Text

Requests for Hearing IAW CFR2.103(b)(2).Supporting Documentation Encl
	ML20129D547
Person / Time
Site:	Turkey Point, 05520726
Issue date:	09/25/1996
From:	Tetrick R; AFFILIATION NOT ASSIGNED
To:	; NRC OFFICE OF THE SECRETARY (SECY)
Shared Package
ML20129D535	List: ML20129D530; ML20129D547;
References
	SP, NUDOCS 9610250070
	Download: ML20129D547 (47)
	v • d • e

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DOCKETED USNPC DOCKETED USPC

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,ce:77 TO: Sedrek: jo- {Er ung y

.the'Odmmission 00CKEiMG E J r VIC' U.S. Nuclea.' Regulatory Commission BRAHM Washington D.C. 2055$

Dear Sir:

Per the letter dated September 12, 1996 I am requesting a hearing IAW 10 CFR 2.103 (b) (2) .

Enclosed you will find:

1. A copy of the original request and documentation.

2. A copy of the denial dated September 12, 1996.

3. My reply to the above dated letter.

Sin rel , -

9 Ralph L. Tetrick 18990 SW 270 Street Homestead, FL 33031 Docket No. 55-20726 9610250070 961007 PDR MISC 9610250065 PDR

4 July 30, 1996 TO: Director Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission, Washington, DC 20555 ;

l

Dear Sir:

Per the letter sent to me on July 19, 1996 I am requesting an informal review of my written examination. Enclosed you will find copies of the questions I wish to be reviewed

along with supporting documentation.

Sincerely, Ralph L. Tetrick l 18990 SW 270 Street Homestead, FL 33031 Docket No. 55-20726

)

SRO QUESTION 24 Which ONE of the following describes the Spent Fuel Pool Cooling (SFPC) system basic operation and connections to the Spent Fuel Pool (SFP)?

The.SFPC pumps normally take a suction on the:

ANSWER:

a. "High" line near the top of the SFP and discharge through a line 1 foot below the top with a 1/2 inch siphon break hole 6 inches below the water level.

REFERENCE:

SD-041, Fuel Pool Cooling, Purification and Ventilation System' page 16. E.O. OF LP 6902141 COMMENT:

Answer (A) is partically incorrect because, (1) the discharge line is routed 10 inches below nominal water level and extends to the middle of the pool (ie 20 feet from top and bottom) and (2) ,

the siphon break is 14 inches below nominal water level not 6 I inches.

Answer (C) is partically incorrect because, (1) the "High" suction line is approximately 3 1/2 feet below the nominal water level and (2) the discharge line is as stated above.

Both answers A and C are equally correct because they indicate the suction is from the high line and that there is a siphon break in the top of the discharge line. An answer of A or C indicates the operator is aware of the design requirement to prevent inadvertant draining of the SFP.

RECOMMENDATION:

Accept answer c as an additional correct answer sence both a and c are partically incorrect.

C. "High" line 1 foot below the top of the SFP and discharge through a line at thr bottom of the SFP with a 1/2 inch siphon break hole 6 inches below the water level.

1

SD 041 01/14/93 Page 16 FUEL POOL COOLING, PURIFICATION AND VENTILATIOM SYSTEM l i

There is a thermal expansion loop in the piping on the discharge of the Goulds SFP cooling pump to accommodate thermal stresses due to pool boiling at 212*F.

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) Spent Fuel Pool Cooling Pumps 3

Three SFP cooling pumps are provided, A, B and emergency. Refer to Fiaure 10. A&B )

l pumps are horizontal centrifugal pumps rated for 2300 gpm at 125 ft. TDH. l Additionally, both pumps are powered from LC C (breaker 0309) via manual transfer panel P-16. There are two switches located on this panel; One switch is for pump A and the other switch is for pump B. They are interlocked such that only one switch can be closed at a time. The pumps are located in the SFP heat exch;,1ger room and are controlled locally.

The third pump, emergency SFP cooling pump is also provided. It is used only when the SFP cooling pumps are not available. Power for this pump is provided by a receptacle in the cask wash area new fuel room. When the emergency pump is used, the SFP purification loop is bypassed. The Emergency Spent Fuel Pit Pump Motor is not normally connected to a permanent 480V AC power source. If its use is required the temporary local motor starter / disconnect stand and attached cables needs to be moved to outside of the Spent F9el Pump Room and the load side cable connected to the Emergency Spent Fuel Pump Motor. The temporary motor starter / disconnect stand is normally stored in the new fuel storage room when not in use. The temporary motor / starter disconnect may be connected to provide standby operations at PS-N !

discretion.

The SFP cooling pumps can take a suction on the SFP through the high suction valve (796) or the low suction valve (797). The high suction line penetrates the SFP near the top and terminates. The low suction line penetrates at a level 6' above the top of the fuel assemblies and extends downward to almost the bottcm of the SFP. Complete j siphon draining of the pit by a break in this line is prevented by a normally locked '

closed valve located at the same elevation as the penetration (797). There are no other connections provided on the SFP. The cooling loop dischargt line penetrates the SFP at approximately l' below the top and extends straight down towards the stored fuel. A 1/2" hole is drilled in the discharge line at approximately 6" below the ,

water surface, it acts as a siphon breaker. .

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SRO QUESTION 63 i

Plant conditions:

I

- Preparations are being made for refueling operations

- The refueling cavity is filled with the transfer gate l valve open.

- Alarm annunciators H-1/1, SFP LO LEVEL and G-9/5, CNTNT SUMP HI LEVEL are in alarm.

Which ONE of the following is the required IMMEDIATE ACTION in response to these conditions?

i ANSWER:

b. - Sound the containment evacuation alarm.

REFERENCE:

3-ONOP-033.2, Refueling Cavity Seal Failure, page 5 l E.O. 6 of LP-6902144 COMMENT:

1 Annunciator H-1/1 is a entry condition for 3-ONOP-033.1 (Attachment 1) . The inmediate actions for 3-ONOP-033.1 is to I verify the alarm is valid.

Additionally, the RCO is required to respond to alarms per ,

0-ADM-219 (Attachment 2). Alarms H-1/1 and G-9/5 are classified '

as priority 3 (BLUE) alarms requiring prompt (not immediate) action. The specified operator actions for both alarms per ARP-097.CR is to verify the alarms (ie containment sump level recorder and spent fuel pit level indication). See Attachment 3 for the ARP actions.

RECOMMENDATION:

Accept answer A as an additional correct answer.

A. - Verify alarms by checking containment sump level recorder and spent fuel level indication.

i

. I

eru .aure m erxxuo rm.: e. ,.:

Sp:nt Fus! Pit (SFP) iponns om:

Cooiing Systsm Melfunction 3-ONOP.033.1 10/21/92 1.0 PURPOSE kTTRchtn e.m \ l?Rh6 i ns O 1.1 This procedure provides instructions for response to off-normal conditions of I the ent Fuel Pit (SFP) and the SFP Cooling System including SFP High/ Low Leve and High Temperature.

