ML20091R824
| ML20091R824 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 06/05/1984 |
| From: | Seidel R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20091R532 | List: |
| References | |
| ODA-102, NUDOCS 8406150252 | |
| Download: ML20091R824 (43) | |
Text
.
COMANCHE PEAK STEAM ELECTRIC STATION OPERATIONS DEPARTMENT ADMINISTRATION MANUAL rea r: awn y
,yl J ,
SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES
,e m
s PROCEDURE NO. ODA-102 REVISION NO. 5 SAFETY 4EATED SUBMITTED BY: DATE: S 6 91 OPERATIONS SUPERINTENDENT
/ .
I)
APPROVED BY: [ Ntb DATE: $ I / 92 Y
/ MANAGER, PLA! OPERATIONS /
B406150252 840609 PDR ADOCK 05000445 A PDR
+:
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL oDA-102 Jfa 07 iss4
'" SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 2 0F 12 ,
1.0 Purpose This procedure describes the required Operations Department shift manning for various modes of operation of the station and delineates the responsibilities and authorities of the members of the shift..
2.0 Applicability
~
This procedure is applicable to all members of the Operations Department shift crews. This procedure becomes effective when issued.
3.0 Definitions 3.1 Senior Licensed Operator - An individual having a current USNRC Senior Reactor Operator License on all station units that have a current facility operating license.
3.2 Licensed Operator - An individual having a current USNRC Reactor Operator or Senior Reactor Operator License on all station units that have a current facility operating license.
_ g]' 3.3 Operating - A reactor unit is considered to be operating when it is in operational Mode 1, 2, 3 or 4 as defined by CPSES Technical Specifications.
3.4 Licensed to Operate - A reactor unit is considered to be licensed to operate if it has a current facility operating license and initial fuel loading has begun.
3.5 Controls - Apparatus and mechanisms the manipulation of which directly af fect the reactivity or power level of the reactor.
4.0 Instructions 4.1 Authority of Licensed Personnel 4.1.1 The station will be operated by USNRC licensed personnel in accordance with 10CFR50, 10CFR55 and the Technical Specifications, Section 6.
4.1.2 All controls will be manipulated by licensed personnel under the direction of senior licensed personnel.
Manipulation of controls by non-licensed personnel is permissible as part of the Replacement Training Program, but must be directly supervised by a licensed individual.
l CPSES ISjUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL 3DN 07 604 ODA-102 SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 3 0F 12 .
4.1.3 The Reactor Operator, Assistant Shift Supervisor, Shift Supervisor or any other licensed member of the station ,
staff assigned to manipulate or supervise the manipulation of the controls of a unit or units has the responsibi.lity and autharity to place the reactor or reactors in a safe condition when he determines that the safety of the reactor (s) is in jeopardy or when operating parameters exceed any Reactor Protection System or Safeguards System setpoints without automatic protection functions occurring.
4.2 Responsibilities of Shift Crew Personnel 4.2.1 General Responsibilities In addition to the specific duties of shift crew personnel
-as delineated in Sections 4.2.2, 4.2.3, 4.2.4 and 4.2.5, all shift crew members have the following responsibilities:
4.2.1.1 The responsibility to believe and respond conservatively to instrument indications unless
('~ )
they are proven incorrect.
4.2.1.2 The responsibility to adhere to Technical Specifications.
4.2.1.3 The responsibility to follow written procedures.
4.2.1.4 The-responsibility to review routine operating data to assure safe operation.
4.2.2 Shif t Supervisor The Shift Supervisor is responsible to the Operations Supervisor for the operation of the station and the management of operating personnel on an assigned shift consistent with administrative and regulatory requirements. Specific duties of the Shift Supervisor include:
4.2.2.1 Supervision of shif t operating personnel to ensure that the station and all associated equipment is operated safely, efficiently, reliably and in accordance with Technical Specifications, approved procedures, regulations and licenses.
O
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAlg Q7 gg4 ODA-102 U) SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 4 0F 12 4.2.2.2 The responsibility to determine the circumstances, analyze the cause and determine that operations can proceed safely before a reactor is returned to power following a trip or an unplanned or unexplained power reduction'.and to provide direction for the return to power.
4.2.2.3 Supervision of the preparation of all routine shift documentation and review of routine operating data.
4.2.2.4 Assumption of complete responsibility for the safe operation of the station in the event of an emergency. The Shift Supervisor shall remain in the Control Room during the emergency until properly relieved. The Shift Supervisor shall maintain a broad perspective of operational conditions during the emergency and should not become totally involved in any single operation.
4.2.2.5 The responsibility for the initiation of the Emergency Plan in the event of an emergency
(/
)
situation and for serving as Emergency Coordinator until relieved.
4.2.2.6 ' The responsibility for the implementation of applicable portions of the Security Plan.
4.2.2.7 Assistance in the review and modification of operating procedures as required.
4.2.2.8 The responsibility to follow radiation protection and control procedures and to manage the radiation exposures of assigned personnel to ensure that they are within administrative and regulatory limits.
4.2.2.9 The responsibility for ensuring that an adequate number of qualified operations personnel are on duty during an assigned shift, consistent with Section 4.3 of this procedure, and for preparing daily shift crew work schedules.
4.2.2.10 Participating in the Requalification Training i Program and maintaining a current USNRC Senior Reactor Operator License.
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL g g7 g ODA-102
\
SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 5 0F 12 .
4.2.2.11 Other specific duties as assigned by the Operations Supervisor and as delineated in Station Administrative Procedures STA-601 through STA-607 and Operations Department .
Administrative Procedures ODA-103 through .
ODA-105 and ODA-301 through ODA-308. ,
4.2.3 Ass-istant Shift Supervisor The Assistant Shift Supervisor is responsible to the Shift Supervisor for assisting in the operation of the station and the management of operating personnel on an assigned shift. Specific duties of the Assistant Shift Supervisor include:
4.2.3.1 Assumption of the duties of the Shift Supervisor in his absence.
4.2.3.2 Supervision of assigned shift operating personnel to ensure that equipment is operated safely, efficiently, reliably and in accordance O
D with Technical Specifications, approved procedures, regulations and licenses.
