Text

Forwards Comments on Four Questions Used on Written Senior Reactor Operator Exam Administered on 890214
	ML20244B616
Person / Time
Site:	Hope Creek
Issue date:	04/07/1989
From:	Hagan J; Public Service Enterprise Group
To:	Russell W; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20244B599	List: IR 05000354/1989004; ML20244B596; ML20244B616;
References
	NTC-89-1043, NUDOCS 8904190291
	Download: ML20244B616 (22)
	v • d • e

. . _ _ _ _ _ _ _

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,. 2% .P ATTACHMENT 2 O PSIEG Public Service Electric and Gas Company P. O. Box A Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station l

l0 NTC-89-1043 I

L

. Mr. William Russell Regional Administrator U.S. Nuclear Regulatory Commission i Region 1 475 Allendale Road

-King of Prussia, PA' 19406

Dear Mr. Russell:

EXAMINATION REVIEW COMMENTS - HOPE CREEK LICENSE EXAMINATION Attached are comments on four questions used on the written SRO examination' administered at Hope Creek Generating Station on February 14, 1989. These comments have been developed following

'the examination. They are in addition to comments provided by Messrs. Sparks and Cirelly during the pre-examination review conducted at the' Region.1 office on February 10, 1989. The number of. post-examination comments is significantly reduced from previous examinations and indicative of a successful pre-exam review.

The-following format has been used to document specific comments:

A. NRC question, answer, and reference; B. facility comment including a recommendation for resolution; and C. support documentation.

These comments are presented in the same order as originally numbered on the SRO examinations.

8904190291 890407 PDR ADOCK 05000354 i V PDC

_!g-The Energy People j l

as a a em u en

.s. .

.W. Ruccclli

' If you have any questions, comments, or need additional

information;.please call L.'Catalfomo, (609) 339-3810.or W. Gott, (609)~339-3769. They:will provide the requested information-orf will see that you are contactedLby the appropriate person.

Sincerely, Joseph J. Hagan b General Manager -

Hope Creek Operations Attachment-C Mr.-Marion Daniels Sonalyst Inc.

215 Parkway North Waterford, CN 06385 BC General Manager - Nuclear Services Manager - Licensing & Regulation Manager - Nuclear. Training I' Operations Manager - Hope Creek Asst. Mgr. - Operations Training Prin. Trng.'Supv. - Hope Creek Ops.

Prin. Trng. Supv. - Sim. Maint. & Upgrade i

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W.-Russoll .

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-If you have any questions, comments, or.need add 3tional

?information; please call L. Catalfomo,.-(609) 339-3810 or W. Gott, (609)-339-3769. They will provide the. requested information or will see that'you are contacted by the appropriate person.

' Sincerely, l'

Joseph J. .Hagan General-Manager -

Hope Creek Operations Attachment C Mr. Marion Daniels Sonalyst Inc.

215 Parkway North Waterford,1 CN 06385-f 1

______.______._.._._.__.____.____________9

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/A.- QUESTION' 5.06 (3.00)-

For each' PLANT' CONDITION listed below, IDENTIFY the appropriateLPROCEDURE(s) to be utilized. If none, STATE NONE. (3.0)

. Consider each plant condition separately. ,

. NOTE: PROCEDURE may be used more*than once or NOT at all.

PLANT CONDITON- PROCEDURE A. Turbine trip from 10% power. 1. OP-EO.ZZ-103 Reactor Building Control B. Reactor water level -40 inches 2. OP-EO.ZZ-100 Reactor C. Suppression Pool Temperature 98 F Scram D. Drywell Temperature 130 F 3. OP-EO.ZZ-101.RI%T Control Z. Reactor level + 10 inches and 4. OP-EO.ZZ-102 Containment Reactor power 57%.

Control F. Reactor Building Diff Press OP-EO.ZZ-202 Emergency 0' inches water 5.

Depressurization

6. OP-EO.ZZs104 Radioactivity Relear.e-ANSWER 5.06 (3.00)

A. NONE B. 3 C. 4 D. NONE E. 3 F. 1 (6 at 0.5 each)

B. An additional answer to items B&E would be #2 (OP-EO.ZZ-100 Reactor j

f Scram). Both of these plant conditions are scram conditions and since reactor scram is the entry into OP-EO.ZZ-100, it should be added to the ANSWER KEY.

