IR 05000352/1988016
| ML20151R255 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/03/1988 |
| From: | Howe A, Lange D, Walker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151R195 | List: |
| References | |
| 50-352-88-16OL, NUDOCS 8808120104 | |
| Download: ML20151R255 (256) | |
Text
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION REGION I-OPERATOR-LICENSING EXAMINATION REPORT EXAMINATION REPORT N0. 88-16 (0L) FACILITY DOCKET NO. 50-352 FACILITY LICENSE NO. NPF-39 LICENSEE: Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 FACILITY: Limerick 1 EXAMINA110N DATES: June 7-9, 1988 CHIEF EXAMINERS: Od%
. Waller, Senior Operations Engineer -
fr/3/&f( Dite aan >c. L A. Howe, Senior Senidr_ Operations Engineer sk/n Dite' APPROVED BY: * W David J. Lange, Rhief," BWR Sd: tion, JP-8-8P Date Operations Branch, Division of Reactor Safety
,
j
!
SUMMARY: Written examinations and operating tests were administered to three . senior reactor operator (SRO) and two reactor operator (RO) candidates. Two l SR0s and one R0 passed the written examinations. One SR0 and.one RO failed the written examinations. Three SR0s and one RO passed the operating examinations.
One RO failed the operating examination. )
8808120104 880805 PDR ADOCK 05000352 V PDC
. . - _ - _
- DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: l RO l SR0 l l Pass / Fail l Pass / Fail l l 1 1 I I I I I I I l Written l 1/1 l 2/1 l 1 l i I l l l l 10peratingl 1/1 l 3/0 l l 1 l l l l l l 10verall l 1/1 l 2/1 l 1 1 I l 1. CHIEF EXAMINERS AT SITE: T, Walker, Senior Operations Engineer A. Howe, Senior Operations Engineer 2. OTHER EXAMINERS: S. Pullani, Senior Operations Engineer T. Fish, Operations Engineer J. Hanek, EG&G (Examiner) M. Parrish, EG&G (Examiner) D. Lange, Chief, BWR Section (Observer) N. Conicella, Operations Engineer (Observer) T. Easlick, Operations Engineer (Observer) 3. The following is a summary of generic strengths ano deficiencies noted on the operating tests. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.
3.1 Strengths
*
Knowledge of Remote Shutdown Panel and associated procedure
*
Familiarity with Piping and Instrumentation Drawings (P& ids)
* Security awareness , '
v
.
- . _ . _
3.2 Deficiencies
* A general unfamiliarity to the Fuel Handling Casualty procedures were noted among the candidates. The facility was requested to check into the adequacy of training in this area.
- A general unfamiliarity with Administrative Procedures _(for :
L example, Shift Turnover Sheets) were noted among some candi-dates. The facility was requested to verify the adequacy of training received by the candidates in this area during their 3 months on-the-job training.
* Annunciator Response Procedures (ARPs) filed in a single binder were not easily accessible to the candidates during the simu-lator examination. . The "required" category of ARPs, identified by a triangle on the annunciator windows were required to be referred to, once such annunciators are received. However, the-candidates were not referring to such procedures during the simulator scenarios.
Inadequate communication among the operating crew were noted during certain events of the simulator test scenarios. Examples of this were: (1) EHC Pressure Regulator Oscillation event and (2) Turbine Trip with Failure to Scram event. The details of these deficiencies were discussed with the facility during the exit interview.
4. The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.
4.1 Strengths
* Knowledge of the effect of changes in related core parameters on Shutdown Margin (SDM). (Question 1.02) *
Knowledge of meaning of "Prompt Jump" and "Prompt Criticality".
(Question 1.04)
*
Knowledge of the effect on Control Rod Worth by changes in related core parameters. (Question 1.06)
* Knowledge of the effect of Xenon transients due to a scram from full power operation and subsequent startup on the radial flux-shape and resultant change in Control Rod Worth of peripheral rods. (Questions 1.07 and 5.03)
. .
*
Knowledge of "Critical Power" and how it changes due to changes in related core parameters. (Question 1.11)
* Knowledge of automatic actions that occur on a LOCA signal while one loop of RHR is in Suppression Pool-cooling and the reactor is operating at power. (Question 2.02) * Knowledge of Diesel Generator trip signals while connected to the normal supply. (Question 2.03) * * Knowledge of trip and interlock functions of SRM, IRM and APRM.
(Question 3.04)
*A deficiency for SR0 candidates. (Question 6.10) *
Knowledge of how RCIC speed controller controls in the turbine speed in various modes of control. (Question 3.11)
*
Knowledge of the guidance (in Administrative Procedure A7,
"%1ft Operation") concerning when a plant shutdown or scram shall be initiated by SLO or LO. (Que'tions s 4.01 and 8.03) , * Knowledge of the requirements in 10 CFR 55 concerning the limitation on who may manipulate the controls of the reactor.
(Question 4.01)
* Knowledge of the entry conditions and immediate operator actions for Special Event Procedure SE-1, "Remote Shutdown" (Questions 4.02 and 7.05) *
Ability to work thermodynamic problems using steam tables.
(Question 5.05)
{
*
Knowledge of Tech. Spec. for core thermal limits. (Question 5.09)
*
Knowledge of plant response due to loss of feedwater flow transients. (Question 5.10)
*
Knowledge of automatic actions initiated by Process Radiation Monitors. (Question 6.07)
*
Knowledge of plant response due to instrument / signal failures in the Feedwater Level Control System. (Question 6.09)
*
Ability to distinguish situations requiring an RWP. (Question
- 7.01)
* Knowledge of station administrative dose limits. (Question 7.02) *
Knowledge of entry condition to TRIP procedure. (Question 7.08) i
<
Knowledge of Safety Limits and Thermal Limits in Tech.. Spec.
~
* (Question 8.01) * Knowledge that Tech. Spec. could be violated to comply with a TRIP procedure. (Question 8.08) * Knowledge and ability to identify Tech. Spec. LCOs and Actions for Primary Containment Ir.tegrity. (Question 8.09).
4.2 Deficiencies
* Ability to identify when the. reacto Ni ~
'ing range after it is critical and on a steady per*, d .- m an 1.05).
- Knowledge-of the effect of Doppler . oc7hc on a mitigation of a power transient. (Question 1.08(a))
* Knowledge of the contributing factors fon Required" and "Available" NPSH. (Question 1.10) *
Knowledge of the automatic response of the Core Spray system pump and valves during a Small Break LOCA. (Question 2.05)
*
Knowledge of automatic response of RHR (valve realignment) from its shutdown cooling mode due to a LPCI initiatiore. (Question 2.06(e))
* Knowledge of the valve actions on a RFP trip.
' (Question 2.09 (b))
* Knowledge of the response of the recirc. pump speed on a conden-
- sate pump trip while operating at power. (Question 3.01(b))
* Knoweledge of the source of error in the indicated level of the reactor level instruments. (Question 3.02) *
Knowledge of the initiation signals for MS!V isolation during reactor startup. (Question 3.03(a)) !
*
Knowledge of the response of the FWCS on a loss of level signal and its effects on plant. (Question J.08) ; '
*
Knowledge that a Group I isolation will be received if the .l reactor pressure is reduced to 900 psig in an attempt to reseat ' a stuck open SRV. (Question 4.03(e))
*
Knowledge and general understanding of Administrative Procedure A41, "Control of Plant Equipment".
- Knowledge of the recuirements for entering a high radiation area. (Question 4.08(b)) j l
l
, . . . _
.
.
* . Knowledge of the requirements for implementing a temporary change to an approved procedure. -(Question 4.09(b))
.
* Knowledge of relative worth of shallow and deep control reds.
(Question 5.04)
* Knowledge of. plant.responseLon a loss of feedwater system.
(Question 5.10)
* Knowledge of the requirements'for a procedure to be considered valid and for implementing temporary changes to an aoproved procedure. (Question 8.04) * Knowledge of th. time. limit for release of.TecF ' Spec. . ,
equipment for surveillance testing. (Question 3.06(d)) 5. Personnel Present at Exit Interview, June 10, 1988 5.1 NRC Personnel N. Conicella, Operations Engineer T. Easlick, Operations Engineer A. Howe, Senior Operations Engineer T. Kenny, Senior Resident Inspector S. Pullani, Senior Operations Engineer 5.2 Facility Personnel E. Firth, Superintendent - Training G. Leitch, Vice President - Limerick Generating Station D. Neff, Licensing Engineer R. Nunez, Operations Training Supervisor D. Weiksner, Instructor S. Wilhelmson, Lead Instructor A. Yarmer, Simulator Instructor 6. Summary of NRC Comments Made at Exit Intervieu
! *
The written examination was conducted on Tuesday, June 7, 1988. The candidates had very few questions during the written examination.
The facility review of the written examination was conducted on Tuesday, June 7,1988, imediately after tiie examination. No signi-ficant prnblems were identified during tha examination review. The facility was reminded to send their written comments to the NRC and EG&G within five working days.
The Training Department was cooperative during the examinatic-process.
l l
-
v v, w
,
!
1
* The operating tests were conductedaon Wednesday and' Thursday, June 8 and 9, 1988. 'The generic strengths and weaknesses noted on the operating -tests (see .Section 3. of this report) were presented.
- Distractions to the. candidates from outside personnel through the.
glass doors and windows of the simulator during the operating test ' were noted by the examiners which were corrected by the Training Department later.
- The need for isolating the candidates from outside personnel in between the~ simulator test scenarios were noted. A separate room could be arranged for candidates to wait-in between scenarios. How-ever, no significant' problems were noted during the tests.
Significant on-the-spot changes were required to be made in the simu-lator scenarios originally prepared by the NRC consultant before they
.
are administered to the candidates on June 8 and 9, 1988. This was because.the Simulator Malfunction Book, sent to the NRC and the consultant and used for the preparation of the original scenarios, was subsequently revised in the numbering and scope of malfunctions.
Several problems concerning the simulator-fidelity were encountered during the execution of the simulator scenarios. The details of these problems were discussed with the facility during the exit meet-ing (see Attachment 5).
There were no problems with access to the plant. .The operations staff was cooperative during the plant walk through portions of the-examinations.
The results of been the written and operating examinations would not be discussed at the exit meeting but would be contained in the_Exami-nation Report. Every effort would.be made to send the candidates' results in approximately 30 working days.
The reference materials provided to NRC for preparation of the written examination were generally adequate. However, the lack of specific Learning Objectives (L0s) to match several Knowledge and Abilities (K/As) in the K/A Catalog (NUREG-1123) were noted by the examiners. The facility perr'nnel stated that they have plans to correct the deficiency before the next license examination is conducted.
.ttachments: 1. Written' Examination and Answer Key (RO) 2. Written Examination and Answer Key (SRO) 3. Facility Comments on Written Examinations after Facility Review 4. NRC Resolution of Facility Comments 5. Simulation Facility Fidelity Report _
. . . - _-- ., - _ _ _ , - . ._ _
' l,
h77ACH ME A> T i
. '...
t / U. S. NUCLEAR REGULATORY COMMISSION
.
REACTOR OPERATOR LICENSE EXAMINATION FACILITY: LiegCiEh_______._________ E REACTOR TYPE: @WB-Qg4__________________ DATE ADMINSTERED: @@4Q6496_________________ EXAMINER: NRQ_BgglON_1_____________ CANDIDATE _________________________ INSIBUGII9N5_IQ_G6ND199IEt Write answers on one side only.
Une separate paper for the answers. sheets. Points for each Staplo question sheet on top of the answer The passing question are indicated in parentheses after theand question. grade of at grade requires at least 70% in each category up asixfinal hours after least 80%. Examination papers will be picked (6) the examination starts.
[
% OF CATEGORY % OF CANDIDATE'S CATEGORY
___@CQBE___ _y86ME__ ______________CSIEQQBy_________.,___ __M86UE_ _IQIBL 23 B t PRIF'?LES OF NUCLEAR POWER _2SAQQ__ _2drd@ ___________ ________ l. '1PERATION, THERMODYNAMICS, PLA HEAT 1ANSFER AND FLUID FLOW 25.zs 23.54 LANT DESIGN INCLUDING SAFETY _9trfR__ _El 12 ___________ ________ 2.
AND EMERGENCY SYSTEMS ZG.6% ________ 3. INSTRUMENTS AND CONTROLS _26 2Q__ _e rt@ ___________ E4.') 5 ti.oq PROCEDURES - NORMAL, ABNORMAL, _EEt2E__ _&@rt? ___________ ________ 4.
EMERGENCY AND RADIOLOGICAL CONTROL
' -
18 95 Totals _k99r2__ ___________ ________% Final Grade All wor k done on this examination is my own. I have neither given nor received aid.
_-____________________'__________.- Candidate's Signatureb
,-
*
l l l
i . l l
. . .
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Ducing the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. R3stroom tr.ips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3. Une black ink or dark pencil only to facilitate legible reproductions.
4. Print your name in the blank provided on the cover sheet of the examination.
5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8. Consecutively number each answer sheet, write "End of Categort __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to categcry and number, for example, 1.4, 6.3.
10. Skip at least three *ines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14.'Show all calculations, methodr, or assumptions used to obtain an answer to mathenatical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE ,
-
QULSTION AND DO NOT LEAVE ANY ANSWER. BLANK. - i d 16. If parts of the examination are not clear as to intent, ask questions of the examiner only. ,
'
17. You must sign the statement on the cover. sheet that' indicates.that the work is your own and you have not received or been given assistance"in completing the examination. This must be done after'the examinat' ion has been completed.
'
._, - -
o
. * . .
e 18. When you complete your examination, you shall
.
a. Assemble your examination as fullows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b. Turn in your copy of the examination and all pages used to answer
,the examination questions.
c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
' d. Leave the examinatio.1 area, as defined by the examiner. If after leaving, you are found in this area while the axamination is still in progress, your license may be denied or revoked.
- i l l
#
)
-.. - . . - , _ . .. , -. - , , . . - , . . _ . , .-
, . ,
is PRINCIPLES _QE_NUQ6588_EQWEB_ELGNI_QEEBOIlON S Pagn 2 ISEBd99XU001GEi_UEGI_IB0NEEEB_009_ELUID_ELOW t .
-QUESTION 1.01 (2.50)'
Concerning Prompt and Delayed Neutrons state whether EACH of the following TRUE or FALSE: a. The percentage of delayed neutrons produced from fission increases as the age of the core increases. (0.5) b. The energy level at which delayed neutrons are produced categorizes them as thermal neutrons. (0,5) c. Neutrons produced from the moment of fission until 10 E-14 seconds are considered prompt. (0.5) d. Delayed neutrons are produced as a result of both thermal fission of U-235 and fast fission of U-238. (0.5) e. Delayed neutrons are the major factor in determining the rate of reactor power decrease immedi atel y (after 1 second) following a scram. (0.5)
.
Ti s
, ;- >
* (
l I
{
l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
-_
_ - - _ _ _ . - . ~ . . _,. -. _ - , _ _ _ _
. . . .. ~. . . .. .
j- .
,, .. .
j i j. .. i QUESTION 1.02 (2.00) ! Will' Shutdown Margin.(SDM). INCREASE or DECREASE for'each of the l' following? a. Poison concentration increases (0.5) b. A number of control rods are inserted into the core (0.5) c.- Moderator Temperature increases (0.5) 4 d. Plutonium 240 concentration increases.- (0.5) s
.
i l l l i l
.
'l
,
i
'
!
,
-
.l'
p,
-
C , ,
, ,
_ ,
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
. . . - - - - . - - - . . - - .
. . , . .. . , . .-- . - - . . .. ., . ....-- .. 2 . ::- .- . :.: . .,
. . - - , i
, 6 D
.
QUESTION 1.03 (1.50) You are performing a normal reactor startup; as K-eff of the reactor increases, STATE the effect (INCREASE or DECREASE) on each of the following pars.neters-for equal reactivity additions, a. The magnitude of the change in count rate -
(0.5)
6. The rate of rise in the count rate (0.5) c. The time it takes to reach a new equilibrium count' rate (0.5)
.
l
!
;
l l i
1
' .
t
.
9 , c.
- p, .
,
, (h.
.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) . _- _ _ _ . _ _ _ . ..
_ . _ . _ _ ._..;.._.. .. .r. . :- .
. . . - _ . . .-. -
.
. .
QdESTION 1.04 (1.00) a. During a rod witndrawal the operator will observe a suriden increase in reactor period initially. This suddsn increase is referred to as _______________________. (0.5) 6. If the reactor were critical on promot neutrons alone power-would increase uncontrollably. This. condition is called ~
(0.5)
_________________. < l l
,
.
{
h r
6
%r
-
<
.
,
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- - - . , _ _ .- _ . . _ . . . . . . _ _ - . . . . . . . . . _ . _ _ - _ _ _ . _ _ - . . - , . . - - . _ - - - . - _ .
.
b -
, ,
)
,
)
'
l
,
l QUESTION 1.05 (2.00) You have just taken the reactor critical from a cold condition and are ! increasing power on an 80 second period.
O. Utiliz.ing control room instrumentation, STATE two methods which will tell you the heating range has been. reached? Rod position and recirculation flow have been held constant. . (1.0) b. In which ONE of the following intervals was the heating range entered? , (1.0)
(1) Interval 1 - reactor power increased by a factor _of 6 in 143.3 I seconds.
(2) Interval 2 - reactor power increased by a factor of 3 in 99.0 seconds.
(3) Interval 3 - reactor power increased by a factor of 5 in 128.8 seconds.
,
I l i
\
l i
.
-
-
'l +
. ,
*
l
'
. L ..
.
'
,
,
'
/ I . !Ts
*
,
..
,
- t ,
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) - -- - - _ . , _ . . . . , _ . ...
. . _ _ . .- . . . . - - - .
.
. . .
OdESTION 1.06 (2.00) Will control rod worth INCREASE, DECREASE, or REMAIN THE SAME for each'of.the following?
(0.5)
a. Increasing moderator temperature (0.5) b. Increasing the percent voids (0.5) c. Increasing the fuel temperature
.
(0.5)
d. Increase in Xenon concentration following a power change
.
;
\
\
!
' .
'
>
:;
(***** CATEGORY 1 CONTINUED ON NEXT PAGE an***)
- - - - - ~ . . . . . - . . .. - _ . _ , l
,
* . .
I QUESTION 1.07 (1.00) l You are perf orming a reactor startup 12 hours following a scram which f occurred after 30 days of full power operation.
WHICH statement below describes the expected effects of Xenon concentration when performing the startup? CHOICES: 1. Thermal neutron flux will be highest in the same areas where the flux was highest during the praavious operational period.
2. Thermal neutron flux will be higher in areas of high Xenon concentration than during the previous operational phase to maintain the same reactor power level.
3. Thermal flux will be pushed to the periphery of the core, making the periphery rods have a high rod worth.
4. Xenon burnup during the reactor startup will make the reactor go critical earlier in the rod sequence than is normal.
l
l
'
l i
. .
l
t-f (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- . - . - - . - - . .. , , - . , ,
l
! . , ,
.
)
QUESTION 1.08 (3.00) During operation at 100% power a f eedwater train automatically isolates due to high water level in a heater.
STATE how each of following coefficients of reactivity will respond to MITIGATE or INCREASE the severity of this transient. Include a brief reason in your answer and consider the entire transient UNTIL the scram occurs.
(1.0) a. Doppler Coefficient Moderator Temperature Coefficient (1.0) b.
(1.0) c. Void Coefficient
.
!
.
c.g i
*
l
, (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) .- - - .. . .. - . - . - . .. -.
'
, .
\
.
l
'
QUESTION 1.09 (2.00) A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being OOC. The GAF's were computed, but'the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE.
DETERMINE if each of-the following statements.is TRUE or FALSE.
a. If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power. , (0.5) 6. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power.is HIGHER than the currently calculated power. (0.5) c. If the steam flow used in the heat balance calcalation was LOWER than the actual steam flow, then the actual power is HIGHER than the currently calculated power. (0.5) d. If the RWCU return temperature used in the heat balance calculation was LOWER than the actual RWCU return temperature, then the actual power is HIGHER than the currently calculated power. - ( 0. 5) s e i,
,
.
,
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- - -- _. -- _
__
.
l
, ,
-
,
l QOESTION 1.10 (2.50)
.
c. LIST FOUR parameters which contribute to AVAILABLE NPSH (Net Positive Suction Head) for a recirculation' pump. Limit your answer to those parameters which are DIRECTLY indicated in the CONTROL ROOM. ,
(1.0)
b. Consider TWO Reactor Plant conditions: 1-Low Power and Low Flow (< 10%) '
'OR High Power and High Flow (>85%). .
1. During which condition is the REQUIRED NPSH for a recirculation pump greater? (0.5) 1
\
2. During which condition is AVAILABLE NPSH for a recirculation pump greater and WHY is it greater? (1.0)
.
',
.
,
,
,. >
'
+
' "y (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
.- - .. - - - - . _ _ . . . . .
QUESTION 1.11 (3.00) i a. Define critical power. (1.0) ' b. For EACH condition (a.-d.) given below, INDICATE whether it will cause an INCREASE, a DECREASE, or have NO EFFECT on CRITICAL POWER.
1., Local peaking factor (LPF) INCREASES - -
(0.5)
2. DECREASE in inlet.subcooling (0.5)
'
3. INCREASE.in reactor pressure (0.5) 4. Axial power peak shifts from BOTTOM to TOP of channel (0.5)
.
'
' '
~ [: , -
. t+,q
*
1 .
*
,,
y.. n
.
Ci
^
, . . t < . (,.
-
'
g >s '
*
..- , .', ,.
-
.-
- y; - 4 9; V; .
.' t ,f qt .p. .j j ' ' 8. ' *
* '
- "" - g , ,g . ' e %'g
.
<
'
.- 4 l
1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) __
.,. . .
_
-. . .. . _ . . . . _.
-
. . . .
i !
-QUESTION 1.12 (2.00)
With regard to MAPRAT: ai WHAT is the relationship between MAPRAT~and MAPLHGR7 (1.O) ,.
,
b) WHAT physical consequence could occur if the MAPLHGR Technical Specification limit is exceeded? (1.0)
,
*
\
.
i
,
!
I i l l l l l l l
~
l l i
\
v
.
s g- .e s
. (***** END OF CATEGORY 1 *****) .. - --- .. . - _ . . - - . , .
.
. .
2i__EL8NI_ DESIGN _IUg69Q1NQ_S@EEIY_8NQ_ENEBQEUgY Pcge 14 SYSIEUS
-
QUESTION 2.01 (3.00)' Answer the following questions concerning operation of the High Pressure Coolant Injection (HPCI) system.
a. STATE the minimum speed that the the HPCI. turbine should be operated at AND TWO reasons for NOT operating the turbine below the minimum speed. , . (1.5) b. The operator is observing the automatic initiation of the HPCI system. The minimum flow valve should close at what system flow?
(0.5)
c. HPCI is being operated in the test mode when a high drywell pressure initiation signal is received. LIST two signals that will cause the HPCI Test Bypass to CST to close. (1.0) s
-
,
d
*
4 s
", .), ,;-)r.
'
'
.- '
< - -
'
. ,
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) . _. __
l
. . .
QOESTION 2.02 (3.00) Reactor is operating at power with the B loop of RHR in cuppression pool cooling. An inadvertent LOCA signal is generated.
STATE if the operator WOULD or WOULD NOT observe each of the following conditions, a. The ESW pumps are running 30 seconds after the diesels start. (0.5)
(0.5)
b. The CRD pumps are running.
c. The RHR SW pump continues to operate to supply cooling to the RHR (0.5) heat exchanger.- d. Turbine building equipment compartment exhaust fan is tripped.
(0.5) e. The control room chillers are operating 2 minutes after~ the diesels (0.5) start.
f. The load center Transformer Breakers Dil4, D124, D134, and D144 are closed 5 seconds after the. signal is generated. (0.5)
.
.: ,
r r . l l
's y e
l I l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) _ . _ - -- , . -- . - _ _. . . ._- - _ . . _ -- -.
_ I
' . .
' '
< .
!
;
QUESTION 2.03 (2.00) , A diesel generator is operating for a surveillance in parallel with.
the normal supply. The diesel trips during operation. List 6 ; possible causes that may have caused the diesel-engine to trip.
(Setpoints not required)
.
l
l
/
l
I l l l I
.
e , a i
-
.
.
-
,.
.
. . -
t L
,
-
.
. - ,
.
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-.... . ,_..... .... . ._..- ...- - _ _ , _ , . , _ _ _ .- ., _ . . -- . . . . _ . - . . . . _ . . . - . -
. . .
e l l QUESTION 2.04 (3.25) l Answer the following questions concerning automatic initiation of the SBLC system. l
c. A reactor high pressure signal of 1093 psig exists. STATE two additional conditions that must exist prior to automatic initiation l occurring. (1.0) ! I b. State the reactor level at which an initiation signal for l automatic initiation of SBLC is generated. (0.25) l c. State the four actions that occur if. an automatic injection of SBLC occurs. Do not include redundant components as separate actions.
(2.0) i l l l
*
,
i l
l i' l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) _. . - - - _ . - -- .. - - - .- -.- .,- . _ . . . .,- - - - . ~ . .
. . . .
QUESTION 2.05 (2.50) A small leak develops which results in depressurization of the reactor at a rate of 30 psi per minute. HPCI maintains reactor water level but drywell pressure increases to 3.2 psig.
c. For each component in the Core Spray system listed STATE the pressure or flow rate at which the component will operate during the depressurization.
1. Minimum flow valve closes. (0.5)
^
,
2. Injection valves open. (0.5) l
.
3. Testable check valve disk opens. (0.5) l 4. Core spray pump starts. (0.5) l b. What is CS rated flow and at what pressure should the operator observe this flow. (0.5) i i l
,
I l
.
. .
t
r; 45, v.
'
; ..
4 .., . ; g, .
.
,
,
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-. . - - . -
. . ,- - , . . . - . . .- , - - - - _,.
. . . .
QUESTION 2.06 (3.00) The reactor is in cold shutdown with loop A of RHR in shutdown cooling. Loop B is inoperable. A break results in a loss of reactor inventory, n. At what reactor level will the S/D cooling isolation occur?
(0.25)
6., LIST the RHR system valves that will close or receive a close signal when the S/D cooling isolation occurs. (1.0) c. At what reactor level will RHR receive a LPCI initiation signal?
(0.25)
.
d. WHAT action (s) would the operator have to take other than arming and depressing the manual LPCI initiation pushbuttons in order for loop A of RHR to inject in the LPCI mode? (0.5) e. LIST the RHR loop A valves that will automatically reposition . I when the LPCI initiation signal is received. (1.0) I J l
l l l
*
.
.
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
-
- _ . - - - - - ..
,, . . . . ._ . . . . . _.. ._ . _ _ . '
e ,
.. .
.: .. . l
~
!
<
-l i
QdESTION: 2.07 (2.00) c .' - LIST two. conditions;that will result 'in an of f gas isol ation. 1
! (Setpoints not required) . -(1.0)
i
;-
~
b. LIST-two automatic actions that occur.due to an:off gas , isolation. ii.O)- . i
;
I
't
~.
*
i
-
,
.
I
. 1
- . .
> ,. ., u
' '
, ag
-
', W. u- ;
4 4 i l I l i
!
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
.t-l
-- r e x- , - 4 , ,, . - - - , - . . , , . ~ , . . . . . . ._......-,..v.. .._,m._.m.. ,,,m,,- .w.y,, 3 , ,c._.,,....-,, ,~, , . , ~ . , - - - , .g.,,,,,..
- _
i
. .
l
-
;
QUESTION 2.08 (2.50) i
!
A complete loss of RECW has occurred due to a rupture'in the suction line from-the head tank.
a. Assuming no operator action, WHAT 3 actions will occur in the RWCU system due to the loss of RECW7 (1.5) b. WILL the RECW pumps automatically trip due to the loss of suction?
(0.5)
c. WHAT system can the ' operators use to supply flow to the RECW system? (0.5)
.
.
.
.
.
!
. .
.
'
'
,
i (***<* CATEGORY 2 CONTINUED ON NF.XT PAGE *****) i
:
-. , - . . . . . . , -.... -.. .. ... -
- - . . - . . . .
_.
.
.
QUESTION 2.09 (3.00) The reactor is operating at 98% power when a RFP trips, a. LIST 4 of the possible causes f or the trip other than manual trips.
Only a single RFP has tripped and no other plant equipment has tripped. (Setpoints are not required) (2.0) D. LIST 5 valve actions ' associated with the tripped RFP that the operator will observe. (1.0) t
)
l i
,
'
e ,
. (***** END OF CATEGORY 2 *****)
. ..- -. - - - - - - ,- . . . - . , , . , - , ,. . . - - - . - .. . - . ,
. .
31__INSIBUdENI5_AND_CONIB065 Page 23
.
QUESTION 3.01 (3.00) STATE what EFFECT each of the conditions will have on 1A and 1B Reactor Recirculation Pump speed. (Values for speed are required.)
a. Both retirc pumps are operating at 35% flow. The master controller is set at minimum. An operator inadvertently places the 1A M/A Transfer Station in Auto. LA<c)
. _..
j,g-b. Both recirc pumps are operating at 95% flow. Condensate pump 'A' trips. LLxn
/, I c. 1B recirc pump is started with its M/A transfer station in automatic and master controller is at 75%. 1A is at 28% speed with its M/A transfer station in manual. (1.0)
&O-l l
!
GJ5
.\ lk '
o# fokl
~
< 4 V' )
. L 7'0 A-l'
.
b' l .0
. (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) - . ---_--- -- . . - . ~ , - - - , .
-
\
.
.
OljESTION 3.02 (2.50) For each of the f ailures listed below state whe'.her indicated level monitored by the cperator is HIGHER THAN ACTUAL level, LOWER THAN ACTUAL level or the SAME AS ACTUAL level.
a. A break occurs in the reference leg. (O.5) b. A wide range instrument is used when reactor pressure in 1000 psig. g (0.5) c. A 50 F increase in drywell temperature occurs. j (0.5) d. Narrow range level instrument is used when reactor pressure (0.5) is 75 psig, e. Reactor pressure is below saturation temperature f or the drywell.
(0.5) s
.w
- 4 q, .
c
,.l s
'.
, , ; 1
. .-
kr
( ft
*+ +
;[- r. :s ,a . .. ,
::
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
-.. ~-. .....
, . - . . . .
PUESTION 3.03 (2.75) a. The mode switch is in startup with reactor pressure equal to 800 psig when a MSIV isolation occurs. LIST the possible signals that could have cauwed the isolation. (Include setpoints)
(2.25)
b. . (TRUE or FALSE) A scram will occur if NSIVs close in only two Main
, Steam lines.- ,
(0.5)
.
I i
'
.
.
1- .
'
i ,
,
P (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) _. _ _ _ . . - _ _ _ _ - _ . _ .
. , _ _ . _ _ .,-, _ _ - _ _ _ . , _ . _ _ _ . _ - . _ _ __
- . _ _ . . _ . _ _ , _ . _ _ , ..
. . . .
OdESTION 3.04 (3.00) For each condition listed below SELECT-the action (s) (listed below) that will occur.
STATE ' no action will occur.
c. A reactor startup is in progress with IRMs on range 2. The operator withdrawing SRM detectors also has channel A of the IRM-
'
selected. (.5) b. The operator ad justs r ecirc flow such that there is a 15% mismatch between loops. (.5) c. Reactor is in the RUN mode and an APRM fails downscale. All IRM are withdrawn and indicating 25 on range 3. (.5) d. Reactor is in startup mode and control roos are withdrawn to 15% power.
(.5) o. An approach tc criticality begins with IRM C on range 2. (.5) f. APRM channel 'A' Mode switch is placed to standby. (.5) Actions: 1. IRM 'A' detector will not withdraw.
2. A scram signal is generated ' 3. A rod block is generated.
l
' ,
w.
M <
'
.
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l l
.. - .. . - . -- ., - .
.,
. .
B QUESTION 3.05 (2.50) a. Reactor water level has decreased to below -129 inches. Drywell pressure is 1.2 psig. FsHR and CS pump interlock is satisfied. ADS valves should open in ______ seconds. (0.5) 6. Reactor water level has decreased to below -129 inches. Drywell pressure is 4.6 psig. RHR and CS pump interlock is satisfied. ADS valves should open in ______ seconds. , (0.5) c. Reactor water level decreased to below -129 inches but recovered to greater than -129 inches prior to either ADS timer timing out.
Will an ADS actuation occur? (0.5) d. An ADS blowdown is in progress. Will the blowdown stop if the operator pl aces t.' ADS Auto Inhibit Switches to INHIBIT 7 (0.5) e. An operator observe that both the green and the amber lights are li't for the acoust.. monitor for an SRV. Describe what informat4cn this provides to the operator concerning the SRV. (0.5)
.
l l l
;
u (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
.- ._ _ _ - _ .
I
. . . .
l
'
QWESTIOr4 3.06 (3.00) l The reactor is at 50% power with the load limit set at 65% and Maximum ' combined flow limiter at 115%. An electrical failure occurs that causes the pressure set signal to' decrease 10 psi.
, . . . DETERMINE the final control valve flow rate and bypass valve flow rate. Refer to the attached drawing of the Electro-Hydraulic ,
'
Control Looic (LOT-0590-6). DESCRIBE how you-determined your answer.
r
%.
+
:l ..-
*:
.
.
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l I
._
-.
,
l
. . *
i OdESTION 3.07 (2.50) STATE the automatic action (s) that will occur when e'ach of the process radiation monitors exceed the condition listed, c. Refueling Area Ventilation Exhaust Duct High Radiation. (0.5) b. Reactor enclosure radiation monitors exceed the Hi-Hi setpoint.
( 1 '. 5 )
*
c. RHR Wea4 combined loop monitor Hi radiation.
Sers.c a (0.5)
'
W.44r'
.
.
.
-
.
i r
,
_J
-
'
;
)
l l
(***** 4JATEGORY 3 CONTINUED ON NEXT PAGE *****) .-- - _ - _ , . - - , - - .-_ . .-. . - . - --. . . - - ,
. . . -
l i QCESTION 3.08 (3.00) ) DESCRIBE the response of the FWCS that will occur if an iinstrument l technician places NR A level instrument in test when it is the ) celected input to FWCB. The instrument technician simulates O inches
~
level. Include the expected response of the Narrow Range level ! indications, RFP speed, feed flow, and any output signals to other i cystems (i.e. Recirc, RPS, etc.). Limit discussionEto the response ) until either a Rx scram or RFP trip occurs. Assume initialipower is l 75% power. : i l
.
I'
*
. C \
'
,
- ,
l
. . . g. :
l l I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l
- - - . .,, , . - . - . .- .. ,. . , ,. . , .. . . . . , . .
_
, . . . .
OlJESTION 3.09 (1.00) An ATWS has occurred and reactor power is 25% on APRM. Reactor water level is 20 inches and drywell pressure is .1. 2 psig. All scram valves opened and the SDV vent and drain valves shut.
- DESCRIBE why the scram cannot be reset in any position of the . reactor mode switch.
- s
-
, <
'
;
.
.
\
.
a
, . . . . .
t ,
,9
.. ..
*j
~
,.'p g. , e
; y 4 *1.s , '$
..
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
_ . - _ - . _ _ - .- ._. _ _ . _._ ,_ _ _ - ;_:-- : = ---. -
- . .
.
. . .
QUESTION 3.10 (1.50) DETERMINE if the following statements are TRUE or FALSE.
a. A reactor startup is-in progress using the A-2 withdraw cequenco. The operator can continuously withdraw a Group 3 rod from position 04 to 08. (0.5) b. RSCS will NOT allow rods to be moved in a manner that would cause the RWM to cause a rod block. (0.5) c. The RWM will NOT allow the operator to continue insertion of control rods with 3 insert errors present. Power is below the LPSP and no withdraw errors are present. (0.5)
.
%
.
'4
:
.
y< s
,
*
..
+ i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
.- - - . ,_-- - ,_ - . . . .- , , . . - . . . . - - . - . . . - - . .. . . -. , .-.,
_ _ _ _ _ _ _ _ _ - c
. . .
QUESTION 3.11 (1.50)- For each mode of operation of the RCIC speed control listed below briefly discuss how speed of the turbine is controlled.
- o. During system startup b. During automatic flow control
.- ,
- c. During manual, flow control
*
.
.
'.
l l l
'
- . 1
'
r. . l
,
. ,y- -
-
: n u ,
r 5: 4
,
'
. ; l
.c \
. . . - .,
.
(***** END OF CATEGORY 3 *****)
- _ _ _ _ _
.
6:__EBgCgQgggg_ _U9Bd6Li_0ENQBd@61_gdgB@ENGY Pega 34 ONQ_B8Qlg6Q@lC66_GQNIBQL
,
QUESTION 4.01 (2.50) a. Per "Shift Operations (A-7)" procedure LIST the four statements of guidance concerning when a plant shut down or scram shall be initiated by a Senior Licensed or Licensed operator. (2.0) b. Per 10CFR55 "Operator Licenses" who may manipulate the controls of the reactor other than Licensed Operators and Senior Operators.
STATE what condition (s) must be met.to allow this person to manipulate the controls. (0.5) I i I l i
,
I
1
,1
'
.
' ,
.~
'
Q ?
+ .
.
(***** CATEGOR( 4 CONTINUED ON NEXT PAGE *****)
.. -. -
. - _ , , . - - .: '-
- _--- . ____ - __ _____ __ _ _ . ..
.,
. . .
QUESTION 4.02 (2.50) Answer the f ollowing concerning Special Event Procedure (SE-1) "Remote Shutdown."
c. LIST three areas of the plant in which a fire.could necessitate the use of the Remote Shutdown procedure. (1.0)
,,.
b. The SRO determines that the Control Room must be abandoned. WHAT immediate actions are required prior to exiting the C.R.? . (1.5)
. '
l l
\
*
.
e l l
,
l
l l
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) - . . . .
. . . .
QUESTION 4.03 (2.50) The operator has observed indications of a stuck open relief valve during power operation. Answer the following questions in accordance with "Inadvertent Opening of a Relief Valve (OT-114)."
o. The operator is reqssired to place ______________ (one/both) loops of suppression rool cooling in service. (O.S) b. If the suppression pool temperature reaches ____________ F then the operator shall place the. mode switch in "Shutdown." (0.5) l c. In addition to the Operational Transient procedure for a Stuck Open Relief Valve, the operator shall enter which procedure at 95 F suppression pool temperature. (0.E)
d. If the stuck open relief valve cannot be shut within _______________ the operator shall place the mode switch in
"Shutdown". (0.5)
e. The operator is directed in the followup steps to "Reduce turbine inlet pressure to 900 psig" to attempt to reseat the valve. WHY does the instruction specify "Turbine inlet pressure" instead of l
"Reactor Pressure"7. (0.5)
l l
\
l
!
I i
,- -
1
*
\
.
'
.
-
,
as .
;
-
, ,
- - .
,s j W
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
.
p -- - , . .- -, , ,a. - - , . - - - - - , -,
. . .
QUESTION 4.04 (3.00) For each set of conditions below LIST which Trip Procedure (s) would be entered.
a. A loss of drywell cooling occurs. . Operators vent the drywell to maintain pressure < 1.2 psig. Drywell temperature is 150 F.
(0.5) b..A reactor scram occurs due to a turbine trip from 45% power.
Reactor level decreases to -10 inches following the scram but is automatically recovered by feedwater. (0.5) c. A MSIV isolation occurs due to improper testing by instrumentation ' technicians. The reactor scrams due to the isolation. (0.5) d. During a reactor shutdown the operator places the mode switch in startup at 20% power. Reactor power decreases to 2% due to PARTIAL , insertion of control rods. (0.5) e. A small leak in the drywell causes drywell pressure to increase to 3.2 psig. (0.5) f. A failure of the EHC system results in a pressure increase which causes high pressure scram. (0,5)
+
s., d
,
*J
"
t
i , %
.^
-
s (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) _ _
._ - - _ _ __
I l
.
,
.
...
!
!
l ObESTION 4.05 (2.50) i
!
Answer the following.concerning "Control of Plant Equipment (A-41)." I a. Who's responsibility is it to determine if independent verification of blocking is required? (0.5) b. What two classes of equipment are required to have independent verification performed when removed from service or restored? (1.0) c. A valve is determined to require independent verification because it is in one of the two classes listed in part b. If the valve is located in a high radiation area is an independent verification required? (0.5) d. The Shift Supervision released a surveillance for testing at 0900 June 6, 1988. How long is permission granted to perform the surveillance ? (0.5) .b
>
s
'
n' s
'
,,
1 o . ,
,
<
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
__ - _ . _ _ , , _ _ _ _ _ _ _ . ___ _ . , _ . _ _ _ - - _
* . . . .
t 00ESTION 4.06 (2.00) Discuss the reasons for each of the following cautions concerning operation of the Feedwater system.
c.-The A RFP should be placed in service first. (0.5) b. Maintaining.the RFPT speed less than 2000 rpm for c.t least one hour
'
following turbine roll. ,
,
, (0.5)
c. Avoid excessive throttling of the RFP Discharge valve that'causes ^ the MGU controller to automatically raise RFP speed (to maintain RPV water level). (Two reasons required) (1.0)
*
.
.
.
.
. , .
, C
, ..
.
..t , 'g. s
'
w? -n.
.: '
'
.
u.- e, r c. v
' '
* '
,, .1 . . ?:s , . ;', s
*
y <,
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
. . . .
QUESTION 4.07 (2.50) Answer the following per "Shutdown Cooling Operation (S51.B.B)" procedure.
a. Placing the RHR system in Shutdown Cooling requires that the operator shut and tag the minimum flow valve. LIST the two purposes f or perf orming this action. (1.0) b. When operating RHR in the shutdown cooling mode the minimum pump flow that is allowed by the procedure is 1500'gpm. EXPLAIN why flow must be maintained greater than this value. (0.5) c. EXPLAIN why flow through.a RHR heat exchanger is limited to less than 11000 gpm. (0.5) d. EXPLAIN why reactor water l evel must be maintained above 60 inches as read on the Shutdown range indicator (LI-42-R605) or 78 inches on the Upset range recorder (LR-42-R608). (Do not explain why the values are different.) (0.5)
'
j.
.
. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) . . - - . -. . . - - - - . . . _ -
' '
! , , o QUESTION 4.08 (2.50) You are directed to operate a valve in a high radiat' ion area.
a. (TRUE or FALSE) The expected dose rate in the area will be >100 mr/hr. . (0.5) 6. In order to enter the area the operator is required to have one of three items. LIST.these three items. '(1.5) c. A RWP _________________ (WOULD/WOULD NOT) be required.for access to the area. - ( 0. 5 )
. ~. - :
<
,
'
.
Q.
', pt y (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- _ . _. . - _ = . . ._. -
_ _ _ - - - _ - .
-
. , . , - - . . . . - - . . .- .... _ . . - . -. . ..
. . . .
QUESTION 4.09 (2.00) Answer the following in accordance with "Procedure for Temporary Changes to Approved Procedures (A-3)."
o. Maintenance has been completed on a valve in the RHR system. The SRO on shift has determined that only part of the RHR valve operability surveillance must be performed in order to determine if the valve is operable. IS a Temporary Procedure Change required
.per "Procedure for Temporary Changes to Approved Procedures (A-3)"
in order to perform the procedure? (0.5) 6. An operator who is reviewing a recently revised procedure prior to performance determines that a step necessary to complete the procedure has been deleted. WHAT requirements must be met to implement a temporary change before the procedure can be performed? (1.5)
,
~
,
-
.,-
, ; y, .,
:ll. '
.
i i. m ,s;
,
.* ,
'; 't
'~
.
"
; g ), ,
-
,
,< ' , y,
'
* *
i c =
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
..- . . - - - - - . . - - . . , - - - - _ , , _ . .- --
- -. . . . , , . . , . - _ ,
_ _ _ _ _ _ _
- . . . .
QUESTION 4.10 (3.25) The operator observes confirming indications that a loss of MCC D114-R-G has occurred.
c. WHAT event procedure would the operator 'efer r to-for actions". (.25) b. While the operator is performing the actions for the loss of MCC
,D114-R-G a loss of off-site power occurs.
