ML20205P365

Forwards Util Comments Re Operator & Senior Operator Exams Administered on 861117.Specific Changes to Listed Questions & Responses Recommended

ML20205P365
Person / Time
Site:	Summer
Issue date:	11/21/1986
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Arildsen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
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ML20205P271	List: ML20205P268 ML20205P311 ML20205P365
References
NUDOCS 8704030248
Download: ML20205P365 (29)

Text

ENCLOSURE 3 gth oNna Electric & Gas Company Da A. n 1 *D Colum 29218 Nuclear Operations SCE&G a . .co. -.,

November 21, 1986 Mr. Jesse Arildsen License Examinor U.S. Nuclear Regulatory Commission Region II, Suite 2900 101 Marietta Street, N.W.

Atlanta, Georgia 30323

SUBJECT:

V.C. Summer Nuclear Station Docket No. 50/395 License No. NPF-12 Operator License Examinations

Dear Mr. Arildsen:

Enclosed are the utility comments to the NRC Operator and Senior Operator examinat{ons administered at V.C. Summer on November 17, 1986. Your consideration is appreciated.

Very truly you"s, D

. A. Nau RCR: DAN:dgb Enclosures cc: J. G. Connelly, Jr. (w/o attachment)

O. S. Bradham (w/o attachment)

M. B. Williams (w/o attachment)

K. W. Woodward (w/o attachment)

A. M. Paglia (w/o attachment)

NPCF File F. Jagger (w/ attachment)

$0 V

ATTACHMENT

• t ,,

QUESTION: [ 0. 3]

l.16 From the words or phrases in parentheses, choose the one which correctly completes each statement below,

d. Calculation of Enthalpy Rise Hot Channel factor assumes core power is (uniform /or/not uniform) and flow through each channel is (the same/or/different) throughout the co re . (0.6.)

ANSWER KEY RESPONSE:

d. Not uniform [0.3] The same [0 3]

REFERENCE:

GP HTFF, pp. 193-198 - Thermal Hydraulic Principles, pp. 13-30 thru 36 CNS Exam Bank SUGGESTED CORRECT RESPONSE:

d. Not uniform [0.3] Delete second part.

REASON: While it is true that the calculation of enthalpy rise hot channel factor assumes the flow rate through each channel is the same (see attached reference TS-13, Design and Operational Charts Limits on Thermal Parameters of the Reactor Core, p. 29 ) , a four percent penalty is used to account for non-uniform distribution of flow (see same reference as above, p. 30, attached). This penalty implies different flow through core channels.

i , .

T t

point during normal operations. These limits on F also g '

. ensure that cladding temperatures will not exceed the self-

sustaining zirconium-water reaction temperatures if a loss of T

l coolant accident and DNS conditions occur. F g, however, does l

! not ensure that DNS conditions will not occur during normal plant

) operations. In order to prevent DNS, limits are imposed on the enthalpy rise hot channel factor.

l Departure from nucleate boiling depends not only on the power density of the fuel, but also on the local enthalpy of the coolant l and the local mass flow rate of the coolant. The fuel rod l i ultimately transferrMg the maximum amount of heat to each unit mass of coolant flowing upward through the core is the rod where DNS is most likely to occur.

~

Since power density is not the only factor determining the onset of DN8, an evaluation of the total enthalpy rise along the flow channel must be made to ensure that DN8 conditions have not been met. In order to calculate the total enthalpy rise along a flow channel, the integrated ;ower for the entire length of the fuel

! rod is evaluated with the channel inlet enthalpy and the flow rate j through the channel.

Since power density throughout the core is not uniform, there is a fuel rod in the core with the highest integrated power.

l Assuming the inlet enthalpy of all channels and the flow rate through, each channel is the same, the highest coolant enthalpy rise occurs along this rod. The flow in this channel will be closest to DNS.

l I

The ratio of the maximum integrated rod power to the average rod power is defined as the enthalpy rise nuclear hot channel factor (F )

I 29 OD29V 1

l

i ,.

N maximum intearated rod power

- IAH " average integrated rod power i

Based on the worst case power distribution and manufacturing tolerances allowed during fuel rod fabrication, the limit imposed on F is equal to 1.55. This limit ensures a minimum DNBR of 1.3 at 100 percent power.

F3 " 1 1.55 Similar to the heat flux hot channel factor, F is measured using the incere instrumentation system. A four percent '

penalty is used to account for the fact that the highest power rod in the core may not be measured. Additionally, another four_ percent _

penalty is used to account for non-uniform distribution of flow.

j Using these penalties, the limit on the measured F AH

~

reduced to i

! N 7

FAH" (measured) = 1 - 1.435 If power is less than 100 percent, the fluid enthalpy rise is lower. With a smaller enthalpy rise, the coolant is further away ,

from a DNS condition and a higher F is allowed. Analyzing I worst case radial and axial flux profiles at various power levels produces the limit ecuation F "Ha 1 1.55 [1 + .2(1 - P)]

where P = fraction of full power The graph of the limit of F as a function of power is shown in Figure FND-HT-101. The shaded area represents the allowed range

of F a to ensure a DN8R > 1.3.

