Text

Forwards Listing of Status of Tech Spec Issues Identified in Section 16 of Ser.Description of Resolution of Issues & Draft Rev 5 to Tech Specs Also Encl
	ML20125E482
Person / Time
Site:	Hope Creek
Issue date:	06/10/1985
From:	Mittl R; Public Service Enterprise Group
To:	Butler W; Office of Nuclear Reactor Regulation
References
	MP85-81-02-1-VW, MP85-81-2-1-VW, NUDOCS 8506130134
	Download: ML20125E482 (75)
	v • d • e

( _.

.s C PS Putsc Serwce Electnc and Gas Company 80 Park Plaza, f Jewark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation June 10, 1985 Director of Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission

'7920 Norfolk Avenue Bethesda, Maryland 20814 Attention:

Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:

SAFETY EVALUATION REPORT TECHNICAL SPECIFICATION ISSUES HOPE CREEK GENERATING STATION DOCKET NO. 50-354 is a current list which provides a status of the Technical Specification Issues identified in Section 16 of the Safety Evaluation Report (SER).

Items identified as

" complete" are those for which PSE&G has provided responses and no confirmation of status has been received from the staff.

We will consider these items closed unless notified otherwise.

In order to permit timely resolution of items identified as " complete" which may not be resolved to the staff's satisfaction, please provide a specific description of the issue which remains to be resolved.

Enclosed for your review and approval (see Attachment 3) are the resolutions to the SER Technical Specification Issues listed in Attachment 2.

Also enclosed (Attachment 4) for your use and incorporation into the Hope Creek Generating Station Draft Technical

-Specifications are five (5) sets of the following revised Hope Creek Generating Station Draft Technical Specification pages:

MP85 81 02 1-vw 8506130134 850610 p

PDR ADOCK 05000354 E

pm The Energy Peopk) 954312 GV) 4-84

Director of Nuclear Reactor Regulation 2

6/10/85 Pages 11 3/4 2-2 3/4 3-43 3/4 8-6 Insert to Pg. B 3/4 4-1 4 3/4 2-2a 3/4 3-45 3/4 11-17 6-7 3/4 2-2b 3/4 3-46 Insert to Pg. B 3/4 1-4

.6-19

-Insert

'A&B to.

Pg. 1-4 3/4 2-2c 3/4 3-47 B 3/4 2-1 1-5 3/4 2-2d 3/4 3-58 B 3/4 2-3 Insert A

.to Pg. 1-5 3/4 3-7 3/4 4-5 B 3/4 4-1 1-6 3/4 3-8 3/4 4-7 Insert A&B Insert A&B to to-Pg. 1-6 3/4 3-13 Pg. 3/4 4-7 1-7 3/4-3-19 3/4 4-9

' Insert-A to Pg.~1-7 3/4 3-27 3/4 6-20

~1-9 3/4 3-32 Insert to Pg.

3/4 6-20 3/4 1-21 3/4 3-34 3/4 6-23 3/4I2-1 3/4 3-35 Notes for Pg.

3/4 8-5 These pages-are submitted as Revision 5 to the Hope Creek Generating Station Draft Technical Specification and have been revised in accordance with the resolutions to the SER Technical Specification Issues provided in Attachment 3, and to incorporate general revisions to the HCGS Draft Technical Specifications.

Also included are copies of the updated Hope Creek Generating Station Draft Technical Specification List of Effective Pages (LEP).

Should you have any questions or require any additional information on these items, please contact us.

Very truly yours, Attachments MP85 81 02 2-vw L :-

I

-Director.of Nuclear Reactor Regulation 3

6/10/85

'C' D..'H. Wagner USNRC-Licensing Project-Manager (w/ attach.)

A. R. Blough' USNRC Senior Resident Inspector'(w/ attach.)

'MP85 81 02'l/3-vw

m 7-7 q

l

- Attachnent 1

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~ iSER TECHNICAL-

'SER,

- R. L. MITIL' 10 ',

-SPECIFICATION. SECTION A. SCHWENCER ISSUE NLNBER SUBJECT STATUS LETIER DATED

-1 2.4.11.2 Service water intake tanperature-Complete 4/10/85 2.

2.4.14 Closing of doors and hatches Canplete 4/10/85:

3 3.9.6 Pressure isolation valves Open 4

4.4.4 Thennal-hydraulic instability Cmplete 4/10/85

'5 4.4.4 Single-loop operation Complete 4/10/85-6

~4.4.5 Crud effects Canplete

4/10/85

-7 4.4.6 Icose parts monitoring system Cmplete 4/10/85 channel operability 8

5.2.2 Safety / relief valve (SRV) test Cmplete 5/13/85 program 9-5.2.5 Reactor coolant-pressure boundary Cmplete 4/10/85 leakage rates 10 5.4.6 Reactor core isolation cooling pmp Complete 4/10/85 testing 11 5.4.7 Residual heat removal system pmp Cmplete 4/10/85.

operability t

12 6.2.1.6 Torus /drywell vacuts breaker and Open vent system testing 13

-6.2.1.6 Vacutn breaker position indication Cmplete 5/13/85 accuracy 14-6.2.3 Testing of inleakage rate and draw-Complete 4/10/85 down time 15

-6.2.4.1 Isakage testing for valves with Open resilient seals l16-6.2.6 Containment isolation valve leakage Open r-MP85 8102 4-vw Page 1 of 4

w.

_ (Cont'd)

--SER R. L. MITIL '10 TECHNICAL SER-

i. SPECIFICATION - SECTION A. SCHWENCER ISSUE NLMBER SURTECT

. STA'IUS IEITER DNIED 17 6.2.6, 6.7, Main steam isolation valve leak rate Open 15.6.5.2

. testing 18-6.2.6-

.Various valve leak rates Open

'19 6.3.4.2.

Dnergency core cooling system Cmplete 5/13/85 (ECCS) subsystem flow rates 20 6.3.4'.2 BCCS subsysten operating sequence Cmplete

-5/13/85

'21 6.5.1.3 Water. seal bucket drain tap' Cmplete 4/10/85 surveillance

.22 7.2.2.3~

Testability of plant protection Cmplete 5/13/85 system at power

'23 7.2.2.8, Anticipated transients without Conplete 6/10/85.

7.6.2.2 scram mitigation 24 7.2.2.9 Reactor mode switch Complete 5/13/85 t

25.

7.3.2.3 Freeze protection of wdter-filled Complete' 6/10/85 lines i

26-7.4.2.3 Remote shutdown system operability Open

[

27 7.6.2.1 Im-pressure /high pressure systems Open.

interlocks t

281 7.6.2.3 Average power range monitor Complete 5/13/85 electrical protection assemblies 29 7.7.2.2 Nonsafety-related equipnent Couplete 5/13/85 operability 30 8.3.1.3 Diesel generator connected loads C mplete 4/10/85 31

.8.3.1.7 Load sequencer logic Cm plete 4/10/85 32 8.3.2.7 DC systen monitoring Complete 4/10/85 MP85 81 02 5-vw Page 2 of 4 l-

.=-

. - - ~.

Attachment.1 (Cont'd)

SER TBCIMICAL.

SER R. L. MITIL 70 SPECIFICATION SECTICN A. SCHWDICER

~ ISSUE NLMBER SUR7ECT STATUS LEITER ERTED 33 8.3.3.3.5

. Testing of breaker time-overcurrent-Canplete 5/13/85 trip characteristics 34 8.3.3.4.1

-Periodic systen testing Complete 4/10/85

- 35 8.3.3.4.2-

'Ioad sequencer testing Couplete 4/10/85 36

. 8.3.3.5.4 Testing of fuses Ccuplete 4/10/85 37 9.1.3 Fuel pool cooling system p m ps Open 38 9.2.1 Station service water pump testing Cmplete 4/10/85 39 9.2.2 Safety auxiliaries cooling system Complete 4/10/85 and reactor auxiliaries cooling system pm p availability 40; 9.2.2 Safety auxiliaries operability to Open ensure diesel generator cooling 41 9.2.7 Control area chilled water system Open availability

' 42 9.3.1, Air quality. testing Complete 5/13/85 9.3.6 43

- 9.3.2 Core damage estimate procedure This is a mnfirmatory iteme

~44.

9.5.1.5

-Fire watch Conglete 4/10/85 45' 10.2 Turbine stean valve inspection Ccmplete 4/10/85 46 10.4.4 Tbrbine bypass valve surveillance Cmplete 4/10/85 47 15.2.

Ibrbine bypass system and level Couplete 5/13/85 8 high-water trip performance 48 15.4.9 Scram speed Complete 5/13/85.

49 15.6.4 Primary coolant activity Conplete 5/13/85 i

[

MP85 81 02.6-vw Page 3 of 4 r

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SER TBOINICAL SER R. L. MITIL 'IO '

SPECIPICATICN SECTION

- A. SCHWENCER ISSUE NtMBER SLRTECT STATUS ETIER IRTED 50' 15.6.4 Main steam isolation valve closure Cmplete-4/10/85 time 51 15.9.3 SRV failure reporting Caplete 4/10/85 52 15.9.3 Autcznatic depressurization system Complete 5/13/85 logic I11D:w M85 8102 7-w Page 4 of 4 f, '

i'

~

4 ATTACHMENT 2

-SER

-- TECHNICAL SPECIFICATION SER ISSUE SECTION SUBJ ECT

-23 7.2.2.8, Anticipated transients without

-7.6.2.2 scram mitigation 25 7.3.2.3 Freeze protection'of water-filled lines

.DFD:az 5/10/85 MP85'96/07 01

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i ;-

ATTACHMENT 3 i

l s

f I-i Ii 1

+

l p

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MP85 96/07 02 i

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m SER_ Technical Specification Issue No. 23 (SER Section

7. 2. 2.8,- 7.6. 2. 2 )

Anticipated transients without scram mitigation The staff will verify that the Technical Specifications include appropriate limiting conditions for operation and surveillance requirements on the RRCS.

