICC instrument.
system to "II.F.2 Attachment 1" entitled, Design and Quali-fication Criteria for Pressurized-Water Reactor Incore
* * ~permocouples.
The following paragraph numbers correspond to the paragraph numbers of Attachment 1:
1.
There are 65 incore thermocouples, with a minimum of eleven in each quadrant and an additional eight thermo-couples shared with the two adjoining quadrant, to pro-vide indication of radial distribution of temperature rise across representative regions of the core.
See FSAR, Section 7.6.
2.
Primary Operator Displays A.
A core map for each quarter of the core is available to the operator on demand on the computer output CRT.
The core map gives the temperature at each core exit thermocouple lo.cation in that quarter of the core.
B.
The core map will give the location of the hottest incore thermocouple.
This hottest incore thermo-couple reading is the basis for the subcooling cal-culations and procedures..
c.
There is direct read-out and hard.copy capability for all thermocouple temperatures.
The range ex-tends from 30°F to 2200°F.-
D.
Trend capability for the thermocouples is available on dema~d on the trend typewriter located on the operator's console.
E.
Alarm capability is provided consistent with operator procedure requirements:
When reactor power
<0.25% alarm at 1200°Fi when reactor power >.025%
alarm at.630.°F.
F.
The operator-display device is human-factor designed as noted in Essex Corp Review of February, 1980.
*. 3.
A backup display is available, but it does not meet the
~tated re~uiremerits as follows:
M P80 123 06/1
, --...... -.. --.~.,......... ~-,..~__,_.~~
.... -~""*~':."'";4-1'. '""...,....~----..... '!'.:'...,............ ~~1'.:""'¢'"'~.* ~-.~'l ~.:**,.,.~-x... ~~-1*r::**] ~:- * ~.,........ -*---
--:.~-,....-:>>:--J'!--...,............. '"""";'f""'"'...., **,. ':" *.,..,..::::..
'..... All thermocouples may be read, but only one at a time.
The range is 370-400°F or 670°-700°F and not required 200°F to 2300°F.
The read-out meter on the flux mapping panel is located in a room adjacent to the control room and may not allow the operator to read 16 thermocouples within a time interval of less than 6 minutes.
The backup display is not referenced in the emergency procedures.
4.
The primary display CRT is located on the control console.
The hard copy and trend displays are located on the operator's computer console which is located directly behind the operator.
5.
The instrumentation does not meet Appendix B, "Design and Qualification Criteria for Accident Monitoring Instrumentation.
See Items 6 -
9 below.
6.
The primary display channel utilizes the station com-puter which is energized from an uninterruptible power supply.
The primary CRT display and operators console which supplies direct digital readout and trend capabil-ity are also energized from the same power supply.
The primary display channel and associated hardware are not
7.
8.
9.
Class lE.
The backup display is supplied from redundant station power sources.
The backup display and hardware are not Ciass lE.
The reference junction heater power is derived from the
. same source as the backup display.
The incore thermocouples, their electrical connectors and reference junction boxes are not environmentally qualified as required in Appendix B, Item 1.
The com-puter and display have not been environmentally qualified or seismically qualified.
The primary display channel uses a computer with an estimated reliability of 98 -
99%.
The overall primary display channel reliability is 90 -
95%.
There is no reliability figure available for the backup display.
The incore thermocouples are not purchased to the stated Q.A. requirements.
M P80 123 06/2
\\
Instrumentation for Detection of Inadeauate Core Coolin in PWRs and BWRs Section 2.1.3.b NRC' Position
1.
Licensees shall develop.procedures to be used by the operator to recognize inadequate core cooling with cur-rently available instrumei::itation.
The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions.
A detailed description of the analyses needed to ~orm the basis for operator training and procedure develop-ment shall be provided pursuant to another short-term
requirement, "Analysis of Off-Normal Conditions, Includ-ing Natural Circula.tion" (see Section 2.1. 9).
In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of cool-ant saturation and condition.
Operator instruction as to use of this meter shall include consideration that is not to be used exclusive of other related plant para-.
meters.
2.
Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed.
for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-inter-pret *indication of inadequate core cooling. A descrip-tion of the functional design requirements for the system *shall also be included.
A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a
schedule for installing the equipment shall be provided.
Response
*.The existing instrumentation available in the Control Room is sufficierit*to recognize inadequate core cooling. The indicatio~s available for determination of core heat removal are:
a.
RCS delta T less than full load delta T.
b~
RCS or co~e exit thermocouple temperatures constant or decreasing *.. *
. '~
M P79 54 01/10 Salem 1 & 2 J.4N. 1 1980
1 I
~
/
c.
Steam generator pressure constant o~ decreasing at a rate equivalent to thci rate of decr~asc of RCS tempera-tures while maintaining steam generator level with continuou~ auxiliary feedwater.
A further guid~ for recognition of inad~quate core cooling is the recent addition of a computer/CIT display for sub-cooling.
The significant parameters which are continuously displayed a~e reactor coolant differential pressure (P act-ual - P saturated) and differential t~~perature (T saturated
- Tactual).
Alarms are set for pressure differen~ial less than 200 psi and temperature differential less than SOOF.
The computer program is predicated on tne hottest in-core thermocouple reading.
The CRT matrix of in-core thermo-couples will display the location of th~ hottest in-core thermo~ouple. A dedicated ~ecorder has been provided at the computer conscle to record differential pressure and temperature.
Additional information is provided in T~ble 0
2.1.3.b-l.
Err.ergency Procedures, EI 4.4, "LOCA", and EI 4.6, n1oss of Secondary Coolant," have been revised to address the use of this computer progra~ to monitor the margiri of subcooling in the Reactor Coolant System.
PSE&G i~ a member of the Weitinghouse Operating Plant Owners' Group.
Westinghouse, under the direction of the Westinghouse Owners Group, is performinglfurther analyses to aid in selection of.more direct indicators of inadequate M P79 54 01/11 Salem 1 &- 2 MAR* 2 & 1960
. *-- -.... _..... -.... *-**---**--...... ~**~...._......
(
core cooling, and to serve as a basis for augmented emer-9ency procedures.
A preliminary report on inadequate core cooling was sub-
~itted to the NRC on October 30, 1979 bi the Owners' Group.
The Salem 2 Em~rgency Procedures will br? revised on an interim basis to spec1fy precautions and operator act~ons to recover from a condition in which the core has experienced inadequate coolin9.
These interim revisions will be avail-able for NRC review prior to power tesring.
The station procedures will be further updated af~~r completion of the final Owners' Group report.
It is our intent to install a device to indicate reactor vessel water level on Salem 2 *. This device will be similar to the proposed Westinghous~ design for VEPCO's North Anna plant.
Installation Of this device will b~ accomplished during the fj~st refueling outage subject to equipment availability*~nd acceptability of the design to provide an
.unambiguo~~ indication 6f inadequate ccre cooling.
\\
M P79 S4 01/12
-B-:
Salem l & 2
/*
MAR 2 8 1980
'".... _....... ___,._..,...................,...~.. ---... - -.... ~-****--***-
--.....,,.-.--..... -~........,--.. ~... *-~--"'~~.,,.-r:_~~.,..-~. "!';'~.~~~:*~ -~~* -.~i;~-:~~-:--~-=~~-~- ~~~"3~~:~-~'."::~:*~:.><<:-:~~~-~.T~Y!:~-~~:--*~'":'~~~*~:- ~-,...=--:.. 4~~~*~*~:~*~~*---:*-r:-* :,"!':J:.*..,..... ~ ~.......
T/\\nI.. l:~ 2.1.3.b-l SUOCOOLING METER INPORM!\\TIOJ ttiselay
Information DisplaJed (T-Tsat, Tsat, Press, etc.)
Display ~
(Analcg, Digital, CRI')
O:>ntinuous or on D::mand Single or Iedundel?*~ Display I.Dcation of Display Alarms (include setpoints)
I OVerall uncertainty (°F 1 PSI)
~ge of Display Qualifications (seismic, environmental,,,;i!EEE323)
Calculator
'fype (process corrpLter, dedicated digital or analog calc. ).
If process corrputer is used specify avail.
ability *. (% of time)
Single or redundant calculators Selection logic (hJ:ghest T.-, lo,.,est press)
Qualif.ications (sei.sinic, environrrental, IEEE323)
Calculational 'lechnique (Stearn Tables, Functional Fit, ranges)
Input Tcrrpcraturc (RrD'~ or T/C's)
Tcnper<<ture (number of sensors and locntions)
Range of tcnpcratu1.*e. scmnors M P79 5*1 01/13
-Sa-Note 1 CRI' and Analog I Pen Recorder Note 1 Note 2 Note 3 Note 4 Note 5 Note 6 None
=-
Process Computer I
(Note 8) 90%.:..95%
O*
(Estimated)
Single Note 7 None Steam Tables
-.I (Note 9) 32<~F<705 12,g>~ia<3204 Chrorni?l/hl~l T/C 65 Incore T/C's 30-2200°1" S..-llcm l & 2
r.
I
~-.... 1 J, I'... ~)
'~ ~ \\
I*
, ~.
(
\\..._
TADLI~ 2.1.3.b-1 (CONTINUED) rtainty* of tenperature sensors ( °F.at 1)
Qualifica~ions (seismic, environmental, IEEE323)
Pressure (specify instn.nnent used)
Pressure (nunber of senso~s and locations)
Range of Pressure sensors Uncertainty* of pressure sensors (PSI at 1)
~alifications (seism~c, environmental, IEEE323j Backup Capability Availability of Teffii? & Press Availability. of Steam Tables etc.
Training of_ ~raters
+ocedures*
!: 5°F None Barton 763 2-#11 Hot Leg 0-3000 psiQ
+/-130 psi (Small Break Conditions)
Qualif ie0, Safety Grade Main Console Indica-tion
~ Cbnversion Curves Completed Completed
Uncertainties are not affected by differences in RCS flow conditions.
Thermocouples are loeated in hottest regions and pressure measurement is independent of fl~~ conditions.
Notes*
. *~...
~.
1. Continuous infonnation displayed on analog pen recorder:
(Tsat-Tact), (Pect-Psat) *. Information available on demand:
(Tsat-Tact),(Pact-Psat), Psat, Pressure.
Tenperature and location of hottest in-core T/C.
2. 'ihe information available on demand can be displayed on the CRT or
_ trend typewriter;_-..
.. ~ *.
M 1>79 S*1 01/14
.. -ab-*
S.'llcm 1 & 2 *
,,,.., n *.* 1 r' n G BC'°H !. l
~!,,
~ i
\\.
TMI,F.: 2.1. 3
b-1
( CCNI'l NU8D)
3.
'!he Cm' is locc ted on the ccntc~lcft portion of the main control console.. 'Ihe o~ber displays are available at the operator's com-
_ puter console.
4
Alarms:
(Tsat-'t'act) - *lC?SS than 50°F subcooling (P-Psat) - less than 200 psi Tenp. *
any T/C greater than 630°.F If these alanns occur, they will be displayed even if the subcooling calculation program has not been r,equirested by the operator.
I
5. overall uncertainty is a factor of the uncertainty in the tempera-ture and pressure measurements and the result~ng potential error when using the steam tables *. '!he uncertainties of these devices are listed separately on the table *
. 6*.- Analog Ranges:
(Tsat-Tact) 0-120°F Di ff; (P-l*sat) 0-1000 psi Di ff CRT Range:
N/A.
7. '!he selection logic used is:
highest in-core T/C reading
average of two reactor ccx:>~ant pressures.-
The operator is provided with the capability t.o reject the selected
.. T/C; the next highest reading T/C will then be* automatically se-
. lected for _the ~alculation.
A reaso~ability check of the two pressure readings is performed 0 by the corrputer. If the readings indicate that one of the measurements is invalid, the corrputer will reject the invaiid reading.
8. '!he process co-:lputer is pc:Mered from a vital bus through an inverter*
with battery ba~up.
9. 'Ihe subcOoling calculations are perfonnad on an 8-10 second basis
M P?9 54 01/15
-Be-Salem 1 & 2 APR... 11980
~
2.1.3.b Subcooling Meter As tndic1ted in our response to NUREG-0578, a computer based sub-coo11ng metering system has been installed in both units. Th1s system uses the tn-core thennocoup1es and redundant reactor coolant system pressure measurements to calculate the margin to subcoo11ng conditions end display (CRT) the pertinent parameters to the operator.
The attached sheets provide the tabular infonnation requested.
~.. ! c,. L ATTACHMENT TO TABLE FOR 2.1.3.b A.
lnfonnation di sp1ayed:
(TSAT-TACT)
TSAT * (P-PSAT) 1 PSAT
Pressure,
'Teffl>era ture and location of hottest in-core T/C.
8. *** The tnfonnation display fs continuous once demanded by the operator.
Display features automatic thirty-second update plus operator recall capability for faster update.
C.
The main display is a single CRT but other display devices are available to provide selectable parameter outputs. These include an analog pen recorder, digital display, and trend typewriter~
D *. The CRT is located on the center-left portion of the main control console.
The other displays are *available at the. operator's computer console.
E.
Alanns:
(TSAT-TACT) - less than so0r subcooHng (P-PSAT) - less than 200 psf Temp. - any T/C greater than 630°F If.these alanns occur. they will be displayed even if the. subcooling cal-culation program has not been requested by the operator.
F.
Overall uncertainty 1s a factor of the uncertainty in. the temperature and pressure measurements and the resul ~ing potential.error when using the steam tabl.es *. The uncertainties of.these devices are 1 isted separately on the tabl,~
G.
The selection logic used 1s:
h;ghest in-core T/C reading
. average of t~o reactor coolant pressures.
. *The operato-r h provided with the capability to reject the selected T/C;
the next highest.reading T/C will then be automatically selected for the
calculation>,*
A reasonab111ty check of the two pressure readings is perfonned by the computer. If the readings indicate that one of the measurements is 1n-valid1 the computer w111 reject the invalid reading. *
~
_.;.,..~,*&iill. ~-----_..
_ _.,....-1:_¥ *., ---*...
H..... -------'----------
~--- ----
,l!tfOR""'llOW Rtyu1R[U ON lH[ su~COOL1N~ M[l[k Dhpl*.X.
-lnfo,..ation Displayed (t-lsat. Tsat. Press. et.c.)
D1sp1ay type (Analog. Digital, tRl)
Continuous or on Uemand Single or aedundant Dhplay Location of ~isplay Alarms (include *etpointsl Over1ll uncertainty (~f. PSll Range of Display
~ua1ifications (sei~ic, environn~ntal, 1EEE32Jl
~efer to attachment CRT -
B -
!D
E
f N/A
CRT None
£_a1culator T1P* (process convuter, oeoicateo oigita1 or ana1og c1lc.I Process Computer lf process cOOl'uter is use* specify 1vai1abl1itJ* I~ of timel 901-95: Estimated Single or redundant calculators