4.2.3.3 The responsibility to determine the ,
circumstances, analyze the cause and determine that operations can proceed safely before a i reactor is returned to power following a trip or an unplanned or unexplained power reduction and ,
to provide direction for the return to power as assigned.,
4.2.3.4 Supervision of the preparation of all assigned shift documentation and review of routine.
operating data.
i 4.2.3.5 Assistance in the review and modification of operating procedures as required.
4.2.3.6 The responsibility to follow radiation protection and control procedures and to manage the radiation exposures of assigned personnel to ensure that they are within administrative and regulatory limits.
4.2.3.7 The responsibility for assisting in on-shift ,
training of operating personnel.
.- -. . .- _ _ - - _ _ _ . _ = _ _ - - .. .
. l l
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL M 07 iM ODA-102 V SHIPT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 6 0F 12 .
s 4.2.3.8 Participating in the Requalification Training .
Program and maintaining a current USNRC Senior !
Reactor Operator License. [
4.2.3.9 Other specific duties as assigned by the Shift Supervisor and as delineated in Station Administrative Procedures STA-601, STA-602 STA-605 and STA-607 and Operations Department t Administrative Prodedures ODA-104, ODA-105 and ODA-301 through ODA-306.
4.2.4 Reactor Operator The Reactor Operator is responsible to the Shift Supervisor or Assistant Shif t Supervisor for operations on an assigned unit or units. Specific duties of the Reactor Operator include:
4.2.4.1 -The responsibility for safe, efficient and reliable operation of equipment in accordance with Technical Specifications, approved I
() procedures, regulations and licenses.
l 4.2.4.2 Maintaining all logs and records for the unit or
, units assigned and reviewing routine operating data.
4.2.4.3 The responsibility for operating equipment Yrom '
the Control Room and locally and assisting in all phases of operation as assigned.
4.2.4.4 Continuous monitoring of Control Room indications on an assigned unit and adjustment of parameters as necessary.
4.2.4.5 Assisting the Shif t Supervisor or Assistant Shif t Supervisor in directing and coordinating the operating activities of the Auxiliary Operators. ,
4.2.4.6 Maintaining necessary communications both within and outside of the plant. ;
l 4.2.4.7 Performing the Auxiliary Operator's duties when
! necessary or when directed by the Shift Supervisor or Assistant Shif t Supervisor.
I e P
r i
! t
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL JUN 0 7 004 ODA-102
. 3
() SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 7 0F 12 ,
4.2.4.8 Assisting the Shif t Supervisor or Assistant Shift Supervisor in tagging and removal of equipment from service and in returning equipment to service when authorized by the, .
Shif t Supervisor or Assistant Shif t Supervisor. !
4.2.4.9 Participating in the Requalification Training Program and maintaining a current USNRC Reactor 1 Operator or Senior' Reactor Operator License. l 4.2.4.10 Assisting in the training of Auxiliary Operators.
4.2.4.11 Assisting in the preparation, review and modification of operating procedures and other operating documentation. l 4.2.4.12 Performing duties required by the Fire Protection Plan and Emergency Plan.
4.2.4.13 Assisting with station housekeeping.
o I
(
N~/ 4.2.4.14 Other specific duties as assigned by the Shift Supervisor or Assistant Shif t Supervisor and as delineated in Station Administrative Procedures l STA-601, STA-605 and STA-607 and Operations Department Administrative Procedures ODA-104 and ODA-301 through ODA-305.
4.2.5 Auxiliary Operator The Auxiliary Operator is responsible to the Shif t Supervisor or Assistant Shif t Supervisor for the safe and efficient operation of equipment required to support overall station operation. Specific duties of the Auxiliary Operator include:
4.2.5.1 Proper, sare and efficient care and operation of station auxiliary equipment and systema under the direction of the Shift Supervisor or Assistant Shif t Supervisor and the Reactor Operator and in accordance with administrative, regulatory and procedural requirements.
4.2.5.2 Periodic inspection of assigned equipment including completing documentation associated with these inspections and adjusting controls as necessary for proper equipment and system
(~')
v operation.
i
CPSES ISSUE DATE PROCEDURE NO.
, OPERATIONS DEPARTMENT ADMINISTRATION MANUAt JUN 0 71984 ODA-102 l I
f 'l N/ SHIFT COMPLEMENT l RESPONSIBILITIES AND AUTHORITIES REVISION No. 5 PAGE 8 0F 12 .
l 4.2.5.3 Assisting the Shif t Supervisor, Assistant Shif t Supervisor or Reactor Operator in tagging and removal of equipment from service and in returning equipment to service when authorized.
4.2.5.4 Maintaining necessary communications with the Control Room.
l .
4.2.5.5 Periodic inspectiods of controlled access areas including the Containment. Auxiliary Building, Fuel Building and Safeguards Buildings as directed by the Shif t Supervisor or Assistant Shif t Supervisor.
4.2.5.6 Performance of other duties and assistance in all phases of operation as required and directed by the Shift Supervisor or Assistant Shift Supervisor.
4.2.5.7 Participation in the Replacement Training Program in order to obtain a USNRC Reactor Operator License.
)
4.2.5.8 Assisting in the preparation, review and modification of operating procedures and other operating documentation.
4.2.5.9 Assisting with station housekeeping.
4.2.5.10 other specific duties as delineated in S,tation Administrative Procedures STA-601 and STA-605 and Operations Department Procedures ODA-104, ODA-301, ODA-302 and ODA-304 4.2.6 Shift Advisor The Shift Advisor functions at the Assistant Shift Supervisor level and is responsible to the Shif t Supervisor for evaluating shif t operating activities and providing appropriate recommendations concerning safe operation.