Response

.The RRCS ATWS detection sensors which automatically initiate the alternate rod insertion subsystem, the standby liquid control system, the RWCU system isolation, and the recirculation pump trip subsystem are reactor vessel pressure and water level.

HCGS Draf t Technical Specification Tables 3. 3. 4.1-1, 3. 3. 4.1-2, and 4.3.4.1-1 have been revised to provide clarification and_to include the timer for the level 2 delay for the recirculation pump motor breaker trip.

DJ D:mr NL 15 1

k, SER Technical' Specification Issue-No. 25 (SER'Section 7.3.2.3) Freeze Protection of Water Filled Lines The Technical: Specifications should include surveillance

-requirements for testing the environmental control and monitoring systems at least once per year before the advent of freezing. weather.

The staff will verify that the Technical Specifications will include 1 surveillance requirements for testing the environ-mental control and monitoring-systems at least once per year before the onset of freezing weather.

In addition, if the analog output of the RTD becomes unavailable, the staff will verify'that administrative procedures will provide for

. verification that the sensing line is not.in danger of freezing.

' Re spon se s -

.HCGS: Draft. Technical Specification Section 3/4.3.3, ECCS Actuation Instrumentation has been revised to include surveillance of the condensate storage tank level sensing line temperature monitoring instrumentation and heat

' tracing.

DJD:bp MP 85 119 03/1 l

L' r:

ATTACHMENT 4 Revision 5 HCGS Draft Technical Specification Pages MP85 96/07 03

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Insert A&B to pg. 3/4 11-18 2

3/4 11-19 2

3/4 11-20 2

3/4 11-21 2

3/4 12-1 2

3/4 12-2 2

3/4 12-3 2

3/4 12-4 2

3/4 12-5 2

3/4 12-6 2

3/4 12-7 2

3/4 12-8 2

3/4 12-9 2

3/4 12-10 2

3/4 12-11 2

3/4 12-12 2

3/4-12-13 2

3/4 12-14 2

Bases for Section 3.0 and 4.0 0

Note to pg. B3.0&4.0 0

B3/4 0-1 0

B3/4 0-2 0-B3/4 0-3 0

B3/4 1-1 0

B3/4 1-2 0

Insert A to pg. B3/4 1-2 0

B3/4 1-3 0

B3/4 1-4 5

Insert to pg. B3/4 1-4 0

B3/4 2-1 5

B3/4 2-2 0

B3/4 2-3 5

B3/4 2-4 0

B3/4 2-5 0

MP85 20/09.12-db Rev. 5 9

.-_r,

6/85 HOPE CREEK GENERATING STATION DRAFT TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES Revision Pages Numbers B3/4 3-1 0

B3/4 3-2 0

B3/4 3-3 0

B3/4 3-4 0

B3/4 3-5 0

B3/4 3-6 0

Insert to pg. B3/4 3-6 0

B3/4 3-7 0

B3/4 4-1 5

Insert A to pg. B3/4 4-1 5

Insert B to pg. B3/4 4-1 0

B3/4 4-2 0

B3/4 4-3 0

B3/4 4-4 0

Insert to pg. B3/4 4-4 0

B3/4 4-5 0

Insert A to pg. B3/4 4-5 0

B3/4 4-6 0

B3/4 4-7 0

B3/4 4-8 0

B3/4 5-1 0

B3/4 5-2 0

B3/4 6-1 0

Insert to pg. B3/4 6-1 0

B3/4 6-2 0

B3/4 6-3 0

B3/4 6-4 0

B3/4 6-5 0

Insert to pg. B3/4 6-5 0

B3/4 6-6 0

B3/4 7-1 0

B3/4 7-2 0

Insert to pg. B3/4 7-2 0

B3/4 7-3 0

B3/4 7-4 0

Insert to pg. B3/4 7-4 0

B3/4 7-5 0

B3/4 8-1 0

B3/4 8-2 0

B3/4 8-3 0

B3/4 9-1 0

MP85 20/09 13-db' Rev. 5

c3 6/85' HOPE CREEK GENERATING STATION DRAFT TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES Revision Pages Numbers B3/4 9-2 0

-B3/4 10-1 0

B3/4 11-1 2

B3/4 11-2 2

B3/4 11-3 2

B3/4 11-4 2

B3/4 11-5 2

B3/4.11-6 2

B3/4 12 2 B3/4 12-2 2

Section 5.0 0

5-1 0

5-2

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' 4 0

5-5 0

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Section 6.0 0

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6-3 0

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6-6 0

6-7 5

6-8 0

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6-10 0

6-11 0

6-12 0

6-13 0

6-14 0

6-15 0

6-16 0

6-17 0

6-18 0

6-19 5

Insert A,B&C to O

page 6-19 6-20 0

6-21 0

MP85 20/09 14-db Rev. 5

Q 6/85 HOPE CREEK GENERATING STATION i

DRAFT TECHNICAL SPECIFICATIONS

. LIST OF EFFECTIVE PAGES Revision Pages Numbers 6-22 0

6-23 0

6-24 0

6-25 0

6-26 0

Insert B to pg. 6-26 0

6-27 0

6-28 0

6-29 0

MP85 20/09 15-db Rev. 5

wg=ww

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DitAFT TECHNICAL SPECIFICATICatS 1

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l!$El DEFINITION 5 -

SECTION DEFINITIONS (Continued)

PAGE z.1 1.y OPERABLE - OPERABILITY.....................................

1-4 ll u

1. M OPERATIONAL CONDITION - C0NDITION..........................

1-4 U 1 4 PHYSICSTESTS..............................................

1-4 l

1,.

PREc gs,e0uNwyappg................................... g,. n n.~.

i, RIMARY CONTAIP9 TENT INTEGRITY..............................

I-5r ll 1.$RATEDTHERMALP0WER........................................

1-5 ll j

3(

1 7 REACTOR PROTECTION SYSTEM RESPONSE TINE....................

1-5 l

sysul-1.)lr REPORTABLE-4G89RRENCF...................................... 1-5 II 37 1.Jr R00 0 ENS ITY................................................ 1-5 l a sacaea, w omg secoacA&Y dod74 nnrd) 1.pt xx 27 6 Ma_(& INTEGRITY............................

i 1-6 ll l

)

J

.1T 1.Js siiUTDOWN M4RG IN............................................ 1-6 l

rn 1.)4 STAGGERED TEST 8A515.......................................

1-6 Il 4

1 5 THE RMAL P0WE R.............................................. 1-6 l

._,,l s.._

1.

TURBINE SYPASS SYSTEM RESPONSE T18E........................ 1-7 l

Ko rw 1-7 l

1.JF WIIDENTI FI ED LEAKAGE.......................................

TABLE 1.1, SURVEILLANCE FREQUENCY NOTATION......................

1-8 4

i TAs LE 1. 2, OPEPTIONAL CONDIT10NS...............................

1-9 l

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OEFINITIONS LOGICSYSTEMlUNCTIONALTEST

1. 6 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, ll s.3-i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERA 8ILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be perfomed by any series of sequential, overlapping or total systes steps such that the entire logic system is tested.

'=:""" TOT ^' ="'MIT rea 1.

77 "JJ::""" "JTAL ":^%:": TA"0: (~"T) ;h11 M tM h ;;;t T."T eic.

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rditi=.)

ll -

-9 IMWA T A MININUN CRITICAL POWER RATIO 17t The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

Af exists in the core.'?x xh :!:n :? fr?).

p Ment 6 OPERABLE - OPERASILITY 1.g A system, subsystem, train, component or device shall be OPERABLE or have l A7 OPERA 8ILITY when it is capable of performing its specified function (s)

}

and when'all necessary attendant instrumentation, controls, electrical s

power, cooling or seal water, lubrication or other auxiliary equipment that are required for the systaa, subsystas, train, component or devica to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CONDITION 1./ An OPERATIONAL COWITION, i.e., CON 01 TION, shall be any one inclusive l

I 88 combination of mode switch position and average reactor coolant c

temperature as specified in Table 1.2.

PNYSICSTQTS, l

1.MPHYSICETESTSshallbethosetestsperformedtoesasurethefundamental l

z.9 nuclear characteristics of the reacter core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) othenrise approved by the Commission.

l i

PRESSURE SOUNDARY LEAKAGE 1.gPRES$URESOUNDARYLEAKAGEshallbeleakagethroughanon-iselablefault l

3, in a reactor coolant system component body, pipe wall or vessel wall.

l i

N @ f_ c heg w.

gev. S

47: (= )

1-4 t_

P l-4 :

Ensee:r K to 9

MEMBER ($) 0F THE PUBLIC I.19 trP MORER(5) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use persions of the site for recreational, occupational or other purposes not associated with the plant.