__ S_i_n..,gl_e _____ _
Selection Logic (hignest 1., lowest press)*
_G ______ _
None Qualifications (seismic, environmental, 1EEE3i3l ta1cu11tiona1 Technique (Steam T1bles, functional Fit, r1ngesl SteaM Tables
,2s.0~<705 12SPsia<3204 J.!'put Te~tr1ture (RlU's or 1/t'sl Te~erature (number of sensors and locations) lange of ttnverature sensors Chromel/Alumel T/C 65 Jncore l/C --
(.
( aerialn\\J* *' '-enture r.enHn 1*r ai' I
~altft~attons (1tt11111t, envtronaen\\.ll, llEt~Z~l
(
frtnYrt h~cU1 tn,trveent u*tOl frt5Mlrt (ft~ber of 1tnsor5 ano \\ocatton*l lange of Prt,,urt 1en5or5
~nctrt..ainty* of prts*urt *tnsors (PSl at \\ l Qualifitattons hthmite environmental, Utt30)
Baclup C1pebiltt1 Av1ilobilit1 of Tt~ I Press Availability of Stea~ laDlts ett.
1ra1ning of operators.
ProttelurU Mone Barton 763 J.. Ill Hot L!9
_o.. 3000 ps 19
+150
-:JOO psi (LOCA Conditions)
Qua11f1ed Main Console Indication Conver2ion Curves Lompleted Completed
Uncertainties are not affected by differences In RCS flow conditions. Thenno-couples are located In hottest regions and pressure measurement Is Independent of flow conditions.
.. \\l..
1.0- PtJRPOSE EMERGENCY INSTRUCTION 1-4.4 LOSS OF COOLANT
<LEAKAGE GREATER THAN MAXIMUM CHARGING FLOW>
1.1 THIS INSTRUCTION PROVIDES THE NECESSARY OPERATOR ACTIONS REQUIRED TO PROVIDE MAXIMUM CORE COOLING TD MINIMIZE CORE DAMAGE FOLLOWING A LOSS OF COOLANT ACCIDE~l.
1.2 THIS INSTRUCTION CONTAINS THE STEPS REQUIRED OF THE OPERATOR TO SWITCH FROM THE INJEC-TION TO RECIRCULATION PHASES OF CORE COOLING AT THE APPROPRIATE TIMES.
1.3 THIS INSTRUCTION INCLUDES THE APPROPRIATE OPERATOR ACTIONS REQUIRED TO COPE WITH THE FOLLOWING FAl~URES.
1.3.l Loss OF A RESIDUAL HEAT REMOVAL PUMP DUE TO EITHER OF THE FOLLOWING:
A.
FAILURE OF THE ASSOCIATED SJ44, SJS SUMP VALVE, TO OPEN.
B.
FAILURE OF AN RHR PUMP.
1.3.2 Loss OF OFFSITE POWER WITH:'
A.
ALL DIESELS OPERATING B.
FAILURE OF A SINGLE DIESEL.
2.0 INITIAL CONDITIONS 2.1 SAFETY INJECTION HAS BEEN INITIATED AND IT HAS BEEN DETERMINED BY USE OF SECTION 5.0,
10ENTIF1cAT10N oF FoLLow-uP AcT10Ns" oF EI 1-q.o, "SAFETY INJEcnoN INITIATioN", THAT A LOSS OF COOLANT ACCIDENT HAS OCCURRED.
3.0 ltr.EDJAIE ACTIONS 3.1 VERIFY THAT ALL IMMEDIATE AND SUBSEQUENT ACTIONS DESCRIBED JN El J-q,Q, "SAFETY }NJECTJO~
INITIATION" HAVE BEEN PERFORMED, (OMPLETE ANY ACTIONS WHICH HAVE NOT BEEN PREVIOUSLY COMPLETED.
3_.2
)F CONTAINMENT PRESSURE REACHES THE HJ-HI SETPOINT OF 23.5 PSIG, VERIFY THE FOLLO"-'lNG AUTOMATIC ACTIONS HAVE TAKEN PLACE BY OBSERVING THE INDICATIONS ON THE STATUS_ PANEL ON RP-ff
3.2.1 CONTAINMENT SPRAY ACTUATION 3.2.2 CONTAINMENT PHASE -*B" ISOLATION 3.2.3 MAIN STEAM ISOLATION 1 OF 37 REV. 10
c
~ !.
1-4.4 3.3 I* CONTAINMENT PHASE "B" tso1.ATIDN IS ACTUATED. TRIP ALL REACTOR Coo1.ANT PUMPS ~!THIN FIVE MINUTES*
Jf,0 S1JBSEQUENT ACUONS - PART l - COLD 1 EG lNJECTlOti
. '~
4.1 (HECK THE FOLLOWING INDICATORS ON THE I
CONTROL CONSOLE TO ENSURE BORATED WATER IS BEING INJECTED INTO THE REACTOR COOLANT SYSTEM.
4.1.l BoRON.INJECTION TANK PRESSURE INDICATING RCS PRESSURE.
4.1.2 CHARGING PUMPS DISCHARGE FLOW 4.1.3 No. 11(21> SAFETY INJECTION PuMP DISCHARGE FLOW WHEN RCS PRESSURE IS < ~ 1500 PSJG, 4.1.4 No. 12(22> SAFETY INJECTION PuMP DISCHARGE FLOW WHEN RCS PRESSURE IS < ~ 1500 PSIG.
4.1.5 No. 11(21) RHR INJECTION fLOW WHEN RCS PRESSURE IS < ~ 170 PSIG*
4.1.6 No. 12(22) RHR INJECTION fLOW WHEN RCS PRESSURE IS < ~ 170 PSIGo 4.2 IF CONTAINMENT PRESSURE l:iA.S. tiQl INCREASED TO THE H1-HI SETPOlNT Of 23.5 PSIG, PROCEED
~ROCEED TO siEP 4.5.
4.3 IF CONTAINMENT_ PRESSURE HAS. INCREASED TO THE HI-HI SETPOJNT OF 23.5 PSIG1 VERIFY THE FOLLOWING BY OBSERVING THE STATUS.
PANEL ON RP*4 AND/OR.THE CONSOLE INDI-CATIONS *.
s~*r~ UNIT 1/UNIT 2 2 OF 37
~**
~;;...:.*
4.0 THE SUBSEQUENT ACTIONS JN THIS INSTRUCTION WILL ADDRESS BOTH THE SMALL LOCA AND THE DBA.
4.2 CONTINUE TO MONITOR (ONTAlNME~*
PRESSURE.
A RELATIVELY SLOW BUILDUP IN
CONTAINMENT PRESSURE IS INDI-CATIVE Of A SPUILL LOCA.
REV. 10
.\\ 1
~.3.1 CoNTAINMENT SPRAY HAS INITIATED
1.
CHECK THAT THE FOLLOWING PUMPS HAVE STARTED, IF A PUMP FAILS TO START, ATTEMPT TO START MANUALLY FROM THE CONTROL CON-SOLE*
No. 11(21) CONTAINMENT SPRAY PUMP No. 12<22) CONTAINMENT SPRAY PuMP 2,
CHECK THAT THE FOLLOWING VALVES HAVE OPENED.
IF A VALVE FAILS TO OPEN, ATTEMPT TO OPEN MAN-UALLY FROM THE CONTROL CONSOLE*
11(2l)CS2 DISCHARGE VALVE 12(22)CS2 DISCHARGE VALVE 1(2)CS16 SPRAY ADD TANK DISCH VALVE 1(2)CS17 SPRAY ADD TANK DISCH VALVE
3.
CHECK THE ADDITIVE TANK LEVEL INDICATOR AND OUTLET FLOW IN-DICATOR ON THE CONTROL CONSOLE TO ENSURE THAT THE SODIUM HYDROXIDE (NAOH) SOLUTION IS BEING INJECTED INTO THE CONTAINMENT SPRAY SYSTEM.
IF THE LEVEL IS NOT DECREASING AND NO FLOW IS INDICATED, DIS-PATCH AN OPERATOR TO VERIFY THE LEVEL LOCALLY AND TO ENSURE THE FOLLOWING MECHANICAL VALVES ARE OPEN.
11(2l)C~20 E.DucTOR SUPPLY VALVE 12(22)CS20 [DUCTOR SUPPLY VALVE 4.3.2 ISOLATION PHASE *B* HAS TAKEN PLACE 1,
(HECK TO SEE THAT THE FOLLOWING VALVES HAVE CLOSED BY OBSERVING THE STATUS PANEL ON RP-~ AND ACKNOWLEDGE 'ON THE APPROPRIATE CONTROL CONSOLE BEZEL*
SALEM UNIT l/UNIT 2 A.
IF ANY VALVE HAS FAILED TO CLOSE, ATTEMPT TO CLOSE IT FROM THE CONTROL CONSOLE BEZEL*
3 Of 37 1-4.4 REv. 10
i* "
~'
'~
CoMPONENT COOLING l<2>CC117 RCP COOL.ING WATER INLET 1<2>CC118 RCP COOLING WATER INLET 1<2>CC136 RCP BEARING OUTLET lC2>CC131 RCP THRM BAR DISCH FLOW lC2>CC190 RCP THERMAL BAR DISCH lC2>CC187 RCP BEARING OUTLET IF ANY REACTOR COOLANT PUMPS ARE RUNNING, THEY MUST BE TRIPPED AT THIS TIME,
~.3.3 filAJN STEAM ISOLATION HAS TAKEN PLACE
}, (HECK THAT THE FOLLOWING VALVES HAVE CLOSED BY OBSERVING THE STATUS PANEL AND ACKNOWLEDGE ON THE APPROPRIATE*
CONTROL CONSOLE BEZEL, IF ANY VALVE HAS FAILED TO CLOSE, ATTEMPT TO CLOSE lT FROM THE CONTROL CONSOLE BEZEL, 11C2l>MS167 No. 11<21) STEAM GEN STOP VALVE 12C22>MS167 No. 12C22> STEAM GEN STOP VALVE 13C23>MS167 No. 13C23) STEAM GEN STOP VALVE l4C24>MS167 No. 14C24) STEAM *GEN STOP VALVE..
llC21>MS18 No. 11C21) STEAM GEN
STOP WARMUP VALVE
12C22>MS18 No. 12C22) STEAM GEN STOP WARMUP VALVE l3C23>MS18 No. l3C23) STEAM GEN STOP WARMUP VALVE
*l'H24H'1Sl8 No. 14C24) STEAM GEN STOP WARMUP VALVE llC21>MS7 No. 11C21> STEAM GEN DRAIN VALVE
.. 12C22>MS7 No. 12C22) STEAM GEN DRAIN VALVE l3C23>MS7 No. 13C23> STEAM GEN DRAIN VALVE l4C24>MS7 No. l4C24) STEAM GEN DRAIN VALVE SALEH UNIT l/UNJT 2
.. OF 37 1-....,
Rev. 10
t *
,q IF THE RWST LEVEL IS DROPPING RAPIDLY
(> "' 2 FEET PER MINUTE) PROCEED TO STEP 4.12.
... s. *tHECK CLOSED THE PowER OPERATED RELIEF
. VALVES, 1C2>PR1 & 2 I
4.6 IF THE PRODAC 250 IS AVAILABLE:
4.6.l INITIATE CRT TEST 41.
IF THE CORE EXIT THERMOCQUPLE IS <1200*f, PROCEED TO STEP 4.8.
IF THE HIGHEST CORE EXIT*THERMOCOUPLE READS ~1200°f, PROCEED WITH STEP 4.6.2.
4.6.2 INJTJATE A SHORT FORM THERMOCOUPLE f'\\AP TO PRINT ON THE TREND TYPEWRITER.
4.6.3 IF FIVE OR MORE CORE EXIT THERMOCOUPLES READ !1200°F A CONDITION OF INADEQUATE CORE COOLJ NG IS DE '/ELOPING, PROCEED AS FOLLOWS UNTIL CORE COOLING JS RE-ESTABLISHED:
A.
ESTABLISH MAXIMUM AUXILIARY fEEDWATER FLOW.TO ALL INTACT STEAM GENERATORS.
.. '.DUMP STEAM AT THE MAXIMUM RATE AVAIL-
.ABLE FROM ALL STEAM.GENERATORS TO w.i1 CH Aux1 LI.ARY FEEDWATER FLOW HAS
~EEN ESTABLISHED.
B.
ESTABLISH MAXIMUM SAFETY INJECTION
-FLOW VIA THE BORON INJECTION TANK.
~
c. 'IF NEiTHER THE CENTRIFUGAL CHARGING PuM*Ps NOR SAFETY INJECTION PUMPS ARE DELIVE~JNG WATER TO THE RCS,
~
STEAM DUMP FROM THE STEAM GENERA-TORS IS NOT AVAILABLE, AND.,
Aux1LiARY fEEDWATER IS NOT AVAIL-ABLE TO ANY STEAM GENERATOR, 5 nF 37 1-4.4 4.4 IF RWST LEVEL IS DROPPING RAPIDLY / IT JS EVIDENT A LARGE LOCA IS IN PROGRESS AND
. UNLIKELY THAT RCS PRESSURE WILL STABILIZE AT AH ELEVATED PRESSURE.
4.5 ELIMINATES PRV's As T~E souRcE
OF LEAKAGE, 4.6.l IF CORE EXIT TEMPERATURE IS.
<12QQ*f, CORE COOLING IS
ADEQUATE, 4.6.2 REQUIRES '\\.2-3 MINUTES TO PRINT OUT.
CRT TEST 13.
4.6.3 f.ORE COOLING IS RE-ESTABLISHED WHEN CORE EXIT TEMPERATURES ARE DECREASING, A.
CONDENSER DUMP JS. PREFERRED.
ATMOSPHERIC DUMP JS ACCEPTABLE.
PUMPS SHOULD BE R!JaiNJNG.
ATMOSPHERIC OR CONDENSER.
Rev. 10
-*-***~"... ** ----- -------------~--==------
ltfE!L OPEN BOTH PRESSURIZER POWER OPERATED RELIEF VALVES, lC2)PRl & 2, TO REDUCE RCS PRESSURE TO THE POINT AT WHICH SAFETY INJECTION FLOW IS BEING DELIVERED TO THE CORE, MAINTAIN 1(2) PRl & 2 OPEN UNTIL AUXILIARY fEED IS ESTABLISHED TO ALL STEAM GENERATORS.
D.
IF NONE OF THE ABOVE ACTIONS RESULT JN A REDUCTION IN CORE EXIT THERMOCOUPLES, Afill, IF COMPONENT COOLING WATER JS AVAIL-ABLE *To THE REACTOR COOLANT PUMP MOTOR COOLERS, ~
START ONE REACTOR COOLANT PUMP 4.7 IF THE PRODAC 250 IS liQI AVAILABLE:
4.7.1 MoN1roR THE WtoE RANGE Rr.s Hor LEG TEMPERATURES.
4.7.2 IF THREE HOT LEG TEMPERATURES INDI-CATE ~700*F A CONDITION OF INADEQUATE CORE COOLING MAY BE DEVELOPING.
PROCEED AS FOLLOWS UNTIL CORE COOLING JS RE-ESTABLISHED:
A, ESTABLISH MAXIMUM AUXILIARY FEEDWATER FLOW TO ALL INTACT STEAM GENERATOR$.
DUMP STEAM AT THE MAXIMUM RATE AVAIL-ABLE FROM ALL STEAM GENERATORS TO WHICH AUXILIARY fEEDWATER FLOW HAS BEEN ESTABLISHED, B.
ESTABLISH MAXIMUM SAFETY INJECTION FLOW VIA THE BORON INJECTION TANK.
C.
JF NEITHER THE CENTRIFUGAL CHARGING PUMPS NOR SAFETY INJECTION PUMPS ARE DELIVERING WATER TO THE RCS,
~
STEAM DUMP FROM THE STEAM GENERA-TORS IS NOT.AVAILABLE, AND.,
AUXILIARY fEEDWATER IS NOT AVAIL-ABLE TO ANY STEAM GENERATOR, J-Lj I Lj INSURE 1C2>PR5 & 7 ARE OPEN, SAFETY INJECTION PUMP SHUTOFF
'\\.1520 PSJG, RHR PUMP SHUTOFF '\\.170 PSJG.
o.
THE CRITERIA FOR TRIPPING RCP's AT 1500 PSJG DOES NOT APPLY.
THE REQUIREMENTS FOR SEAL INJECTION FLOW, SEAL LEAKOFF FLOW, THERMAL BARRIER CCW FLOW AND No. 1 SEAL tP DO NOT APPLY, Li.7.1 ON RfcoRDERS ON 1C2)RPq, L!.7.2 (ORE COOLING JS RE-ESTAB-LISHED WHEN CORE EXIT TEMPERATURES ARE DECREASING.
A.
CONDENSER DUMP IS PREFERRED ATMOSPHERIC DUMP JS ACCEPTABLE.
PUMPS SHOULD BE RUNNING.
ATMOSPHERIC OR CONDENSER
i f: C.
~
OPEN IOTH PRESSURIZER POWER OPERATED RELIEF VALVES, 1C2>PR1 & 2, TO REDUCE RCS PRESSURE TO THE POINT AT WHICH SAFETY INJECTION FLOW IS BEING DELIVERED TO THE CORE, f'IAINTAIN 1<2> PRl & 2 OPEN UNTIL AUXILIARY fEED IS ESTABLISHED TO ALL STEAM GENERATORS, D,
IF NONE OF THE ABOVE ACTIONS RESULT IN A REDUCTION IN CORE EXIT THERMO-COUPLES, Arill.,
IF COMPONENT COOLING WATER IS AVAIL-ABLE TO THE REACTOR COOLANT PUMP MOTOR COOLERS, IH£tL START ONE REACTOR COOLANT PUMP.
~.8 SAFETY INJECTION MAY BE TERMINATED IF A.l..L OF THE FOLLOWING EXIST, ti..
~
1.
RCS PRESSURE JS >2000 PSIG AS INDICATED ON 2/3 PRESSURIZER PRESSURE INDICATORS, ~
2.
ACTUAL PRESSURIZER LEVEL JS >20%
AfilL
3.
TOTAL AUXILIARY fEEDWATER FLOW TO ALL STEAM GENERATORS >42xl04 LB/HR OR ACTUAL LEVEL JN AT LEAST ONE STEAM GENERATOR !5% AS INDICATED ON 2/3 NARROW RANGE INDICATORS, AND THE JNCORE THERMOCOUPLES AND WIDE RANGE Tij __ ~RE STABLE OR DECREASING, Afi.lL
~. RCS TEMPERATURE IS MORE THAN so*F SUBCOOLED AS INDICATED BY EITHER THE WIDE RANGE TH INDICATORS OR THE IHCORE THERMOCOUPLES,.
q,8.1 IF ALL OF THE ABOVE ARE MET, PROCEED TO El 1-4.2, *RECOVERY fROM SAFETY INJECTION*.
SALEM UNIT l/UNtT 2 7 OF 37 I-If,If INSURE lC2>PR6 & 7 ARE OPEN.
SAFETY INJECTION PUMP SHUTOFF
"-1520 PSIG, RHR PuMP IHUTorr ~170 PSJG,
o.
THE CRJTERIA FOR TRIPPING RCP's AT 1500 PSIG DOES NOT APPLY, T~E REQUIREMENTS FOR SEAL INJECTJON FLOW, SEAL LEAKOFF FLOW, THERMAL BARRIER CCW FLOW AND No, l SEAL tP DC NOT APPLY i4, 8 f"toNITOR FOR THESE CONDlTJONS THROUGHOUT THIS INSTRUCTION.
VERIFY RCS PRESSURE BY ~OM PARING IT TO BIT PRESSURE, BJT PRESSURE SHOULD BE SLIGHTLY
HJGHER,
2.
CORRECT PRESSURIZER LEVEL FOR REFERENCE LEG HEATUP IAW APPENDIX 2,
3.
CORRECT STEAM GENERATOR LEVELS
¥0~ REFERENCE LEG HEATUP IAW APPENDIX 2.
q, IF TEMPERATURE ~MAINS BELOW THE TSAT -so*F CURVE ON THE PRESSURE-TEMPERATURE CURVE, so*F SUBCOOLING JS ASSURED
4.8.1 EJ 1-~.2 HAS ALL THE STEPS REQUIRED TO TERMINATE SAFET~
INJECTION AND THE CRITERIA F RE-INITIATION, REv. 10
4.8.2 IF RE-INITIATION OF SAFETY INJECTION WAS REQUIRED IAW THE CRITERIA JN El J-4.2, PROCEED WITH THIS INSTRUCTION FROM THIS POINT, 4.9 IF REACTOR COOL.ANT PR.ESSURE STABILIZES ABOVE THE SHUTOFF HEAD (~170 PSIG) OF THE.RESIDUAL HEAT REMOVAL PUMPS, PROCEED AS FOLLOWS:
4.9.l RESET SAFETY INJECTION BY DEPRESSING
. BOTH TRAIN *A" AND TRAIN *B" SI RESET PUSHBUTTONS ON THE SAFEGUARDS ACTUATION BEZELS ON THE CONTROL CONSOLE.
4.9.2 RESET THE SAFEGUARDS LOADING SEQUENCE BY DEPRESSING THE EMERGENCY LOADING RESET PUSHBUTTON ON THE CONTROL CONSOLE FOR lA,, lB1.AND 1C:C2A,, 2B AND 2C) DIESEL GENERATORS.
S.t1.LEM l!?HT l/UNIT 2 8 OF 37 1-lf."
4.8.2 JK1 tiQl ATTEMPT TO TERMINATE SAFETY. INJECTION AGAIN.
fOR CERTAIN SMALL BREAKS t\\AXIMUM CHARGING FLOW MAY NOT KEEP RCS PRESSURE
>2000 PSIG.
4.9.l IF AT ANY TIME AFTER THE SAFETY INJECTION AND CONTAINMENT SPRAY StGNALS ARE RESET, A BLACKOUT SIGNAL IS RECEIVED, THE VITAL BussES WOULD BE STRIPPED AND THC: 8LAC_KOUT LOADS WOULD BE SEQUENCED ON BY THE SEC TttE RHR, SAFETY INJECTION, AND CONTAINMENT SPRAY PUMPS AND THE CoNTAtNMENT fAN CotL UNtTs l'!lJ.L tiQI BE RESTARTED.
THESE ~
Bf MANUALLY RESTARTED ONCE THE LOADING SEQUENCE IS COMPLETE AS INDICATED BY *LOADlNG cor~LETE.
LIGHTS ON THE lA, lB, 1CC2A, 2E 2C> Dt ESEL BEZELS ON THE CONTR:
- CONSOLE*
ltttS IS TO BE ACCOMPLISHED lN SUCH A MANNE~ AS TO PREVENT OVERLOADING THE DIESELS.
THE LOADS SHOULD BE APPLIED AT
~ 10 SEC. INTERVALS.
00 tiQI RESTART THE EQUIPMENT I MANUALLY INITIATING SAFETY IN-JECTION OR C~Alt~MENT SPRAY,.
THIS MAY RESULT IN UNDESIRABLE VALVE OPERATIONS
4.9.2 Tt4E STATUS LIGHTS ON RP-4 AS-
. SOCIATED WITH SEC LOADING SHOULD GO OUT.
REv. 10
"'a**-*---**
---~-*~ -_____.. -
~
(
11.9.3 STOP No. 11 I 12C21 I 22) RESIDUAL HEAT REMOVAL PUMPS.
tAUllOli:
IF REACTOR CooLANT PRESSURE DECREASES TO BELOW THE SHUT-1
OFF HEAD (...,170 PSIG) FOR THE RESIDUAL HEAT REMOVAL PUMPS, RESTART THE PUMPS.
q,9,q OPERATE THE SAFETY INJECTION PUMPS AS REQUIRED TO MAINTAIN PRESSURIZER LEVEL BET~EEN 50% AND 90%.
4.10 COMMENCE TAKING THE PLANT TO COLD SHUTDOWN CONDllIONS BY COOLING DOWN AS FOLLOWS:
11.10.l f'\\ANUALLY CONTROL THE AUXILIARY FEEDWATER CONTROL VALVES CAf21) TO MAINTAIN ACTUAL STEAM GENERATOR LEVELS AT APPROXIMATELY 33'1.
tiQif.
IF No. 13(23) AFW PuMP 1s RUNNING IT WILL BE NECESSARY TO CONTROL THE Afll VALVES*
~.10.2 PLACE THE STEAM DuMP IN MAIN STEAM PRESSURE (ONTROL MODE AND PERIODI-CALLY, REDUCE THE PRESSURE SETPOl~T OF THE MAIN STEAM PRESSURE CONTROLLER BY DEPRESS I NG THE SE1P01N1 DECREASE PUSHBUTTON I 11.10.3 As APPLICABLE, TAKE THE PLANT 10 (OLD SHUTDOWN CONDITIONS IAW 01 1-3.6,
Hot.STANDBY TO (OLD SHUTDOWN*,
....,*.~..