2.0 SYMPTOMS 2.1 Annunciators e 2.1.1 H 1/1, SFP LO LEVEL l 2.1.2 H 1/2, SFP HITEMP 2.1.3 H 1/3, SFP HI LEVEL i' 1 2.2 Indications i 2.2.1 High/ low SFP, as indicated on LI-3-651. (VPB) or by local visual l j

inspection using level placard located at the Southwest corner of the SFP (normal levelis 56 - 10"- 57'2")

I 2.2.2 Low SFP Cooling Pump Discharge Pressure (PI-3-651B; 651A) 2.2.3 SFP Filters High AP tDPI-3-151 A, B, C),

AP across filters should be < 10 psid 2.2.4 Low SFP Skimmer Pump Discharge Pressure (PI-3-671A) 2.2.5 SFP Skimmer Pump Filters High AP (DPI-3-150A, B, C),

AP across filters should be <10 psid 2.2.6 SFP Demineralizers High AP (IMLET PI-3-655 A - OUTLET PI-3-655B) 1 (N/A if RWST is on Recirc through the SFP Demin)

AP across Demin should be <35 psid 3.0 AUTOMATIC ACTIONS 3

. NOTE .

I l Bkrs for SFP pps are locatedinside Unit 3 SFP Hx Roorn.

l l l a. 3A SFP Pp Bkr3P212A l

. =

' 38 SFP Pp Bkr 3P212B (Power supply to 3NP212 panel is fed

. b. .

l from Bkr30309) l t._. ._.-._._._. ._._._. . .-.-. . ._._._. .-._.J 3.1 Possible SFP Cooling Pump Breaker trip on overload.

3.2 Possible SFP Skimmer Pump Breaker trip on overload. (Bkr 30777) 4.0 IMMEDIATE ACTIONS 6 4.1 Verify annunciated alarm is valid.

. w;c s.,,

. mas 7. ^ z vue.- paea:-

Approval Date:

0-ADM-219 Annunciator Response Procedure Usage 3/12/96 kPCBM:.NT 2 (9RGE 169P_3 3.0 RESPONSIBILITIES 3.1 Nuclear Plant Supervisor (NPS) - The NPS shall provide technical guidance for event mitigation when ARPs are in effect.

3.2 Assistant Nuclear Plant Supervisor (ANPS) - The ANPS should direct the detailed event minganon strategy for the affected unit unless otherwise directed by the NPS when ARPs are in effect.

3.3 Nuclear Watch Engineer (NWE) - The NWE should direct non-licensed operators, if necessary, to determme the cause of the alarm condition and the performance of corrective actions when ARPs are in effect.

3.4 Affected Unit Reactor Control Operator (RCO) - The affected unit RCO is responsible for the followmg when ARPs are m effect:

3.4.1 Respond to alarms based on color code priority and plant conditions.

s

>- 3.4.2 Reading the ARP in effect and performing the event mitigation strategy for alarms received in the Control Room.

3.4.3 Transition to the appropriate procedures if required by the ARP.

3.4.4 Inform the unit ANPS of abnonnal alarm conditions.

3.4.5 Coordinate actions with non-licensed operators when the alarm condition occurs at local annunciator panel in the field.

3.5 Non-affected Unit Reactor Control Operator - The non-affected unit RCO should mamtam the non-affected unit m a safe condition which does not threaten the event mitigation strategy on the affected unit when ARPs are in effect.

3.6 Third Licensed Operator -

The third licensed operator should assist the affected unit (s) RCO m performance of the event mitigation strategy when ARPs are in effect.

3.7 Non-Licensed Operator (NLO) - The NLO is responsible for the following when ARPs are in effect on either unit:

3.7.1 Read the ARP in effect and perform the event mitigation strategy for alarms received at local annunciator panels.

3.7.2 Inform the affected unit RCO of the alarm condition.

3.7.3 Performing actions requested from the affected unit RCO or NWE to correct the alarm conditions in the Control Room.

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un " " '

bc aw. 'lan:- . ..

~

m homer., . ,, .

Approval oste:

0-ADM-219 Annunciator Response Proctdure Usag2 3/12/96 hTTfWme.NT 2 (2. N 2.3 4.0 DEFINITIONS 4.1 Annunciator Response Procedures (ARPs) v Plant procedures that specify the operator actions required to mitigate the consequences of transients that cause plant parameters to exceed alarm setpoints.

4.2 Local (Locallv)

An action performed by an operator outside the Control Room.

4.3 Manual (Manuallv)

An action performed by the operator in the Control Room. His does not include automatic actions which take place without operator intervention.

4.4 Priority 1 (Red): Nuclear Safety These alarms require immediate response and reflect a potential challenge to plant safety and require protective systems to activate. These alarms include:

Si and Reactor / Turbine / Gen Trips.

4.5 Priority 2 (Yellow): Power Production Availability -

These alarms require immediate response and reflect a challenge to plant ,

equipment or systems that may affect continued plant availability or timely recovery. Imtaediate response to these alarms would be deferred only if action were required by a Priority 1 alarm. Failure to properly respond to a Priority 2 condition may lead to or contribute to a higher level condition.

4.6 Priority 3 (Blue): Investment Protection R ese alarms require prompt response and provide information that, if unattended, may result in a threat to higher level actions. Prompt action to this level of alann may reduce the consequences of the problem by minimizing equipment damage or material waste.

4.7 Priority 4 (White): Status /Information D ese alanns require non-priority response and reflect equipment status, transitions, or conditions to be corrected, but do not threaten the unit availability. Because they are not strictly "Information Only" items, they may warrant , operator action. Priority 4 items are deferred in the face of higher pnonty items.

4.8 Transition A change from one place to another in the procedures, either from one step to another step or from one procedure to another procedure.

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aeoem.. v aeo n sn . ,,,

Central Room Annuncistor Rssp::nsa ** lTiOY3 l 3. ARP.097.C R LWESTMENT PROTECTION H L1 l !

BLUE t

51 gyggmervra 1EX )_ q l I \ D:l?!!'

- - * ~ ~

g7p 2 l 1

l 3 ! l I gggy LO LEVEL 5 , ! , I l

' gg gpgj 6 ; I l J l l l 1 23 4 5678 9 DEVICES: SETPOINTS:

Level actuator at 56'10" north end of SFP LT 3-651 OPERATOR ACTIONS:

1. Verify alarm by checking the following:

@ a. LI-3-651 (VPB)

2. Corrective actions:

a. Dispatch operator to check:

(1) Spent ' fuel level indication LY-3-651 (behind VPB) '

(2) Locallevel at the SFP. ;

(3) Power to LT-3-651 (LP-50. Bkr 19 . east wall of cable spreading room. ;

b. Refer to 3-ONOP-033.1 SPENT FUEL PIT (SFP) COOLING.

LF in a refueling configuration with the SFP transfer tube omen, THEN l c.

terminate refueling operations and refer to 3-ONOP-033.2. 3efuehng Cavity Seal Failure System Malfunction for cavity seal failure required actions.

d. Refer to TS 3.9 for additional actions.

. NOTES I I '

If SFP coohng has to be secured, monstor SFP temperature per Attachment 3 of l.e

. 3.OP.033 notify Reactor Engineersng.

I I '

If annuncsatoris 005, refer to 0-ADM-214. .

i.._._...._._...._._._._._._.._._...._.a CAUSES:

1. Actuallow levelin SFP (Evaporation, leakage or SFP system valve misalignment)

2. Loss of power to LT.3-651

3. Instrurirentation failure

REFERENCES:

i

1. FPL DWG 5613-M-3033 Sh 1

2. Tech. Spec. Section 3.9

. oc< .- :: su

l

'- G 391 aw,.i o.s. .