P.

I- 'l 3 Tusee.r a t.

S OFF5ITE 00$E CALCULATION MANUAL (nncM) 1 26 M The OFF5ITE 005E CALCULATION MANUAL shall contain the current methodol and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alare/ trip setpoints, and in the conduct of the environmental radiological sonitoring program.

=

l l

a O

kev,S

OEFINITIONS PRIMARY CONTAI MENT INTEGRITY

+ 1,4 san.T( A1.) PRIMARY CONTAllelENT INTEGRITY shall exist when:

l 13 a.

All primary containment penetrations required to be closed during accident conditions are eitner:

1.

Capable of being closed by an 0PERABLE primary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in.its closed position, except as provided in Table 3.5.3-1 of Specification 3.6.3.

b.

All primary containment equipment hatches are closed and sealed, c.

Each primary containment air lock is in compliance with the requirements of Specification 3.5.1.3.

d.

The primary containment leakage rates are within the limits of Specification 3.5.1.2.

e.

The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

]

f.

The sealing mechanism associated with each pr mary containment penetration; e.g., welds, bellows or 0-rings, is OPERA 8LE.

RATED THERMAL POWER 1.dRATED THERMAL POWER shall'be a total reactor core heat transfer rate to l

3F the reactor coolant of $31WT.

REACTOR PROTECTION $YSTEM RESPONSE TINE 1.M REACTOR PROTECTION SYSTEM RESPONSE TILE shall be the time interval from '

l l

38 when the monitored parameter exceeds its trip setpoint at the channel i

sensor until de-energization of the scras pilot valve solenoids. The l

response time may be asasured by any series of sequential, everlapping or l

total steps such that the entire response time is sensured.

1 REPORTA8LE 000We#ENef & vr#1" 1.gAREPORTABLE b

! shall be any of those conditions specified in l

% ';=4 *Wt'... G.:.1.; 15.t. '.t. Sachee. SO 716e /0 CFt ind 30 ROD DENSITY 1.[ 300 OENSITY shall be the number of control rod notches inserted as al 77 fraction of the total number of control rod notches. All rods fully inserted is equivalent to 1005 R00 OENSITY.

t?L...PPE' 1-s s-,

[ev..F

\\

InstAT A ' to P.

I-f 9

PROCESS CONTROL PROGRAM (PCP) 1.31 4,41 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, testa, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive wasta.

PURGE - PURGING

l. :pA 2,.4a PURGE or PURGING is the controlled process el discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

em N

I 4.v. s

4 OEFINITIONS

[

gy m m aus ( s ecwpae ( <ennawnt)

% :Z Z?"" Z 3 C.. 7 INTEGRITY l

asAeros h tpjg6 "' INTEGRITY shag 1 exist when:

.l 1.M %:Z J -" Z:...._

All':coardpc WitA!3 z: sN N

T- ; __--

2 penetrations required to be closed during a.

accIdentconditionsaree,ither:

Capable of being closed by an OPERA 8LE :ned.e k atL*s3 c; t

t re :::t_.

i 1.

automatic isolation system, or l

d automatic (:_?r:) (:-) {damperl (.[- - ;?f::ri:) secured in its Closed by at at one manual valve lind flange, or deactivated 2.

1 l

l i

closed position. :n--t _,...___

" ::tfi xtt:- 2.0.0.".

2sfa>[

(see in <=ab' *)

All :eemebp. LJtic

_ rj ::: a ::: ^ hatches and blewout panels are closed b.

and sealed.

sig w % hnlw m A a.4 V=A M m c.

The :t C, ;:: ^ n^n ^ system is in compliance with the requirements of Specification 3.6.5.3.

.f g.

The sealing mechanism associated with each ^,O:r,%A%s z; t,

]

penetration, e.g., welds, bellows or 0-rings, is OPERABLE.

5M c) /.

The pressure within the c::us,:< L*d11 4 _ d::_ r; :: _. __... is less than er equal i

fa to the value required by Specificatten 4.6.5.1.a.

l 5HUTDOWN SRGIN 1.M SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is l

31 subcritical er would be subcritical assuming all control rods are fully inserted except for the single control red of highest reactivity worth which is assumed to be fully withdrawn and the reacter is in the shutdown condition; cold, i.e. 64*F; and menon free.

--, inn AT &

1 STAGGERED TEST-8 ASIS 1.JIASTAGGERED.TESTSA515shallconsistof:

l a.

A test schedule for n systems, sesystems, trains or other designated campenents obtained by dividing the specified test interval into n l

'l equal subintervals.

i b.

The testing of one systes, subsystem, train er other designated component at tne beginning of each seinterval.

THERMAL POWER 1

l 1.8 TNun4t,0wtR shan b. th.==i reacar c.re h t transf.r rat. a th.

l g reacter coolant.

5

s. wit 1-6

...s_-.,,

,A%..S*

INSERT A TO TG. 1-6:

d.

For double door arrangements, at least one door in each access to the reactor building (secondary containment) is closed.

e.

For single door arrangements, the door in each access to the reactor building (secondary containment) is closed except for routine entry and exit.

B +. P. l~ 0 :

l'a ssa.T S

5!TE 80UNOARY I.40 1.E The $1TE B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet westes into a fore that meets shipping and burial ground requirements.

SQLlRCE CHECK A 5OURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

key. C

DEFINITIONS

-- =

= h =

g

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1.

The TOTAL PEAKING h

' ^;

r o of local LMGR for any ll

~

specific location v

core average LNGR associ-g: ;r:r g bundles of the same type opera core average r

ll TUR$!NE BYPASS SYSTEM RESPONSE TIME (f

1. X The TURSINE SYPASS SYSTEM RESPONSE TINE 2:!' 5: tt:t t'n '-': ::t *:n

' t st r: n t;-' ' :t tt :t: =;1 l dx tt: (x ' u nd ;:ri-: t r :x x t:

r = r) (t e' = ig n: :r =? ='t ;r:nte : to'- ign: n?= <!=

';.;!) x^ti tt: t-tt= ig:x nha innt ^: ^2:'

n; 'nd ;n'*e r:,

Tu n:; n n t!= a 5: 2 :r nd tj =j n ':: :" x ;x t':?, n ;rl: ;* ;

t^:? ^^:;; x 2 tt:0 tt: x^' : n:;:n: t' n

r r n d.

UNIDENTIFIED LEAKAGE 1.NUNIDENTIFIEDLEAEAGEshallbeal'1leakagewhichisnotIDENTIFIEDLEAEAGE. l u

y IN3cnT A edV oh. h ab m Mdd w eed

, c4 he. % bWe. 4ConsisIta, ok ho c.ovwhM d,

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% =3 nes E 3.epevh.k\\, w\\g bL\\. 4, SLA.

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I~ 7 I

..hsen.T A To VS 4

UNRESTRICTED AREA I

/.47 h4d An UNRESTRICTED AREA shall be any area at or beyond the SITE 800NOARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXNAUST TREATMENT SYSTEM A 98 i

W A VENTILATION EXHAUST TREATIENT SYSTEM is any system designed and installed to. reduce gaseous radiofodine or radioactive asterial in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorters and/or NEpA filters for.the purpose of removing todines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be

)

i VENTILATION EXHAUST TREATMENT SYSTEM components.

1 VENTING

/,4f.

h4g VENTING is the controlled process of discharging air or gas from a con-finement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner.that replacement air or gas is not pro-vided or required during VENTING. Vent, used in systas names, does not imply a VENTING process.

4 W

I l

Ru. s

._m.,_

DEFINITIONS

~

TA8LE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1.

POWER OPERATION Run Any temperature 2.

STARTUP Startup/ Hot Standby Any temperature 3.

NOT SHUTDOWN Shutdown #'***

> 200*F l

4.

COLD SHUTDOWN Shutdown #'##'***

1 200*F l

5.

REFUELING" Shutdown or Refuel **'#

1 140*F l

N wJ nJ re /ete8 inSY'"**a

The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlockhfunctions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

ffThe reactor mode switch may be placed in the Refuel position while a single control rod-drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the coactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

I

- 5ee Special Test Exceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERABLE.

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VOLUME - CONCENTRATION

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Y v3 VOLUliE 13.4 y

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MINIMUM REQUIRED CONCEllTRAfl0l4 LillE g

5 0

=

4f 53 4990 5066 12.0 l

l I

4400,

4500 4600 4700 4800 4900 5000 5100

5200 g

V - NET VOLUME (GALLO!!S)

.s n

figRE3.1.5-1 S@ldg YENDfB0RRTC CoffyrgtyfgAl AS a papcnod of UEf (DM/9lf p

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i 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Floures 3.2.1-1, 0.0.1-0,. 4 0.0.1-2.

3. 2. / -J.,.r. 2.1 -3, 2. 2. t - 4 ll And 3 2./-f.

APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 325% of RATED THERMAL POWER.

ACTION:

s.2.1-1 s.1.1-3, s.s./- y a, s. s.i. 5-With an APLHGR exceeding the limits of Figure 3.2.1-1, 0.0.1,2, n 3.2.1-0, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than T253%ofRATEDTHERMALPOWERwithinthenext4 hours.