C. See'next page for supporting documentation.

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P8/T 6 RS/R-5 R8/L-7 OPERATOR CAUTIONS rs

' if slgnals of hlgh suppression pool

. IF WHILE EXECUTING THE FDLLOWING level (78.5 inches

HPCl only) or STEPS AN ENTRf CON 0lil0N FOR k.*[M.l./, $..j '

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CP-E0.22 101 ExtSTi, THEN EXIT THIS low condensate storage tank level (22.558 gallons

HPCI and RCIC)

PROCEDURE At(Q ENTER 0 5 .22 101 occur, confirm automatic transfer i L of or manually transfer HPCl l 5-4 and RCIC suction from the condensate ;

storage tank'to the suppression pool.

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j r *'i VERsFY THE SCRAM - 10 Do not secure or place an ECC5 in -

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Indications",I (1) misoperation l In automatic mode is confirmed ,or !

PLACE THE MODE SWITCH (2) adequate core cooling is assured.

IN 5HUTOOWN If an ECCS ls placed in manual mode ,

L it will not initiate automatically, Meke f requent checks of the initiating i

i 53 I parameter. When manual operation is INSERT THE 5RMs AND IRMs (SE) no icnger required, restore the system AND SELECT THE IRMs ON THEIR to automatic / standby mode If posslble. j L RESPECTIVE RECORDERS Do not operate HPCI or RCIC below 5-4 2150 RPM.

j Wg GENERATOR LOAD 15 APPROXIMATELY e T,KE,N, TRIP THE MAIN TUR8INE

, - i 5-5 VERIFY GENERATOR LOCKOUT (MA)

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l RESTORE AND MAINTAIN RPV WATER LEVEL BETWEEN

+12.5" AND +54" WITH ONE OR MORE OF THE FOLECWING SYSTEMS:

o CONDENSATE (AD) 0-720 PSIG e FEEDWATER (AE) 720-1200 PSIG ,

e CR0 (BF) 0 1500 PSIG 1 e RCIC (8D) 65 1250 PSIG I e HPCI (8J) 100-1250 PSIG e CORE $ PRAY (BE) 0-380 PSIG l L e LPCI (8C) 0 340 PSIG 57 CONTROL RPV PRES $URE BELOW 1037 PSIG

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REACTOR SCRAM OP-EO.22-099

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A ., . QUESTION 6.05 (2.00) I A. IDENTIFY the signal (s) that initiate (s) the <

End-of-Cyle Recirc Pump Trip.

(1.0)

B. IDENTIFY the SYSTEM that supplies the signal (s). _(0 5)

C. WHEN is this trip AUTOMATICALLY BYPASSED.

_ ( 0. 5)'

ANSWER 6.05 (2.00)

A. 1. Turbine stop valve closure

2. LOW EHC oil pressure-(TCV closure < 530 psig)

B. Reactor Protection System C. <30% as sensed by 1st stage turbine pressure (0.5 each)

REFERENCE LP _ 302HC:19 (019-04) book 3 page 57 OBJ 12 KA 212000K103 (3.6) 202001K413 (3.7/4.0) 212000K103 202001K413 ..(KA's)

B. An additional answer to item B should be considered. Although the sensors belong to the Reactor Protection System and sensors send signals, the SYSTEM which supplies the signal can well be considered to be 530 psig EHC oil pressure and 5% closure of the EHC Turbine Stop Valves. Electro Hydraulic Control (EHC) should be considered an acceptable alternate answer.

.C. References are attached and are pages 22, 35, and 36 of the EHC Control Oil Lesson Plan 302HC:50.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ I

i LES9ON NAME: EHC-CONTROL OIL

, INSTRUCTIONAL CONTENT:

L . KEY / AIDS valve. This action removes ETS from the disk dump and the shutoff valve. The-area above the disk dump valve is ported to FCD. The disk dump valve unseats (spring pressure) and valve operating springs rapidly close the turbine control valve (s). The shutoff valve is positioned to block FAS.