Centrol Reem Annunciator Rsspon>3 8/6/92

! 3-ARP-097.CR INVESTMENTPROTECTION ' G 9/5 l

! BLUE

! G45 ATTACHMENT 7 P ; 50 of 54 I

l 1 l l 3r:10 - CNTMT SUMP 2 l l j 3 Il E EEL j

d I <kTeement 3 (TR65 2W2> -

a 6 \

l l 1 234 56.789-DEVICES: SETPOINTS:

R-1418 30" l

i (unit 4 VPA)

OPERATOR ACTIONS:

1. Verify alarm by checking the following:

.> a. CNTMT sump recorders R-1418 (unit 4 VPA), R-6308A/B, and DDPS point 1

DDPSA102-3.

2. Corrective actions:

a. Verify proper operation of the containment sump pumps.

b, Pump c own the sump as required.

c. Momtor RCS parameters for indications ofleak, if applicable.

d. Perform 3 OSP-041.1 to determme the RCS leak rate,if applicable.

._._._. ._._._._g g .-.-._. 3 g.-.-.

tiannunustor us oos. refer to 0-Aou-2 u. l l

t._._._._._._._._._._._._._._._._._._._._._._._.;

CAUSES:

1. RCS leak.

2. Instrument malfunction.

l

REFERENCES:

1. FPL DWG 5610-M-12 !

e 15/JGKJct

. o SRO QUESTION 84 Which ONE of the following is the basis for step 1, " Verify Reactor Trip", of FR-S.1, Response to Nuclear Power Generation /

1 ATWS?

ANSWER:

a. - To ensure that only decay heat and reactor coolant pumps are a

adding heat to the RCS.

REFERENCE:

3-BD-EOP-FR-S.1, Response to Nuclear Power Generation /ATWS, page

8. E.O. 6 of LP-6902346 W

COMMENT:

A review of the corresponding Step 1 of 3-EOP-E-O and 3-EOP-FR-S.1 with respect to reducing reactor power indicates a difference. Rods are manually inserted in FR-S.1 but not E-0. While the basis documents for both procedures discuss decay heat and reactor coolant pump heat, only Basis Document BD-EOP-FR-S.1 discusses the need for taking further corrective action if the reacto is not tripped. (ie manually insert control rods) (see attached basis documents-and procedures)

RECOMMENDATION:

Accept answer C as an additional correct answer.

C. - To alert the operator to take further corrective action if the reactor is NOT tripped.

Procoeure fitta: Page

. Proceduro taa.s 7 Apprgv3L Dat@:

3-EOP-E-0 REACTOR TRIP OR SAFETY INJECTION 06/22/95 ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED STEP l l

l 7...........................................................................,

j NOTES i i i i . Steps 1 through 14 are IMMEDIATE ACTION steps. i i i i e Foldout page shall be monitored throughout this procedure. i i i 6............................................................................a 1 verify Reactor Trip: Manually trip reactor. ir reactor i power is areater than 5% OR l

= Rod bottom lights - ON intermediate range power is NOT stable or decreasing, THEN perform )

e Reactor trip and bypass the following: ;

breakers - OPEN

a. Direct operator to monitor

. Rod position indicators - AT Critical Safety Functions using ZERO 3-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES.

j

. Neutron flux - DECREASING

b. Go to 3-EOP-FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS, Step 1.

O

Page 9 3-BD-EOP-E-0 REACTOR. TRIP OR SAFETY INJECTION 08/01/91 BASIS DOCUMENT WOG Procedure Step: 1 PTN Procedure Step: 1 BASIS:

Reactor trip must be verified to ensure that the only heat bein5 added to the RCS is from decay heat and reactor coolant pump heat. The safeguards systems that protect the plant during accidents are designed assuming that only decay heat and pump heat are being added to the RCS, If the reactor cannot be tripped, a transition is made to FR-S.1, RESPONSE TO NUC M POWER GENERATION /ATWS, to deal with ATUS conditions.

STEP DEVIATION FROM WOG GUIDELINE:

TYPE DESCRIPTION 8 The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology. I 1 The RNO was changed so that a transition to FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATUS, will only be made if the criteria from l

~

the Critical Safety Function status tree for suberiticality is satisfied. This provides the operator with a clear definition of what constitutes a reactor trip, and eliminates the need for the operator to make a decision under stress. This change complies with the intent of the RNO column provided in ERG Feedback item DU-88-033.

9 The WOG guidelines require initiation of Critical Safety Function status tree monitoring whenever exiting E-0. The RNO was modified to provide procedural guidance for performance of this task so that the need to memorize User's Guide requirements is eliminated.

PLANT SPECIFIC SETPOINTS:

5% - Reactor power level just in the Power Range. (EOP Setpoint P.2)

Procedur-3 KO.! Profecura 11tle: P:g)

ADDPCral Data: ,

3-EOP-FR-S.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS 03/30/95 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l 3

'-*-*-*--*""*-*-'**-'-**'*--*-*-*-*-**1 ri * - - - * - * - ' * * "" ' *" * - * - ' *" "" - * - ' - * ** * *"N O T E i i

i i Steps 1 through 2 are IMMEDIATE ACTION steps. i i i

. L.._.......................................................a 1 verify Reactor Trip: Manually trip reactor. g reactor ,

will NOT trip, THEN manually l

. Rod bottom lights - ON insert control rods.

Reactor trip and bypass breakers - OPEN

Rod position indicators - AT ZERO l

= Neutron flux - DECREASING 2 verify Turbine Trip:

a. All turbine stop valves - CLOSED a. Manually trip turbine. E turbine will NOT trip, THEN manually run back turbine.

IF steam flow to turbine causes l E controlled RCS cooldown, THEN I l'

close main steamline isolation and bypass valves. .

i l

b. Close MSR Main Steam Supply b. Close main steamline isolation !

Stop MOVs and bypass valves. !

c. Reheater timing cam - AT ZERO c. Remove timing cam to close

timing valves. E any timing valve can NOT be closed, THEN close main steamline isolation and bypass valves.

d. MSR Purge Steam Valves - CLOSED d. Manually close MSR purge valves. E any MSR purge valve can NOT be closed, THEN close main steamline isolation and bypass valves.

3 Check AFW Pumps - ALL RUNNING Manually open steam supply valves.

4

Page 8 3-BD-EOP-FR-S.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS 08/01/91

. BASIS DOCUMENT.

WOG Procedure Step: 1 .

PTN Procedure Step: 1

, BASIS: l Reactor trip must be verified to ensure that the only heat being added to

! th2 RCS is from decay heat and reactor coolant pump heat. The safeguards j systems that protect the plant duringaccidentsaredesignedassumingthat) only decay heat and pump heat are being added to the RCS. If the reactor <

cannot be tripped, then the control rods should be manually inserted into )

the core in order to decrease reactor power. l STEP DEVIATION FROM WOG GUIDELINE:

TYPE DESCRIPTION i 1

8 The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

PLANT SPECIFIC SETPOINTS:

N/A

^

l l

i

=. -_ - - - .. . . . _ . . _ _ _ - -

SRO QUESTION 96 Which ONE of the following is the lowest level of position responsible for ensuring entries are made in the Technical Specification Related Equipment Out-Of-Service Index?