I 11.10.11 1F THE PRODAC 250 COMPUTER IS AVAILABLE1 INITIATE CRT TEST No. 111, *toRE TEMPERATURE/
PRESSURE MONITOR PROGRAM**
S&Ji~ Ui~lT 1/UNlT 2 9 OF.37, 1-4.11 4.9.3 THE RHR Pu~s SHOULD NOT BE ALLOWED TO RUN ON RECIRC FOR MORE THAN ~30 MINUTES SINCE THERE 1s NO COMPONENT tooLlHG TO THE HEAT [XCHANGER.
4.9.~ Do NOT ATTEMPT TO THROTTLE THE DISCHARGE FLOW.
11.10.l (ORRECT STEAM GENERATOR LEVELS FOR REFERENCE LEG HEAlUP IAW APPENDIX 2.
4.10.2 DECREASING THE SETPOINT INCREASES STEAM FLOW.
IF THE CONDENSER IS HOT AVAILABLE, USE THE f'S-lO's.
4.10.3 DUE TO ABN'JRMAL PLANT coN-DITlONS SOME STEPS MAY NOT BE APPROPRIATE AND SOME EQUIPMENT NOT AVAILABLE*
IT KAY BE NECESSARY TO MAINTAIN FLOW THROUGH THE BIT SINCE MAXIMUM CHARGING FLO'*i KAY NOT MAINTAIN RrS INVENTORY UNTl L PRESSURE 1 S
$1GNIFICANTLY REDUCED*
4.10.11 THIS PROVIDES A READILY AVAILABLE DISPLAY OF suB-COOLlNGo REV. 10
~i
.JI
('
~ ;
~*
. \\...
. "e' 4.10.5 MAINTAIN SUBCOOL1NG >sn*F BY INCREASING STEAM FLOW AS REQUIRED AS DESCRIBED IN STEP 4.10;2 ABOVE 4.11 WHEN CONDITIONS PERMIT, RETURN THE 4KV VITAL BussES TO NORMAL av:
4.11.l STOPPING THE EMERGENCY DIESEL GENER-ATORS IAW 01 1v~16.3.l, *EMERGENCY POWER - DIESEL OPERATION",
4.11.2 START OR STOP VITAL BUS LOADS, AS REQUIRED*
4.12 CLOSELY MONITOR RWST LEVEL*
As IT APPROACHES THE LOW LEVEL ALARM, PREPARE
. TO CHANGE FROM THE INJECTION PHASE
TO THE CoLD LEG REc1RcuLATION PHASE.
PROCEED AS FOLLOWS:
4~12.l RESET SA.FEtY INJECTION BY DEPRESSING BOTH TRAIN ;,'A" ~ND TRAIN *B" SI RESET PL1SHBUTTONS ON THE SAFEGUARDS AcTuATioN BEZELS ON THE CoNTROL CONSOLE**
11..... ? -
10 OF 37 1-4.4
)F THE COMPUTER IS NOT AVAIL-
. ABLE OR THE INDICATIONS ARE CONSIDERED NOT RELIABLE DUE TO ADVERSE CONTAINMENT CON-DITIONS, SUBCOOLING CAN BE DETER"JNED FROM EITHER THE PRESSURE-TEMPERATURE CURVE OR FROM THE STEAM TABLES.
4.12.l IF AT ANY TIME AFTER THE SAFETY INJECTION AND CoN-TAJ NMENT SPRAY SIGNALS ARE RESET, A BLACKOUT SJGNAL IS RECE1VED, THE VITAL BussE~
WOULD BE STRIPPED AND THE BLACKOUT LOADS WOULD BE SEQUENCED ON BY THE SEC.
THE RHR, SAFiJ'Y INJECTION, AND CONTAINMENT SPRAY PUMPS AND THE CoNTAINMENT fAN Co11 UNITS ltllJ.. tlQ.I BE RESTARTED
. *. l~ESE tuJil BE MANUALLY RE-STARTED ONCE THE LOADING SEQUENCE IS COMPLETE AS INDICATED BY THE *LOADING tOf"J'LETE* LIGHTS ON THE l.A.
lB, 1C(2A, 2B, 2C) DIESEL 8FZELS ON THE CONTROL CONS REv. 10
c
.."~*'
4.12.2 RESET THE SAFEGUARDS LOADING SEQUENCE BY DEPRESSING THE EMERGENCY LOADING RESET PUSH-BUTTONS ON THE CONTROL CONSOLE FOR lA, lB, 1CC2A, 2B, 2C>
DIESEL GENERATORS.
4.12.3 RFSET CONTAINMENT SPRAY, IF CONTAINMENT PRESSURE JS LESS THAN 23.5 PSIG ON 3/4 CHANNELS, BY DEPRESSING TRAIN *An AND TRAIN *a" SPRAY ACT RESET PUS.HBUTTONS ON THE SAFEGUARDS ACTUATION BEZELS ON THE CONTROL CONSOLE.
4.~,q WHEN CONDITIONS PERMIT, RETURN THE 4KV* VITAL 8USSES TO NORMAL BY:
A.
S.TOPPING THE EMERGENCY DIESEL. GEN-ERAT()RS IAW 01 IV-16.3.1, *EMERGENCY
. POWER - DIESEL OPERATION" I B,
START OR STOP VITAL BUS LOADS, AS REQUIRED.
IF A LOSS OF OFFSJTE POWER HAS OCCURRED IN COINCIDENCE WITH THE LOCA, ALIGN THE ELECTRICAL SYSTEM IN ACCORD-ANCE WITH APPENDIX }, PRIOR PROCEEDING WITH PARTS II OR JJI OF THIS PROCEDURE, 4.13 PROCEED TO SECTION 5.0, PART II - COLD LEG Rf CIRCULATION.
lJ OF 37 THIS IS TO BE ACCOMPLISHED IN SUCH A ~NNER AS TO PRE-VENT OVERLOADING THE DIESELS.
THE LOADS SHOULD BE APPLIED AT~ 10 SEC, :NTERVALS, DO fiOI RESTART THE EQUIPMENT BY MANUALLY INITIATING SAFETY INJECTION OR CONTAINMENT SPRAY AS THIS ~AV RESULT IN UNDESIR-ABLE VALVE OPERATIONS WHICH MAY RESULT IN EQUIPMENT DAl"~GE.
4.12.2 THE STATUS LIGHTS ON RP-4 ASSOCIATED WITH SEC LOADING SHOULD GO OUT,
. '.,.~-*..
REV. 10
PART JJ 1-4.11 5.0 SUBSEQUENT ACTIONS - PART II - COL] LEG RECIRCULATION ce CAUTION THE CHANGEOVER FROf'I THE SAFETY INJECTION t<n'1ENIS c
PHASE TO COL>> LEG RECIRCULATION MUST IE DONE QUICKLY TO PRECLUDE EMPTYING THE RWST *. IF ANY VALVES FAIL TO RESPOND OR COl'IPLETE THE REQUIRED MOVEMENT, CON-TINUE WITH THE SEQUENCE AND INITIATE ANY CORRECTIVE ACTIONS WHEN THE CHANGEOVER IS COMPLETED.
5.1 VERIFY THAT THE FOLLOWING NORMALLY CLOSED VALVES ARE CLOSED:
11C2l>SJ40 11(21) DISCH VALVE TO Hor LEG 12C22>SJ40 l2C22> D1scH VALVE to Hor l..EG 11C2l>SJ113 SI (HG PUMPS X-OvER VALVE 12C22>SJ113 SI CHG PUMPS X-OvER VALVE 11C2l)SJ45 RECIRC fSOL VALVE TO SI PUMPS llC2l>CS36 FROM 11(21) RHX VALVE 12C22>CS36 FROM 12(22) RHX VALVE 1<2>RH2 RHR COMMON SucrtoN VALVE 1C2>RH1 RHR COMMON SUCTION VALVE 11C2l)SJ44 SIS SUMP VALVE 12C22>SJ44 SIS SUMP VALVE 1(2)RH20 RHX BYPASS VALVE lC2>RH26 Hor LEG ISOLATION VALVE 11C21>RH29 11(21) RHR PUMP BYPASS 12C22>RH29 12(22) RHR PUMP BYPASS 12C22)SJ45 SUCTION FROM RHX (TO CHARGING PUMP) 5.2 OPEN 11C2l>CC16 AND 12C22>CC16 RHR HEAT
[XHCANGER 0uTLET VALVES.
5.3 VERIFY THAT THERE IS AN ADEQUATE WATER LEVEL IN THE (ONTAJNMENT SUMP AS INDICATED BY AN ENERGIZED AVAILABLE NPSH LIGHT ON THE CONTROL CONSOLE.
5.1 THE RH-29's WILL BE CLOSED ONLY IF RHR FLOW JS >1000 GPM PER PUMP, IF CONTAINMENT SPRAY HAS NOT BEEN ACTIVATED THE RHR PUMP NEED NOT BE ALIGNED TO THE SPRAY HEADER.
5.4 WHEN RWST Low LEVEL ALARM ACTUATES 5.4 ANNUNCIATOR D.J.6 ON UNIT 1
<D-35 ON UNIT 2). REcEI PT oF
RWST Low LEVEL ALARM INSURES SUFFICIENT NPSH FOR THE RHR p STOP THE FOLLOWING PUMPS If THEY ARE
RUNNING, No. 11(21> RHR PuHP No. l2C22) RHR PuHP No. 11(21) CS PuHP OR No. 12<22) CS PuHP, IF CONTAINMENT SPRAY ACTUATION HAS OCCURRED, SALEM UNIT l/UNIT 2 12 Of 37 ONE CS PUMP SHOUL.>> CONTINUE.
OPERATE UNTIL THE RWST Low-l LEVEL ALARM JS RECEIVED OR i SPRAY ADDITIVE TANK EMPTIES.
CONTAINMENT PRESSURE IS >23 Rtv. 10
-*-* -* -**------ -- ---*-* ~..-.---'-
c HQlt:
ALIGNING THE RHR PUMPS AS DESCRIBED IN THE FOLLOWING STEPS WILL PROVIDE FLOW TO THE CHARGING AND SAFETY INJECTION PUMPS AND THE CONTAINMENT SPRAY HEADER, r
JF ONE RHR PUMP IS NOT AVAIALBLE TO PROVIDE FLOW THE OTHER PUMP WILL SUPPLY THE CHARGING AND SAFETY INJECTION PUMPS AND THE CONTAINMENT SPRAY HEADER WITH NO ADDITIONAL VALVE OPERATION, HOWEVER, THE COLD l.EG INJECTION FROM THE OPERATING RHR PUMP WILL HAVE TO BE ISOLATED BY CLOSING THE APPROPRIATE SJ-49.
IF LOSS OF OFFSITE POWER HAS OCCURRED CONCURRENTLY WITH LOCA, ~EE APPENDIX l FOR INSTRUCTIONS ON SECURING CoNTAINMENT SPRAY PUMP.
5.5 CLOSE 11C2l)RH4 RHR PUMP SUCTION VALVE, IF
No. 11<21> RHR PuMP 1 s AVAi LABLE.
5.6 CLOSE 12C22)RH4 RHR PUMP SUCTION VALVE, IF No. 12(22) RHR PUMP IS AVAILABLE.
5.7 CLOSE THE BREAKER FOR lC2)SJ69 AND CLOSE THE VALVE, AS SOON AS POSSIBLE.
5.8 REMOVE THE LOCKOUT ON RP4 AND OPEN 11{21)
SJ44 SIS SUMP VALV~, IF No. 11C21> RHR PUMP IS AVAIALBLE, 5.9 REMOVE THE LOCKOUT ON RP4 AND OPEN 12C22)SJ44 SIS SUMP VALVE, IF No. 12C22> RHR PuMP 1s AVAILABLE, 5.10 CLOSE llC2l)RH19 11(21) RHX CROSS DISCH VALVE, r 5.11 CLOSE 12C22)RH19 12C22)RHX (Ross DISCH VALVE.
~
5.12 START No. 11 & 12C21 & 22> RHR PUMP.
5.13 REMOVE THE LOCKOUT ON RP4 AND CLOSE 1(2)
SJ67, 12(22) MINI FLOW.ISOLATION VALVE, AND 1(2)SJ68 11(21) MINI FLOW ISOLATION VALVE.
13 OF 37 1-lf.lf PSIG IT WILL NOT IE POSSIBLE TO STOP EITHER CONTAINMENT SPRAY PUMP.
5.5. 11C2l)RH4 MUST BE CLOSED IN ORD:
5.6 5.7 TO OPEN 11C2l)SJ44.
12C22)RH4 MUST BE CLOSED IN ORD TO OPEN 12C22)SJ44, lC2)C WEST 230\\' VITAL VAL.VE CONTROL CENTER.
THIS CAN BE COMPLETED IN CONJUNCTION WITH THE REMAINING STEPS, 5,13 THERE ARE REDUNDANT SWITCHES ON 1(2)RP4 TO OPERATE 1C2)SJ6 AND 68,
[ITHER THE PUSHSUTTO ON THE CONTROL CONSOLE OR THESE SWITCHES WILL ALLOW FUl REV. 10
~] ~
5.14 OPEN 12!22lSJ45 SUCTION FROK RllX, l2C22) RHR PUMP IS AVAILABLE.
IF No.
5.15 OPEN llC2l)SJ45 REClRC ISOLATION VALVE TO SI PuHP1 JF No. 11C21> RHR PuHP ts AVAILABLE.
5.16 CLOSE THE BREAKER FOR 1(2)SJ30 AND CLOSE THE YALVE1 AS SOON AS POSSIBLE, 5.14 5.15 5.16 Ii OPERATION Of THE VALVES ONCE THE LOCKOUT JS REMOVED.
To OPEN l2C22)SJ451 THE FOLLOWING VALVES MUST BE POSITIONED AS LISTED BELOW:
1C2>RH1 OR lC2>RH2 - CLOSED 1(2)SJ67 OR lC2>SJ68 - CLOSED 12C22>SJ44 - OPEN To OPEN 11<2l>SJ451 THE FOLLOWING VALVES MUST BE POSITIONED AS LISTED BELOW; 1(2)RH1 OR 1C2>RH2 - CLOSED 1C2)SJ67 OR 1(2)SJ68 - CLOSED llC2l)SJ44 - OPEN lC2)C WEST 230V VITAL VALVE CONTROL CENTER.
THIS CAN BE COMPLETED IN CONJUNCTION WITH THi 5.17 OPEN 11C2l>SJ113 SI CHARGE PUMPS X-OvER VALVE.
5.18 OPEN 12C22>SJ113 SI CHARGE PUMPS X-OvER VALVE.
5.19 CLOSE 1C2)SJ1 RWST TO CHARGE PuHP.
5.20 CLOSE 1C2)SJ2 RWST TO CHARGE PUMP, 5.21 5.22 5.23 5.24 VERIFY THAT No. 11C21) SI PuHP 1s OPERATING PROPERLY BY ovsERVING No. 11C21> SI PuHP DISCHARGE PRESSURE INDICATOR AND No. 11(21)
SI PUMP DISCHARGE FLOW INDICATOR.
VERIFY THAT No. 12C22) SI PuMP JS OPERATING PROPERLY BY OBSERVING No. 12C22> SI PuMP DISCHARGE PRESSURE INDICATOR AND No. 12(22)
SI PUMP DISCHARGE FLOW INDICATOR.
VERIFY THAT No. 11(21) CENTRIFUGAL CHARGING PUMP IS OPERATING PROPERLY BY OBSERVING THE CHARGING PUMP JDSCHARGE FLOW INDICATOR AND BORON INJECTION TANK DISCHARGE PRESSURE INDICATOR, VERIFY THAT No. 12(22) CENTRIFUGAL CHARGING PUMP IS OPERATING PROPERLY BY OBSERVING THE CHARGING PUMP DISCHARGE FLOW INDICATOR ANO BORON INJECTION TANK DISCHARGE PRESSURE INDICATOR,
,,.. ~\\ u~uT l/UNIT 2 llf OF 37 REMAINING STEPS, 5.21 PRESSURE:
175 - 1520 PSIG FLOW:
-..400 - 100 GPM 5.22 PRESSURE:
175 - 1520 PSIG FLOW:
"-400 - 100 GPM 5.23 PRESSURE:
175 - 25(X) PSIG FLOW:
"-500 - 100 GPM 5, 24 PRESSURE:
175 - 2500 PS JG 1
,1 FLOW:
'\\,500 - 100 GPH J
Rev. 10 j
-~~~-'.;-,:t;,,..o~~... ~~~*~>;-0"~.~:..;.~.:...... -.... ---......,., *--*
_I
-"*,-.,-.,,,....~.~-----*----*..., ~ :,:" -*.:- 7""'-.'"' """'.""'*-* ----"".--........ *-*--.--=-,.---"*-.*~-- ::**"*..
I I
I I
I I
I I
(
HOIE:
IF CONTAINMENT SPRAY HAS NOT BEEN ACTUATED, DELETE STEPS 5.25 THROUGH 5.28.
5.25 WHEN THE RWST Low-Low LEVEL ALARM ACTUATES STOP THE FOLLOWING PUMP:
No~ 11<21> CONTAINMENT SPRAY PuMP OR No. 12(22) CONTAINMENT SPRAY PuMP, WHICH-EVER PUMP JS STILL RUNNJNG.
5.26 REMOVE THE LOCKOUT AND CLOSE 12(22)SJ49 RHR PUMP D1scH lsoL VALVE, CllC2l>SJ49 JF No. 12<22> RHR PuMP 1s NOT AVAILABLE).
5.27 OPEN THE* FOLLOWING CONTAINMENT SPRAY VALVES:
12C22>CS36 CllC2l)CS36 IF No. 12(22> RHR PUMP IS NOT AVAILABLE),
5.28 (LOSE THE FOLLOWING CONTAINMENT SPRAY VALVES:
llC2l>CS2 12<22>CS2 5.28.l CONTINUE SPRAY OPERATION FOR A MINIMUM PERIOD OF 22.5 HOURS IN ORDER TO ASSURE CONTAINMENT IN-TEGRITY AND REMOVAL OF AIRBORNE FISSION PRODUCTS FROM THE CONTAIN-MENT ATMOSPHERE, 1h.*~T ]/UNIT 2 15 OF 37 1-4.4 S.25 ANNUNCIATOR D-44 BoTH CONTAINMENT SPRAY PuMPS SHOULD BE IDLE AT THJS TIME.
IF CONTAINMENT PRESSURE HAD NOT DECREASED TO BELOW 23.5 PSIG, THE CONTAINMENT SPRAY PUMP CANNOT BE STOPPED FROM THE CONTROL CONSOLE.
To STOP THE PUMPS, IT WILL BE NECESSARY TO TRIP THE BREAKERS LOCALLY ON THE lA AND lCC2A AND 2C) 4KV VIT~
BUSSES BY TURNJNG OFF THE 125 voe CONTROL POWER AND DEPRESSING THE MANUAL TRIP BUTTON INSJDE THE BREAKER CABJHET, 5.28.l THE EMERGENCY (ORE COOLING SYSTEM IS NOW ALIGNED FOR CoLD LEG RECIRCULATION As FOLLOWS:
RHR PuMP No,....llC21> IS SUPPLYING WATER FROM THE CONTAINMENT SUMP DIRECTLY TO RCS LOOP 11C21) AND
}3(21) (OLD LEGS VIA VALVE 11C2l>SJ~9 AND TO THE suer*
OF THE SAFETY INJECTION PUMPS THROUGH VALVE 11(21)
SJ'45.
REV. 10
(
5:29 CLOSELY f"QNITOR THE CONTAINMENT H2 CONCENTRATION ON RP-5.
WHEN THE CONCENTRATION EXCEEDS ~ 21, PL.ACE THE HYDROGEN RECOMBINERS IN SERVICE lAW 01 11-15.3.1, *HvDROGEN RECOM-BINERS - NORMAL OPERATION*.
5.30 VERIFY CORE COOLING IS MAINTAiNED BV OBSERVING THAT THE CORE EXIT THERMOCOUPLES ARE STABLE OR SLOWLY DECREASING,
!!.,,.., 1/UNtT 2 16 OF 37 RHR PUMP No. 12<22> ts SUPPLYING ~ATER FROM THE CONTAINMENT SUMP DIRECTLY TO THE CoNTAINMEHT SPRAY HEADER AND TO THE SUCTION OF THE CHARGING PUMPS THROUGH VALVE 12(22)SJq5, REV* 10
(
PART JJ J
.0 SUBSEQUENT ACTION - PART Ill - HOT LEG RECJRCU~TJON AFTER APPROXIKATELY 22.5 HOURS OF COLD LEG RECIRCULATION, REALIGN THE SAFETY INJECTION SYSTEM FOR Hot LEG RECIRCULATION.
THE SEQUENCE FOR THE CHANGEOVER FROM COLD LEG RECIRCULATION TO Hor LEG RECIRCULATION IS AS FOLLOWS:
fiQif IF A LOSS Of OFFSJTE POWER HAS OCCURRED IN COINDIDENCE WITH THE LOCA AND ONE OF THE DIESEL GENERATORS HAS FAILED TO START, REFER TO SECTION 4.0 OF PART II, Ill OR IV OF APPENDIX 1, AS APPLICABLE.
6.1 CLOSE 12C22)CS36 FROM 12(22) RHX VALVE, 6.2 OPEN 12C22>RH19 RHX CROSS DISCH VALVE, '
6.3 CLOSE THE BREAKER ON 1(2)C [AST 230V VITAL VALVE CONTROL CENTER AND OPEN 1C2)RH26 HOT LEG ISOLATION VALVE.
6.4 REMOVE THE LOCKOUTS ON RP-4 AND CLOSE 11 & 12C21 & 22>SJ49 RHR PUMP DISCH IsoL VALVES.
. 6.5 STOP No. 11C21> SAFETY INJECTION PuMP.
6, 6 CLOSE 11C21> SJ134 11(21) S.l PUMP D 1 SCHARGE TO COLD LEG.
6,7 CLOSE THE BREAKER ON 1(2) A EAST 230V VITAL VALVE CONTROL (ENTER AND OPEN 11C2l)SJ40 DISCHARGE VALVE TO Hor LEG.
6.8 START. No. 11C21> SAFETY INJECTION PUMP.
6.9 VERIFY THAT No. 11<21> SAFETY INJECTION PUMP I~ OPERATING PROPERLY BY OBSERVING No. 11<2:
SI PUMP DISCHARGE PRESSURE AND FLOW INDICATORS (A PRESSURE OF 175 TO 1520 PSlG AND A FLOW OF~ 400 TO 100 GPM SHOULD BE INDICATED),
6.10 STOP No. 12<22> SAFETY INJECTION PuMP.
6.11 CLOSE 12<22>SJ13~ 12C22) SI PUMP DISCH TO COLD LEG
..
6.12 CLOSE THE BREAKER ON lC2)C WEST 230V VITAL VALVE CONTROL CENTER AND OPEN 12C22)SJ!J0 D1scH VALVE To Hor l.EG.
J 6.13 START No. 12C22> SAFETY INJECTION PuHP.
6.1~ VERIFY THAT No. 12<22) SAFETY INJECTION PUMP IS OPERATING PROPERLY BY OBSERVING No.
12(22) SI PUMP DISCHARGE PRESSURE AND FLOW INDICATORS (A PRESSURE OF 175 TO 1520 PSIG AND A FLOW OF~ 400 TO.100 GPH SHOULD BE INDICATED) *
. 6.15 VERIFY CORE COOLING IS MAINTAINED BY OBSERVING THAT THE CORE EXIT THERMOCOUPLES ARE STABLE OR DECREASING.
17 OF 37 REV. 10
c
~
'.{
KDIE THE RESIDUAL HEAT REMOVAL PUMPS AND SAFETY INJECTION PUMPS ARE NOW ALIGNED FOR HOT l.EG RECIRCULATION AS FOLLOWS:
1) No. 11(21) RHR PUMP IS SUPPLYING WATER FROM THE (ONT A 1 NMENT SUMP TO THE suc*T 1 ON HEADER OF THE SAFETY INJECTION PUMPS.
2)
No. 12(22) RHR PuMP 1s SUPPLYING WATER FROM THE CONTAINMENT SUMP TO THE REACTOR COOLANT SYSTEM THROUGH RCS LOOPS 13(23) AND 14(24)
HOT LEGS AND TO THE SUCTION OF THE CENTRI-FUGAL CHARGING PUMPS.
3)_No. 11(21) SAFETY INJECTION PUMP IS SUPPLYING COOLING WATER TO THE REACTOR (OOLANT SYSTEM THROUGH RCS LOOPS 13(23> AND 14(24) Hor LEGS.
~) No. 12(22> SAFETY INJECTION PuMP 1s SUPPLYING COOLING WATER TO THE REACTOR (OOLANT SYSTEM THROUGH RCS LOOPS 11(21) AND 12(22) Hor LEGS.
5) No. 11(21) AND 12(22) (HARGlNG PUMPS ARE SUPPLYING COOLING WATER TO.THE REACTOR (OOLANT SYSTEM THROUGH THE B11 VIA THE (OLD LEGS.
PREPARED BY ____
...::.J:....:*~B:..:a::..:i:..:l::..:e::..yL-___ _
l<Ll<t~
MANAG{R - S{LEM GENERATING STATION REVIEWED BY ____
_.G~.__.c_.a...,n..,n~o
..... ri------
SORC MEETING No. ___
7_4;..,..-_e_o ______ _
REV. 10
~**~ UNIT 1/UNIT 2 18 OF 37
-~~