4-HM-0010 Note:eETSfluidtpressureds1 monitored %

cddition of .afterithe fast acting solenoidtvalveg4 icolation i(diskidumpivalvelcavity: pressure)Jfor valves for- each'ofLtheiturbine controlivalves'..d pressure lThese; pressure switchesilocatedion/the switches , control valve lcontroltpactinitiatel.a reactor.' scram,and'EOC;RPT(signals if

, pressureTdecreases"to'<530"psig.]

Obj 2 6) Disk Dump Valve a) With the exception of the turbine. bypass valves,.all turbine steam valve control pacs have disk dump valves. As the name implies, the disk dump valve is utilized to " dump" hydraulic fluid-(FAS) pressure from the hydraulic operating cylinder to allow valve operating springs to rapidly close the steam valves.

b) With the turbine trip system in a RESET condition, emergency trip system (ETS) fluid is available to the valve control pacs. The~ PTS fluid and spring pressure seat the di A dump valve to seal the hydraulic operator (below piston) from the FCD. This attion allows the FAS fluid to pressurize the hydraulic cylinder to effect valve operation.

c)= Any turbine trip signal will trip the main turbine hydraulic trip system and remove ETS fluid pressure. (Note that this action is not the same as energizing the fast acting solenoid valves.) The disk dump valves unseat and the steam valves rapidly close.

7) Solenoid Operated Test Valves

, Page 22 Date: 11/15/88 302HC:35. Rev.: 4 >

ILESSON NAMES- EHC CONTROL OIL INSTRUCTIONAL CONTENT:

KEY / AIDS Obj 4c Two pressure switches (PS-100 A,B) monitor !

emergency trip system (ETS) pressure. These pressure switches seal in any turbine trip signals. Additionally, if the ETS fluid pressure decreases.to 800 psig, a turbine trip signal.is initiated.. These pressure switches are located in the front standard (west side; dry pocket area).

Obj 4d c.. Turbine Control Valve Fast Closure Under conditions which indicate a generator load rejection, the EHC system energizes the fast acting solenoid valves on the turbine control valves to rapidly initiate valve closure. This action is necessary to prevent excessive turbine overspeed under conditions in which generator load is rapidly removed (power to load unbalance).

. Energizing the fast acting:solenoidivalves .

removes?ETS. pressure"from thetcontrolivalves/,8 The;ETS pressure-atsthe valvesEis monitored:byla*

Lpressure switch' located"onithe; valve' control ^ pac

!When this pressure decreases lto <.530 psig'.;a;

~

reactor: scram signal"islinitiated. A~one-?out-of two-twice logic must be.satisfi'ed'forithe_

scram-to occur. ' The" scramis initiated ascan /

~

" anticipatory" fmeasure'. 'That is, the scram is

, - initiated ; to ., reduce L reactorL poweri( flux ) 'in -

anticipation of the reactor pressure; increase andg i subsequent: power increaseL(void collapse)Twhich

~

results from'rapidTcontrol valve closure.~ This l- scramLsignal'is automaticallyfbypassed;if reactor I . power:is less than 304 as sensed by~ turbine'first stage' pres'sure. .

V. SYSTEM INTERRELATIONS A. Support System

1. Turbine Auxiliaries Cooling System (TACS)

The Turbine Auxiliaries Cooling System is required to maintain EHC fluid temperature during system operation.

2. Non - lE AC Power Page 35 Date: 11/15/88 302HC:35 Rev.: 4 n

EHC CONTROL OIL

' LESSON NAMES INSTRUCTIONAL CONTENT:

1 i

KEY / AIDS Tcble 1 The non - lE AC power system is required to supply operating electrical power to EHC control oil )

components as shown in Table 1 Obj 5 B. Interconnecting Systems

1. EHC Control Logic The EHC control logic provides the control signals for the components of the valve control pacs as well as turbine trip signals to the hydraulic trip system. ,

Additionally the mechanical trip solenoid valve and master trip solenoid valve are powered from power sources within the EHC logic cabinet 10C363.

2. Reactor. Protection System (RPS)

Actuatibn'offcontrol' valve lowLETS fluid pressure,f switches.provides scram input. signals'into the RPS.

-The signal-is automatically bypassed when reactorf

power is' below 30% as sensed' by turbine ~first "stige^

pressure.i

3. Main Turbine Oil System Main turbine lube oil is used to test the mechanical overspeed device. Bearing header oil is used to reset the mechanical trip valve in the turbine front standard.