ANSWER:

b. Assistant Nuclear Plant Supervisur

REFERENCE:

0-ADM-213, page 10 ;

COMMENT:

l 0-ADM-200, Conduct of Operations states that the Nuclear Watch i Engineer (NWE) is responsible for " Routinely relieving the ANPS of the Control Room command function". By relieving the ANPS i the NWE assumes all responsiblities of the ANPS thereby becoming the lowest level able to make entries into the EOOS index. Additionally FP&L Training Dept. recently resolved a j similar test question on our Contractor Exam (Question 88, see attached) and ruled similar as this request (ie accepted ANPS and NPS since ANPS can relieve the NPS).

RECOMMENDATION:

Accept answer d as an additional correct answer.

d. - Nuclear Watch Engineer

Process No.: N- -+re reue-I8 Approvu cat 7 0-ADM-200 Conduct cf Operations 2/22/96 3.4.6 Review and approve all unit Plant Work Orders. prior to work commencing and ensuring the NWE and RCO are aware of all work outside of the Control Room.

3.4.7 Maintain the equipment out-of-service book in accordance with I 0-ADM-213 Technical Specification Related Equipment and i Risk Significant SSC Out-of-Service Logbook I 3.4.8 Coordinate the on shift training of licensed operators.

3.4.9 Maintaining a thorough knowledge and understanding of the following:

1.

He duties and responsibilities of the ANPS required by the facility operating licenses.

2.

Conditions and limitations contained in the facility operating licenses and Technical Specincations.

3. Operating procedures for the nuclear units.

4. Plants' status at all times. [ Commitment - Step 2.3.3]

3.4.10 Notifying the NPS when any Technical Specification Limiting Condition for Operation is entered.

3.4.11 Notifying the NPS when any Risk Significant SSC is removed I from service. I 3.5 Nuclear Watch Eneineer (NWE) - One Nuclear Watch Engineer will be assigned to assist the NPS in coordinating the activities of Licensed and Non-Licensed personnel during routine. complicated. or infrequent evolutions.

He NWE reports to the NPS and is responsible for:

3.5.1 Performing duties assigned by the NPS o: his designee for each unit.

3.5.2 Coordinating the activities of the Control Room with other operations and plant personnel to achieve safe. reliable, and efficient unit operation as directed by the NPS/ANPS.

3.5.3 Supervising and coordinating the operation of plant equipment and systems when assigned by the NPS.

3.5.4 Actina as the Fire Brigade Chief or Shift Communicator. when

~

assigned, but not both.

3.5.5 Routinely relieving the ANPS of the Control Room command function to enable the ANPS to leave the Control Room. In an emergency, function as the NPS if required.

\

13MBSittltr/dt

QUESTION 88 RO & SRO Which one of. the following is correct regarding who has control and

' responsibility for the issuance of ICCS keys?

a. NPS b NPS/ANPS

c. NPS/ANPS/NWE

d. NPS/ANPS/NWE/RCO ANSWER: A

. ~ .

REFERENCE:

294001K1.05 h'.4/3.6)

ADM-205, Section 10, Key Control ADMf2.05%ssstf6EES f4 hf Key jV Control:, clearly states- ICCS arehunder theVbontirb1$of$thesNPSMThe'hquestion . asks about issuance, nothing .

eliseMEJNO? CHANGES %REQUIREDf v way.v.w4vn. %v W4M%vh wnM+e-A%ww.w.+iw. .vM.-Aw. - w .w.w. .a

' QCY&sVi'sysdisb6ssshdstibdFcohsider'ed and final: decis' ion:Jmadeito addspt5Af6E!$fBlaisicorfsetsbe'causee.of :-numerus ~ exceptions. to'.tthe2NPS 1

iss6eissddifsmsht9PiANPS$mapibe acceptable forcissue of'ICCS'akeys .

undericsktsin166hditsi'onhiidADM--205 does not have*any' provisionsifor 2

t_vli_eI_R..C~O?6_fmelNWE(Everfi,ssuingJICCS keys-i i

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- . . w- =* _. a . _. A

/ 'g

UNITED STATES O NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20656 4001

\

- - <*l September 12, 1996 i

Mr. Ralph L. Tetrick 18990 SW 270 Street Homestead, FL 33031

Dear Mr. Tetrick:

In response denial issued to to youyour htter on July 19, of July 30, 1996, we have reconsidered 1996, the pro examination adminis',ered to you on Juneand reviewed the grading of the written 14, 1996, you supplied.

result of our revier is enclosed.We find that you did not pass The the written' Consequently, the proposed denial of your license ap .

days as discussed below, the preposed denial will become a finalYou denial.

may thenconditions:

following reapply for a license in 3ccordance with 10 CFR 55.35, subject to the a.

Because you did to you on June 14,not 1996, pass the written examination administered written examination, you will be required to retake the b.

Because June 19 - you 21, did pass your operating test administered on 1996, you may request a waiver of that portion.

A waiver would be valid for up to 1 year from your test date. !

c. j You may reapply for a license 2 months from the date of this letter.

facility management.A reexamination will be scheduled upon request by your If you do not accept the proposed denial of this letter, request a hearing pursuan,t to 10 CFR 2.103(b)(2).you may, within request, in writing, to the Secretary of the Commission, U. S. NuclearSubmit your Regulatory Commission, Washington, D. C.

20555, with a copy to the Assistant General Counsel at the same address. for Hearings and Enforcement, Office of the General Counsel, Failure on your part to request a hearing within 20 days constitutes a waiver of your right to demand a hearing and, for the purpose of reapplication under

A

! Mr. Ralph L. Tetrick 1 -

4 10 CFR 55.35, renders this letter a notice of final denial of your application, effective as of the date of this letter.

If you have any questions, please contact Mr. Stuart A. Richards, at 301-415-1031.

Sincerely, i

t W .-

Bruce A. oger, Director i Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation Docket No. 55-20726

]

Enclosure:

As stated cc w/ encl.: Mr. W. Lindsey, Licensed Operator I' Training Supervisor, Turkey Point l

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l NRC REVIEW FOR RALPH L. TETRICK - SRO APPLICANT - TURKEY POINT

! In response to the applicant's letter of Aly 30, 1996, the NRC reconsidered the proposed denial issued on July 19, 1996, and reviewed the grading of the )

written examination administered to the applicant on June 14, 1996. Giving

due consideration to the information supplied by the applicant, the NRC has

determined that the applicant failed the written examination. The results of i

the NRC review are outlined below.

REVIEW OF WRITTEN EXAMINATION OUESTIONS j EXAMINATION QUESTION #24

4

) Which 0NE of the following describes the Spent Fuel Pool Cooling (SFPC) system j basic operation and connections to the Spent Fuel Pool (SFP)?

i i The SFPC pumps normally take a suction on the:

a. "High" line near the top of the SFP and discharge through a line 1

) foot below the top with a 1/2 inch siphon break hole 6 inches below l the water level.

b. " Low" Ifne near the botton of the SFP and discharge through a line 6 inches below the top with a 1/2 inch signon break hole 1 foot below i the water level. '

i

, c. "High" line 1 foot below the top ur the SFP and discharge through a f line at the botton of :he SFP with a 1/2 inch siphon break hole 6 inches below the water level.