30 0029V i

I

- .. ,i QUESTION: [1i50]~ 1 2.22 Of the signals used to automatically;close the Feedwater

~

Isolation Valves (1611A, B, C), which two are specifically designed to' prevent water hammer.in the feedwater piping and steam generator inlet connections. Include setpoints and coincidence (logic) where applicable.

ANSWER KEY RESPONSE:

1.- Lo-Lo-Lo S/C level [ 0. 2] -- 2/3 [0.13 --- 5% [0.1].

2. The signal which is the coincidence oft [0.1]  ;

a. Lo S/G pressure [0.2] -- 2/3 [0.1] -- 625 psig

[0.1]. x s

b. Lo feedwater temperature [0.2]'< 225 F [0.13

c. Lo feed flow [0.2] < 20% [0.1].

REFERENCE:

VCS, TB-7, p.22 SUGGESTED CORRECT RESPONSE:

s Either the above answer or:

1. Lo-Lo-Lo S/G level [0.2] -- 2/3 [0.1] -- 5% [0.1].

2. The signal which is the coincidence or! [0.1].

a. Lo feedwater temperature [0.3] < 225*F [0.2]. .

i

b. Lo feed flow [0 3] < 13% [0.2]. '

REASON: Recent plant changes have removed the lo isteam pressure interlock and changed the lo feed flow setpoint to < 13%.

(See attached B-208-045 electrical drawing, notes ' 10 &

11). Training materials have not been revised to-reflect these changes..

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QUESTION: [2.00]

3 12 a. Describe the instrument coincidence and setpoints necessary to insert the Reactor Coolant low flow trips into the protection system for 1 and 2 loop loss of flow.

~b. Explain how and why the undervoltage and underfrequency low flow reactor trips operate differently. t ANSWER KEY RESPONSE:

a. 1. Single loop loss of flow will insert if 2/4 power ranges are above 38% (P-8). (0.5).

I

2. Two loop loss of flow will insert if 2/4 power ranges

> 105 [0.2] or; [0.1],1/2 impulse pressures > 10%

. [ 0. 2 ] .,

b. The undervoltage trip is to provide a trip signal on loss-of power to the RCP's. [ 0. 2 ] The reactor will trip and RCP's continue to provide coastdown flow. [0.2] The underfrequency trip provides protection for a grid disturbance. [0.2] The RCP breakers and reactor are tripped [0.2] to prevent deceleration and loss of coastdown flow. [0.2]

REFERENCE:

VCS IC-9, pp. 46, 47 SUGGESTED CORRECTED RESPONSE:

Either the above answer or:

a. 1. Single loop loss of flow will cause a-trip if 2/3 flow elements in one loop sense less than 90% flow with reactor power above P-8 (38%). [0.5]

2. Two ;'op loss of flow will'cause a trip if 2/3 flow .

elementt *n each of two loops sense less than 90%

flow with reactor power above F-10 (10%). [0 5J

REFERENCE:

VCS 10-9, pp. 46 (attached)

^

REASON: Because of the length and wordiness of the question, it lends itself to different interpretations of what is being asked.

b. While the flow trip provided by the loop flow transmitters supplies the basis protection required when flow is lost in one or two loops, undervoltage and underfrequency sensors associated with each RCP electrical supply provide the more rapid protection ,

required if flow is lost in l l

l

'l

- - .-..s , , . ,.c - .,-c - - . , ,

QUESTION: (3.12b continued) [2.00]

two or more loops. (Reference 10-9, Reactor Protection and Logic, pp. 46, 47, attached ) . These rapid loss of i flow reactor trips are provided to protect against DNB.

Reliance upon the actual loss of flow trip does not provide adequate protection against DNB if flow is lost-i in two or more loops, so the anticipating underfrequency l and undervoltage trips are required. [1.0] (Reference VCS

! FSAR, p. 15.3-6).

l l The above answer is offered as an alternate because the explanation described by question 3.12b could be interpreted as a requirement to explain how and why UV.

and UF reactor trips operate differently from the low flow trips mentioned in part a of the same question.

1 l

l i

1

- . = . _. _.

a. ..