4. Main Steam and Extraction Steam The EHC hydraulic trip system trips turbine steam admission valves and feedwater heater extraction steam non-return check valves (and associated drain valves) if a turbine trip signal is generated.

Obj 14 VI. TECHNICAL SPECIFICATIONS i

Review the following Technical Specifi.ation as they relate to the EHC Control Oil System:

. 1 A. Section 2.2 Limiting Safety System Settings B. Limitino Conditions for Operation

1) 3/4.3.1 Reactor Protectior System Instrumentation Page 36 Date: 11/15/88 302HC:35 Rev.: 4

.. i ' ' .,

A. QUESTION 6.15 (3. 00)'

In*accordance with procedure OP-AP.ZZ-110~(Q)."Use and Development of Operat ing Logs":

A. The overall status of the Radwaste. systems is recorded in.

the Log.

(0.75.i B. LALL Operations Logs in use during'a shift shall be reviewed

. by '. (TITLE). (O.75 C. Calls to the NRC concerning.significant events shall be recorded-in.the ,_ Log.

(0.75 .

D. Implementation of the Emergency Plan is recorded in the Log.

(0.751 ANSWER 6.15' (3.00)

A. Shift Support Supervisor B. Senior. Nuclear Shift Supervisor C. Control Room (Narrative)

D. Senior Nuclear Shift Supervisor (4 0 .75 each)

REFERENCE-LP 301HC/2:25 (113-01) book 13 page 6

, OBJ El KA 294001A106 (3.4/3.6) 294001A106 ..(KA's) l B. Another correct answer to consider for part "D" is the " Control Room Narrative Log".

C. See attached for supporting documentation.

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u' OP-APl. Z Z- 110 ( 0 )

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l The . types of : entries into. the NSS log. 'l are, but not . limited 'to,. the following:

a '. RX power changes greater than 5% ,

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b. Major equipment status change, c.. Major : system and . equipment testing', i y

d. Conditions that: limit unit operation,

e. Personnel ' accidents or injuries,'

f.. ' Control problems associated with major equipment, h '. Entering or , leaving a Technical Specifications Action Statement,

i. Reactor ' trips and. reasons , . if known,-

j. Automatic protective actions 1and reasons,

( .. Potential ' reportable occurrences ,

- cu-429Y k.

1. Security' incidents ,

m. Out-of-specifications chemistry results,

n. Pertinent miscellaneous information.

-5.223.M' Control [ Room ?NarrAtiive' Log

~

The Control Room Narrative provides a chronological record of . day to day evolutions and significant. events within the plant. The first entry in the log for each shif t shall be the-personnel assignments and plant status. SubsequentGentries into the -

'J1ogishall include? but<not restricted' .to the following:.

l

a. Changes in generator output,

b. Starting and stopping of equipment 4

and the reasons why, 17 Rev. 4 OP-AP.ZZ-110(Q) 1

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OP-AP.ZZ-110(0)'

I L c.. Performance of surveillance : tests and the results of the tests, ,

l I: d. Instrumentation or equipment

- malfunctions.or failures,

e. Relay operations and , targets ,

p f. Chang es ' to 4 KV , 7 Kv , 13 KV , and

.500KV electrical alignments,

g. Reactor trips and the reasons why,

h. Reactor startup and shutdown, a

1. Mo'de changes and the ~ reasons why,

j. Starting and stopping ~ of liquid or gaseous radioactive releases including the release permit number, fliEtaip5EiissEsEliAI26s 76M7theirgency17 7laEaiiillt:hsIreasonitor$suchy actiiodWJ f '.- 1. Of f-site calls to/from NRC or upper management concerning significant events,-

m. Pertinent miscellaneous information.

5.2.3.4 Shift Support Supervisor Log The Shif t Support Supervisors log shall

. record overall status of the Radwaste systems. The log entries shall describe radwaste plant evolutions during a shift in chronological order.

The Radwaste system status section shall be completed at the beginning of each shift.

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18 Rev. 4 OP-AP.ZZ-110(Q)

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A. = QUESTION '6.17- ( 3. 0 0)'

Using.the furnished Event' Classification Guide, IDENTIFY the: Reporting requirements for each of the following events. .If,NONE are required, STATE NONE. . .