The Shift Advisor is assigned to assist the Shift Supervisor but reports to the Operations Supervisor. He has direct access to plant and corporate management and is responsible for pursuing the resolution of disagreements affecting safe operation through successive management levels.
O)
\_
- Specific responsibilities and authorities includes l l
l
, N. ,
1 -.
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAt .JUN 0 7 d64 ODA-102
(.
(/ SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 9 0F 12 ,
4.2.6.1 Respoacible for assisting in the determination of the circumstaucce, analyzing the cause and
' determining that operations can precead safely before recommending a return to power following a trip or an unplanned or unexplained power.
reduction.
4.2.6.2 '
Assisting in the review and modification of operating procedures.
4.2.6.3 Assisting in the preparation and review of shift documentation and operating data.
4.2.6.4 Assisting in on-shift training of operating personnel.
4.2.6.5 Participating in Shift Advisor training, including recurrent training.
4.2.6.6 Other specific duties as assigned by the Shift Supervisor. The duties assigned cannot include those requiring a license.
4 ('_')
s s 4.3 Shift Complement The minimun on-duty shift complement for various modes of single and dual unit operation shall be as shown in Attachment 1 and as follows:
4.3.1 A USNRC Senior licensed Shift Supervisor shall be onsite
, at all times when at least one unit is. loaded with fuel.
When the Shift Supervisor is absent from the Control Room during routine operations, he shall be relieved by a qualified and USNRC Senior Licens0d member of management.
This is normally an Assistant Shift Supervisor. The Shift Supervisor's relief shall assume the Control Room command function as well as the complete responsibility and authority as is normally assigned to the position.
4.3.2 One USNRC Senior Licensed Operator shall be in the Control Room at all times when in Modes 1, 2, 3 or 4.
m 5,3.3 One USNRC Licensed Operator shall be in the Control Room at all times for each reactor containing fuel.
.4.3.4 Two USNRC Licensed Operators should be in the Control Room for each reactor while undergoing a startup, scheduled shutdown or reactor trip recovery.
[)
CPSES ISSUE DATE PROCEDURE NO.
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL JUN 07 E84 *A-102 C' SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 10 0F 12 4.3.5 Two USNRC Senior Licensed Operators shall be onsite at all times with both units loaded with fuel.
4.3.6 In addition to the operators specified in 4.3.1, 4.3,2, ,
4,3,3. and 4.3.5, an additional USNRC Licensed Operator shall be onsite at all tince and available to serve as relief operator for the Control Room if either unit is in Mode 1,, 2, 3 or 4.
4.3.7 Shift crew assignments during periods of core alterations
< shall include a USNRC Senior Licensed Operator to directly supervise the core alterations. This operator may have fuel handling duties but shall not have other concurrent operational duties.
4.3.8 With one unit licensed to operate (mode 5 or 6), each shift crew shall have at least three members including one Shift Supervisor and one USNRC Licensed Operator.
4.3.9 With one unit operating (mode 1, 2, 3 or 4), each shift crew shall have at least six members including one Shift Supervisor, one Assistant Shift Supervisor and two USNRC
-() Licensed Operators.
4.3.10 With two unita licensed to operate (both units in mode 5 or 6), each shift crew shall have at least six members including one Shift Supervisor and two USNRC Licensed Operators.
4.3.11 With two units licensed to operate and ona or both operating (mode 1, 2, 3 or 4), each shif t crew shall have at least eight members, including one Shift Supervisor, one Assistant Shift Supervisor and three USNRC Licenwed Operatcre.
4.3.12 In addition to the personnel specified in 4.3.8, 4.3.9, 4.3.10 and 4.3.11 above and with fuel in the reactor, one Radiation Protection Technician and one Chemistry and Environmental Technician shall be on site at all times..
4.3.13 With one or both units operating (mode 1, 2, 3 or 4), a Shift Technical Advisor shall be on site.
4.3.14 With one or both units operating (modes 1, 2, 3 or 4), a Shift Advisor shall be on site.
O
\
CPSES ISSUE DATE PROCEDURE NO.
ODA-102 OPERATIONS DEPARTMENT ADMINISTRATION MANUA1 M 07 4 SHIFT COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 11 0F 12 .
4.3.15 A site Fire Brigade of at least 5 members shall be maintained onsite at all times. The Fire Brigade shall not include the Shif t Supervisor and the 2 other members of the minimum shift crew necessary for safe shutdown,of the unit and any personnel required for other essential functions during a fire emergency.
4.3.16 Encept for the Shift Supervisor, the Shift Crew composition may be one 1cce t han the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew composition to within the minimum requirements.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
5.0 References 5.1 CPSES Final Safety Analysis Report, Section 13.1 O 5.2 Procedure ODA-101, " Operations Department Organization and Responsibilities" 5.3 USNRC Standard Review Plan, Section 13.1.2 5.4 USNRC Letter, " Interim Criteria for shift Staffing, July 31, 1980 5.5 NUREG-0578, 2.2.1.a 6.0 Attachments 6.1 Minimum Shift Crew Composition, Attachment 1 s
CPSES ISSUE DATE PROCEDURE NO.~
OPERATIONS DEPARTMENT ADMINISTRATION MANUAL ODA-102 JUNO't ed4 SHIFT. COMPLEMENT RESPONSIBILITIES AND AUTHORITIES REVISION NO. 5 PAGE 12 0F 12 ,
' ATTACHMENT 1 PAGE 1 0F 1 MINIMUM SHIFT CREW COMPOSITION e
MODE ,
eisa LICENSED TO OPERATE UNIT 1 , UNIT 1 AND 2 1 S. S. 1 S. S.
ONE OR BOTH UNITS 1 Ass't. S. S. 1 Ass't. S. S.
IN MODE 1, 2, 3, OR 4 2 R. O. 3 R. O.