~J SURVEILLANCE REQUIREMENTS l

4.2.1 All APLHGRs shall be verified to be equal to or less than the limits

.:.;-,.c4.2.1-::

l determined from Figures 3.2.1-1p3 t..I-1, 3.s.I ~2, 3. A,/ -tf, an/ 3* 2./ ~S~ l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at l

1 east 155 of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a' LIMITING CONTROL A00 PATTERN for APLHGR.

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o I

I m

CHAMEL y(

OPERAT10114L CHAIBIEL FUNCT10114L CHAISIEL COISITIONS FOR lAIICu

i FINICTICIIAL isIIT I

CHECK TEST CAlltRAT1011,)

SURVEILLAIICE REQUIRES 1.

1 Intermedtate Range p ftors:

10 a.

IIeetron fleet - Nigh 54,5,g SN, W R

2 l

1 i

5 W

R 3,4,5 j

b.

Inoperative 11 4 W

IIA 2,3,4,5 2.

Average Pouer Range flanfterNI:

go a.

Itsetron Flum -

54,5,y SN, W SA j

2

)

upscale, Setdown 5

w SA 3,{t b.

Flaw Blased Simulated Q

j Theres1 Power - W ale 5,0N8)I M,X MdII'I,5A,(RIA){

i Y

1 i

c.

Flmed IIeetron Fleet -

Y Upscale 5

M. it M*I,5A 1

4 d.

Insperative 11 4 '

iPQ IIA V

1,2,3,[5[

}

Je.

Somescale 5

WR SA 1{

l 3.

neecter Vessel Stean sene Pressere - nigh 35f

-4 18/g 1, 2 4.

asector vessel tinter Level -

Law, Level 3 ISS M8 f[

1. 2 l

i9 5

5.

IIsla Steam Line Iselatlen m

Valve - Closure 11 4

4 R

1 6.

91 min Steam Line Radiatien -

1 High 5

Q R

1, 2(g) 1 7.

JPriesty Cantainment) A c!!,

{

Pressure - Nigh jsf

+4

' {A)g 1, 2 i

)

1

i e.v i

i TABLE 4.3.1.1-1 (Continued)-

REAC18E POSTECTION SYSTEN IIISTEWElffATION SURVEILLANCE Equ1REleffs

%,2.

7 CWISEL OPERATICIIAL l

i O CIWWWEL FisICTICIIAL DIAIBIEL COIGITIGIIS FOR 18tICH f,

i l 4 FWETIBIAL WIIT CIIECK _

TEST CAL 10 RAT 10Il SURVEILLAIICE ItEqulREO I

n 5c 2

t a.u - M'rhE, tinter e.e -15F

4 da 1, 2, s l)

'l Scran Sischarge vel L

l %k i

h s.

tadA.

03

k. %t suiA.ws MA Q.

It t 1,6 9.

Tertine Stap Valve - Closure 4)%

4 IIII 1

19. Terhine Centrol Valve Fast Closure Valve Trip System l

Sil Pressure - Law 44MA WQ 181

)

1 i

i l

11. Reacter IInde Switch 1

Shutdeun Positten IIA R

IE 1,2,3,4,5 w

12. Nonmal Scran' 11 4

.WW IIA 1,2,3,4,5 i

neutron detectors may be encluded from OIAISEL'CALIORAllGN.

w (a)

The INI and SWI channels shall be determined to everlap for at least 1%R decades during each startup en (b) after entering SPERATIGIRL CSIBITISII 2 and the 150 and APWI channels shall be determined to everlap for at least145 decades during each controlled sIndesun. If not performed within the previous 7 days.

^10;.'" E. p ' r t: -t=-*

. if

^ ; - ' c f J ^4 0 "

_ _ _ 7 L..

4e)

This calltration shall consist of the adjustment of the APWI h1 to confers to the peuer values (d) calculated by a heat balance during SPERAT10114;. CSISITIGII 1 when TENEL POWER > 25E of IIATES Adjust the APWI channel if the absolute difference is greater teen 2K of RATED TIENEL TIEmmL PSER.

PSER. Any APWI channel gain adjestment made in compliance with Specification 3.2.2 shall not be r

l m

included in determining the absolute difference.

l i

Q (e) This calltration shall coasist of the adjustment of the APWI flew biased channel to confers to a l

l callkreted flew signal.

{

(f) The LPWIs shall be calibrated at least once per ites effective full peuer hours (EFrel) h

[

rec;m.idi= leg (W using the TIP system.

(

I aes #3..y

.f.

j (g) Verify measured core fleu[to be greater than er equal to established core flew at the existing,

.^

, /4w 3(h) This calibration shall consist of { 4'- ^- ;", = ;4 ;;f, :")(verifyingl the 615 second h

I slamisted thermal power time constant.1

~

(

(i) This function is not required to be OPERA 0LE when the reacter psure vessel head is removed per l

v:

1; j

l Specification 3.10.1.

1 (j) With any centrol red withdraun. Itet appilcable to control rods removed per Specification 3.9.10.1 l0 l

i j

er 3.9.10.2.

Ng @ f

[

j (2 talNe % subtid lead **ce. W t

TABLE 3.3.2-1 (Continued)

, IG ISOLATION ACTMil0N INSilRSENTAil0N g

1 ;; ^

VALVE ACTUA-TION GROUPS HINIMUN APPLICABLE 1 :

I OPERATED BY OPERABLE CHANNE OPERATIONAL SIGNAL (d) *PER TRIP SYSTEN gI-U TRIP FINICTION CONDITION ACi GH 1 f

f 5.

REKTORCOREf50LATIONCOOLINGSYSTENISOLATION 1

a.

ACIC Steam Line & Pressure -Htgh G

J1) %)ve.

1, 2', 3 23 b.

Reu:. She 1.1,4. A Pogoge. Mgh,Ti.or (s

i vahp0 '

I, S,3 23 CK KIC steam supply 4

2/ Valve. (O 1, 2. j 23 Pressure - Low-4 K.

KIC Turbine Emboust 6

2/Va\\vt '}

1,2,3 23 I

elaphrop Pressure - High M

E% >

[

.t)f.

KIC r;'

M neem 115/y.lve C')

1, 2, 3 23 M

,Tamperature - Nigh P-ene f.)(,

KIC E;*;--r asem W,hildin h25 6

fli/V=IW.CO 1,2,3 23 A Temperature - High P

KIC 7.c. P:my ^ 'q Area it)4/Valvt(c) f-37

(,

1, 2, 3 23 Teeperature - High g

m i;c.ypet\\ Pmsyn t 3@)

h<

Hi

, 7:s 1..a

(,

m a/valvt 1, 2, 3 23

/ y, 1 T

=.

n':.

M.

RE!'"-:;-;":

f--h-44-)-

L, 2, 3

-d-Temperet e

%',0.

(,$h%

1,2,3 X25J J.

Manuai Initiation A. x,. w w m -. y

/,J) g.3 23 yp,l-r 4

+

e-

?

-m m

N e

TM4.E 3.3. 2 -1

'130 LPrtio N

&cTUPrT80Al INSTR 0Wh7:W n$luTS U@

h 3

)

s P i

k blMtt.G 8

b TRIP FUNCTION TRIP SETPOIW yAt UE. 8

't t$

5.

REACTORCORdISOL'ATIONCOOLINGSYSTEMISOLATION n

a.

RCIC Steam Line A Pressure High i 118,0, H o 5 127.0 %O 3

b, Reic sam um A Pvegure. 4

, Time 2 3.0 scumds 6 rs.o sand.s (K

RCIC Steam Supply 2 (,q,5 psiy 2 %.f pg9 Pressure - Low-a dK.

RCIC Tuttine Exhaust i10.0 ps;9 g go.o p;p Diaphragm Pressure - High Em2

[

.t)0 RCIC E W eent Room (lahr)

(Inkr)

Temperature - High

M3 A

Pamp k%

RCIC5;t---IRoomVe'nhNionhdi (laltr)

[laMr)

A Tamperature - High D

. P 4

Skr)%

RCIC ?jrees '.; Area

~ &o *sF h

3.K.

f

.y Temperature - High s

br 1

%_ ygg(j P n;9,t Hi3,-.

6 l Ce8 psip 6 f.7P fsi3

/ Fe r y,Ei

. _. _......f - -=, @h g

Temperstm - :::#,

j t.

Manual. Initiation NA-NA

-h

%s Area Ti,yerdum-Ily4 p3o p *

,3+

A O

i, N

l!