< 1. LIST 2-plant components that.the operator is required to verify as starting. (Do not list redundant components in separate divisions) (1,0)
.
2. WHAT 2 system lineups must be performed by the operator in order to supply cooling to drywell coolers. (1.0) c. Subsequent to the loss of off-site power a loss of all AC occurs.
- LIST the 2 actions that the operator is required to verif y. (1.0; I l l l l l l l
'
l
. I
. ". t 74
.
.
j
..
-
*- s ,, , /;3
, . s l
+
.
'
l (***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
..
_ _ _ _ _ _ _ _ _ _ _ . -
. . .
1&__EBINGIELES_QE_NQ9LE88_EQWEB_EL8NI_QEEBBI19N2 Pega 44 IEEBbgDYU601GE2_BE01_IB8NEEEB_8ND_ELUID_ELQW
.
ANSWER 1.01 (2.50)- a. 'Falso b. False c. True d. True e. True C5 e 0.5 ea.] (2.5)
.
REFERENCE LGSt LOT-OB60, PP. 4 & 5 Lesson Objective 5. LOT 0870 PP. 4 & 5, Lesson Objectives 3, 4&5 KAI 3.2 292OO1K102 ..(KA's) ANSWER 1.02 (2.00) a. SDM Increases b. SDM Increases c. SDM Increases d. SDM norrer :: Iscecag,9 g4 4 0.5 ea.] (2.0) s AEFERENCE LGS: LOT-0950, PP. 6 & 11 Lesson Objective 6. -
o. 1. Period indication becomes lon'ger.
2. Indicated power on SRMs/IRMs is leveling off (due to power
'
overshoot). [2 0 0.5 ea.0 (1.0)
>
1 u b. Interval 2 From P=Poe(t/T). (In interva1J2.the peri'odthas' lengthened from 80 seconds. The other, intervals.have BO.second peri ods. ) (1.0)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1 I
!
, , ,
1,.__ER I N C I P L E S_QE_NQQLg68_EQWEB _E(@NI_QEEB@IlgN 2 Pcgo 46 IUEBdQ9YNedlGSi_UEBI_IB8NSF.EB_8ND_EbulD_ELQW
.
REFERENCE LGS: LOT-1430 Learning Objective 4.
LGS: Normal Plant Startup GP-2 Appendix I PP. 4 & 5. LOT-1430 P. 5.
KAI 3.6 3.8 . 292OOOK113 292OO8K112 ..(KA's)
,-
ANSWER 1.06 (2.00) - a. Increase b. Decrease.
c. Remains the same.
d. Decrease C4 G O.5 ea.3 (2.0) REFERENCE LGS: LOT-1490 PP. 5-12, Lesson Objectives 4 & 5.
KAI 2.5 292OO5K109 ..(KA's)
.
ANSWER 1.07 (1.00) 3 (1.0) REFERENCE LGS: LOT-1510 PP. 7-9, Lesson Objective 6. GE BWR Acdemic Series Chapter 6, PP. 6-10a to 6-12a.
KAI 2.8 '.
292OO6K108 ..(KA's) -
#
_
'
s p
*
:
,. ,
l
. .
l l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1
!
l l
- -
. . ,
li__EBINGIELES_9E_UYGLE68_EQWEB_ELOUI_QEEB8IlgN 1 Pcgs 47 ISEBdQ9YuedIGei_SEGI_ISBNSEEB_0U9_ELU19_E69W
.
.
ANSWER 1.08 (3.00) c. Doppler will mitigate the transient [0.53 because of the increase in fuel temperature. [0.53 (1.0) b. Moderator Temperature Coefficient will increase the severity, CO.53 due to increase in core inlet subcooling. CO.53 .* (1.0) c. Void coefficient will increase the severity of the transient CO.53 because of the increase in core inlet subcooling. CO.53 (1.0) REFERENCE LGS: FSAR FP. 15.1-1 thru 15.1-4, Figure 15.1-2 KAI 3.3 3.2 3.4 295014K206 295014K204 295014K203 ..(KA's) ANSWER 1.09 (2.00) a. False b. False c. True
.
d. Falso C4 0 .5'each] (2.0) i
, i REFERENCE l LGS: LOT-13OO PP. 1-6. ,
KAI 2.7 l t 293OO7K111 ..(KA's)
~
j
..
L
*
\ f
,] *
I
'
'e s
!
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
.- - - _
. _._i
, . .
11__EBING1ELEE_9E_NYGLE86_EQWEB_ELONI_OEEBOIlgN2 Pegn 48 IEEBdQDYN8dlGEi_SEGI_IB8NEEEB_6ND_ELUIR_E69W
.
ANSWER 1.10 (2.50)
a. Feedwater temperature - Feedwater flow RPV pressure RPV water level C4 8 0.25 ea.] (1.0) b. 1. High flow, High power (0.5) 2. High flow, High power CO.503, due to the increased -inlett subcooling from the increased feodwater flow. CO.503 (1.0) REFERENCE LGS: LOT-1290 PP.8-9 KAI 3.9 3.3 29tOO4K106 202OO1K402 ..(KA's) ANSWER 1.11 (3.00) a. The assembly power that would cause OTB at some point in the
*
assembly. o.se < F + bw ndit i 4. s. k 9bl y w'. fl 655sMfg (1.0) b. 1. Decreases 2. Decreases 3. Decreases
'
4. Decreases [4 9 0.5 ea.] (2.0) REFERENCE
-' '
LGS: 1370 PP. 8 Learning Objective 1, 2. ' -
- " '
KAI 2.9 2.7 2.7 2.6 '
,
-
293OO9K126 293OO9K125 293OO9K124' 293OO9K122 ,
.f(KA*m)
,
, , .s
,.
. . . , ,
,
; , ,
.
,
' ~ '
vh . " i ,,, ;; ' k -
-
>
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
. ' . ,
11__EBINGIELES_9E_NWGLE8B_ERWEB_EL8NI_QEEB@IlgN2 'PagGL49 IBEBd9DXN601GS2_Ug@I_IB9NSEEB_8ND_ELUID_EL9W
. . ANSWER 1.12 (2.00)
a. MAPRAT is the ratio of APLHGR TO APLHGR Limit OR the ratio of , MAPLHGR(act)'to ,M6PLHGR(LCO) (Either answer acceptable for full credit.) (1.0) b. The clad temperature can exceed 2200 degrees F. during a DBA LOCA (1.0)
'
REFERENCE s LGS: LOT 1410 P.4 Lesson Objectives 3 & 4.
KAI 2.9 2.8 293009K111 293009K112 ..(KA's) i r
.
s
,
# .
.
!
!
(***** END OF CATEGORY 1 *****)
_ - . . - . . .. . . . - . . - - . - .
I
. +
21-_ELONI_DESISN_INGLWDING_39EEIY_8NQ_Ed5B@gNQX Pcqa 50 l l , SYSIEUS L i
.
ANSWER 2.01 (3.00) o. 2200 rpm (0.5) Minimize water hammer in exhaust line (0.5) and prx er operation of hydraulically operated valves which the attached lube oil pump supplies (0,5) b. 600_gpm (0.5) c. Injection valve to CS (FOO6) is not fully shut. (0.5)
*
Initiation signal. (0,5) h684P t o _d tw (svel Or M8 ' d ' /,# [,,. ,, ,, [ ,
.
4, y.,j g il p r e o ., i REFERENCE LGS: LOT-340 P. 6, 14, AND 15.
Lesson Objectives 6.
KA (3.2) (3.4) (3.2) 206000K418 206000K411 206000K407 ..(KA's) 2. . '[ ANSWER 2.02 E.^9) a. Would not b. Would not.
c_ u2ui g , t -a
, g , [ ,, j. Q d. Would.
o. Would not. , f .. Would.
' z.. C .
[p 8 .5 each] M) , -
J l
.
. .
.
,
.
o -
,
~
l
< (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
!
_ _
- . _
. _ . , - , ,.
Zi__ELONI_DEE19N_INGLUDINQ_E9EEIY_6ND_EdEB9ENRY Pcom 51-EYSIEUS
.
REFERENCE LGS: LOT-660 P. 12, 13.
Learning Objectives 5 and 6.
KA (3.3) (3.2) (3.1)
' .262OO1K403 264000K405 264000K408 ..(KA's)
C
,
i.' ' ANSWER 2.03 (2.00) Jacket Coolant Temp High Jacket Coolant Low Press Generator Ground Neutral Overcurrent Lube' Oil Low Pressure - Lube Oil High Temperature Fire Protection Actuation Engine Overspeed Diesel Generator Differential Overcurrent (2.0) (6 0 .33 each)
REFERENCE LGS: LOT-670,P 11 and 12 Learning Objective 5.
KA (3.5) (4.0)
,
264000K402 264000K401 ..(KA's)
, ;
.
..
-
..
-u . l
.
. . i 1 ,
.p
*
*
'.c %
l
, -
i l l i l (***** CATEGORY 2 CONTINOED ON NEXT PAGE *****) _ _ _
-
_ _ _ _ _ - _ _ - .
.. . . .
2t__ELONI_QESl@$_INQLUQ1NQ_SdEEIY_68Q_ENEB@gNQX Pega 52 EYSIEUS
.
ANSWER 2.04 (3.25) a. No APRM downscale (0.5) ll$ y& second time delay (0.5) b. -38" (0.25) c. All pumps start (0.5) . * * All squib valves open (0.5) er 6.s t.4 p t tA W s W '.i k I ' O( ' 'Y( * RWCU isolates (0.5) 10 minute reset timer-actuates. (0.5) De c.r e.e.fi q mk l e v e.\ best essbR Reo.ekth r= ?**J 't REFERENCE LGS: LOT-0310 p. 16 and 1/. Learning objectives 9, 10, and 11.
KA (4.2) 211000A30B ..(KA's) ANSWER 2.05 (2.50) a. 1. 775 gpm 2. M M psig 3. 330 psig (accept 310 to 350) 4. 455 psig (2.0) (.5 for each setpoint) 6. 3175 gpm (.25) at 250 psig (.25)
.
REFERENCE 4:, LGS: L OT-0350 p. 6, 8, 10, 11 Learning Objectives 5, 7, 9, 11. 1 KA (3.8) (3.6) (3.8) (3.5) (3.7) a
,
.
-
209001A304 209001A303 209001A302 209001A301 209001K408
..(KA's) . (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
( _ ..
__ _____ _ _ _
.
.. .
.-
. .
Bi__ELONI_QEgl@N_IUCLUDINQ_g8EgIY_6ND_EdEBQENCY Pcgm 53 SYSIEMS
.
ANSWER 2.06 (3.00) i l
'
' e. 12.5 inches (.25) I
b. Inboard Shutdown Cooling Suction Valve (FOO9) (.25) Outboa-d Shutdown Cooling Suction Valve ' (FOO8) (.25) . Shutdown Cooling Injection Valves (FO15 A,B) (.25)
-'
Head Spray Velves (F023, F022) (.25) .
(Accept either;the valve names or numbers)
l c. -129 inches (,03) , d. The suction of <HR pumps will nave to be aligned to the l suppressior, pooi t.b).
I e. RHR HX bypass valve (F 048) would open (.5) and the L'PCI injection val v+2s (F017) would open (.5) REFERENCE LGS: LOT-0370 p. 10, 22, and 23 Learning Objective 6, 7, 8, 10.
KA (4.2) (4.4) (4.1) (3.8) (3.6) 205000A205 205000K403 203000A300 203OOOA216 203OOOK401
..(KA's) .
ANSWER 2.07 (2.00) a. Low Process flow (0.5) High Recombiner Outlet tamp (0.5) 6. Isolates steam supply to first stage air ejectors. C '. 5 ) Isolates suction piping from condenser (.47 -
<
t' e
*'
ei
, (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) . . . . ... - _ . - -- _ -
, _ ,_ _- _ _ . __ . ' . s :3i__EL8NI2REH10N.1NGLup1N0_5BEFIY_8ND_EDEB@gNGy -Pago,54 SYSIEde .-
REFERENCE t LGS: LOT-0510 p. 9, 10.
Laarning objectivo 2h. 4. , KA (3.1) _ m ' 271000K408 ..(KA's) 1
'
,
.s 2.0
"
'
'l ANSWEh 2.08 Siihr60)
a. Outboard isolation valve rhuts. (.5) RWCU. pumps trip (.5) , Demin hold pumps start (.5) 6. No (.5)
,
c_ m' '.5M d e! dad REFERENCE
,
LGS: LOT-0460 p. 8. 11, 12.
Learning objective 9.
KA (3.5) (3.1) (3.3) (3.2) 204000A201 295018A101 295018K301 295018K101 . . (K A 's ).-
+
n ; , ,,. _ l a , :. M \
.. . ~
.;.
- @ A;l.. l
= ,3.
1:
'a: , g.=_ -
,
i e s
, ,
,,o
, I ti. n,
,t ?8 - f)( 11 .f5
'~~, , l* -
; g -j g.$.k4 p. s . jf. j)/f
'
b '
.-
M . s e ,4;.,
, < -- n i; , c y .;s _&b,L-l l
I o 1 (u**** CATEGORY 2 CONTINUED ON NEXT PAGE *****) j i l _ _ . - . n ?
*
,
l
,
A. ELANT DESIGN INC8,UDINg_g8EgIy_8ND_EdgBQgNCY Pcgm 55 .l SYSIEdS
.
ANSWER 2.09 (3.00) _ c. Low suction pressure Low bearing oil pressure ~ Inactive thrust bearing wear high ,
' . Active thrust bearing wear high .RFPT High vibration r O p u.u. ^ . ' /.U Y "NO.
0) each) RFP Discharge Check valve will shut (.2[)
'
b.
HP and LP Stopvalves (.2jd and control valves (.2 will shut HP and LP below neat drains will open ( M (.o.1) g) LP Steam supply valve upstream drain valve opens ( P26) lo T.)
Min Flow recirc valve will shut if open (CAF if this valve would be open at the given power level) M % M.4 g d M hut
.f Y J4 % & 95 REFERENCE (5 e o.z. e,.Q LGS: LOT-0540 p. 13, 22.
Learning Ob. active 4.
KA (3.4) 259001A310 ..(KA*s)
<
, .
M T- I
, .
.
,
s.. ,:r -
'
s i l l l
(***** END OF CATEGORY 2 *****)
l
+ t , , . . I
_ _ , , , , _ . - . _
. .
L.__INSIBUdENIS_6ND_CONIBOLS Pcge 56
.
2.oo ANSWER 3.01 (JArtr> (,o . '15) e. 1A recirc pump will speed up to 45% (fv757.
W 1B will be unaffected
;
M " Lo.1s) 19 (, o p** O o) 6.-Both recirc pumps will decrease to J"f/.. (,1.<<r1
*
*,*g Mc %r.,en JM wkl a. W- %
'
& M &c. Mt s circ pump is >90% the pump will
/. When the discharge valve1A forwill 1B be unaffected (0.25).
speed up to.75% (0.75)~ *..
. Qda.fid REFERENCE LGS: LOT-OO40 figure 2, p. 6, 8.
Learning objectives 2, 3, 4 and 12.
KA (3.0) (3.5)
202OO2K604 202OO2K402 ..(KA's)
!
ANSWER 3.02 (2.50) I l I a. higher than actual b. same as actual oe h'. cg k + / NA o Ma ( c. higher than actual d. higher than actual s e. higher than actual (2.5) (0.5 each) REFERENCE- j,' o LGS: LOT-OO50 p. 28 and 29 Learning objective 7 g KA (3.0) (3.4) (3.2) (3.2) . j
; , 4 ; jy t ,
'
216000A211 216000A208 21oOOOA207 216000A203 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) * . . , . . - . + - . - ~ -
__ , . . . . _ _
_ _ _ _ _ _ - -
' . .
Oz__INSIBydENIg_6ND_GQUISQLS Pcgn 57
.
l l ANSWER 3.03 (2.75) I a. High Steam Line flow 140% of rated in a steam line Low reactor vessel level -129 inches MSL High temperature Tunnel Temp 192 or Turbine enclosure temp 165 l MSL High Radiation 3 x NFPBG l Anal .
' (2.25) (.25 f or each signal and . 2[g)for each setpoint)
6. False (0.5) l l REFER.~.NCE l LGS ' LOT-0120 p. 1B ' Learning Objective 9.
KA (4.0) (3.8)
.
239001K401 230001K127 ..<KA's) ANSWER 3.04 '.00)
- .
a. 3 (.5) 6. No action occurs (.5)
'. c. 3 (.5) (.C)
d. 2 (A) M L.Mrt l o. 3 (.5 ) f. 2 (.25) and 3 (.25)
.
(3.0) (6 G .5 each) >
.
.
t l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ - - - _ _ _ _ _ _ - _ - _
_ - _ _ _ _ _ _ _ _ _
' . ,
IL__INSIBudhNIS_8NQ_ggNIBQLS Paga 58.
. REFERENCE LGS: LOT-0250 p. 10. Learning Objective 10.
LOT-0270 p. 11, 12, and 13. Learning Objective 7. ' KA (3.7) (3.7) (4.1)
.
215005K401
'
215005K402 215003K401 ..(KA's)
,
e % ANSWER 3.05 (2.50) - 57.5 _ o. 4!iKt- (O.5) , b. 105 (0.5) c. no (0.5) d. no (0.5) , p. The valve has opened (0.25) but is presently closed (0.25).
REFERENCE LGS: LOT-0330 p. 9, 12, 13. Figure 6.
Learning Objectives 2, 5, and 6.
KA (3.7) (3.8) (3.8) (3.7) 218000A303 218000K501 218000K402 218000K401 ..(KA's)
.
-
, .. ,
-~
, .
,
,
2 er nn ,
'
I
-
3 - l
,
<,
-l
.r .; a J l
*
j' 1.: , <
- r + p . ' ,,_ _,
_
.
, ,%-
.
j (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l _ _ - _ _ - _ - _ _ _ _ . _ _ _ _ _ _ _ _ - . .
'
'
. .
at__INSIBudENIg_6NQ_GQUIBQL@ Pcgs 59
.
ANSWER 3.06 (3.00) - 2[ Output from pressure summer A increases to Ad, psi (.5). ,The HVG passes this signal to the Pressure / Flow converter (.5). The Pressure / Flow converter output will increase to 83% (.5). The load limit will limit the signal to the governor-valves to 65% (.5).
Total flow is less than the maximum combined limiter (.5) .so the bypass valves will opentopass_17) ow (.5). , CREFERNCE LGS: LOT-0590 p. 6, 7, 8, 13, 15, 16.
Learning objectives 3, 8.
KA (3.7) (4.1) (3.7)
REFERENCE i 241000K308 241000K306 241000K305 ..(KA's) l
ANSMER 3.07 (2.50)
.
a. Refuel is lates. (. ) 6 (a ~f d a floor
+ets supply i 'f any a,ligk sd+ exhaust o +Le r s. ( $(ce r- . ?.9)
b. Supply and exhaust ventilation isolates. (.5) Standby gas treatment starts. (.5) Reactor enclosure recirculation starts (.5) c. Trips RHR service water pumps. ,
(.5)
REFERENCE LGS: LOT-0720 p. 10, 11, 18.
Learning Objective 2.
LGS: LOT-0180 p. 23 Learning Objective 2. - KA (3.6) (3.2) (3.6) (3.6) (3.7), !
!
272OOOK402 272OOOK109 272OOOK108 272OOOK106 , 272OOOK101
..(KA's) . .
-
y t
,. (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- . * . .
Tt__INSIBydENIS_GNQ_CQNIBgLg Fcga 60
.
ANSWER 3.08 (3.00)
^
NR A (.375) and t'he NR Recorder (.375) will-indicate O inches.
.All remaining level instruments will track upward as feedflow' increases (.375).
The RFP will accelerate to the high speed setting ( .' 37 5 ) . E Feedflow will increase (.375). ' Recirc pumps wil'1 runback to 28% (.375). . The RFP (.375) and Main turbine trip on high level (0.375). - - REFERENCE LGS: LOT-0550 p. 18, 19.
Learning Objective 6, 7c.
KA (3.2) (3.8) (3.7) (3.4) (3.5) (3.6) 259002A203 259002K605 259002K307 259002K302 259002K301 259002K115 ..(KA's) ' ANSWER 3.09 (1.00) The SDV scram bypass will only operate in Shutdown or Refuel. (.5) The scram cancot be reset in startup, refuel or shutdoNn because of the high APRM flux. (.5) REFERENCE LGS: LOT-03OO p. 8, 9, 14.
Learning objective 7, 8. 7 , KA (3.9) (3.8)
,
a,
.c ,a y ,
I- * rf ,
, r ;
-
212OOOK219 212OOOA404 ..(KA's) :~ ~'
, ,
,'. ::4 n . :p l ,.
l
- ~ -* .
n. [ ? i >
; hQ. nE
,
.
)i:Q ..) ' % y. . . l
~ ;'
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l
!
. _ . , . . . . . - , ,_. _ , . _ _-
. ._ *
p . , Is__INSIBWMENIS_8ND_QQNIBg(S Pcgo 61
.
ANSWER 3.10 (1.50) a. False (.5) (Continuous Withdraw Inhibit;will, prevent continuous withdrawl.) . b. False (.5) (RWM divides RSCS group into more than one group)
,
c.'True (.5) s
.
REFERENCE LGS: LOT-0100 p. 10, 16.
Learning Objective 6.
LGS: LOT-OO90 p. 12, 16.
Learning Objective 3.
KA (3.1) (3.3) (3.3) (3.3) (3.4) (3.5) (3.3) (3.2) 201006A301 201006K403 201006K402 201006K401 201006K106 201006K406 201004K403 201004K402 ..(KA's) ANSWER 3.11 (1.50) a. a ramp generator controls turbine speed. (0.5) 6. Flow controller produces an electrical signal proportional to the difference between desired flow and actual flow. (0.5) c. Speed is adjusted by the operator. (0.5)
.
l i REFERENCE LGS: LOT-0380 p. 12 , Learning Objective 10. I KA (3.7) (3.6) , (3.,) 6 (3.7) . .. , I
' ' & l '
i' 217000A401 217000A304 217000A302 217000A105 ..(KA's) i
!
l ,
(***** END OF CATEGORY 3 *****) l l
.
, _
. . . - - - .. . - .
* . .
4. PRQgEDQBEQ_ _NgBd6L 3_8pNgBd@L3_EMEBGENGy Pcga 62 BNQ_S691Q69 Gig @E_ggNIBQL
,
ANSWER 4.01 (2.50) a. observations of plant conditions and equipment'indicater, a safety hazard. (0.5) . doubt'as to whether safe conditions exist (0.5) e when RPS parameters have been exceeded without a scram' (0.5) approved procedures so' direct (0.5) b. License trainees (0.25) under direct supervision of a licensed operator. (0.25) REFERENCE 10CFR55.13 LGS: Shift Operations (A-7) p. 11, 15 and 16.
LGS: LOT-1570 Learning Objective 2 and 3.
NRC Information Notice No. 88-20.
KA (3.7) 201001 GOO 1 ..(KA's) ANSWER 4.02 (2.50) a. Main Control Room (.33) Aux. Equipment Room (.33) Cable Spreading Room (.33) b. Scram the reactor (.5)
'
Trip the main terbine (.5) - Close the MSIVs (.5)
. .
*
*
,.
*.
~
,
a
* * '
a
~,
@_ .:, ..-(-:
.r - s .c ,
i . l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) J l
,
l
. , _ - . -
- - . _ _.%. '7'
,,
_ . . _ _ _ _ - *
. ,
sc__EB9CEQUSES_:_NQBd862_8BNQBd861_EMEBGENQY .Pcgo 63 BNQ_B8g1Q6QQlg86_QQNIBg6
.
REFERENCE LGS: LOT-1563 p. 3 Learning objectives 1, 2.
LGS: Remote Shutdown (SE-1) p. 1 and 2.
KA (3.8) (4.1)
.
295016G011 295016G010 ..(KA's)
.
ANSWER 4.03 '?.50) a. both (.5) b. 1f0 (.5) . c. T-102 (Primary Containment Control) (.5) (Accept either procedure name or number) d. 2 minutes (.5) e. Reactor pressure CANNOT be reduced to 900 psig without receiving a GP I isolation (.5)
'
REFERENCE LGS: LOT-1540 p. 30, 32, and 33. Learning objectives 1 and 2.
LGS: Inadvertent Opening of a Relief Valve (OT-114) p. 1.
LGS: Inadvertent Opening of a Relief Valve (OT-114) Bases p. 3. i
.
KA (3.0) (3.8) (4.1) (4.2) (3.3) (3.G) 239001K401 241000K504 239 ' + . > 15 239002G014 239002 GOO 1 239002A201 ..(KA's) - s . n-
,
,
.>
'
- , ;
.s .
-
- s
r- > .
, ,
;
. 4 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) . . - . _ .__ _ , _ . . . . . . , . . . _ , .. . , , . . _ .
i
- !
-
. .
9t__EBQGEDUBES_:_NQBd@6,_@QUQBd@61_gbEBQENQY Pega 64
,
BNQ_B@Qlg6QQlg@6_QQNIBQL
.
l ANSWER 4.04 (3.00)
-
* !
c. T-102 (Primary Containment Control) (.5) ; '
~
b. T-100 (Scram)- "(.5) f ;
~ .. .
c. T-101 (RPV Control) (.5) C - d. T-100 (Scram), (.5) a o. T-101 (RPV Control) (.25) and T-102 (Primary Containment Control)
(.25)
f. T-101 (RPV Control) (0.5)
(Accept procedure name or number)
REFERENCE LGS: LOT-1560 p. 7 Learning Objective 3 LGS: RPV Control (T-101) LGS: Primary Containment Control (T-102) LGS: Scram (T-100) KA (4.3) (4.2) (4.2) (4.3) (4.2) (4.4) (4.3) 295037G011 295031G011 295030G011 295029G011 295028G011 295006G011 295024G011 ..(KA's)
.
2.0 ' ANSWER ~ 4.05 (4 ,6 64
, < ,
.,.
-
, 's
*
a. Chief Operator (.5) i b. Safety related (.5) or tech. spec related (.5) -i a
'
!
se*Pt 4 thhd e . P m a.) .Ar No. ( . 5) .
(dk; "* b't '~ e,
* @
% s; (;<
l i
' = . , 'r: T ,
d. Until 0800 June 7, 1988. (.5) ,
-
,
,, q;?T l
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) ;
)
l
. - . . ~ , - , - . . - , - . . . .- .a..
---
O - _.
- '
. .
.d=__EB9GEDUBEE_:_N9Bdebi_8aN9Bd8L2_EMEBGEN9Y Page 65 BND_B89196901G86_GQNIBQL
.
' REFERENCE LGS Procedure f or Control of Plant Equipment _(A-41) p. 6, 8, 12.
LGS: LOT-1570 p. 44. Learning Objective 3.
-- KA (3.9) ,
, ,
*
29400iK102 . ..(KA's)
' <
,
,
,
ANSWER 4.06 (2.00)
, -
a. Makes it possible to make up to the RPV using the Startup Bypass valve. (.5) b. Allows for sufficient RFPT pre-warming. (.5) c. Can cause unstable flow (.5) and excess valve wear (.5).
REFERENCE LGS: Placing a Standby RFP in Service (SO6.1.C) p. 3 LGS: Removing RFPs from service to a Standby Condition (SO6.2.C) KA (3.6) (3.9) (3.2) 259001G010 259001A402 259001A401 ..(KA's) i
.
l l ANSWER 4.07- (2.50) ) I
.
a. Ensur: 211 fica i n te th- "errel 'O. 5) rad prevent inadvertant draining of'the vessel %.o), , 4
~
b. Prevent pump overheating (0.5)' , . r-c. Prevent damage to the heat exchanger (0.5) a d. Ensure proper natural circulation-(0.5) . c
, . e ' < ,; .
I I
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
-
. .- t. ...: : =.:
-
;-. .. - ..~ .- . -.
.
- . ' , .
Sc__EB9GED9BES_:_N9BdeL2_8BN9Bd862_EdEBGENGY Pago 66 869_B89196991G66_G9 NIB 96
. .
REFERENCE LGS: Shutdown Cooling Operation (S51.8.b) p. 3 and 5.
KA (3.2) (3.6)
'
-
.-
205000K102 205000G010 ..(KA's) _
'
i- . ANSWER 4.08 (2.50)
,
'
a. True (0.5) - b. Radiation monitoring instrument continuously operating (0.5) An alarming dosimeter (0.5) A qualified HP technician (0.5) c. would (0.5) I REFERENCE LGS: Technical Specification 6.12 LGS: LOT-1760 p. 7, 13, and 15. ) Learning Objective 1, 3, and 4.
KA (3.3) l 294001K103 ..(KA's) ANSWER 4.09 (2.00) 2. No. (.5) s.
&g j
',*" / ^/
,. -, ~,,.) Q : 4 '.., .
b. 1. Determine:t the change does,not change lthe, intent'tof.the'- i procedure. .sg ,
,
'
t. r 1 2. Document the change ( pB) , 7.: ,2 , Have the, change approved by the. shift.. super.intendent t( . 51, ,. and an individual knowledgeable 'in >the', hreas Jaf f ecte4 by
~
i ,
. procedure.(.25) who is a member of PORC or previouslyy , .tc~ Jhei s
'
'- designated (.25). . ', c,
.
, , -
1;
,
,. .
.
v.g .
; . .
g_
/ r 4 : 1j _ : q. 4# ft ' (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) .__ . .
_
.
- . , , l l
4* __EBOGEQUBES_:_N9Bdebi_0BN9BdeLi EMEBQENQL Pago 67 l BNQ_B@Qlg6QQ1Q@L_QQUIBQ6 ;
, .
REFERENCE LGS: Procedure for Temporary Changes to Approved Procedures (A-3) p.
3.
LGS: Technical Sp ec i f i c at i ons ~6'.'B. 3.1. J'.T .;T ~. _1'_
'
.,
LGS: LOT-1570 p. 16.-Learning' objective'2.-- * 1 'I 4 --
','
KA-(4.2). &> c
. . . -
,,
.
-
,
,
294001A102 ..(KA's)
. m
'
f
.
ANSWER 4.10 (3.25) a. E-D114-R-G (.25) b. 1. Diesels (.5) ESW pumps (.5) 2. Line up ESW to RECW (.5) Lino up RECW to Drywell coolers (.5) c. Rx scram (.5) MSIV isolation (.5) REFERENCE LGS: LOT-1566 p. 2, 4, 6. Learning Objectives 2 and 3.. LGS: Loss of All AC Power (Station Blackout) (E-1) p. 1.
LGS: Loss of Off-Site Power (E-10/20) p. 1 and 2.
LGS: Loss of MCC D114-R-G.
KA (3.9) 295003G010 ..(KA's)
,, Y ;> J
.e a , ,-
-3 I
- . .'- - . ;
. :. .c -
,, s c4 O. : n - .,
a
'
.i ; ., 4 A >g M + - ' ;> ,
,
-
; :
,
. r . .. . .y
-
y w .>
'
y, , ,
'
,.: .} _
*
'x 1
<
'e a
l (***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
__ _ __ J
.
ATTA CH MEAM 2 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _ LIMERICK;1______________ REACTOR TYPE: _pWR-GE4_________________ DA'.'E ADMINISTERED: ,@@f96f9Z_,______________ EXAMINER: _NRg_REGlgN_,I____________ CANDIDATE: _____@_6_$_T@_6___________ IN@IRUCllgNS_IQ,,C9NQ1991El Uso separate paper for the answers. Write answers on one side only.
Steple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY _ _ V_ A_ L_ U_ E__ _ _ T O_ _T_A__ -L-_SC_ O_ R_ E_ _ _ __ V_ A_ L_ U_ E_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ C_ A_ T_ E_ G_ O_ R_ Y _ _ _ _ - _ _ _ _ _ _ _ _ _3Et99__ _29199 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND , THERMODYNAMICS l 24 50 2H.To j
'
, _ 2 5 : 2 1 _ _ _-E E z 9 2 ___________ _ _ _ _ _ _ _ . 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25 SD zg.so _2E222__ _EE 92 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _3E199__ _3E199 ___________ ________ B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199:99__ ___________ ________x Total s Final Grade All work done on thi s examination is my own. I have neither given nor received aid.
______________________________.____ Candidate's Signa.ture
e NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
.
During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penal ti es.
2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3. Une black ink or dark pencil gnly to facilitate legible reproouctions.
4. Print your name in the blank provided on the cover sheet of tne nx ami nat i on . S. Cill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least thtgg lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets fcce down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or ussumptions used to obtain an answer to mathematical problems whether indicated in the question or not. ' I 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. ]
16. If parts of the examination are not clear as to intent, ask questions of I the examiner only.
17. You must sign the statement on the cover sheet that indicates that the , work is your own and you have not received or been given assistance in l completing the examination. This must be done after the examination has ) been completed.
i l l l i I l
_. _ __
.. . - - _ - _.
. . - .- . . . . .. v 18.'When you complete your examination, you shall a. Assemble your examination as follows:
(1) Exam questi ons on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
5. Turn in your copy of the examination and all pages used to answer the ex ami nati on questions.
c. Turn in all scrap paper a r.d the balance of the paper that you did not use for answering the questions.
d. Leave the examinati on area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
.
-. r +n ,-e.- - ,, . ~ - --,,e,- ,, ,e-we w w w,. - .-c.- ,,,,.w,-we-,-r,m e,--e,,.,,m.- - , , n r-,. <w. v.. r -
51__IHEgBy_gE_ NUCLE @B_EgWEB_E(@NI_gEEB@llgN 3 _E(ylp@t_@N9 PAGE 2 ) IHEBMggyN@MIC@
- J l
QUESTION 5.01 (2.50)
!
For each of the f ollowing events, STATE which coefficient of rocctivity would act FIRST to change reactivity.
a. Control rod drop at 100% power (0.5) b. SRV opening at 100% power (0.5) l c. Loss of shutdown cooling while in Cold Shutdown (0.5) d. One recirc pump trips while at 50% power (O.5)
.
e. Loss of one feedwater heater (extraction steam isolated) (0.5) QUESTION 5.02 (2.25) l Following a normal increasa in power from 75% to 100% wi th recircul ation , flow, HOW will each of the following parameters change (INCREASE, DECREASE, or REMAIN THE SAME) and WHY? l a. The pressure difference between the reactor and the turbine steam chest (0.75) b. Condensate subcooling at the exit of the main condenser (0.75)
c. Feedwater temperature (at inlet to the reactor vessel) (0.75) , l QUESTION 5.03 (3.00) Rocctor power was decreased from 100% to 50%. c. Briefly EXPLAIN WHY the xenon concentration will i ncr ease following the manuever. (1.00) b. How will peripheral control rod worth be affected (INCREASE, DECREASE, or REMAIN THE SAME) during the xenon peak? Briefly EXPLAIN your answer. (1.50) c. Will the new (50% power) equilibrium xenon reactivity be MORE l THAN, LESS THAN, or EQUAL TO one half the 100% equilibrium value? (0.50)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) . . -. .- - ..
F __IHgggy_g[_NgCbE@g_[gyE5_E6@NI_gPgg@IlgN1 _{Ly]Qgt_@Np PAGE 3 IHERDODYNBDICS
.
QUESTION 5.04 (2.00) The reactor is operating at 75% rated power and the operator is withdrawing control rods.
WILL the withdrawal of a central control e od from notch 04 to notch 08 have a LARGER or SMALLER affect than withdrawal of the same rod from notch 36 to notch 40 on EACH of the f ollowing parameters? c. Overall core thermal power (0.50) b. Axial flux distribution (0.50) c. Radial flux distribution (0.50) I d. Local power surrounding the rod (0.50) QUESTION 5.05 (2.00) Using the Steam Tables or Mollier Diagram, calculate HOW LONG it will i take to cooldown from 1000 psig to O psig at the maximum allowable cooldown rate allowed per GP-3," Normal Plant Shutdown". (2.00) ; QUESTION 5.06 (2.25) Fol l owi ng an AUTO INITIATION of HPCI at a reactor pressure of1000 psig, reactor pressure decreases to 500 psi g.
HOW are each of the following parameters affected (INCREASES, DECREASES, REMAINS CONSTANT) by the change in reactor pressure? BRIEFLY-EXPLAIN your choice.
ASSUME the HPCI System is operating in automatic as designed.
a. HPCI flow to the reactor. (0.75) b. HPCI pump discharge head (assuming NPSH remains constant). (0.75) c. HPCI turbine RPM. (0.75) ,
(***** CATEGORY 05' CONTINUED ON NEXT PAGE *****) ,, - -- .
D __IHEQBy_gE_NyCLEgB_EgWgB_EL9NI_gEEB911gN 1 _ELgipS1 _9NQ PAGE- 4 IHE6DQQyN@DICS
.
QUESTION 5.07 (2.00) EXPLAIN HOW it is possible to produce an increase in power - as control rods are inserted into the core (Reverse Power Effect). Include in your answer under WHAT conditions this is possible. (2.00) QUESTION 5.08 (2.00) A reactor heat balance was performed (by hand) during your shift due l to the Process Computer being out of service.
STATE whether each of the following statements is TRUE or FALSE.
a. If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater f l ow rate, then the actual -
)
powe- is HIGHER than the currently calculated power. (0.50) b. If the reactor recirculation pump heat input used in the heat I balance calculation was OMITTED, then the actual power is HIGHER than the currently calculated power. (0.50) c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is HIGHER l than the currentl y cal cul ated power. 'O.50) I l d. If the RWCU return temperature used in the heat balance calculation was LOWER than the actual RWCU return temperature, then the actual power is HIGHER than the currently calculated i power. (0,50) ! I OUESTION 5.09 (2.00) A periodic core performance edit (P-1) has just been completed by the i Process Computer. After r evi ewi ng the output, you notice a MAPRAT value aqual to 1.002.
a. WHAT is the relationship between MAPRAT and MAPLHGR 7 (0.50) b. Have any thermal limits been exceeded? If so, WHICH one(s)? EXPLAIN your answer. (1.00) c. WHAT physical consequence could occur if the MAPRAT limit is exceeded? (0.50) i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-- _ -_ . - - _._ . , _ _ _ _ . _. ._.
_ ' Ez__IHEggY_QE.,NgCLE86_EgWEB_[L@NI_ GEE @@IlgNt_ELylQS 1_@NQ .PAGE 5 i IHE6DgpVN@DICS
~
i QUESTION 5.10 (3.00) Whsn the reactor is at full power, a spurious trip of all'feedwater pumps occurs. Using the attached FSAR Figure 15.2-9, EXPLAIN WHY the following parameters respond for the periods st ated bel ow.
c. WHY does the reactor water level decrease between O seconds (0.25) and 7' seconds? b. WHY does reactor water level continue to decrease following the scram at approximately 7 seconds? (2 reasons) (0.50) c. WHY does the reactor pressure decrease between O seconds and 7 seconds and WHY does the rate of pressure decrease INCREASE
'(0. 75)
after 7 seconds? d. WHY does reactor pressure increase f ollowing the MSIV closure at approximately 18 seconds and WHY does the increase stop at appr ox i mat el y 32 seconds? (0.75) e. WHY does the core (i nl et ) flow SLOWLY decrease between O seconds and 15 seconds and WHY does it decrease at a FASTER rate after (0.75) 15 seconds?
(
-l I
i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-
9 __IHEggy_gE_Ngg(E@B_EgWEE_E(@NI_gPE6@llgN 1 _E(gips _@Ng t PAGE 6 IHE6dggyN@ gigs . QUESTION 5.11 (2.00) c. Choose WHICH ONE of the f ollowing events is the most likely to produce water hammer. (1.00) 1. Core Spray pumps A and C are running in full flow test lineup, CS pump C is stopped 2. Core Spray pumps A and C are running in full flow test lineup, the test line orifice becomes blocked with debris 3. Core Spray pump A is running in full flow test lineup, Core Spray pump C is started 4 Core Spray pump A is started with the suction valve cl osed b. Choose WHICH ONE of the pairs listed below will complete the f ollowing statement. (1.00) With Core Spray pump A running in full flow test lineup, starting Core Spray pump C will _____________ CS pump A flow and ___________ CS pump A discharge pressure.
1. increase, increase 2. increase, decrease 3. decrease, increase 4. decrease, decrease
.
(***** END OF CATEGORY 05 *****)
_, -. - . , . . - _ . -
. ..,_.
be__P6@NI_gYgIEgg_gEgigyz_CQN]@ght_@yg_lNgl@gggNI@IlgN PAGE 7
-
i i QUESTION 6.01 (1.00) HPCI has .aut omat i cal l'y initiated, taking a suction from the suppression pool and injecting to the reactor vessel. The suction screen becomes pertially clogged, causing the pump suction to reach IB" Hg vacuum.
Aacuming no operator action and the i nitiation signal, present, the racponse of HPCI will be to: (CHOOSEONE) c. Continue to inject b. Trip c. Isol at e d. Trip then restart (1.00) QUESTION 6.02 (2.50) Reactor water l evel has decreased to below -129 inches. Drywell pressure is 1.2 psig. The RHR and CS pump interlock is satisfied. i n. WHEN will the ADS blowdown commence? (0.50) b. WHY is a low level (12.5") signal used along with a low level i
'
initati on signal (-129") to initiate ADS? (0.50) c. During blowdown the operator depresses the ADS D1V I Logic Reset l Push Button. DESCRIBE the response of the ADS system. (0.50) d. During blowdown the operator turns ADS DIV I and II Auto Inhibit ! Switches to the INHIBIT position. DESCRIBE the response of the ; ADS system. (0.50) o. An operator observes that both the green and amber lights are lit for the acoustic monitor for an SRV. DESCRIBE what information this provides to the operator concerning the SRV. (O.50) l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) . er -- wgge e P - - * err- M*
.
6 __PL@N1_SYSIEMg_ DESIGNt_ CON 189Lt_@ND_lNSIBUMENI@llgN PAGE B l l
.i '
R l QUESTION 6.03 (2.00) Th9 plant is operating at 23% power and both Recirc Pump M/A Transfer Stations are in MANUAL with the Master Controller set at 50% speed D d3 mand. The "Rerirc Flow B Limit" annunciator is CLEAR.
For each of the following situations, STATE HOW the speed of Recirc Pump "B" wfli change (INCREASE, DECREASE, or REMAIN THE SAME), WHICH COMPONENT (S) of the control system will limit or prevent the speed change and WHAT SPEED the Recirc Punip is limited to.
NOTE: Figures T-LOT-OO40-2 L 8 are provided for reference.
a. Vessel pressure oscillations cause feedwater flow (actual and indicated) to oscillate plus and minus 5%. (1.00) d eu-nae. PAk% b. Recirc Pump "B" M/A Transfer Station manual gatan t i on.etee i s turned fully in the cuarter clochw&e- ' -- *:^^ (1.00) M bui Gr., s a tem.w.g ca u.a p . QUESTION 6.04 (3.00) A reactor high pressure signal of 1O93 psig exists.
a. WHAT two (2) additional conditions must exist for an automatic initiation of SBLC to occur? (1.00) b. STATE the four (4) actions that occur upon an automatic initiation of SBLC. Do not include redundant components as separata actions. (2.00) 2.50 QUESTION 6.05 '3.00) Tho mode switch is in STARTUP with reactor pressure equal to 800 psig when a MSIV isolation occurs.
a. LIST the possible signal s that could have caused the i sol at i on.
(Eetpoints not required) (1.00) b. The MSIVs required closing time is three (3) to five (5) seconds.
WHAT are two (2) reasons For each the minimum and maximum time requirements? (1.00) I c. TRUE or FALSE? A scram will occur if MSIVs close in only two of the Main Steam lines. (0.50)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l l
. . . _ .
-. ..-
6 t__P(@NI_@y@lEdg_pEgigN t_CgNIBQ(t_@NQ_lN@l@UDENI@llgN PAGE 9
. .