Primary Coolant Sysi!em Trips low Reactor Coolant Flow (Sh. 5)

Three redundant differential pressure transmitters are used to measure the flow in the cold leg of each reactor coolant loop. The low-pressure taps for each transmitter penetrate the RCS at the inner piping elbow located downstream of the steam generator primary outlet. The high-pressure taps for each transmitter join at one common line that penetrates the RCS at the outer piping elbow.

Each of the transmitters for a loop has a low-flow bistable associated with it. The bistables are set to trip if the differential pressure being sensed falls to a value that corresponds to less than 90 percent loop flow. If 2 of 3 bistables in a loop trip, a single loop loss of flow signal is sent to the Solid State Protection System. A reactor trip will occur only if 2 of 4 power range channels exceed the P-8 setpoint (38 percent). If a loss of flow con-dition is sensed in more than one loop, a reactor trip will occur if 2 of 4 power range channels exceed the P-7 (P-10 and P-13 ) . Conversely, the reactor will not trip on any condition of low flow when the plant is operating below P-7.

Reactor Coolant Pump (RCP) Undervoltage (Sh. 5)

While the flow trip provided by the loop flow transmitters supplies the basic protection required when flow is lost in_on_e or two loops, undervoltage_sen. sors.

aNociated with each RCP electrical supply provide the more rapid protection required if flow is lost in two or more loops. Each of the RCP electrical buses is supplied with three undervoltage sensing devices. These W sensors are located on the motor side of the RCP circuit breakers to detect a W condition should the breaker (s) open. When all three W sensors associated with one bus sense that pump supply voltage has dropped to 70 percent of normal (4830 volts),

a W signal exists for that bus. Signals from the W sensors are sent through time delay energize devices. This prevents spurious reactor trips caused by  !

l short-term voltage perturbations. The W trip is one of the low-power trips.

As such, the trip is automatically blocked when the plant is operating below the P-7 setpoint.

l 46 1155S:4 Rev 2 6/85 l

i i

's ..

_ Reactor Coolant Pump Underfrequency,(Sh. 5)

There is one underfrequency sensor monitoring the frequency on each of the RCP.

The trip is designed to provide fast response to a low-flow condition in two or more loops. A major disturbance on the grid could result in a frequency reduction. Since the RCPs are driven by ac induction motors, pump speed and

[ thus primary flow would be dragged down by such a frequency condition. If the sensor detects an underfrequency of 57.5 Hjz on any of the RCPs it produces

! an underfrequency signal . A reactor trip will occur if reactor power is greater than P-7. '

l The underfrequency condition also causes the RCP breakers to trip. This allows the pumps to achieve the coastdown time duration that is analyzed in the loss of flow analysis. With the plant operating below P-7, the pump bus

~

underfrequency reactor trip is automatically blocked. The signals from the ,

underfrequency sensors pass through time delay energize devices so that short-term frequency dips will not cause a reactor trip.

l

. Overtemperature AT Trip (Sh. 5)

This plant trip prevents the core from approaching the condition of departure from nucleate boiling (DNB). A setpoint is continuously and separately calcu-lated for each of the three loops. The setpoint is in terms of percent of rated full power loop AT. It basically takes into account the factors that affect the margin to DNB. Should the actual loop tenperature difference exceed the present loop setpoint, an OTAT trip signal will exist for that loop. If 2 of 3 loops exceed their setpoint, a reactor trip occurs. The  ;

OTAT trip cannot be bypassed.

Overpower E Trip (Sh. 5)

This trip is designed to limit core power and thus minimize fuel tempera-tures . Again, an individual setpoint is continuously calculated for each loop. The setpoint is in terms of percent of full power loop AT. The parameters used to determine the limit are T0 avg? and axial flux difference.

The input from axial flux difference is not presently being used for setpoint 1155S:4 47 Rev 2 6/85. I i l l

15.3.4 COMPLETE IASS OF PORCED REACTOR COOLANT FIAW 15.3.4.1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps. If the reactor is at power at the time of the accident, the_inunediate_effect, of loss of coolant flow is a rapid increase in the coolant _ temptrAtutex This increase could result in DNB with sub_s_equent fuel damage,_i_f_the reactor were not tripped,promptly. The following reactor tr_ip,s_ provide necessary protection against a loss of coolant flow accident:

l

1. Reactor coolant pump power supply undervoltage or underfrequency.

i 2. Low reactor coolant loop flow.

The reactor trip on reactor coolant pump undervoltage is provided to protect against conditions which can cause a loss of voltage to all reactor coolant pumps, i.e., station blackout. This function is blocked below approximately 10 percent power (Permissive 7).

The reactor trip on reactor coolant pump underfrequency is provided to trip the reactor for an underfrequency condition, resulting from frequency disturbances on the power grid. Reference (7] provides analyses of grid frequency disturbances and the resulting nuclear steam supply system protection requirements which are generally applicable to current generation Westinghouse plants.