, d. " Low" line located with a 1/2 inch siphon break hole 1 foot below l l the water level of the SFP and discharge through a line 1 foot below the top.

l ANSWER: a Applicant's Contentions (summary): The applicant asserts that choices "a" and "c" are both partially incorrect because the discharge path enters at the top of the SFP but then extends down 20 feet to the middle of the pool, where the returning fluid is actually " discharged." The applicant also points out some other minor discrepancies in the elevations specified in the question. The applicant recommends that the NRC accept choice "c" as an additional correct answer because both choices are partially (in) correct.

NRC Analysis and

Conclusion:

Answers "b" and "d" are clearly incorrect because the " low" line suction is normally locked closed to prevent inadvertent draining of the SFP. The suction points specified in choices "a" and "c" are essentially the same, so the decisive difference between the choices rests in the return flow path "high" (i.e., one foot below the top ENCLOSURE

1 l.

1 4 I of the SFP) in choice "a" and " low" (i.e., at the bottom of the SFP) in choice

!' "c." Although the intent of the question was to evaluate the applicants' general knowledge of- the piping penetrations through the wall of the SFP (not

! the specific elevations of each penetration with respect to the top of the pool or the nominal water level), the NRC agrees with the applicant that the i

wording of the choices (i.e., " discharge through a line") could also lead the reader to believe that the question was concerned with the location of the outlet of the SFP cooling system discharge pipe. Under this interpretation neither answer "a" or "c" is correct; the discharge pipe terminates at an j

elevation approximately half way between the bottom and the top of the SFP.

i Therefore, the NRC concludes that no answers are completely correct and the

! question should be deleted from the examination.

s j EXANINATION QUESTION #63:

) Plant conditions.

l Preparations are being nade for refueling operations.

The refueling cavity is filled with the transfer tube gate valve open.

Alars annunciators H-1/1, SFP LO LEVEL and G-9/5, CNTWT SUMP HI LEVEL are in slann.

) Which ONE of the following is the required IMMEDIATE ACTION in response to 1 these conditions?

a. Verify alanes by checking containment sump level recorder and spent l fuel level indication.

I b. Sound the containment evacuation alarm. !

1 l

{ c. Initiate containment ventilation isolation. i i

d. Initiate control roon ventilation isolation.

J i ANSWER: b j Applicant's Contentions (summary): The applicant asserts that the reactor

' control operator (RCO) is required to respond to alarms per 0-ADM-219,

" Annunciator Response Procedure [ARP] Usage," by reading the ARP in effect and performing the event mitigation strategy for the alarms received in the control room. The applicant also states that, individually, these alarms are priority 3, which require prompt rather than immediate action, and that the appropriate response for each alarm, per the associated ARP and off-normal operating procedures, is to verify the alarm. Therefore, answer "a" should be accepted as an additional correct answer.

NRC Analysis and

Conclusion:

Reactor operators and senior reactor operators are expected to analyze alarms and determine the appropriate course of action 2

based upon the specific plant conditions and indications. This expectation is

reflected in step 3.4.1 of 0-ADM-219, which directs the RC0 to respond to alarms based on color code priority and plant conditions. Furthermore, steps 5.1.15 and 5.6.8 of 0-ADM-200, " Conduct of Operations " direct on-shift licensed operators involved in abnormal or emergency o,perations to believe and respond to their instrument indications until the instruments are proven to be

{

i incorrect. The plant conditions and indications specified in this question

! (i.e., the refueling cavity filled and the transfer tube gate valve open with coincident SFP LOW LEVEL and CONTAINMENT SUMP HIGH LEVEL alarms) are mutually supportive, confirmatory, and sufficient to enter Off-Normal Operating Procedure 3-0NOP-033.2, " Refueling Cavity Seal Failure." That procedure has only one IMEDIATE ACTION - to sound the containment evacuation alarm. In accordance with step 5.2.1 of 0-ADM-211. " Emergency and Off-Normal Operating Procedure Usage," operators shall be capable of performing steps identified as IMEDIATE ACTION steps from memory, and step 3.5.1 requires the RCO to ensure that all immediate operator actions of the procedure in effect are performed.

Furthermore, step 5.13 of 0-ADM-211 states that plant operating procedures have the following order of priority: Functional Restoration Procedures, Optimal Recovery Procedures, and Off-Normal Operating Procedures. That information plus the fact that both of the existing alarms were classified by their respective ARPs as priority 3 (blue) alarms that require prompt rather than immediate response supports the argument that there is only one correct IMEDIATE ACTION in response to the stated conditions. Therefore, the NRC concludes that choice "b" is the only correct answer to this question.

EXAMINATION QUESTION #84:

Which DNE of the following is the basis for step 1, " Verify Reactor Trip", of i FR-S.1, Response to Nuclear Power Generation /ATWS?

a. To ensure that only decay heat and reactor coolant pumps are adding heat to the RCS.

b. To ensure shutdown margin is within Technical Specifications limits for HOT STANDBY. i

c. To alert the operator to take further corrective action if the reactor is NOT tripped. i

d. To verify that all automatic reactor protective features have functioned as designed. ;

Answer: a Applicant's Contentions (suimmary): The applicant compares the corresponding steps and the difference in the basis documents for procedure FR-S.1,

" Response to Nuclear Power Generation /ATWS," and procedure E0P-E-0, " Reactor Trip or Safety Injection." In FR-S.1, the control rods are manually inserted, 3

1 1

i i

b$ttheyarenotinsertedinE0P-E-0.

l The basis documents for both procedures j discuss decay heat and reactor coolant pump heat, but only the basis document for FR-S.1 discusses the need for inserting control rods if the reactor is not tripped.

4 Therefore, accept answer "c"-as an additional correct answer.

NRC Analysis and

Conclusion:

The basis documents for both FR-S.1 and E0P-E-0 clearly indicate that the first step of each procedure (i.e., " verify reactor trip") is necessary to ensure that the only heat being added to the reactor coolant system is from the reactor coolant pumps and the radioactive decay of fission products.

The safeguards systems that protect the plant during accidents are designed on the basis of that assumption. Although the basis document for FR-S.1 goes on to state that the control rods should be manur.11y inserted into the core to decrease reactor power if the reactor cannot be tricped, this does not change the fact that the reason (i.e., basis) for the (aup>matic and manual) action associated with this step of FR-S.1 (i.e.,

tripping the reactor or inserting control rods) is to reduce the heat load to vithin the capacity of the safeguards systems. Therefore, the NRC concludes that choice "a" is the only correct answer to this question.

EXANINATION QUESTION #96:

Which ONE of the following is the lowest level position responsible for ensuring entries are made in the Technical Specification Related Equipment Out-0f-Service Index?

a. Nuclear Plant Supervisor

b. Assistant Nuclear Plant Supervisor

c. Senior Nuclear Plant Operator

d. Nuclear Watch Engineer ANSWER: b

\

Applicant's contentions (summary): The applicant recommends that choice "d" be accepted as an additional correct answer because procedure 0-ADM-200,

" Conduct of Operations," makes the Nuclear Watch Engineer (NWE) responsible ,

i for routinely relieving the Assistant Nuclear Plant Supervisor (ANPS) of the control room command and control function to enable the ANPS to leave the control room. Consequently, the NWE becomes the lowest level responsible for making entries in the Technical Specification Related Equipment Out-of-Service i Index.