QUESTION 6.06 (3.00) Under LOCA conditions with a LOSS OF OFFSITE POWER,,. Diesel Generator Dll output breker has just closed on to Safeguard Bus,Oll.
a. WHAT four (4) loads will be automatically sequenced on to the bus? LIST the loads in the sequencing order. (Times are not required) (2.0) b. LIST two (2) other loads on the bus that will not automatically sequence onto the bus. (0.5) c. WHAT operator action is required to restart the loads that do not automatically sequence onto the bus? (0.5) OUESTION 6.07 (2.50) STATE the automatic action (s) that will occur when each of the process radiation monitors exceed the condition listed.
a. Refueling Area Ventilation Exhaust Duct High Radiation. (0.50) b. Reactor enclosure radiation monitors Hi-Hi setpoint. (1.50) Scro u w h c. RHR -Hest combined loop monitor Hi radiation. (0.50) I QUEST.;ON 6.08 (3.00) l The reactor is at 50% power with the Lead Limit set at 65% and Maximum Ccmbined Flow Limiter at 115%. An electrical failure occurs that causes the pressure set signal to decrease by 10 psi.
DETERMINE the final control valve flow rate and bypass valve flow rate. Refer to the attached drawing of the Electro-Hydraulic Control Logic (LOT-0590-6). DESCRIBE how you determined your answer. (3.00)
(***** CATEGORY 06' CONTINUED ON NEXT PAGE *****)
-_ - - . _ _ _ __ __ - . _ _ _ - _ . _ . _ ._
...
6 PLANT SYSTEMS DESIGNg_CQNTRQLt_ANQ_INSTRUMENTAIIQN PAGE 10
*
QUESTION 6.09 (2.00) Tha Feedwater Control System is being operated in 3-Element Control using reactor level detector channel "A". Reactor power is at 85%. For each of the following instrument or control signal failures, STATE HOW recctor water level will INITIALLY respond (INCREASE, DECREASE, or REMAIN CONSTANT) and briefly EXPLAIN what happens in the Feedwater Control System to cause the Desponse.
NOTE: A block diagram of the Feedwater Control System is attached.
a. Channel "A" reactor level detector signal fails downscale. ,
(1.0)
l 6. "B" reactor feed pump speed c ontr ol l er fails low. (1.0) l l
.
QUESTION 6.10 (3.00) ,
For each condition SELECT the action (s) from the list below that will ' occur. If no action will occur. state NONE.
a. A reactor startup is in progress with IRMs on rance 2. The ,
'
operator withdrawing SRM detectors also has channel A of the IRMs selected. (0.50) b. The operator adjusts recirc flow such that there is a 15% mi smat ch bet, ween loops. (0.50) c. Reactor is in the RUN mode and APRM 'D' ' ails downscale. All IRMs are withdrawn and indicating 25 on range 3. (0.50) d. Reactor is in STARTUP mode and control rods are withdrawn to 15% power. (0.50) e. An approach to criticality begins with IRM C on range 2. (0.50) f. Reactor is in RUN mode and APRM 'A' is bypassed, APRM 'F' mode switch is placed in the "STANDBY" position. (0.50) ACTIONS: 1. Half Scram 2. Full Scram 3. Rod Block (***** END OF CATEGORY 06 *****)
. ?t__E89EE998EE_ _U98U9hi_9EU98U9hi_EUg@ggggy_ANp PAGE 11 889196991C86_CgNIBg6 l l - t l l l i QUESTION 7.01 (2.00) In accordance with HP-310, "Radiation Work Permits", STATE (YES/NO) whother an RWP is required for each of the following conditions e. Who.le body radiation level = 25 millirem /hr (0.50) b. Average removable surface contamination of 5000 dpm/100 sq.cm , beta-gamma (0.50) c. Average removable surface contamination of 500 dpm/ 100 sq.cm, alpha (0.50) d. Neutron dose of 25 millirem /hr (0.50). i l
!
l l I I QUESTION 7.02 (2.00) l According to Health Physics Procedure itP-102, "Administrative Dose l Limits. Dose Extensions, and Notification Hequirements": a. WHAT is the PECO administrative whole body dose limit for 1 year with a current NRC Form-4 on file? (0.50) i b. Based upon 10CFR20, WHAT is the maximum allowable whol e body accumulated dose for a 30 year old person? (1.00) c. WHAT whole body exposure, if exceeded, requires immediate notification (1 hour) to the NRC? (0.50) l l l
;
l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
._ _- ,
Zi__PBgggpuBgg_ _yggd@6t_99NgBU@61_gDgBQgNgy_AND PAGE 12 6091969G1986_99NIBQ6
.
QUESTION 7.03 (2.00) c. Concerning procedure GP-5, "POWER OPERATIONS": 1. Power in decreased to 8% rated thermal power. The reactor operator informs you that the control rod pattern is-NOT latched in the RSCS. What action per GP-5 do you tske? (0.50) 2. Why are you cauti oned to avoid prolonged recirculat_on pump operation at resc,iance speeds? (0.50) i b. Concerning GP-3, "NORMAL PLANT SHUTDOWN": 1. Why are ret rculatico pump speeds required to be maintained within 5% of each other? (Assume v.ower is 50%) (0.50) 2. At about 60% power when shutting down the first condensate pumn, WHY should the operator be prepared to take manual control of the incividual Reactor Feed Pemp MGU crotrollere? (0.50) GUESTION 7.04 (2.50) In accordance with procedure SSI.8.b, "Shutdown Cooling Operation": e. Placing the RHR system in Shutdown Cooling requires that t!.e oper ator shut and tag the minimum flow valve. LIST the two purposes for performing this action. (1.0) D. Wher operating AHR in the shutdown tooling mode the minimum pump f4ow that is allowed by the procecure is 1500 gpm. EXPLAIN why flow must be maintained greater than this value. (0.5) c. EXPLAIN why . low through a RHR heat exchanger is limited to las than 11000 gpm. (O.s) I d. EXPLAIN why reactor water level must be me'itcaned above 60 inches as reed on the S;)u Jown range indicator O i 42-R605) or 78 i nches on the Upset range :r- fr (LR-42-R608). (Do not explain why the values are differe 1 (0,5) 1
l ,
* (***** CATL ' Ai -
NUED ON NEXT PAGE *****) ! , l . e e- -- c -- .p 9
ZL__P69g@Dyg@g_;_ggggg(g_g@NQSU@(t_EDE6@@dgy_AND PAGE 13 88D1969GIC66_CgNI696
.
QUESTION 7.05 (3.00) In accordance w4th Special Event Procedure SE-1, "Remote Shutdown": a. For each of the f ollowing situations STATE (YES/NO) whether entry into SE-1 is required.
1. Fire in a Diesel Generator Room (0.50) 2. Fire in Auxiliary Equipment Room (O.50) 3. Cable spreading Room uninhabitable due to noxious fumes (No fire) (0.50) b. LIST the three (3) immediate operator actions to be performed in the Control Roem, prior to evacuation, when a remote shutdown is required. (1.50) 3.co OUESTION 7.06 12.501-During a reactor startup, the unit is at 28% power and 50% core flow
. Shen reactor water level begins tc oscillate (+ 5 inchen).
a. WHICH Operational Tr ansi ent Pr oc ed u r 9 ( r. ) should be entered? If none, state NONE. (1.00) b. WHAT ACTIONS would you, as the Shift Supervisor, direct the operators to take? Include any immediate actions required by procedure. (1.00) c. If reactor water level reaches 100" the operator, by procedure, is directed to scram the reactor and close the MSIVs. WHAT is the bases for each of these actions? (1.00)
/ (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
.Z:__PBQCEgUBEM_;_NgBM@61_@@dgBM@bz_EMEBGENCY_@Ng PAGE 14 689196901C86_CgNI606
.
QUESTION 7.07 (2.50) For each cf the f ol l owi ng condi ti ons, determine whether entry into the Off Normal (ON) Procedures i s r equi-ed. If entry is not required, otete NONE. STATE any ON(s) that would be entered. The ON Procedure Index is attached for your reference, c. An unexplained increase in reactor power accompanied by an unexplained decrease in core flow indication (0.50) b. ROD OVERTRAVEL alarm when rod is f ully withdrawn (0.50) c. CRD CHARGING WATER LOW PRESS alarm (0.50) , d. Total loss of SRMs in STARTUP mode with IRMs in range 3 (0.50? e. Standby Gas Treatment System surveillance test results indicate that SGTS can maintain the Reactor Enclosure at -0.18 inches of water with a flow rate of 1350 SCFM (0.50) OUESTION 7.08 (3.00) For each set of condi ti ons bel ow, STATE which, if any, Trip Procedures (100 Series) should be entered. If none, state NONE.
a. A loss of drywell cooling occurs. Operators vent the drywell to maintain pressure below 1.2 psig. Drywell temperature is 150 degrees F. (0.50) b. A reactor scram occurs due to a turbine trip from 45% power.
Reactor level decreases to -10 inches following the st'am but is automatically recovered b/ feedwater. (0.50) c. A MSIV iso.ition occurs due to improper testing by ILC technicians.
The reactr scrams due to the i sol at i on. (0.50) d. During a reactor shutdown, the operator places the mode switch in STARTUP at 20% power. Reactor power decreases to 2% due to partial insertion of control rods. (0.50) :
i o. A small leak in the drywell causes drywell pressure to increase to l 3,2 psig. (0.50) f. A failure of the EHC system results in a pressure increase which causes a high pressure scram. (0.50)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
l l I l
,
--
_ Ti__6699ggu6[@_;_Ug6d@[t_@@yg6d@bg_{DgBQ[ygy_AND PAGE 15 B09196991C06_CQN18g(
.
QUESTION 7.09 (2.50) Procedure T-ll3, "Blowdown Cooling", directs the operator to steam cool the reactor. This is accomplished by opening one SRV. If RPV pressure drops below 700 psig during steam cooling, the procedure directs the operator.to T-112, "Emergency Blowdown".
c. WHY must Emergency Blowdown be performed in place of steam cooling when pressure drops below 700 psig? (0.50) b. HOW MANY SRVs are required for performing the Emergency Blowdown? (0.50) c. After Emergency Blowdown is comp.ete,
'
it is assumed that injection from at least one system will be cuccessful. WHAT is the preferred system and WHY is this system preferred over other systems? (1.50) QUESTION 7.10 (3.00) Using the attac'hed Emergency Plan Implementing Procedure EP-101,
"Classification of Emergencies", CLASSIFY the following events.
a. A cable fi re started in the HPCI Room and was promtly exinquished by the Fire Brigade. Damage to HPCI control cables suspected. (0.75) b. During steady state operations, SJAE Discharge radiation monitor levels increased from 20 R/hr to 210 R/hr over 30 minutes. (0.75) c. A total loss of Control Room annunciators occurrai concurrent with a reactor scram. (0./5)
$
d. All Diesel Generators started and picked up emergency loads during a blizzard which caused a loss of all off-site power. Sustained wind speeds of 75 mph are indicated on OBC 699. (0.75) k (***** END OF CATEGORY 07 *****) f l _ _ _ _ _ _ __ ___ ___- _- ___-
_ _ _ _ _ _ _ _ _ _ _ _ ai__69 MINIS 16911'gE_P69CEQUBEs t _CgNQ111gN@t,@Np,LIMll@llgN@ PAGE 16
.
QUESTION 8.01 (2.50) STATE whether a SAFETY LIMIT or a THERMAL LIMIT would be violated for EACH of the f ollowing operating condi tions. If none, state NONE.
1. Core flow is 20%, thermal power is 33%, pressure i s 755 psig. (0.50) 2. All relief valves open, reactor pressure is 1315 psig. (0.50) 3. A P-1 edit shows C'MFLPD is O.09. (0.50) 4. MODE switch in run, all rods in, level is -150 in. (0.50) 5. Reactor pressure i s 900 psi g, core flow iu 30%. MCPR is 1.03. (0.50)
*
QUESTION 8.02 (1.50) For each of the following automatic protective functions, STATE the Tech Spec BASES for the function. Include any applicable Saf ety Limits.
a. Turbine Stop Valve Closure SCRAM (0.75) , b. Reactor Vessel Water Level-Low SLR R fW (0.75) OUESTION 8.03 (2.50) a. In accordance wi th Admini strati ve Procedure A-7, "Shift Op er a t i on s" , LIST the four (4) si tuations that require a Senior Licensed Operator or Licensed Operator to scram or shutdown the plant. (1.00) b. Fol l owi ng an unscheduled shutdown of the plant: 1. WHOSE (by title) approval is required for a restart? (0.50) 2. WHO shall direct the return to power? (0.50) l i 3. An unlicensed Shift Technical Advisor wishes to operate l the controls to withdraw rods prior to criticality under the direction of a the Chief Operator. Is he allnwed to do this? WHY or WHY NOT7 (0.50) l l ,
(***1* CATEGORY 08 CONTINUED ON NEXT PAGE *****) - _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ - _ - 9:__99DINigIB@llyg_E699gguggg3_gggg111gNgz_9Ng_LldlI@llgNg PAGE 17
.
QUESTION 8.04 (3.00) While on shift in the Control Room as the Shift Supervisor, you are roviewing the procedure to perf orm an operability surveillance following completion of maintenance on a valve in the RHR System, c. LIST the four (4) requirements that must be met for the procedure to be considered valid for use. (1.00) b. You have determined that only part of the RHR valve operability-surveillance must be performed for post maintenance testing. Is a temporary change required per A-3, "Procedure for Temporary Changes to Approved Procedures" in order to perform the partial surveillance? (0.50) c. Assumi ng a temporary change is required, WHAT requirements must be met to implement a temporary change before the procedure can be performed? (1.50) DUESTION G.05 (2.00) In accordance with Admi ni strative Procedure A-43, "Surveillance Testing Program": e. LIST two (2) of the three (3) responsibilities of Shift Super vi si on in regards to Surveillance Testing (ST). For each responsibility, STATE whether or not it may be delegated to the Chief Operator (CO) or Assistant Control Operator (ACO) (1.00) b. LIST two (2) responsibilities of Shift Supervision, if a ST performed on the shift failed and Technical Specification requirements cennot be met. (1.00)
. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****?
l
~ -- - _ _ - _ _ _
si__@gMIN1@IBBIlyE_P8QCEQUBES t _GQNQlIlgN@t_@NQ_LidlI@IlgNS PAGE 10
.
QUESTION B.06 (2.50) In cccordance with Administrative Procedure A-41, "Procedure f or Control of Plant Equipment": c. LIST the two (2) classes of equipment that are controlled
- by this procedure. (1.00)
b. WHOSE (by title) permission is required to remove equipment from service? (0.50) c. STATE one instance where an independent verification is NOT required when returning equipment to service. (0.50) d. A piece of equipment was released for surveillance testing at 0900 June 6, 1988. How long is permission granted to perform the surveillance? (0.50) QUESTION 8.07 (2.50) In accordance wi th Administrative Procedure A-31, "Procedure for Notification of NRC", STATE (YES/NO) whether the following events WOULD or WOULD NOT be a 1 hour reportable event to the NRC. Use tho attached Emergency Plan Implementing Procedure EP-101 for cl assi fi cation of the event, if required.
c. Thermal Power of 3290 MWt (0.50) i
'
b. A maintenance worker gets injured and contaminated while repairing a feedwater turbine and must be taken to the hospital. (0.50) c. A spurious initiation of Core Spray at 100*/. power during , testing of the system logic (0.50) l d. Operations resulting in a condition requiring a change in an approved procedure (0.50) l e. Operations resulting in an Unanalyzed Condition ,
(0.50)
i l
.
QUESTION B.08 (1.00) A step of a TRIP procedure violates a Technical Specification roquirement. Should you perform the TRIP procedure step or comply with the Technical Specifications? EXPLAIN YOUR CHOICE. (1.00)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
,-
.. 9 __0951Nig16@llVE_P69CEQU6E@t_CQNgillgNgt,,@Ng,LlDil@llgNg PAGE 19
-
!
.
l
!
QUESTION B.09 (3.00) ) l The plant is operating at 85% power. During a surveillance cycling test of the suppression chamber to drywell Vacuum Breakers, you are informed that the "A/B/C/D/Drywell VACUUM RELIEF VALVE OPEN" alarm will not clear, cnd the OPEN AND CLOSED valve position indication lights are lit for the
"A" vacuum breaker.
a. Are Primary Containment integrity requirements met in accordance with Technical Specifications? (Consider your answer in terms of Tech.
Spec. definitions) (1.00) b. Can the plant continue to operate? If yes, under WHAT conditions? If not, WHY not? (2.00) NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS (TS) TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE TS THAT YOU USE TO DEVELOP YOUR ANSWER.
QUESTION B.10 (2.50) Whcn you assume the midnight to eight A.M. shift, the plant is at 85*/. power and all conditions are normal with the following exceptions:
- APRM channel 'A' is bypassed for maintenance - APRM channel 'B' is failed low and bypassed.
Two heJrs into the shift, IRM channel 'C' fails downscale.
In accordance with the Technical Specifications:
- Are the applicable LCO's satisfied ? (1.00) - Can the plant continue to operate ? (O.50) -
If so, under what conditions ? If not, why not ? (1.00)
*** Justify your answers ***
NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS (TS) TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE TS THAT YOU USE TO DEVELOP YOUR ANSWER.
!
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
I i
,. - -- . - - - . . - , .
ai__8DMINi@lB@l1VE_EBQCEQU6E@t_CQNQlligN@t_@gQ_(1MlI@IlgN@ PAGE 20
.
QUESTION 8.11 (2.00) Limnrick Unit 1 is in COLD SHUTDOWN. A unit startup is scheduled to commence on your shift. The Maintenance Supervisor reports the fcilure of the outboard blower for the MSIV Leakage Control System.
Ha ostimates that it will take two (2) days to return the blower to ccrvice.
In view of the above malfunction, CAN you proceed with the startup? If you can, under WHAT conditions? If you cannot, WHY not? (2.00) NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS (TS) TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE TS THAT YOU USE TO DEVELOP YOUR ANSWER. ,
:
l l I (***** END OF CATEGORY OB *****)
(************* END OF EXAMINATION ***************) .- - . . _ . , . -..
-5___IUE96Y_9E_dyC6Ee5_E9WE5_P6@yl gPgy@!!QN3,[bylgSz,@hp -PAGE 21
'
ISE800DyNAdlCS
' ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I !
l I l i
ANSWER 5.01 (2.50) a. Doppler or fuel temperature b. Void c. Moderator temperature d. Void e. Moderator temperature (0.5 each) (2.5) REFERENCE LOT-1440, p. 1, LDs 1, 2&3 K/As 292004: K1.01 (3.2/3.2), K1.05 (2.9/2.9), K1.10 (3.2/3.2), & K1.14 (3.3/3.3) 292OO4K101 292OO4K105 292OO4K110 292OO4K114 ...(KA'S) ANSWER 5.02 (2.25) c. Increase [+0.253. Higher steam flow results in a higher pressure drop C+0.50]. b. decrease [+0.253. Higher heat input to condenser with same cooling [0,503 c. Increase [+0.253. Extraction steam energy to feedwater heaters increases faster thar. feed flow [+0.503.
REFERENCE LOT-1270, LO 73 p. 10-13 LOT-1190, LO 33 p. 7-21 K/As 293006 K1.03 (2.4/2.5), 293005 K1.05 (2.7/2.8) L 292000 K1.20 (3.3/3.4) 292OOOK120 293OO5K105 293OO6K103 ...(KA'S)
;
l i l
-
,
o g ~ THEORY OF NUCLEAR POWER PLANT OPERATION g,FLUIQ@g_ANQ PAGE 22 IHEBUQQyN@dICg
. ANSWERS - LIMERICK 1 -CB/06/07-NRC REGION I
:
ANSWER 5.03 (3.00) a. The decrease in the burnout term [0.53 with-the production of xenon'from iodine still'at the higher power rate dominates CO.53 causing the xenon concentration to increase. (1.0) b.- Peripheral rod worth will increase CO.53 because the highest xenon concentration will be in the center of the core where the highest flux existed previously CO.S3, This will suppress the flux in the center of the core and increase the flux in the area of the peripheral. rods, thereby, increasing their worth CO.53. (1.5) c. More than half the value at 100*/.. (0.5)
.
.
REFERENCE LOT-1510, LOs 3, 5 & 61 p. 6-10 LOT-1490, LO 4; p. 10 K/As 292006 K1.11 (2.6/2.7); 292005 K1.09 (2.5/2.6) 292OO5KlO9 292OO6K111 ...(KA*S) ANSWER 5.04 (2.00) , a. Larger , b. Smal l er c. Larger d. Smaller (0.50 Each) REFERENCE LOT-1490 LO 43 p. 4-12 KA 292005 K1.12 (2.6/2.9) 292OO5K112 ...(KA'S)
,
j
;
l l l
5i__IUEggy_gF_NyC(E@@_[gWEQ_f(@N1_g[E@@llgN g _{(ylpS t
,@NQ PAGE 23D IMERdODyN@dlCS
. ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I ANSWER 5.05 (2.00) O psig = 14.7 psia = 212 deg F (O."O) 1000 psig = 1014.7 psia = 546.3 deg F (0.50) GP-3 Limit = 100 deg F/hr (0.50)- 546'.3 - 212 = 334.3 deg F (0.25) 334.3 deg F / 100 deg F/hr = 3.3 hours (0.25)
'
REFERENCE LOT-1160, LO 1; p. 4-6 Steam Tables GP-3, p. 12 K/A 293003 K1.23 (2.8/3.1) 293OO3K123 ...(KA'S) , ANSWER 5.06 (2.25) a. Remains constant (0.25). Flow is controlled by the HPCI flow controller which will attempt to maintain a constant output flow regardless of reactor pressure (0.50).
6. Decreases (0.25). The flow controller functions to maintain a constant flow, thus pump discharge pr essure is decreased along , with the decreasing reactor pressure to maintain constant flow. < OR Since the flow controller maintains a constant flow to the ! reactor, as reactor pressure decreases, the pump discharge head must decrease to maintain a constant flow. (0.50), c. Decreases (0.25). To maintain a constant flow, turbine PPM must also decrease (0.50). ) REFERENCE LOT-1290, LO 53 p. 13 LOT-0340, LO 5; p. 6 K/As 291004 K1.05 (2.8/2.9); 293006 K1.08 (2.5/2.6) 291004K105 293OO6K108 ...(KA'S)
. .
_ _ _ m-90%) condition (0.5) by tho insertion of shallow rods (rods below the core midplane) (0.5),
thG resul ting posi tive reactivity f rom local decrease in void formation in the lower portion of the control cell (0.5) would more than offset the negative reactivity of the low-worth rod incertion (0.5).
REFERENCE LOT-1490, LO 91 p. 5-l'd K/A 292005 K1.12 (2.6/2.9)
292OOSK112 ...(KA*S)
ANSWER 5.Oe (2.00>
a. FALSE b. FALSE c. TRUE
*
d. T-R44G FT E S c )
(0.50 Each)
REFERENCE LOT-13OO, LO 21 p. 4-6 K/A 293007 K1.13 (2.3/2.9) 293OO7K113 ...tKA*S)
.
ANSWER 5.09 (2.00) l ( 0.5 0) a. MAPRAT = MAPLHGR actual /,dAPLHGR-LCO '!.00! '
(MAPRAT is the Maximum Average Pl anar Ratio. It is the comparison between the actual APLHGR to the MAPLHOR limit as programmed into l the Process Computer.) i l
b. Yes (0.25), we have exceeded the Tech Spec limit for APLHGR (0.50).
The value of MAPRAT should never be greater than 1.0 (0.25).
c. The clad temperature can exceed 7200 F during a DB LOCA (0.50) REMERENCE I LOT-1410, LOs 2-41 p. 4 TS 3/4.2.1, Average Planar Linear Heat Generation Rate K/As 293009 K1.10 (3.3/3.7) & K1.13 (3.1/3.6) 293OO9K113 ...(KA'S) i I
l
-, ,
l
5 __ISEggy_gE_NgCbE93_EQWE5_EL@N1_gEEg@llgN t _{(ylgS _@ND t PAGE 25 ISE6dODYN951CS ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION 1 ANSWER 5.10 (3.00) c. Reactor water level decreases because the feedwater pumps are no Innger feeding the reactor and steam is still being removed (0.25).
b. The l evel continues to decrease f ollowing the scram because of void collapse (0.25) and because steam is being removed via BPVs and SRVs for decay heat removal. (0.25) c. Rx pressure decreases as reactor power decreases (0.25). Power decreases initially due to the decrease in inlet subcooling (0.25) and decreases at a faster rate when the reactor scrams (0.25).
d. After the MSIVs close decay heat causes the pressure to increase (0.375) until the relief valves open (0.375) e. Core inlet flow decreases slowly due to loss of feedwater (0.375) The rate of decrease increases due to the trip of the recirculation pumps (on low level) (0.375) REFERENCE LOT-OO40, LOs 3, 4; p. 6-10 LOT-0550, Los 6, 7; p.8-21 LOT-0580, LO 4; 17-19 K/As 295001 K3.01 (3.9/3.9) 295001 K3.12 (3.8/3.9) 295001 K4.11 (3.5/3.5) 259001K301 259001K312 259001K411 ...(KA'S) , ANSWER 5.11 (2.00) a. 2 (1.00) or i b. 3 (1.00) REFERENCE LOT-1260, p. 16 - 18, LO 4 K/As 293006: K1.05 (3.2/3.3), K1.13 (2.6/2.7) 293OO6K105 293OO6K113 ...(KA'S) _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - _ _ _ _ _ -________ ___ __ _ _ __- __
. _ _
,
65__ELONI_@y@lEdg_pE@}GN t _QQNI69Lg _@Np_lN@lgydENJ@lIQN) PAGE 26 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I
.
ANSWER 6.01 (1.00) d. (1.0) REFERENCE LOT-340, Objective 9 KA 206000 K4.01 (3.8/3.9) 206000 K4.03 (4.2/4.1) 206000K401 206000K403 ...(KA'S) ANSWER 6.02 (2.50) , c. after the 420 second timer times out (0.5) b. Low level (12.5") signal is a confirmatory signal that prevynts inadvertant initiation following a single instrument failure (0.50) c. Blowdown will continue. (0.5) (If both DIV I and II were reset, blowdown would be interrupted.)
d. Bl owdown will continue. (0.5) (These switches do not terminate blowdown once initiated) o. The valve has opened (0.25) but is preaently closed (0.2S) I REFERENCE LOT-0330, LO 2, 5 & 63 p. 9-13 & Figure 6 K/As 218000: K4.Ol (3.7/3.9), K4.02 13.8/4.0), K4.03 (3.8/4.0) K5.01 (3.8/3.8), A3.03 (3.7/3.8) 21800K403 ...(KA*S) ANSWER 6.03 (2.00) e. Decrease (0.5) <20% feedwater flow places the 28% speed limiter in service (0.25) educing speed to 28% (0.25).
( o - no e bA- u . oo'>') b. Decrease (0.5); Scoop Tube Positioner Stops (Electrical or Mechanical)
(0.25) limit the decrease to 20% (0.25)
REFERENCE LOT-OO40, LOs 11 & 12; p. 5, 6, 10 & 11 K/A 202002 SG.09 (3.8/3.5) 202OO2 GOO 9 ...(K4'S) l
!
i
I l l i l
._____--_--_ _ __--___- - _-____ - __-__-_____ -__--____ :__-___ ___-____ _ ____-____ __ ___-______ ______-___________ - ____ --__ ___ _- _.- __
61__E66NI_gy@ Igd @_QEgigN 3 _ggylgg(3_Byg_IggIggdEyIBIIgy PAGE 27 ANSWERS -- LIMERICK.1 -88/06'07-NRC REGION I ANSWER 6.04 (3.00) c. No APRM downscale (0.5)
(0.5) -Eg- second time del ay b. All SBLC pumps start (0.5?
All squib valves open (0.5) o c- conhw.% bgM s to.gwnh RWCU isolates (0.5) 10 minute reset timer actuates. (0.5) A% ar m bt A m e e g 4 A W .\ oc powo 6a crew REFERENCE LGS: LOT-0310 p. 16 and 17.
Locening objectives 9, 10, and 11.
K/A 211000 A3.08 (4.2/4.2) l
'
211000A308 ...(KA*S) ANSWER 6.05 s) a. High Steam Line flow (0.27$ Low reactor vessel level ( 0. 27) . MSL High temperature (Tunnel or Turbine enclosure) (0. Z6) MSL High Radiation (0.2%) Manani (o.t) b. Fast enough to:
- Limit coolant loss during SLB outside containment - Limit radioactive release during gross fuel failure j - Limit radioactive release during a major leak inside containment (Any 2, & O.25 each)
Slow enough to:
- Minimize damage to valve and piping i
- minimi z e transi ent in boiler l
!
(0.25 each)
c. FALSE (0.5) REFERENCE LOT-0120, LO 2.f L93 p. 13 & 18 TS P.4SES 3/4.4.7 K/As 223002 SG.06 (2.9/3.9); 239001 K1.27 (4.0/4.1) & K4.01 (3.0/3.8) 223OO2 GOO 6 239001K127 239001K401 ...(KA'S)
!
I I l
- - - - , ,_
, ,
6i__PL@NI_SY@ led @_DE@lGN t _ggN18Q6t_9ND_lN@l89dENI@llQN PAGE .28 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I ANSWER 6.06 (3.00) c. 1. RHR pump 1A (0.4) 2. Load Center Bus / Transformer D114 (0.4) 3. CS pump 1A (0.4) 4. ESW pump OA (0.4)
(0.4 for correct sequence)
b. 1. RHRSW pump OA (0.25) 2. TB Equip. Ccmpartment Exhaust Fan 1A (0.25) c. Manual reset (0.5) REFERENCE LOT-0660, LO 6; p. 14 K/As 264000 K3.03 (4.1/4.2), K4.05 (3.2/3.5) , 264000K303 264000K405 ...(KA*S)- ANSWER 6.07 (2.50) a. Refuel floor supply and exhaust i sol at e. (O.5) 5 6 (.T s e.V cA 4 cA. g +c RP N S o a. sW y*,s. % (. o . s t) b. Supply and exhaust ventilation i sol at e. (0.5! Standby gas treatment starts. (0.5) Reactor enclosure recirculation starts (0.5) c. Trips RHR service water pumps. (0.5) REFERENCE LGS: LOT-0720 p. 10, 11, 18. I. earning Objective 2.
LGS: LOT-0180 p. 23. Learning Objective 2.
KAs 272000: K1.01 (3.6/3.8), K1.06 (3.2/3.3), K1.08 (3.6/3.9), K1.09 (3.6/3.8), K4.02 (3.7/4.1) 272OOOK101 272OOOK106 272OOOK108 272OOOK109 272OOOK402
...(KA*S)
.6___PL@NI_Sy@ led @_9E@lGN t _CQN16QLt_@NQ_lN@l6QUENI@llgN PAGE- 29 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION 1
.
ANSWER 6.08 (3.00)
Output f rom pressure su.nmer A increases to JM7 psi (0.5). The HVG pcsses this signal to the Pressure / Flow convertegt(0.5). The Prcssure/ Flow converter output will increase to M/. (0.5). The load limit will limit.the signal to the control -valves to 65*/. (0.5).
Total flow is less than the maximum combined limiter (0.5) so the bypass valves will open to pass (0.5).
{7. flow REFERENCE LGS: LOT-0590 p. 6, 7, 8, 13, 15, 16.
Learning Objectives 3, 8.
KAs 241000: K3.05 (3.7/3.7), K3.06 (4.1/4.1) & K3.08 (3.7/3.7) 241000K3OS 241000K306 241000K3OB ...(KA'S) ANSWER 6.09 (2.00) c. Causes reactor level to increase ( 0. 5) due to the level control system having a l evel error, (level set > indicated level) (0.25) resulting in an increase in the speed of the reactor feed pump turbines (0.25).
b. Reactor level decreases (O.5) because the RFP turbine "B" decelerates to low speed setting (0.5). (The level decrease causes a flow error which causes the master contrclier to raise the speed of turbines "A" &
"C". The l evel will stabilize at the original setpoint.)
REFERENCE LOT-0550, LO 71 p. 18-20 K/As 259002 K6.05 (3.5/3.5); K4.06 (3.1/3.2) 259002K406 259002K605 ...(KA'S) ANSWER 6.10 (3.00) c. 3 (0.5) ;r NeuE b. NONE ( 0. -D c. 3 (0.5) u. 2
' (0./3)(and 3)40.23)
O. '3.5? f. 1 (0.25) and 3 (40 . 2 5 ) REFERENCE LGS: LOT-0250 p. 10. Learning Objective 10. * LOT-0270 p. 11, 12, and 13. Learning Objective 7.
K22 215003 K4.01 (3.7/3.7), 215005: K4.01 ~3.7/3.7) & K4.02 (4.1/4.2) 215CO3K401 215005K401 215005K402 ...(KA*S)
Zi__P69CEpyBES_;_NQ@d@(i_@BNQBd@Lt_EDEggENCy_@NQ PAGE 30 5991969GICe6_GQNIBQ6
"ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I
.
ANSWEH 7.01 (2.00) . c. NO (0.50) b. NO (0.50) c. Y'c. G (0.50) d. YES (0.50) REFERENCE LOT-1760, LO 5 & p. 21, 22 K/A 294001 K1.03 s3.3/3.8) 294001K103 ...(KA'S)
.
ANSWER 7.02 (2.00) n. 4500 milli-Rem / year (0.50) b.(5(N-18) = 5(30-18/)= 60 Rom &G.3G im l cr a.u i . L O."C lm u.icul:t:;r' C a .co) c. 25 Rems (0.50) REFERENCE LOT-1760 LOs 1, 2, SRO-2 & p. 4,5 K/A 294001 K' 03 C . 3/3. 8) 294001K103 ...(KA*S) - ANSWER 7.03 (2.00) a.1. (Immediate) SCRAM (0.5) c.2 To prevent damage to the pump internals (0.5) b.1. To ensure adequate core flow coastdown on a LOCA (0.5) b.2 Because ther e is a potential for the loss of the master l evel control signal. (0.5) REFERENCE LOT-1530, LO 2 GP-5 p. 4, 53 GP-3 p. 4, 5 T/S B3/4 4-1 K/As 201004 G.1 13.9/4.1), 202001 G.10 (3.5/3.7) & G.6 (3.0/4.1), 259002 G.10 (3.3/3.4) 201004 GOO 1 202OO1 GOO 6 202OO*G010 259002G010 ...(KA'S)
PAGE 31 f ZE__E09EEE98EE_!_N98UOhi_9pNgggg(g_ggggggggy_ANg 68919LQG1C8L_CQNI@QL .
ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION 1-
;
ANSWER 7.04 (2.50) a. Ensure all flow is to the vessel terS7 and prevent inadvertant draining of the vessel M O. cod b. Prevent pump overheating (0.5) c. Preve-t damage to the heat exchanger (0.5) d. Ensure proper natural circulation (0.5) REFERENCE LGS: Shutdown' Cooling Operation (S51.8.b) p. 3 and 5.
KA 205000 G.10 (3.2/3.3) 205000 K1.02 (3.6/3.6) 205000G010 205000K102 ...(kA*S) ANEWER 7.05 (3.00) c. 1. NO (0.50) 2. YES (0.50) 3. NO (0.50) b. 1. Scram the reactor (0.50) 2. Trip the main turbine (0.50) 3. Close the MSIVs (0.50) REFERENCE LOT-1563, LDs 1, 2 & p. 3 SE-1, "Remote Shutdown", p. 1 and 2 K/As 295016 SG.11 (4.1/4.2) & SG.10 (3.8/3.6) 295016G010 295016G011 ...(KA*S) l I I
..
. ._ _ - - . _ _ _ _ _ - _ _ _ _ _ _ _ _
Z __PBgCEDyBES,;_Ng8M@L _@BNgBM@bt_EMEB@ENCy_@ND t PAGE 32 88 Dig (gGIC@(_CgNI@g( ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I 2h J.ew ANSWER 7.06 '2.50t-( i . c) c.(OT-100, Reactor Low Level -fer5) and OT-110, Reactor High Level +OrErt
" - ' - ' ' ' - '^ ~^'
b.
(If a RFP controlle- malfunction exists,) take manual control of the RFP(s) to control level f0. N Ream Mes J.i wW Wott ure e a (c.s) If a scram condition occurs, enter Procedure T-100 M H-3 &) L o. C) c. Scram Turbine Stop Valve Closure Scram is bypassed when power is less than 30*/. (as sensed by the first stage shell pressure) so, a turbine ' trip (D +54") would not have scrammed the reactor (0.25) and a scram i s necessary to limit the heat input tn the suppression pool (0.25) Close MSIVs To protect aginst unnecessary flooding of MSLs (MSL supports downstream of outboard MSIVs are not designed for flooding)
(0.5)
REFERENCE LOT-1540 LOs 1, 2, 3 L p. 3, 4, 25, 26 OT-100, "Reactor Low Level", p. 1 OT-110 "Reactor High Level " , p. 1, 2, 3 K/A 295031 SG.10 (4.0/3.8) L 295008 AK3.01 (3.4/3.5) 295031G010 ...(KA'S) ANSWER 7.07 (2.50) a. None (0.5) b. ON-104, 7
"Control Rod Problems" (0.5)
c. ON-10E, "Control Rod Drive System Problems (0.5) d. Men !O,5F oN-scA,a m a teu cA sex,i4,% ,,A9M q w " (, c 5 ) 9. ON-111, "Loss of Secondary Containment" ~(0.50) REFERENCE LOT-1550 LO 1 & p. 3-20 ON-100, "Failure of a Jet Pump", p. 1 ON-104, "Control Rod Problems", p. 1 ON-107, "Control Rod Drive System Problems", p. 1 ON-109, "Total loss of SRM, IRM, or APRM Systems", p. 1 l ON-111, "Loss of Secondary Containment", p. 1 K/As 202001 K3.03 (3.9/3.9) 201003 A2.02 (3.7/3.8) 201003 A2.OB (3.0/3.7) l
'
215004 K3.G4 (3.7/3.7)
- - _ _ _ _
. --______ _ - -
7'. PROCEDURES - NORMAL _ABNORMALt_ t EMERGENCY _AND PAGE 33 69DIQLQGlC@(_CQNI6QL kNSWERS -- LIMERICK 1 -88/06/07-NRC REGION I ANSWER 7.08 (3.00) c. T-102, "Primary Containment Control" (0.50) b. T-100, "Scram" (0.50) c. T-101, "RPV Control " (0.50) (T-100, "Scram") d. T-100, "Scram" (0./5) o. T-101, "RPV Control " (0.25) and T-102, "Primary Containment Control" (0.25) (T-100, "Scram") f. T-101, "RPV Control" (0.50) (T-100, "Scram") REFERENCE LOT-1560, LO 3 & p. 7 K/As 295006 SG.11 (4.3/4.5), 295024 SG.11 (4.3/4.5), 295026 SG.11 (4.2/4.5) 295028 SG.11 (4.2/4.4), 295029 SG.11 (4.2/4.5), 295030 SG.11 (4.3/4.5) 295037 SG.11 (4.4/4.7) 295006G011 295024G011 295028G011 295029G011 295030G011 295031G011 295037G011 ...(KA*S) ANSWER 7.09 (2.50) e. insufficient steam flow for adequate core cooling exists with only one SRV open below 700 psig (0.50) b. Five SRVs (0.50) e r- h u. M9s c. One of the Core Spray systems (0.50). Spray heat transfer is preferred because the core temperature is elevated during steam cooling (to provide the temperature differential so steam can carry away core heat) (0.50). Coro Spray spargers (system) can safely lower core temperature with a reduced possibility of core damage (0.50).
REFERENCE LOT-1560 LO 5 & p. 11, 12 K/A 295031 SG.12 (3.9/4.5) 295031SG12 ...(KA'S) ANSWER 7.10 (3.00) c. Alert (0.75) b. Alert (O.75) oc- LM e d beA Lo.t,$') c. Site Emergency (0.75) d. Alert (0.75) REFERENCE LOT 1520, LO SRO-1, p. 5-7 EP-101, p. 2, 8, 10, 13, 14
- - - - - _ - _ _ - _ - _ - _ _ . _
'*
.. m-
'ai 4 '
h__EBQGEQ(18g@_ _NQ8dO_BgNQBdh_gligBQgNQy_BNQ PAGE . 34- 1 ' 7 .-
'
689196991986_GQNIBQ6 h
. ANSWERS -- LIMERICK 1 -08/06/07-NRC REGION I
.
.
i
> .K/A:294001 A1.16 (2.9/4.7)
294001A116 . . . (KA*S) , f
f e
i i b .
,
I I , I l l I i
!
.)
,
0 __0RDINi@l6911yE_E69CEgyBE@t_CQNQlligNgt_@Np_LlDil@llgN@ PAGE 35 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I . ANSWER 8.01 (2.50) ScScy 1. '5 r:21 Limit (0.5) 2. NONE (0.5) 3. NONE (0.5) 4. NONE (0,5) 5. Safety Limit (0.5) REFERENCE T/S 2.0, 3:? ~ LOT-3820 Objective 2, LOT-1840 Objective 3.b ANSWER O.02 (1.50) c. Anticipates the pressure, neutron flux and heat flux increase which result from the closure of the stop valves. (O.A) Icotect: th: eacter veusci pin =mu, u - c. d fuel ther c /h J.-ulim nofity !!-it=_ 'o 751-Ewe % A N u 3 .% cre u 1 4 t.c.25) b. U rd ir +-mn-irnt encl ysi s drelin; "i t h cc;1:rt i-"-atar" d-r c= ' o "A P stmai. U ,e iu l ih m. ...e l / h 7 draulim ..J v== sal precrue= = a f e t '/ 'i-itra Fcf eg beAow %\ epera geoei k ed.d agio L,ps (c.g @ u 40. 2Z Wg. eg cho,>c. b(vel % ** %A pro \echa b- W M greux lO (o.4c) REFERENCE T/S BASES Sec. 2.0, L DT-1830 Obj ective 4 KA 212000 G.6 (3.4/4.3) 212OOOGOO6 ...(KA*S) i ANSWER 8.03 (2.50) i i c. 1. Safety hatards indicated j
'
2. Doubt of safe condition 3. RPS parameters exceeded without a scram 4. By procedure direction (0.25 Each) I b. 1. Station Superintendent *e McM h6*/F ** des gaki abd (OM 2. Senior Licensed Operator (SLO) (0.50 each) or Shd+ SweuJii.* j 3. No. Only licensed operators or unlicensed operators in a licensed j operator training program may operate the controls (0.5) l i REFERENCE *
!
LOT-1570, LOs 2, 3, SRO-1, 2 l A-7, "Shift Operations", p. 11, 15, & 16 l 10 CFR 55.13(a)(2) l K/As 294001 A1.09 (3.2/4.3) & A1.11 (3.3/4.3) ! l l
,
!
I _ _ _ _ ___
'01__0951NISIB9IlyE_P89CEpW8E@2_CgypillgNgi_@Np_Lidll@llgNg PAGE.'36 ANSWERS -- LIMERICK 1 -C8/06/07-NRC REGION I
.
294001A109 294001A111 ...(KA'S)
.