The reactor trip on low primary coolant loop flow is_provided to_ protect, against loss of flow condittons which affect only one reactor coolant loop.

~

This' function is generated ~by two out of three low flow signals per reactor coolant loop. Above approximately 39 percent power (Permissive 8), low flow in any loop will actuate a reactor trip. Between approximately 10 ,

percent power and 39 percent power (Permissive 7 and Permissive 8), low i flow in any two loops will actuate a reactor trip. If the maximum grid frequency decay rate is less than approximacley 2.5 Hz/second this trip function will protect the core from underfrequency events. This effect is fully described in Reference (7].

i Normal power for the reactor coolant pumps is supplied through buses from the unit auxiliary transformer which is connected to the main generator downstream of the main generator breaker. Continuity of power to the pump buses is maintained by backfeed from the main power transformer in case of a trip of the main generator output breaker. On a loss of the main power transformer, the buses are automatically transferred to the emergency auxiliary transformer.

15.3.4.2 Analysis of Effects and Conseeuences J

15.3.4.2.1 Nethod of Analysis This transient is analyzed by four digital computer codes. First the-PHOENIX [8] Code is used to calculate the loop and core flow during the 15.3-8

QUESTION: [0 5]

4.14 During a serious emergency,. operators may be called upon to assist in search and rescue or recovery operations in the-plant.

a. In such cases, according to Health Physics manual, what dose could your receive:

1. To bring an injured worker to safety? [.05]

ANSWER KEY RESPONSE:

a. 1. 100 REM [0.5]

REFERENCE:

VCS HP Manual, pp. 5-13 SUGGESTED CORRECT RESPONSE:

a. 1.' 100 REM or.75 REM [0 53 REASON: Emergency Plan Procedure limits on exposure to save a human life differ -from the same exposure limits described in the HP manual.- (Reference EPP-Oll, p. 1 of 5, and EPP-020, p. 2 of.5 -- attached).

l

i '

EPP-011 REVISION 4 7/14/86 1.0 PURPOSE To provide guidelines for performing a search and rescue j' operation to find missing individuals.

2.0 REFERENCES

2.1 " Virgil C. Summer Nuclear Station Radiation Emergency Plan."

2.2 NUREG 0654, Criteria for preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants.

l 2.3 EPP-009, "On-Site Medical".

2.4 EPP-020, " Emergency Personnel Exposure Control".

2.5 EPP-008, "Oncite Assembly" -

l 2.6 EPP-012, "Onsite Personnel Accountability and Evacuation" l l

2.7 EPP-016 " Emergency Pacilities Activation and Evacuacion"  !

($)' 3.0 CONDITIONS AND PREREQUISITES 3.1 Personnel accountability nas been performed and it has been determined that a person on-site may be missing. This i procedure assumes that a missing person (s) is injured until proven different.

3.2 At least one memoer of the Search and Rescue Team will be qualified in Health Physics monitoring techniques and one i

member will be at least Multi-media First Aid qualified. I 3.3 The leader of the Search and Rescue Tbam should be the "B" Nuclear Reactor Operator or his designee as availaole.

Other members of the Search and Rescue Team should be:

available Auxiliary Operators, Maintenance Personnel, Chemistry Pe'esonnel, etc. , as per EPP-016 Attachment II.

3.4 If necessary, the members of the Search and descue Team mag, be authorized to receive 75 rem to save the life of the missFn~g fifdividdalfsT~iiis per Reference 2.4.

PAGE 1 0F 5

H- -

l EPP-020 REVISION 6 5/27/86 3.2.1 Normal Administrative exposure limits =ay be suspended or modified verbally by the Interim Emergency Director / Emergency Director, Radiological Assessment Supervisor, or Shift Supervisor. Verbal exposure extensions should be followed-up with the appropriate documentation as defined by reference 2.4. .

3.2.2 Federal limits of 10 CFR 20 will not knowingly be exceeded, except as described in Section 3.2.4 of <

this procedure.

l 3.23 Maintain exposure ALARA. )

J 3.2.4 Planned exposures of up to 75 Rem to save human life l

~or prevent the serious endangerment of_ human lifA or up to 25 ReE to save or delgate significant damage to vital equipment and/or facilities must be approved by the Interim Emergency Director / Emergency Director or Radiological Assessment Supervisor. -

Persons performing the planned actions should be SCE&G employees who are volunteers broadly familiar l with exposure consequences. '

l NOTE: This does not apply to " spontaneous reactions" by on the scene personnel in a threatening situation. ,

O 3.2.5 Non-company employees such as medical support i

personnel should be limited to:

a. 3 Rem if there is an adequate number of personnel so that rotation may be accomplished.

b 5 Rem if the number of personnel is limited such

• i that rotation can not be acomplished. 1

c. 25 Rem to save a life.