Tnsts s.3.2-4

!l TsoLhrtoAl MTUA-T/oeV *Ns7RumENfh7'iod t

TRIP FONtTIDAJ T. RG#.,T0t. tore, TsoL&fioN C00LtN6 VALVES Cl.05Eb BY S16ML q

SYSTEM IsolkTI0d i' i

4. Itcic sb LM b Prr.ssvg - 49h 6 (HV-F007, HV-Fo76 HV Fool) l f

b. ReCC Sb um b Pragart.- Rif,hv 6 ( BV-Fdo?, HV-Fo7G, 4V-Food) 4 l

l f

C

c. RCit Skam Supp y regure,- Loss

d. Rtic TurL,u. E.xLset 6 ( HV-Foo?, HV-Fo76 j4v-Fet) 1 w

bigp(rapn Pressort. - H 74 M

d. RC)t Fm floom %eperdvrt, -ifi A 6 (Hv-Foo?, SV-F076, RV -Fa7t) t

f. ItCIC Pump Room'(edI/dion bulx G (H-V-Foo?, HV-fo76, HV-Foot) b Tt, fav0vtt - $1 h 9

i 3, R CI C 'ft/ c d a g j Arm 6 ( Hy-F007, Hv-F076, HV-Foof)

.,j

~

% w p u d ort. < N / 6 l

7 h, %s Am T4caiwe -. Hif

& (MV-f'co?,h-fo16,HVf**!)

i g. brydl Pressu,<. - 4i;6 6 (HV-F062, HV-Fo&V )

.] Z hwod Tnilidien 6

('HV-POOP) i

~

s TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREMENTS

.O e

1!S CHANNEL OPERATIONAL i 5 CHANNEL FUNCTIONAL

' CHANNEL CONDITIONS FOR 141ICH

g i

TRIP FUNCTION CHECK TEST CALIBRATION SURVEllt!ANCE REQUIRED it i

5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line A Pressure -H_gh N6 M

R I, 2., 3 t

b, Reit Sb L.1,4. h Pvessage. 4%,Timy M A.

m

((

1, ~1, J

Cji, RCIC Steam Supply MA M

M 1,23 f

Pressure - Low-4K.

RCIC Turbine Exhaust NA M

Ie % 3 Olaphragm Pressure - High 4

[

.tyr.

RCIC E M ment Room M

f6 I, 2,3 Temperature - High Pom f)6 RCIC 5 ;!;e---t Room Vinhklion hdi NA-M G.

I,2.,3 A Temperature - High o

y ej %

RCIC. -

E".t.; Area NA-M G.

L,2,3 I

s Temperature - High 1

brygg,() Pvn;9..n.. Hi3,

w 5

m kco t 2., 3

.1. Xy, p r Q - g 3.2.

SC M.

4CIC !=,.: :;.t r f rir Temperet e - l:16

j. [

Manual. Initiation NA M*

NF 1, 2, 3 h.

> Aren Tenyerstove.-Ihy b jw}

M q

1, 2., 3

,v

/

3 i,

4 l

pg TABLE 4.3.2.1-1 k

f lk ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIRENENTS

.s CHANNEL OPERATIONAL l 'm j

CHANNEL FUNCTIONAL

CHANNEL CONDITIONS FOR idHICH 2

'l d;

TRIP FINICTION.

CHECK TEST CALIBRATION SURVEILLANCE REQUIRE 0 I

j 7.

RHR SYSTEM 58R1T01301 COOLING BRIDE ISOLATION-a.

Reactor Vessel Water 5

M R'O l, 2., 3

)i Level - Lou, Level 3

:

Manuel Initiatten 9A m

yg i,.z., 3 g

c o

4 s

\\

O' f

w;; _ w Ra nu+.- Lillias to 8 Dheii'lia~nillTng irradiated fuel in thehecondary containment)and during CORE ALTERATIONS and operations M

CJ1 with a potential for draining the reactor vessel.

N

When ag %bint dog %\\vt igreder Abn (Wo open ang/oy gn 41t key-locks # bypss spkh a in ec. Noem pos dmt, (a) Manual initiation switches shall be tested at least once per 18 months during shutdown. All other circuitry *N 1
associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at. least once per 31 days as p. art of c'ircuitry required to be tested for automatic system istlation.
l'
.(b) Each train or logigcpi shall be tested at least every other 31 days.
(O Calibmh hip Uw8M leabponcs, per 31 days.
j l
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u j j f rW6D')
INSTRUMENTATION ~ ~
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATI 3/4.3.3 LIMITING CONDITION FOR OPERATION The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERA 8LE with their trip setpoints 3.3.3 set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and.with ENERGENCY CORE COOLING SYSTEM RESPONSE TIM APPLICA8ILITY_: As shown in Table 3.3.3-1.
ACTION:
With an ECCS actuation instrumentation channel trip setpoint less '
t conservative than the value shown in the Allowable Values colum a.
Table 3.3.3-2, declare the channel inoperable untti the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
f With one or more ECCS actuation instrumentation channels inoperable, b.
taite the ACTION required by Table 3.3.3-1.l'"' t'~fe'~f*rs "' **
f o"n s sto e kh level M3fn3 with k condensafeinstrun,entst,w ingetesbis%veetly t%t fle.sensiny lint hed traerny is eneyer/
c.
y leurs,.alene* *r tAe outsids nie tenyerntsee. Is less tism er
<+ /sont e,,ee,,er-
I"#
i SURVEILLANCE REQUIREMENTS Each ECCS actuation instrumentation channel shall be demon OPERABLE by the performance of the CNANNEL CHECK, CHANNEL FUNC 4.3.3.1 1
CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and frequencies shown in Table 4.3.3.1-1.
LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation
)
4.3.3.2 all channels shall be performed at least once per 18 months.
The ECCS RESPONSE TIME of each ECCS trip function sho Each l
shall be demonstrated to be within the limit at least onc 4.3.3.3 l.
are tested at least once every N times 18 months where N is the total number I
of redundant channels in a specific ECCS trip system.
habs~e,khm to,stau</ n 4.3.3,+ The Aest trac;ny ad femperntwe nuni toe,,,3/cre/.sensiny kne shtl fe
+4e
e. /r,, sate s+ysy e.
bnk A ge der,,e,,rMef b he ofer*Lle at least once. pee yn,- 7,,.o r.
onse7 of -Nect/g won b.
I l
\\
K gev, f l
%ees. ettsu 3/4 3-9
--- '~
.,in,,
7 a..
s.~
- - \\
~
TABLE 4.3.3.1-1 (Continued) g,f E8ERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRENENis r
i CHAIRIEL OPERATIONAL g{
CHANNEL FUNCTIONAL CHANNEL CON 0lil0NS FOR WHICH
' 4 TRIP FUNCTION CHECK _
TEST CALIBRATION SURVEILLANCE REQUIRED
;'m 4
I 4.
AUTOMATIC DEPRESSURIZATION SYSTEN -
1 ReactorVesselWaterLevel-h*q l
a.
Low Low Low, Level 1 5
M R
1, 2, 3 i M i
4R 1, 2, 3 b.
Drywell Pressure - High 151 c.
ADS Timer NA M
, i,i,
Q 1, 2, 3 1
d.
Core Spray Pump Olscharge Pressure - High (57 M
(R1 1, 2, 3 j'
e.
RHR LPCI Mode Pump Discharge j
l Pressure - High
$5[
M JKJ 1, 2, 3 4
J f.
Reactor Vessel Water Level - Low, 1, 2, 3 Y
Level 3 S
M R
/M444 NA 1, 2, 3 I
l
[
Manual Initiation MA hO5.yLdSS PN 5
856 [50hk N.
N.
f[
a.
4.16 kw Emergency Bus Under-l p
voltage (Loss of Voltage)
NA NA R
1, 2, 3, 4**, 5**
5 b.
4.16 kw Emergency Sus Under-l
.g.
voltage (Degraded Voltage) 5 M
R 1, 2, 3, 4**, 5**
8
)
All other' l
i
$(f) Manual initiation switches shall be tested at least once per 18 months during shutdown.
i circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL, TEST at least once mar 31 daw an==rt of circuitry reeutred to be tested for automatic system actuattan 2
~
l When the system is required to be OPERABLE per Specification 3.f 2.
Required OPERA 8LE when ESF equipment is required to be OPERABLE.
Not required to be OPERABL whenpeactorsteamdonepressureislessthanorequalto$100)psig.'l l
A M Cohbrde. hp udh,{T e$d once pen-1 3\\
s.
I 9
I l
as E
W i
m 5
I?
t E *z 3
$s 13 Ia W
q a
[
wa
=
5 B-b 4
S m
t l
1 g
~
3 7
0; I
5 k
~
5,,
W3 3
2
-. 5 1.
]
II v
g a
8 5
5')
E l-1 0
l E.
T 8
14
=
Tt
'6" 3-41 36 2
- B -l E
A R ggs N
y
~
I 3
xn j"g'j St
-1
'g 5
si uf 9
3-3 4
s
~
g x
x la a
a e
n 2
3 a
y 4
4S m
ll n
0 OT C' ?./0) l 3/4 3-N e Rev. C HOPE CAEEA l.
L
M
.g t
E.
8-
'A 2
y s.
E w*
ss v
to Al vl E
E N
5;"
g B
2
=
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=
a u
6 ay e
1 E
ag
:s a
W!
=-
Y e-Al v3 2
h E
i
.t I
5 s
i L
J i:
5 i
7
'c f
5 g
5
=
g
.s a
M$
~b 3
,1, c
=
-J ja I 5
,i t
3 i h"
. I TF
. S. s
.i R g' &
%H t+
5 sa u
u1 3g 3
3 =i e
4 M
Ws W
43 3 ]L t
4 j
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=
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u 4G jbeecuGA 3/4 3-see gy,g l
__ 07: 'rn'O
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O v
TABLE 4.3.4.1-1 ATWS RECIRCULATICII Ptse TRIP ACTUATICII INSTRtBIENTATION SURVEILLANCE REQUIREENTS
on T
CHAIRIEL CHA000EL FUNCTIONAL CHAIRIEL TRIP FLBICTIGII CHECK TEST CALIRRATION i
M' WM 1.
Reacter Wessel Water 1p1 -
S M
R*
Low Low, Leve112
(*XbXc) 3.f.
Reactor Vessel Pressure - High
$7 M
1RI.*
l.
.t.
Re.d.r Vessel Wafec' Level -
riA M
g Lev el 4 Tt=* e y
trip sef *d f
t b5p h5 ;d Ifad oweg, per Md w
g ks (a) This instrumentation activates logic which provides a 25 second time delay after which, if sufficient power reduction has not occurred, feedwater runback is initiated for the RRCS.
(b) This instrumentation activates logic which provides a time delay of 5312 seconds after which, if sufficient power reduction has not occurred, automatic injection of SLCS is actuated and the inboard / outboard RWCU valves are isolated for the RRCS.
(c) This instrumentation initiates the Alternate Rod Insertion (ARI) function of the RRCS.
n j
O
g TABLE 4.3.5.1-1 o
$N.
REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRMNTATION SU h,,
CHANNEL
! i
CHANNEL FUNCT10NAL CHANNEL "f
3 CHECK TEST CALIBRATION i
FUNCTIONAL UNITS ReactorVesselhterLevel-S M
R a.
(Low Low, Level 2) b l
5 M
b.
Reactor Vessel Water Level - High, Level (8) i c.
Condensate Storage Tank Level - Low TSI M
M d.
^::::*r 7 :12:t= L::? -
5
Id?
(O M
IU s.,
l d g. Manuel Initiation MA p-(Q NA All ot l
(a) Manual initiation switches shall be tested at least once per 18 months during shutdown.
circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once LdayLasJrt of circuitry..pired to be tested for automatic system actuation.1
' - - ~
l l
r
% Cdhde 46p uwfjgTfd oce,er u 4.p,.
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l 10 J AN 1983 r
REACTOR COOLANT SYSTEM RECIRCULATION PtMPS
.s..-
~ s LIMITING CONDITION FOR OPERATION l
Recirculation pump speed shall be maintained within:
3.4.1.3 55 of each other with core flow greater than or equal to 70% of l
a.
rated core flow. * *
105 of each other with core flow less thin 705 of rated core flow.
ll b.
l APPLICABILITY: OPERATIONAL C0tCITIONS 18 and 2*.
ACTION:
With the recirculation pump speeds different by more' than the specified limits,. either:
Restore the recirculation pump speeds to within the specified limit a.
within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or Declare the recirculation loop of the pump with the slower speed not l
in operation and take the ACTION required by Specification 3.4.1.1.
b.
SURVEILLANCE REQUIREMENTS Recirculation pump speed shall be verified to be within the limits ll 4.4.1.3 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
c l
=see special Test Excepti'on 3.10.4.
4 ge,,,pe,4)