A.- An accidental criticality during refueling. (0.75)

B. -Inadvertent. start and injection of the HPCI system during plant operations. (0.75)

C. Presence of-loose parts discovered in the reactor. (0.75)

D.,

A1 Reactor Scram caused by low reactor level during a plant startup.

(0.75)

ANSWER' 6.17 .(3.00)

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months report B.; 4~ hour report C.'24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months report' D. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . report REFERENCE ~

LP Guide Section 18 OB7 KA 294001A116 (2.9/4.7) 294001A116 295006G002 (3.0/4.5) 206000G003 (3.1/4.4)

-295006G002 206000G003 ..(KA's)

B. -Another possible answer for part "B" is a one hour report. This event occurs in two. sections of the ECG: Section 18 (as per the answer key) and Section 10. Section 10 requires this event to be classified as an unusualievent and hence must be reported to the NRC as soon as

-possible, but in all cases within one hour.

The examinees /did ask a question during the written exam concerning "What reporting-requirements?" The examiner's answer was: "NRC report times consistent with 50.72 and 50.73." Emergency Plan implement:ation does appear in 50.7.2.

C. See attached for supporting documentation.

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y SECTION 18 w

-3 3 NON-EMERGENCY /

. INITIATING CONDITIONS INITIATING CONDITIONS INITIATINl E2 ACTUATION.0F ENGINEERED F. REACTOR SCRAM EXCEPT . G. ANY EVENT F(c SAFETY FEATURE EXCEPT PREPLANNED , THAT WOULD C

PREPLANNED- -

DEGRADED PLd

, Note:

J Refer to Sect 2cn i, Introduction i for explenetion of en Engineered Sefetg Feeture

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IF IF fagevent,fod reactor ic shd L st been founs Ang event or conditton &ct would have ed results in manuel or autometto. Manuel or automette Reector Plent, includ2X cetuation-of eng Screm except sefety berriej Engineered Safety Feature (ESF) part of a preplanned sequence being seriod including the during testing or operetten. I Reector Protection System (RPS). !

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comprotnises l' 2 ~

THEN THEN t i

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Note: Actuetsen of en ESF er RPS that is part of a preplanned

__ sequence during testing or __,

operetten m_sy not be reporteble.

Refer to Att.8 of SA-AP.ZZ-226@

for clarif2:etsen.

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Note: Refer also to Section la, !

Technical Specif2cotion Items, ---

prior to elessificetton.

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I REFER TO L ,, .

ATTACHMENT 23 FOUR HOUR REPORT

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- M ...., M, ECG SECTION

! PORTABLE EVENTS PG.2 OF CONDITIONS INITIATING CONDITIONS INITIATING CONDITIONS 60 5'HILE SH1JT00VN H. EVENT / CONDITION THAT ALONE COULD 1. PRESENCE OF LOOSE PART IN THE fE SERIOUSLY HAVE PREVENTED CERTAIN SAFETY REACTOR t ' FUNCTIONS 1

F IF IF U l ehile ths Any event ce condition that sown, that, hed elene could have prevented the wing operation, ,(

fulf211 ment of the safety '

]ted in tha function of structures or sts principle systems that are needed to:

Shutdown the reactor end Presence of a loose part in gdegradad maintain it in a safe shutdown the reactor.

3 conditon enzlgrad OR signifteentig Remove residual heet ent safety. ga Control the releese of redsoect2ve mater 2e1 OR THEN S1 H2tigets the consequences of on ec=2 dent. APERTURE !

CARD i

THEN Also Availab!e On i Note: Events covered in EAL 18.H obove meg include one or more:

Aperture Card \ !

procedural errors, equipment f**I"'"

l dtscovery of des;gn, onelysis, febescetion, cons tructon, *

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end/or peccedurel 2nedequecies.

Note: Refer to Att.2 of b !

SA-AP.ZZ-225(0) concerning 4 2ndividual component re21ures which meg not be reporteble.