ANSWER 8.04 (3.00) a. 1. Must be stamped in red, "Controlled Copy" to.25) 2. Station Supt. or Alternate signature and date (0.25)
' 3. DA Supt. or Alternate signature and date (0.25)
4. Date of use must be later than effective date (0.25) b. No. (0.5) c. Determine that the change does not change the intent of the procedure (0.25) Document the change (0.25) Have the change approved by the Shift Superintendent (0. 5' and an individual knowl edes.bl e in the areas affected by the procedure (O.25) who is a member of PORC or previosly designated (0.25) REFERENCE LOT-1570 LOs 1 L 2; p. 5, 6 L 16 A-3, "Procedure for Temporary-Changes to Approved Procedures", p. 3 Technical Specifications h.8.3 K/As 294001: A1.01 (2.9/3.4) L A1.02 (4.2/4.2) 294001A101 294001A102 ...(KA*S) ANSWER 8.05 (2.00) nd a. 1. Permi ssi on to perform ST - mayabe delegated 2. Signs off completed ST - may be delegated 3. Comply with Tech. Spec., (if ST failed) - may NOT be delegated (Any 2 O O.50 Each - O.3 for responsibility, 0.2 for delegation) b. 1. Deter mine system operabili ty 2. Determine if LCOs satisfied 1 3. Scheduling of any additional tests required l 4. Notify agencies required ! 5. Notify Plant Staff (Any 2, 0.50 Each) < REFERENCE LOT-1570, LOs 3, SRO-1, 2; p. 4B A-43, p. 2, 7 K/A 294001 A1.03 (2.7/3.7) 294001A103 ...(KA*S)
;
Ei__690lNi@lB@llyE_EggCEQU6E@t_CQNQlligN@t_@NQ_Lidll@llgN@ PAGE 37 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I ANSWER B.06 (2.50) c. 1. Safety 4Related equipment (0.50) (Q List) 2. Other equipment required to be operable or in surveillance by Tech. Spec. (TS equipment) (0.50) b. Shift Supervision (0.50) c. High radiation areas (0.50) o v- 44 chuk O@ Lnk uM be peMM d. 0800 June 7, 1988 (0.50) REFERENCE LOT-1570, LOs 2, 3, SRO-1 & p. 44 A-41, "Procedure for Control of Plant Equipment", p. 1, 6, 7, 8, 9, 12 K/A 294001 K1.02 (3.9/4.5) 294001K102 ...(KA'S) ANSWER 8.07 (2.50) a. no (0.50)
*s. yes (0.50) (Unusual Event declared)
c. no (0.50) d. No (0.50) a. yes (0.50) REFERENCE LOT-1570, LOs 3, SRO-1, 2 & p. 39 A-31, Attach. 2 K/A 294001 A1.03 (2.7/3.7) 294001A103 ...(KA*S) ANSWER 8.08 (1.00) Follow TRIP (0.5) TRIPS are the governing document. (0.5) REFERENCE LOT-1560 Intro To TRIPS, Objective #1, SRO1 KA 295025 G.7 (3.5/3.7) 295025 G.12 (3.9/4.5) 293025 GOO 7 295025G012 ...(KA'S) _ _ __ _ _
.. . .
Bi__9901NJ@IB911ME_PBgCEQUBESz_CQNQJIlgNgi_@ND_L]dlI@IlgNS PAGE 3D' ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I
~
\
l, ANSWER B.09 (3.00) c. Containment integrity requirements are met. (1.0)
!
b. Yes (0.5) T/S 3/4.6.4 ACTION b. applies. (1.0) Conditions: (ACTION b.) the other valve in the pair must be verified closed wi thin 2 hrs. and the open valve must be restored to closed within 72 hrs or be in hot shutdown within the next 12 hrs and cold S/D within the following 24 hrs. (0.5) ,
(The valve cannot be verified closed. Thus, the more conservative approach is to assume that the valve is open. Until some means other than the position indication shows the actual valve position, the posi ti on indication must be used- since there is no inf ormation given to place the indication in' doubt.)
REFERENCE Tschnical Specifications 3/4.6.4 LOT-1840 Objective 3.b, 3.c KA 223001 G.11 (3.3/4.2) 223001 K6.09 (3.4/3.6) 223OO1G011 223OO1K609 ...(KA'S)
!
ANSWER 8.10 (2.50)
'
TS, Table 3.3.1-1, requires that there shall be a minimum , of two (2) operable channels per trip system. In this situation .RPS Channel A does not meet the specification (1.0) (for downscale trips).
(The ACTION statement requires that you be 'in STARTUP within 6 hours.
BUT) - LCO 3.3.1 allows that if the minimum number of operable channels is not met in one channel, then that channel may be placed in the tripped condition within one hour.
THEREFORE - in this condition the plant may continue to operate (0.5) with a half-scram inserted on RPS Channel 'A*.(1.0) REFERENCE LOT-1940 Objective 3.b, 3.c TS 3/4.3.1 K/A 215005 SG.11 (3.4/4.1) 215005G011 ...(KA'S) __ _ _ ~ - - _ . . , - . - _ , -
- -.- . - - - - - -__ .- _ _ , - - - .
9 __0901Ni@l@@llyE_E6QCEQQBE@t_CQNyll[QNSg _@NQ_Lidll@llgN@ PAGE 39 ANSWERS -- LIMERICK 1 -88/06/07-NRC REGION I . ANSWER 8.11 (2.00) NO (0.5). (A reactor startup would represent a violation of the Tsch Specs.)
TS 3.6.1.4 requires that two MSIV LCS be operable in Op Con 1-2-3.
The action statement for TS 3. 6.1. 4 all ows 30 days of continued operation, (0.75) but Tech Spec 3.0.4 does not allow entry into an Operational Condition while relying on an action statement. (0.75) REFERENCE Technical Specifications 3/4.6.1.4, 3.O.4 LOT -1840 Obj ecti ve 3.b, 3.c K/A 223001 SG.11 (3.3/4.2) 223OO1SG11 ...(KA'S) i i
l l l
l
. . . - . _ - . . __ .-_ _. . . - . .
_ _ _ _ _ _ _ .
*
TEST CROSS REFERENCE PAGE 1' QUESTION VALUE REFERENCE - :----- ------ ---------- 05.01 2.50 SVPOOOO435 05.02 2.25 SVPOOOO438 05.03- 3.00 SVPOOOO441 05.04 2.00 SVPOOOO439 05.05 2.00 SVPOOOO444 05.06 2.25 SVPOOOO445 05.07 2.00 SVPOOOO440 05.08 2.00 SVPOOOO442 05.09 2.00 SVPOOOO443 05.10 3.00 SVPOOOO437 05.11 2.00 SVPOOOO436
------
23.00 06.01 1.00 SVPOOOO476 06.02 2.50 SVPOOOO456 06.03 2.00 SVPOOOO457 06.04 3.00 SVPOOOO458 06.05 . 3r00E5 SVPOOOO464 , 06.06 3.00 SVPOOOO459 06.07 2.50 SVPOOOO462 06.08 3.00 SVPOOOO463 06.09 2.00 SVPOOOO461 06.10 3.00 SVPOOOO460
------ ~3. 00 245 07.01 2.00 SVPOOOO455 07.02 2.00 SVPOOOO454 07.03 2.00 SVPOOOO447 07.04 2.50 SVPOOOO446 07.05 3.00 SVPOOOO450 07.06 2<S03 0 SVPOOOO448 i 07.07 2.50 SVPOOOO449 l 07.08 3.00 SVPOOOO451 )
07.09 2.50 SVPOOOO452 07.10 3.00 SVPOOOO453 ;
------
M 2.S.5' 08.01 2.50 SVPOOOO472 08.02 1.50 SVPOOOO475 03.03 2.50 SVPOOOO466 08.04 3.00 SVPOOOO465 03.05 2.00 SVPOOOO469 03.06 2.50 SVPOOOO468 08.07 2.50 SVPOOOO467 08.08 1.00 SVPOOOO471 4 08.09 3.00 SVPOOOO470 I 08.10 2.50 SVPOOOO474 l
l l l i
.. . . . . .. . . . . _ . . . - - - - , . . = .
_
,l
..
'
'*
: TEST CROSS REFERENCE !
P GE l2- 'l
-
r.
QUESTION ~ VALUE REFERENCE -
'
- :. . :------ ---- ----- ,.
'l 08.11 2.00:. ..SVPOOOO473 , - -25.00 -l --- ----- '100.00'
DOCKET NO 352-1 i l i
'
i
+
.
.
* O e
r
. =
E%% m IP MRa'Wi$ i M&"nNdifm"
' * * * '""
>=-
NiMFfd.! iMX2 3 !!! ice, 150.
b M W. - - 50. 4_^I
-
J
- ,
w a i N/
-
k.. E_ De. ,e t ,,, ,g .
. . -
.
mg. . . . . . . .,,.
,,,,,,,,. *=-
lif?Js F iU"a 3 r . *
-
!EA T Mus"@n
- ruwiisi i i=-
'M.W W1N /
/
/ f
-
, /
, ,
._.
, L sco, m< __ - m J l
. ..gm,,.
W M
. - .
a[ _
:
q. . , ,g7 . .
- -
o ,,
- * n se,, ,, ,,
a eta, a,
,
.
l SEE NOTE ON TABLE 15.211 f
LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSl3 REPORT l LOSS OF ALL FEEDWATER FLOW
'
FIGURE 15.2 9 R EV. 32. 05/84 l
. - . , . . - _ _ -. - _ . . - _ - - -
l
.
-
!
jfi;l
~
gl- -5 1 !! i p5]
-
g
--
s 1 2
- o m t T -1 l ' ' '
.
< !l= ~
? o '
. .
'
r 1 - l J :
. , g .l . 11 r, J
~
T ~I! }J e .f-" El< - b si I ' f*e
"
}A J",i
, , "
<
{"k i
.
-- .
.
i w ! {m j
.
_ L gi
! k j #
,
1 - - -* !a$ i
!I{l i- o f ! ) f):
p
-
__ s - i , lid
.. -,
5g I m 1 o 3 "
. } $$
33, _ g + -
.
-
.
,
I 4 _
. ; - " q1j< $]< f j, a 1,-
1-
., 'E* (,7 g- _ '
b '
*i * 'A
, 4
,g o - j j<
,
-
i
1 i
- I
1 j i gg,
.
_ g T
~
",.
L_ i i -- V+
' -
u. y 11) !! 1
- - - :,
it, - 2 ,3 .: s C; IE 1__3>>
, , _ 0_
I
. .
al" i1
,
,. !! @
e
- ; i *1I]. 2] '
*
8 7
]< -G l ~ 1.;,l GB I es
=
' 'i '
-
=$
f t. g- i g. n, __ _ _. j]]} .
!^
=
j g-as - il* 1 1_ _ J 1 _ i
.
x 1..m
.- _ _
1.{ 1311
- - - - - --- - --- - - -- - - - - - - - - - , , -
O e gj } jf Yi 7 i j j li!
]vl -
-y; p- ,
,
I lf h)
-
ql )!!
-
-
_ . . _
~ ~ .
"
4AA 1 :tym i l-G).5 yk1 1l t i
^
!
1_%= ij
- ,
r < +... - L 3 1.- f .o 8 . M<' : ..
* '
e t-M E I. k 11 A1115$1L$$3gj {4 N
,ep= l' - -
j . , e y,t'
-
#1 in =1,a s 1 1 _
-
I
*
g
% % - i + l' , 5
< o
+ 7
,
i epusr7 i ni e 1
!1 CD,@lll
'
11 11 1
- -
.
l
- l AAA.a&
Q
.
Meeter ControDer )
'
. - , - ,
'
'
l
-
speed Demand , Q Limiter l Hi 102.6% 1 O to d6% i l l F I r
!, Control For Other MG
[,(A"g -
Opens When Reactor Water In nt < 1 2.5". l
--
~~ . Qoses After Manual Roset
,, Of Rectre. Raabech
~ Oosed When Pump Duch.
Gen. Startup Valve la Fully Open speed T==h-O si . c.. 35 45% O timiter Qosed When reini reedvoie, 28 % -- Signal _
#1 Flow is > 20%
Proportional To Gen.
[TQosed When Rosetor Wat
" " Icel le > 12.5" Speed U
,
' "
i O Speed timitw -- 1 75,%
,
1 g .-
.2 ""3
.. 4
..
3.
ak A
F coeed After Manual Reset
--Of Rectre. Rumbach
,, l 1 Cosed when feedwater Dow 00**d la each line is >20%.
, _- FieldWhen Bk is Gen;_ Cased When Geb- Qosed When Gen. Field Field Bb is --
a cased b Opes e B b is Q osed 2 Comed when level >27%".
,
3 comed when total feedwater g Sow <D0%.
-.
Control Loop with cased When -- Prop. Band, Reset 4 Comed when aD three Gen. Field Bb 3 and Rate condensate pump brothere Is open are elooed.
~ scoop Tube Posetioner I
~ ~
i M
- -
0 l
-
T- 1.t)T- Oti t 11- 2 _ . . _ _ _ _ _ - . -
._ . . . _ - _ . . - _ _ _ _ . - - . - .
. . . . _ ,
.
.
.
-
---
120 - O*
# ** gcS 10 0 -
*
'
p%s.,.4-80 - N.s # s,#' ts0
I h O / 96
: so -
sy etS OD (u CONSTANT PUMP SPEED LINE
.j, 500 b 4 4o -
x NATURAL - CIR CUL ATION I LINE MINIMUM POWER LINE !
' - TYPICAL STARTUP PATH
_lx l 1 - fl O 10 20 " 30 40 50 60 70 80 90 10 0 PERCENT CORE FLOW OPERATING MAP T- L OT- OO40- 8 _. ,__ _____ _ ___ - __ _. . _ _ _ _ _ _ _ _ . _ - - . . _ _ _ _ _ - _ _ ._ .______._ .__ - _ - . . _ _ _.
. Ptga 1 of 1 Revision 2 , 0 3 -16- R 8 I IMrm 's 'N EN FRt.T ; -
.
$1AIION OFF-NORMAi, PROCLDUHt.S INDEX
-
..._-
,2=========,-===========u====.======================.=====
FROCEDMRE HEV. DATE OF LAST NU*tBER NO. T I T f,E DATE SIGNED PERIODIC
,.======.==.:====r= .,==============================BY SUPER. REVIEW
===============,=>====
N-100 2 Pai1ure of a Jet Pump 12/01/87 12/01/P~ _ r.,r: 101 CW ! 0 2 2 Loss of Isolated Phase Bus Cooling 12/01/87 12/01/87 4 /.i r Ejector Discharge fligh 02/25/88 02/25/88' Radiation uu2iol 1
~ Control of Sustained Combustion ,i.n ,th_e_ Off-gas System 12/22/86 12/22/86- 'jN Au4_ _ ,,2 _._qontr_ol Rod Problems _ ,_._03/07/88 03/0</8) _
U N '. % , , _ _ __ Cancelled - - . . . . . . . - . . . . - _ _
. " i t; Cancelled Y
_ _ ____
. _
Oh -i 3 e 4 Control Rod Drive System Problems 03/07/88 03/0'//" _ O!:- 10E Ca nce ?. led _ ON 10 ') '. l'etal Loss of the SP.w, IRM, or APRM Systerts 02/25/88 02/25*
._
ON-110 3 Loss of Primary Containment 03/02/88 03/02/_eE_ ON-ill 3 Loss of Secondary Containrrent 02/25/88 02/25 -l, _ O!. , J 4 Loss of REch
.w. 's > r 02/25/88 02/25,'" . c, or < t t t r: dater CoolTiig 02/25/88 02/25< ~
Runoack (g'f ,)_11 4 _ Loss of Control Enclosure Cooling _ 03/02/88 03/0? f [_ 0:1-116 _ 1, High Reactor Water Condu_ctivity 02/19'/88 12/30/8 ON 1 Loss of TECW 4 02/25/88 02/2',fF'. o."- l_18_ North Stack High Radiati(n 12/31/87 12/31/tf -
.p .
*
.
i ___ __-_ ___ v
__-__ _
.
, .
.
*
l EP-101, Rev. 5 l Page 1 of 2t MJR:mla PIIII.ADSLPHIA Et,ECTRIC COMPANY ' LIMERICK GENERATING STATION /d o# EMERGSNCY PLAN IMPLEMENTING PROCEDURE EP-101 CLASSIFICATION OF EMERGENCIES 1.0 PARTICIPANTS 1.1 Shift Superintendent or designated alternate shall assume the role of Emergency Director and implement this procedure, until relieved.
l 1.2 Plant Manager or designated alternate shall relieve the Emergencp Director, assume the role of Emergenc Director, and continue implementing this procedure,y if necessary. - 2.0 ACTIONS-IMMEDIATE CAUTION:
,
IMPLEMENTATION OF THIS PROCEDURE DOES NOT CONSTITUTE PLAN IMPLEMENTATION OF THE EMERGENCY
.
CAUTION: THE JUDGEMENT OF THE EMERGENCY DIRECTOR IS VITAL IN PROPER CONTROL OF AN EMERGENCY AND TAKES PRECEDENCE OVER GUIDANCE IN THIS EMERGENCY PLAN PROCEDURE
- *
R O
. .. .
e
"
- - ~
. . . . .. [.': y
[ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - - - - - ---
- _ _ - - - - - - - - _ - - - - - , ,
. l , EP-101, Rev. 5 l Page 2 of 20 MJR:mla 2.1 Emergency _ Director shall: 2.1.1.1 Select categories related to station events or conditio s.
Page Number Hazards to Station Operation 6 Environmental 7 Loss of Power 8 Personnel Injury 9 Fire
Radioactive Release 11 Evacuation oC Control Room 12 Damage of Fuel 13 Instrument Failure 14 Scram Failure 15 Bour.dary Degradation /LOCA 16 Unusual Shutdown 18 Loss of Hot or Cold Shutdown Capability Security 19
2.1.2 . Beginning at the indicated page in Appendix EP-101, review the Emergency Action Levels (EALS) for categories selected.
2.1.3 Classify and EALS.the event based on the selected category
~
IF EVENT TRIGGER IS KNOMd TO BE SPURIOUS, DO NOT CLASSIFY EVENT I.E., FALSE HI READING, FALSE CHLORINE MONITOR READINGS ETC.
2.1.4 If the most severe events or conditions are - citssified as an Unusual Event, implement EP-102,
"Unusual Event Response."
2.1.5 If the most severe events or conditions are classified Response."
as an Alert, implement EP-103, "Alert l
.
2.1.6 If the most severe events or conditions arc classified as a Site Emergency, implement EP-104, ,
"Site Emergency Response." !
2.1.7 If the most severe events or conditions are classified .?s a General Emergency, implement EP-105,
"General Emergency Response. "
.
. _ . . . ._ - . . - . , , - - - . . _ - . - - -- . , . _ . ._
.
l EP-101, Rev. 5
,
Page 3 of 20 l MJR:mla 3.0 ACTIONS-FOLLOW-UP 3.1 If the event is classified as an Unusual Event, the Emergency Director shall have a written summary sent to the NRC within twenty-four hours of closeout in accordance with EP-106, Written Summary Notification.
3.2 If event is classified as Alert, Site Emergency, or General Emergency, the Energency Director shall: 3.2.1 Periodically evaluate the event classification as listed on attached Appendix EP-101. Based upon results of corrective action taken to recover from - the emergency situation, escalate or de-escalate the emergency classification. (It is preferable, but not mandatory, to obtain concurrance from the Site Emergency Coordinator and Corporate Headquarters prior tu classification reduction).
The NRC and appropria'.e off-site authorities shall be inforced of the decision to move from one emergency class to the next. As appropriate, agencies or personnel listed in phone lists of * Appendix 1 of EPs 102, 103. 104, and 105 shall be notified within 15 minutes once the emergency level is declared.
3.2.2 Have a written summary sent to the NRC within eight hours of closecut or de-escalation of_the emergency classification in accordance with EP-106, Written Summary Notification.
3.2.3 When the emergency has been controlled and the power plant and auxiliaries have been placed in a safe shutdown condition, only then will a decielon be made as to whether a recovery phase is justified.
Enter the recovery phase after the emergency or accident situation is considered no longer in effect, obtain the concurrence of the Site Emergency Coordinator and the Emergency Support Officer at Corporate Headquarters as required per EP-410, Recovery Phase Implementation. The recover is a departure from an emergency situation.y phase The Site Emergency Coordinator and yourself should evaluate plant operating conditions as well as the in-plant and out-of-plant radiolgical conditions when making this decision. Notifications to the various individuals and agencies that the recovery phase has been implemented is the responsibility of the Site Emergency Coordinator.
, t
. , . . . , . _ , . . . _ _ _ - . - . _ , _ _ ._ , _ _ _ _ _ - . . _ . _. ,, . - , . _ _ , . , . . _ _ _ _ _ _ _
_ _ _ _ _ - _ _ -_
. . * .. , .
l
* EP-101, Rev. S l
Fage 4 of 20 MJR:mla 4.O APPENDICM
'
4.1 EP-101-1 Hazards to Station Operation 4.2 EP-101-2 Environmental I EP-101-3 4.3 Loss of Power 4.4 EP-101-4 Parsonnel Injury 4.5 EP-101-5 Fire 4.6 EP-101-6 Radioactive Release 4.7 EP-101-7 Evacuation of Control Room 4.8 EP-101-8 Damage of Fuel 4.9 EP-101-9 2nstrument Failure 4.10 EP-101-10 Scram Failure 4.11 EP-101-ll Boundary Degradation /LOCA
.
4.12 EP-101-12 Unusual Shutdown 4.15 EP-101-13 Loss of Hot or Cold Shutdown Capability 4.14 EP-101-14 Security 5.0 SUPPORTING INFORMATION 5.1 Purpose The purpose of this procedure is to provide guidelines for classifying an event or condition into one of four emergency clarsifications as described in the Emergency Plan. Additionally this procedure details the method to change from one emergency action level to another and to enter the recovery phase, if applicable.
5.2 Criteria For Use 5.2.1 This procedure shall be implemented whenever the Shift Superintendent becomes aware of conditions which meet or exceed the Emergency Action Levels in EP-101, Classification Tables.
,
.
9
- - - - . _ . _ _ _ . _ _ _ _ _ . _ _ - - - - -
.
.
!
EP-101, Rev. 5 g Page 5 of 20 MJRamla 5.3 Sp;clat Equipment - None 5.4 References 5.4.1 Limerick Generating Station Emergency Plan 5.4.2 NUREG 0654 Criteria for Preparation and Evaluation Rev. 1 of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 5.4.3 EP-102 Unusual Event Response 5.4.4 EP-102 Appendix 1 Unusual Event Notification Message
,
5.4.5 EP-103 Alert Response ' l 5.4.6 EP-103 Appendix 1 Alert Notification Message , 5.4.7 EP-104 Si.te Emergency Response l 5.4.8 EP-104 Appendix 1 Site Emergency Notification 5.4.9 EP-105 General Emergency Response l 5.4.10 EP-105 Appendix 1 General Emergency Notification l 5.4.11 EP-106 Written Summary Notification 5.4.12 EP-410 Recovery Phase Implementation
.
0
-- - -
m - , --4.- ,
.-1 r, . , , - - - +. .e___ - s- y
.
. ... ,
,
..
!
. EP-101, Rev. 5 Appendix EP-101-1 Page 6 of 20 l MJR:mla HAZARDS TO STATION OPERATION _ _ . . _ _ - _ . . _ _ _ _ _ . . . . . . . - . _ _ _ ,_ UNUSUAL EVENT ALERT 1. Aircraft crash on-site or 1. Entry of toxic, flammable gases unusual aircraft activity over the site, or chlorine into the power block with subsequent habitability problem indicated by: 2. Train derailment within the Visual observation, direct site boundary, measurement or notification received by' Control Room.
3. Explosion within or near the site boundary. WHEN BOTH UNITS ARE IN COLD SHUTDONN 4. Nearby or on-site release of 2. Aircraft crash or missile impact potentially harmful quantitles on the Reactor Enclosure, Control of toxic, flammable gas or Enclosure, Turbine Enclosure, chlorine.
Diesel Generator Enclosure or Spray Pond Pump House.
3. Known explosion damage af'ecting l plant operation, l I SITE EMERGENCY GENERAL EMERGENCY 1. Entry of toxic, flammable gases or chlorine into vital areas, i where lack of access constitutes i a reactor safety problem, l indicated by: - A. Shif t Supervision evaluation AND B. Visual observation, direct measurement, notification received by Control Room.
WHEN EITHER UNIT IS NOT IN COLD SHUTDOWN 2. Aircraft crash or missile impact on the Reactor Encloadre, Control ,
!
Enclo'Aure, Turbine Enclosure, * Diesel Generator Enclosure or Spray Pond Pump House.
3. Known explosion damage affecting plant operation.
. D
. . ..
. l . EP-101, Rav. 5 Appendix EP-101-2 l Page 7 of 20 MJR:mla ENVIRONMENTAL
- . . - . -. . . . - _ . - - - - -
. .--. -
- - . -
UNUSUAL EVENT,
. ALERT 1. An actual earthquake detected 1. An actual earthquake detected by the Seismic Monitoring by the Seismic Monitoring System (00C693) at or below System (00C693) beyond the operating basis earthquake operating basis earthquake (.0759 ). (.0759 ).
2. A tornado is observed within 2. Tornado strikes the Reactor or near site boundary. Enclosure, Turbine Enclosure, Spray Pond Pump House, Control 3. A National Weather Service Enclosure or Diesel Generator Enclosure.
hurricane warning is issued for Montgomery County. 3. Sustained high winds greater than 70 mph as indicated on OBC 699.
.
.
.
-
SITE EMERGENCY GENERAL EMERGENCY 1. Sustained high winds greater 1. Earthquake beyond the safe than 90 mph as indicated on shutdown earthquake (.159) OBC 699 if either unit is not or other natural disaster which in Cold Shutdown. causes massive damage ' leading to other General Emergencies. ; 2. An actual earthquake detected I by the seismic Monitoring System (00C693) beyond the safe shutdown earthquake (.159) if either unit is not in Cold i Shutdown.
.
.
.
.
.. .
. . l - EP-101, R3v. 5 Appendix EP-101-3 l Page 8 of 20 MJR:mla LOSS OF POWER
-- _ - -.-. - -
.- _ _ _ .-
UNUSUAL EVENT ALERT
~
1. Loss of all off-site power or N/A loss of all on-site AC power for greater than 60 seconds.
SITE EMERGENCY GENERAL EMERGENC-1. Loss of all on-site AC power N/A and loss of off-site power
,
-
2. Loss of all safety-relat&d DC power as indicated by: a) 250VDC MCC Out of Service ' alarms AND b) IPPA1 D3 Distribution Panel undervoltage alarms
.
*
S _ _ _ _ _ _ _ _ _ _ _ - - - --
-_.
L, . . - . .. : ~ . : .. .. . ..,
-I .
I EP-101, Rev. 5 Appendix EP-101-4 Page 9 of 20 I MJR:mla PERSONNEL INJURY _ . . . _ . _ _ . . _ . _ _ , - - . _ . . . . _ _ _ . _ . . . _ _ _ . _ _ . . . . . . _ _ _ . . . _ . . _ _
,_
UNUSUAL EVENT
- - - ALERT -
-
1. Transportation of contaminated N/A injured individual from site to off-site hospital.
9
.
SITE EMERGENCY GEN'ERAL EMERGENCY
'
N/A N/A i
.
_-- - - . - - - . - .. - -- --
. . . .. . . . . . . . .
l
.
EP-101, RSv. 5 Appendix EP-101-5 l.
t Page 10 of 20 MJH:mla FIRE
, . _ _ . . - _ ._ _ _ . . _ . . . . _ _ _ _ . . _ _ _ .
UNUSUAL Ey,E,N_T ALERT _ l. Fires involving permanent plant 1. Fire which could make an structures within the protected ECCS inop as indicated by area lasting 10 minutes or more observation.
after initial attempts to extinguish li.
.
.
SITE EMERGENCY GENERAL EMERGENCY 1. Fire which makes an ECCS 1. Fire which causes massive' inop and requires or causes . damage leading to other immediate plant shutdown as General Emergencies.
indicated by observation. , l
.i '
l
. . , - - - , .- _, -,. --- - < - -- ,
, ,
. . . . . . . .. .. ,
l , EP-101, RGv. 5 Appendix EP-101-6 RADIOACTIVE RELEASE * Page 11 of 20 l MJR:mla _ _ . _ _ _ _ _ _ _ _ -
,__ , ,, UNUSUAL EVENT ALERT 1. Report indicates liquid 1. Radiological effluents release effluent release exceeds greater than 0.5 mR/hr at site technical specification boundary as indicated by an un-3.11.1.1 or 3.11.1.2. controllable release for greater than 20 minutes with: '
2. Report indicates gaseous a) North stack effluent radiation effluent release exceeds monitor exceeds 1.ON2 uCi/cc or technical specification 3.11.2.1 or 3.11.2.2 or b) South stack effluent radiation 3.11.2.3 monitor exceeds 1.2N2 uCi/cc.
, .
.
.
SITE EMERGENCY GENERAL' EMERGENCY 1. Radiological effluent release 1. Radiological effluent release greater than 50 mR/hr at site greater than 500 mR/hr at site boundary as indicated by boundary as indicated by an i an uncontrollable release for uncontrollable release for greater than 20 minutes with: greater than 20 minutes with a) North stac:t effluent radiation a) North stack effluent radiation monitor exceeds 1.0 uCi/cc. monitor exceeds 10 uCi/cc.
2. Projected whole body dose 2. Projected whole body dose greater than .1 rem or thyroid greater than 1 Rem or thyroid dose greater than .5 Rem at or dose greater than 5 Rem at or beyond the site boundary over beyond the site boundary over course of the event utilizing course of the event utilizing RMMS procedure calculating RMMS procedure calculating offsite doses, offsite doses.
.
. . --- , - _ - - - _, , , - , -
~..
.
. . . . .
.. -
. :
l - EP-101, Rav. 5
*
Appendix EP-101-7 Page 12 of 20 l MJR:tala EVACUATION OF CONTROL ROOM
..._ ___. . . . . _ _ . . . . . __.- __ __.. - .. _ . ...-. _. .... _ _ ..__. _ __
UNUSUAL EVENT ALERT N/A 1.. Evacuation of Control Room anticipated or required with control established ..t. remote shutdown panel.
9
. *
SITE EMERGENCY GENERAL EMERGENCY " 1. Evacuation of Contol Room N/A and control of shutdown systems not established from remote shutdown panel in 15 minutes.
.
9
.
e
9
.
- - ..
I
EP-101, Rev. 5 Appendix EP-101-8 l Page 13 of 20 DAMAGE OF FUEL MJR:mla _ __ _ _ .. _ ._.__ _ . _ _ . . _ _ _ . - . _ _ . - . - _ _ _ _ _ . UNUSUAL EVENT ALERT 1. Steam Jet Air Ejector Discharge 1. Steam Jet Air Ejector Discharge radiation monitor exceeds radiation monitor exceeds 2. lP4 mP,'h r . 2.lPS mR/hr 2. Steam Jet Air Ejector Discharge 2. I-131 dose equivalent in the radiation monitor has an un- reactor coolant exceeds 300 uC1/g expected increase of 4000 mR/hr from sample and main steam over 30 minutes. line high-high radiation with resultant scram.
3. 1-131 dose equivalent in the 3. Spent fuel damage res..lting in a reactor coolant exceeds refueling floor area ventilation i 0.2 uCi/g from sample exhaust monitor alarm.
analysis. ,
!
4. Containment Post LOCA Radiation : Monitors greater than IP2 R/hr. l _.
SITE EMERGENCY GENERAL EMERGENCY 1. Major damage to spent fuel: 1. Containment Post LOCA Radiation a) Observation of major damage Monitors greater than IP4 R/hr.
to spent fuel QR b) Water loss below fuel level in spent fuel pool.
2. Containment Post LOCA Radiation Monitors greater than IP3 R/hr.
~
. _ _ _ _ _ _ _ _
e .
. .
. . . .
)
' l EP-101, Rev. 5 Appendix EP-101-9 Page 14 of 20 l MJR:mla INSTRUMENT FAILURE _ _ . . _ . . . _ _ . _..._.... . __
- _ _ . _ .
-- .._ _ _
UNUSJAL EVENT ALERT 1. Complete loss of all Main 1. Loss of all-annunciators.
Control Room communication equipment.
2. Significant loss of assesement capability in the Main Control Room as indicated by: Simultaneous loss of the - plant process computer and the ERPDS for a period greater than 24 hours.
.
.a
.
SITE EMERGENCY GENERAL EMERGENCY 1. All annunciators lost and N/A plant transient initiated or in progress.
.
. *
_ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . - - . - - -
.,. . .. ., -
!
i EP-101, Rev. 5 Appendix EP-101-10 Page 15 of 20 I SCRAM FAILURE
-- - . . . . . - . . _ . _
_ _ _ _ _ _ , UNUSUAL EVENT ALERT
.
N/A 1. Failure of the Reactor protection system to automatically initiate and complete a scram AND Scram fails to bring Reactor suberitical as indicated by APRM's greater than 4%, one - minute after scram initiates.
.
.
SITE EMERGENCY GENERAL EMERGENCY 1. Transient requiring standby 1. Transient requiring standby liquid control system to liquid control system to initiate with failure to initiate with failure to scram. Failure to Scram scram is indicated by APRM'S AND greater than 4% one minute after a scram Reactor does not become initiates, sub-critical as indicated by APRM'S greater than 40 10 minutes after scram initiates.
f
.
.
)
-
_
)
. *,. * *
. e.
. l EP-101, Rev. 5 Appendix EP-101-11 Page 16 of *.0 I MJR:mla BOUNDARY DEGRADATION /LOCA PACE 1 of 2
- . . . _ -. .-. - _ - - . .. ,_ UNUSUAL EVENT ALERT 1. Failure of a main steam relief 1. Scram with small leak as valve or ADS valve to close indicated by following reduction of a) Scram alarm applicable pressure. AND b) '.teactor level less than -129" , AND c) Containment pressure greater than 1.68 psig and pressure is increasing.
2. Reactor coolant leak rate exceeds 60 gpm total leakage averaged over any 24 hour period as indicate by surveillance test report. . 3. High airborne contamination in the Reactor Enclosure as indicated by: . 2. Reactor coolant leak rate a) Reactor Enclosure vent exhaust exceeds 30 gpm total RAD monitor A/B or C/D leakage as indicated by Hi-Ri ALARM on 10C800 (20C800) OR
~~~
surviellance test report.
b) 1000 fold increase of airborne radiation in a major area of the reactor enclosure as determined by health physics.
. S
.
e
*
- - - - ---
_ __ _ _ , , _ - . .
* .. . .
,
, ,
' l EP-101, Rev. 5 Appendix EP-101-11 l Page 17 of 20 MJRamla BOUNDARY DEGRADATION /LOCA PAGE 2 of 2
. _ _ .
SITE EMERGENCY GENERAL EMERGENCY 1. Scram with LOCA as 1, Scram with LOCA & no ECCS as indicated by: indicated by: a) Scram alarm a) Scram alarm AND AND b) Reactor level.less than -129" b) neactor level less than -129" AND AND c) Containment pressure greater c) F.tth r e to bring Reactor level than 10 psig ate.-e -129" after 3 minutes
. AND ,
d) Containraent pressure greater than 20 psig 2. Main steam line break outside 2. Scram with LOCA & Containmt si, containment without isolation Failure as indicated by: as indicated by: a) Scram with Reactor level less
'than -129" AND a) High Main Steam Line Flow (108.7 psid)
AND b) Reactor Enclosure Vent Exhaust Rad Mcnitor A/B or C/D b) Bigh Steam Tunnel Temp Hi-Hi alarm on 10C800 (20C800)
(165 deg F)
AND c) Main Steam Line Low Pressure (756 psig)
.
.
4
, -- - , -- -- ., -
,,,
'O '
. * * , ,, -
l EP-101, Rev. 5 Appendix EP-101-12 Page 18 of 20 l MJR:mla UNUSUAL SHUTDOWN
--
_
--
UNUSUAL EVENT ALERT 1. Controlled shutdown due to N/A failure to meet limiting ' condition of operation.
2. Shutdown other than normal controlled shutdown AND for the purkJse of placing the
.
the plant in a safer condition.
3. Cooldown rate exceeds technical specification limits.
. SITE EMERGENCY GENERAL EMERGENCY N/- N/A
.
.
.
.
.
, -
~.--.-,.v. ,m.. pe ,
' . . . . .
l o . . ~ ; l EP-101, Rev. 5 .' Appendix EP-101-13 ,
'
l Page 19 of 20 M. ~1a LOSS OF HOT OR COLD SHUTDOWN CAPABILITY UNUSUAL EVENT ALERT N/A 1. Complete loss of the ability to establish and maintain the plant in a Cold Shutdown condition.
Symptomized by: a) Inability to establish Reactor Coolant Temperature of less than 200 degree.s F in a timely manner with the mode switch in Shutdown OR b) Loss of all means of Primary and Alternate Decay Heat Removal when shutdown such that Reactor Coolant Temperature cannot be maintained below 200 - degrees F.
. SITE EMERGENCY GENERAL EMERGENCY 1. Actual inability to reduce 1. Inability to reduce Reactor Reactor Coolant System Coolant System Temperature with Temperature while the Plant the potential for the release of is Shutdown. Indicated by: large amounts of Radioactivity.
Indicated by: a) Reactor mode switch'in Shutdown AND a) Inability to maintain Reactor
,
Level greater TAF b) Reactor Coolant Temperature OR greater than 200 degrees F substantial Fuel Damage has and rising occurred AND ,AND b) Inability to maintain c) Suppression Pool Temperature Suppression Pool Temperature greater than 120 degrees F below the upper limit of the and rising.
Heat Capacity Curve (SP/T-21 on T102 Procedure.
.
- , ww+-,
' .
.. ,
. : .
'
.
..
~
- EP-101, Rev. 5 Appendix EP-101-14 Page 20 of 20 MJR:mla SECURITY
_ _ . . . . . . _ _ . . . _ l UNUSUAL EVENT ALERT 1. Security threat 1. Ongol_ng security compromise QR Attempted entry QR Attempted sabotage as illustrated by:
.
Event 1 - Sabotage or Bomb Threat Event 2 - Intrusion and Attack Threat Event 7 - Suspected Intrusion Event 8 - Actual Intrusion Event 9 - Suspected Bomb or Sabotage Device Discovered Event 15 - Guard Strike Event 16 - Onsite Hostage Situation
.
I __ l SITE EMERGENCY GENERAL EMERGENCY 1. Imminent loss of physical 1. Loss of physical control of control of the plant. the facility. , i i
.
.
-m -g e- - e-ww -
r- r 4
~ - '
-
. . . .
~\
__
- - ., _ --
- - - .
., _.
- . . _ _ _ . i
- _ _ ___
- .g. - - - .-. .
- - - - ___ ,
- .
- - _.
-- -- n
=- .,87
:- ? lu,lGz0
, - - -
.
- -- :
,
__ _- - _ _ = . - -- _ - - .. ___
,_
~, SAFETY LIMITS A .) LIM **IN", SAr[ y- SYSTEw SEUING$
- ----
SECTION . FAGE _ 2.1 SAcETY LIMI~5 THERMAL POWER, Low Pressure or Low Flow..................... 2- 1 _
"-
THERMAL POWER,,High Pressure and High Flow........,......... 2-1 Reactor Coolant System Pressurt............................. 2-1 Reactor Vessel Water Level.................................. 2-2
.
2.2 LIM} TING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints......... 2-3 Table 2.2.1-1 Rea: tor Protection System Instrumentation Setocints...... . ...... 2-4 BASES
,
2.1 SACETY LIMITS THE RMAL POWE R , Low P re s s ure o r Low F1 ow. . . . . . . . . . . . . . . . . . . . . B 2-1 THERMAL POWER, Hign Pressure and High Flow.................. B 2-2 Left Intentionally B1ank.................................... B 2-3 Left Intentionally B1ank.................................... B 2-t-l ; Re a cto r Cool ant System P re s s ure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-5 Reactor Vessel Water Leve1.................................. B 2-5 1 l 2.2 LIMITING SAKETY SYSTEM SETTINGS Reactor Protection Systes Instrumentation Setpoints. . . . . . . . . . B 2-6 L1HERICK - UNIT 1 iv Amendment No. 7
- f.ljG 1419M l l
l
.
- ., , , , _ . , ,. . _ , ,
- - - - . _ - -
.- -
.. _ - _
.
. - -
. . .- -
- .* *
_ . - . ,
- - - . ._.
_
-
= _ .. _
_ . _. , ww M- e
:- . , -
_
-
_ 3875101020 _ _ . . - -- _ _ _ m _
.
_ , _
-
..
.
-
. - -
-- -
.
INDEX LIMITINO CON 0!T10N5 JOR OPERATION AND SURVEILLANCE REQUIRENfMTS SEC*!ON PA E _- 3/4.0 APPLICAE M17Y.............$ ............. ...... .... .... 3/4 0-1
~
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN...................... . . . . . . . . . . . . . . . . . 3/4 1 1 3/4.1.2 REACTIVITY ANOMALIE5..................................... 3/4 1-2
3/4.1.3 CONTROL ROD 5
!
l
Control Rod 0peracility.................................. 3/4 1-3 Control Rod Maxime Scram Insertion Times................ 3/4 1-6 l Control Rod Average Scran Insertion Times................ 3/41-7 l
Four Control Rod Group Scram Insertion Times............. 3/4 1-8 l l
'
C o nt ro l Ro d Sc ram Accoul a tors . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 Control Rod Drive Coupling............................... 3/4 1-11 Control Rod Position Indication.......................... 3/4 1-13 Control Rod Drive Housing Support........................ 3/4 1-15 3/4.1.4 CONTROL RCD PROGRAM CONTROLS Rod Worth Minim 12er...................................... 3/4 1-16 Rod Sequence Control Systas.............................. 3/4 1-17 Rod Block Monitor........................................ 3/4 1-18 3/4.1.5 STAND 8Y LIQUID CONTROL SY$TDl............................ 3/4 1-19 Figure 3.1.5-1 Sodie Pentaborate solution Temperature / Concentration Requirements........................ 3/4 1-21 Figure 3.1.5-2 Sedim Pentaborate Solutten Vol me/ Conc e ntrati ca Requi rements. . . 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMIT $ 3/4.2.1 AVERAGE PLAMAR LINEAA. HEAT GDIERATION RATE...............
~
3/4 2 1 Figure 3.2.1-1 Maximus Average Planar Linear Heat Generation Rate (MAPLMGR) Versus Average Planar Emposure Initial Care Fuel Types PSCIt274............ 3/4 2-2 ,
-
l LIMERICK - UNIT 1 v ! fd)5 8196 l
- _ - _ _ _ _ _ . _ _ _ _ _ _ .-. - - - . _ - - _ . . . . . . . , - - , . - -- , - , - . .-,.__,\
~
"
. _ _ . - - - -
_ _ _
.~* - - --
,, ., ,
.--
c.- _ _ .
. ._..-- _ . - ' ~ -
--
- -. .___
_ _ _3 8_7.5101~0 20
'
_--- _ ____ _
..
~~~ ~ ' ~
(
-
__ - .-
-
. _3 -
-
_._
.
_ _ _-
._
_ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANl PAGE l SECTION l POWER DISTRIBUTION LIMITS (Continued) - Figure 3.2.1-2 Maximum Average Planar Linear Heat-Generation Rate (MAPLHGR) Versus ~ l Average Planar Exposure Initial
.
Core F uel Types P8CIB248. . . . . . . . . . . 3/4 2-3 Figure 3.2.1-3 Maximus Average Planar Linear Heat Generation Rate (MAPLHGR) Versus f Average Planar Exposure Initial Core fuel Types P8CI8163........... 3/4 2-4 l Figure 3.2.1-4 Maximus Average Planar Linear Heat ' Generation Rate (MAPLHGR) Versus * Average Planar Exposure Initial l Core Fuel Types PSCIB094........... 3/4 2-5 j l l Figure 3.2.1-5 Maximum Average Planar Linear Heat l Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial . 3/42-6-
'
Core Fuel Types P8CIB071........... Figure 3.2.1-6 Maximus Ave' rage Planar Linear Heat , Generation Rate (MAPLHGR) Versus Average Planar Exposure For Fuel Type BC320A (GEBX8EB).............. 3/4 2-6a 3/4 2'.2 APRM SETP0!NTS.......................................... 3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RATI0............................ 3/4 2-8 Table 3.2.3-1 Deleted l Figure 3.2.3-la Minious Critical Power Ratio (MCPR) Versus (P8X8R/SP8X8R Fuel). . . . . . . . . . . . . 3/42-10l Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR) Versus (GE8X8E8 Fue1). . . . . . . . . . . . . . . . . . 3/42-10al l 3/4 2-11 Figure 3.2.3-2 yFactor.............................. 3/4 2-12 3/4.2.4 LINEAR HEAT GENERATION RATE............................. 3 /4. 3 IM$TRtBENTATION 3/4 3-1 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRupENTAT!0N............... Table 3.3.1-1 Reactor Protection System I nstrumentati on. . . . . . . . . . . . . . . . . . . . . 3/4 3-2 Table 3.3.1-2 Reactor Protection Systes . Response Times...................... 3/4 3-6 I Table 4.3.1.1-1 Ructor Protection Systaa Instrumentation Surveillance Requirements...................... 3/4 3-7 vi Apencoent No. 7 LIMERICK UNIT - 1 AU614 IM7
- - - . -. . - .- - -. . -- . _ . . . . _ . - _ - _ - - . _ -
- "-
_ _ _ .. _ _ _ ._ -
-
-- - -
,_
*
-
- -
. -
, -)
. _ . -
- -
- .- . _ _. . - -
--
-. --
. _ _
.