. 3.2.6 If an individual receives an annual dose equivalent in excess of twice the Annual Dose Equivalent Limit, ,

the case should be referred to a medical physician for review ~(as per ICRP Report 26).

PAGE 2 0F 5 ,

I

QUESTION: [0.25]

4.15 c. To what value can the Administrative Guideline for whole body be raised to and who (by title) must approve this limit increase?

A"SWER KEY RESPONSE:

, c. 2 R/qtr.:not to exceed 5 R/yr. [0.25]

i Plant Manager j

REFERENCE:

VCS HP Manual,, pp. 5-13 SUGGESTED CORRECT RESPONSE:

Either the above answer or:

c. 2.5 R/qtr. not to exceed 5 R/yr. [and 5(N-16)J. [0.25]

} Director, Nuclear Plant Operations i REASON: Health Physics Procedure - 153, Administrative Exposure

Limits, contain'the 2.5 R/qtr. limits. (See attached HPP-153, p. 3, and Attachment I, p. l'of-1). Also, the title Plant Manager has been changed to Director, Nuclear.

1 Plant Operations. (Attached reference HP-153, Attachment I. p. 1 of 1).

l f

1 l

l I

l

y . _ _ .

HPP-153 REVISION 5 10/9/85 d

4.0 GENERAL 4.1 The current quaEterly allowed exposure for each individual is listed on up-to-date printouts available in the Health Physics Lab and at key locations in the plant.

4. 2 The limits addressed within this procedure shall be enforced via the use of Self-Reading Dosimeter Cards, as describ ed in HPP-152. These cards reflect current accunulated quarterly exposure received and the remaining ,

( quarterly exposure allowed.

4.3 Additional controls established to ensure that administrative limits are not exceeded include:

, 4.3.1 Review of all current exposure records by Health Physics.

j 4.3.2 Computer printouts (" Daily Dose Printout") of l

individual exposures for personnel to review. .

4.3.3 Reports forwarded to Managers and Supervisors t6 inform them of personnel who are approaching quarterly exposure Ibnits.

1

, 4.4 Health Physics performs audits of exposure extension i

(s%) requests to identify cases where supervision is requesting

? higher extensions of administrative limits than necessary for the performance of the work. Upon identification of such cases, the responsible supervisor will be contacted to justify these extension requests.

5.0 LB4ITS AND PRECAUTIONS 5.1 Exposure limits for new personnel on site may vary depending on the individuals prior site, current quarter exposure and will be set as described in Reference 2.1.6.

Without an extension, the administrative limit for whole l 5.2 l body exposure received at the V. C. Summer Station is 1.0 ' )

- Rem /qtr. Female employees without a doctors verification t )

of infertility will be 10mited to 500 mrem /Qtr.

5.3 Quarterly Administrative Limits are reviewed to ensure no, individual will exceed 2500 mRC4 during a quarter arter ,

_ previous site dose is added to the V.C. Summer Station (

allowable Administrative Dose for the quarter. The 1

-l Quarterly Administrative Ltnits are also reviewed to ensure l

j

!' the annual administrative limits (5000 MREM for SCE&G 1

! personnel-'aiiF1FOT6 iiiREM for contractors) and liretime l

p)

( exposure limits (5 x (age-18)) are not exceeded.  ;

'j PAGE 3 0F 5 GENERAL REVISION l

(

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! \

l HPP-153 ATTACHMENT I PAGE 1 0F 1 REVISION 5 AUTHORIZATION TO EXCEED QUARTERLY ADMINISTRATIVE EXPOSURE LIMIT Added Exposure needed by Da t e/ Tim e '

Nrmer SSN:

Company: TLD: DOB(N):

g TM/ MUN/U A Y N Current V.C. Summer Qtr. Limit: mR I

T This request is to extend the individual's whole body exposure tot mR I

A YE S NO b T . .

} O 1 1 I L_J Is this extension or the Administrative Whole Body Ex po sure limits absolutely i

R . . . . necessary?

2 I I I I Has every reasonable effort been made to maintain this individuals exposure to within his current administrative limit?

3 I I I I Has ex posure been distributed am ong workers to the extent possible?

t 4 i i i i Are other qualified individuals with less ex po sure available?

5 I i i t Will this exposure extension minimize collective ex po sure?

Purpose:

Associate Man ag e r/ Supe rv iso r Date/Ilme Total Exposure Cur r ent to: Verified By:

Uste/Ilme

~

This qtr.