1..~'..i!...----.
I~
.. -*....,.:.,,,_. __ *...,;;,.....::...,~~-*--.. ~-'"-1-0:.,...:....:..'...!;.,.._;.;:_~..;.:..,,
1' 1 L
~.
~ l
-~ :
(*
-~
1-4.4 APPENDIX 1 IF A LOSS OF OFFSITE POWER HAS OCCURRED IN COINCIDENCE WITH THE lQtA, THE DIESEL GENER*TORS WILL IE. SUPPLYING POWER TD THE VITAL aussEs.
DURING THE RECIRCULATION PHASE, IT IS NECESS*RY TO RUH TllE (oMPONE NT too LI NG PuHPS AND THE HYllROGEN RECMOB 1 NERS
IN ORDER TO Attol'lllA TE TH IS ADD m ONAL LOAD* OTHER EQUIPMENT MUST BE STOPPED BEFORE. THE (OHPONENT (ooLING PUMPS AND HYDROGEN REtOHBINERS ARE STARTED TO PREVENT OVERLOADING THE DIESEL GENERATORS*
UJSCUSSlOH AFTER THE SAFETY INJECTION AND SEC ARE RESET, PROCEED WITH THE APPROPRIATE SECTION*
l - ALL DIESEL GENERATORS OPERAJlliG 1.0 STOP THE FOLLOWING EQUIPMENT:
BOli Do NOT STOP BOTH No. 11(21) AND 12(22) CONTAINMENT SPRAY PuMPS UNTIL THE RWST Low-Low LEVEL ALARM ACTUATES.
WHEN ENTERING COLD \\.EG REClRC., STOP ONLY ONE CONTAINMENT SPRAY PUMP, EITHER 11(21)
OR 12(22),
1.1 EQUIPMENT ON 1A(2Al VITAL Bus (POWERED BY 1A(2Al DIESELIGENERATORl
l. No. 11(21) CONTAINMENT SPRAY PuMP
2.
No. 11(21) AuxtLIARY BUILDING EXHAUST FAN
3)
No. 11(21) SwITCHGEAR RooM SuPPLY FAN
4.
No. 11(21) CHILLER
. 1.2 £gu1PHENT ON.1B(2B) VITAL Bus (POWERED BY 1Bt2Bl DIESEL/GENERATOR)
l. No. 12(22) CoNTAINMENT FAN Coll UNIT
2.
No. 14(24) CONTAINMENT FAN Coll UNIT 1.3 [QUIPHENT ON ltt2ti VITAL Bus (POWERED BY lt<2Cl D1ESELIGENERATORl
l. No. 12(22) CoNTAINMENT SPRAY PuMP
2.
No. 11<21> Aux1LlARY ButLDING SuPPLY fAN
' \\ :> *:
. *. 2.0 : START THE FOLLOWING EQUIPMENT:
tAUJ10ti WHEN ENTERING CoLD LEG REClRC* START ONLY ONE.COMPONENT CooLING PuMP.
ENSURE THE CoMPONENT*cooLtNG PuMP to 1E STARTED IS.ENERGIZED FROM THE SAME VITAL Bus AS NAS THE CONTAINMENT SPRAY PUMP SECURED IN THE ABOVE.STEP*
19 Of 37 REV. 10
2.1 Eou1P11ENT ON 1A<2Al VITAL Bus (P.. ERED BY 1A<2Al D1ESELf6ENERATORl
1. No. 11 COMPONENT CooLtNG PuMP, OR
2. 2 toul Pl1E NT DH lC <2tl v IT AL Bus (POWERED BY lC ( 2Cl DIE SE LIGE NE RA TOR 1
1. No. 13<23) COMPONENT CooLING PuMP llOli IF IRRADIATED FUEL 1S STORED IN THE FUEL HANDLING BulLDlNG, START NO. 11 & 12(21 & 22) FHB EXHAUST FAN.
3.0 OPEN 1J(2l)SW122 AND 12<22)SW133 TO SUPPLY SERVICE WATER TO COMPONENT COOLING.
~.o RETURN To SUBSEQUENT ACTIONS PART 11, sTEP s.1, of THIS INSTRucTioN..
11 - FAILURE Of 1AC2A) DlESEl GENERATOR
},0 STOP THE FOLLOWING EQUIPMENT:
1.1 Eou1PHENT ON lB<2Bl VITAL Bus (POWERED BY 1B<2Bl D1ESELIGENERATORl c
l. No. 11(21) CHARGING PUMP
2. No. l2C22> OR No. 14C24) CONTAINMENT FAN CoIL UNIT
3.
No. 12C22> AuxILIARY BuILDING SUPPLY FAN 1.2 Eou1P11ENT ON 1C<2Cl VITAL Bus (POWERED BY 1C<2Cl DIESEL/Gr*ERATORl
1. No. 12<22> CONTAINMENT SPRAY PUMP wH*N RWST Low-Low LEVEL ALARM ACTUATES.
2.0.START No. 12(22) COMPONENT COOLING PUMP AND OPEN 12(22lSW122 TO PROVIDE SERVICE WATER TO COMPONENT COOLING.
liQ.li
~ -.
1* !.
~.
~~
IF IRRADIATED FUEL IS STORED IN THE FUEL HANDLING ButLDING, START No. 12C22) FHB EXHAUST FAN.
3.0 THE FOLLOWING SHOULD BE THE ALIGNMENT FOR (OLD LEG RECIRCULATION.
3.1 THE FOLLOWING PUMPS SHOULD BE RUNNING:
. -. 1. No. 17(22> RHR PuMP
2.
3.
No. 11<22.) CHARGING PUMP No. 12C22> SAFETY INJECTION PuMP No. 12<22) CoNTAINHENT SPRAY PuMP. UNTIL RWST Low-Low LEVEL ALARM ACTUATES*
REv. 10
!(>.t\\';.M UNIT 1/UN1T 2 20 Of 37