2 A. O. 3 A. O.
1 Shift Advisor 1 Shift Advisor 1 S. T. A. I S. T. A.
1 R. P. Tech 1 R. P. Tech 1 C. E. Tech 1 C. E. Tech TOTAL 10 12 1 5. S. 1 S. S.
BOTH UNITS IN 1 R. O. 1 Ass't. S. S.
MODE 5 OR 6 1 A. O. 2 R. O.
1 R. P. Tech 3 A. O.
1 C. E. Tech 1 R. P. Tech 1 C. E. Tech TOTAL 5 9-
, POSITION (1) USNRC LICENSE SHIFT SUPERVISOR -
S. S. SRO ASSISTANT SHIFT SUPERVISOR - Ass't. S. S. SRO REACTOR OPERATOR -
R. O. R0 AUXILIARY OPERATOR -
A. O. NONE SHIFT TECHNICAL ADVISOR -
S. T. A. NONE SHIFT ADVISOR NONE (1) Any qualified and USNRC Senior Licensed member or managment may 'oe p used to satisfy the minimum Shift Supervisor or Assistant Shift Q Supervisor requirement. Any qualified and USNkc Licensed individual may be used to satisfy the Reactor Operator requirement.
-, - , ..r ..--we w 4- --,w-- - - , - - - - - - - + -.,..---,-----r-- - - - -,*e.- , - - - - --- r**e- - ~. * - - -
py%-y-L h-3
[ Attachment 4
[- TRA-299 Shift Advisor Training and Qualifications t
?-
e 5-t i
f lh b _.,. _ . -. __. . . _ , - _ _ . . . _ . _ , - _ - __ _ - _ _ . _ - -__ _
I EXAM KEY COURSE NAME SHIFT ADVISOR TRAINING WEEK 2 Date MAY 18, 1984 TOTAL POINTS 32.75 _
(
I i
l .
}
l SUBMIi IED BY : 5W DATE : I7 #kN8 i
APPROVED BY : A /"' DATE : #b!#
/ ,
NOT-105-1 Fcrm C Rev. 0
1 RPS 1 Q. Briefly describe the operation of (1.5) the Reactor Protection System by tracingcthe path of components for a reactor trip signal.
2 RPS 1 A. From the sensor to the bistables, to the input cabinets and input relays to the solid state logic cabinets, to the UV coils on the reactor trip breakers.
5* RPS 3 Q. Describe the functions provided by (3) the following permissives/ inter-locks.
- a. P-4 Reactor Trip
- b. P-7 At Power
- c. P-12 Low-low Tavg _
- d. C-9 Condenser interlock
- e. C-5 Low Pcwer interlock f .- C-8 Turbine T-ip 6 RPS 3 A.
NAME FUNCTION
. P-4 _ Reactor Trip 1. Trips Main turbine
- 2. Trips FRV.W/Lo Tavg
- 3. Prevents reactuation of SI after a manual reset P-7 At Power Automatically unblocks PZR low press., PZR hi level, all flow trips.
P-12 Low-Low Tavg Interlocks steam dump below setpoint. Cooldown valves may be bypassed.
. C-C Low power. interlock Stops outward rod motion in auto only.
I i
m . . . . . _ , . - . - . . , - - - ,e-,- % , , - , - , , , - . . . . ,- , - - - - _ , . - - . _ _ - , , , - - - , , , - . , .
- . - , - ~ - - .
C-8 Turbine trip Arming of steam dump after turbine trip-Train."A".
-Train "B" shifts steam dump
, valves from output of LOL to the T.T. Controller.
C-9 Condenser interlock Condenser available for steam dump.
1 ERG 1 Q. What types.of procedures make up (1) the Emergency Response Guidelines?
2 ERG 1 A. 1. Emergency Operating Procedures (EOP)
- 2. Emergency Operating Sub-Procedures (EOS)
- 3. Emergency Contingency Actions (ECA)
- 4. Functional. Restoration Guide-lines (FRG's)
(.25 each) 3 ERG 2 Q. What is the purpose of the
(.5) Functional Restoration Guidelines?
4 . ERG 2 A. The FRG's are procedures designed to maintain the plant in a safe condition without regard to initiating events.
7 ERG- 4 Q. If a Critical Safety Function
(.5) Status Tree contains a " Red" path, do you:, (circle correct answer) a.
Finish the step you are on and then take actions defined by the CSFST.
- b. Immediately take actions as determined by the CSFST.
- c. Complete the procedure you are in and then take actions defined by the CSFST.
8 ERG 4 A. b. Immediately take actions as '
determined by the CSFST.
31 SOP 16 Q.- How many RCP's may be started at (1.5) one time? What.are the starting duty limitations on the RCP's?
32 SOP 16 A. l.RCP at a time. (.25)
L Maximum of 3 starts in a-2-hour-period with at least 30 minutes rest between starts or attempted starts.-- (.75)
A fourth start should'first allow a 1-hour rest period. (.5)
O
'r
1 I
5 .ABN 3 Q. List three operator actions which ,
(.75) are performed following verification of the failure of the
'91 seal in a RCP.
(ABN-101A) 6 ABN 3 A. Any three for full credit (.25 pts each)
- 1) Reduce reactor power to less than 45% within 30 minutes.
- 2) Within 5 minutes of discovery and verification of the failed seal, close the #1 seal leakoff isolation valve for the affected RCP.
- 3) After reaching 45% reactor power, secure the affected RCP.
- 4) Place the Unit in Hot Standby within one hour after stopping the affected RCP.
- 5) Consult CPSES Tech Specs Section 3/4.4.1 for any applicable LCO's.