L this site mR mR T previous site. mR H To t al mR 2500 mR -

P This year (excluding c'ur r e n t qtr.) mR mR H Previous Ye ar s ~

mR Y Total Lifetime to date mR mR 5 (N -18 ) '

S Of Current Quarter To t al mM rocket Dosimeter Latimate mR TLD I

'C Co uld this exposure result in the individual receiving greater than S 5000 mR for the current year (including this qtr) yes no If the answer is yes, then the initiator will complete and include At t a chm ent II.

S This individual's Whole Body Exposure Limit is extended to mR  ;

I C

N s-A In d iv id ual Da te/ line Associate Manag er , He al t h Physics ate /ilme T

U R Supervisor Daterline Group Manager Date/Ilme E

m Associate Manag er/ 5upervisor of Ops Date/line Director, Nuclear Plant 0g3/ _

Date/Ilme

Manager Da te/ lim e Vice President, Nuclear Operations Da te/ Ilm e w- -- - - , - , - - - - - . ---n , , , , , , - - .

QUESTION: [ 2.0]

~

4.16 According to SOP-401, Reactor Protection and Control System, list four' of the five Instrument Channels that need not be -

placed in the trip mode when they are declared to be ou t-of-se rvice .

ANSWER KEY RESPONSE:

1. Source Range Excores

2. Intermediate Range Excores

3. Containment Pressure (High-3)

4. RWST Levels 5 Emergency Feedwater Suction Pressure Low.

[any 4, 0.5 each]

REFERENCE:

VCS, SOP-401, p. 4 SUGGESTED CORRECT RESPONSE:

1. Source Range Excores

2. Intermediate Range Excores 3 Containment Pressure (High-3)

4. RWST Levels 5 Emergency Feedwater Suction Pressure Low

6. RCS Flow Protection (Underfrequency)

7. RCS Flow Protection (Undervoltage)

8. Control Channel THOT RTD 9 Control Channel TCOLD RTD ,

2

10. Pressurizer Pressure Control l

[any'4 8 0.5 each]

REASON: (See attached SOP-401 pages). None of the additional responses (6-10) require tripping-bistables.

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

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SOP-401 REVISION 6 2/21/do B. CONTROL CHANNEL HOT LEO RTD FAILS LOW 1.0 SYMPTOMS 1.1 Abnormally low indication on one of tne following indicators:

a. TI-411A, T-A/G*F.

b. TI-421A , T-AVJ'F.

c.. TI-431A, T-AVG'F.

1.2 Aonormally low indication on one of the following indicators.

a. TI-411B, T 4.

b. TI-4218, T %. ,

c. TI-4313, T%.

13 dCS TAVG DEV dI/LO annunicator.

1.4 RCS T DEV HI/LO annunicator.

2.0 AJTOMATIC ACTIONS None 3.0 IMMEDIATE CORRECTIVE ACTIONS 31 verify instrument failure by comparing control channel T and Tavs indication with Protection Channel T and Tavs indication for the faulty loop.

PAGE 15 0F b3

SOP-401 REVISION 6 2 /21/do 4.0 FOLLOWGP ACTIONS 4.1 Place dI f AdC INdI8IT switch to the loop wnien nas tne failed instrument.

4.2 Place HI TAVG AJC IddIBIT switch to the loop witn the failed instrument.

NOTS 4.3 RCS T DEV HI/LO annunicator will not clear and the indications will remain abnor.nal until the channel is repaired.

43 Confirm that RCS TAVG DEV HI/LO annunicator clears.

4.4 Have the failed channel repaired and verify proper '

operation prior to returning it to service.

PAGE 16 0F 63

802-401 dd/ISION 6 2 /21/do D. CONTROL CHANNEL COLD LEG RTD FAILS LOW l.0 SYMPTOMS 1.1 Abnormally low indication on one of the followind indicators:

a. TI-411A, T-AVG *F.

b. TI-421A, T-AVG *F.

c. TI-431A, T-AV3*F.

1.2 ADnormally nigh indication on one of the followind indicators:

a. TI-4113, T%.

o. PI-421B , T %.

c. TI-4313, T %.

1.3 Hign readings on one pen of ZR-403, ROD POSITION /dOD INSERTION LIMIT Recorder.

1.4 RCS Tavg DEV HI/LO annunicator.

15 dCS T DEV dI/LO annanicator.

1.6 CRB INSERT LMT LO annunicator.

1.7 CRs IdSERT LAT LO-LO annunicator.

2.0 AUTOHATIO ACTIONS None i

PAGE 20 0F 63 1

S0P-401 RSVISION 6 2/21/86 3.0 IMMSOI ATE C0ddSCTIVE ACTIONS 3.1 Verify instrument failure by comparing Control Cnannel T and Tavs indication with Protection Channel T and Tavs indication for tne faulty loop.