*---_,,A.. -

(*
I-~.~
3.2 CLOSE VALVES l2C22)RH19 AND 12C22)$J49 TO PREVENT FLOW TO THE COLD LEGS AND TO INSURE ADEQUATE FLOW TO No. 12<22> CHARGING PUMP AND No. 12C22) SAFETY INJECTION PuMP SUCTIONS AND TO INSURE FLOW TO THE CONTAINMENT SPRAY HEADER THROUGH 12C22)C$36 WHEN IT IS OPENED.
3.3 THE COLD LEG RECIRCULATION FLOW PATH WOULD BE AS FOLLO~S:
1. No. 12(22) RHR PUMP TAKING SUCTION ON THE CONTAINMENT SUMP AND DISCHARGING TO THE suc110Ns OF No. 12C22> CHARGING PuMP AND No. 12C22> SAFETY INJECTION PuMP THROUGH 12C22>SJ45, 12C22>SJ113, llC21>SJ33, AND l2C22)SJ33.
2.
No. l2C22) CHARGING PUMP DISCHARGE THROUGH THE BORON INJECTION TANK TO ALL FOUR COLD LEGS.
3. No. 12(22) SAFETY INJECTION PUMP DISCHARGE THROUGH 12C22)SJ134 AND 1(2)SJ135 TO ALL FOUR COLD LEGS.
4, 12(22) CONTAINMENT SPRAY PUMP TAKING SUCTION FROM THE RWST AND DISCHARGE TO THE SPRAY HEADER.
4.0 RETURN TO SUBSEQUENT ACTIONS PART II, STEP 5.1, OF THIS INSTRUCTION.
5.0 PROCEED AS FOLLOWS FOR Hor LEG RECIRCULATION:
5.1 CLOSE 12C22)CS36 TO STOP CONTAINMENT SPRAY, 5.2 5.3 5.4 STOP No. 12C22) SAFETY INJECTION PuMP AND No. 12<22) CHARGING PuMP.
(LOSE 12C22)SJ134 TO ISOLATE (OLD LEG RECIRCULATION.
CLOSE THE BREAKER ON 1(2)C WEST 230V VITAL VALVE CONTROL CENTER AND OPEN 12C22)SJ40 TO SUPPLY Hor LEG RECIRCULATION.
5,5 START No. i2c22> SAFETY INJEcT10N PuMP.
5.6 THE Hor LEG RECIRCULATION FLOW PATH WOULD BE AS FOLLOWS:
}, No. 12C22) RHR PUMP TAKING SUCTION ON THE CONTAINMENT SUMP AND DISCHARGING TO THE sucr10Ns OF No. 12<22> CHARGING PUMP AND No. 12C22> SAFETY INJECTION PuMP THROUGH 12C27)SJ~S, 12C22)SJ113, 11C2l>SJ33, AND l?C22)SJ33 *
2. No. 12(22) SAFETY INJECTION PuMP DISCHARGING THROUGH 12C22>SJ40 TO No. 11 & 12
<21 & 22> Hor LEGS.
6.0 RETURN TO SUBSEQUENT ACTIONS PART III, STEP 6.1, OF THIS INSTRUCTION.
.. ' -~ 1/UN IT 2 21 OF 37 REv. 10
.. -.:.-----:::-.-.~-
9 FAILURE OF JBC2B> DIESEL GEflEMIQR
( !.STOP THE FOLLOWING EQUIPMENT:
IOIE Do NOT STOP BOTH llC21) AND 12C22) CONTAINMENT SPRAY PuHPs UNTIL THE RWST Low-Low LEVEL ALARM ACTUATES.
WHEN ENTERING COLD l.EG'RECIRC,1 STOP ONLY No. l2C22)
CONTAINMENT SPRAY PUMP.
l.l EouJPHENT ON lAC2A> VITAL Bus (POWERED BY 1AC2A> DIESEL/GENERATOR)
1. No. llC21> CONTAINMENT SPRAY PUMP WHEN RWST Low-Low LEVEL ALARM ACTUATES.
1.2 [QUIPMENT ON 1CC2C> VITAL Bus (POWERED BY 1CC2C> DIESELIGENERATOR}
1. No. 12<22) CONTAINMENT SPRAY PuMP
2. No. l2C22> SAFETY INJECTION PuMP 1-ll. ll 2.0 START No. l3C23) COMPONENT COOLING PUMP AND OPEN llC21)SW122 TO PROVIDE SERVICE WATER TO COMPONENT COOLING, filIT.E IF IRRADIATED FUEL IS STORED IN THE fUEL HANDLING BUILDING~ START No. 12<22> FHB EXHAUST FAN.
THE FOLLOWING SHOULD BE THE ALIGNMENT FOR COLD LE.G RECIRCULATION:
3.1 THE FOLLOWING PUMPS SHOULD BE RUNNING:
l. No. 11(21> RHR PuHP
2. No. 12C22> CHARGING PUMP
3. No. 11C21> SAFETY.INJECTION PUMP q, No. 11C21> CONTAINMENT SPRAY PuHP1 UNTIL RWST Low-Low LEVEL ALARM ACTUATES.
3.2 CLOSE VALVE 11C2l)RH19 AND 11C2J)SJ~9 TO PREVENT FLOW TO THE (OLD LEGS AND TO INSURE ADEQUATE FLOW TO No. 12C22> CHARGING PuMP AND No. 11C21> SAFETY INJECTION PUMP SUCTIONS AND TO INSURE FLOW TO THE CONTAINMENT SPRAY HEADER THROUGH 11C2l)
CS36 WHEN n JS OPENED.
3.3 THE COLD LEG RECIRCULATION FLOW PATH WOULD BE AS FOLLOWS:
.. 1. No, 11(21) RHR PUMP TAKING SUCTION ON THE CONTAINMENT SUMP AND DISCHARGING TO THE
sucTIONs OF No. 12C22> CHARGING PuMP AND No. 11<21> SAFETY INJECTION PUMP THROUGH 11C21>SJ'451 1JC2l>SJ1131 llC2l)SJ331 AND 12C22)SJ33.
2. No. l2C22> CHARGING PuMP DISCHARGING THROUGH THE BoRON INJECTION TANK To ALL FOUR COLD LEGS.
'. VUr*UT 2
~ *~.. '.
-~-*~~--......-
-*-**.o;~*--'I'-*,***...
22 OF 37 REv. 10
ce 3, No. 11(21) SAFETY INJECTION PUMP DISCHARGING THROUGH 11C2l)SJ134 AND 1(2)SJ135 TO ALL FOUR CoLD UGS, lf. No. llC2J) CONTAINMENT SPRAY PUMP DISCHARGING TO THE SPRAY HEADER, UNTIL RWST Low-Low LEVEL ALARM ACTUATES.
4.0 RETURN TO SUBSEQUENT ACTIONS PART II, STEP 5.1, OF THIS INSTRUCTION.
s.o PROCEED AS FOLLOWS FOR Hor LEG RECIRCULATION:
5.1 CLOSE 11C2l)CS36 TO STOP CONTAINMENT SPRAY, 5.2 STOP No. 11<21> SAFETY INJECTION PuMP.
5.3 CLOSE 11C2l)SJ13~ TO ISOLATE COLD LEG RECIRCULATION,
-~.'f 5.~ CLOSE THE BREAKER ON lC2)A [AST 230V VITAL VALVE CONTROL CENTER AND OPEN 1JC21)$J40 TO SUPPLY Hor LEG RECIRCULATION.
5.5 START No. 11(21> SAFETY INJECTION PuMP.
5.6 THE Hor LEG RECIRCULATION FLow PATH wouLD BE AS FOLLOWS:
l. No. 11(21) RHR PUMP TAKING SUCTION ON THE CONTAINMENT SUMP AND DISCHARGING TO THE SUCTION OF No. 11(2]) SAFETY INJECTION PUMP THROUGH llC2l)SJ45, 11(21)SJ113, llC2l>SJ33, AND 12C22)SJ33.
2. No. llC21> SAFETY INJECTION PUMP DISCHARGING THROUGH 11<21)$J40 TO No. 13 & 14
<23 & 24> Hor LEGS.
3.
No. l2C22> CHARGING PuMP DISCHARGING To ALL Rr.S CoLD LEGS.
6.0 RETURN TO SUBSEQUENT ACTIONS PART III, STEP 6.1, OF THIS INSTRUCTION.
IV - FAILURE OF 1CC2C> DIESEL GtNERATOR 1.0 STOP THE FOLLOWING EQUIPMENT:
l.l EouJPMENT ON 1AC2A> VITAL Bus (POWERED BY 1AC2A> D1EsELIGENERATOR)
1. No. 11<21> CONTAINMENT SPRAY PUMP, WHEN RWST Low-Low LEVEL ALARM ACTUATES.
2.* No. llC21> AuxILJARY fEEDWATER PuMP 1.2 [QUJPMENT ON 1BC2B> VITAL Bus (POWERED BY 1BC2B> D1EsELIGENERAToR)
1. No. 12<22> RHR PUMP START No, 12(22) COMPONENT COO.LING PUMP AND OPEN 11C2l)SW122 TO PROVIDE SERVICE WATER TO COMPONENT CooLING, 23 OF 37 REv. 10 1
~-----*--..
BOIE IF IRRADIATED FUEL IS STORED IN THE FUEL HANDLING Bu1LDING, START No. 11 & 12(21 & 22> FHB EXHAUST fAN.
I-Li*"
c-~
3.0 THE FOLLOWING SHOULD BE THE ALIGNMENT FOR COLD l.EG RECIRCULATION:
3.1 THE FOLLOWING PUMPS SHOULD BE RUNNING:
.. 1. No. 1H21> RHR PuMP
2. No. 11(21) CHARGING PUMP
3. No. llC21) SAFETY INJECTION PuMP q, No. ll<2ll CONTAINMENT SPRAY PUMP, UNTIL RWST Low-Lo* LEVEL ALARM ACTUATES.
3.2 CLOSE VALVES 11<21JRH19 AND lH2llSJ49 TO PREVENT FLOW TO THE CoLD LEGS AND TO INSURE ADEQUATE FLOW To-No. 11(21> CHARGING PuHP AND No. 12C22l SAFETY INJECTION PuMP sucTIONS AND TO INSURE FLOW TO THE CoNTAINHENT SPRAY HEADER THROUGH 11<2llCS36 WHEN IT.rs OPENED.
3.3 THE COLD l.EG RECIRCULATION FLOW PATH WOULD*BE AS FOLLOWS:
1. No. 11121) RHR PuHP TAKING SUCTION ON THE CoNTAINHENT SUMP AND DISCHARGING TO THE sucTIONS oF No. llC2ll CHARGING PuHP AND No. 11C2ll SAFETY INJECTION PuHP THROUGH llC2l)SJL15, llC2l>SJ113, 12C22>SJ113, 11C2l>SJ33 AND 12(22)SJ33.
2. No. 11(21) CHARGING PUHP DISCHARGING THROUGH THE BoRON INJECTION TANK TO ALL FOUR CoLD l.EGS.
3. No. 12(22) SAFETY INJECTION PUHP DISCHARGING THROUGH 12C22JSJ13'1 AND 1(2JSJ135 TO ALL FOUR COLD LEGS.
q, Ho. 11<21l CoNTAINHENT SPRAY PuMP DISCHARGING TO SPRAY HEADER, UNTIL RWST Low-Low LEVEL ALARM ACTUATES.
q,o RETURN To SUBSEQUENT ACTIONS PART 11, sTEP 5.1, oF TH1s 1NsTRuc110N 5.0 PROCEED AS FOLLOWS FOR HoT LEG RfCIRCULATION:
5.1 (LOSE 11C2l)CS36 TO STOP (ONTAINMENT SPRAY.
5.2 STOP No. 11(2]) SAFETY INJECTION PuMP.
5.3 (LOSE 11(2l)SJ13Li TO ISOLATE COLD LEG RECIRCULATION.
5,q CLOSE THE BREAKER ON 1(2)A [AST 230V VITAL VALVE CONTROL CENTER AND OPEN 1J(2JlSJ40 TO SUPPLY HoT LEG RECIRCULATION.
5.5 START No. 11C21> SAFETY INJECTION PuMP.
REV. 10 211 OF 37
. -~
I 1-4.4 5.6 THE HoT LEG REtlRCULATION FLOW PATH wouLD BE AS FOLLOWS:
l. No. 1](2)) RHR PUMP TAKING SUCTION ON THE CONTAINMENT SUMP AND DISCHARGING TO THE SUCTIONS OF No. llC21) CHARGING PUMP AND No. 11C21) SAFETY INJECTION PuHP THROUGH 11C2l>SJ45, 11C2l>SJ113, 12C22>~Jll3, 11C2l)SJ33, AND 12C22>SJ33.
:2. No. 11C21> SAFETY INJECTION PUMP DISCHARGING THROUGH 11C2l>SJ40 TO No. 13 & 14 (23 & 24) HoT LEGS.
3. No. llC21) CHARGING PuMP DISCHARGING to ALL RCS CoLD LEGS.
6.0 TRANSFER THE SECURITY SYSTEM TO THE EMERGENCY POWER SUPPLY ON lA 230V VITAL Bus.
7.0 RETURN TO SUBSEQUENT ACTIONS PART Ill, STEP 6.1, OF THIS INSTRUCTION.
25 OF 37 REv. 10
\\
t l,.
l r l l 1**
}
J
s APPENDIX 2 DETERMINATION Of ACTUAL PRESSURIZER AND STEAM GENERATOR LEVELS lo DETERMINE ACTUAL PRESSURIZER OR STEAM GENERATOR LEVELS, UTILIZE THE ATTACHED TABLES AS DE-SCJBED IELON:
1.
2.
3.
4.
TABLES 1-5 ARE TO IE USED FDR THE STEAM GENERATORS TABLES 6-11 ARE TO BE USED FOR THE PRESSURIZER OBTAIN THE AVERAGE (ONTAINMENT TEMPERATURE FROM COMPUTER POINT Ul30LI REFER TD THE APPROPRIATE TABLE FOR THE PRESSURE IN THE STEAM GENERATOR AND PRESSURIZER.
LOCATE THE DESIRED ACTUAL LEVEL UNDER THE COLUMN FOR THE AVERAGE CONTAINMENT TEMPERATURE.
fROM THE COLUMN HEADED *INDICATED LEVEL (%)* DETERMINE THE INDICATED LEVEL WHICH SHOUIJ) BE MAINTAINED TO ENSURE THE ACTUAL LEVEL IS MAINTAINED AT OR ABOVE THE DESIRED VALUE.
BOii IF THE COMPUTER IS UNAVAILABLE, MAINTAIN INDICATED LEVEL ~35% ABOVE THE DESIRED ACTUAL LEVEL*
""'. 1/UNIT 2 26 OF 37 REv. 10
.l i
.J l
l l
.~
l I
-1
\\
> 1
I l
L~YELt1.) Al 200~111 ACTUAL S/G LEYELt?.> AT TEMPEP.ATURE 275
'350
!"... !',.. ~ *.
125 150 17~
200 225 250 o.oo
-11.8*
-18.74
-11.6&
-20.73
-21.es
-23.14
-24.50
-25*
-z1.51
-z*.1' s.oo
-13,6l
-14.46
-15.41
-16.46
-11.61
-18.87
-20.z2
-21.68
-23.23
-z4.87 lo.oo
-*.34
-10.1*
-11.13
-12.1&
-13.33
-14.5?
-15.,5
-17.41
-18*
-zo.s*
1s.oo
-s.01
-s,,1
-6.&6
-1,,1
_,,06
-10,32
-11.6&*
-13.13
-14.68
-1*.JZ INDICATED LEYELtX>
N....
0 w
zo.oo
-0.1'
-1,64
-2.s8
-3.63
-4*7'
-6.04
-1,40
-8,86
-10.41
-1z.~
zs.oo 3,48 2.64 l,6, o.64
-o.sl
-1.11
-J.13
-4.58
-6.14
-1.11
... CJ2 J.7b 2.5ll
-0.31 J.'16 a.21t
-1*.~*
2.41'
-3.SO 0.11 5.09
33 30.00 35.00 40.00 lt5.00 so.oo 55.00 60.00 65.00 10.00 75.00 90.00 85.00 7.76 12.03 u.. 31 20.sa 21,.85 29.13 33.40 37.68 41.'15 54.77 5'1.05 b3.32 l'l.73 24.0l 2&.28 32.56 36.83 41.10 45.38 se.20 62.48 1&.79 23.0b 27.34 31.bl 35.88 48.71 52.'18 57.26 bt.53