7 ABN 4 Q. List 4 conditions that could result (1) in a gross failed fuel monitor
" alert" alarm. (ABN-102A) 8- ABN 4 A. 1) Gross Failed Fuel Monitor Malfunction
- 2) Depleted resin in Letdown ion exchanger
- 3) Crud burst causing activity
- 4) Actual failed fuel (0.25 pts. each)
.. --.y- , - - - , , --r , . ,
i i
DIFFERENTIAL INTEGRAL INTEGRAL i
i a e .
l
+,
as E E N o o E
a or 4
i l
o R
l Rod Position, Steps Rod Position, Steps Rod Position, Steps FIGUllE TDB-1 i
1 1
1 6
A Unacceptable Operation A portion - protection i th against saturation i
im B conditions in the B rtion - protection against DNB conditions i
Acceptable Operation -
J j Power i FIGUllE TDB-2 i
i
1 TDB 1 Q. Draw an integral and differential (1) rod worth curve and explain the reasons,for the observed shape.
e (Label'ach axis with correct units
- actual-numerical values are not required.) -
2 TDB 1 A. The reactivity of any absorber at any position within the core is proportional to flux squared -
since the axial flux is a cosine shape the axial. worth is propor-tional to (cosine): and since the worth is related to the flux at the tip of the rod the axial differen-tial worth will be (cosine) 2 or as.
shown above - the integral of this function will be a sigmoid as shown in Fig. TDB-1.
3 TDB 2 Q. Draw a family of curves similar to (1) the CPSES safety limit curves and explain the rationale for each portion of the curve. (Label each axis with correct units - actual numerical values are not required.)
4- TDB 2 A.
I
+
i-
,5 TDB . 3 Q. Figures 3.2 and 3.3 (Heatup'and (1.5) Cooldown curves) in the Tech. Data Book show a parameter "RT ~
NDT briefly explain what is meant by this term; also which is more liiniting heatup or cooldown and why?
6 TDB 3 A. RT - reference temperature for NDT the transition from ductile to brittle fracture.
Cooldown is more limiting since the
' tensile forces on the inside surface of the vessel are all additive (i.e., no counteracting compressive forces) and the total tensile force comes closest to reaching the maximum allowable design stress where a failure may occur.
'3 TAA 2 Q. Provide the bases for the Rod (3) Insertion Limits as specified by_
the CPSES Technical Specifications?
4 TAA 2 A. 1. Ensure adequate shutdown margin. (1)
- 2. Promote more even power distribution. (1)
- 3. Minimize effects of ejected rod accident. (1) l e -
-+- -- . . - r .r --,,-- ,..e. ,. --. . -y .,-,--,,-v . ,- ,- , , - - . _
5 TAA 3 Q. What conditions must be maintained (4) by operators to ensure that the hot channel. factors limits are not exceeded?
6 TAA 3 A. 1. Delta Flux limits are observed as prescribed by Axial Flux Difference Target Band. (1)
- 2. Rod Insertion Limits are observed. (1)
- 3. _ Observe proper bank sequencing with overlap. (1)
- 4. Maint?in rods in a bank within i 12 steps of cach other. (1) 19 TS 10 Q. Define " shutdown margin."
(1) 20 TS 10 -A. The instantaneous amount of reactivity by which-the reactor is suberitical (.r o would be.suberit-ical) if all rod clusters were inserted except the cluster of highest reactivity worth.
57 TS 29 Q. CIRCLE THE CORRECT ANSWER (S).
(.5) Per Technical Specification Bases, the limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that:
- 1. The design limits on peak local power density and minimum DNBR are not exceeded.
- 2. A coolable core geometry is maintained.
, . - - , - , - ,- ---,,-a , , ~ ,---n---,n- - - - - . - - - , - - -
, g , _
- 3. The DNB parameters are not l exceeded. ,
- 4. In'the event.of a LOCA the peak clad temperature will not exceed:the 2200*F'ECCS acceptance criteria limit.
- 5. All.the above.
- 6. None of the above.
58- TS 29 A. 1. The design limits.on peak local power density and minimum DNBR are not exceeded.
- 4. In the event of-a LOCA the peak clad. temperature.will not exceed.the 2200*F ECCS acceptance criteria limit.
73 ITS 37 Q. .At.the Comanche Peak Steam Electric.
(1) Station, no credit was taken.for the Source Range and Intermediate Range-Rx. Trip in the Final Safety Analysis Report for a startup s
accident event. Explain why. ,
74- TS - 37 A. The Source and Intermediate Range-trip circuits can be manually bypassed by operators by utilizing the Level Trip Bypass' switches on the front of the instrument drawers. Since the Rx Trips can be physically bypassed, they were not
.used in the FSAR.
4 i ;-
---m ,.-,.,ar,..-. ..,,w n.,~,-, se.,,,-,w---.- ,~,,w,-,,-.---mm,~,...---w-. -.--,,,,e, - - ---n--,,,-,---,.,e.,w~~--e-..m.,, ,
T 75 TS 38 Q. a. Define the following (4)
- 1. DNBR (1) 2.-'
Hot Channel Factor (F)
(1)
. 3. Critical Heat Flux (1)
- b. What Rx protection signal is designed specifically to prevent DNB for all combinations of pressure, power, coolant temperature and axial power distribution? (1) 76 TS 38 A. a. 1. DNBR is defined as the ratio of the heat flux that would cause DNB at a
~
particular core location to the local heat flux at that same core location.
- 2. Hot Channel Factor (F) is a peak (maximum) to
' average ratio of something, e.g. for local
. power density:
Peak kw/ft Average kw/ft
~
- 3. Critical Heat Flux is the heat flux. (Q) necessary to depart from the nucleate boiling region, or the heat flux needed at DNB.
- b. OTN-16 trip.
-to 1
li -
77 TS 39 Q. According to Technical (3) Specifications, what requirements are necessary for a high radiation area in'which the intensity of radiation-is greater than 100 mrem /hr but less than 1000 mrem /hr?