u k s sfer+s9 e idle rectrealdGs p mf, tAtt reg a;renant how folkwiny hile p mp res hr t, he en s be suspenleal fer up
+o six(G)4.ues f.,. +Ls eahirJten

T %is reyin~mf mnyset mecAanical and efec+ria l 5+ar$-
of tAs Mce C'
h. 0 M D.,
3/44-g
/ev.
5"~
r. -.,.
-r,
.-n,-,------,,------,--....--.,-..-,,,,--,------n--,,,-
REACTOR COOLANT SYSTEN 3/4.4.2 SAFETY / RELIEF VALVES SAFETY / RELIEF VALVES LINITING CONDITION FOR OPERATION 13 j
(At t.:t (tu ) n : ^:r r -? ;' y:'
rf: n'-ty
!=: rf) ghe l
~
safety valve function of at least Mo:f the following1 reactor coolant 3.4.2.1 system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:"
180B
.. _,.. _ _., M, s _ _, _..
f m
)M,
"."Z!' ~4. D... '.._1..**"1H.. ; W..,.4..,. w 4 IS) safety-relief valves 9 M sia 71E ng 5(H safety-relief valves 9 f449H psig 1%
S tS) safety-relief valves G f400H psig 71%
nao APPL'ICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
~
inset?
~
A*
ACTION:
Lu]
d m:-* r :--?:-t :""- :f: M'
h n r.:2) the saisi,('5': 'r ;.;1"- functwW"rMthe above a T.
W11.n '
y 13 ~ required safetyf_re14f ::Mi-T15 iip ~erab e, v. r s' 1==<t HOT SHUTDOWN i
.q, iz nours and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
I
= ":ty :
= :r) safety / relief valves stuck open./ "O With one or more (r d:
provided that suppression pool average water temperature is less than feQ'F, close the stuck open (rf: =":ty "in ri'Or) safety relief valve (s);
if unable to close the stuck open valve (s) within 2. minutes or if sup-1 pression pool average water temperature is (90'F or greater, place the,
i reactor mode switch in the Shutdown position. Ho h.
W tp one or more saJty/ relief valve (^ ; inoperable (:x;t(n - f '; ; t :r : :r't;tn) 0 poustic rsrinoperable, restore the l . O onitor(s OPERABLE status within 7 days or be in at least NOT SHUTDOWN within the next 12 hours ano in COLD SHUTDOWN within the i following 24 hours. I" _ hnS n-salaix lnnaf hacHe o f gree. er more,+ $ -r,u,+u.,, ASve luted emlves infernbic,' je in sf' / gest h>T* SM7y g/C /Z Aeurs a,,/ in Cela Surpul wlfAin tie e. auf 24 kotrS. l psen.T 6 "The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. ,~-9 r.-. 3/4 4 ' y ~~ ^ 1 -,.-..,e ,,.,,,,,,, _,. ~ _, _ _ _..,... -,. _.. _ - _ _, _. _ _ _ _.. -. -
INSERT A TO PG. 3/4 4-7: I a. With the safety / relief function of two of the 14 above listed valves inoperable, restore at least one inoperable valve to an OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. INSERT B TO PG. 3/4 4-7:
SRVs which perform an ADS function must also satisfy the OPERABILITY requirements of Technical Specification 3/4.5.1, ECCS - OPERATING.
sav, wA:cA perform a.
l.w - l e w set (. ct;.n mu t also sabsfy the oesp alt.lTY repirements of Technical Speciftedt ~ 3/4.1. 2. a. SertTY/AtLler vns.vu Low-Lo ser faNc.roen. e Ae v. C
c REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The =11:' =?= *=dir =d th low-low set function of the following l reactor coolant system safety / relief valves shall be OPERABLE with the following settings: ~ Low-Low Set Function lief Function Setooint" (psio) t 1% Se oint* (osic) 1% Valve No. Open Close Do CTose IOl? 90$ /~o/3 # (4000) (405) F o /3 P /8W7 (4096) (406)9E~ 441993-49463- / \\ / 4 tits) (%~J / \\ ^ -(&tt9) -(046) / N\\ APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. s ACTION: j a. With the low-low set function of one of l, the abo e required reactor coolant system safety / relief valves inoperable, v g~ restore the inoperable -nli:' =?r "r tf r M lot-low set function l to OPERABLE status within 14 days or be in at least NOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. With S: =11:' =1= ";rdir ri'r the low-low set function of espeds//. -4 hen-one-of the above required reactor coolant system safety / relief valves l b. inoperable, be in at least HDT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24' hours. SURVEILLANCE REQUIREMENTS 4.4.2.2.1 The nii ' =1= ";;di= =d S; low-low set function pressure l actuation instrumentation shall be demonstrated OPERABLE by performance of a: fnp retpean+, l s. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit at least 3 once per 31 days. \\ b. CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least once per 18 months. t l "The lift setting pressure shall correspond to ambient conditions of the (. valves at nominal operating temperatures and pressures. Now essex 9 Z-STS (5 "/4) 3/4 4 Y /78v. # ..n.--n, --,,,,.n~ .n, ,,,m,--
l CONTAlletENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHApSER A LINITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERA 8LE with: a. The pool water: ll8 ooo 3 tihoco 3 1. Volume between (*j 5^^} ft and (20,000) ft, equivalent to a i .indiccded level between (2? ,and-(?f g,and a ^" 2. Maximum average temporar.ure of $95P F during OPERATIONAL ) CONDITION 1 or 2, except that the maximum average temperature 1 may be pemitted to increase to: a) 1105YF during testing which adds heat to the suppression i chamber. I b) (110FF irith THERMAL POWER less than or equal to {1}% of 1 RATED THERMAL POWER. t c) (120l*F with the main steam line isolation valves closed following a scram. b. (0;;.:ll-te :gr;;sien et --t:r types; leekege les; then er eqeel-to-10" af the ecceptable A/,T n; iga vehe ;f (0.00) ft'.) (A total leakage between the suppression chamber and drywell of. lass than the.* equivalent leakage throug$M)Oiinch diameter orifice at a ha differential pressure of psig.T j APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: a. With the suppression chamber water level outside the above limits, restore the water level to within the limits within 1 hour or be in t least HOT SHUTDOWN within the next 12 hours and in COLD SHUTD0 ,in the following 24 hours. Q w b. In OPE IONAL CONDITION 1 or 2 with the suppression chambe verage water temp ature greater than(95PF, restore the aver temperature l /gT to less than equal to {951*F within 24 hours or be n at least HOT SHUTDOWN wit the next 12 hours and in COLD DOWN within pgWP the following 24 ho except, as permitted ab e: 1. With the suppressio haaber average er temperature greater l than (105PF during te ng which s heat to the suppression rchamber, stop all testing ich dds heat to the suppression chamber and restore the aver temperature to less than {95)*F within 24 hours or be in less T SHUTDOWN within the next 72 hours and in COLD 5 withi he following 24 hours. 2. With the suppress chamber average wat temperature greater than: a) 19 or more than 24 hours and THERMAL R greater than 4 of RATED THRMAL POWER, be in at least H0 TDOWN thin 12 hours and in COLD SHUTDOWN within the n 24 hours. F, place the reactor mode switch in the Shutdown I {110$*ionandoperateatleastoneresidualheatremoval1 posit in the suppression pool cooling mode. ~ NorE EMEk l -GE-STS-(9WR/4P 3/4 6-45 2.* p,,9,y_ w w -v--w,m--w-~w--w--_~wr=mw-+v,ww~,r -w-,----ww,w- ,---~,=---,w,-