.,.1 q 13.ltb 11.11a 22.01 26.2'l 30.5&
34.84 J'1.11 43.38 47.6&
8.04 12.Jl tb.51J 20.86 25.13 2'1. 41.
JJ.68 J7.'1b 42.23 ltb.51 5q.JJ
b. 78 11.05 15.JJ 1'1.bO 23.88 28.15 32.43 Jb.70 40.CJ7 45025 58.07 62:.J~
IP.70 1a.21t 22.52 26.7CJ Jl.07 35.34 J'J.61 43.8'1 48.16 52.lt4 56.71 Ml. C?C?
12.51 lb.7' 21.06 25.33 2.,.61 JJ.88 J8.lb 42.43 46.71
50. CJ8 55.25
'!5'?. '!53 lo.**
1S.2ft l'P.51 ZJ.78 21.0*
3!.33 36.61 40.88 45.lb 13.60 17.17 22.19 26.ft2 J0.70 34.'P7 3*.25 lt3.52 1t1.1*
52.07 56.Jlt
I
I I
, I
I~
o
I
. t
I.
w
,*d******~-.....
JND!CATED STEA" CENERATOP. LEVELCX) VS. ACTUAL LEYEL<Y., AT 400~111
!NOtCATEO LEYELOO
\\:
~
12, 1~0 ACTUAL SIC LEVEL(~) AT TE"PERATURE 175 200 225 250
~1~~*-.i,,1 275 JOO 32S 3SO o.oo
-1a.s2
-19.41
-20.1,2
-21.54
-22.11
-21,.10
-25.SJ
-21.06
-2s.10
-3D.4*
s.oo
13.98
-14.88
-15.89
-17.01
-18.23
~19.56
-20.99
-22.53
-24.17
-25.93 10.00 15.00 20.00 25.00
-9.4S * -10.J~
-11.35.**12.47
-13.70
-15.0J
-16.46
-17.99
-19.64
-21.40 30.00 3S.OO 40.00 ltS.00 SO.OD ss.oo
-4.91
-0.38
8.69 13.22 17.76 22.29 26.83 31.36 60.00 3S.8*
6S.OO
. 40~43 70.00 44*
75.00 49.,0 ao.oo
~4.0J as.oo
~8.57 90.00 63.10
"~-00 67.64
-s.a1
-1.21 3.26 7.79 12.JJ 16.86 21.40 25.93 30.47 35.00 39~54 44.07 48.60 SJ.14 57.67 62.21 66.74
-6.82
-2.29 2.25 6.78 11.32 20.39 24.92 29.46 33.99 38.52 43.06 47.59 52.13 56.6b 61~20 65.73
-7.94
-J.41 1.13
~.66 10.20 14.73 19.27 23.80 28.JJ J2.87 37.40 41.94 46.47 51.01 55.54 60.08 64.61
-9.11
-10.49
-11.,2*
-13.46
-1s.10 -1*.1*
-4.63
-5.96
-7.39
-8.,2
-10.57
-lZ.33
-0.10 4.44 8.f17 13.~51 18.04 22.ss 27.11 31.64 36.18 lt0.71 45.25 49.78 54.32 58.85 63.38
-1.42 J.11 7.65 12.18 16.71 21.2s 25.7&
30.32 34.85 39.39 43.'12 48.45 52.'19 S7.S2 62.06
-2.85 1.68 6.22 10.7S 15.28 19.12 24.35 37.96 42.49 47.03 51.56 S6.09 60.63 b~.16
-4.39 O.lS 4.68
.21 13.75 11.28 22.82
-6.03
-1.~o 3.04 7.57 12.10 16.64 21.17 27.JS 2S.71 Jl.89 30.21t 36.42 34.71 40.*6 3*.31 4S.4' 43.85 50.02 48.38
~4.S6 sz.*:n 59.09 57.45 63.63
-1.1*
-3.26 1.21 s.11 10.34 14.81 1*.41 23.9S 21.41 33.02 37.55 42.09 46.62 60.22 N
I I
I I
~, :
fl~OtCATED
.EYELOO INDICATED ITEAft CENERATOP. LEVELCX) VS, ACTUAL LEVELCY.> AT 600~111 ACTUAL SIG LEVEL(?.) AT TEMPEP.ATUR~
125 150 17~
200 225 250 275 300 329 3SO o.oo
-1,.21
-20.19
-21.25
-22.44
-23.73
-25.12
-26.61
-28.21
-2*.*2
-31.77 s.oo
-14.50
15.42
-16.49
17.67
-18.96
~20.Jb
-21.85
-2~.44
-25.16
-27.00 io.oo
-9.74
-10.66
-11.12
-12.,1
-14.20
-15.59
-11.os
-18.68
-20.39
-2z.24 lS.00
-4.97
-5.89
-6.96
-8.14
-9.43
-10.82
~12.Ji"* -13.91
15.63
17.47 20.00 25.00 30.00 35.00 40.00 45.00 so.oo ss.oo 60.00 65.00 70.00 7'5.00 80.00 e~.oo
,o.oo
"'r. nn
-0.21 4.56 23.62 28.3, 33.lS 37.,Z 42.61 47e4S 52.22 bl.75
66. ~Sl 71.28
-1.13 3.64 a.40 13.17 17.,4 22.70 27.47 32.23 37.00 41.76 46.53
'51.30 56.06 60.83 65.541 70.36
-2.19*'
2.57 7.34 12.11 16.87 21.64 26.40 31.17 35.93 40.70 45.lt7 50.23 55.00 59.76 64.53 69.2?
-3.38 1.39 6.16 10.92 20.45 25.22 2*.98 34.ns 39.52 44.28
"'. 05 53.81 58.58 63.Jlt 68.11
-4.67 0.10 4.86.
'1.63 14.40 19.16 23.93 28.6*
33.46 38.22 42.99 47.75 52.52 57.29 62.05 66.82
-6.06
-1.29 3.47 e.24 13.00 17.77 22.Sl 27.30 32.07 J6.&J 41.60 46.Jb 51.13 55.89 60.66 65.43
70. 1?
-2.78 1.98 11.Sl 16.21 21.04
. 30.58 35.34 40.11 44.87 4,.,,..
54.40 5'1.17 63 I <<13 68.70
-4.38 0.38 5.15 14.68 1,,,.,
24.21 28.98 33.74 38.51 43.28 S7.S7 62.34 67.10
-10.86
-6.10
-1.33 3.44 1.20 12.'il7 17.73 22.so 27.26 32.03 36.11 4l,S6
,6.33 51.09 55.86 60.62 6~.3~
-1z.11
-7.94 ll.12 15.89 20.6S 25.42 30.11 3'*'s.
3*.72 S4.01 58.78 I
ii w/
I I
I
I ;
INDICATED STEA" CENERAT~R LlVELC7.) vs. ACTUAL llVEL(~) Al eoo~***
atCATEO
',:YEL C 1. >
ACTUAL SIG LEVELC?.> AT TEMPERATURE 175 200 225 250 o.oo
-20.20 * -21.11
-22.1s
-2J.39 s.oo
-15.20
-16.11
-17.18
-18.39 10.00
-10.20
-11.11
-12.18
-13.39 1s.oo
-s.20
-6.11
-1.18
-8.39 zo.oo
-0.20
-1.11
-2.1a
-J.J9 2s.oo 4.80 J.89 2.82 1.61 30.00
,.80 8.8, 7.82 6.61 3S.OO 14.80 13.8' 12.82 11.61 40.00 19.80 18.89 17.82 16.61 4S.OO 24.80 23.89 22.82 21.61 so.oo 29.80 28.89 27.82 26.61 ss.oo 34.80 JJ.89 32.82 31.61 60.00 39.80 38.89 37.82 36.61 65.00 44.80 43.89 42.82 41.61 70.00 4,.80 48.89 47.BZ 46.61 75.00 54.80
~J.89 52.82 51.61 80.00 59.80 58.8~
57.82 56.61 es.oo 64.eo 63.89 62.e2 61.61 90.00 69.80 68.89 67.92 66.61 74.80 73.89 72.82 71.61
-24.72
-1'1.72
-14.72
-9.72
-4.72 0.28
~.28 10.28 15.28 20.28 25.28 J0.28 35.28 40.28 45.28
~0.28 55.28 60.28 65.ZB 70.28
-26.16
-21.16
-16.16
-11.16
-6.16
-1.16 J.8.ft 8.84 13.84 18.84 23.84 28.84 33.81t 38.84 43.Blt 48.84 53.84 SB.84 63.Blt 68.84
~-..
27S 300 32!
-21.11
-29.36
-22.71
-24.36
-17.71
-19.36
-12.11
-14.36
-7.71
-9.36
-2 *. 71
-4.36 2.2'1 0.64 7.2*
5.64 12.29 10.64 17.29 15.64 22.29 20.64 27.29 25.64 32.29 30.64
37. 2 1i
35. 61t 42.29 40.64 47.2.,
45.64
~2.29 50.64 57.29 55.64 62.29 60.64 67.29 M5.64 7'/_. 29 70.6't
-31.12
-26.12
-21.12
-16.12
-11.12
-6.lZ
-1.12 3.88 8.88 13.&8 18.88 23.88 28.88 33.88 38.11 43.88 48.88 53.88
!58.88 63.88 68.88 3SO
-33.00
-21.00
-23.00
-11.00
-13.00
-1.00
-3.00 2.00 7.00 12.00 17.00 22.00 27.00 32.DO 37.00 42.00 47.00
. s2.oo S7.00 62.00 67.00 r
! !I t4 NE
~
I I
ti fl ii I
~
I i'
i/
ii
J?CATED
'fl.ELOO
~**
INDICATED STEA" CENE~AlO~ LEVELCX> VS. ACT L LEVELCY.t AT i000~1l1
~.....
. *II\\ lik *..
ACTUAL SIC LEYELC~> AT TEMPERATURE 125 150 11~
zoo z2s z~o 275 JOO
... 32s o.oo
-21.23
-22.20
-23.J~
-24.63
-26.04
-27.56
-29.18
-J0.92
-32.7*
-34.7*
s.oo
-16.00
-16.97
-18.12
-19.40 10.00
-10.77
-11.1~
-12.88
-14.17 ss.oo
-5.54
-6.51
-7.65
-&.94 20.00
-0.31
-1.28
-2.42
-J.71 25.00
. 4.92 3.95 2.81 1.52 30.00 10.lS 9.18 8.0~
6.75 JS.OD 15.38 14.41 13.27 11.99 40.00 20.62 19.64 18.50 17.22 45.00 25.85 24.87 23.73 22.45 so.oo 31.08 30.11 28.96 27.68 s~.oo 36.31 35.34 34.19 32.91 60.00 41.54
.40.57 37.43 38.14 65.00 46.77 45.80 44.66 43.37 10.00 s2.oo s1.0J 4?.e?
4&.6o 75.00 57.23 56.26 SS.17 53.83 ao.oo 62.46 61.49 60.Js s1.06 85.00 67.69 66.72 6~.~8 64.JO 90.00 72.93 71.?5 70.81 69.53
,5.00 78.lb 77.18 76.04 74.76
-20.&1
-22.33
-1~.58
-17.09
-10.35
-11.&6
-5.11
-6.63 0.12
-1.40 5.35 J.8J 10.58 9.06 15.81
.. 14.2'1 21.04
19. ~2 26.27 24.75 J1.!j0 29,98 J6.7J 35.22 41.96 40.45 1,1.20 43.68 52.43 50.91 57.66 S6.14 62.89 61.37 68.12 66.60 73.35 71.83
.,.-y
-18.72
-20.46
-113.4;** -15.23
-1.26
-10.00
-3.03
-4.77 2.20 0.46 7.43 5.69 12-.66 10.92 17.90 16.16 23.13 21.3, 28.36 26.62 JJ *
Jl. 85 38.82 J7.08 44.0S lt2.Jl 49.2!
47.'54
~It
S l S'Z. *11 59,74 5B.OO 64.'17 63.23 70.21 68.47 7~L44 73.70
-11.09 -1*.o*
-11.86
-13.16
-6.63
-1.63
-1.40
-3.40 3.83 1.13 9.06 7.06 14.29 12.29 19.52 17.52 24.75 22.76 29,,,
27.99 JS.22 33.22 40.45 31.ltS 45.68 43.68 S0.91 48.91 56.14 54.14 61.J7 s,.J7 66.60 64.60 71.RJ 64'.e3 I I I £ I f I
ro I ~
I
,~
i:
"'. t
. t
~
I' f:
~ I
~ I
~.
f*
F f
~ l f I i i
'i I
.J
.1MOICATED Us.EYEL C X) w N
0....
w o.oo s.oo 10.00 15.00 20.00 zs.oo 30.00 3S.OO 40.00 4S.OO so.oo
,5.00 60.00 65.00 70.00 75.00 ao.oo es.oo 90.00 q'5.00 INDICATED PRE99URJZER LEVELCY.) VS, LEYEL<X> AT 500~111
-~~~~~
125 190 ACTUAL PRZ LEYELCX> AT CONTAI"ENT TE"PERATURECDet F, 4'dllai"'
175 200 225 250 275 JOO
-10.2s
-11.16
-12.20
-1J.Js
-14.61
-1s.*1
-17.43
-1,.00
-20.**
-1.1s
-a.06
-9.10
-10.25
-11~s1
-12.s1
-14.33
-1s.*O
-11.s*
-4.0S
-4.96
-6.0D
7.15
-8.41
-9.77
-11.23
-12.81
-14.SO
-0.95
-1.86
-2.90
-4.06
-5.Jl
-6.67
-1.14'**
-*.71
-11.40 2.15 1.24 tt.20.
-0.96
-2.21
-3.57
-S.04
-6.61
-8.JO s.2s 4.34 3.Jo 2.14 o.88
-o.4s
-1*'"
-J.s1
-s.20 8,35 7.43 6.40 S.24 J.98 2.62 1.16
-0.41
-2.10 11,44 10.53 Y.49 8.34 7.08 5.72 4.26 2.6*
I.CO 14.Slt 17.64 20.74 23.84 26.*4 30.04 33.14 36.24 39.33 42.43 48.63 13.63 16.73 19.83 22.93 26.03 29.13 32.23 35.33 38.42 41.52 44.62 47.72
12.:S9 15.69 18.7'1 21.89 24.'19 26.09 Jl.1'1 34.29 37.JI!
40.48 11.ltft
14. 54 17.64 20.74 23.83 26.93 JO.OJ 33.13 36.23 39.33 42.43 45.53 10.1a 13.28 16.JS 19.48 22.sa 25.68 28.77 31.87 Jlt.'17 38.07 41.17 44.27 8.82 11.92 15.02 18.12 21.22 24.32 27.41 30.51 33.61 Jb.71 41..0l 7.36 10.46 13.56 16.66 l,.7S 22.85 2S.9S 29.05
32. lS 35.25
~.79 8.89 ll.'19 15.08 18.18 21.28 21t.J8 27.48 J0.58 3*3.68 36.78 3,,88
42. '11 4.10 7.20 10.JO ll.3' 16 *.\\9 l9uS9 2z.**
25.7*
28.89 31.99 no.
-22.s1
-1*.42
-16.3!
-12.22
-1a.1z
-0.12 2.27 5.37 1.47 11.57 1~.67 17.77 Z0.87 23.*7 27.01 30.16 33.26 36.36 J9,lt6
INDICATED PRESSURIZER LEVEL(~) YS. ACTU 1000.-1..
INO!CATEO
, LEVELCX>
125 lSO ACTUAL PRZ LEVELC?.J AT CONTAI"ENT TE~PERATURE(D*f Ft.,...,.,_.
175 200 225 250 275 300 325 3~0
.iw jw
' io "j
w f
!~
I l l I
o.oo
-11.4a
-12.46
-1J.61
-14.90
-16.32
-11.as
-19.48
-21.23
-23.11
-2s.1z s.oo
-a.oo
-a.9&
-10.13
-11.42
-12.aJ
-14.Jb
-1b.oo
-11.15
-19.62
-21.63 io.oo
-4.~l
-~*"'
. -6.64
7.9J
-9.35
-10.ee
-12.~u
-14.26
-16.14
-11.1s lS.00
-1.~3
-2.01
-J.16
-4.4S
-5.&7
-7.39
-9.oi*
-10.78
-12.65
-14.67 20.00 2.45 l.48 o.JJ._
-0.97
-2.Je
-J.91
-s.~s
-1.30
-9.17
~11.11 25.00.
5.94 4.96 J.81 2.52 1.10
-0.42
-2.06
-J.81
-5.69
-1.10 30.00 35.00 40.00 45.00 so.oo ss.oo 60.00 6~.oo 70.00 75.00 80.00 85.00 90.00
'15.00 9.~2 12.91 16.39 19.87 23.36 26.84 30.33 33.81 37.30 40.78 44.26 47.75
~1.'ZJ 54.72 8.44 11.93 15.41 1&.90 22.38 25.87 29.35 32.83 36.32 39.&0 46.77 so.2~
~J.74 7.30 10.78 14.26 17.75 21.23 24.12 28.20 Jl.68 35.17 38.65 42.14 45.62 4?.10 6.00 9.49 12.97 16.46 t*.94 23.42 26.91 30.39 33.88 37.36 40.84 44.33 47.81 51.30 4.59 8.07 11.S6 15.04 18.S2 22.01 25.49 2&.98 32,46 35.94 3'1.43 42.91 46.40 49.88