NRC Analysis and

Conclusion:

Procedure 0-ADM-213, " Technical Specification Related Equipment and Risk Significant SSC Out-of-Service Logbook," states that the ANPS is the lowest level position responsible for entering inoperable equipment in the subject index. When the NWE relieves the ANPS, he becomes the ANPS; he is not authorized to make entries in the subject index unless he is acting in the capacity of the ANPS. Therefore, the NRC concludes that choice "b" is the only correct answer to this question.

4

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SUMMARY

OF NRC REVIEW i

In summary, the NRC has concluded the following based on its review of the j applicant's contentions:

i QUfiTION NO. NRC CONCLUSIONS I I

24 This question has been deleted because there is no correct answer.

63 No change, choice "b" is the only correct answer.

84 No change, choice "a" is the only correct answer.

96 No change, choice "b" is the only correct answer.

APPLICANT'S ORIGINAL EXAMINATION GRADE: 78% (78 of 100)

FINAL NRC GRAQE1 78.8% (78 of 99)

The applicant's final grade of 78.8% remains below the minimum passing grade of 80%. Therefore, the NRC has concluded that the applicant failed the written examination.

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1 In response to the letter dated' September 12, 1996 applicant contends the following:

EXAM QUESTION #63 Plant conditions:

-Preparations are being made for refueling operations.

-The refueling cavity is filled with the transfer tube gate valve open.

-Alarm annunciators H-1/1, SFP LO LEVEL and G-9/5 CNTMT SUMP HI LEVEL are in alarm.

Which one of the following is the required immediate action in response to these conditions?

A. Verify alarms by checking containment sump level recorder and spent fuel level indication.

i B. Sound the containment evacuation alarm.

C. Initiate containment ventilation isolation.

D. Initiate control room ventilation isolation.

ANSWER: B The NRC analysis and conclusion contends that reactor operators and senior reactor operators are expected to analyze alarms and determine the appropriate course of action based upon specific plant conditions and indications.

Applicant contends that performing an action based solely on annunciation alone is not the proper way to operate. Even though SFP LOW LEVEL and CONTAINMENT SUMP HIGH LEVEL alarms are mutually supportive and sufficient to enter 3-ONOP-033.2 " REFUELING CAVITY SEAL FAILURE" The annunciators should be verified by additional supportive information to preclude the possibility of annunciator failure. Additionally CONTROL ROOM ANNUNCIATOR RESPONSE procedure 3-ARP-097.CR states that for all alarms the ARP shall be consulted. Applicant therefore contends that answer "A" verify alarms is also a correct answer.

j .

Procedure No.: Procedure Tele: PaSe:

, 8 i , Aprovat oase:

j . 3-ARP-097.CR Control Room Annunciator Response 6G96 1

5.0 SUBSEOUENT ACTIONS

._.11OTSC l

1)

ANNUNCIATOR RESPONSE GUIDELINES Unit ANPS/NPS SHALL be made fully awaretognizant.of all Annunciators at aH lI I l I limes (whether they have deared or are lockedin). 1I

.l l

2)- RCO- Upon recept of an annunciator, take immedunte corrective actions as lI I necessary, inforrning ANPS of any corrective actions. 1I

$ l I I

3) Daily AnnunciatorResponse Pro dure Usage:

i i II

\ l For expected alarms such as I&C woMng in Racks, actual opening of lI

. ARP' sis not required. =I i l ll

$ l For common or frequent alarms (WBP, Blender Deviation) use of the ARP is lI i .

required for the Hrst annunciation on the particular shiR for the day. =I l

I Subsequent annunciation does not require ARP consultation. lI

! l l For ALL other alarms the ARP SHALL be consulted as weII as any other =I

! W M P mcedures. !

l l._._._._._._._._._._._._._._._.

1 5.1 Annunciator on Panel A 5.1.1 Perform Appropriate Attachment 1, Page 15 j i 5.2 Annunciator on Panel B

! 5.2.1 Perform Appropriate Attachment 2, Page 69 i ,

l j 5.3 Annunciator on Panel C j

j 53.1 Perform Appropriate Attachment 3, Page 123 ,

j !

5.4 Annunciator on Panel D j 5.4.1 Perform Appropriate Attachment 4, Page 177 j 5.5 Annunciator on Panel E 5.5.1 Perform Appropriate Attachment 5, Page 231 j 5.6 Annunciator on Panel F l 5.6.1- Perform Appropriate Attachment 6, Page 285 i

I 5.7 Annunciator on Panel G 5.7.1 Perform Appropriate Attachment 7, Page 339 i

J

1/JGK/dt/dt/dt 1

l EXAM QUESTION #84 ;

1 Which one of the following is the basis for step 1, " VERIFY REACTOR' TRIP", of FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS?

A. To ensure that only decay heat and reactor coolant pumps are adding heat to the RCS.

B. To ensure shutdown margin is within technical specifications limits for hot standby.

C. To alert the operator to take further corrective action if the reactor is not tripped.

D. To verify that all automatic reactor protective features have functioned as designed.

ANSWER: A The NRC contends that the basis for E-0 step one and FR-S.1 step one are the same and that there is only one answer.

The applicant contends that FR-S.1 is a FUNCTION RESTORATION PROCEDURE And that it gives guidance to restore CRITICAL SAFETY FUNCTIONS. SInce the reactor was verified not tripped in E-0 step one you are sent to FR-S.1 Where the operator is directed to insert rods because the reactor is not tripped. Because FR-S.1 is a FRP and gives guidance the applicant contends that the basis for FR-S.1 is twofold, (1) to ensure only decay heats is added and (2) To direct corrective actions. Therefore the applicant asks that answer "C" also be accepted as a correct answer.

. l

. Procedura No.: Procedura Titis: Pd ge:

a 8

4 Emergency and Off-Normal Operating Approva! Dus:

0-ADM-211 Procedure Usage 8/23/95 4.0 DEFINITIONS 4.1 Action Verbs I j

All action verbs used in two-column format procedures are defined in 0-ADM-101, Procedure Writer's Guide.

1 4.2 Critical Safety Function i An activity which serves to protect the integrity of one or more of the physical barriers against radiation release.

J 4.3 Emergency Operatine Procedures (EOPs) l Plant procedures that specify the operator actions required to mitigate the consequences of transients and accidents that cause plant parameters to exceed reactor protection system setpoints, engineered safety features serpoints, or other appropriate technical limits. The EOP network consists of all Optimal Recovery Procedures and Function Restoration Procedures.

4.4 Faulted Refers to any steam generator with an unisolable leak in it's secondary pressure boundary of sufficient size to require Safety Injection.

4.5 Functional Restoration Procedures (FRPs)

Those procedures which respond to Critical Safety Function challenges.

Guidance is provided to restore the Critical Safety Function to a satisfied condition. Typically, actions are based on the severity of the challenge and may not correspond to " good operational practice". These procedures are identified by the procedure identifier F or FR.