IAIMRT To Pde. 3/Y G -20 ACTIOM: ' a. With the suppression chamber water level outside the above limits restore the water level to within the limits within 1 hour or be k - at least HOT SHUTDOWN within the next 12 hours and in COLD 5% / within the following 24 hours, 95 b. In OPERATIONAL CONDITION 1 or 2 with th pression chamber aveq water temperature greater than or equa F, restore the.am temperature to less than or equal to F within 24 hours or be t at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTm[ within the following 24 hours, except, as permitted above: 1. With the suppression chamber average water temperature greats, than 105'F during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than or equal to F within 24 hours or be in at least HOT SHUTDOS within the next 12 hours and in COLD SHUTDOWN within the folf 24 hours. 2. With the suppression chamber average water temperature greats than 110*F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop ' the suppression pool cooling mode. 3. With the suppression chamber average water temperature greate than 120*F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours. + s.c.s -* +- 3/* s > r., secs c.7 -n. r = t Aev. f ~
10 JAN 1983 CONTAINMENT SYSTEMS 7 ,UPPRE5510N POOL W. 2^.' T_'.! SPRAY LIMITING CONDITION FOR OPERATION The suppression pool (:r drf:11) spray mode of the residual heat consisti{ngo: system shall be OPERA 8LE with two independ 3.6.2.2 removal RHR I J0ne'(Tu:) OPERABLE RHR pump (+h and a. An OPERABLE flow path capable of recirculating water from the b. suppression chamber through an RHR6W-heat exchanger and the suppression pool,( ::d dr/_11)-spray spargerfe-b APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. f ACTION: With one suppression pool (rd'er d,tell)-spray loop inoperable, J restore the inoperable loop to 0PERA8tE status within (72 5::r:) s. 17 days [or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. With both suppression pool (rd'Or d,S !!) spray loops inoperable, t (restore at least one loop to OPERABLE status within 8 hours cry be b. in at least HOT SHUTDOWN within the nextl 12 hours and in COLD SHUTDOWN
within the (Olhub;)((nextj 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool (: d dr,1:11) spray mode of the RHR system shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve, manual, power j operated or automatic, in the flow path that is not locked, sealed a. or otherwise secured in position, is in its correct position. By verify.ing that each of the required RHR pumps develops a flow of gpaonrecirculationflow{throughtheRHRheat b. foo "at least'- and suppression pool spray sparger when tested pursuant exchange to Specification 4.0.5. At is. 4 ~** ner 18 months % pefformance of a s stem f"eth.. e system turn;;5-"t I fc. which includes -f:::...." - ""'- ar+""S. vi J. M:r;--~ -----t':; _ _M and ver s",jk; 'hmt each ::t:rt' e valve it: x h ;q:":: t:-its M position.) 1? :;r:y ?; ; r':rrer " 55 5- er erh *!r t :t Of th: dr;: (d.==1:: :t 1:::t :::: ;:r 5 ::r: :n' =r"yS; th:t =:h ;r:y r.,,;;h h =:htr=t:d.) i "Whenever both RHR subsystems are inoperable, if.unabic to attain COLD $HUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. HokcGek 374 g.gg-D geu g .-.... s.~,-s s
SUPERSCRIPT NOTES FOR PAGE 3/4-8-f' (1) Test is per IEEE 308-1974, Section 6.4, to demonstrate that the Class lE loads can operate on the preferred power supply. (2) IEEE 308-1974, Section 6.4, requires demonstration that the standby power supply is independent of the preferred power ~ supply. The surveillance requirement for verifying that the diesel generator circuit breaker is open after the diesel generator had been synchronized-with the offsite power source, transferred its leads to that source and re-stored to its standby status provides this demonstration. 9 = 9 N 0 hv. c \\ t
Il0 JAN 283 ELECTRICAL POWER SYSTEMS ~ SURVEILLANCE REQUIREMENTS (Continued) N'M. Verifying that the following diesel nerator. lockout features,wkm 4eded, prevent diesel generator starting * - p" "- ---".'--^ l '( b.in ;;; ; }--d.)-
)
m (R. ;;;;; :t-. 4) ^ /#. At least once per 10 years or after any modifications ich could [l affect diesel generator interdependence by starting Mith diesel generators simultaneously, during shutdown, and verifying that bothall diesel generators accelerate to at least (906).rps in less than l 84 or equal to GB)t* seconds. j p: At least once per 10 years by: ll 1. Draining each fuel oil storage tank, removing the accumulated sedimentandcleaningthetankusinga{sodiumhypochlorite) solution, and 2. Performing a pressure test of those portions of the diesel fuel
oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section 11 Article IWD-5000.
l 4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.1. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1. August 1977. If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is greater than or equal to 7, the report shall be supplemented to include the additional information reconmended in Regulatory Position C.3.b of l Regulatory Guide 1.108, Revision 1, August.1977. ' ah ng., eoec37eeA, perder dbred% =A low 6be. e on p es.s.wre, C re p r tec.kod rek,M se,Q, h 3 Q1!5ch7 geweredor derenM. )mA gemedor cuer (w&.9 w b d reb,M st.s. 3 4 4 he1 1 ud \\ecked vehp-reg ce,, e } eh Cmer Ao. bee 69 ( re %,m e m. m ..\\, ..=...t W l
)
RADI0 ACTIVE EFFLUENTS 2 :5 = d t; tit: tzd : 4 : - W ?d ?") EXPL0SIVE 4AS MIXTURE ' y;t: : 6 Appropriate alternatives to the ACTIONS below can be accepted if they provide incentive for timely repair of monitors and for compliance with GDC 3 ff4,e fi;^:rf r), LIMITING CONDITION FOR OPERATION 3.11.2.6 The concentration of hydrogen er-enygen in the main condenser offgas treatment system shall be limited to less than or equal to 45 by volume. APPLICA8ILITY: At all times. ACTION: s. With the concentration of hydrogen ee-emygen in the main condenser offgas treatment systaa exceeding the limit, restore the concentration to within the limit within 48 hours. t. "ith ;= tin;;;;
c. iter; ic.;;;..;,1;,.;;;;;; g. 3 ; ;;;r.; i.;;; ire;
':r : ;;;i:2 =
==.: 2 L ;.
be The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.6 The concentration of hydrogen es-emygen in the main condenser offgas treatment system shall be determined to be within the above limits by con-tinuously monitoring the waste gases in the main condenser offgas treatment systashmith the hydrogen openygen monitors required OPERABLE by Table 3.3.7.12-1 of Specification 3.3.7.12. Cwh m = u. 4. , e...vatr s,, % 7 :.,,,. Q n l t .pera en e ew..c -#R-STS-t-- 3/4 u-1T N J n-, a ~. - - -,.

. -.-...n...