4.0 FOLLOWUP ACTIONS 4.1 Place HI T AUC INHIBIT switch to the loop which has tne failed instrument.

4.2 Place HI TAVG AUC INdIBIT SWITCH to tne loop witn the failed instrument.

NOTE 4.3 RCS T DEV HI/LO annunicator will not clear and

• the indicators will not return to normal until the channel is repaired.

4.3 Confirm that RCS Tavs DEV HI/LO.

4.4 Have the failed enannel repaired and verify proper operation prior to returning to service.

PAGE 210F 63

S0P-401 dEVISI0d 6 2/21/86 I. PRESSURIZER PRESSURE CONTROL CHANNEL PAILS HIGH 1.0 SYMPTOMS 1.1 Abnormally hign indication on one pen of PR-444, PZR PSIG RECORDER.

1.2 Abnormally high indication on one of the following indicators:

a. PI-444, PZR PRESS CNTRL CHAN.

b.' PI-445, PZR PRESS CMTRL CHAD.

13 PZR CNTRL PRESS dI annunicator.

1.4 PZR PRESS dI/LO annunicator.

1.5 PZR PRESS dI annunicator. .

1.6 PZR PCS HI annunicator.

LO AUTOMATIC ACTIONS 2.1 If channel PT-445 failed, the following valves open if in AUTO:

a. PCV-445A, P4R RELIEP.

o. POV-4458, P4R RELISP.

2.2 If channel PT-444 failed, PCV-4443, P4d RELId? opens if in A'JTO .

23 If channel PT-444 failed, tne folic wing spray valves open if in AUTO.

a. PCV-4440, PZR SPRAY.

c. POV-4440, PZR SPRAY.

PAJE 34 0F 63 J

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• • 'o' SOP-401 dEVISION 6 2/21/66 i

NOTE 3.0 When pressurizer pressure reaches 1985 psig, automatic reset or manual block on high pressurizer pressure (P-11) will shut the power relier valves.

2.4 It channel PT-444 railed, the following heaters turn off:

a. BU GRP 1.

b. 80 GRP 2.

c. CNTdL GRP.

3.0 IMME0 Card C0dddCTIVE ACPIONS 31 Verity the failure of the Control channel oy comparing PI-444, PZR PRESS or PI-445, PZR PRESS To the following:

a. PI-455, PRESS PSIG.

b. PI-456, PRESS PSIG.

c. PI-457, PRESS PSIG.

32 It channel PT-445 nas failed, close the following valves:

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a. POV-445A, P4R dELIEP.

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b. PCV-4453, PWR RELIEP.

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PAGE 33 0F 03

. . . . . *o SOP-401 dEVISION 6 2 /21/d o 1

33 If channel PT-444 nas failed, do the followind;

a. Close POV-4448, P4R RELIEF.

o. Set PZR PRESS dASTER CONTROL in MAN, and operate to re-establish Reactor Coolant System pressure at 2235 115 psig using functioning pressure enannels.

4.0 POLLOWUP ACTIONS 4.1 If channel PT-445 has failed, verify that the Reactor Coolant System pressure has increased to 2235 115 psig by automatic action of enannel PT-444.

4.2 If channel PT-444 has failed, verify that the Reactor Coolant System pressure has been re-established and can be maintained by manual operation of PZR PRESS dASTER CONTROL.

4.3 Have the failed channel repaired and verify proper

• operation prior to returning it to service.

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l PAGE 36 0F 63 1

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• • 4'.' SOP-401 l REVISIOd o 2/21/8o J. PRESSURIZER PRESSURE CONTROL i CHANNEL FAILS LOW l 1.0 -SYMPTOMS 1.1 Abnormally low indication on one pen of PR-444, PZR PSIG RECORDER.

1.2 Abnormally low indication on one of the following indicators:

a. PI-444, PZd PRESS CNTdL CdAN.

b. PI-445, PZR PRESS CNTRL CHAN.

13 PZR PRESS dI/LO annunicator.

1.4 PZR PCS LO 80 DTRS ON annunicator.

2.0 AUTOMATIC ACTIONS NOTE 2.1 Wnen channel PT-445 indicates a pressare of 2335 psig, the following valves will open:

a. POV-445A, PWR RELISP

b. POV-445B, PWR RELIEP 2.1 If channel PT-444 fails, both control and backup pressurizer heaters will come full on.

2.2 If channel PT-445 fails, no automatic action will tEce place.

PAGE 37 0F 63 l

• .t'o' , SOP-401 RSVISION 6

, 2/21/06 3.0 IMMEDI ATE C0ddECTIVE ACTI0dS 3.1 verify tne failure of tne Control Channel by comparing PI-444, PZR PRESS or PI-445, PZR PRESS to the following:

a. PI-455, PRSSS PSIG.

o. PI-456, PRESS PSIG.

c. PI-457, PRESS PSIG.