_
.IN?EF . . _
~
. - __
-
- LIMITING CONDITIONS FOR ODERATION AND SURVEILLANCE REOUIREMENTS ~ _ _ _
' '
-
SECTION ,,,. PAGE 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION..................... 3/4-3-9 Table 3.3.2-1 Isolation Actuation Instrumentation..................... 3/4 3-11 _..
'
Table 3.3.2-2 Isolatio'n Actu"ation - Instrumentation Setpoints........... 3/4 3-18 _ Table 3.3.2-3 Isolation System Instrumen-tation' Response Time................ 3/4 3-23 Table 4.3.2.1-1 Isolat' ion Actuation Instrumen-tation.' Surveillance Requirements...................... 3/4 3-27 3/4.3.3 EMERGENCY CORE COOLING SYSTEA ACTUATION . INSTRUMENTATION......................................... 3/4 3-32 Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation........... 3/4 3-33
'
Table 3.3.3-2 Emergency Core Coo' ling System Actuation Instrumentation Setpoints........................... 3/4 3-37 Table 3.3.3-3 Emergency Core' Cooling System
. Response Times...................... 3/4 3-39 Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation -
Surveillance Requi rements. . . . . . . . . 3/4 3-40 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION
.
ATVS Recirculation Pue:p Trip System Instrumentation..... 3/4 3-42
,
Table 3.3.4.1-1 ATWS Recirculation Pump Trip ! System Instrumentation............ 3/4 3-43 l I Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints......................... 3/4 3-44 Table 4.3.4.1-1 ATWS Recirculation Pump Trip . Instrumentation Surveillance Requirements...................... 3/4 3-45 End-of-Cycle Recirculation Pump Trip System
*
l Instrumentation......................................... 3/4 3-46
' ' -
LIMERICK - UNIT 1 vii k% 8 IN5 g
.
* O- enum * ye a, , m egg enee .. - . .@ Sib.p.g ee. e a de geGeee
- - . . . . . . --.. - - - , - -
...g . .. . . . . .
l l
. ,. ..., , . ,- -- .__ - - . , , - .
' '
, . - -
~ . ~ -
. . , . . . . _
_
-
-' - - - ~ ~
- _. ... . _ ,
__. ..
- _ _ . _
_
-- - - .- . - .
-
-- ._.
- __ __- - _
. . . __ _
__ _
~~
~
."
- -
INDYX' _
.
l
- --
__ __ _ ,
,
_
~
i LIMITING CONDITIONS FOR 0PERATION AND_ SURVEIL' LANCE REQUDIEMENTS
-- -
SECTION PAGE INSTRUMENTATION Continued)
...
Table 3.3.4.2-1 End-of-Cycle Recirculation Pump _- Trip Syster Instrumentation....... 3/4 3-48_ _.
Table 3.3.4. 2-2 End-of.-Cycle Recirculation Pump ..
.
- Trip Setpoints.................... 3/4 3-49
:
Table 3.3.4.2-3 End-Of-Cycle Recirculation Pump Trip System Response Time......... 3/A 3-50
'
Table 4.3.4.2.1-1 End-Of-Cycle Recirculation ! Pump Trip System Surveillance Requirements.................... 3/4 3-51
,
l 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION......................................... 3/4 3-52 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumenta-tion................................. 3/4 3-53 Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints........................... 3/4 3-55 ,, Table 4.3.5'.1-1 R'eactor Core Isolation Cooling
*
System Actuation In.strumentation Surveillance Requi rements. . . . . . . . . . 3/4 3-56 . 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION....................... 3/4 3-57 l Table 3.3.6-1 Control Rod Block Instrumenta-tion.............. .................. 3/4 3-58 Table 3.3.6-2 Control Rod Block Instrumenta- ! tion Setpoints....................... 3/4 5-60 (
Table 4.3.6-1 Control Rod Block Instrument 4- l 3/4 3-61
'
tion Surveillance Requirements. . . . . . . 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.............. ..... 3/4 3-63 . Table 3.3.7.1-1 Radiation Monitoring Instrumentation................... 3/4 3-64
.
!
LIMERICK - UNIT 1 viii ~ US 8 I!6 g
. . __.. . . _ . . . . _ . . . _ _ . . . . . . .. .... -.. _.... ___
.
. - - - - - - - - - - - -
. . . . . . . . . . . . . . _ , . . . . . . . . . -
.-- - , - .-. , , ,, - - - . _ . - - - _ - . _ . . . . . - - , -. - . .
- _
. -
_- - - - . - - - - - l .
.
-
.- - , - _
--
. . . .
_ - - - - > _ -
- - - -
.
-- 1
' - - - ~ -- - - -
_
! ~ ~~ -
_ - - INDEX _
~~~~
TIM! TING CONDITIONS FOR OPERATION AND_ SURVEIL 1.ANCE- RE0VIRE>iERTS -
--- - -- - - . -~ ~
__ SECTJ ON_ _ _
._
-
.
PAGE
- -- ---
INSTRUMENTATION (Continued)
-
-
Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance e Requirements...................... 3/4 3-66 3/4 3-68
* ' ~~
Seismic Monitoring Instrumentation..,................... .
- -
..
Table 3.3.7.2-1 Seismic Monitoring Instrumentation.................... 3/4 3-69-Table 4.3.7.2-1 Seismic Monitoring - Instrumentation Surveillance Requirements....................... 3/4 3-71 .
' Meteorological Monitoring Instrumentation. . . . . . . . . . . . . . . 3/4 3-73 ~ Table 3.3.7.3-1 Meteorological Monitoring .
Instrumentation................... 3/4 3-74 Table 4.3.7.3-1 Meteorological +lonitoring Instrumentation'5urveillance . Re q u i reme nts . . . . . ... . . . . . . . . . . . . . . . 3/4 3-75 Remote Shutdown System Instrumentation and Controls... .. 3/4 3-76
~ Table 3.3.7.4-1 -Remote Shutdown Systen' ' .
Instrumentation and Controls...... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System - Instrumentation Surveillance Re q u i rem nts . . . . . . . . . . . . . . . . . . . . . . 3/4 3-83 Acci der.t Monitoring Instrumentation. . . . . . . . . . . . . . . . . . . . . 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation............................ 3/4 3-85
~ ' Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements.... 3/4 3-87 Source Range Monitors................................... 3/4 3-88 Traversing In-Core Probe System......................... 3/4 3-89 Chlorine Detection System............................... 3/4 3-90
_ Toxic Gas Detection System.............................. 3/4 3-91 ! Fire Detection Instrumentation.......................... 3/4 3-92 LIMERICK - UNIT 1 ix , , , ,
, , . . s
-
. ,
...
j l
)
. - - . - - - , . .. . . _ . . . . -.--
-- . . . _ . . . . . .
. - . . . . ..: ' '
*
*
. . ...: . ....:.---.------------ = - - - -
-- .
._ . .~ . . . - , - - . ~ , _ _ _
. -. _. ,
-
-
- .
e ---
- _ ~ E __.- . ~ -
._ _. . _ _
, -
.-- - -_ _
.
_. .- -- _
._._
__.
_
- -
. . _
_
. ___ . _ - _ _._ - -
_ _ _
-
- TNDEk- - ~ -
.
.
-
_ LIMITING CONDITIONS FOR OPERATION AND SU_RVEILLANCE REQUIREMENTS , j
~~
- '
SECTION PAGE
-
-
-
INSTRUMENTATION (Continued) Table 3.3.7.9-1 Fire Dstection Instrumentation.... 3/4 3-93 _ _
~
Loose-Part Detection System.......*...................... ,3/4 3-97
'
Radioactive Liquid Effluent Monitoring Instrumen- : * tation.................................................. 3/4 3-98 Table 3.3.7.11-1 Radioactive Liquid Effluent . Monitoring Instrumentation....... 3/4 3-99
< Table 4.3.7.11-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements........ 3/4 3-101 Radioactive Gaseous Effluent Monitoring Instrumen-tation.................................................. 3/4 3-103 Table 3.3.7.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation....... 3/4 3-104 Table 4.3.7.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements........ ' 3/4 3-107 ,
3/4.3.8 TURSINE OVEP. SPEED PROTECTION SYSTEM...................... 2/4 3-110 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUAi10N 3/4 3-112
~
I NST RUME NTATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.... 3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip
. System Actuation Instrumen-tation Setpoints.................... 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip I System Actuation Instrumenta-tion Surveillance Require-ments............................. 3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM -
3/4. 4.1 RECIRCULATION SYSTEM
~ ~
Recirculation Loops..................................... 3/4 4-1 , i LIHERICK - UNIT 1 x 'yg E W3 s l
:
._.. . _
--
= .
._. _ . . .. . . - . _ . . . . _ . . . . _ _ . _ . _ . _ _ . . . . . . _ . _ . _
. . . - . . . . . . .
-- - - -, _ _. _ _ , _ _ _ _ _I
_.
.._
- ~ ~'. . _
_
---.
___
- - - - __
~ _. ___ _ _ I
--
-- - _ __ . __ _..- -
.
----_
.JrOEX _ --
- - -
.
~ ~
~~~~~ ' ' ~ -
-LIMITING CONDITIONS FOR ODERATION AND SUR[EILLANCE REQUIREMENTS.__ --
__ _ __
- -~~~~~ 7 SECTION .- - -
_ PAGE __ _
-- REACTOR COOLANT SYSTEM (Continued)
-
Figure 3.4.1.1-1 Thermal Power versus Core F1ow............................. 3/4 4-3
,
et Pumps.....'.... ..._.................................. 3/4 4-4
~~
Recirculation Pumps..................................... 3/4 4-5 Idl e Reci rcu1, ation Loop Startup. . . . . . . . . . . . . . . . . . . . . . . . .. 3/4 4-6 2/4.4.2 SAFETY / RELIEF VALVES.................................... ' 3/4'4-7
'
3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................... . 3/4 4-8 Operational Leakage..................................... -3/44-9 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves................... 3/4 4-11 3/4.4.4 CHEMISTRY............................................... 3/4 4-12 Table 3.4.4-1 Reactor Coolant System Chemistry - - Limits..............................
,
.
3/4 4-14.
3/4.4.5 SPECIFIC ACTIVITY....................................... 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity
-
Sample and Analysis Program. . . . . . . . . 3/4 4-17 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-18 :
,
Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel l Metal Temperature Vs. Reactor , Ve s s el P re s s ure. . . . . . . . . . . . . . . . . . 3/4 4-20 ' Table 4.4.6.1.3-1 Reactor Vessel Material Surveil- .
-ance Program - Withdrawal l Schedule........................ 3/4 4-21 Reactor Steam Dome...................................... 3/4 4-22 l
3/4.4.7 HAIN STEAM LINE ISOLAT. ION VALVES. . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-23
,
l 3/4.4.8 STRUCTURAL INTEGRITY..~ ................................. 3/4 4-24
~~
LIMERICK - UNIT 1 xi lE S ISE t l l __
)
,
l
. . . . - . . . - . . . . . . . . ... . . _ --.. _ -- - - - .-..- - - .
l
. - -
. -- . - - _ . - - .. .. --.
_ _ _ _ _ _ _
. .
- . ,_.
-
-- - .u - -. . _.
,
. _ _
-
.- - ._ .
-- - .
.
- -
-
- _ . .-. __.__ . ,
-- . .- ___ . . _ _ _._
- -_ ___
, -- . _
-
- _ _
- -
-
~
2_ INDEX
-
~_ .
_ , _
.-
.
- . _ _ . _ .
"
~
-
~~
LIMITING CONDITIONS FOR ODERATION AND SURVEILLhNCE REDUIREMINTS
--
.. ._
~~ ~~ ~ '-
--
SECTION
-
BGJ 3/4.4.9 RESIDUAL HEAT REMOVAL , Hot Shutdown............................................ 3/4 4-25 . Co(dShutdown................:.......................... 3/4 4-26. ,
.
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - 0PERATING........................................ 3/4 5-1 3/4.5.2 ECCS - SHUTD0WN......................................... 3/4 5-6-
.
3/4.5.3 SUPPRESSION CHAMBER..................................... 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS , 3/4.6.1 PRIMARY CONTAINMENT Primary Contai nment Integri ty. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/46-1 Pri mary . Contai nment Leakage. . . . . . . . . . ., . . . . . . . . . . . . . . . . . 3/4 6-2 Primary Contai nment Air Lock. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 i MSIV Leakage Control System,............................ 3/4 6-7 ~ , {
~
Primary Containment St'ructural Integrity. . . . . . . . . . . . . . . . 3/4 6-8 Drywell and Suppression Chamber Internal' Pressure....... 3/4 6-9 , Drywell Ave rage Ai r Tempe rature. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-10 Drywell and Suppression Chamber Purge System............ 3/4 6-11 3/4.6.2 DEPRESSURIZATION SYSTEMS S uppre s s i on Chambe r. . . . . . . . . . . . . . . . . . . . . . . . . ........... 3/4 6-12 Suppression Pool and Drywell Spray. . . . . . . . . . . . . . . . . . . . . . 3/4 6-15 Suppression Pool Cooling................................ 3/4 6-16
,
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . 3/4 6-17
. .
Table 3.6.3-1 Primary Containment Isolation Va1ves.............................. 3/4 6-19
-
,.
I LIMERICK - UNIT 1 xii AV6 8 DI3 %
.-
..
.
. . - . .. - - . . - .
- - .. . . . - . . _ . . .- -. ._.
. . . . . . . . , . . . . . . . . . .. ~=....:-...-.~-.- -- - , , _ - _ - . .- .,, , . . . . . . , . . - - . . . - - . - _ . - - , . . ,. . . _ . . . . - - - . . _ . - _ ~ . _ . .
- - .
-
-
...
-
~
_ , - ._.-
. ,,_ - .. - -. j
__ .-. - --
--
. ---
--
_ _. 1
- .-
u.oD. . .-.
.-
LIMITING TONDITIONS FOR OPERATION 7ND SURVEItLANCE REhJIREMENTS - l
-.
?._ PAGE
--I
..
SECTION -_
- ~ CONT INMENT SY5TEMS (Continued)
3/4.6.4 VACUUM PELIEF
'
-
Suppression Chamber - Drywell Vacuum Breakers. . . . . . . . . . . ~ . 3/4 6-44
'
b 3/4.6.5 SECONDARY CONTAINMENT
. Reactor Enclosure Secondary Containment Integrity. . . . .. . 3/4 6-46 ,
Ref ueling Area Secondary Containmen;t Integrity. . . . . . . . . . 3/4 6-47-Reactor Enclosure Secendary Containment %utomatic Isolation Va1ves........................................ 3/4 6-48 Table 3.6.5.2.1-1 Reactor Enclosure Secondary ; Containment Ventilation l System Automatic Isolation Va1ves.......................... 3/4 6-49 I Refueling Area Secondary Containment ' Automatic Isolation Va1ves........................................ 3/4 6-50 Table 3.6.5.2.2-1 Refut"ing Area Secondary Contain- l ment Ventilation System Automatic- ' - Isolation Valves................ 3/4 6-51 Standby Gas Treatment System............ .............. 3/4 6-52
-
Reactor Encl osure Reci rculation System. . . . . . . . . . . . . . . . . . 3/4 6-55
3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Primary Containment Hydrogen Recombiner Systems. . . . . . . . . 3/4 6-57 D rywel l Hyd ro ge n Mixi ng System. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-58 Drywell and Suppression Chamber Oxygen Concentration.... 3/4 6-59 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS . Residual Heat Removal Service Water System. . . . . . . . . . . . . . 3/4 7-1 - Emergency Service Water System.......................... , 3/4 7-3 Ultimate Heat Sink...................................... 3/4 7-5
'AD6 81935
~~
g LIMERICK - UNIT 1 xiii
. . - - - - .
.. .. .. -.-
.. . .. . . . . . . . . . . . . . . . . .
- . . .
. . , . .. .. . . . . . . . . . . .
.)- . . - . . . - . , . . . . . . . . . . . . . _ . .
, , , _ . -- ,,. . - . . _ . . .
.- . ..-. = . l
_ - _
-
.
- -
-
--
_ , _ , . _ , _
-
J
; - - _ __
_ ; _ _ .
-
_ _.
. __ .- . _ . - ,
-_ _ . . , ,
_ - - - - ._ . _ _ . . . _ ._
- _- ___ _
. _ _._ _. __.
_.
j
-. - .
* '
-
_ _ _ _ _ _ _ _ INDEX _ _ .. -- ._ _ __. .__ _
(_
LIMITING COND!TIONS- FOR ODERATION AND SURVEli. LANCE REOUIREMENTS ' ~~ -
. - . -_ -
SECTION -
._. PAGE PLANT SYSTEMS (Continued)
3/4.7.2
.
CONTROL ROOM EMERGENCY. FRESH AIR SUPPLY SYSTEM.......... 3/4 7-6 _ _.
3/4.7.3 REACTOR CORE I.JLATION'C00 LING SYSTEM....T.............. 3/4 7-9
'
3/4.7.4 SNUBBERS................................................ 3/4 7-11 Figure 4.7.4-1 Sa'mple Plan 2) For Snubber Functional Test.................... . 3/4 7-16 3/4.7.5 SEALED SOURCF CONTAMINATION............................. 3/4 7-17 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System......'..................... 3/4 7-19 Spray and/or Sprinkler Systems.......................... 3/4 7-22 CO 2 Systems............................................. 3/4 7-24 Halon Systems........................................... 3/4 7-25 Fire Hose Stations...................................... 3/4 7-26.
.. Table 3.7.6.5-1 Fire Hose Stations................ 3/4 7-27 Yard Fi re Hydrants and flydrant Hose Houses. . . . . . . . . . . . . . 3/4 7-29
.
Table 3.7.6.6-1 Yard Fire Hydrants and Associated Hydrant Hose Houses............... 3/4 7-30 3/4.7.7 FIRE RATED ASSEMBLIES................................... 3/4 7-31 3/4.8 ELECTP.ICAL POWER SYSTEMS l 3/4.8.1 A.C. SOURCES i A.C. Sources - Operating................................ 3/4 8-1 l
Table 4.8.1.1.2-1.D'iesel Generator Test Schedule........................ 3/4 8-8 ;
* 1
'
A.C. Sources - Shutdown...................'.............. 3/4 8-9 3/4.8.2 D.C. SOURCES . D.C. Sources - Operating................................ 3/4 8-10 8 _ LIMERICK - UNIT 1 xiv MB B B33 t
. __ .. _ - . . . . . . __ - - . . - . - . - - -
..... ... ~ . . - - . .
..
_ .. . . . -
. _ . _ _ __ _ _. , . _ _ _ _ . ____ . . _
__
. . .
_ .: -- - -
-.
_ _ _ _ _ -
-
--
..-
. . - . .
,
,,
.
._ .. .. .
-- - .
..
.
. . . . . .
_- ,
,
.yy -
- -.
:
.. -_. .-
~
* -- --
LIM' TING C055!TIONS FOR OPERATION AND SURVEILLANCE _REOUIREMENTS -
-
-.
SECTION
~
.I PAGE
-
-
ELECTRICAL' POWER"SYSTEMS (Co:.'.inued) Table J.8.2.1-1 Battery Surveillance . Requirements...................... 3/4 6-13 - _ _ _ .
,- -
D . C . S o u rc e s - S hirtd own . . . . . . . . . . . . . . . . .T. . . . . . . . . . . . . . . 3/4 8-14 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS ,
-
'
Olstribution - Operating................................ 3/4 6-15
'
D i s t r i b uti o n - S h utfiown ! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-18 3/4.E.4 E:.ECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overcurrent Protective Devices...................................; 3/4 8-21 Table 3.6.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective Devices........ .................. 3/4 8-23 Motor-operated Valves Thermal Overload Protection..... .. 3/4 8-27 Reactor Protection Sys'tec Electric Power Monitoring..... 3/4:8-28 ,- i 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH..................................... 3/4 9-1 . 3/4.9.2 INSTRUMENTATION................................'......... 3/4 9-3 3/4.9.3 CONTRCL ROD P0SITION............_....................... 3/4 9-5 3/4.9.4 DECAY TIME.............................................. 3/4 9-6 3/4.9.5 COMMUNICATIONS.......................................... 's/4 9-7
'
3/4.9.6 REFUELING PLNTF0RM...................................... 3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L. . . . . . . . . . . . . . . . . . 3/4 9-10 3/4.5.8 WATER LEVEL - REACTOR VESSEL. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-11 , 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L. . . . . . . . . . . . . . . . . . . 3/4 9-12 .
LIMERICK - UNIT 1 xv f AUS S B85 5
.. . - .. . . - . . ... . . . . . . . . .
-
'
,..-.+...--..~m------- * ~ * *
,,= . .. .
.y. . l
/
,
. . - - - . . .-
.
_
,
.
_,
,-
-
- - -- -
.. . ....- - - - - - _ _
_
- -. .- _
- ..
. - - _.
. . - . - -- . -
- -- ._.. . - - - . -
-
* - . .
-
-
- :-- ._ _
~~ *
_ -- _ _
- - -
INbEX ___ _ -. _
.(-
,
--- -
.- .- _
-
LIMITING CONDITIONS FOR OPERATION AND' SURVEILLANCE REOUIREMENTS _
- . __
SECTION -
-
-
_ PAGE 3/4.9.10 CONTROL ROD REMOVAL
,-
Single Control Rod Remova1.............................. 3/4 9-13 __ Multiple Control Rod Removal..... 2 .................... 3/4 9-15 ,
'
3/4.9. n RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HighWateri. eve 1........................................ 3/4 9-17 Low Water Leve1............;............................ 3/4 9-18 3/t.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY........................... 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM............................. 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS. ........................ , 3/4.10-3 3/4.10.4 RECIRCULATION L00PS..................................... 3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION.................................... 3/4 10-5
.. -
3/4.10.6 TRAINING STARTUPS....................................... 3/4 10-6 3/4.11 RADIOACTIVE EFFLUENTS -
.
-
.
3/4.n.1. LIQUID EFFLUENTS Concentration........................................... 3/4 11-1 Table 4.n.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program........................ 3/4 n-2 Dose.................................................... 3/4 n-5 . ' Li qui d Radwaste Tre atment System. . . . . . . . . . . . . . . . . . . . . . . .. 3/4 n-6 Li q u i d H o l d up . T a n ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 n-7
.
3/4.n.2 GASEQUS EFFLUENTS Dose Rate............................................... 3/4 n-8 (
.
ADB B BIS LIMERICK - UNIT 1 xvi .
*
. . . - .. . . . . - . - _ . .-. . . . . . - . . . . . . . . . . . _
.. . - --
. . . .. . . . . . . . .
_, .
_
.- _
- - .
_ .
. .
_ _
-
-
. . . _ --
- - -
.- 1
_.- .
..
_.
- *-~
.-- _,
,
- - - - ~ ~ 1
.__
~=
387.5101C20
~ - - l
__
- . - . .
---' *- - _ , .
. 3nogx -
l
_
- - -
-- ._ _
j
'
-- . - - -
--
-- . L I w ! ' NO CON !!!ONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS j
- . -
.
PAGE - SECT!0N -
..
..
~.-
RADIOACTIVE EFFLUENTS (Continued) Table 4.11.2.1.2-1 Racioactive Gaseous Waste i Sampling.and Analysis !
-- Program................... .... 3/4 11-9 Dose - Noble Gases...........................:.......... 3/4 11-12 Dese - Iodine-131. Iodine-133, Tritium, and Radionuclices in Particulate Form..................... 3/4 11-13 -
i
)
Ventil ation Exnaus t Treatment System. . . . . . . . . . . . . . . . . . . . 3/4 11-14 i Explosive Gas Mixture................................... 3/4 11-15 l Main Condenser................. ........................ 3/4 11-16 ,
.
Ve nt i ng o r P u rg i n g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-17 l I 3/4.11.3 S0'.10 RADWASTE TREATNENT........ ....................... 3/4 11-18 3/4.11.4 TOTAL DCSE.............................................. 3/4 11-20 3/4.12 RADIDLOO! CAL ENVIRONMEG AL MONITORING 3/4 12-1 3/.4.12. 1 MONITORING PR0 GRAM...................................... Table 3.12.1-1 Radiological Environmental Monitoring Program................. 3/4 12-3 Table 3.12.1-2 Reporting Levels For Radio-activity Concentrations in Environmental Samples.............. 3/4 12-9 Table 4.12.1-1 Detection Capab(11' ties For Environnertal Sample Analysis...... 3/4 12-10 3/4 12-13 3/4. 12.2 LAND USE CENSUS......................................... 3/4 12-14 3/4.12.3 INTERLABO RATORY COMPARISON PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . l
' - _
LIMERICK - UNIT 1 xvii g gg
- ,
I l
. . . -. .
- _ _ - _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ________
-
--.
. -
._
..
, -
- - _ _
-
..
_ _ _ . _ __ -. . . _ - - - _ - . _
--
__ _ --- - - - - _
~~~1 _
--
~ ~ -
- ~
___- 3875101f C.
- - _- _.
_
._
- -
-
_ _ __ _ _ .
-~ -
_ __ __ __ _ ' - _.
. ._.
_.
8ASES PAGE __ SECTION
~ - 3/4.0 AP P LI C A81 L I TY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . W 3/4 0- 1
_ 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SNUTDOWN MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 "i- 1 3/4.1.2 RE ACT IVITY AN0MLI E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 1- 1 3/4.1.3 CONT RO L R00 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 1- 2 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS......................... 8 3/4 1-3 , 3/4.1.5 ST AND8Y LIQUID CONTROL SY5 TEM. . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION 8 3/4 2-1 RATE................................................. LE FT I NT ENTI ONALLY 8 LANK. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 2- 3 3/4.2.2 A P RM S ET P0! NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 2- 2 3/4.2.3 MINIMUM CRITICAL POWER RATI0......................... 8 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE.......................... 8 3/4 2-5 3/4.3 Ih5TRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUNENTATION. . . . . . . . . . . . 8 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUIENTATI0N. . . . . . . . . . . . . . . . . . 8 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRupENTATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 3- 2 3/4.3.4 RECIRCULATION PUlf TRIP ACTUATION INSTRUMENTATION.... 8 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...................................... 8 3/4 3-4 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION.................... 8 3/4 3-4 _ 3/4.3.7 MONITORING INSTRtMENTATION _ Radiation Monitoring Instrumentation................. 8 3/4 3-4 i __ LIMERICK - UNIT 1 xviii Anwndment No.7 AM i41NI
. _ - . - _ . -_ .. .. .. - .- _ - -_- _.- -.- - -... - ..-.. - . . . - . -
.. . . . . _ . . . .s...... .
.m.,.._. . . . . . . . . . . . . .. ._
- - - -
._
- -
- -
_
-. --- -
_ __ _
. - - -
___. . . . - . .
, _. _ . -- - ___
_ _
-- -
- - ~ ..
___ _
-. -
--
- - -
:.. - - - -
_
-_
--
. . _ - .
.
- --
.-.
__
- - . --. _.-. -_ . . - .-_ - - - . .- - --- -. ..
_ _
-
.- _
--
-.
. - _
.
.
.
.
SECTIONS 3.0 and 4.0 LIMITING CONDITIONC FOR OPERATION
.
AND SURVEILLANCE REQUIREMENTS
..
.
-
l l
!
*
i I
- -
h t
$1 1 6a 4 '
- - - - - - - - - . . . . .m. .. y ._ . . . . . .. ... .._
m,_.. . . . _ _ . ___ . _ . _ ,
- - - -
-
,
_.
_ _ : _ _ ___- _ ._ ._ . . _ _
--
- -
.
-
_ _ __ _
~ -
_ . -
~
:- -
._. _
._
_ 7 _
-
3/4.0 --APPLICABILITY -_ - - _
--
~
-
LIMITING CONDITION FOR OPERATION
- ~
---
.
3.0.1 Compliance with the Limiting Conditions for Opera' tion contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.--
-
-
__
' 3.0.2 Noncompliance with a Speci.fication shall exist when the requirements of the-Limiting Condition for Operation and associated ACTION requirements are not met within the specified tin intervals. If the Limiting Condition for Operation is restored prior to e: )iration of the specified time intervals, completion of the Action requires ents is not required.
3.0.3 When a Limiting Condition "or Operation is not met, except as provided in the associated ACTION requiremiits, within one hour action shall be initiated to place the unit in an OPERATIONA, CONDITION in which the Specification does not apply by placing it, as applica le, in: a. At least STARTUP within the next 6 hours, b. At least HOT SHUTDOWN within the following 6 hours, and c. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Speci-fications.
This Specification is not applicable in OPERATIONAL CONDITION 4 or 5. - 3.0.4 ' Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or'to OPERATIONAL CONDITIONS as
,
required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.
.
.
e
*
.~ _
%
,_
.
LIMERICK - UNIT 1 3/4 0-1 t
@
. . . . . . , . . . . . . . . ... . ~ . - . - - - - - . - . . . . - - . - -
:.h . -
. . ... .
l _ __
-
.
-
.-
--
_--
- - -- - ,- __
_ __ _ _ _ ._
. _ . _ --_
,
-
.
*
-*= _ .
*
l
- ---.
-- APPLICABILITY . . _ . _-- -
. . _ . _ ___.
_
~
-
-
_ _. - _ _ _ _ _ _ SURVEILLANCE REQUIREMENTS - ~ _ _
-_-_ ,
_ _ - . _ 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other condit. ions specified for individual Limiting Conditions.
for Operation unless otherwise stated in an individual Surveilla.nce Requirements._
~
4.0.2 Each Surveillance Requirement.shall be performed within the specified ._ time interval with:
- .
a. A maximum allowable extension not to exceed 25% of the surveillance interval, but b. The combined time interval for any 3 consecutive surveillance inte.rvals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requir,ement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to thes'e requirements are stated in the individual Specificatons. Surveillance requirements do not have to be per-formed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condi-tion shall not be made un'less the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, & 3 components shall be applicable as follows: a. Inservice inspection of ASME Cnde Class 1, 2, and 3 components and inservice testing of ASME Code ,Cla'ss 1, 2, and 3 pumps and valves shall be performed'in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, - Section 50.55a(g) (6) (1).
b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications: ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice - inspection and testing inspection and testina activities activities Weekly At least once per 7 days
-
Monthly At least once per 31 days . Quarterly or every 3 months . At least once per 92 days Semlannually or every 6 months At least once per 184 days _ Every 9 months- At least once per 276 days Yearly or annually At least once per 366 days l l LIMERICK - UNIT 1 3/4 0-2 t
@ i
-
'
L. ____. ._
'
-* '- ~ " ' --
!* _
l _
-- - - -
_.
-. \
-
_ _
-
-
i
-- -
__
,
_ _
- - . - 3/4.3 INSTRUMENTATION --- - - - - _
_ _. . __- - l _
" -
~
-- 3 /4. 3.1 REACTOR DROTECTION SYSTEM INSTRUMENTATION -
,
j
~
---- ~~
iIMITINGCONDITIONFOROPERATION
-- -
3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2. i APPI '.CABI LITY: As shown'in'. Table 3.3.1-1. -
- - .$T10N: e a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place.the inoperable channe'1(s) and/or that trip system in the tripped condition" within 1 hour. The provisions of Specification 3.0.4 are not applicable, b. With'the number of OPERABLE channels less th n required by the Minimum OPERABLE Channels. per Trip System requirement for both trip systems, place at least one trip systom** in the tripped conditian within 1 hour.
and take the ACTION required by Table 3.3.1-1.
' SURVEILLANCE REOUIREMENTS
.
4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL r.ALIBRATION operations for the OPERATIONAL - CONDITIONS and at the f requencies shown in Table 4.3.1.1-1.
~ 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip , functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its l limit at least once per 18 months. Each test shall include at least one ' channel per trip system such that all channels are tested at least once every ! N times 18 munths where N it the total number of redundant channels in a i specific reactor trip system.
l
"An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the -
*
tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both
. systems have the same number of inoperable channels.,olace either trip system-in the tripped condition.
.- LIMERICK - UNIT 1 3/4 3-1 O 616 t
.
-
-- .
. . . . - - - . . . - . - . - .. . ... . . . .
. . . . . . . . . . ... ._._... ._:..-- . .=-- -
- , , , - - - .-- --. ~ . , -
*
4l l l-
-
i l: l l t !
-
i
. ,
, .
*
TABLE 3.3.1-1 , f'i'**i, t.
j REACTOR PROTECTION SYSTEM INSTRUMENTATION i , I i '3 C.
i n APPLICABLE MINIMUM
~
. l l.,,
,
.
*
j l OPERATIONAL OPERABLE CHANNELS
- *
E FUNCTIONAL UNIT -CONDITIONS PER TRIP SYSTEM (a) ACTION l )
'
-
-
,l ; ,
! S*l 1. I,ntermediate Range Monitors (b) , ,
'
, ,
- l '
i i i l'
. ; a. Neutron Flux - High 2 3 1 i ,ii l 3, 4 3 2 i~* <
l 5(c) 3(d) 3 .
(i. I , ! 1 .l' l *
b. Inoperative 3, 4 2 3'
1
1 i ' ll : 3 I i
.
-
'5 3(d) .
.
i 3' . j . . R Average Power Range Monitor I '): - I I' ', *- 2. , t
'
;i T a. Neutron Flux - Upscale, Seldown 2 2 1 ,
" 3, 4 2 2
-'
.
*
' l' i i 5(c) 2(d) 3~ ,
'
I -i s !' b. Heutron Flux - Upscale " l ' 2 4 1) Flow Blased 1 )} 2) liigh Flow Clamped 1 2 4 l
'
l. i!
'
Inoperative 1. 2 2 1
! c. ,
'
3, 4 2 2 l -l ,[ -
)
5(c) 2(d) 3
,
-
l'l [
' '
d. 1(g) 2 4 Downscale i I ,'
,
l I : , j l
. .' [
. ' 3. Reactor Vessel Steam Dome Pressure - High 1,2(f) 2 1
] -
M . ' l 4. Reactor Vessel Water level - low, It p 2 -
- 1. t - - Level 3 1, 2 i l, 4 : I".
! . 5. Male Steam Line Isolation Valve - ' f.
1(g) ,1/valv'e 4 ; ( ., l i Closure , l. ; , i
-
I
, ,
.
,
'
,%
'
.
., ,j
" '
; .
i
.
'
:i.,
.
.t <
,
l: .
-
..
I , , r-TABLE 3.3.1-1 (Continued)
.
,
-
l;f L , i j ,
. REACTOR PROTECTION SYSTEM IliSTRUMENTATION , , , , , .
l
. / .l
5 n l , 1 * '
' APPLICABLE MINIMUM 1 4 -
OPERATIONAL OPERABLE.CHAhNELS i !! E FUflCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION .
'
- p .
- -
.
'
l I q 6. Main Steam Line Radiation - ,
' liigh 1,2(f) 2 5 i
,' ! i
',
.
,
, l-
,
l j 7. Drywell l j.
i Pressure - liigh 1, 2(h) 2 1 .
,
li
-
;
-.
l ! 8. Scram Discharge Volume Water I- l.
Level - filgh -
.
l t
'I
.
- ' .
'
I ' a. Level Transmitter 1, 2 2 - 1 i
:.
R> 5(f) 2 3 .
> '.;
-
Y
*
b. Float Swltch 1, 2 2' ' I
*
r 5 (1) 2 3 l-
'
'
9. Turbine Stop Valve - Closure 1(j) 4(k) 6 . !
- I .
,
*
l . , 10. Turbine Control Valve l'ast Closure. - - Ik) l Trip 011 Pressure - Low 1(j) 2 - 6 i ;
-
l 11. Reactor Mode Switch Shutdown j i l i
*
Position 1, 2 3, 4
2
- 7
'
.
i
- d j. [ '
;.
' 5 2 3 i ,
i . i . 12. Manual Scram 1. 2 2 1 -i.' ,! M 3, 4 2 8 j
i
* -
'S 2 9 I .
- ; l'
m
'
@ i j
.'
,
-
. :. -
, ,
i l
- .
i
>
I l* l .
.. i
: , ..
,
. 'r
-
- . .
^.
.
, . l
- 1
- ..' 7, _ . _., - _
. _
. . _ . - . _ _ . w. . .._ _ .
_ ..
=- . .
*-
.--. - -
.--- _
,_,
. _ . .-_ .
_ _
- -- -
._.
_ . . . ,
~
. .- .
-
-
-._. _ '~-~~~~__
-- --
. . . . __
. ._
_
--
-
__ ..
~ ~
'- -- ~
--
-
- TABLE 3.3.1-1 (Continued) _i_ _,
_.
_
~'
~
-
-
REACTOR PROTECTION SYSTEM INSTRUMENTATION
- -
ACTION STATEMENTS-ACTION 1 - Be in at least HOT SHUTDOWN within 12 hours.
ACTION 2 - Verify all in'sertable control rods to be inserted in the core-and lock the reactor mode switch in the Shutdown position within 1 bour. -
,
'
- . .
ACTION 3 - Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour.
ACTION 4 - Be in at least STARTUP 'within 6 hours.
ACTION 5 - Be in STARTUP with the main stean; line isol'ation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours. l
.
\
ACTION 6 - Initiate a reduction in THEPJiAL POWER within 15 minutes and , reduce turbine first stace pressure until the function is !
.
automatically bypassed, within 2 hours. ) ACTION 7 - Verify all insertable control rods to be inserted within I hour. l ( < ACTION 8 - Lock the reactor mode switch in the Shutdown position within ! I hour. I ACTION 9 - Suspend all operations involving CORE ALTERATIONS, and.
insert all inse-table controi rods and lock the reactor mode switch in the SHUTDOWN position within I hour. .
.
l l l l
'
.
.
.
.
LIMERICK - UNIT 1 3/4 3-4 g5 8 M5 g
. - . _
. . _
, . _ - . -. -- - - _ _ ,
,
_.- _. ,
. -
_- _
.-
_ _
-
_
..
. - - - -
_L --
~~
.
--
-
^~ -
.:...
~~~
~--__ __
_
~ -
_.
T
~ ~
_ - -TABLE- 2. 3.1-1 (Continued)., - _
--- -
_
-
._
-
--REACTOR PROTECTION SYSTEM INSTRUMENTATION
-
--
_ _. _
~
TABLE NOTATIONS
- ~
-
l (a) A channel may be placed in an inoperable ' status for up to 2 hours for required surveillance without placing the trip system in the tripped 2-. condition provided at least one OPERABLE . channel in the same trip system i is monitorin0 that parameter. ,
.(b) This function shall be automatically bypai; sed when ih'c reactor rode switch is in the Run position and the associated APRM is not downscale.
(c) The "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn * and shutdown margin I demonstrations performed per Specification 3.10.3.
~
- l (d) The noncoincident NHS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, -
the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs and 2 SRMs.
. .
(e) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(g) This function shall be automatically bypassed when the reactor mode switch
.is not in the Run posi. tion. .
,
(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT
,
IKTEGRITY is not required.
~
(i) With any control rod withdrawn. Not eppl cable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER.of less than 30% of RATED THERMAL POWER.
(k) Also actuates the EOC-RPT system.
'
. *Not' required for control rods removed per Specification 3.9.10.1 or 3.9.10.2. l I
LIMERICK - UNIT 1 3/4 3-5 _ C E12 g
!
, -. _ . . . _ _ . _ - .... -- _ . . . - __.=
. . . . . . . .. .
. _ _ _ _ _ _ -
-_
_ _ _ _ . . _ _ _ _____ __ _ _ _ - - _ _ . _
_ _
'
.
.
-
.k .' ! ,
j'
,
,
t
'
l' l - TABLE 3.3.1-2
. ! * i i_ 1
, r- .- .
' -
l
*ii REACTOR PROTECTION SYSTEM RESPONSE TIMES \
m
. % RESPONSE TIME i E
~
'
'
l . FUNCTIONAL UNIT (Seconds) _ , . I
,
;
g 1. Intermediate Range Monitors: ! f T ii l ]z
. " a. Neutron Flux - liigh N.A. .
e ,;
.
,
, w * b. Inoperative N.A.
,
, ' '
2. Average Power Range Monitor *:
'
:
I a. . Neutron Flux - Upscale, Setdown M.A.
l l
- ,
l b. Neutron Flux - Upscale t
*
t C 1) Flow Blased 10.09 ]l l i -
!l i 2) liigh Flow Clamped 10.09 r l
c. Inoperative N.A. l l { N.A. i .. 2 . w . d. Downscale l 7-* i E g .i
I 3. Reactor Vessel Steam Dome Pressure - High $ 0.55 - , 4. Reactor Vessel Water Level - Low, level 3 < 1.05 9 i ; .a l ' ' ' ' ' 5. Main Steam Line Isolation Valve Closure
-
3'O.06 l , l ' '
Main Steam Line Radiation - High N.A. 'l ;
;
6.
, l
.
N.A. ! $ 7. Drywe?1 Pressure - High . .. i. .; , >
..-
' 8. Scram D Scharge Volume Water I.evel - High l ;1 .'.j ,' a. Leve? Transmitter N.A.
I , Float Switch N.A. . e i b. j - i p i ' l
:F, 9. Turbine Stop Valve - Closure 1 0.06 I
I o,' 10. Turbine Control Valve Fast Closure, . l
$ 0.08" ,
,
j g Trip 011 Pressure - Low i
*
11. Reactor Mode Switch Shutdown Position N.A.
N.A. . fl . I 12. Manual Scram
-
l ' ! ! ' l ' i
*
* Neutron detectors are exempt from response tiea testing. Response time shall be measured e
.-
l from the detector output or from the input of the first electronic component in the channel.
- ,
,
.
** Measured from start of turbine control valve fast closure. .
l
-
.
;
,
ji
,
. .
- -
..
.
,i
!
.
{. ..
.
ii.j
, TABLE 4.3.1.1-1 , , ' i C.
,7, REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS -
. :n
'
'
'
h CllAliNEL OPERATIONAL
, ,
FUNCTIONAL UNIT CllANNEl; CllECK
. FUNCTIONAL TEST CilANNEL CALIBRATION I '} CONDITIONS FOR W!!ICll I] *
SURVEILLANCE REQUIRED ,
, .
g -
! .
,
{
'
N 1. * Intermediate Range Monitors:
'
!
l
' - 6' a. Neutron Flux - liigh S/U.S(b) S/U(c) W R 2 . . . j i i S W(j) R 3,4,5 I
,
-
i . l
'
) l b. Inoperative N.A. k(j)' N.A. 2,3,4,5 , '[ 2. Average Power Range MonitorII): l I-l * i a. Neutron Flux - S/U.S(b) S/U(c),W SA 2 W(j) 4 Upscale, Setdown
-
S
~
SA 3, 5 . ll.
l i i
!i -l -
.,
#
! g b. Neutron Flux - Upscale .
, , :
1) Flow Blased 5,D(g) 'S/U(c),W W(d)(e),SA, ,1 ' l
} ; ,
3 2) liigh Flow Clamped S S/U(c),W W(d)(e). SA 1
,
i, j.
~
.
I c. Inoperative N.A. W(j)- it. A. 1, 2, 3, 5 i i , ' '
'
*
j L- . ' d. Downscale S W SA 1 l - e l 3. Reactor Vessel Steam Dome i Pressure - High S M R 1, 2(h) i
,
'I'
' '
.
l e
'
I 4. . Reactor Vessel Water level - i ,
,
.R
,
1 Low, Level 3 S M . 1, 2 ,
' I j
'
t G )
"' ,.,,
5. Main Steam Line Isolation Valve - Closure N.A. M R 1 ,
)'\ : t l
! SY 6. Main Steam Line Radiation - I
,' *
i
! liigh 5 - M R 1, 2(h)
,
; ,
-
I 7. Drywell Pressure - Illgh S M R . 1, 2 , l
- ,
.
I iI'
. . .
~
,
! l
, , - : ;
.
l
' - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ - _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
..
.