78 TS 39 A. (.5 pts. each)
- 1. Area shall be barricaded
- 2. Conspicuously posted as a high radiation area
- 3. Entrance controlled by requiring issuance of RWP
- 4. Individuals entering shall be provided with, or accompanied by one or more of the following:
(a) A radiation monitoring device which continuously indicates dose rate, or (b) A radiation monitoring device which continuously integrates the dose rate and alarms when a-preset limit is reached, or~~
(c) A HP' individual (qualified) with-a dose rate monitoring device who is responsible for positive control over the activities in the area.-
79 TS 40- Q. What is/are (3)
- a. Axial Flux Difference (1)
- b. Quadrant Power Tilt Ratio (1)
- c. Subcooling Margin (1) 80 TS 40 A. a. Axial Flux Difference is the-difference in power (expressed in %) between the upper and lower halves of the core, e.g.
P ~
- O "**
TOP BOTTOM Measured by excore detectors in each of four quadrants of the core.
- b. Quadrant Power Tilt Ratio is the ratio of the maximum calibrated upper detector
~
output to the average of.the upper. detector outputs or the maximum calibrated lower detector output to the average of the lower detector outputs; whichever ratio.is greater.
(1)
- c. Subcooling margins is T SAT ~
T operating r the margin between the hottest fluid temperature-in a system and the saturation temperature for the system pressure, e.g. Pzr Temp at saturation = 653'F and THot'at full power a 620*F.
Subcooling margin a 33 F. (1)
. . , . - - - , . , , - . . -%,. - ,..- --- - - - - - , , - -- - - - y , %, -, ,- -,-a ,-,,--,----%, ---%. --m ,
B l
EXAM KEY l COURSE NAME SHIFT ADVISOR TRAINING WEEK 4 l
Date l JUNE 1, 1984 l
TOTAL POINTS 5 YY I l
l l
I 1
l SUBMI: iED BY : h- =
DATE: A l APPROVED BY : ! / DATE : 'C d!!
/ '
i NOT-105-1 Form C Rev. O
- .; 5 1 CG 1 Q. What concentration of hydrogen is (1) . considered an explosive concen-tration in air?
.2 CG- 1 A. > 4%-
t t
, 3 CG 2 -Q. List three (3) sources'of hydrogen (1.5) productionLin containment following a LOCA.
4 .CG -2 A. 1. Hydrogen present in the reactor coolant released on depressarization. (.5)
- 2. Zirconium / water reaction in
-the core. (.5)
- 3. Radiolysislof water in the core and containment sump.
'(.5)
- 4. Aluminum and' zinc corrosion-of plant materials. (.5) 37 EPP 19 -Q. State the purpose for Protection ,
(.5) Action Guides.
'- 38 EPP 19 'A. To provide guidelines to the Emergency Coordinator for evaluating post-emergency condi-tions' and for -making recommenda-tions-to offiste agencies concerning protective measures that
~
could be implemented to insure minimum doses of radiation to the public following offsite releases of radioactive nuclides.
~
39 EPP 20- Q. Define: ,
-(l)- !
- a. Plume Exposure Emergency Planning Zone ,
- b. Ingestion Exposure Emergency Planning Zone 40 EPP 20 A.- a. The Plume Exposure EPZ.is a zone 10 miles in radius from plant centerline. Used in determining protection measure to protect public from excessive doses received from plume passage. (.5)
- b. The Ingestion Exposure EPZ extends to a radius 50 miles from plant boundaries. Used in determining protection measures to limit ingestion of radioactive nuclides from food supplies and to limit public exposure from ground level contamination. (.5) 3 105 2 Q.
What type radiation detectors are (1) used in Process and Area Monitors?
4 RM 2- A. Process + Scintillation GM Tube Area + G-M' Tube (Low Range)
+ Ionization Chamber (High Range)
m 7 101 4 Q. How do we monitor for fission-l(1) product activity in the RCS by the Radiation Monitoring System? !
-8 RM 4 A.- Use of failed fuel monitors 9 5 O. a. Otherk %*'8gf th major M LOCA, TAA' NDP wh* events at the CPSES are 2 ost limiting with respect to DNBR. (2 pts.)
- b. Define DNBR. (1 pt.)
- c. What parameters are monitored to ensure that DNBR is maintained greater than a value of 1.30? (2 pts.)
10 TAA 5 A. a.- Rod ejection ac nt; continuou n le rod with-draw ccident and RCP locked' r or shaft break accident.
(2)
- b. DNBR = h' eat flux to cause DNB at some pt.-in the core actual heat' flux at that point in-core c., DNBR > 1.3 Temperature (.5)
~
Pressure (.5)
Potier (.5)
Flow (.5) e, di - .
4 , -, ,- + , , 7
. , - - . , .- , .n.-m. ,, c,w, , , , - - - - - ,
(' 27 4 TAA (5) 14 Q. Draw graphs of nuc ear power, pressurizer presliure, Tavg, Loop AT j
)
and DNBR form he ten minute period l following n increase in feedwater ;
flow aco eng M
28 TAA 14 O/ Q A.
80 i i l i .i l I 70 E-a-
60 b 2400 ES 50 7 Sa 8- 40 1 2300 -
5;i 30 w j Cd
'g J~ _
y
'N 2200 0-10
.E .
ar 10 O
=( , '
600 5 .
= 2000 se E I I I I E
< 590 I I 1 .
1900 -
g 150 175 200 g-0 25 50 75 100 125 TlHE (SEC) [3
=a i e
5 570 -
~
E g 560 u
1.2-I i >! l i l i 550 1.0 4 a= 4.0
-0.8 g
-- 3.5 gg_ 0.6 m
if 0.4
,'o
- 2.s 1_ . 3.2 0.0
( 2.0
' I 1.5 , ; l l l l 1.3
- n v m
21! TAA 11 Q. State three (3) methods of decay (3) heat removal following a turbine trip. (Assume the steam dumps are not operable.)
a 22 LTAA- 11 A. a. Pressurizer reliefs and safeties
- b. Steam generator reliefs and safeties
- c. Auxiliary feedwater flow to S/G's 27* TAA 14 Q. Draw graphs of nuclear power, (5) pressurizer pressure, Tavg, Loop AT and DNBR for the ten minute period following an increase in feedwater flow accident.