4.6 Local (Locallv)

An action performed by an operator outside the Control Room.

4.7 Manual (Manuallv)

An action performed by the operator in the Control Room. This does not include automatic actions, which take place without operator intervention.

4.8 Optimal Recoverv Procedures (ORPs)

Those procedures which provide guidance to recover the plant in the most efficient manner to a safe and stable end state. Typically, actions correspond to " good operational practice". These procedures are identified by the procedure identifiers E, ES, and ECA.

/J8S/dt/dtir

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, Page 9 !

3-BD-EOP-E-0 REACTOR. TRIP OR SAFETY INJECTION 08/01/91

BASIS DOCUMENT

~

i WOG Procedure Step: 1 PTN Procedure Step: 1 l l

BASIS: I l

Reactor trip ':ast be verified to ensure that the only heat being added to '

i the RCS is from decay heat and reactor coolant pump heat. The safeguards systems that protect the plant during accidents are- designed assuming that j only decay heat and pump heat are being added to the RCS. If the reactor cannot be tripped, a transition is made to FR-S.1, RESPONSE TO NUCLEAR-POWER GENERATION /ATWS, to deal with ATUS conditions.

STEP DEVIATION FROM WOG GUIDELINE:

~

TYPE DESCRIPTION 8 The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

1 The RNO was changed so that a transition to FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS, will only be made if the criteria from the Critical Safety Function status tree for suberiticality is ~~ l satisfied. This provides the operator with a clear definition of what constitutes a reactor trip, and eliminates the need for the operator to make a decision under st.ress. This change complies with the intent i of the RNO column provided in ERG Feedback item DW-88-033. -

9 The WOG guidelines require initiation of Critical Safety Function status tree monitoring whenever exiting E-0. The RNO was modified to provide procedural guidance for performance of this task so that the need to memorize User's Guide requirements is eliminated.

PLANT SPECIFIC SETPOINTS:

5% -

Reactor power level just in the Power Range. (EOP Setpoint P.2) l l

t .

)

Page 8

/

, 3-BD-EOP-FR-S.1 i

RESPONSE TO NUCLEAR POWER GENERATION /ATWS 08/01/91 l

1 BASIS DOCUMFNT i

f j WOG Procedure Step: 1

, PTN Procedure Step: 1 i BASIS:

a Reactor trip must be verified to ensure that the only heat being added to the RCS is from decay heat and reactor coolant pump heat. The safeguards

- systems that protect the plant during accidents are designed assuming that '

only decay heat and pump heat are being added to the RCS. If the reactor cannot be tripped, then the control rods should be manually inserted into ,

j the core in order to decrease reactor power.

l STEP DEVIATION FROM WOG GUIDELINE:

i

TYPE DESCRIPTION

8 The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

3 PLANT SPECIFIC SETPOINTS:

i i

N/A 1

4 d

a 4

d 3

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a . _ . _ _ _

EXAM QUESTION #96 Please review this question as stated in the original request.

6 4

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THe following question was discovered to be wrong after the first I request for review was sent. The applicant wishes for the question to also be considered.

EXAM QUESTION #90 i

When draining the RCS using 3-OP-041.9, REDUCED INVENTORY OPERATIONS, the reactor vessel head and pressurizer are both vented to containment atmosphere.

Which one of the following describes the effects on reactor vessel level indication if an adequate vent path is not provided?

(Assume the reference leg remains full).

A. A vacuum in the res loops will result in level indication being lower than actual levels.

B. A vacuum in the rcs loops will result in level indication being higher than actual levels.

C. A positive pressure in the res loops will result in level indication being lower than actual levels.

D. The level instruments automatically compensate for positive or negative pressure.

ANSWER: A

REFERENCE:

3-OP-041.9, REDUCED INVENTORY OPERATIONS, Page 25, Section 5.2.2.3 Caution E.O. 3 OF LP-6902121 The assumption that the reference leg remains full makes this question invalid. At Turkey Point the drain down level indication has dry reference legs. This condition is verified by 0-FMI-041.110. Applicant requests that this question be deleted.

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

. Procedure No.: Proceoura Tas:

Page:

)

s 25 i Approv2 Date

j 3-OP-041.9 Reduced Inventory Operations 6/6/95 i !

!' INITI A LS j CK'D VERIF 5.2.1 (Cont'd) i 6. Both RHR Pump Discharge Isolation valves have been j

throttled to limit maximum RHR System flow to 3200 gpm.

}

, a. RHR Pump A DischIsol,3-754A i

i b. RHR Pump B Disch Isol,3-754B i

! 7. RHR Ex Bypass Flow valve, FCV-3-605, has been adjusted to i

maintain between 3100 and 3200 gpm RHR flow.

i

8. Verify one Source Range Nuclear Instrument audible i

count rate is on in the Control Room when fuel is in the i

ReactorVessel.

5.2.2 Procedure Steps

1. Station an operator at Drain Down Level Indicator Hose, LI-3 3422 and verify direct communication with Control Room in order to commence logging level every 15 minutes using Attachment 1. [ Commitment - Steps 2.3.6 and 2.3.8]

2. Commence logging reduce inventory parameters using Attachment 2.

3. Place Letdown Diversion Valve, TCV-3-143 to DIVERT.

CAUTIONS e

RCS levelindication may be lower than actuallevel during RCS draining unless large ventpaths are provided.

e RCSlevelindication is connected to Loop A intermediate leg. At high RHR flow \

rates, the indicated level will be different than actuallevel at the RHR hot leg 1 suction. Refor to Enclosure 2 for minimum required RCS levelindication.

4. Verify open RHR Letdown Stop,3-205B.

5. Open RHR LTDN to CVCS, HCV-3-142.

6. Throttle Low Pressure LTDN Controller, PCV-3-145 as necessary to maintain RCS Drain Down Level indication on LIS-3-6421 within 4.0 percent of LIS-3-6423 during RCS draining. [ Commitment - Step 2.3.8]

st:2dBS/vm/mbp/bc

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! Florida Power & Light Company i l Turkey Point Nuclear Plant *2NEWj cs '

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Asis ec 3s5 1 i

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-h 0-PMI-041.110

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et i e.t sEG S sii R6 &

r32 5&a

Title:

RCS Drain Down Level Calibration Safety Related Procedure Responsible Department: Maintenance Revision ApprovalDate: 3I619 6 Periodic Review Due: 8/29100 RTS 95-0620. 96-0144 OTSC 0450-95 PC/M 95-150

/OK/ev/sw/ev

Proceoure No.: Procecure Titta: 1 Page:

J Approval Dite:

0.PMI-041.110 RCS Drain Down Level Calibration 4

8/30/95 i i

INITI A L S

CK'D VERIF 6.2 Level Transmitter LT-6421/23 Calibration Check l

. C A UTION i

Care shall be taken not to break the neck seal between the sensor .,odule and the electronics housing.

t*****""'*""*""**-*-*""**"*"""*----*"T'"ES'"-*-*"'"*****"*""*""*"""*"""*-*1 I

NO .

I l.

Transmitter output Test terminals (local) are located inside the transmitter housing. l To gain access, housing side cover identified as Terminal side (see nameplate) .