3.2 If channel PT-444 nas failed, set PZR PRESS i4ASTSR CONTROL to MAN and operate to re-estaolish Reactor Coolant Systeta pressure at 2235215 psig using functioning pressure channels.

4.0 FOLLOWUP ACTIONS 4.1 If enannel PT-444 has failed, verify that tne Reactor Coolant System pressure of 2235115 psig nas been re-established and can be maintained by manual operation of PC-444A, (PZR PRESS MASTER CONTROL).

4.2 dave the failed channel repaired and verify proper operation prior to returning to service.

PA35 38 OP 63

k SOP-401 ATTACHMENT I PAGE 3 0F I4

, a REVISION 6 i INSTRUMENT F AI LURE REFERENCE MANUAL j

I NSTRUMENT SISTA8tES BlSTABLE SYSTEN FUNCTION NO. ASSOCI ATED LOCATI ON TRIP STATUS LIGHT TECH SPECS STP j 111.F. RCS FLOW FT-426 F8-426A C3-733-GS-1 CHAN 111 RCS LPS FLO LO TABLE 2.2-1 ITEM 12 302.017 i PROTECTION YmLE 3.5-1 I TEM 12 345.013 (LOSS OF FLOW) i TA8tE 3.5-2 ITEN 82 OATC RCS FLOW LOG III.G. FT-434 F8-434A C l-434-85-1 CHAN I RCS LPC FLO LO TA8LE 2.2-1 ITEM 12 302.012 PROTECTION TA8LE 3.5-1 ITEM 12 345.008

, (LOSS OF FLOW)

TA8LE 3.5-2 ITEM 12 04TC l LOG 111.H. RCS FLOW FT-435 F8-435A C2-734-85-1 CHAN 11 RCS LPC FLO LO TA8LE 2.2-1 ITEM I2 302.015 PROTECTION ,

TABLE 3.5-1 ITEM 12 345.011 (LOSS OF FLOW)

TABLE 3.5-2 ITEM 12 OATC LOG III .I . RCS FLOW FT-436 FS-436A C3-73445-1 CHAN lli RCS LPC FLO LO TA8LE 2.2-1 ITEM 12 302.018 4

PROTECTION

' TABLE 3.5-1 1 TEM 12 345.014 (LOSS OF FLOW)

TABLE 3.3-2 ITEM 12 OATC

,A LOG lit.J. RCS FLOW 81 (RCPA) XPN-60ll

/N/A CHAN l IEP SUS UF TABLE 2.2-1 ITEM 16 506.003

, PROTECTI ON 81 (RTS) [ N/A MPN-6012 CHAN ll RCP SUS UF TABLE 3.5-1 ITEM 16 4 506.007 (UM)ER FREQUENCY) 81 (RCPC) N/A XPN-6013 CHAN III fCP SUS UF TABLE 3.5-2 ITEM 16 Ill.K. RCS FLOW 27 (RCPA) M/A XPN-60ll CHAN I RCP BUS UY TABLE 2.2-1 ITEM 15 506.002 PROTECTION 27 (RCPB) NPN-6012

.{ N/A CHAN 11 RCP BUS UV TA8LE 3.3-1 ITEM 15 506.006 (UM)ER VOLTAGE) 27 (RTC) \ 1(/A MPN-6013 CHAN III IEP BUS IN TABLE 3.5-2 ITEM 15 IV.A.

1 I

TAVE TE-412 8/D M2-8-8 C 1-421-85-1 CHAN I LPA (P T TABLE 2.2-1 1 TEMS 7, 8 302.001

. PROTECTI ON 78-412 4 -2 Cl-42145-2 CHAN I W/DRWL ROD 8LCK TASLE 3.3-1 ITEMS 7, 8 345.001 TB-412-C-I C l-421-65-3 CHAN 1 LPA OT T TA8LE 3.5-2 ITEMS 7, 8 TB-412-C-2 Cl-42345-4 CHAN I W/DRWL ROD BLCK TABLE 3.5-4 ITEMS 4.d, i

TB-412-0-1 Cl-422-85-2 CHAN I RCS LPA TAVG LO

! 9.b TB-412-0-2 C l-42245-3 NONE TABLE 3.5-4 ITEMS 4.d, TB-412E Cl-422-GS-1 CHAN I LPA P-12 9.b TABLE 3.5-5 I TEM 8 i .

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QUESTION:- 6.12L (same as. Question:2.22)

~. QUESTION: 7 16a.1- (same'as Question 4.14a.1)

QUESTION:: 7.17c '(same'as Question.4.15c)-

QUESTION: 7.18' (same as Quescion'4.16)-

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