- l
'
'f['~
TAPtE 4.3.1.1-1 (Centinued)
-
,
' '
. REAC10R PROTECTION S) TEN INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
'
p
*
- .
r-CilANhEL OPERATION.il . ;
*i! .l - ?
g * CllANNEL FUNCTIONAL CllANNEL CONDITIONS FOR WHTCH i l-
'
' FUNCTIONAL UNIT CilECK 1EST CALIBRATION SUrtVEILLANCE REQUIRED M i[ Scram Discharge Volume Water I ! I 7 8. .
;
e , level - Higli 1, 2,jS II) !
,
Level Transmitter
-
,I a. S M R 5 . lJ ,'
-'
w b. Float Switch N.A. M R 1, 2,.5(g) i i i
'
'
.
;
, l 9. Turbine Stop Valve - Closure N.A. M R 1 l- l,
-
.t , l' 10. Turbine Control Valve Fast .
-
.
.. .
4 .
! l '! I - ,".
l Closure, Trip 011 ~ . Pressure - Low N.A. M R 1 l l l ' l i 11. Reactor Mode Switch i lI i.
' Shutdown Position N.A. R N.A. 1,2,3,4,5 l N.A. 1,2,3,'4,5 l $ 12. Manual Scram N.A. M-l 4 ! [ (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. .
(b), The IRM and SRM channels shall .be determined to overlap for at least % decades during each startup l ;
i
.E l'
lI
. after entering OPERATIONAL CONDITION 2 and'the IRM and APRM channels shall be determined to overlap for at least h decades during each controlled shutdown, if not performed within the previous 7 days.
i I , l.
bi
,
i l (c) Within 24 hours prior to startup, if not performed within the previous 7 days. ,
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values A
h
't
calculated by a heat balance during OPERATIONAL CONDITION I when THERMAL POWER > 25% of RATED ha.j
'
* '
l THERMAL 99WER. Adjust the APRM channel if the absolute difference is greater tiiad.2% of. RATED T'HERMAl[ l -
'
POWER. Any APRM channel gain adjustment made in compilance with Specification 3.2.2 shall not be ,
'
,
'I
;
Included in determining the absolute difference. . ~
'
!
(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a j l- ,
j g
..
calibrated flow signal. -
(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system. ji t ,
,
,
[.
l.,'
'
l (g) Verify measured core flow (total core finv) to be greater.than er eqeal to estabitshed core flow at l' '
' "'
, the existing loop flow (APRM % flow). ; i j $ (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed'per i I J.
, Specif! cation 3.10.1. I
l (1) With any control rod withdrawn. Not appilcable to control rods removed per Specification 3.9.10.1' : or 3.9.10.2. .(',
-
. t ,.
- (j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, .
! I and no control rod shall be moved from its existing position. .
. , I
'
3l
- . - 4
'
- '
ll .. . _
.
. ..
.
-
.
.
.
l-
.
';
.
~ -
_ .-n.-. _ .
~~
T387-200559'2r-~ --
, _ _ . . _ _ _ . _ _ -
-
-;- -. -
-
._
_ _. _
~~
___ _ -INSTRUMENTATION
[ -
_,
.--
_ , _ -
..
'
-
3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION __ . --_
-
LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.. APPLICABILITY: As shown in Table 3.3.6-1. --
'
ACTION: a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
D. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION requirec by Table 3.3.6-1. ,g SURVEILLANCE REOUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentction channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.
. _
.
LIMERICK - UNIT 1 3/4 3-57 AUS S Sti ,
I 1, .
'
'
. j' ,
'
387200,5592 l( , l TABLE 3.3.6-1 ,
'
l CONTROL ROD BLOCK INSTRUMENTATION g I E MINIMUM APPLICABLE 5-
'
OPERABLE CHANNELS OPERATIONAL l CONDITIONS ACTION 9 TRIP FUNCTION PER TRIP FUNCTION l i
'
1. ROD' BLOCK' MONITOR I ")
,
-
l.
i 2 1* 60 E a. Upscale 2 1* 60 l G b. . Inoperative 1* 60 l1 i w c. Downscale 2 i i I
,
2. APRM I I a. Flow Blased Neutron Flux - 1 61
,
Upscale *
i Inoperative 4 1,2,5 61 ,
b. 1 61 t 4 , c. Doenscale 2, 5 61 'g d. Neutron Flux - Upscale, Startup 4 ! 3. SOURCE RANGE MONITORS *** 3 2 61
'
li R a' . Detector not full in(b) 2 5 61 i
, * '
3 2 -
Y b. Upscale (c) 2 5 61 . g , I I c. Inoperative IC) f f h l 3 2 61 I d. Downscale(d) 2 5 61 I i t 4. INTERMEDIATE RANGE MONITORS I' 6 2, 5 61. f a. Detector not full in 6 2, 5 61 l
'
b. Upscale 2, 5 61 j ,
' c. Inoperativ 6 2, 5 61 d. Downscale'g) i; k 2 5.
, SCRAM OISCHARGE VOLUME 2 1, 2, 5** 62 l fN 3 -
"
a,
. ,
Water Level-High < , i i I, t l g 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW 2 1 62 l a. Upscale 2 1 62 , a b. Inoperative 2 9. - 1 62 l i, 3 c. Comparator i 3 '4 63 % 2 , ) 7. ~ REACTOR MODE SWITCH SHUTDOWN POSITION ; ,- ,
. i
' ,
, _ _ _ _ _ __
- ____
,_
~
_
- ' *
_.-_ _
--
_.-..
- - . . _ _ .
- -- - - - - __
-
,. ~~
JQ~5 ; 9
~ ^
- ~- . .- ,
-
_ _- - _ _ __. _
~ ..-
-
--_ _. __
~
-
TAoLE 3.3.6-1 (Continvec)- - --
- COCRDL R00 V?TH3RAVAL BLOCK INSTRUMENTATION _
ACTION STATEMENTS ACTION 60 - ~ Declare the RBM iniperable and take the ACTION required by
'
Specification 3.1.4.3.-
~
ACTION 61 - With the number of.0PDLABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function recuirement, place at least one inoperable channel in the tripped condition within one hour.
ACTION 62 - Vith the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
ACTION E3 - Vith the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requireeent, initiste a rod block.
NOTES
* With THERML POWER > 30% of RATED THERML POWER.
** Vith more than one control red withdrawn. Not at.plicable to control rocs removec per Specification 3.9.10.1 or 3.9.10.2.
'" These channels are not required when sixteen or fewer fuel assemblies,
.ecjacent to the SRMs, are in the core.
;
\
(a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference A*RM channel indicates less than 30% of RATED THERML POWER.
(b) This function shall be autoestically bypassed if detector count rate is
> 100 cps or the IM channels are on range 3 or higher.
(c) This function is automatically bypassed when the associated IM channels are on range 8 or higher.
(d) This function is automatically bypassed when the IM channels are on l range 3 or higher. l
(e) This function is automatically bypassed when the IM channels are on I range 1. !
l _ _
;
~
. l LIMERICK - UNIT 1 3/4 3-59 Asenerent Ho d
- MAY 11 IMI l
- ._ . ._ .- .- . - . .. . - -
'
I l
-
( s i
o
:
3875101020 h , i 5 c5 TABLE 3.3.6-2 .I
4 CONTROL R00 BLOCK INSTRUMENIAll0N SETPOINTS
;
'
n
'
'
7 TRIP TUNCTION 1 RIP SEIP0lMi Alt 09A8tE VALUE , i k- 1. R00 BLOCK MONITOR upscaie i
. *
e.
1. flow blased 1 0.66 W + 41%, with a maximum of 2 5 0.66 W * 44%, with a maximum of, l ' t {. l 11. high flow clamped < 107% < 110% l '
,
b. Inoperative H.A. R.A.
c. Downscale > 5% of RATED TitERMA1. POWER > 3% of RATED THERMAL POWER.
-
*
l l 2. APlet a. Flow Bla:ed Meet on Flux - Upscale < 0.58 W + 50%* -< 0.58 W + 53%) l 'l I b. Gnoperative R.A. M.A.
> 3% of RATED TilEllMAL PCWER ; w c. Downscale i w d. betron Flux - Upscale, Startup
-> 4% of RATED THERMAL POWER 512% of RATED TitERMAL POWER
~
< 14% of RATED'iHERMAL POWER I
'l A 3. SOURCE P.*4E3E IE)MITORS ,
s. Detector not full in M.A. M.A. l , b. Upscale < 1 x 105 cps < 1.6 x 105 cps c. Inoperative h.4 R.A.
!
) 3 r.pT** > 1.8 eps** j
'I d. Downscale 4. INTEIBEDIATE RANGE MONITORS a. Detector not full in M.A. M.A.
i
'lI i
}i il b. Upscale < 108/125 divisions of < 110/t75 divisions of !
l c. Inoperative Tull scale M.A.
Tull scale N.A.
,
'
',j d. Downscale > 5/125 divisions of full > 3/125 divisions of full l i' * i.
. scale scale GEg * , o ' 5. SCRAM 015CHAAGE VOLUME l l i j ~{
^ '
a. Water Level-High 1 257' 5 9/16" elevation *** $ 257' 7 9/16" elevation
.
@a"x a. Float Switch .
,
,
.
i w
! i + ,;
! - i
~ . .
, .
t I e.
,- J-
. l 3872003.7a3 l l
,
;e-g TABLE 3.3.6-2 (Continued) 3 t
c
,
'
:
'S ' CONTROL ROD BLO4K INSTRUMENTATION SETPOINTS '
M l
* TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION
'
l h 6. REACTOR COOLANT SYSTEM RECIRCULATION
.
M FLOW Upscale < 111% of rated flow < 114% of rated flow t ( :i .
w a.
b. Inoperative R.A. R.A. ' i l l f '.
-
l
'
, c. Comparator 1 10% flow deviation i 11% flow deviation i ,
,
^
.
7. REACTOR MODE SWITCH SHUTDOWN I I
,;
N.A. N.A. , I P55TT10N 1 ( l,
' The Average Power Range Monitor rod block funci. ion is varied as a function of recirculation loop flow L
'
(W). The trip setting of this function must be maintained in accordance with Specification 3.2.2. ' e w (
i 2 **May be reduced to 0.7 cps provided the signal-tn-noise ratio is > 2. , Y *** Equivalent to 13 gallons / scram discharge volume. i I ' i E 'I
~ . .
t
-
I
- -
l . ii j-
,,
'
l g i t !
'
l f. - O l l
-
, '. ,i
. - '
" ,
'
~ ~
.
l :\ I
,5 i
w 'j
~
i . . !
. _ . . - _ .
. . _ _ _ -
* M- , - - -
- - . . _ .._7
. - .-
__
-
. -. - .
-
~~r- 3872003733 ~
~
= - ---
_ _ e
--
_ - . _ . _.
.
" ' W W W . - - ---. -
_ . __ row
g s-w -M
* '*
w.
_ e M.
,-m e we INTENTIONALLY LEFT BLANK
,
I l
.
.
!
l l l
1
.
* ee
, - , ,
. . ' j, l 2
! .l
>I .
*
'
TABLE 4.3.6-1 j l
,
,_
.-4
,
.
'I r) CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS, y l
;l ;; .
i 3 - CNANNEL OPERATIONAL t ;
;~ e CilANffEl- FUNCTIONAL CllANNEL CONDITIONS FOR WIIICll
, j! g TRIP FUNCTION CIIECK TEST CALIBRATION (a) SURVElltid8CE REQUIRED
"
. Z 1. ROD Bl.0CK MONITOR I
.
'
s l
! a. Upscale N.A. S/U(b)(c) (c) 33 3,
, ,
;i ,; ,
'
b. Inoperative N.A. S/U(b)(c), (c) N.A. la .'
' I' ., ,
c. Downscale N.A. S/U(b)(c) (c) 33 7, ,
,
j"* 2. APRM - i I t i;' a. Flow Blased Neutron Flux - . -' "~ - ' '
! I .
Upscale N.A. S/U(b) M'
, SA 1 i 8 .i . b. Inoperative N. A. . S/U ,M N.A. 1, 2, 5 It is c. Downscale N.A. S/U(b),M SA 1 L Neutron Flux - Upscale, Startup '
1: d. N.A. S/U ,M SA 2, 5 l
{ ,
*
w 3. SOURCE RANGE MONITORS b'
~
-
I:
'
a. Detector not full in N.A. ' S/U((b) y- N.A. I 2, 5
} :!I '
b. Upscale N.A. S/U b) W
. SA 2,5- i .
'
I c. Inoperative N.A. S/U ,W N.A. 2, 5 l d. Do'wnscale N.A. S/U ,W SA 2, 5 ;
'
*
4. INTERMEblATE RANGE MONITORS . I
' a. Detector not full In N.A. S/U(b) W
, N.A. -
2, 5 , i
'; b. Upscale N.A. S/U ,W SA ,
2, 5 I ,
, l'
c. Inoperative N.A. S/Ug),W N.A. 2, 5 l l-: d. Downscale N.A. S/U ,W SA
'
2, 5. < t 1 5. SCRAM DISCHARGE VOLUME . .
, ;
l;' g ,
.
I a. Water level-liigh N.A. M R 1, 2, 5** ,
'
i , r 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW ..
'
M , a. Upscale N.A. S/U ,M SA 1 ' ',
.,' ., b. Inoperative N.A. .
S/U(b),M N.A. 1 ', , 8 ,
, . c. Comparator N.A. S/U ,M SA 11 g
. 7. REACTOR MODE SWITCH Sil0TDOWN l li
.
l POSil10N N.A. R N.A. 3, 4
*
I ; - i i ,
, . .
'
- . 1
,
'
.e -
.. - .m_ - ..
, ' _ - , y. . _
-
s*' . . e. : L,,,,_.- . 3- - *. _, _--
-- - - - . . -
- .-
._. . - - ~
_, --
.. _
-
-
-_ .--
_
.
- ---- .
' '
e
' 7_h ^-
__. _,, _ --- .- . TABtE4.3.f-I (ContigUEd7 .._- -
' -
-
CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE-REQUIREMENTS
.
TABLE NOTATIONS
.. _,_, . __
(a) Neutron detectors may be excluded from CHANNEL f.ALIBRATION. _,
~
~
(b) Within 24 hours prior to startup, if not performed within the
- previous 7 cays. .
'
(c) Includos reactor manual control multiplexing systes input. ..-
' '
* With THERMAL POWER > 30% of RATED THERMAL POWER.'
** With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. -
-
, .
. .
,
..
.
.
* -
.,
.
.
.
.
.
.
. ,
,
-
.
-
..
.
LIMERICK - UNIT 1 3/4 3-62 AUC 8 155 -t
- .. ..
. - . . . . . . .. - .. . . . . - -
- . . . . - . . . - .. _ .
_. _ - . _ _ _ _ __ , __
.;-. ,
[
- - . . . . , . . . -- . . .- ._... ~
: ;: ---- _ - *~:_. .
~-
.
. _ . -
_
-
* _ -_. _ . . _~-- - -
._.
- _.
-. ___
- - - - ~ . -
. - _
~
- -
__
^ ~ --
T ~~ - 2 /b. 6 CONTAINMENTShSTEMS. _ _
-- - ~
3/4.6.1 PRIMARY C0NTAINMENT ~ , g_. --
-
.
PRIMARY CONTAINMENT INTEGRITY.
- . LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: OPERATIONAL- CONDITIONS 1, 2*, and 3. .
.
*
' ACTION:
Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour or be in at least HOT SHUTDOWN within thr next 12 hours and in . COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMRY CONTAINMENT INTEGRITY shall be demonstrated: a. After each closing of each penetration subject to Type B testing, except the primary containment air locks, if opened following Type A or B test, by leak rate testing the seals with gas at P,, 44.0 psig,. and verifying that when the measured leakage rate for these seals is added to the leakage rates detemined pursuant to Surveillance Requirement 4.6.1.2d. for all other Type B and C penetrations', the combined leakage rate;is less than or equal to 0.60 L,. b. At least' once per 31' days by verifying that all primary containment penetrations ** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident * conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provideo in Table l 3.6.3-1 of Specification 3.6.3. l
,
c. By verifying the primary containment air lock is in compliance with ! the requirements of Specification 3.6.1.3. I d. By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
"See Special Test Exception 3.10.1
**Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed, or otherwise secured in the c'iosed position. These penetrations shall be verified closed during each COLD -
SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days. - - b LIMERICK - UNIT 1 3/4 6-1 t
- -
. . .-- _._ . .. -- . _ ... ... . ..=
- ' -
-- -
. . . . .. .. .. .. ...-. - .. .- . - .- . - --
.F
-
~ " * " *
, - ; : . .g.; .
*
- ~p * *~-
. [ . T*[ ., :-
-
. .; =~ y q_. . Q *
_ _
----
. .._ .. _ -
_._.. _
. _ _ ._ _ _
-
.. .- - ~ ---.
-
- -
_ _ _ _
-- -
__ _ _ . _
' - .. _._ .. _-. .
'
~
CONTAINMENT SYSTEMS
-
--
~ ~ ' '
-- _I
- ~ ~ - - -
PRIMARY CONTAIN5ENT LEAKAGE._ __ _.__ - _ .
- - LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
a An overall , integrated leakage t' ate of lest than or equal to L,, 0.500 percent by weight of the containment air per 24 hours at P,, 44. 0 psig. .
, ' '
5. A combined . leakage rate of less than or equal to 0.60 L, for all j penetrations and all valves listed in Table 3.6.3-1, except for main steam line 1 solation valves" and valves.which are hydrostatically
~
tested per Table 3.6.3-1, subject to Type B and C tests when pressurized to P,, 44.0 psig.
'
~
c. "Less than.or. equal to 11.5 scf per hour for any one main steam . line through the isolation valves when tested at P t 22.0 psig.
d. A combined leakage rate of less than or equal to 1 gpm times the i total number of containmer.t isolation valves in hydrostatically
'
tested lines which penetrate the primary containment, when test'ed at 1.10 P,, 48.4 psig. .
,
APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY 'is required per . 3pecification 3.6.1;1. . ACTION: .
. . '
With: . .
-
a. The measured overall integrated primary containment leakage eate exceeding 0.75 L,, or 1 b. The measured combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves" and valves which stre hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests exceeding 0.60 L,, or c. The measured leakage rate exceeding 11.5 scf per hour for any one main steam line through the isolation valves, or l d. The measured combined leakage rate for all containment isolation ; valves in hydrostatically tested lines which penetrate the primary l containment exceeding 1 gpm times the total number of such valves, restore, i
. 4 a. The overali integrated leakage rate (s) to letss than or equal to -
*
0.75 L,, and
- ;
'
* Exemption to Appendix J of 10 CFR Part 50. -
LIMERICK - UNIT 1 3/4 6-2 F.UB B 135 .,__,t
. . . - - . . . . -
w - ,-_ w n . --- c.
Z
--- --- ' - - - - - - ~ - -
_ _ . . - . _ ._
~
l' n _ r .- .
'-
.
~
3-Sf2&80420
~ '" "NTAfNhNTSVSTEMS __
_
. , _ ,
-
-
3'! TING CONDITION FOR MERATli)N;(Continued) .
._
~
- -
ACTION: (Continued) ' _.
_
~
b. The comoined leakage rate- for 'all penetrations and all valves listed ~
- -
in Table 3.6.3-1, except for main steam line isolation valves" and - valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests to less than or equal to 0.60 L,, and c. The leakage rate to less than or equal to 11.5 scf per hour ~~" for any ~ one main steam line thro _ ugh the isolation valver, and d. The combined leakage rate for all containment isolation valves in
-
hydrostatically tested lines which pentrate the primary containment to less than er equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200*F.
SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criterh specified in Appendix J of 30 CFR Part 50 using the methods and provisions of ANSI 45.4-1972 and BN-TOP-1 and verifying the result by the Mass Pcint _ Methodology described in ANSI N56.8-1981: a. Three Type A Overall Integrated Containment Leakage Rate 7.ests shall be conducted at 40 210 n.onth intervals during shutdown at F,, t.4.0 psig, during each 10 yece service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection.
b. If any periodic Type A test fails to meet 0.75 La, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L3, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L,, at whict. time the above test schedule may De resumed.
c. The accuracy of each Type A test shall be verified by a supplemental test which: 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,. The formula to be used is: [L, + L ,- 0.25 L,] 5, L c
$ [L, + L,, +'O.25 L,] where L, = supplemental test result; L,=
superimposed leakage; L ,= measured Type A leakage.
2. Has doration sufficient to establish accurately the ch3nge in leakaga rate between the Type /4 test and the supplemer. cal test.
3. Requires the quantity of gas injected into the containment or . bled from the centsinment during the supplemental test to be between 0.75 L, ^d 1.25 L,,
. "Exemption to Appendix "J" to 10 CFR Part 50.
AUS 81985 LIMERICK - UNIT 1 3/4 6-3
._. .
-
-
-
--
_
.. _ . . - - .
-
-
-
.- -.
-
T __ a- 2 .- -
. _
-
..
--
.-
^3862080420 2 1
-
-
CONTAINMENT SYSTEMS - - _ __ _
-
_l
--- =. . ._
. 1
~
~ SURVEILLANCE REOUIREMENTS (Continued) - :- -
'
. T ( l
-
_ . _
._. . -
d. Type B and C test ~s shal.1 be condu:xed with gas et P,, 44.0 psig* _
. . . . .at intervals no greater than 24 months ** except for tests involving: l 1. Air locks, l ' 1 2. Main steam line isolation valves, _. __
3. Containment isolation valves in-hydrostatically tested lines --
-
which penetrate the primary containment, and e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
f. Main steam line isolation valves shall be leak tested at least once per 18 months.
g. Containment isolation valves in hydrostatically tested lines which penetrate the primary containment shall be leak tested at least once per 18 montns.** I h. The provisions of Specification 4.0.2 are not applicable to Specifica-tions 4.6.1.2a., 4.6.1.2b., 4.6.1.2c., 4.6.1.2d., and 4.6.1.2e.
.
.
.
"Unless a hydrostatic test is required per Table 3.6.3-1.
- A Type C test interval extension to Mayf26, 1986 is permissible for primary containment isolation valves identified by an asterisk in the inboard and outboard isolation barrier columns of Table 3.6.3-1, Part A, as discussed in Application for Aaendment of Facility Operating License dated December 18, _
1965. , 4 e MAR 3li!i
,
LIMERICK - UNIT 1 3/4 6-4 Amendment No. 2 l
. - -
i
~ ~
z ._'-
. . . . . . . . _ = w.w.-_ _ _:.-. . .- . . _ _ - . - . . _ .
,
. , . . . _
-_ . . - _ .
. ..
-
_
.- : - . . - - -- . ._
_
~ ~~
-
-_--
CONTA!hMENT SYSTEMS - ._ __-_. _
'
^" -
-
MSIV LEAKAGE CONTROL SYSTEM -
,
-
'
'
' ~
LIMITING CONDITION FOR OPERATION 3.6.1.4 Two independent MSIV leakage control system (LCS) subsystems shall be OPERABLE.
~~~ ' APPLICABILITY: OPERATIONAL CONDITDNS 1; 2, and 3.
. . ACTION:
'With one MSI'/.leakige control system sub' system inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in' COLD SHUTDOWN within the following,24 hours.
SURVEILLANCE REQUIREMENTS - 4.6.1.4 Each MSIV leakage control system subsystem shr*,1 be demonstratad OPERABLE: a. At least once per 31 days by: 1. Stari.1:4 the blower (s) fros the control room and operating the blower (s) for at least 15 minutes.
2. Energizingtheheatersandverif91ngatemperatureriseindicat-ing heater operation on downstream piping. 1 b. During each COLD SHUTDOWN, if not performed within the previous 92 days, by cycling each motor , operated va've through at least 'one complete cycle of full travel... .
. .
.
*
c. At least once per 18 months by: 1. Perfonnance of a functiona'l test which includes simulated actua-tion of the subsystem throughout its operating sequence, and * verifying that each interlock and timer operatas as designed, each automatic valve actuates to its correct position and the blower starts.
2. Verifying that the blower (s) develops at least t.se below l required vacuum at the rateid capacity: a) Inboard valves, 15" H2O at 100 scfm.
b) Outboard valves, 15" H2O at 200 scfm. -
~
d. By verifying the opereting instrumentation to be OPERABLE by perfonnance of a: l 1. CHANNEL CHECK at least once per 24 hours, ,
*
2. CHANNEL FUNCTIONAL TEST at least once per 31. days, and 3. CHANNEL CALIBRATION at'least once per 18 months. , j
.
.
LIMERICK - UNIT 1 3/4 6-7 ' E 6M-** "'--
% ,
l
.
. . - . - . . - _ . . ... ._ . ..... . - - -
- . .- ... .. . - . . _ . _ . _ _ _ _ . .
l
- -
_ - - ---
.. , _ _ __
' ~ - ~
, .. .. - .. - -. ~ -- : .- -- ~ : - = Y-
. - - ~" - =
~, ~ . . , . , . . . . . .
f~ ' _ _ al? ' ' _
-' - - ~
-
~
- - . . .
-
- -_'
. _ _ _
,. ~._'-'"
~
_ _.
-
-
..
-
._ _ .
.
_ _
. _ . _
'
.
,
-
. _ . _ ._ CONTAINMENT SYSTEMS ~
~
_.
- _
-
.
'
(.
_
- . -
. _ _
DF.IMARY CONTAINMENT STRUCTURAL INTEGRITY _i ._
. - -
_.
_ LIMITING CONDITION FOR OPERATION
,
*' 3.6.1.5 The structurtl integrity of the, p,rimary containment shall be maintained at a level consistent with the acceptance criteria in. Specification
*
.
4. 6.1. 5., APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
- ACTION: .
. .
With the structural inteprity'of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits witnin 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDO='N within the following 24 hour,s.
SURVEILLANCE REOUIREMENTS
. .
4.6.1.5.1 The structural integrity of the exposAd accessible interior and i exterior surfaces of the primary containment, including the liner plate, sna11 be detersined during the shutdown for each Type A containment leakage rate test by a visual inspection of those surfaces. This inspection shall be perfomeo
.
prior to the Type A containment leakage rate test to verify no apparent changes ~ in appearance or other abnormal oegradation. - .
.
. .
,
4.6.1.5.2 Reports Any abnomal degradation of the primary containment structure cetected during the above required inspections shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within - 30 cays. This report shall include a description of the condition of the liner and concrete, the inspectf on procedure, the tolerances on cracking, and the corrective actions taken.
.
~
!
i
.
.
.
.
*
(
._
LIMIF.ICK - UNIT 1 3/4 6-8 2 E !!U - -- %
.
_.
. . .... _ . .. . .. . . . . . . . _ _ . . . .
-
' ........-.:- - = :.- . = .:.: = ~ r ..:: ~ ~ ~.. . .~. :~'". - ,:.~.~ . L .. ..? ~. ... ,.. - . * J -- '.. .- ..
_-. . . - . _a - - - .
.--
_
~ ~ ~
_
..._-- . . . -
_ _ _ _ __
-- ._._ -
_ CON;AINMENT SYSTEMS
'T --
,__
.
.
..
_ __
.. . . - ._DRYWELL '-
AND SUPPRESSION CHAMBER INTERNAL PRESSURE -
.
- .
*
LIMITING CONDITION FOR' OPERATION
"
3.6.1.6 Drywell and suppression chamber internal pressure shall be maintained
- _between 0.0 and +2.0 psig. l
-
.
. APPLICABILITY: OPERATIONAL CONDITIONS 1,.2, and 3. -
ACTION:
.
With the drywell and/or suppression chamber internal pressure outside of the specified limits, restore the internal pressure to within the limit within 1 hour or be in at least HOT SHUTDOWN within the ~next 12 hours and in COLD SHUTDOWN within the following 24' hours.
'
. ..
.
-
.
.
SL'RVEILLANCE REOUIREMENTS 4.6.1.6 ' The drywell and suppres'sion chamber internal pressure shall be - determined to 'be within the limits at least once per 12 hours.
.
. ,
,
,
i
[
i
.
.
- -
.
.
' ($ Sliii LIMERICK - UNIT 1 3/4 6-9 - t
,
. . . . . . . . . . - _
-
. . . . . . . . . . . -, . .. . . .. . . _ - _ . . - . -
.-. .
. .._.. ... . . . - . - - . .
.- .-- , .- -- * - . - , . -
. .... .. .- _
_.
~--. .,:.... .. . - . _ , - . 2.,,_...._..-....a.......,... 1..-.-.- . _ __ . :_. - .-- --
~ ~
_
,_-- _
7 -- _ _ _ .
- - ._
~
CONTAINMENT SYSTEMt-
~
..
__.
~~ - __ ( '~
-
_ __ _ ,_ _
- -
_ .-
- '
-
DRYWELL AVERAGE AIR TEMDERATURE - .
-
-
,
LIMITING CONDITION FOR OPERATION
.. -. ,
3.6.1.7 Drywell average air temperature shall not exceed 135'F.
.
*
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. .
.
ACTION: With the drywell Everage alf temperature greater than 135'F, reduce the average air temperature to within the limit within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
l l SURVEILLANCE REOUIREMENTS
.
.
4.6.1.7 The drywell average air temperature shall be the volumetric average of the temperatures at the following locations and shall be deterinined to be * within the limit at least once per 24 hours: .
..
.
Elevation Azimuth * a. 330' 45', 50', 225* * I b. 320' 105', 225*, 345' c. 260' 50', 165*, 285* d. 248' 11', 74', 150', 182', 25?', 337'
.
. l
*
*At least one reading from each elevation is required for a volumetric l average calculation. ,
.
.
-
3/4 6-10
-
LIMERICK - UNIT 1 n'3 6 US$ t
. . . ..._- -__. ..... . . . . .. ..... . . . - .. ..... . - - .......
. . ~ . - . .
a.-.: .
-
..-. .
-
. . . . . . . . . - - . --. .. _
. . ..____ ...
. .. - _ . . . . _ . _._
.; 4y. - -
_ _
-
~~ ~
~~' ~
-~
-.._. __ _
.
_
-
.
== ,- .
- --
- ~ ~~
--
-
CONTAINMENT SYSTEMS ~ _ 3/4.6.3 PRIMARY CONTAIAMENT ISOLATION VALVES , __ _
. ~
LIMITING CONDITION FOR' OPERATION - 3.6.3 The primary containment isolation valves and the reactor instrumentation line excess flow check valves shown in Table 3.6.3-1 shall be OPERABLE with isolation times less than or equal to those shown in Table 3.6.3-1; APPLICABILITY: OPIRATIONAL CONDIT1'ONS 1, 2, and 13. _
' *
ACTION: a. With one or more of the primary containment isolation valves shown in Table 3.6.3-1 inoperable, maintain at least one. isolation, valve OPERABLE in each affected penetration that is upon and within 4 hours either: 1. Restore the inoperable valve (s) to OPERABLE status, or 2. Isolate each affected penetration by use of at least one de-activated automatic valve secured in the isolated position," or 3. Isolate each affected penetration by use of at least one closed.
r.anual valve or blind flange."
4. The provisions of Specification 3.0.4 are not applicable provided that within 4 hours the affected"penetration is isolated in , acccrdance with ACTION a.2. or a.3, above, and provided-that ' the associated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are performed.
. Otherwise, be in. at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN wittiin the following 24 hours.
' b. With one or more of the reactor instrumentation line excess flow l check valves shown in Table 3.6.3-1 inoperable, operation may l continue and the provisions of Specifications 3.0.3 and 3.0.4 are
-
not applicable provided that within 4 hours either: 1. The inoperable valve is returned to OPERABLE status, or 2. The instrument line is isolated and the associated instrument is declared inoperable.
Otherwise, be in at least HOT SHUTDOWN,within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
l i l
* Isolation valves closed to satisfy t%se requirements may be reopened on an intermittent basis under adminiF' rative enntrol. -
.
.
LIMERICK - UNIT 1 3/4 6-17 AU6 ~819!5 i
. .
y.e. g e6 * mg e e p *e - 6N** * * " . * ' e. m ee e . G mammbe og. g
.e+ .O
_ _ . . _ _ _ _ . _ _ __ -
:--- ~ * ^- ' ~
~~ -
. . . . . . . . . . - - - . -
_ , _
;- _.
,
. -.. _ :. .. .. -... . - . . . . . . . . . - - . _
. .< ,
-
a _ _
- - .
-- -
- * ~ -
-
-
T'~ --
.
_ _ _
--
_ m- _. .
- -
_ _
(_
- - ~~
- - -
_ CONTAINMENT' SYSTEM 5 - .. _ _
~ ~ ~ --
SURVE!LLANCE REOUIREMENTS'
-- _
_ .
. -
4.6.3.1 Each primary containment isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve to service after mainte-nance, repair'or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at_least one complete cycle of full travel and verifying the specified isolation time.
4.6.3.2 Each primary containment automatic isolation valve sh%n in .
, , Table 3.6.3-1 shall be demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 18 months by verifying *that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.
4.6.3.3 The isolatio'n time of each p'rimaiy containment power operated or i automatic valve shown in Table 3.6.3-1 shall be deterinined to be within its : limit when tested pursuant to Specification 4.0.5.
4.6.3.4 Each r'eactor instrumentation line excess flow check valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE at least orce per 18 months by . verifying that .the valve checks flow, i 4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE:
'
. . j l
a. At least once per 31 days by verifying the continuity of the explosive i charge. . b. At least once per 18 months by removir.g the explosive squib from the explosive valve, such'that each explosive squib in each explosive .. .. valve will be tested at least once per 90 months, and initiating the ! explosive squib. The replacement charge for the exploded squib shall l
'
be from the same manufactured batch as the one fired or from J.'.,inet batch which has been certified by having at least one of that u tch . successfully fired. No squib shall remain in use beyond the expirat4n of its shelf-life and/or operating life, as applicable.
I
.
*
.
~
-
.
l l l LIMERICK - UNIT 1 3/4 6-18 - g g g- *
- - - .. . . . - . . - ---.- ..- , ~ - .. .. . . . - . . . - . - . . . - . . . - , , . - , , , _ _ , , _. , - . , . , . _ - _ _ ,
_, _.__,_,..,y _ . - , ,
, , ..
.' I l '
3862080420
-
TABLE 3.6.3-1 . 1,. PART A - PRIMARY CONTAINMENT ISOLATION VALVES , l'i
,
INROARD OllTBOARC . ISOL. t ; M ISOLATION ISOLATION MAX.ISOL, SIGNAL (S), i NOTES;' P&ID
* PENETRATION FUNCTION ' BARRIER BARRIER TIME.IF APP. IF APP.
NUPEER ',
(SEC)(26) (20)
y .-
)
i
--e NA 59 w 0038 CONTAINMENT INSTRUMENT 59-1005B (CK) ,
HV59-1298 C ,II,5 I GAS SUPPLY - HEADER 'B' / l t NA 003D-2 CONTAINMENT INSTRUMENT 59-1112*(CK) 59 HV59-151B* 45 M l CAS SUPPLY TO ADS VALVES I il .
, ,
5* C.D.E.F,P,Q 6 I41 007A(B,C,0) MAIN STEAM LINE IIV41-1F022A j
A'(B,C,0) (B,C,D) ,
IIV41-1F028A 5* C,D,E F.P,Q 6 I.
i ! I (B,C,0) i w IN40-If0018 45 EA 6' i ,, I' h (F,K,P) m NA G,1 ',
-
(XV40-101B '
5'
,'. (F,K,P)
-
'
* ' , .
SEE PART B,
!
TilIS TABLE) ' f li s 30 C.D.E.F,P.Q ' 4 41 MAIN STEAM LINE DRAIN llV41-1F016 C D,E,F,P,Q 008 !!V41-1F019 30
-
,
-
41-If010A(CK) N/. 4 1 009A FEEDWATER ,i HV41-1F074A(CK) HA I 41-1036A(CK) NA . , llV41-1308 45
,
i '
' ilV41-133A 45 .
E
%
llV41-109A HV41-1F032A(CK) NA NA li 32 l
'"z E ~ IIV55blF105 30 ; 7 '
,
5w IN44-If039(CK) NA I
*
(X-98) ,
31: . m @ 41-1016(X-98, NA j ; a : X-44) ,
-
<
i
., i, . , _ _ _ _ _ _ _ _ -
'
.
tl.
386h080420
-
;
TAOLE 3.6.3-1 (Continued)
,
, i ,
i; F' ART A - PRIMARY CONTAINHENT ISOLATION VALVES
*
>
'
, l 'I ,
e INBOARD OUTBOARD ISOL.
p ISOLATION ?%0LATION MAX.ISOL. SIGNAL (S )', HofES PAID PENETRATION FUNCTION TIME.lf APP. IF APP. ,. , BARRIER B"RRIER ,: 8 c NUMBER *
,
iSEC)126) (20) f _
" I NA l3 41-1F010B(CK) t 41 w 0098 FEEDWATER
!!V41-1F074B(CK) NA
.
41-1036B(CK) NA l* ' IIV41-130A 45 , ilV41-133B 45 l
'
IIV41-1098 NA 32' i , HV41-1F032B(CK) NA ' LFCC l llV49-1F013 23 ilV44-1F039(CK) NA l
'
(X-9A) 3k 41-1016(X-9A, NA il
.
w
,
l.
X-44) .
. -
1 .
{49
, 7. 2 * K, KA 5 Ifv49-If007 K, KA I E 010 RCIC STEAM SUPPLY HV49-1F008 7.2* -
HV49-1F076 45 K, KA l
; (
'
12* L, LA 5I 55 ; HPCI STEAM SUPPLY HV55-1F002 L, LA 011 HV55-IF003 12* 45 L LA >
>, li
' HV55-1F100 ,
,
-
100 A,V 9,22 251' IIV51-1F009 {3 012 RHR SHUTOOWN COOLING PSV51-155 NA A,V I lg SUPPLY * IIV51-If008 100 ! 11 i 9,22 51 g NA A,V HV51-1F0504*(B*)
" 013A(8) RilR SHUTDOWN COOLING (CK) A,V
-
F RETURN llV51-151A*(B* )
A,V I I '
*
" HV5}-1F015A(B) l B.J,Y 44'
10* RWCU - SUCTION HV44-1F001* ; B.J,Y ;
$ 014 ' HV44-If004* 10*
f j n
'
i I
.. . _ - __ _ _.
i
'
I
, t
^
i !'
, l' j
* .
I
,
3874086670 ,
.
,
TAllLF 3.6. .l-1 (Cc.nt inued) / , r- I: y PART A - PRIMARY CONTAINHINI 150 tall 0N VALVES
'
! ,
O l,
;q IN8OARD OUTBPARD 150L. , * PENETRATION FUNCTION ISOLAIION 150LAll0N MAX. IS08.. SIGNAL (5), jNOTES~
P&l0 I
*
faseER BARRIER BARRl[R TIME.lf APP. IF APP. l E ___ _ (SECJ(26) [20) _ i .
'
i M y 016A CORE SPRAY INJECTION llV52-1F006A(CK) NA 9.22 52 HV52-1F039A 7 9,22 i
'
liv 52-IICOS In l 0168 CORE SPRAY INJECTION HV52-1f006B(CK) NA 'l9,22l 52 ' - l . ' liv 52-If 0398 / y,22 IIV52-108(CK) flA i
*
lI ; l ' RPV HEAD SPRAY HV51-IF022 60 AV l 4,9,22 51 i 017 NA 9,224 PSV51-122 , ; HV51-IIO23 135 AV
$ NA , 15 - ,
SERVICE AIR TO DRWELL 15-1140 ' h 021
'
i 15-1139 NA ,
,
- .
4 w ' 022 DRWELL PRESSURE IIV42-14 7C 45 10 42 [ :i
{
INSTRUMENTAT!ON ' HV13-106* 40 11,28, 13 I,l 023 RECW SUPPLY TO 29 RECIRC PUMPS 11,28 i IIV13-In8" 30
; 29j I
llV13-109* NA 11,13, ; F 40 4 11,28{ 13 l IIV13- 101" ,
$ 024 '
RECW RETURN FRON i i29
'l RECIRC PUMP 5 30 t11,28 I, * llV13- 111 *
z 129
*R I I', I .) t I ilVI)-110* NA ,
jf , F%& I'
.. . .
i i lk' l: 4 =i >
' I
'
i
< ,
i; . 3874086670 t i,
.
IABLE 3.6.1-1 (Continued) , g
'
h PART A - PRIMARY CONTAINM[Ni 150LAT10N VAtVf5 ,
' '
I 1 IN80ARO OlllB0ARD g50t, l x PENE1 RATION FUNCTIGN ISOLAil0H ISOLATION MAX.150L. SIGNAL (5), NOTES . P&l0 , '
*NlsSER BARRIER BARRIER IIME.If APP. IF APP. l g (SEC)(76) (70) . "
DRWELL PURGE SUPPLY HV57-121(X-201A) 5** 8,H,5,U,W R.T 3.11,14l 57
" 025 HV57-123 5** R.H,5,U,W,R,T 3'11,14 IIV57- 163 9 3fil,14 i IIV57-109 6**
B,il,R,5 8,H,5,U,W,R,1 11 ! l, (X-201A) IIV57-131 $** 8,ll ,5',U ,W . R . T 11 lI l
(X-201A) ;
.l i (
'
<
HV57-135 6** 8,H,5,U,W,R T 11 g 026 ORWELL PURGE EXHAUST HV57-Il4 5** 8,H,5,U,W,R T 3.11,14.33 57 w IIV57-Ill 15** 8,H,5,U,R I 5,11: * HV57-161 B,H,R,5 3,11,14 ' I
) S ^ . SV57-139 5 i 8,H,5',U,W,R,T 10 !
IIV57-ll5 6** 11,33
" IIV57-117 5** 8,H,5,U,R,T 11 [ l,
'
'V57-
, 145 5 B H,R,5 11 l
'
'
027A CONTAIMIENT INSTRUMENT 59-1128(CK) .NA IIV59-151 A 45 M * , GAS. SUPPLY TO A05 VALVES
; j ,
H.M.S5
,
HV43-If019 10 B,0 " [43 028A-1 RECIRC LOC? SAMPLE B,0 'I HV43-lf020 10 i i
'
h
"
i SV57-132 5 B.H,R,5 11 57 ( 028A-2 DRWELL H2/02 SAMP.LE 8,H R,5 11 SV57-142 5 j
'
B II.R,5 57' sE
~~ . 028A-3 DRWELL H2/02 SAMPLE SV57-134 SV57-144
5 B.II.R,5 11'
,
, l al ' i
,
:
'
l I' .i
: '
.
'
386208042 ' TABLE 3.6.3-1 (Continued) '
.
1 I l t
', 8 PART A - PRIMARY CONTAINMENT ISOLATION VALVES I
*
:= ISOL.
INBOARD OUTBOARD E ISOLATION MAX.ISOL. sir,NAl(S) . NOTES
* PENETRATION FUNCTION ISOLATION I'P&ID
;
BARRIER BARRIER TIME.IF APP. If APP. '
' NUPSER
*
*
(SEC)(26) (20)
E ,
,
U SV57-133 5 B,il,R,5 11 ! 57 0288 DRYWELL 112/02 SAMPLE B,II.R,5 11 SV57-143 5 l l.
SV57-195 5 B,II,R,5 11, ;;
.l IN42-147A 45 42 '
0308-1 DRYWELL PRESSURE l 10 . *
1 , INSTRUMENTATION s ; . NA
' ,' 59 035A TIP PURGE 59-105G(CK) i !
(DOUBLE "0" RING) B,II,5 16 .
IIV59-131 7 ll
.
NA B ,Il 11,16.21
59 l TIP DRIVES XV59-141A-E
; 1 035C-G (DOUBLE "0" RING)
,
'
. XV59-140A-E NA ., 11,16 4 I 47'
'
NA . 12 CR0 INSERT LINES BALL CllECK 12 :
037A-D HCU HA ,
~,
'
'
NA 12, 47 IlCU ' 038A-0 CRD WITHDRAW LINES 25 30 XV47-If010 ! SDV VENTS & DRAINS 30 30 XV47-If180 , ' XV47-1F011 25 I 30 30 30 XV47-1F181
!
l 160 4,11 51 DRYWELL SPRAY llV51-1F021A(B) i 11 , } ', 039A(B) IIV51-]F016A(B) 160 ,i
< , j
*
ilV42-147D 45 ' 10 42 ' 040E DRYWELL PRESSURE , INSTRUMENTATION 'g ,
.