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i REFER TO '

REFER TO ATTACHMENT 13 ATTACHMENT 14

.3 HDUR REPORT j

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24 HOUR REPORT -

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INITIATING CONDITIONS INITI ATING- CONDITIONS B.cENORMAL COCLANT TEMPERATURE AN0/0R cc,E55UcE OUTS:CE C

NTol '

A.(OSj f p;AL ef i ca.,-N.. e.c cg.ej,1T,f{siONDguigNGSr.UTCCWN a.A LIMITS-OR LOSS OF ENG:NEERED SAFETY FEATURE OR FIRE PROTE: TION SYSTEM FUNCTICN RECUlRING EHUTC0hN SY TECHNICAL SPEC!FICATIONS te.e.. beceuse of malfunetten, personnel eerce, or procedural inadequee).

OR

ER PLANT CONDITIONS EXIST THAT l

, . he .:'E- ?LANT SHUTDOWN UNDER TECHN! CAL SPECIFICAT;ON REQUIREMENTS

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q p IF IF Shutce.n menuelly inatteted to Shuteewn tnts:eted to comply Shutdown a cemele with Technical v:th Technicel Specif:cattens with Tec.1 Specificettons Action Statements Aetten Statements 3.4.6.1 efterex def:ned 1 3.4.6.2 4-l -

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UNUSUAL EVENT

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PG.1 0F 3' INIT1 ATING CONDITIONS IN1TIATING CONDITIONS

'iECHNICAL SPECIFICATION . C. EXCEEDING 10Dir.E SPI <E (T.E. LIMITS) LEVELS CN ' D.EME:,0ENCY .CC:.E -CC13.G EYSTEM '

REACTCR COCLANT I 131 EQUIVALENT ACTIVITY EC:5)Ni ATE 3 $N3 OISCHARGE T3 AND/0R LEVELS IN OFF GAS VE5SEL HOTE:If SNS3/EDO Judgement indicates pecheble fuel f ailure completten of chemis.rg semple should net restrict /deley clessificetten. .

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IF- IF IF attleted to comply Leel Spsetiteettone A_ng of the felle tng: T.S. 3.5.1 Aetten Statement g been enteredc .

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Off Gas redietton monitor h88,3 0:ing Safety Ltmits es Alert eierm (RM11 Rc621/Rc622)

E" EPV 2"J8C22 "L OR Off Ges treeted rediet:en mentter T.S. 2.2 Alert olerm (RM11 R 525/Rc626)

OR Reector Coelent Activitu creeter then 4 uCt/gm Dese Equiv'elent I-131(as determined by comple)

NOTE Refer to Section 5 for esceletang conditions l<

MEN THEN THEN !

SI - 1 APERTURE CARD d

- Also Available On Aperture Card REFER TO REFER 70 REFER TO ATTACHMENT 1 ATTACHMENT 1 ,

STTACKYENT 8 UN'JSUAL EVENT -. UNUSUAL EVENT-JNUSUAL EVENT 1 f

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ATTACHMENT 3 NRC RESPONSE TO FACILITY COMMENTS The following represents the NRC resolution to t.he facility comments (listed in Attachment 2) made as a result of the current examination review policy. I Comments made that were insignificant in nature and resolved to the satisfaction of both the exsminer and the licensee during the pre-examination review are not listed (i.e.: typographical errors, relative acceptable terms, minor set point changes).
Question 5.06: Comment rejected. The PLANT CONDITIONS listed are Emergency Conditions requiring the use of symptom based Emergency Operating procedures.
Question 6.05: Comment accepted. Revised answer key to accept "EHC System" Question 6.15: Comment accepted. Revised answer key to accept " Control Room (Narrative) Log.
Question 6.17: Comment accepted. Revised answer key to accept "I hour report"
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ATTACHMENT 4 SIMULATION FACILITY REPORT Facility License: DPR-16 Facility Docket No.: 50-354 Operating Test Administered on: February 15-16, 1989 This attachment reports examiner observations for information only. These i observations do not constitute audit or inspection findings and are not, '
without further verification and review, indicative of non-compliance with 10 CFR 55.45(b). The observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were observed.
ITEM DESCRIPTION SPDS The SPDS displayed erroneous information during all ooerating examinations.
P-1 When the simulator Initial Conditions.were established to commence the operating exams, the P-1 edits were in error. The indicated APRM power level and the P-1 power level were not within 2%. The applicants would reference the plant Technical Specifications and contact I&C to adjust the GAFs. This item caused an unnecessary delay during the operating exams.
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ML20244B616

Text

Dear Mr. Russell:

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