~-

J.06 6.54 10.03 13.51 11.00 20.48 23.96 27.45 J0.9J 34.42 37.90 22.33 25.81 29.JO 32.78 36.26 39.75 43.23 46.72
~0.20
-0.JJ 3.16 6.64 10.12 13.61 17.09 20.sa 24.06 27.SS 31.0J 34.Sl 38.00 41.48 44.'17
-%.20 1.28 e.2s 11.73 15.22 18.70
_,.21
-0.73 Z.75 6.24
.7Z 13.21 16.6*
22.1*
20.11 ZS.67 23.66 2*.16 27.14 32.64 30.63 36.12 34.11 J~.61
. 37.60 41.08 44.56 I
I
.... ~..... ~ ** ",....,: *. :... :.,, *....:.~.j~:..,:.:.;...;,f~:~-~~:,!~;:.:.~.;..;.~;::;.*~:~:-~*7~~":~:..;;,.::~~,.,_,,.,_:.,.,~......... ~... ~*~ut-1"... '..~'f"'"" _.....,_,"'"';...,....,.,.... __..Tl..........,.,..... _~
...,..... ~.**.*..-...... ~;:,--..*-.** *****-**** *-..,....... ~. -~.-,.... - _,.
re
. ~-.. *---
nc1u-YtLl7.I ce l
INDICATED rF.ESSUIUZEt. LlV£L no vs.
A1 1500Plll
~..........,,,..
:1mICATEO ACTUt\\L PRZ t.EV!:L(~l nr CO.NTt\\lt1ENT fEt1PEP.AfURE(0~~ F,..,*c*.:*
.t:YE\\.(Y.)
125 150 1 *1~
zoo 22~
£'50 i1s
mo 325 350 o.oo
-1J.bl
-11t.73
-16.03
17. 4A
-1?.07
20. 1?
22. 64
'Zlt.61
-26.71
-28.'fS 5.00
-fl.67
-10.80
-12.0'I
-1J.S4
-1~.lJ
-16.B'!'.J
-18.70
-20.68
-22.78
-25.0l 10.00
-5.71t
6.Sb
B.16
?. 61
-11. 20
-12. *n
-14.77
-1b.71t
-18.8..
-21.os 15.00
1.80
-2.'n
.,,
22
-5. 67
-.,
26
- 8. '18
-10.eJ
-12.eo
-11t.*H
-17.llt 20.00 2.11t 1.01
-*0. 28..
-1. 73
-J.32
-S.05
-6.8'1
-8.87
-10.'17
-13.21 25.00 6.07
". '15 3.b'5 2.20 O.b1
-1.11
-2.'lb
-1t.'1J
-7.0lt
-'J.27 30.00 10.01 8.88
1. 5*1 6.1'9 4.SS Z.83 o.CJ8
-1.00
-J.10
-S.33.
w 35.00 lJ.'14 12.e2 11.52 10.01 8.48
b. 76
". q 1 2.'14 O. Bit
-l.40 t
0 lt0.00 11.ee 16.75 lS.46 1... 01 lZ.42 10.1u 8.es 6.87 lt.11 2.51t Ht w
lt5.00 21.e2 20.6'1 19.3'1 17,qt, 16.36 14.bJ 12.7'1 10.e1 8.71 6.lt7 50.00 25.75 24.63 2J.J3 21.88 20.29 ULS7 16.72
14. 75 12.61't 10.41 55.00 29.6'1 28.'56 21.21 25.82 24.23 22.s1 20.66 18.68 16.58 llt.35 60.00 JJ.62 32.so
H. ZfJ 2'1.75
28. l(J Z6.44 2... 5'1 ZZ.62 20.'52 18.28 65.00 37.56 36.43
~'~. 14 JJ.6'1 32.10 JO.JB 28.53 2b.'55 21t. a,5 22.22 10.00 i.1.1t'I lt0.37 J?.07 37.62 Jb.03 Jlt.Jl 32.46 JO* c,q 21.~I' 26.lS 7'5.00
45. lt3
.,,.
30 1, 'J. lH
41. '56 3'1. en 38.25 Jb.40 J4. ltJ 32.32 JO.O'J eo.oo 4CJ.J7 4B.21t 46.?4 45.SO 4J.91
'z. lB 40.34 Jft.36 Jb.26 Jlt.02 e!i.OO
'SJ.JO
~2.1e
'50. r.l\\
4'1.43 1,7.~lt t,b. 12
'*". 2"/
t.Z.30
40. lq 37,q6 qo.oo S7.Zlt
'56.11
'.) 't *
~* z SJ.J7
~1. 78 SO. Uh lt8. 21 46~~3 44.13 "1. qo q~.oo bl.17 60.05
~8. 7~
'57.JO
~~. *11
'53.'1'1 52.14 so.11 48.07 45.83 I. r:._
5,. *IJ 5(,,08
~.... 10
'52.00
.,,,_ 77
INDICATED PRESSURIZER LEVEL nn vs. nCTUAL LEVELOO AT 2000tttll INDICATED ACTUAL PRZ LEVELC?.) AT CONTAI"ENT TEMPERATUP.E<O*t F, LEYELOO 125 150 175 200 225 250 275 JOO 325 350 o.oo
-17.39
.-18.69
-20.19
-21.85
-2J. 67
-25.64
-27.75
-J0.00
-32.40
-34.,7 5.00
-12.!3
-14.1J
-15.63
-17.2'9
-19.11
-21.08
-2J.19
-25.44
-27.84
-30.41 10.00
-8.27
-'l.~7
-11.07
-12.73
-14.SS
-lb.52
-18.6J
-20.88
-23.28
-25.85 15.00
-3.71
-5.01
-6.51
-8.17
-'9.'19
-11. 96
-14.07
-16.32
-18.72
-21.2*
I 20.00 0.85
-0.4!5
-1.75
-J.61
-5.43
-7.40
'1.~1
-11.76
-14.16
-16.73 I
25.00 5.41 4.11 2.61 0.'15
-0.87
-2.84
-4.'95
-7.20
_,.bO
-12.17 I
30.00 9.'97 8.67 7.17 5.51 J.69
l. 72
-0.3'9
-2.61t
-S.04
-7.61 1~
35.00 11t.53 13.23 11.73 10.07 8.25 6.28 4.17 1.'92
-0.48
-3.05
~
1 I:
40.00 19.09 17.79 16.2'1 14.63 12.81 10.84 8.73 6.48 4.08 1.Sl f;
23.65 6.07 45.00 22.35 20.85 1'9.1'1 17.37 15.40 1J.2'9 11.04 8.64
~o.oo 28.21 26.91 25.41 23.75 21.93 19.96 17.85 lS.60 13.20 10.63 S5.00 32.77
31. 47 2'9.97 28.Jl 26.4'1 24.52 22.41 20.16 17e76 15.1*
I I 60.00 37.JJ 36.0J JltoSJ 32.87 Jl.05 2?.oe Z6.97 24.72 22.32 1,.75 65.00 41.8' 40.59 39.09 37.43 35.61 JJ.61t 31.'53 29.28 26.88 24.Jl I
~
70.00
46. 4:5 45.15 43.65 41.9'1 40.17 38.20 Jb.09 JJ.84 31.44 28.87
. I I
I '
75.00 51.01 49.71
'*B.21 46.'5'5 44.73 42.76 40.6~
38.40 36.00 JJ.43 l
80.00
'55.S7 54.27 52.77
~1.11
't9. 2?
47.JZ 45.21 lt2.96 40.56
37. '1'1 s~.oo 60.13 58.83
'57.33
'55.67
'53.CS 51.88 49.77 47.52 4S.12 42.55 90.00 64.b9 6J.J9
61. 8?
60.2J 58.41 Slt.ftlt 54.JJ 52.08
""
68 47.11 q!_..00 b9.25 b7.'1'5 66.4~
64.79 62.97 61.00 58.89 56.64
~lt.24
'51.67
-~
bJ.45 bt.20
~tL80
~b.23
~~?**~ !*:~ f ~t)
~ f \\'t t ~ ~:)
~*. t.tt.* '
!U.mJ l '!*. Ul*
20.rm
JU.O*J J~.Ott
't0.00 5f]. *J*.l
/*!
l)fJ
1 ~*.0U
r. ~* I ('f'
-~. '*'
., "-\\*1
.. - ~*
Zl.3J
'.. !*. l~i' '
\\.:-
~. ' f'..
J
1 'J
t J '* * -*
!1 '!.., ~ *~
I
"\\"*
flt: P'.
'~*~~! ' *:*~* I,., '
., I
'.'M !., I~*.~.*'
!'."~*
~*~*
~.
.-~*-*
..! ~ ': *.~ ~-.
t *
=... * ~
J ** ~.. *~ :
I I
o I
I; I : *
'* i
'~ *I: * *,
I *.
I*
~ ** J *
.. :1
J '~
'.~. !. (.
,..... l."/
~. *.. *. t.. 1~
'~ '*.... ~~
.'.~I * /~ *:
~I.,I *. ! o ('I I,'* * **':
l* f,I * !,I'
~**.* * $ *:-
.... II!.*
'.' /.
. /.
o I
I*
I
. / ** Pt
~'l.!*l
.:'*.' I {1 f',
'** :~.
~ '.'
/, "'.
~. *,:
.~.:. *'*.:
~I *.1' I
( o f,'1 f '"
I 11
.,... *J I
f I
* *I:. I. :
. ~- '.I * '~ :'
f:.* ( {.
'J * *y~
t' I, f.,
~... '* '*
I
':* *.: *. i *~~
I ** t I
~. 'i
.',*t. t I~ '.
~' I ~I'.'
/fl 1 *;;-1~"::-'~'lll:~*.:r*~*~'! '°
.*/~*'
/*l'l'
.n.. n,
..,,~.~:s
'.' /.,
,* t,
. '.. '.' * '..I
.*,:t *./';.
. u.. *.1~:.
u. I 7::
r.;
I I I~ *,* I
~,.,
I~*~"./'t t. t
~
.t *~
\\* *...
~J1.*.n
.. ::: :' * *H
. :.-*:::*. !.H
t. I
tJ !
12.01
. /
U'J
- i'. l'l' t.* Ott
~J.I.)!
JJ. tn
'* ~:
t* ~**
... 'l'.
.) ~... -:..
'*'1. !6
.. m.1:,
-lf.J.l:S
-1~. lS
.,_0.;!5
'*
t£
.,
3-!J 1 ~. tl*
1 '.'. \\J l
.~*-'.a:~
l't' *.I * ~' \\!
\\
I nmt*=~r*