28 TAA 14- A.
29 TAA 15 Q. Compare the differences in severity
-(3) of a large steam line rupture with forced Reactor Coolant flow versus
. initiation of'the accident ~
accompanied with loss of off-site power.
30 TAA 15 A. Large steam line rupture cooldown event is faster acting on core when the pumps are running since there is fa' ster coupling of. cold water to the core from the S/G. with no pumping power, the S/G cooldown that results is not seen as soon by the core because of the reduced flow. -
Therefore, maintaining pump power is a more severe transient.
- - - .-. . .~ .- -
s 37 TAA 19 Q. The CPSES reactor is operating at (3) 100% power with all control systems in automatic when an inadvertent dilution event occurs. Assuming BOL conditions and the dilution continues indefinitely, explain what will happen to plant conditions over the first 15 minutes. If a reactor protection signal is generated, what will cause it to occur?
38 TAA 19 A. As dilution begins, positive reactivity will be added to the core, causing reactor power to increase. As reactor power rises above steam demand, T hot and, subsequently, T cold will increase causing Tavg to increase. When Tavg increases to 1.5*F greater than Tref, rods will step inward to maintain Tavg with Tref. Rods will probably move inward intermittently as dilution continues until the rod ~
insertion limit low and low low alarm points are reached and
. pussed. Delta flux will start to be pushed negative which will force.
operation outside the target band and also cause a penalty f (A flux) to the OTN-16 trip setpoint, reducing the setpoint. As OTN-16 trip setpoint is reduced toward the actual N-16 in two of four loops turbine runback / rod stop will occur (3% difference) and quite probably
- a reactor trip will follow.
Dilution will continue until
-stopped by operator or system action.
4
.y. . . _ - - -. _ _ . - . , . - . . . - . . . . - , , , , , . , , . . . . . . - - . - . , - - - . , . . , _ . -
s 69* MNui 35 Q. The CPSES Reactor is operating at (3) 100% power with rods in manual when a 50% load rejection occurs.
Assuming no reactor trip and proper steam dump operation, describe the plant conditions once they have
, stabilized. Include in your discussion the transient effects on power, temperature, pressure and S/G level.
70 TAA 35 A. Answers will vary based on assumptions for values of aT and Doppler power coefficient. Assume BOL values: aT - 5pcm/*F; Doppler
,' power coeff - 10 pcm/% power Ideal case: With rods in manual, the steam dumps will accommodate 40% of the load rejection. When load is lost, Tavg will increase adding - p to the core, causing power to decrease which adds + p due to Doppler which balances the reactivity.
With a 5*F deadband on the load rejection controller before actuation, Tavg will use from 588'F (full power program) to 593*F to actuate dumps. Meanwhile Tref will be reduced to 572.5'F as-a function of turbine steam pressure.
The 5*F increase in Tavg will reduce power about 2.5% to 97.5%
power. Steam dump p>p open actuation will account for 40% of steam demand, so the remaining 10%
will be lost due to Tavg increase.
10% is -100 pcm which would result in a 20*F increase in Tavg.
- a. Tavg would approximate 613*F.
- b. Nuclear power would approximate 87.5%.
- c. Steam dumps would be wide open dumping 40% steam demand.
lT .
- d. Turbine load would approximate 50%.
- e. S/G levels would shrink but are assumed to return to normal.
- f. Pressure would exercise but return to normal (sprays).
39 TAA 20 Q. a. What is a reactivity anomaly?
(2)
- b. What action (s) is/are required if a reactivity anomaly were to occur?
40~ TAA 20 A. a. An anomaly is essentially something that deviates in excess of normal variation.
In the case of reactivity anomalies, they are specif-ically defined as a 1%
deviation of the actual boron concentration versus the predicted boron concentration over core life. (1)
- b. A reactivity anomaly must be reported to the NRC immediately (uithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />),
but does not require plant
- shutdown, unless subsequent evaluation dictates shutdown
. as a necessary action. (1)
L-,_
m
- 15 TAA- 8 Q. What are the indications for a loss
-(3) of natural circulation flow?
16 TAA 8 A. a. Increasing loop delta T's indicate-that natural circula-tion flow is decreasing. With insufficient flow through the core, core exit temperature will rise. Core exit thermo-couples and wide range T hot will be the first indications of insufficient core cooling and their indications will start to increase.
- b. Increasing T cold indicates insufficient heat is being removed by the steam gener-ator. If actions are not taken to increase secondary-cooling, T hot will also start to increase. This will result in higher secondary. steam pressures and lower nargin to saturation in the reector coolant.
17* TAA 9 0 Discuss the effects of an (2) inadvertent initiation of ECCS during power operation.. Assume the spurious SI signal does not cause a reactor trip / turbine trip. Discuss
~
the effects of this transient on major RCS parameters and overall unit load.
18 TAA 9 A. If the Reactor Protection System does not produce an immediate trip as a result of the spurious SIS uignal, the reactor experiences a negative reactivity excursion due to the injected boron causing a decrease in reactor power.
a
- a. The power mismatch causes a drop in Tavg and consequent coolant shrinkage in the system results in pressurizer pressure and water level drop.
- b. Load will decrease due to the effect of reduced steam pressure on load after the turbine throttle valve is fully open.
- c. If automatic rod control is used, these effects will be lessened until the rods have moved out of the core. The transient is eventually terminated by the Reactor c Protection System low pressure trip or by manual. trip.
- d. The time to trip is affected by initial operating condi-tions including core burnup history which affects initial boron concentration, rate of change of boron concentration and Doppler and moderator coefficients.
91' TAA 46 Q. In performing transient analysis, (1) results of~such analysis are compared to criteria specified in 10CFR100. What are the criteria?