-,,---.nn__-,. ,-----e
o INSERT TO PAGE B3/4 1-4 .... and other piping systems connected to the reactor vessel. To allow potential leakage and imperfect mixing, this concentration is increased by 25%. The required concentration is a hieved by having a minimum available quantity of 4853 gallons of sodium-pentaborate solution containing a minimum of 5750 lbs. of sodium-pentaborate. This quantity of solution is a net amount which is above the pump suction, thus allowing for the portion which cannot be injected. The pumping rate of 41.2 gpm per pump provides a negative reactivity insertion rate over the permissable pentaborate solution volume range, which adequately compensates for the positive reactivity effects due to temperature and Xenon during shutdowri. The temperature require-ment is necessary to ensure that the sodium-pentaborate remains in solution. e-O Rev. 5
10 JAN W 3/4.2 POWER DISTRIBUTION LIMITS l BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. j 3/4.2.1 AVERAGE PLANAR' LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times (LOJS is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification i AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in Figures /3.2.1-1,f,2,.'.-2 j.g2.],q:,),._ % g 3,,,, _ f,
\\
s.u-a 3.n.)-si s.tJ-y e d s.z.t.c j The calcuIstional procedure used to establish the APLHGR shown on Figures A.2.'-2 :d 3.2.P3) is based on a loss-of-coolant accident analysis. l 13.2.1-1 2 Ine analysis was performed using General Electric (GE) calculational models A which are consistent with the requirements' of Appendix K to 10 CFR 50. complete discussion of each code eeployed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses can be broken down as follows. a. Input Chances 1. Corrected Vaporization Calculation - Coefficients in the vaporizat-ion correlation used in the REFL000 code were corrected. 2: Incorporated more accurate bypass areas - The bypass areas in the t$ guide were recalculated using a more accurate technique. 3. Corrected guide tube thermal resistance. 4. Correct heat capacity of reactor internals heat ' nodes. 9 e b6 T ) B 3/4 2-1
n Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters: 3V30 Core THERMAL P0WER..................... (00:0) Mwt* which corresponds to4105]Eofratedsteamflow q Vessel Steam Output. ................. (1. s e.10; x les lha/hr which l corresponds to (105)% of rated steam flow VesselSteamDomePressure...y........(1055} psia Design Basis Recirculation Line Break Area for: .l l ~ a. Large Breaks (4.11fts l b. Small Breaks M ft2 o.oq Fuel Parameters: PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT. AXIAL CRITICAL FUEL BUNOLE GENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kw/ft) FACTOR RATIO Initial Core {8x85 $13.47 41.47 ftrt95
f. Ao A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection '1".
F of the FSAR. G.s.1
This power level meets the Appendix K requirement of 102%. The core heatup l
calculation assumes a bundle power consistent with operation of the highest powered rod at 3102)E of its Technical Specification LINEAR HEAT GENERATION RATE limit. l Hoes assw_ - : ;T (;U- '0; 8 3/4 2-3 8ev. f
O 4H rese.4er/re4hr valves cperale A !es a r<dalded MS c"ni 60 pecvia The sdeh coobd system Ae b<lrs p:ssvriscd skova a. de& allowdle y,ja of 13 ?5*ps,g in assordam ee. will 1L Ae AshfEC,ede. G. folal of /3 OPERABLE CCEANT SYSTD' #"f*bl/##/lff V"I '#af uA ve c. penents to w?"k.= ASME B aHeaabte dedy V IPU$ UC IIlI /m/ 3/4.4 REACTC: Pre al BASES Y". ~ 1 .a ara ano.
A? *0*r *PPr0P'!* h ASaf M r
- - a g ;, y ;; q tt(--.ent. 3/4.4.1 RECN.CULATIG" SYSTEM Operation with one reactor core coolant recirculation loop'inoperatie is prenibited until an evaluation of the performance of the ECCS du-ing are i::: operatfor nas teer performed, evaluated and determined te Le a::eptacle. msenT B '+An in:pe able,).st pump 1s not, in itself, a sufficient rassen tc ca:iare a recirculatica lo6p inoperable, but it does, in case of a cerig.-basis-a::ident, increase the blowdc n area and reduce the capability of refic: ding the co a; thus, the rec;irer. ant for shutdown of the facility with a jet pu : irece i.le. Jet pum failure ca.. te detected by ::nitoring jat pump pe-fer.ma. e en a pres:ribed senedule for significant degradation. Recircula-ion : :: steed misestch limits are in cc :liance with the ECCS LOCA analy:is cesig. cri.s-ia. (ine limits will ensure a a:e:;uate : re ih l 1 coas.c:wn fr:m either recircuiation io:: foll: wing a LOCA.4 ~ In crcer tc prevant undua stress on ths sessel nc::les and botter nead - region, tne re:irculatic. locp ter:eratures shall be within J50) F of ea:- other prier to star..p of an idle icc. The loc; terperature cust also be within - $501 F of the reactor pr:sse e vessel c:olant temperature to prevent tht-al shock to tne recirculation pump and recirculatien nc::les. Since the cc:lant in tre 00-t: of -he vessel is at a icwer tercerature than tne ccolant in e e :er regions o' tne core, un:.:e stress on the vessel w:al: result if the temp &rature difference was greater inan U.00*F. Ms 3/4.4.2 5AFETY/ RELIEF YALVES (T5:.:.^g:ter :::-lan cy ter ::de ::'ety valve: 2nd) the safety v 1 function of tiesW valves operate to prevent E oor coolant . 1. of (1:25)*psig in~ lM f system frem being pressurized a = a of w, ~ {:sfety :nd/O-) l accordance wi.h the AS".E C safety-relief s re:;uired to limit reactor pressure ~ ~ SME III. 211___.: e:lu:: for th: acr:;t case upset trar.sier... Demonstra-[onofthe{::: ::fety v:lv: 2nd) safety relief valve lift settings wi-11 occur only during shutdown and will be performed in accoroance mis ca.T-with the provisiems of Specification 4.0.5. A--y ^: leu-ica ::t sy::: ensure: thet ::f:t-yhl4:f vah: disch: g:: 2-^ minimized. econd opening of these valves, follo ing any sure trar.si ent. T.is is e.. ad by autonatically lower' - .- c osing, set ain. cf .) valves following tna initial (7) valves and lowering the one.. eto '
opening.
In tnis way, the fre~ n.%.sgnttude of tne containment b h down duty cy:1e is subst ..j reduced. Sufficient remmQncy is provice: fer + the low-1~ .. system su:n t at failure of any One valve to - or cicse at it.. : ::: :-.pcint d:::.: "ic!sted th: denigr tatic. M ce ,. ere...a. n , /,v L J 5F Rev.S
a INSERT A TO PAGE B3/4 4-1: The low-low set system ensures that safety / relief valve discharges are minimized for a second opening of these valves, following any overpressure transient. This is achieved by automatically lowering the closing setpoint of two valves and lowering the opening setpoint of two valves following the initial opening. In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis. REV. S
= f 6.5 ggyggw An AW IT STATION OPERATIOMS REVIEW COMMITTEE (SORC) 6.5.1 FUNCTION 6.5.1.1 The Station Operations Review Committee shall function to advise the General Manager - Hope Creek Operations on operational matters related to nuclear safety. COMPOSITION 6.5.12 The Station Operations Review Committee (SORC) shall be composed oft Gen.es/ Manger -Hop Creek. Operations Chairmans-: Assistant General Manager - g.,,,f,e.,,/ vtce. cA.wm., Hope Creek Operations Member and Vice Chnirman Operations Manager Member and Vice Chairman: Technical Manager Member!: f Tic; M ir.zz., Maintenance Manager Member Operating Engineer Member I & C Engineer Member Senior Nuclear Shift Supervisor Member: Technical Engineer Member: Maintenance Engineer Members -Radiation Protection Pfansitr-Member Chemistry Engineer Member: Manager - On Site Safety Review Group or his designee. ALTSRNATES 6.5.1.3 All alternate members shall be appointed in writing by the SORC Chairman. l a. Vice Chairmen shall be members of Station management. b. No more than two alternates te members shall participate as voting mem% rs in SORC activities at any one meeting. c. Alternate appointees will only represent their respective department. d. Alternates for members will not make up i part of the voting quorum when the member the alternate represents is also present. i l l RwS I Neer eacwie -NRM 1/02 4-
? R ADMIN!$TRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS 5.8.1 Written procedures shall be established, implemented, and maintained [ covering the activities referenced below: a. The applicable precedures recommended in Appendix A of Regulatory ll Guide 1.33, Revision 2 February 1978. NUREG-0737.The applicable procedures required to implement the requirem l l c. Refueling operations. d. Surveillance'and test activities of safety related equipment. I e. Security Plan implementation. (' f. Emergency P1an tuplementation. l Joserf Fire Proteption Program implementat. ion. l g.
A" -
> s see resort *B" (T.S.27:nt ;c.nd_re ef*!;n'f t;;tte 3.3.1, nf ct r.g.; ti .et., 2._;; ;,, 9 ' ^ 7d ir_:d by O.: '""""'s: :h;11-Le ;;re.e4 4, tr.e g %..^. 4 _,,....;... J,,-;.. i te 1 ;1:_ :c.".:tionfand redewed-peciedfeeily- = ;e $ er;t, ',,.r bie^..^,0.e p.. " N g m =_ ' /---- I e- ^ ^ ^ h *,'
M'*'
l r ~r-n a -- T(- =-/' changes to procedures of Specifjestion 5.2.1 may be nede pr] l 6.8.3 g g.pt vided: a. The intant of the original procedure is not altered; l ~ b. The change is approved by two members of the unit management staff, at least one of whom holds a Senior Operator license on the unit affected; and,C,, tp,,,ep f $ee Insert f- ~ -4 :- 4: in
:, = :" :_z r; L4 %,",...:
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l 6.8.4 The following programs shall be established, implemented, and asintained: l l a. Primary Coolant Sources Outside Containment A program te reduce leakage free these portions of systees outside j confainment that could contain highly radioactive fluids during a l serious transient er accident to as low as practical levels. The l l ' systys include the '""CI, 00, ^^::, "CIC, hy n;:r rn:-if re,
n=

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t,nf O'z-2. ::: t n.'_ _.;"':y: - :
.f l The program shall include the following: 1. Preventive maintenance and periodic visual inspection requirements, and l
2. A ::"z;n";f leak test x _-t:rnt; for each system at refueling cycle intervals er less.
etw on.*seld:n-tenhey res;/a4 hee,f ree. oaf,+cere l,y,3,/f; &ressenre,,cookd in tekj,TJ as en. analyse ,gone a n g,,,
-sui,(ent sam e
sisay s perk)pt,s le,nes.:,, t,e afran *s (feraan sy Ae v..S.
g. g Hott CMEK J
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ML20125E482

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Response

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