*~
Lt.\\'t.L ~~O
1 * *.m
V~. 1
!~I: f II !\\I I * * ;> *1 1 'I
~--* ~--*...... ~.....,......................... -
' '.~ '
{* I
~**.**.*\\*r ~ It
'** *..:" *.: ** f. :~ ~
t\\l' 1:0Nrr~'-'*~*:r1~*
r*.:M.P*:~n n~:": ~ *~~'! F'
1~*
!l'l'
JfJ * '~ ~:
.* ~ '**! * **~ '!
... I ~* **'
!-'I, 1 Z
f.11, 1.'i.'.
~ ** tm
2~.* tm
U.:rn iu.uu
-1J./j l~.OQ
r.. u.
Z*.l.*.m JO.u*l 3.1:5 1:-0. J 1*'. *n so. \\)lj 6U.o*J tr~*. m.*
I: fi ' ~!
,,~. t)tj
~J.!!
3\\1. O*J
ti I*
* '* I
JI), l)IJ
"~*. '*"*'
,. ~ * ~" 1 1.,.., *,*
41]. 32
~.-,.t.t"*
~:.!
S'~
I I,
, ~ '*
~~*~
'* ~- * ~. f',
'* '-'. '-~~*
',.... *.~
". ;
I *.:
.'* *.:** * * :.f J. "' !
..... ~. : '*
'.*~*.,. (
'* t I,*~
~/.'
1'.!.* M*
. :t. '.'*l*
t*. t.r*.
1....
l'3. ~
': '.! * ~*'-'
~.* :~.* f'* '*
I., * '*
~- 1. I,, I.
o f*
I
.~*
' f, *. '*:.
,, I
I I
,1 '
I *-i
-.** t -
.. *.n
(,. (. ~*
~-*
'.~.'
~ *'. !'* ':'
'.*' *~,. '*
(I.
t.11 T'
.. *
I o ;
I, o ~
I**
. *~,*
. ' r I. ;
,. ~
,, :~.
'~ *'
/I: 'l*':'
. "'
t*i'.
I
* ~L
.: <*. rt*
.'*._, * *.:I;
,, '.. ') ~=
': *.1. ~ r*.
'..\\I
I "
,.~,. *J'~
~. (. t.
I :
'.-' :1
'*, bU
! * 'l'!.
I: l* * ~,£,
z.~.

~
'.,I * !'* l' I '
I "
J -
t -
(.~*. (11*
1 '.i * &6
.. I
.*/.
. ~ :~. '* '~
I:.'.. ! (.
.*~**II\\'
(.(.* :*1~
' *. ' ~~ ':
32~
!'.'. r:1
ltU, '!J
JS. J 1
.*~!,.IJ!.
. 18.ft~*
'! * ! I
"'
ll':'
11.1!.
~J **,:;
zo.3'.'
JS. l *~
JZ.t.!
J.,. r,,
'd.J.'
'. *~. ~'~*
.).*.,, :
ti I
t
- \\
I i
\\
I
\\
\\
I
-I I
I i
~
. ~
ITEM II.K.3.17:
REPORT ON ECCS OUTAGES
RESPONSE
Tables 1 and 2 provide the information required by this item for Salem Unit No. 1.
Table 1 is a list of all ECCS outages which were inspec-tion related in the last 5 years.
Table 2 is a list of all ECCS outages which were not inspection r~lated.
For Salem Unit No. 2, there have been no ECCS outages.
TABLE 1 INSPECTION RELATED ECCS OUTAGES DATE iNSPEC.I SYSTEM/COMP.
DURATION REASON/CAUSE COPPECTIVE ACTION 2/16/78 100138 11/12 SI Pump 4.0 hrs.
Motor Brg. Oil Change Changed Oil 3/22/78 100168 11/12 CH Pump 4.0 hrs.
Motor Brg. Oil Change Changed Oil 4/23/78 100116 11 RHR Pump 3.0 hrs.
Motor Brg. Oil Change Changed Oii 4/25/78 100116 12 RHR Pump 3.0 brs.
Motor Brg. Oil Change Changed Oil 8/20/78 100138 11/12 SI Pump 4.o hrs.
Motor Brg. Oil Change Changed Oil 10/21/78 100116 11/12 RHR Pump
5. O hrs.
Motor Brg. Oil Change Clanged Oil 2/13/79 100138 11/12 SI Pump 4.0 hrs.
Motor Brg. Oil Change Changed Oil 3/22/79 100168 11/12 CH Pump 4.0 hrs.
Motor Brg. Oil Change Changed Oil 2/22/79 100036 11/12 CH Pump 4.5 hrs.
Motor Brg. Oil Change Changed Oil 9/10/79 100493 21/22 CH Pump
. 5 hrs.
Motor Brg. Oil Change Changed Oil 7/17/79 100138 11/12 SI Pump c.o hrs.
Motor Brg. Oil Change Changed Oil 8/21/79 Motor Brg. Oil Change Changed Oil 100036 11/12 CH Pump 4.0 hrs.
10/19/79 100116 11/12 RHR Pump
4. O hrs.
Motor Brg. Oil Change Changed Oil 1/21/80 100038 11/12 SI Pump 5.0 hrs.
Motor Brg. Oil Change Changed Oil 2/6/80 100036 11/12 CH Pump c.o hrs.
Motor Brg. Oil Change Changed Oil 3/22/80 100168 11/12 CH Pump 4.o hrs.
Motor Brg. Oil Change Changed Oil 4/1/80 100506 21/22 RHR Pump 6.0 hrs.
Motor Brg. Oil Change Changed Oil 7/10/80 100506 21/22 CH Pump
6. 0 hrs.
Motor Brg*. Oil Change Changed Oil 7/17/80 100038 11/12 SI Pump 5.0 bra.
Motor Brg. Oil Change Changed Oil 8/C/00
,*,'1.00lii-*
11/12 RHR Pwr.p c.o hra.
?IOtor Brg. Oil C?Mnge Changed Oil B/11/80 100036 11/12 CH Pump 4.o bra
Motor Brg. Oil Change Changed Oil 10/3/80 100506 21/22 RHR Pump 6.0 bra.
Motor Brg. Oil Change Changed Oil 10/19/80 100116 11/12 RHR Pump 5.0 bra.
Motor Brg. Oil Change Changed Oil
DATE 12/1/76 12/3/76 1/10/77 5/6/77 11/8/77 2/3/78 6/G/78
~
8/20/78 8/20/78 11/27/78 ll/27/78 11/28/78 12/22/78 12/22/78 12/22/78 12/22/78 l/12/78 TABLE NON -
INSPECTION RELATED ECCS OUTAGES IRI IFAPP SYS/COMPONENT 1-76-63
, 11/12 Charging Pumps 1-76-63 11/12 Charging Pumps 1-77-12
lSJl/lSJ2 1-77-88 11/12 SI Pumps 1-77-168 lJ SIS Accum.
1-78-020 BIT 1-78-080 11 Charqinq Pump 1-78-222 11 SI Pump 1-78-223 12 SI Pump 1-78-367 BIT 1-78-369 11 RHR 1-78-374 BIT 1-78-415 12 SI Pump 1-78-416 12 SI Pump 1-78-417 11 Charing Pump 1-78-418 12 Charing Pump 1-79-16 BIT DURATION 6.5 hrs.
/
4.3 hrs.
20.0 hrs.
3.5 hrs.
7.2 hrs.
1.7 hrs.
10.3 hrs.
1.5 hrs.
0.8 hrs.
4.1 hrs.
9.3 hrs.
6.5 hrs.
l.5 hrs.
1.9 hrs.
.8 hrs.
1. 3 hrs.
1.0 hrs.
CAUSE/REASON of OUTAGE Improper valve line-up Improper valve line-up Defective Mov motor9 Pumps C
T'd High Accum. Pressure Low Boron Concentration Auxiliary Oil Pump Maint.
Oil Change Oil Change
. Low Boron Concentration Inoperable Breaker Low Boron Concentration Replace Zinc's in L.O. Cooler
Replace Zinc's in L.O. Cooler Replace Zinc's in L.O. Cooler Replace Zinc's in L.O. Cooler Low Boron Concentration
\\.
CORRECTIVE ACTION MODE Charging Procedures 3
~
Changing Procedures 3
Replaced Motors 1
Made pumps operable 3
Reduced 2
Restored Concentration 1
Repaired AOP 3
Changed Oil l
Changed Oil 1
Restored Concentration 3
Replaced Breaker 3
Restored Concentration 3
Replaced ZincQs 1
Replaced Zinc's 1
Replaced Zinc's 1
Replaced Zinc's 1
Restored Concentration 1
IRI IPAPP SYS/COMPONEH'l' DURATION 1112/79 1-79-11 BIT 14.8 hrs.
h 2/13/79 1-79-105 11 SI Pump 3.2 hrs.
2/13/79 1-79-107
~
12 SI Pump 1.8 hrs.
2/15/79 1-79-119 11 Chargin9 Pump 7.0 hra.
2/20/79 1-79-130 12 Charging Pump 8.2 hrs.
2/22/79 1-79-144 11 Charging Pump 1.3 hrs.
2/22/79 1-79-397 12 SI Pump
2. 3 hrs.
11/14/79 1-79-397 12 SI Pump 9.5 hrs.
11/15/79 1-79-404 11 SI Pump 13.7 hrs.
11/28/79 1-79-422 12 SI Pump 6.5 hrs
9/15/80 1-80-335 11 Charging Pump 3.4 hrs.
7/31/80 2-80-273 RWST 5.75 hrs.
7/22/80 l-I0-257 11/12 Charging Pump 0.9 hrs.
7/16/80 1-80-246 12 Charging Pump 12.75 hrs.
7/15/80 1-80-241 12 Charging Pump 14.0 hrs.
7/8/80 1-80-216 11 SI Pump 14.0 hrs.
6/26/80 1-80-202 12 Charging Pump 37.8 hrs.
CAUSE/REASON of OUTAGE Low Boron Concentration Oil Change Oil Change Inspect Zinc:s Inspect Zinc's Oil Change Oil Change Vibration Check (Motor)
Adjust Pump Seals Packing Leak 12SJ40 Repair Servi~~ Water Leak Low Boron Concentration Faulty Disch. Press. gage during surveillance test.
Repair AOP Pressure Switch Speed changer Oil Level low Water Oil/Failed L.O. tube coder Replaced L.O. Coder (Failed tube)
CORRECTIVE ACTION MODE Rest~red Concentration l
~
Changed Oil l
Changed Oil 1
Completed Inepectic..1s 1
Completed Inspections 1
Changed Oil l
Changed Oil 1
Checked vibration 3
Adjusted Seals 3
Repacked Valve 3
Repaired Leak 1
Added Boron 3
Repaired gage 1
Repaired pressure switch 1
added oil 1
Plugged L.O. Cooler 1
Replaced L.O. Cooler 1
r' IRI IP'APP SYS/COMPONENT DURATION 6/12/80 1-80-185 12 BAT 7.5 hrs.
5/21/BO l-80-156 12 SI Pump l.3 hrs.
3/7/80 1*80-102
'12 c1'aring Pw.p 2.6 hrs.
3/5/80 1-80-98 11 BIT 1.0 hr*.
2/28/llO 1-10-89 Vital Heat Tracing 5.5 hrs.
2/27/IO 1-80-88 Vital Heat Tracing 10.5 hr*.
2/6/80 1-80-62 12 Charging Pwap 1.8 hrs.
2/5/BO 1-80-61 11 Charging Pwap 6.0 hrs.
1/23/80 1-110-47 BIT 5.6 hrs.
1/21/80 1-80-41 11/12 SI Pump11 3.5 hr* each 1/15/80 1-80-27 BIT 3.5 hrs.
1/14/80 1-80-26 BIT 5.5 hrs.
1/10/80 1-80-19 11 RHR Pump 11ech.
22.5 hrs.
heat exchanger 1/7/110 1-80-14 11 Charging Pump/
47.0 hr*.
12 RHR pump mech.
seal heat exchanger l/4/80 1-80-14 12 Charging Pump 23.5 hrs.
CJ\\ USE/REASON of OUTAGE Low Bat Level Erroneous 1U1111eter reading Inspect Zinc'*
Repair Flow avitch Shorted primary circuit Shorted Primary Circuit Oil Change Oil change Repair recirc f l<'.'V met~r Lo Flow alarm circuit Oil Change Check recirc flow meter Check recirc Flow meter Heat Exchanger supports Heat exchanger support*
CC Pipe Support*
and CORRECTIVE ACTION MODS Restored I.9vel l
Replaced Amleter 1
Completed In*pection 1
Repaired avitch l
Cleared 11hort 2
Cleared Short 3
Chan9ed Oil I
1 Changed Oil 1
Repaired Same 3
Changed Oil 1
Completed check 3
Completed check 3
Repaired *up~rt*
l Repaired *up~rt*
1 Repaired 11upportm l

ML18085A447

Contents

Text

RESPONSE

Response

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Response

RESPONSE

Navigation menu

ML18085A447

Text

RESPONSE

Response

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Response

RESPONSE

Navigation menu

Search