92 TAA- 46 A.. Whole body dose < 25 REM Thyroid dose < 300 REM Both for a two hour stay time at the exclusion area boundary.
~
t t
- 95. TAA 48 Q. For a steam generator tube rupture (1) what principal indication (s) distinguish this event from other RCS inventory loss transients?
96 TAA 48 A. Steam Jet Air Ejector Rad Monitor Alarms or Condenser Off-Gas Rad Monitor Alarm.
Steam Flow / Feed Flow mismatch.
Main steam line rad monitor alarms.
S/G Blowdown rad monitor alarm.
Increased S/G Level on affected S/G.
99 TAA 50 Q. Upon identifying that an ATWT event (1) has-occurred, there are three things an operator must do to mitigate the event irrespective of the type of ATWT. List the three basic actions.
100 :TAA 50 A. 1. Try to trip the reactor
- 2. Try to trip the turbine
- 3. Insure that a heat sink is
[ available L .
i 61- MCD 31 Q. .a . What are the RCP trip criteria (2) specified by EOP 0.0 Rx Trip or Safety Injection?
f~ b. Why is it recommended to trip
[- RCP's with the above criteria on a small break LOCA?
l l
l
62 MCD 31 A. a. Component Cooling is lost and upper or lower bearing temperature is greater than 200*F.
-OR
. SI is on and RCS pressure < ~
1700 psig.
will tend to pump water out of the system for some break locations. Also, their power supply is not safety related c <
and, therefore, not guaranteed. During a small i~ break LOCA, a loss of the RCP's.could cause a more severe condition than
~ ~
analyzed. . .
5 89 MCD 45 J. What are the RCS operational (3.0) leakage limits as specified by Technical Specifications?- No action statements required!
I
-90 MCD 45 A. (.5 each) a.. No pressure boundary leakage
- b. Max of 1 gpm unidentified leakage
- c. 1 gpm S/G primary / secondary leakage total or 500 gpd/S-G
=
- d. 10 gpm identified. leakage
- e. .40 gpm controlled leakage at 2235 i 20.psig j 1 f. 1 gpm.from any RCS Pressure p Isol. Valve-at 2235 20 psig
?
f-'
f
.s e n-e ,w,,. , ,,,y-- ,..e , , , ,--er---~~, -e
-. n. 4 > , - - - - - - ~ , -,, .e e-w ,,..-m-,-----r- ,-,,,-w, ,, w.
y ..
115*- MCD 58 Q. a. What are the six Critical
- (3. 5) - Safety Functions in order of priority from most important ,
(1.5) to least important?
, b._ Choose-two CSF items and
. . explain what parameters /
systems must be monitored to determine if the CSF is satisfied. (2.0)
-116 ,,
MCD 58 A. a.- (.25 each)
- 1. Subcriticality
- 2. Core Cooling
- 3. RCS Integrity-
- 4. Heat Sink
- 5. Containment
- 6. Inventory _
- b. (1 each)
- 1. Subcriticality 2. Core Cooling
'SR, IR, PR Scales Subcdoling Margin:
SR, IR SUR meters. -Thermocouples, loop wide
~
range temperature, Pzr pressure, Loop pressure
.3. 'RCS Integrity 4. Heat Sink RCS pressure -(Pzr or Loop) RHR Sys Parameters-vs. Thermocouples RCS temperature Aux Feed ~ Flow (Thermocouples, Loop S/G Narrow Range Levels temperatures.Th, Tc) _ S/G~ Pressures Steam Dump or atm relief availability 5, . Containment .6 . Inventory Pressure Pressurizer Level Radiation Level Sump Level-3 LI
9 .,
37 ERG 19 Q. In the' Emergency Contingency Action (1) procedures related to " Loss of All AC Power" and its recovery, special attention is given to the Reactor Coolant Pump Seals. Why?
38 ERG 19 A. The RCP Seals are the only normal unisolable RCS leak paths (0.5).
They_are very sensitive to quick temperature changes and can there-fore be damaged due to thermal shock causing a large increase in RCS leakage. (0.5) 67' ERG 34 Q. State 4 different methods of (1) restoring the primary heat sink or other means of removing heat from the primary as described in FRH-0.1
" Response to Loss of Secondary Heat Sink".
68 ERG 34 A. 1. Establish AFW flow to the steam generators (.25)
- 2. Establish MFW flow to the steam generators (.25)
- 3. Feed a steam generator with a condensate pump (.25) 4.- Remove heat from the RCS using fccd and bleed - SI to RCS and bleed through the PORV's (.25)
O
- . . es 21 ERG 11 Q. List four (4) actions that must be (1) taken to isolate the ruptured steam generator (s) per EOP-3.0 (SGTR).
22 ERG 11 A. a. Isolate auxiliary feedwater.
(.25)
- b. Close main steam-isolation valve, bypass valve and drains.-(.25)
- c. Verify S/G PORV's closed.
(.25)
- d. If appropriate, close steam supply valve from affected steam line to AFW pump 9 DG 5 Q. List the automatic start signals (2) for the diesel. Can the diesel be-auto started with the mode selector switch in local?
.10 DG 5 A. Undervoltage on the respective safeguards bus.
Undervoltage on the preferred-power sourCO.
Safety injection.-
When the selector switch on the local diesel control panel is in local, the automatic start signals are blocked.
[ r-s - . .
Shift Advisor Examination Results COURSE SHIFT ADVISOR EXAM 1 EXAM 2 EXAM 3 AVERAGE 2
D. E. Burton 83 1 80 73 2 79 D. R. Campbell
. 79 80 69 76 H. C. Crummey 90 90 85 88 2
L.,J. Ryan 87 85 75 2 82 S. Stevens 88 89 79 85 F. D. Pauli 3 NOTESi 1. Upgrade Training Completed.
- 2. Upgrade Training Scheduled.
- 3. Training Not Completed.
m