,
<=
'
45 C,ll,5 5 ' 59 * CONTAllMENT INSTRUMENT llV59-101 C,H,5 ,I ,-
- 040F-2 llV59-102 7
' * GAS -SUCTION ,,
s ( __ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ .. .
.
i l
'l TABLE 3.6.3-1 (Continued) .
i PART A - PRIMARY CONTAINHENT ISOLA 110N VALVES i 386208042C ,
.
{
m l I l OUTBOARD '150L. i 5 IhPGARD MAX.ISOL. SIGNAL (S), NoiES ' P&I0' FUNCTION 150LAT10N ISOLATION i E PENETRATION IF APP.
TlHE.IF APP.
' BARRIER BARRIER NUMBER (SEC)(26) (20) ~h g
- .-
i'I 60-1057 NA * 5,11' 6 60
" 040G-1 "
ILRT DATA ACQUISITION 60-1058 NA , 11 ' i' 5,11 i 160 l' i l 60-1011 NA 040G-2 ILRT DATA ACQUISITION 11
'
60-1070 NA ' lI. , fig' 'i 59-1005A(CK) NA ' 040H-1 CONTAIMENT INSTRUMENT IIV59-129A 7 C ,11,5 GAS SUPPLY - IIEADER ' A' ; 48-1F007(CK) NA 48i 042 STAND 8Y LIQUID CONTROL 60 29 (X-Il6) IIV48-If006A l l . w 3 10 8,0 41 0438 MAIN STEAM SAMPLE ffv41-Ir084 ' i HV41-Ir085 10 8,0 T
'
' '
! i 7I ,
41-1017 NA i i 5,31 ,41 044 IlWCU ALTERNATE ,' 41-1016(X-9A, MA ; RETURN : X-98) I PSV41-ll2 NA . ,
..
NA l l9,2/ ! LPCI INJECTION 'A'(6,C,D) Ifv51-lF041 A*(8,C* , 045A(5,C,D) I 3 D*)(CK) I 9,22
.
i , HV51-142A*(8,C*, 7 ,
,
E
" D*) i llV51-IF017A* 38 k"
,
!
,
(8,Ca,0*) ,
1 IT IIV42-1478 45 10 - 42 I 050A-1 DRWELL PRESSURE ' - ' N INSTRUMENTATION A l m I C.H 11 87
'
IIV87-128* 60 053 DRYWELL CHILLED WATER 60 11,28, ;
'A' ilV87-120A* ,
u SUPPLY - LOOP 29
:3 ,, 31V87-125A* 60 ,11,,28,29 ,[ *,
'
~
.
,-
.
i ;- ,
;
_ . _ _ _ _ _ _ - _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
'
i f I i:
, i I
1l I 387408667d ,
,
i T ABL E 3.6. 3-1 (Cont inucil) , l *; 4
,
PART A - PRIMARY CONTAIHttfHi 150lAll0N_VALVf5 '
=
E IN8OARD OUTBOARD ISOL. i .
, * PENETRATION FUNCTION ISOLATION 150lAll0N MAX.ISOL. SIGNAL (5), NOTE 5j P3.lD.
'
'
MSSER BARRIER llARRl[R IIMI.lf APP. If APP.
E (SEC)(26) (20)
-
,
N ! I
'
~ 054 DR M LL CHILLED WATER HV87-129* 60 C,lt 11 , 87 RETURN - LOOP 'A' ilV87-121A* 60 11,28,.
'29 '
l ' ' ' ilV87-124A* 60 11,28,' 2l l 29 - ' I i I
;iV87-122" 60 C.H 11 87 i 055 DRWELL CHILLED WATER '
SUPPLY - LOOP 'B' lN87-1208" 60 11,28,> I I
IIV87-125fl* 60 ;11,28,29 l l s IN87-123a 60 C.H' 11 87 5056 DRWELL CHILLED WATER ilV87-121na 60 ,11,28,29' l , RETURN - LOOP '8' 11,28,29 I I HV87-124H" 60 ,r
"us '
NA 15 061-1 RECIRC PUMP 'A' SEAL 43-1004A(CK) '
(XV43-103A - NA i 1 i PURGE '
SEE PART 8, .
' I illIS TABLI) f
, NA 15-
; 061-2 RECIRC PUMP '8' SEAL 43-10048*(CK)
NA
;
'
43 ll PURGE (XV43-10311 -
{* Sif PARI II,
.
[ ,
3 Illis T Af!LI) ', ! f. .
't I
S ,
[
; N 8,II R,5 11 57 SV57-150(X-220A)
,
'
"" 062 DRYWELL H2/02 SAMPLE 5 8,II.R,5 11 SV57-159 g RETURN, M2 MAKE-UP (X-220A)
1 '
*
30** II,II, R ,5 11
* ,
IIV57-Il6 C i (X-220A) fl .II . R ,5 11 j,, sv5 7- 190 5 T
* (X-270A) - ,
L
.,
,
l
- __ -__ __. .-. _
I
:' i
,
'g i
i
{ r
.
I i 3874086670 l
'
IA8tE 3.6.3-1 (Continued) l ; {' , j
!
C
PART A - PRIMARY CONTAINMENT ISOLATION VALVE 5 , x INBOARD OUIBOARD isot, l i[
'
x PENETRATION FUNCTION 15ULAll0N '.,0LATION MAX.150L. SIGNAL (5), NOTES P&lD
* NLSSER BARRIER BARRIER ilME.lf APP. IF APP. i ,
(SEC)(26) (20) j'
c
[
ew
* SV57-191 5 B,H R,5 11
>
" (X-220A) l i, .
'
48-lF007(CK) HA 48 l 116 STAND 8Y LIQUID CONTROL l t I (X-42) HV48-ITC068 60 79 ,
,
I ,l DRYWELL RADIATION SV26-190A 5 8,H,R,5 11 26 i , 1178-1 MONITORING SUPPLY Sv26-1908 5 8,II,R,5 11 l
'
{'
Sv26-190c 5 8,H,R,5 11 <26 l 1178-2 DRVWELL RADIATION SV26-1900 5 8,H.R,5 11 ! l w IGNITORING RETURN A IIV57-124 5"* 8,H,5,U,W,R.T 3,Ik.14 57 i - 201A SUPPRESSION POOL PURGE 8,H,5,U,W,R.T HV57-131(X-25) 5** 3.11,14 ; 4e
$UPPLY llV57-164 9 B,H,R,5 3.11,14 3 IIV57-109(X-25) 6** 8,H,5,U,W.R.T 11 ! ';
ilV57-14 7 6** 8,H,5,U,W.R,1 11 IIV57-121(X-25) 5 * *. 8,H,5,U,W.R.T 11
, ,.
HV57-104 5** 8,H,5,U,W.R.T 3.11,14.33 57 f; 202 SUPPRESSION POOL PURGE 8,H,5,U,R,i 5,11 1 IW57-105 15** l EXHAUST IN57-162 9 8,H,R,5 3,11,14 tE 8,H,5,U,W.R.T ,11, 33
" HV57-Il2 6** I IIV57-Il8 5** 8,1,5,U,R,i 11 =
SV57-185 5 8,II,R,5 11 , g I 4.22, .i 51 I l I 203A(8,C,0) RHR PUMP SUCTION llV51-II004A(B. 240 19,29
* C,D) ,
o 22 1, j P5v51-lF030A(n. NA ' r 4) j; ______ ____ _________________________ __ - ___________ -__ -__ ______ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
'
'
386208042b' . i' TADLE 3.6.3-1 (Continued) l ,
{.
l h PART A - PRIMARY CONTAltlMENT ISOLATION VALVES l3 i 9 INDOARD OUTBOARD ISOL. 1 i5 M ISOLATION ISOLATION MAX.ISOL. SIGNAL (5), NOTES ! P&ID
^ PENETRATION FUNCTION ' '
NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
.- (SEC)(26) (20) E '
,
. , i
'
llV51-125A(8) 180 4,22,29 51 .
"[204A(B) RilR PUMP TEST LINE AND CONTAllmENT COOLING
! ,
,
llV51-1F027A*(B) 45 CG 11 , 51 ! l 205A(8) SUPPRESSION POOL SPRAY IIV52-1F001A 160 6 4,22,29 52 206A(8,C,D) CS PUMP SUCTION (8,C,0)
*
IIV52-1F015A(B) 23 C.G
$
5,22' 52 207A(8) CS PUMP TEST AND FLUSil
'}
!!V52-1F031B 45 LFCil 5,Ih2,29 52 w 2008 CS PUMP MINIMUM RECIRC i I .
2 ' 4,22 ' 55 HPCI PUMP SUCTION llV55-1F042 160 L.LA
. 209 llV55-1F072 120 4,22,29 55 210 HPCI TURBINE EXilAUST 1 ,
liv 55-1F071 40 B ,Il 4,22 55 212 HPCI PUMP TEST AND FLUSil HV49-1F031 60 l 4, 2,29 49 214 RCIC PUMP SUCTION 215 RCIC TUR8INE EXHAUST llV49-1F060 80 ' .; 4,22,29 45, l IIV49-1F019 8 LFRC 5,,22l * 49 216 RCIC MINIMUM FLOW
'
l ': k i
>
sw
! i'.
s ". y
>
.,
'
FR
,
.) ;
.
'
h I
I , i; .' , j
'
.
_____-___- __ - ___ - - _ _ ______ _ _ _ - . ._. _ -- _ _ _ __ _- _-
,
1 Y. i
#
" 3862000420 i
>
IABLE 3.6.3-1 (Cor tinued) ; PART A - PRIMARY CONTAINMENT ISOLATION VALVES l
.-
2 :o
;'; INBOARD OUTBOARD ISOL.
'
.l j'
{~ ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES 'P&lD
* PENETRATION FUNCTION 11ME.IF APP., If APP.
' NUMBER BARRIER BARRIER l# I~
'
(SEC)(26)_ (20) ;
,
l E ,_
, ~%
llV49-1F002 60 E,29 ' 49 ,1 e 217 RCIC VACUUM PUMP DISCil '
"
49-If028(CK) NA '
, ii .,
-
, P NA ', 159 218 '7NSTRUMENT GAS.TO 59-1001(CK) ' l C ,II . S :
:fv59-135
,
j VACUUM RELIEF VALVES ' 4 . j INSTRUMENTATION -
--
HV55-121 45- 10
'
l 55' i 219A ') i SUPPRESSION POOL .
,
LEVEL -
-- ifV55-120 45 ., 10 55 w 2198 INSTRUMENTATION - i1 -
SUPPRESSION POOL
- ) LEVEL l 9 h,11,RS' il 57 SV57-191(X-62) 5 ,
I l N 220A ' H2/02 SAMPLE RETURN SVS/-190(X-62) 5
'
B,II,R,5 Il i.
HV57-116'iX-62) 30** 8,II,P,5 11 l SV57-150(X-62) 5 8,ll,R,5 l l 11 '
- SV57-159(X-62) 5 B,II,R,5 l 11 ,
i - 10, 57
--
SV57-101 5 'I 2208 INSTRUMENTATION - j I- -
*.
SUPPRESSION POOL PRESSURE SUPPRESSION POOL LEVEL B,il,R,S il 11 I
'
,.7I SV57-181 5 l 221A WETWELL H2/02 SAMPLE 5 B,II,R,5 11 ;
' SV57-141 SV57-184 5 B,H,R,5 11 i i l '; ' i [ " 5 B ,II, R , S . 11 57 WETWELL H2/02 SAMPLE SV57-183 .. B,H,R,5 i. 11 2218 SVd7-186 5 ,
, 3"' l- l'
J. ' i
=
_.
x i * ! , ,
.
i
'
.
'
,
j.
- j ,. i 4
,
j i:
'
i TABLE 3.6.3-1 (Continued) ,
, ; lj:l1-I r l j g '
PART A - PRIt!ARY CONTAINHENT ISOLATION VALVES ' i i l d n INBOARD OUTBOARD - ISOL. . FUNCTION ISOLATION ISOLATION HAX.ISOL. SIGNAL (S), NhiES,iIt.10 P i . '. 7 PENETRATION BARRIER TIME.IF APP. IF APP. l ! - NUMBER BARRIER ISEC)(36) (20) i 'i,I h-4
.__
. li g-i
. 9 225 RIIR VACUUH RELIEF SUCTION llV51-130 60 B.H : 4,11 51 i i
ilV51-131 60 B,11 11 .
,
IIY51-105A 40
'
4,22,29 51 b.. ! i 226A RilR HININUll RECIRC a i ?cj RilR HINIMUM RECIRC llV51-1058 40 4,22,29 '51 i
' li.
226B , . f 60 '" ! 227 ILRT DATA ACQUISITION 60-1073 NA 5 f SYSTEM 60-1074 NA
' l'
,
.i-llPCI VACUUH RELIEF IN55-1F095 40 H,LA 4,11,24 55 .
Y
* 228D 11,24 A llV55-1F093 40 H,LA ' '
T *
,
INSTRUMENTATION - DRWELL
--
liv 61-102 45 1 U 230B
21,2339
,29
, SUHP LEVEL IIV61-112 HV61-132 45 23,29 'l l1l
. .
.- .
I ORWELL FLOOR ORAIN llV61-110 30 B,H 11;,22 6i ' 231A '
,
B ,fl. 11,22 SUMP DISCllARGE . IIV61-111 . .30 j h. . L- , HV61-130 30 B,H 11,22 61 , f.
231B DRWELL EQUIPMENT DRAIN 11l,22 -t i flV61-131 30 B,H -
.
TANK DISCilARGE HV52-1F031A 45 5,22,29l'52 .
!
235 CS PUMP HINIHVH RECIRC ! l $I , i HV55-1F012 15 LFHP
-
I 236 IIPCI PUMP HINIMUM RECIRC 5,d2l
'
: ,I , ,
,. .
r
'
'
:r-
;
! , e
.
* iii'
, . ;
4 ! . l I ,
! '
: 9
l j.I .. -
*
-
-
'
. ; !iit
,
- __ _
'
.
.
;
- e y
! ! *I
'
! r-TADLE 3.6.s-1 (Continued) , I
. .
k PART A - PRIMARY CONTAINMENT ISOLATION VALVES .
'
h !;
' -
p
.
INBOARD OUTBOARO ISOL.
.
, PENE1 RATION FUNCTION ISOLATI0fl ISOLATION MAX.ISOL. SIGNAL (S), NOIES lI P&ID :
, c NUMBER BARRIER BARRIER TIME.IF APP. IF APP. I i
,. ,
*
*
(SEC)(26) (20)- g i I. j,'
i .
' ** 237-1 SUPPRESSION P0OL CLEANUP llV52-127 60 B ,11 4,11.22 52 ,,
.
'
: PUMP SUCTION PSYS2-127 NA -
11,22 . ;i
! IIV52-128 60 8 ,11 , 11,22 i :
237-2 SUPPRESSION POOL llV52-139 45
!
10 . 52' i ;'.i'if i LEVEL INSTRUMENTATION SV52-139 6 10 ' l.!
!
238 RilR RELIEF VALVE flV-C-51-1F104B 18 C,G 51 *! . li DISCllARGE
*
-
PSV51-1068 PSYSI-1F0558 NA NA
19 i
, .
m 19 l
.
} 'PSV51-1018 NA
' '
2 239 RilR RELIEF VALVE IIV-C-51-1F103A 18 . CG .i 1 '. '
, o DIStilARGli PSV51-106A NA , 19 ^ l
' PSY51-1F055A NA i 19 i
PSV51-101A NA 19 . l
-
.
.' '
l %
'
240 RilR RELIEF VALVE ~PSV51-1F097 NA 19 518 '
* l '
DISCilARGE L' ' I . 241 RCIC VACUUM RELIEF liv 49-1F084 40 N KA 4,J1,24 49l i- , llV49-1F080 40 H,KA 11p24 .
,l l , ,
'
;
i il i !. i
.-
, .a n
-
.
-
li i i r.;;
,
' I l
= .
i ;. E ,
! '
t
, ; .
. -l .g ',
; e
-
!
'
: i
'
8 * I - l
,
% .
'1. .. i
: .
-
,;
.
-
- _ _ _ _ _ - . _ - _ _ _ - _ _ _ _ _ _ - - _ _ .
'
.
.
*
. . _/
; - l}: ;;
1 ,
,
l .
'
.i;
'
TABLE 3.6.3-1 (Continued)
-
! ,
-
* . c i
'
,l M PART B - PRIMARY CONTAINf1ENT ISOLATION' EXCESS FLOW CHECK val.VES'
a
-
. .
' -
ISOL. l l R IN00ARD OUTBOARD
' , PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&lD . '
c- HUMBER BARRIER BARRIER TIME.IF APPc IF APP. * l 2.i f
'I; (SEC)(26) (20) _
>
} .
.
I p t ,-
" 003A-1 INSTRitMENTATION 'D' --
XV41-1F070D 1 i 41
'
HAIN STEAM LINE FLOW XV41-1F073D
;, ,
003A-2 INSTRUMENTATION 'A' -- XV43-IF003A 1 43 r s.
l! '- ! i
. RECIRC PUMP SEAL PRESSURE !
003C-1 INSTR. - HPCI STEAM FLOW -- XV55-1F024A. - . 1 iS5 j ,- ,g
;- .
' '
C 003C-2 INSTR. - HPCI STEAM FLOW -- XV55-1F024C 1
' >!
l55 , .) .
'
w
'A' HAIN STEAM --
XV41-1F070A < 41 1
) 003D-1 INSTR.
, e, i LINE FLOW XV41-1F073A j l1 .
.i
- *; a, H
-
007A(B,C,0) INSTR
.
'A'(B,C,D) MAIN
-
(llV41-1F022A(B, 5 C,0,E F,P,Q 6
'
I 41: t
!
' STEAM LINE PRESSURE C,D) SEE PART A (HV41-1F028A 5
*
C,D,E,F,P,Q 6 l
] ! ' TilIS TABLE) (8,C, 0) SEE . l' J!
l 6
!
~
. .PART A ,
! I
' '
.THIS TABLE)
.I
'
"
.
I (IIV40-1F0018 {1 3. { l'
'
, '
'
l
;
'
(F,K,P) SEE '
f.
PART A THIS j.- i i '
! TABLE) . l, F XV40-101B(F, 1 i j~
!
: l; K,P) I
,
' g ('
'
' -, . t i.
~
' ' --
XV42-1F045B 1 42 ' 020A-1 INSTR - RPV LEVEL
- , ,, : > 1 51 ;4'
.'
. E 020A-2 INSTR 'B' LPCI DELTA P --
XV51-1028 . i .. s
.g
* '
. .
.l*-
'
. XV51-lb3B ! 1 j $ 020A-3 INSTR 'D' LPCI DELTA P -- , i ;
, . *
13 l
" ~ 1 ,42 XV42-1F045C I[ .
--
. 020B-1 INSTR - RPV LEVEL 1i 51
*f'
'
0208-2 INSTR 'C' LPCI DELTA P -- - XV51-102C
.
p
. - i I
*
.
,' ,
. , i
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - - - . _ _ - _ _ _ _ - . - - _ _ - - _ _ - . _
i
.
i . i : . ,; j. . .
-
j .'ii i jl !.1,
.
TABLE 3.6.3-1 (Continued) , , C i i g PART 8 - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CIIECK VnLVES : il l-x t t
: U INBOARD OUT00ARD ISOL. )
SIGllAl(5),
'
FUNCTION ISOLATION ISOLATION MAX.ISOL. NOIES PAID J' 7 PENETRATION ' NUMBER BARRIER BARRIER TIME.IF APP. IF APP. l e '; 3 _
(SEC)(26) (20)
-4 w 0278-1 INSTR - IIPCI FLOW --
. XV55-1F0248 ,
1 I 55 ;
'
XV55-1F0240 55 -I 0270-2 INSTR - IIPCI FLOW --
: 1 l
t .' 1,27
'
029A INSTR - RPV FLANGE -- XV41-1F009 41 LEAXAGE
-
'
I k' . i '
, L!
~
l
:
0290 030A INSTR - CS DELTA P INSTR 'D' MAIN STEAM
-- --
XV5'2'-1F018A' XV41-1F071D 1 i
'
,
l 52 I
lk
'
' Y.
5.. <
* '
FLOW XV41-1F072D
{ ,
, ,
T 0300-2 INSTR 'C' HAIN STEAM -- XV41-1F071C 1 41 'l ' j M LINE FLOW XV41-1F072C j
'
. 031A INSTR - JET PUMP FLOW -- XV42-1F059B 1 t, 42. ,'
;,
*
.(JP1) i . j'
j-
'
*
.XV42-1F0590 (JP2) g', .!-
..
' '
*
XV42-1F059F l l l
,
. (JP3) , ,
'
0318 INSTR - JET PUMP FLOW - - . XV42-1F059H I 42
(JP4) -
XV42-1F0518 l
: (JP5) ,
-
w XV42-1F0538 l 31 l . * U '
(JP6) f' . * * . lj l a n - ) i k :I i
l t'
! .; i
'
*
* ;,
,
i i .
' ' . . . , . , #
*
.
'
'
t. .
.
; u
.
-
TABLE 3.6.3-1 (Continued) ,
. t l # : '.
C g
=
I . PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES , , f,i , I 5 IN00ARO OUTBOARD ISOL.
PEN 5iRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), N0iES P& i- ; .i 7 BARRIER BARRIER TIME.IF APP. IF Ar?. .i l
.
NUtt3ER (20) ! E (SEC)(26) :
,
$'-j
.,
. - , '
>
032A INSTR - JET PUMP FLOW -- XV42-1F059M 1 4
(JP6) - <
,
*I XV42-1F059P ' ;-
(JP7)- .
'
i XV42-1F0595 , 14 (JP8) ,
,-
' i INSTR - JET PUMP FLOW -- XV42-1F059U ' 1 42 . i' 032B !
'*
I ! (JP9) ;
'
,
T ' l XV42-1F0510 l '
(JP10)
w XV42-1F0530 i
'
I , ' 1 (JP10) *
-
o i t
.
XV42-1F055 1 42 h ' !' "O 03M. INSTR-PRESSURE ABOVE
--
.
,
I CORE PLATE i i. XV.42-1F076 , g. ,
'
.
' 42
'
XV42-1F061 1
-
--
033A-2 INSTR-PRESSURE BELOW
,
CORE PLATE r.. .! I
-
.
-- XV49-1F044A,C 1 49 i 0338 INSTR-RCIC STEAM FLOW ,
I O .. XV41-1F070C 1 1 42,
- i 034A INSTR 'C' MAIN STEAM --
'
LINE FLOW XV41-1073C l ,
. " --
XV43-1F00SC 1 . 43 3 0340-1 INSTR - RECIRC FLOW .
'
, ,
g XV43-1F010D l II i , i
'
--
'
XV43-1F0090
'
1 43 l M
* 0348-2 INSTR - RECIRC FLOW .
g g XV43-1F010C ; . 't
.
l . l o
*
l , ,. : ' i'
';J
,
. ; ;
'
*!'
,! .
.
'
.
.
.
T
,
.
r
I ii
.,
I l. .
,
TABLE 3.6.3-1 (Continued) , ! L;
-
C l l~. : g , PART B - PRIMARY CONTAINNENT ISOLATION EXCESS FLOW CllECK VALVES
'
,
' ~
' ~ ~
ihBOARD' OU1BOARb' ISOL.
' 7 PENElliAt luN NUMBER PNCTION ISOLATION BARRIER.
ISOLATION
. BARRIER
!!AX. ISOL.
TIME.IF APP.
SIGNAL (S), IF APP.
NOIES P&iD
.,
lh(i c
a ($EC)(26) (20) i i l
~'-
e,. j 9 0404 INSTR - JET PUMP FLOW -- XV42-1F059L 1 42 I
(JP15) .
'
XV42-1F059N i I
(JP17)
-
,
XV42-IF059R ! l ', i (JP,18) ' i 040B INSTR - JET PUMP FLOW -- XV42-1F059G . 1 42
!
*(JP14) ,
, ,
XV42-1F051A l * R
*
(JP16)
XV42-1F053A i
'
I i (JP16)
:
U . I ' 040C INSTR - JET PUMP FLOW --
.
X*.'42-1F059A (JP11)
.'
~4 2 f' f
'l .
, XV42-1F059C , ,
-
, (JP12) i
~
XV42-1F059E i.*
-
(JP13)
. \ .
'8
-- XV42-1F057 1 42 l 0400-1 INSTR - PRESSURE BELOW
& .,
CORE PLATE ' l ' 0400-2 INSTR - RWCU BOTTOM DRAIN-
--
XV44-170 1 , 44. ,
- FLOW XV44-171 i - l'
,. ' I F' . gi
,
si . i .
,,
-
l
. =
i
,
+
. ,
! , e -
,.
.
.
-
I, ,
'
i l
. ' ' - ' - - . .
- -
.
-
.. ,
l .,
.
,
, .
-_ _ - - - _- - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
-, !
.l ..
q' +l'
-
I;[. . TABLE 3.6.3-1 (Continued). ' g
,
l 'l -
~
'
g PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CllECK. VALVES
'
l Q . INBOARD OUTBOARO ISOL.
' . '
; PENETRATION . FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (5), N0hES l P&lU e NUMBER BARRIER ,
BARRIER TIME.IF APP. IF APP. 'l -
-'
(SEC)(26) (20) 'l
_
).
j
'
'
i -
' '>040F-1 ' INSTR - RCIC STEAM FLOW --
XV49-1F044B 1 . 49 I XV49-1F044D l , ,
' -
,
,
04011-2 INSTR 'B' RECIRC -- XV87-15GB 17 l 87 l PUMP COOLER FLOW XV87-1578 .
'
i .- i 041-1 INSTR - RWCU FLOW -- XV44-102A,B -
l ldl4
}
041-2 INSTR 'A' LPCI DELTA P --
*
XV51-103A. 1i ;l $1 i
-
' i ! r. .- '
$043A INSTR - RECIRC LOOP --
XV43-1F040A,C 1 431 l l, , 'A' OELTA P :
,; ,
'
i 047 INSTR - RWCU FLOW -- XV44-1020 1 44 .
~
I 048A-l' INSTR - RPY LEVEL -- XV42-1F065B l' i ' XV42-1F0478 l 42 i 'p i'
'l 048A-2 INSTR - CS DELTA P --
X 52-IF0188 -
j 52 [R'.
.
' l'
*
! 048B INSTR - RPV. LEVEL --
XV42-1 FOSSA 1 24 2 ,
'
i XV42-1F047A l
' -
I i
:*l .
I l 049A,B INSTR 'A' AND 'B' MAIN --
.XV41-IF071A,B 1, . 41
,
'
'
STEAM LINE FLOW XV41-1F072A,8.
, ,
. . ,.
,
i 050A-2 INSTR 'B' RECIRC FLOW -- XV43-1F0lIA 1 43 1 ,,
! .,, XV43-1F0128 ; i l
i s1 .
,
','s -
m
'
5,
-
-
.
. .
. fl !
,
.; - 7,
,
,,
i
' i I I
>
l ,i f.
.
'
#
t
'
I
,
-
.
,
i
; .
' ,'
,
: i
*
_ TABLE 3.6.3-1 (Continued) , i
*
i
,
~
g PART B - PRIMARY CONTAINMENT ISOLATION EXCCSS FLOW CHECK VALVES I :lg .
!' :* '
'
$ l g INBOARD OUTBOARD .
ISOL. . .,
, PENETRATION FUNCTION ISOLATION ISOLATION HAX.ISOL. SIGNAL (S), NOIES PAID ,
c NU'fBER DARRIER, BARRIER TIME.IF APP. IF APP. ,
*
.,
(SEC)(26) (20) :i 5 i i
,
'
*' 050A-3 INSTR 'B' RECIRC FLOW --
XV43-1F0118 . 1 l43 # XV43-1F012A
.
'
0500-1 ' INSTR 'A' RECIRC PUMP -- XV43-1F004A 1 43 l SEAL PRESSURE i + .,
,
. l .
y e
*
0508-2 INSTR 'A' RECIRC PUMP -- XY87-156A b 87 i COOLER FLOW XV87-157A . I ' f ,051A-1 INSTR 'A' RECIRC LINE -- XV43-1F009A 1l 43
*
s XV43-1F0100 .'.
- *-
FLOW il i 'nt i
! 'T OSIA-2
$
INSTR 'A' RECIRC LINE -- XV43-1F009B XV43-1F010A
I 4 I 'hb t FLOW
'
-l
. ,
'
051B INSTR - JET PUMP FLOW -- XV42-1F059T 1 42 l '; .s
! -
(JP19) .
* -
,
s ,; 'l
,
'
i XV42-1F051C <
'
:
! ,I (JP20) l
'
I, jI '
, XV42-1f053C 'l i* ; ,5 (JP20) , l , ~ll 052A INSTR 'B' HAIN STEAM --
XV41-1F0700 1
41 *
'l
:! LINE FLOW XV41-1F073B ', , j
't b528-1 >
. INSTR 'B' RECIRC -
XV43-IF011C,D t 1 . 43 , j i
-
LINE FLOW ;
**
y -
.
l ', I * 0528-2 INSTR 'B' RECIRC -- _XV43-1F012C,0 1 43 l i = LINE FLOW
'
, j'
! G l
! * 057 INSTR - RWC0 FLOW --
XV44-102C 1 44 i i i .
'
, ;
'
'
l ,
*
i'
} ,
- 'k
.
I ,
. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ - _ _ _ _ _ _ _ ___
.
i :
'
,
*
'l'$
'
l :i g ; TABLE 3.6.3-1 (Continued) ,
,
l i C -i t i g PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CllECK VALVES
i
!
'
I 5 ISOL. '
!l i i1 Q -
FUNCTION (NBOARD ISOLATION OUTBOARD ISOLATION HAX.ISOL. SIGNAL (S), NOTES PtID 'PlI
, PENETRATION NUMBER BARRIER BARRIER TI.lE.IF APP. IF APP. Il !'
*
c (SEC)(26) (20) : _
.
l $
-4 l,t'; ' '
,
-- XV43-1F040B 1 43 u 050A INSTR - RECIRC LOOP '
.
*
j 'B' DELTA P ' .
(43-1004A(CK) -
IS 41 IY
!. '
061-1 RECIRC FUMP SEAL PURGE See Part A of XV43-103A 1 l j'; '
a-this table) . , s r.< 15 ' 43 $.:
(43-1004B(CK)
.
061-2 RECIRC PUMP SEAL j.. l
-
-
i ,
; '
PURGE See Part'A of this table) XV43-103B i :
,
.
;; I,
' '
s
'B' XV43-1F0400 1 -41 [
* 063-1- INSTR - RECIRC LOOP
--
i
'
DELTA P
, [ i1
- I
'XV43-1F004B 1 II 43 I .
08i3-2 INSTR 'B' RECIRC PUMP --
'
SEAL PRESSURE XV43-1F003B i h.
'
'
' i 1 42
-- -
NV4'2-1F0438 l 065A INSTR - RPV PRESSURE }. ' 1 42 ,. l INSTR - RPV PRESSURE
-- XV42-1F049A
. 065B
'
l' , 42 INST-RPV LEVEL
-- XV42-1F045D 066A-1- i j J
'
1 'l ,l , INSTR 'B' LPCI DELTA P
--
XV51-1020
<
066A-2 i {c
< XV51-1030 ) ! l,
; 1 42 '{
XV42-1F045A
-
--
"
0668-1 INST - RPV LEVEL '
'
l .
! E --
XV51-IO2A 1 61 ,
? " 0668-2 INST ' A' LPCI DELTA P i 4 . .
XV51-103C - i m
= 1
*
i
li l f.' l1
-- XV42-1F049B
. - 3 067A INSTR - RPV PRESSURE ! . ; i s: - I
, . .
j9
. t
. - - _ . -- - - _- _____________ -
~ " ~ *
.. .
*
~~ J, -
, , _ ..g- ._...
.;*"* ^ * ~ . . m
.s . s.*.
- * * *
,_ ; ,
. -
y
*
,,
n- ,
>
,
_ s ,.~- ~ s . .t
..4...... + . - - ~ ~ . = = = - * - " * * - ' + * * - - ---..
.. . .. -
.
- -
..=.=.=.. , . _ _ _ .
__. ~~ _ - - -
--~ ._
- . -
. ~
,_._
*
Q -- ., .
--
e5 N N N N
._ c.
- v v -W_ * etr,
- ---
-
f __
-
, - .
J
- -.
- - - ._ . -
-.- ~
M ~
~ W ._
---
-
>= .
M M M M -~ -
. n.
_
.
m 4A v . J Ch. - eCA
, .J Z *EC m .
O t3 O M *=e L N *
-
M M m-e v ,
. m W >
J Ch.
. * C .
> .A *EC e
*
J c lhd OWN U M m-e v W o== .m
= .WV W >< Z W <~A -
3 Z 9- V O J ^ 6 t.D ^ A M 1D
*C M A Ch.
C: m *C -7 7
::3 W en M V v C 4.J Z v v *C CD = X OO O O m m e.J W E M CE 6 6 CD CD C eC W W FW e=4 9M P4 O Z O C *-e t I a 9 .
O O CD J CE N N N N v w *OE et* v e4" etP .
<J D to 4C > > > > - *A O m CD M x x x e O *
tn m. e=e = W
. W m Z W
a b
-
.
.
CD - Z . C. eC C p- s- OME y * Z Ca: > W C * cec = U CJE CD O E
>= Z W eC e a I e ,
CC M M CD 4 9 I 4 k
-
E CE A CE O O >= l r== U U C CD W aC W CE W E 8=. D E E M J *
< M W * Ch.
A W > Cl T Ck: W Z 5 A .J D Cn.
A
>
CL c
> 4-.
>=
W CC CE W 7 Z 7 0 8 8 I
*=e I W CC J .J U e- CC.
P= >= W >= W Z th M LA > th > D Z Z ZW ZW 6 e-o m m .J m .J _ s O m 9'"
< v-4 N l E.c.WCe
. i i
-
WQ Z CD
%
CD
%
eC N % I l W$ cZ to C to C C
**
O M ,_.. t
'
'
LIMERICK - UNIT 1 3/4 6-38 Al4 & IMS
.. ... **' * w v 9 - y .,.-eg.g m+c-9 s-- - , - -,-aw,, er 4y-a--r--w* .- *m- ,---w ep .
i ijiii)lf
! ,
- ' , ' 1
.
'
.l- !
iI
,
1*Ii! -
;?
i
-
' 8
. ,
i J
,g
;1 I
l i
-
.l
'
.
'
, .
'
I i! * i
!
l> ' I ,I i
; .
D I ! j I. i 0 0 0 0 0 0 0 '
;
0 0 0
:T ,i A
P 6 6 6 6 A 6 6 '6 6
-
6 6
'
i
' j fik
- ' l8 1 i
'Ji
:
l l
' ' S ' ,
E 8 ' e
.i j T
- 1, i !
'
' O 2 2 8 8 8 8 8 '8l l
N 2 2 2 ,a
-
i
, )
S - ( . lP :
.AP - <
!' .
LNA) OG 0 '
- - - - - - - -
iSIF2 - - - - ,'
*
- - - -
ISI( - 7
,
-
. ;
P P
.A) ) L 6 B OF2 SI( ~
E .)
) P I.
XME EC -
,
d Y - - - - - - - e T AI S - - - -
- - - - -
u ( MT( - - i n S t N n O . o I C T ( A _ R N ' 1 T DO ~ _ _
- E RIR _
3 N ATE _ E OAI _ 6 P BLR TOR - . - - -. - - - _ 3 T USA - - - - '
- - _
N OIB - - - - - - - '- _ _ E E _ L M _ B N _ A I T A G G T G G G _ N N N N N N _ I I O I I I R R , . _ C R R R - N - - - 0 0 0 R R R R R R - Y O 0 0 E E E E E E R DIR - E T T T T A RTE E E E E T T S S S S M AAI L L L L L S S - I I I I OLR B B B B B I I I N R BOR U U U U U N N N N N P NSA O O O O Q A A A A A A , D D D D D C C C C C C IIB
-
C M T E R T S R A ~ Y O R P R R T E O O S E O O E A W L G D DK L G C - O O N C 0 l i N R S I WP R A S SO 1 C I E L D O T L S SL I N A T A R O W O DA NN N I LL O N O F E E SR C C CL M I I T P AG D C CC I I N UT A A AN S L N P ES O ON E N S A O M R I EO G lE T TO E V H U UL T NC A N I N NS C O P TE I A T
.
E ER C M N AV S LE L O k C RE O LG O I l M ME A E O EA V T E P PP R R R EL P P CT - C W I I D T I W N Y U UD A D U C MW D R SL IO O U R Q QN E R E E EO _ F D E EA II C N R TL C MV L - N O i I . T _ A D D D E C _ _ RR - 8, - - - _ _ TE -
'
ED A A A A A A _ _ 6 0 1 3 4 5 6 _ NM 1 2 4 0 0 0 0 0 0 _ _ EU A 0 0 0 0 1 1 1 1 ; ifI k PN N 0 0 0 0 1 1 ,M a i@ _ rk,Co7 ezl P
- -
d I* _
' _
_ _ _ _ J .
:lI ~ g.
,. ', .li* 1: .i !-
, I - - _
:
.'j ,' ;1
.
,
s $ ,- ;I .' ! 6 :i i. ! :
;dil; ; : i .: ~ : I! _
_
,!: !' _
_ _
! ; )-
,g . .
jt
, j ;' l' j. 6 - i
,i.'g-
'
'-
-
,' .* i1i . !i1 4 ,- . ..
?h.
) -{- e J
- -[ ;;
i.\ l -
.
l I!Ii
.
,
( '
I
., ! l j ' i /
N . l i
. i
,
-i'l l i D ,
l
& !
j0 q' '0 ,
.
I P 6 6 6 i ;
.
:! '
- 1
!I'
:I i
-
I
~
L '
'
e i , S i
,.
- ,
\
.'
E *
'
;
T \ !
,f O
'-l N 2 8 8 i ' '
!
l
. , )
S ( .
! 4 .LP LNA)
AP . i OG
;
0 - - . SI F2 - - - . ISI( - - - -
,
.
.
P . P
.A) ) L 6 .
B OF2 SI( E I .)
. P .EC Y XME .
T AI S - - -
( MT( - - -
S
) N d O ,
e I u T . n A i R N - t ! OO n E PI R o ! i ATE C E OAI -
( P BLR TOR 1 T USA - - - * - N OIB - - - ,
3 E
. H l
6 t l 3 A T G E N N . L O I . B C R . . A N - . T Y O 0 R R R DIR E E A RTE E T T - M AAI L S S I OLR B I I R DOR U N N P NSA O A A . IIB D C C - - C T R A L P O R T . - N R . _ O T C S N D I N H A E C G T N U A O A ' N I I I G O T I S A N T S C I
-
C E I A - N C D R - U C N T F A I S .
- ..
_ _ N i _ O I T A RR 8, TE EB A A NM 0 2 0 EU 0 2 3 PN 2 2 2 ne, - [ Cggn7 c34 w - , ,4
$ 7 g; .
, !,- i
:1 ;I- ;;! i !!1i
! ,
. : ! 4 ' '
'
-- " ~
- ;.
? . . .~.- , _ ,
..
_ . _ . .
-
m . .._ _ .. .-.3.-.-.
. - . . . . . . . . . . - -
_ __ _ _
~" ~ ~
-
~
_ _ _ __ . j
-
( -
._
_ i l . l
~~
'
. . - . - TABLE 3.6.3-1
-
~~
PRIMARY C0h'*AINHENT ISOLATION VALVES ~ . - NOTATION. -
.- ' ~ ' ' -
NOTES 1. Instrumentation line isolation provisions consist of an orifice and excess flow-check valve or remote manual ~ isolation valve. The excess
- -
flow-check valve is subjected to operability testing, but no Type C test -- is perfobned or required. The line does not isolate during a LOCA and can leak only if the line or instrument should rupture. Leiaktightness of ,
- the line is verified durirtg the integrated leak rate test (Type A test).
2. Penetration is sealed by a blind flange or door with double 0-ring seals.
These seals are leakage rate tested by pressurizing between the 0-rings.
- 3. Inboard butterfly valve tested in the reverse direction.
4. Inboard gate valve tested in the reverse direction.
5. Inboard globe valve tested in the reverse direction.
6. The MSIVr'and this penetration are tested by pressurizing between the valves.
Testing of the inboard valve in the reverse direction tends to unseat the valve and is therefore conservative. The valves are Type C tested at a test pressure of 22 psig.
7. Gate valve tested in the reverse direction.
8. Electrical penetrations are tested by pressurizing between the seals.
.
.
9. The isolation provisions for this penetration consist of two iso 1a' tion valves and a closed system outsice containment. Because a water seal is maintained in these lines by the safeguard piping fill system, the inboard valve may be tested with water. The outboard valve will be pneumatically . tested. l 10. The valve does not receive an isolation signal but remains open to measure containment conditions post-LOCA. Leaktightness of the penetra- ' tion is verified during the Type A test. Type C test is not required.
11. All isolation barriers are located outside containment.
12. Leakage monitoring of the control rod drive insert and withdraw line is ) provided by Type A leakage rate test. Type C test is not require'd.
'
'
13. The motor operators on HV-13-109 and HV-13-110 are not connected to any power supply. j 34. Valve is provided with a separate testable seal assembly, with double concentric 0-ring seals installed between the pipe flange and valve flange f acing primary containment. Leakage through these seal.s is included within the Type C leakage rate for this penetration.
__ AUS 8 lHS LIMERICK - UNIT 1 3/4 6-41 t
-
. . . _ _ _ . . _ _ . . . _ - . . . . . . . . . . . . . , .
,_
.
..
. . . _ . . - .-
, ,,
.- -. .- -
_ - _ _ - - _ . - .- - _
-
_~
. '.L '~ '. ;. . .T ._
~
J~~.-~' '. ; . . .
~
.... . , . 7I ;. >w .. .. Z.T.i _."..T.E ^ ~ Z.
--
. -- 3 _
_
-
_
-
_ _
.
._ : -m : .
_ _
._ __ .
. _- -
.___ _
.___
_
- - .
_
(-
-
- -
TABLE 3.6.3-1 -- .__
-PRIMARY CONTAINMENT IS0tATION VALVES _.
-
- -
*
,- NOTATION ,
_
.w
-
-
- .
-
_ -- J NOTES (Continued)
- -
_- 15. Check valve used instead of flow orifice. ) 16. Penetration is sealed by a flange with double 0-ring seals. These seals j
- are leakage rate tested by pressurizing between the 0-rings. Both the TIP__ ,
Purge Supply (Penetration 35A) and the_ IIP Drive Tubes (Penetration 350-G)' I
'
are welded to their respective flanges. Leakage through these seals is
- included in the Type C leakage rate total for this penetration. The ball ; - valves (XV-141A-E) are Type C tested. It is not practicable to leak test i the shear valves (XV-140A-E) because squib firing is' required for closure.
, Shear valves (XV-140A-G) are nomally open. ,
'
17. Instrument line isolation provisions consist of an excess flow check valve.
Because the instrument line .is connected to a closed cooling water syster inside containment, no flow orifice is provided. The line does not isolate during a LOCA and can leak only if the line or' instrument snould rupture.
Leaktightness of the line is verified during the integrated leak rate test (Type A test).
18. In addition to double "0" ring seals, this penetration is tested by pres-surizing volume between doors per Specification 4.6.1.3.
19. The RHR system safety pressure relief valves will be exempted from the initial LLRT. The' relief valves in these lines will be exposed to contain- f ment pressure during the initial ILRT and all subsequent ILRTs. In addi- i tion, modifications will be perfomed at the first refueling to facilitate local testing or removal and bench testing'of the relief valves during sub-sequent LLRTs. Those relief valves which are flanged to facilitate removal will be equipped with double 0-ring seal as.semblies on the flange closest ' to primary containment by the .end of the first refueling outage. These . seals will be leak rate tested by pressurizing between the 0-rings, and the results added into the Type C ' total for this penetration.