, I. must be removed. I I

. . l I

Zero and Span adjustment screws are accessible externally and are located behind the transmitternameplate. I 1

' l

. . l J The transmitter output increases with clockwise rotation of the adjustment screw.

l

' 1 l*

O-rings shallbe replacedif housing coveris removed.

l I

i Level Transmitters LT-642W23 are located inside containment on the 14 foot level I l outside the bio-wall.

. l 3

s...................-..-...J LT4421 LT4423

. 1

6.2.1 Shut transmitter high, low, and equalizing valves at 3-valve manifold.

6.2.2 Open transmitter vent valves slowly to release any trapped pressure.

l r..-...........-....-.-..-.-.-..,

NOTES .

I i

l. Transmitter and impulse line liquid contents should be collected in accordance l with RWP requirements.

I l

l. Equalizing valve should never be opened as low side is dry.

l

6. . . . . . . . . . . . . . . .-.-.-.-.-.-.-.-..a 6.2.3 Remove caps from test fittings.

6.2.4 Allow both sides of transmitter to drain.

6.2.5 Connect pressure source and test gauge to high pressure test fitting.

6.2.6 Close vent valve on high pressure side.

/OK/ev

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

Procedur3 Nr Procmura

Title:

Page: I s

1 0-PMI-041.110 Approval Date:

RCS Drain Down Level Calibration 2/22/96 INITI A L S i CK'D VERIF i

! LT-6421 LT4423

)

t 63.7 Ins:all cover and hind tighten.

i 6.3.8

Attach special transmitter cover tool to torque wrench.

l INDEPENDENT VERIFICATION POINT Independent Verifiershall

Verifyproper torque application in Step 6.3.9. I

\

Initialappropriate space on Data Sheet.

,6.3.9 Torque cover to 200 in-lb.

Acceptance Criteria:190 in-lb to 210 in-lb f 6.3.10

~

Record O-ring part number and attach QC tag to PWO.

63.11 Disconnect and remove pressure test set.

1.

Replace defective tr,1smitter test fittings (i.e., Swagelok), if required.

6.3.12 Reinstall test fitting caps and tighten properly.

6.3.13 Reinstall vent valve caps finger tight, do not torque at this time.

6.4 Placine Level Transmitter in Service 6.4.1 Verify , that pressurizer.

Operations has established a vent path through the 6.4.2 Remove the cap and connect a hose or place a poly bag to catch any fluid from the dry leg low point drain valve below the transmitter three-valve manifold.

6.4.3 Open the dry leg low point drain below the transmitter three-way valve.

6.4.4 Place a poly bag to catch any fluid from the three-valve I manifold on test tee downstream of PRZR Safety Valve t

-551 A (*-551B) Loop Drain *-545A (*-546A). !

/OK/ev/dplsw

' ~ ~

Proceoura No.: Proceours Tale:

Page: '

, s i Approvat oats 0.PMI.041.110 RCS Drain Down Level Calibration 2/22/96

-m j- INITI ALS l 3 CK'D VERIF 1

LT4421 174423 i

' ' 6.4.5 Disconnect dry leg tubing at B valve on three-valve manifold I or remove test 'T' cap downstream of *-545A (*-546A). I

,._............._........._.3 NOTE I .

I t

l !' test tee is used, have Operations close 'B' valve on three-valve manifold.

l s........................a.

6.4.6 Connect a source of dry nitrogen or instrument air to the (

test tee on dry leg tubing on the 58 foot level and blow down I

to the drain valve on the 14 foot level until all moisture I is removed from the line. I 6.4.7 If test tee was not used, disconnect the blowdown connection i I from the dry leg tubing. l NOTE 3 i

i -

I l If test tee is used, have Opbrations close "B' valve on three-valve manifold.

. l j

~

s.........._..............a.

6.4.8 Place a poly bag to catch any fluid below the pipe cap or

I i drain valve downstream of *-545A (*-546A). I !

! 6.4.9 Remove the pipe cap or open the drain valve downstream of

I

-545A (*-546A) and drain any moisture. I 1 6.4.10 Replace the pipe cap or close the drain valve downstream of I

-545A (*-546A). I k,

6.4.11 If test tee was not used, connect a source of dry nitrogen I

or instrument air to the B valve on three-valve manifold I downstream of *-545A (*-546A). I

l. Blow into the pressurizer until all moisture is removed from I the line.

6.4.12 Disconnect the blowdown connection.

I 6.4.13 Reconnect the dry leg tubing to the B valve connection or I replace test "T" cap on the three-valve manifold. I j 6.4.14 Verify closed equalizing valve on threeJ/alve manifold at transmitter.

6.4.15 Remove cap from transmitter low side vent.

loKlev/dja/sw 4

- - _ . _ . _ - . ~ - - - __

ProcedurJ No.: Procecura

Title:

- Page:

16 0.PMI.041.110 Approval Date:

RCS Drain Down Level Calibration 2/22/96 i

INITI ALS i CK'D VERIF '

LT4421 LT4423 i

l

._._._._._._._._._._._g.,_._._._._._._._._._._.7 I Dry nitrogen or instrument air may be used to assist in draining the low side of the.I transmitter. Direct allflow down from transmitter v I l . _ . _ . _ . _ . _ . _ . _ . _ . _ . _ . _ . _ . _ .ent to lowpoint drain. ,

_._._._._._._._._._.J 6.4.16

> Slowly open low side isolation valve and allow transmitter to drain.

6.4.17 4

Close the dry leg low point drain valve below the transmitter three-valve manifold.

6.4.18 Remove the hose if used, and replace the cap on the low point drain valve.

6.4.19 Replace cap on transmitter low side vent.

6.4.20 Remove cap on high side test tee above transmitter three-valve manifold. Install a length of hose from the test tee into a poly bag.

4 I

._.._.._._._._..._.) .

It may be necessary to bleed several gallons of fluid before all air is removed from the l I line. .

I

)

t._._._._._._._._._._._._._._._._._._._._._._._.J :

6.4.21

. Slowly open the high side isolation valve and drain RCS fluid I i

into poly bag until all air is removed from line. Open the valve i fully to obtain the maximum flow rate.

I 6.4.22 Close the high side isolation valve.

6.4.23 Remove the hose and replace the test tee cap.

6.4.24 Remove the transmitter high side vent cap. -

6.4.25 Slowly open the high side isolation valve and fill the transmitter.

A 6.4.26 Close the high side imation valve.

6.4.27 Replace cap on transmitter high side vent.

loKlevidja4w

ML20129D547

Contents

Text

Dear Sir:

Dear Sir:

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REFERENCE:

REFERENCES:

REFERENCES:

REFERENCE:

REFERENCE:

REFERENCE:

Dear Mr. Tetrick:

Enclosure:

Conclusion:

Conclusion:

Conclusion:

Conclusion:

SUMMARY

REFERENCE:

Title:

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Title:

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ML20129D547

Text

Dear Sir:

Dear Sir:

REFERENCE:

5 m '5" ma-*" yi(, .<=A. a=

REFERENCE:

REFERENCES:

REFERENCES:

REFERENCE:

REFERENCE:

REFERENCE:

Dear Mr. Tetrick:

Enclosure:

Conclusion:

Conclusion:

Conclusion:

Conclusion:

SUMMARY

REFERENCE:

Title:

Title:

Title:

Navigation menu

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