20. See Sp.ecification 3.3.2, Table 3.3.2-1,.for a description of the PCRVICS . isolation signal (s) that initiate closure of each automatic isolation valve.
In addition, the following non-PCRVICS isolation signals also initiate closure of selected valves: EA Main steam line' high pressure, high steam line leakage flow, low MSIV-LCS dilution air flow LFHP With HPCI pumos running, opens on low flow in associated pipe, closes , when flow is above setpoint ' I LFRC With RCIC pump running, opens on low flow in associated pipe, closes when flow !s above setpoint -
.
LFCH With CSS pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve N11y closed or RCIC turbine stop valve fully nond ' All power verated isolation valves may be openea or closed remote manually. , LIMERICK - UNIT 1 3/4 6-42 D E 195
,
. _ . . . . . . . - . . _ . , . . -
. - - - _ . .. .... . . . _ ~ _.
- -- - w ,w i,w - w--w - w- w-1.,-w- ,,,w w w rw +- --my v wirs--w ww we- w- aw -w--- ww-* -,--- vw -ww-v,-- --
-w -<g-e---y ,w-,r-'
_ __ _--___ _ _
~ -
-:
~ - -
.
. _
-
._ _ i-- -
_ - . - -
-
_
--
__ _ --
-- . .. _
~
.
387408"6670 _ __ _2
~~
yme3 m _-
- .
_ PRIMARY C0hiiTiKEt NOTAT!^M 15b[IT106 VALVES- - - _ _
- --. _ _ ~~'- - - -
__ _ _ . __
'
NOTES (Continued) 21. Automatic isolation signal causes TIP to retract; ball vaive croses when - probe is fully retracted.
. 22. Isolation barrier remains water filled or a water seal remains in the Isolation valve may be tested with water. Isolation
~
line post-LOCA. ~
--
valve leakage is not included in 0.60 La total Type B & C tests.
23. Valve does not receive an isolation signal. Valves will be open during - Type A test. Type C test not required.
24 Both isolation signals required for valve closure.
Deleted l 25.
26. Valve stroke times listed are maximum times verified by testing per Speci ' ficatiur. 4.0.5 acceptance criteria. The closure times for isolation valves ' in lines in which high-energy line breaks could occur are identified with a single asterisk. The closure times for isolatior valves in lines which provide an open path from the containment to the environs are identified with a double asterisk.
27. The reactor vessel head seal leak detection line (penetration 25A) excess flow check valve is not subject to OPERABILITY testi*g. This valve will not be exposed to primary system pressure except under the unlikely con-ditions of a seal f ailure where it could be partially pressurized to reactor pressure. Any leakage path is restricted at the source; therefore, this valve need not be OPERABILITY tested.
28. Automatic isolation logic to be added by the end of the first refueling outage.
29. Valve may be open during normal operation; capable of manual isolation from control room. Position will be controlled procedurally.
30. Valve normally open, closes on scram signal.
31. Valve 41-101G is an outboard isolation barrier for penetrations X-9A, 8 and X-44 Leakage through valve 41-1016 is included in the total for , I penetration X-44 only.
32. Feedwater long-path recirculation valves are sealed closed whenever the The reactor is critical and reactor pressure is greater than 600 psig.
valves are expected to be opened only in the following instances: a. Flushing of the condensate and feedwater systems during plant startup.
' b. Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to cosusencing plant startup. l Therefore, valve stroke timing in accordance with Specification 4.0.5 is not required.
33. Valve also constitutes a Refueling Area Secondary Containment Automatic Isolation Yalve as shown in Table 3.6.5.2.2-1.
.
-
JUL 8IM7 LIMERICK - UNIT 1 3/4 6-43 we w 6 __.
v -
'
-
_ ,_- -- -. _-
,
-
- - - . - .
_-_
'
_ - - _ ,
--
-
=_ - - -
,
. - - - - _
-- *
_ _ _ _ -- - - -
.__.
_
-- . _
__
~ ~
-38740866[
-
--
~ ' - .
..
CONT AINMENT - SYSTEMS- --_
- . - _ _ ~ - ._
__ 3/4.6.4 . VACUUM RELIEF .- _
' -
SLPDRESSION CHAMBER - ORYWELL VACUUM BREAKERS _ tlM! TING CONDITION For OPERATION
*
.- _
'
l
~
3.6.4.1 Each pair of suppression chamber - drywell ucuum breakers shall be OPERABLE and closed. . _ ADDLICABILITY: OPERATIONAL CONT *TIONS 1, 2, and 3.
ACTION: a. With one or more vacuum breakers in one pair of suppression chamber - drywell vacuurs breakers inoperable for opening but known to be closed, restore the inoperable pair of vacuum breakers to OPERA 3LE status within 72 hours or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours, b. With one suppression chamber - drywell vacuum breaker open, verify the other vacuum breaker in the pair to be closed within 2 hours; i restore the open vacuum breaker to the closed position within 72 hours or be in at least HOT SHUT 00WN witnin the next 12 hours and in COLO j SHUT 00WN within the following 24 hours, c. With one position indicator of any suppression chamber - drywell vacuum breaker inoperable- , 1. Verify the other vacuum breaker in the pair to be closed within 2 hours and at least once per 15 days thereaf ter, or i 2. Verify the vacuum breaker (s) with the inoperable position l indicator to be closed Dy conducting a test which demonstrates
'
that the AP is maintained at greater than or equal to 0.7 psi l for one hour without makeup within 24 hours and at least once I per 15 days thereaf ter.
Otherwise, be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.
I
. - .
W 8% LIMERICK - U,i!T 1 3/4 6544
. - - __ _
. - - . . . -. -
.
__. . _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ . _
- - - .- -_ .- =
_ __
--
, . - . . .
-
- . :- 2
; -
-
. . _ _
_ z ---_ __ _
~=38740_8,6_670.__ -
T ._--
~
=~~~~ _, CONTAINMENT SYSTEMS .
- --
_ ,
- - - _ _.-_ _
- -
~ ' ~ ~ _~~ SURVG LLANCE REQUIREMENTS
- __
_ _ 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be: a. Verified. closed at least once per 7 days.
b. Demonstrated OPERAELE: 1. At'least once per 31 days and within 2 hours af ter any discharge of steam to the suppression chamber from the safety / relief valves', by cycling each vacuum breaker through at least one complete cycle of full travel.
2. At least once per 31 days by verifying both position indicators OPERABLE by observing expected valve movement during the cycling , test.
3. At least once per 18 months by; a) Verifying each valve's opening setpoint, from the closed position, to be 0.5 psid : 5%, and b) Verifying both position indicators OPERABt.E by performance of a CHANNEL CALIBRATION.
c) Verifying that each outboard valve's position indicator is capable of detecting disk displacement 10.050", and each inboard valve's position indicator * is capable of detecting ~ disk displacement 10.120".
l
-
\
)
Atl6 81985 ' LIMERICK - UNIT 1 _ 3/4 6-45 l l
._. .. . . -
' - ~: ^
_
, - _
-- ._.
~ ~ _ ~ ' ~ T
~
--: . _ .. .---
--
,_ - :
_ -- - _ _
. _
- __ ___
-
,
. - _._
-
J -. _ - ~~ __ _ __
~~ ~ ~
3B7408T57: CONTAINwENT SYSTEMS ~~ -
-
_ __ _. _
~~~[
-
3/4.6.5 SECONDAR%-CONnMENT _ -
-
-
REACTOR ENCLOME-SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained.
ADDLICASILITY: 0?! RATIONAL CONDITIONS 1, 2, and 3.
_ ACTION: Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within A hours or be in at least HOT SHUT:)0WN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
'
.
SURVEILLANCE REOUIREMENTS 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by: a. erifying at least once per 24 hours that the pressure within the
.
reactor enclosure secondary containment is greater than or ecual to 0.25 inch of vacuum water gauge.
b. Verifying at least once per 31 days that: 1. All reactor enclosure secondary containment equipment hatches and blowout panels are closed and sealed.
2. At least one door in each access to the reactor enclosure secondary '! containment is closed.
3. All reactor enclosure secondary containment penetrations not capable of being closed by OPERABLE secondary containment auto-matic isolation dampers / valves and required to be closed during l accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers / valves secured in - position.
c. At least once per 18 months: 1. Verifying that one standby gas treatment subsystem will draw down the reactor enclosure secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 121 seconds with the reactor enclosure recirc system in operation, and 2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inch of vacuum water gauge in the reactor enclosure secondary containment at a flow rate not exceeding 1250 cfm.
JUL 8 tM1 Wt 2. 6 LIMERICK - UNIT 1 3/4 6-46
- - _ __ -- -. . _ _ - - - _ . . .
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
ATTACHMENT 3 PHILADELPHIA ELECTRIC COMPANY t.lMERICK GENER ATING STATION P. O. BOX A S AN ATOG A, PENNSYLV ANI A 19484 (215) 3271200. EXT. 3000
"".U.7.', .'.'U * *
vi u................." June 13, 1988 Mr. Robert M. Gallo, Chief Operations 5tuch . Division of Reactor Safety I U.S. Nuclear Regulatory Commission Region 1 475 AllendO e Road King of Prussia, PA 19406 i Subject: LO/ SLO Written Exam Comments for NRC Inspection #88-16
Dear Mr. Gallo:
Attached you will find the Limerick Operations / Training Section responses to selected questions and answers associated l with the NRC RO/SRO written examinations recently administered on ) 06/07/88 at the Limerick Generating Station. These comments are ! submitted in the interests of clarifying those areas where i alternate correct answers should be considered as well as providing updated Limerick specific information which can be used in the process of evaluation of these exams. Wherever possible, references have been indicated or added which justify the comment made or question / answer clarified. Also, a separate sheet is attached to indicate discrepancies in the point values of exam questions and/or answer keys.
! The written examination for both the RO and SRO candidates did appear to be quite lengthy in that all of the candidates took the entire six (6) hours or more with little time for review of their answers. Compared to the last Hot License exams, which were administered on 10/20/86, it appears that these recent exams had many more subparts within each question. For example, the
.
, - - . .
___ _ _
. _ ,
. _ . -___
M Juno 13, 1988 Page 2 of 2 l 10/20/86 RO exam had 47 questions with 81 total subparts, whereas the 06/07/88 RO exam had 42 questions with 136 total subparts. ) i Now, although there is absolutely no concern about exam ; comprehensiveness, there is a concern that the many question subparts in both exams may have caused undue stress on several candidates to complete the exam within the six (6) hours 1 allotted. I If there are any questions concerning these comments, please contact E. G. Firth, LGS Superintendent-Training, at 327-1200 (x2080).
Very 1 urs,
-
EGF:mgd Attachments ; c.c. M. J. McCormick, Jr.
J. Doering E. G. Firth R. A. Nunez J. F. Hanek - EG&G Idaho, Inc.
i l e
.
' t _ , - - _ , __ , m.___ . _ _ _ _ - - . _ . _ - . , - - _ . - _ , _ , - . . . - . . . - - - - - -,-
- _ - . . - _ _ _ _ _ - _ - - .
NRC SENIOR REACTOR OPERATOR EXAM (06/07/88) POINT VALUE DISCREPANCIES EXAM Section 5 - None Section 6 - Question 6.05 Total Point Value Given =
(3.00)
Breakdown Point Value =
(2.50)
Category 6 Total Points would =
(25.00) if worth (3.00) 1 Section 7 - Question 7.06 i
Total Point Value Given =
(2.50)
Breakdown Point Value =
(3.00)
Category 7 Total Point would =
(25.00) if worth (2.50) .
Section 8 - None -
,
ANSWER KEY Section 5 - Question 5.09 part (a) states (1.00), should state (0.50) Section 6 - Question 6.05 - same as exam Section 7 - Question 7.06 - same as exam Question 7.08 part (d) should be worth (0.50) vice (0.25) .
.
>
4f y
. . _. - . - , - - , - . . - - . - , - , . , - , , , - - . _ _ . . - , -,..,.,. .-- --- .
. . .
WWWWM 8 NRC REACTOR OPERATOR EXAM (06/07/88) RESPONSES _ Category 1: Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow 1.01 d. Answer can be either TRUE or FALSE. The phrase "as a result of" is vague, and can be interpreted two ways by the candidate.
1) Delayed neutrons are produced upon decay of delayed neutron precursors, and not from fission of U-235 and U-238 directly, tPerefore the answer is FALSE.
2) Delayed neutrons are eventually produced after the fission of U-235 and U-238, therefore the answtr is TRUE.
REF: LOT 0870 f.C.2.a
.
1.02 d. Answer should be SDM INCREASES. Increasing Pu-240, an absorber of epithermal neutrons, reduces kegg in the core, which results in an increasej SDM.
REF: LOT 0950 Typical Question 6.d
!
1.05 a. Additional correct answer: INCREASING TEMPERATURE OF RECIRC PUMP SUCTION. During startup, Recirc Pump suction Temperature is monitored, and is considered to , be Reactor Water Temperature for an operating Recirc pump. Reactor Water temperature is recorded and monitored to the heating range per GP-2.
REF: GP-2, Sect. 3.3.25 l l 1.05 b. Interval 2 and Interval 3 are both correct answers.
The period for Interval 2 is 90.11 sec.
The period for Interval 3 is 80.03 sec.
Both are greater than 80 seconds, and either could be an interval in which the heating range was entered.
REF: LOT 1430, Obj. 4
.
_ _____ ,__ __ _ , . , - _ - _ _ - - - _ _ _ _ , , _ . . _ _ . _ _ _ , _ _ _ _ _ _ _ _ _ . _ _ , _ , _ _ _ , , , .. . . _____. , _ _ _ _ . - __
(y, w -
,
.,
;bsterOporotorExcnRosponaOc(Continu2d) -
Pogo 2 of 9 8 Candidates may discuss how the reactivity coefficients change when feedwater temperature is reduced, as well as how the coefficients affect the transient, and should not be penalized for this.
REF: LOT 1450; LOT 1460; LOT 1480 1.08 c. FSAR analysis assumes a decrease in Feedwater temperature of 100 degrees F. The isolation of a feedwater train would result in less of a change in temperature than 100 degrees F, and with appropriate operator action, a scram on high flux can be avoided.
Operator action per OT-104 "Unexplained Reactivity Insertion" would result in voids decreasing followed by voids increasing.
REF: OT-104; LOT-1460, Typical Question #1; FSAR pp 15.1-1 through 15.1-4
.
. 1.09 This question is similar to SRO 5.08. The wording is awkward and extremely confusing. The question is not based on a Learning Objective REF: LOT 1300 1.10 a. Additional correct answers are: REACTOR POWER - Class discussion included graphing NPSH vs Reactor Power for Recirc pumps. Reactor Power is a parameter available to the operator on the Process Computer.
RECIRC SUCTION TEMPERATURE - Saturation temperature is directly related to NPSH, and is a parameter available to the operator on the Process Computer.
RECIRC PUMP SPEED - Friction loss is affected by pump speed and is directly related to NPSH. It is available on a meter on the C601 panel in the main control room.
REF: LOT 1290, pp 8 & 9
. - , .. . , . . -- . . - ..- --.- . . - -
. _ _ _ _ _ _
pscctor Op3 rotor Exca R20ponacs (Continu3d) Pcg2 3 of 9 1.11 a. "Bundle" and "Assembly" are used interchangeably, and both should be accepted.
REF: LOT 1370, p 5 1.12 a. Additional correct answer: MAPLHGR/APLHGRLCO Although LOT 1410 uses MAPLHGRLCO, Tech Spec 3.2.1 defines the LCO for APLHGR, and not MAPLHGR.
REF: Tech Spec LCO 3.2.1
.
e
.
l
. l l . _ _ _ . .__ . . _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ . _ . _ . _ . . . _ . , _ ____..._..._._.
_
- _ _ _ _ _ _ _ _ _ _ _
, NRC Rocctor Op3rator Exc3 R0sponsOs (Ccntinu0d) Pcg3 4 of 9 l Category 2: Plant Design Including Safety and Emergency Systems 2.02 c. Both WOULD or WOULD NOT are correct answers, depending t on candidate's assumptions.
1) If B RHRSW is providing cooling water to the B RHR heat exchanger, it will trip on a LOCA signal, since it is powered by the Unit 1 D-12 Bus, and WOULD NOT is the correct answer.
i 2) If D RHRSW is providing cooling water to the B RHR heat exchanger, it will not trip on a LOCA l signal, since it is powered by the Unit 2 D-22 l Bus, and WOULD is the correct answer.
l l REF: LOT 0400, p 9 l l 2.04 a. Correct answer is 118 seconds. MOD 86-189 changed this time delay from 58 seconds to 118 seconds. This { i is referenced in S36.1.B, section 8.2 NOTE: "116 to
'
i 120 seconds after manual initiation a 10 minute timer is enabled."
REF: S36.1.B, step 8.2 Note, step 8.3 l l 2.04 c. Additional correct answers.
1) Decreasing tank level 2) Decreasing reactor power 3) Continuity Lights extinguish REF: LOT 0310, p 17
'
l l 2.05 a. 2. Correct answer is 455 psig.
REF: LOT 0350, p 8 2.05 a. 3. Question should be deleted. Check valve will ' open when discharge pressure is greater than reactor pressure. Discharge pressure is not required knowledge for RO.
REF: LOT 0350, p 8 *
. . _. . . - . - - _ ,
-_ -_. __ _ _ _ _ _ _ _ _ _ _ _
NRC Reactor Op3 rotor Exc.C R0aponses (Ccntinucd) Pago 5 of 9
!
l 2.05 b. Design pressure should be deleted from this question.
Design flow is required for RO knowledge. Design , i pressure for rated flow is not required knowledge for j j RO. i ! REF: LOT 0350, Obj. Sa ! l 2.06 e. Suppression pool spray and Full flow test valves also close. Candidates should not be penalized for including these in the answer.
l REF: LOT 0370, p 14 2.08 c. Answer should be either NO SYSTEM or ESW, depending on candidate's interpretation of question. ESW can be used 'to cool the RECW heat exchanger, and can provide l cooling to the Recirc pump seal and motor oil coolers which are normally cooled by RECW, but ESW cannot * supply flow to the entire RECW system. Depending on assumptions made by the candidate, either NO SYSTEM or ESW is an appropriate answer.
REF: LOT 0460, pp 11-14 2.09 a. Additional correct answers: OVERSPEED LOW CONDENSER VACUUM
Each overspeed trip is independent of the other l feedpumps, and could have caused one feedpump to trip, i Each RFP turbine exhausts to a different shell of the l main condenser, independent of the other turbines, and could have caused one feedpump to trip.
REF: LOT 0540, pp 6 & 13
.
_ _ . . _ - . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ .
__ - .- _ _ _ _
. . - .,
NRC Roactor Op3 rotor Exc3 Rosponsos (Continued) Pcgo 6 of 9 _ Category 3: Instruments and Controls 3.01 a. Delete this question. The Recirc Pump Master Controller is not used at LGS, and the M/A Transfer Station is not used in auto. This question should be deleted since it does not apply to LGS.
REF: S43.1.a, step 8.20; S43.2.A, step 8.1; LOT 0040, pp 6 & 7 3.01 b. 75% flow is equivalent to approximately 60% pump speed. Either is an acceptable answer.
. REF: LOT 0040, p 8
.
3.01 c. Delete this question. The Recirc Pump Mastec Controller is not used at LGS, and the M/A Transfer Station is not used in auto. This section should be - deleted since it does not apply to LGS.
REF: S43.1.A, step 8.20; S43.2.a, step 8.1; LOT 0040, pp 6 & 7
.
3.02 b. Both HIGHER TRAN ACTUAL and SAME AS ACTUAL are correct answers. Wide range is calibrated with no jet pump flow, and would read the same as actual assuming no jet pump flow. Assuming jet pump flow, wide range would read higher than actual.
REF: LOT 0050, pp 9 & 10 3.03 a. (Similar to SRO 6.05) Candidate should not be prnalized for including LOW CONDENSER VACUUM 10.5 PSIA as an additional answer, since this could be bypassed - or active in startup, depending on assumptions made.
Candidate should not be penalized for including MANUAL, due to the wording of question.
REF: Tech Spec 3.3.2 notation ** on page 3/4 3-16; LOT 0120, p
. - - . _ - . . - _ _ _ . . _ . . . , . . _ . , _
_ _ _ _ _ . . . _ _ _ _ _ _ . _ , _ _ _ , . _ . _ _ _ . _ _ . . _ _ __ _ _ _ _ _ _ _ . . _ . _ .
_ _ _ _______
,
NRC Rocctor Op3rator-Exc3 RosponsGs (Continu0d) Pago 7 of 9 3.04 d. (Similar to SRO 6.10.d) Action 3 is not a correct answer. If the reactor is in startup mode and power reaches 15%, a scr'am signal is expected. A rod block should have been generated at 12% power. If rods can be withdrawn to allow power to reach 15%, the rod block signal has failed, and a scram signal is now expected. Action 2 is the only correct answer.
REF: LOT 0270 pp 11-13, Obj. 7 3.04 e. No ACTION is also a correct answer. Since criticality can be reached at any time, a rod block will not always be generated in this condition.
REF: GP-2, App.1, steps 3.1.S; 3.1.6; 3.1.7; on pp 3 & 4
.
3.06 Numbers and calculations on LOT 0590-6 should be accepted I in lieu of discussions. Candidate should not be penalized for continuing the transient until the plant stabilizes.
REF: None 3.07 a. Standby Gas Treatment System will also start if aligned to the Refuel Floor Ventilation system.
REF: P&ID M-76, sh 5 & 6
.
I l l l
4 I f
- - - ., - - - - _ . - . . . - --- . - _ . --- -,
_ _ _ _ _ - _ - _ _ -
'
NRC Rocctor Op3rctor Exc0 R0sponses (Continu0d) Pcgo 8 of 9 Category 4: Procedures - Normal, Abnormal, Emergency, and Radiological control 4.01 a. Directions in A-7 for an RO to shutdown the plant or scram are covered in two sections of A-7, using different wording. Section 5.1.2.4 is as appropriate as section 5.2.8 as an answer.
REF: A-7, sections S.2.8; 5.1.2.4 4.05 b. Safety Related Equipment is equipment contained on the Q List or QAD, and is commonly referred to as "Q-listed equipment".
REF: A-41, section 4.1; SRO exam 8.06 a.
. 4.05 d. The permission to release equipment for the . surveillance is an SRO responsibility. This question should be deleted from this RO exam.
REF: A-41, section 5.1.2 4.05 c. YES is the correct answer. This question is vague.
A-41 allows IVOR "need not be performed in cases where such verification would result in significant radiation exposure, as determined by Shift Supervision." A "valve in a high radiation area" may or may not result in significant radiation exposure. This would be determined by Shift Supervision, and would not be automatically excluded from an IVOR.
REF: A-41, section 5.3 l l l
)
l l
'
l
!
l i _ _ _ - - . - , - -- ' __ , __
*
NRC Rocctor Op3 rotor Exca Rosponsos (Continusd) Pcg3 9 of 9 j 4.07 a. Answers on answer key are redundant: Additional correct answer: TO REMIND OPERATOR THAT THE MIN FLOW VALVE IS BLOCKED CLOSED.
Tr.is is because of the basic purpose of the Blocking Procedures, i.e., assure safety of workers and avoid equipment damage.
REF: LOT 1860 Obj. 1; p 2
.
,
e
.
I I
. \
l l l l I e l __
, . . . . , ,_. _ . . _ .
_ _ _ . . .
. .
NRC SENIOR REACTOR OPERATOR EXAM (06/07/88) RESPONSES Category St Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics S.04 a. Answer can be either LARGER or SMALLER depending on assumptions (which were not required by the question).
When dealing with how moving a control rod will effect overall thermal power, many variables come into play and the final answer will be dependent on assumptions.
REF: LOT 1490, pgs 4-12 S.0S SROs should not lose credit if the problem was worked both at 1000F/hr and at less than 100 F/hr 0 cooldown rates since GP-3 note states cooldown is limited to < 100 ~ degrees F per hour.
REF: GP-3, p 16
.
S 08 c. Can be confusing to the candidates since steam flow is not directly inputted in a manual core thermal power evaluation as per RE-101. Consider alternate answer l depending on assumptions. l REF: RE-101, Core Thermal Power Evaluation
!
S.08 d. Answer is incorrect. The correct answer should be FALSE vice TRUE. The Reactor Operator Exam question 1.09 part (d) is the same and it's answer is given as FALSE. This answer is correct and supported by the heat balance calculation.
REF: Reactor, Operator Exam, Question 1.09 (d) l
S.09 a. Additional correct answer: MAPLHGR/APLHGRLCO Although LOT-1410 uses MAPLHGRLCO, Technical Specification 3.2.1 defines the LCO for APLHGR, and not MAPLHGR.
REF: Technical Specification 3.2.1
. .___.___. .._ ._. _ _ _
____- _.
7-~ ve- , , NRC'E2nior R0 actor Op3rator Excm RGsponsGO (Continucd) Pago 2 Cf 9 5.10 Answers should be accepted depending on SRO candidates supporting information. The candidates were not provided the note from Table 15.2-11 which is referenced on Figure 15.2-9 which they were given. In addition, Table 15.0-2 was not provided to the candidates which lists the input parameters and initial conditions for transients. Also, in accordance with the objectives for LOT-1580, BWR Transient Analysis, the candidates were not held responsible for a Loss of All Feedwater Flcw Transient. Due to these reasons, a range of answers should be accepted based on student's assumptions / justifications.
REF: FSAR, Figure 15.2-9, Table 15.0-2, Table 15.2-11; LOT-1580, P1 5.11 a. Subjective question dealing with water hammer.
Candidates could justify choices 1, 2, or 3. Many flow changes (starting flow, cessation of low) can be linked to cases of water hammer. For this reason, . acceptable choices are 1, 2, or 3.
REF: LOT 1291, pgs 1-7
:
I l i
. . \
l
:
_. - . _ . . _ _ - , . . . _ __ _ _ . . . , _ . , . - _ _ ,
, '
j NRC S nior R3cctor Op3rctor Excm RooponsGs (Continu0d) Pago 3 of 9 i Category 6: Plant Systems Design, Control, and 1 Instrumentation 1 6.01 Question is werded such that it could be misinterpreted since the HPCI turbine trip is at 15" Hg Vac and the question implies that the HPCI pump is still working with a suction of 18" Hg Vac. Therefore, answer (a) "Continue to inject" can be justified due to the fact the pump did not trip at 15" Hg Vac, REF: LOT 0340, pgs 11, 12 6.03 a. Alternate answer is NOTHING. Given the conditions stated in the question, recirc pump M/A stations may already be at 28% or lower. The operating map referenced in the question does not specify "speed" only flow which can be due to natural circulation and/or forced circulation. For this reason, if the - recire pumps are already at or below 20% speed, nothing will occur when feedflow drops below 20%. REF: LOT 0040, pgs 25, 26, 27 6.04 a. Correct answer is 118 seconds. MOD 86-189 changed this time delay from 58 seconds to 118 seconds. This is referenced in S36.1.E section 8.2 note, "116 to 120 seconds after manual initiation, a 10 minute time is enabled".
REF: S36.1.B section 8.2 NOTE, step 8.3 6.04 b. Additional correct answers: 1) Decreasing tank level ,
,
2) Decreasing reactor power 3) Continuity lights extinguish REF: LOT 0310, pgs 10, 17 l
. !
l l _ -- -. - _ .
'
E nr ,
'
NRb Senior R0 actor Op3rator Exc2 Rosp3nsos (ContinuGd) P ge 4 Of 9 6.05 a.- Additional acceptable answers are: Condenser vacuum low s Manual
'
Since the question did not specifically set other conditions or state "Automatic", these answers are acceptable in accordance with Technical Specifications depending on candidates interpretation of the question.
REF: Technical Specifications LCO 3.3.2, Table 3.3.2-1, Table 3.3.2-2 6.07 a. Standby Gas Treatment System will also start if aligned to the Refuel Floor Ventilation System via the Slide Gate Dampers.
REF: P&ID M-76, shts 5 and 6 - 6.08 During post exam review, the examiner was made aware of the error in the answer key dealing with the EHC numbers.
Answer should state 25 vice 40 and 18% vice 17%. 6.10 a. Question was worded such that interpretation of the phrase "...also has channel A of the IRM's selected" could be the IRM/APRM selector switch on the C603 panel. This is due to the use of the word "channel" instead of the word "detector". Alternate answer in this case would be NONE.
REF: LOT 0250, pgs 10-14; LOT 0270, pgs 11, 14 6.10 d. Alternate acceptable answer would not include the rod block as part of the answer. This is due to the fact that the candidate 'aay interpret the question as those
.
actions which occur at 15% power. Since the rod block occurs at 12%, they may only say scram signal (which I is generated at 15% in startup).
REF: Technical Specifications, Table 2.2.1-1, Table 3.3.6-2 i i _ __ -__ _ . _ _ _ _ _ _ , , _ _ _ _ _- _ . _ . . , . . . _ . . - - - _
-
- .
NRC Sanior Roactor Cp3rutor Exca Rospons0s (Continusd) - l Pago 5 of 9
_.
Category 7: Procedures - Normal, Abnormal, Emergency, and Radiological Control 7.02 b. Answer key breaks down point values into credit for calculation and credit for formula. The question does not ask the SRO candidate to show this in his answer.
Kequest that the formula and calculation not be required for full credit.
REF: SRO Exam Question 7.02, Answer Key 7.02 I 7.04 a. The question asks the candidate for two purposes. The , caution in S51.8.B shows that the two purposes given l in the answer are in essence saying the same. Either should be counted for full credit and other answers j givers by the SROs should not cause a loss of credit since the question asked to give two different purposes. . REF: S51.8.B, pg 5
!
7.06 a. and b. Confusion exists as to what the question was asking for versus what the answer key was answering. If the question meant to state level was oscillating +5 inches, did this mean from I normal to +5 and back to normal; or did it mean from normal level +5 inches (question only lists
+5 inches, which would not be an oscillation so it's assumed it was meant to be +5 inches).
In either case, the following OT procedures could be entered based on assumptions.
, OT-100 Low Level OT-110 Hi Level OT-104 Unexplained Reactivity Insertion The SRO's answer to part (b) will then be in accordance with part (a). The answer key lists two OTs in part (a) but only addresses one in part (b).
Acceptable answers will be according to question it.terpretation.
Also, part (b) question states "What actions 1 would you, as the Shift Supervisor, direct the ' operator to take?" The Shift Supervisor would l l
- . . - . - ,. . - - - . , _ - - - _ _ .
n _ _ - . RI
. ,.,- /
a,:Q M w s:; mmMMWm,y ;,w,mwmm ymg gs g y: m n gyym y ;g m y' e; -
. , .*~ -
pggog Meactor np] rotor Exam R;sponse,s/(Continued)ff Q J M i
~ '
a of f
'
p,ge ;
! .- 1 not direct the operator to enter T-100 if a scram )
occurs but rather would enter T-100 and direct the operators accordingly. Therefore, this step should not be counted as part of the answer to part (b).
In addition, the SRO candidate may enter OT-104, Unexplained Reactivity Insertion, in response to the power increase which the reactor may undergo ! when level oscillates so this answer should be accepted.
REF 97-100, pg 1; OT-104, pg 1; OT-110, pg 1 L' ' Answer key typographical error. Answer should read 7.07 ON-10*/ vice ON-105, for "Control Rod Drive System Problems".
. REF: grr N rmal Procedure Index d* Acceptable answer is entry into ON-109. "Total Loss of 7.07 the SRM, IRM, or APRM Systems". Cnis is because Technical Specifications require the SRM Upscale and Inoperative TRIPS be operable until the IRMs are on range 8 or higher in order to meet LCO 3.3.6 "Control Rod Block Instrumentation".
REF Technical Specifications 3.3.6, Table 3.3.6-1, Table 3.3s6-2 . l' ' Alternate acceptable answer is 3. In T-ll2, steps EB-7.09 7, EB-8 and EB-9 ask for 5 ADS valves or ADS /SRV valves. However, in step EB-10, it asks the operator i
"Are at least 3 ADS /SRVs open". For this reason either 3 or 5 are acceptable answers.
p37, T-ll2, Emergency Blowdown t I
'
1
-- , .. - - - . .
NRC Senior Rocctor Op3rctor Exc3 Rocponsos (Continucd) Pago 7 of 9 7.10 b. Acceptable answer according to EP-101 is classifying this as an Unusual Event since for an Alert classification, Steam Jet Air Ejector Discharge Radiation monitor exceeds 2.lPS mR/hr. Question states 210 R/hr. For conservatism, the SRO candidate may answer the same as the answer key, which states Alert.
However, strict interpretation could lead the SRO candidate to only classify as an Unusual Event. Both answers should be accepted.
REF: EP-101, pg 13, Damage of Fuel
,
o e l l l i
. ,. , .,,,n-- -, , . - - - - - , - - - - - -- - ------ - - - - - - - - - ~ - ' -
'
~$aC Senior Rocctor Op3rator Exco Responses (Continusd)
Pcgo 8 of 9 Category 8: Administrative Procedures, Conditions, and Limitations 8.01 1. Answer key is incorrect. This is a safety limit violation in accordance with Technical Specifications for Thermal Power, Low Pressure or Low Flow.
REF: Technical Specification 2.1.1 8.03 a. Alternate correct answer should include the portion of A-7 dealing with job responsibilities of the Control Supervisor (section 5.1.2.2) and Control Room Operators (section 5.1.2.4). Although the wording is different than section 5.2.8 of A-7 daaling with Initie. tion of a Scram or Shutdown, tho intent is the same.and the SROs should be given credit.
REF: A-7, pgs 9, 19, 11, 17 - 8.03 b. (1) Alternate acceptable answers should be his designated alternate or Plant Manager. Due to the reorganization within PECo, the title will change to Plant Manager vice Station Superintendent. A-7 states Station Superintendent or his designated alternate.
REF: A-7, 5.2.12 8.03 b. (2) Alternate acceptable answer should be Shift Supervision. They are Senior Licensed Operators.
REF: A-7, 5.2.12 8.05 a. Answer key states that permission to perform an ST may be delegated. In accordance with A-43, this may not be delegated. Shift Supervision's permission to perform an ST must be given as well as having the responsible CO/ACO's permission. This permission can not be delegated.
REF: A-43, p 2, part 2.4
. - . _ . _ - , _ _ . . _ . - . . _ . _ _ - _ .
_ _ _ _
~'
NRC S3nior R3 actor Oparator Exts RGsp2nsGs (Continusd) j Pago 9 of 9
8.06 c. An acceptable alternate answer is (COL) in that an independent verification of a permit does not have to 1 be done if the equipment check off list (COL) is going to be performed. This is in accordance with procedure A-41 suctions 5.4.1 and 5.4.4 REF: A-41, pg 12, parts 5.4.1 and 5.4.4 ' 8.10 Candidate answers should be graf.ed based upon what interpretative Tech Spec logic they used. The answer key takes the conservative approach by placing the "A" channel in the tripped condition. However, a strict interpretation of Technical Specifications would lead the candidate to come to the conclusion that no actions are required since Tech Spec Table 3.3.1-1 does not require the IRM. neutron flux high trip to be OPERABLE in OPCON 1, and table notation (b) states that the function is automatically bypassed when the mode switch is in RUN and the associated APRM is not downscale (given in the question). - REF: Technical Specification 3.3.1 l
,
l
'
l l
. _ . . - _ _ _ . - . _ , . _ - , - _ _ . . _ _ - . _ , ,. _ _ _ . _ _ .l
-
.
. ATTACHMENT 4 NRC Resolution of Facility Comments on SR0 Examination Administered on June 7, 1988 Category 5: Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 5.04(a). Answer can be either LARGER or SMALLER depending on assumptions (which were not required by the question). When dealing with how moving a control rod will effect overall thermal power, many vart-ables come into play and the final answer will be dependent on assumptions.
REF: LOT 1490, pgs. 4-12 NRC Resolution: Rejected (This question compares deep rod movement with shallow rod movement. Movement of a deep rod has larger effect on overall core thermal power. No questions were asked about other assumptions during the examination).
5.05. SR0s tnvuid not lose credit if the problem was worked both at 100*F/hr and at less than 100 F/hr cooldown rates since GP-3 note states cooldown is limited to 5100 degrees F per hour.
REF: GP-3, p 16 NRC Resolution: Rejected per GP-3 note, maximum cool down rate is 100 F/hr.
However, partial crcdit will be given for method used for assumptions other than 100 F/hr.
5.08(c). Can be confusing to the candidates since steam flow is not directly inputted in a manual core thermal power evaluation as per RE-101.
Consider alternate answer depending on assumptions.
REF: RE-101, Core Thermal Power Evaluation NRC Resolution: Rejected (Candidate should know that steam flow enters into the heat balance, although RE-101 assures steam flow = feedwater flow).
5.08(d). Answer is incorrect. The correct answer should be FALSE vice TRUE.
The Reactor Operator Exam question 1.09 part (d) is the same and it's answer is gisen as FALSE. This answer is correct and supported by the heat balance calculation.
REF: Reactor Operator Exam, Question 1.09 (d) NRC Resolution: Accepted (changed answer key) 5.09(a) Additional correct answer: _W MAPLHGR/APIHGR
% ~
LOC Although LOT-1410 uses MAPLHGR/APLHGRLOC, Technical Specification 3.2.1 defines the LCO for APLHGR, and not MAPLHGR.
l
- -_-__--__-_-___-__-__________-______-__-_-__ _ _____-__ -__-___--__ _- _ _-__ - __-__ _ - _ _ - - - - _ _ _ _ _ _ - _ _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ . _ _ _ _
.-
. Attachment 4 2 REF: Technical Specification 3.2.1 NRC' Resolution: Partially accepted (The answer key changed to the correct definition in TS, i.e., MAPLHGR/APLHGR-LCO.
5.10. Answers should be accepted depending on.SR0 candidates supporting information. The candidates were not provided the note from Table 15.2-11 which is referenced on Figure 15.2-9 which they were given.
In addition, Table 15.0-2 was not provided to the candidates which lists the input parameters and initial conditions for transients.
Also, in accordance with the objectives for LOT-1580, BWR Transient Analysis, the candidates were not held responsible for a loss of All Feedwater Flow Transient. Due to these reasons, a range of answers should be accepted based on student's assumptions / justifications.
. REF: FSAR, Figure 15.2-9, Table 15.0-2, Table 15.2-11;' LOT-1570, p 1 NRC Resolution: Accepted (Exact wording not required, candidate's correct assumption / justification will be considerate in the grading) , 5.11(a). Subjective question dealing with water hammer. Candidates could justify choices 1, 2, or 3. Many flow changes (starting flow, cessa-tion of low) can be linked to cases of water hammer. For this reason, acceptable choices are 1, 2, or 3.
REF: LOT 1291, pgs. 1-7 NRC Resolution: Partially accepted (Question asks to identify most itkely cause of water hammer. Answers 1 and 2 are more likely than Answers 3 and 4.
Changed Answer Key to 1 or 2).
Category 6: Plant Systems Design, Control, and Instrumentation.
6.01. Question is worded such that it could be misinterpreted since the HPCI turbine trip is at 15" Hg Vac and the question implies that the HPCI pump is still working with a suction of 18" Hg Vac. Therefore, answer (a) "Continue to inject" can be justified due to the fact the pump did not trip at 15" Hg Vac.
REF: LOT 0340, pgs. 11, 12 NRC Resolution: Rejected (the wording of the question is cleir to indicate pertinent parameter:. An assumption of a failure to trip is not indicated on the question).
6.03(a). Alternate answer is NOTHING. Given the conditions stated in the question, recirc pump M/A stations may already be at 28% or lower.
The operating map referenced in the question does not specify "speed" only flow which can be due to natural circulation and/or forced circulation. For this reason, if the recirc pumps are already'at or below 28% speed, nothing will occur when feedflow drops below 20%.
' Attachment'4 3 REF: LOT 0040, pgs. 25, 26, 27 NRC Resolution: Accepted (Changed Answer Key)
-
6.04(a). Correct-answer is 118 seconds. Mod 86-189 changed this time delay from 58 seconds to 118 seconds. This is referenced in S36.1.B section 8.2 note, "116 to 120 seconds after manual initiation, a 10 minute time is enabled".
REF: $36.1B section 8.2 NOTE, step 8.3 NRC Resolution: Accepted (Changed Answer Key) 6.04(b). Additional correct answers: 1) Decreasing ' tank level
'
2) Decreasing reactor power 3) Continuity lights extinguish.
REF: LOT 0310, pgs. 10, 17 NRC Resolution: See resolution.to similar RO Question Changed the Answer Key to include the additional answer 6.05(a). Additional acceptable answers are: Condenser vacuum low Manual Since the question did not specifically set other conditions or state "Automatic", these answers are acceptable in accordance with Technical Specifications depending on candidates interpretation of the question.
REF: Technical Specifications LC0 3.3.2, Table 3.3.2-1, Table 3.3.2-2 i NRC Resolution: Partially accepted (Condenser vacuum low is not acceptable because safety valves, which bypass the scram signal, are not open until after 2 BPVs'are open. This would not occur until the pressure is greater than 920 psig. Reference: F01.1.a and GP-2. However, Manual is an additional accept-able answer. Answer Key modified to include this and revised point value distribution). ,
.
.l 6.07(a). Standby Gas Treatment System will also start if aligned to the Refuel l Floor Ventilation System via the Slide Gate Dampers. ,
l l _- _ _-__________ _ - _ __ _ _ _ _ - - - - = _ _ _ _ . --
._. _
i Attachment 4- 9 REF: -P&ID M-76, shts 5 and 6 NRC Resolution: Accepted (Changed Answer Key) 6.08. During post exam review, the examiner was made aware of the. error in the answer key dealing with the EHC numbers. Answer'should state 25-vice 40 and 18% vice 175 NRC Resolution: Accepted (Changed Answer Key)
<
6.10(a). Question was worded such that interpretation of the phrase, "... also has channel A of the IRM"s selected" could be the IRM/APRM selector switch on the C603 panel. This is due to the use of the word
"channel" instead of the word "detector". Alternate answer in this.
case would be NONE.
REF: 10T 0250, pgs. 10-14; LOT 0270, pgs. 11, 14 NRC Resolution: Accepted (Changed Answer Key) 6.10(d). Alternate acceptable answer would not include the rod block as part of the answer. This is due to the fact that the candidate may inter-pret the question as those actions which occur at 15% power. Since " the rod block occurs at 12%, they may only say scram signal (which is generated at 15% in startup).
REF: Technical Specifications, Table 2.2.1-1, Table 3.3.6-2 J - NRC Resolution: Accepted (Changed Answer Key) Category 7: Procedures - Normal, Abnormal, Emergency, and Radiological , Control < 7.02(b). Answer key breaks down point values into. credit for calculation and , credit for formula. The question does not ask the SRO candidate to show this in his answer. Request that the formula and calculation not be required for full credit.
, REF: SR0 Exam Question 7.02, Answer Key 7.02 NRC Resolution: Accepted (Changed Answer Key).
7.04(a) The question asks the candidate for two purposes. The caution in S51.8.B shows that the two purposes given in the answer are, in essence, saying the same. Either should be counted for full credit and other answers given by the SR0s should not cause a loss of credit since the question asked to give two different purposes.
REF: S51.8.8, pg. 5 _ - -__ _ _ _ _ - - ___ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -
Attachment 4 5 NRC Resolution: Accepted (Changed Answer Key to reflect this comment. To prevent inadvertent draining of the vessel is the main idea which will get full credit. Other reasonable answers will not cause a loss of credit.
7.06(a). and (b). Confusion exists as to what the question was asking for versus what the answer key was answering. If the question meant to state level was oscillating 15 inches, did this mean from normal to 15 and back to normal; or did it mean from normal level 5 inches (question only lists 5 inches, which would not be oscillation so it's assumed it was meant to be 15 inches).
In either case, the following OT Procedures could be entered based on assumptions.
OT-100 Low Level PT-110 Hi Level OT-104 Unexplained Reactivity Insertion The SR0's answer to part (b) will then be in accordance with part (a). The answer key lists two OTs in part (a) but only addresses one in part (b).
Acceptable answers will be according to question interpretation.
Also, part (b) question states "What-actions would you, as the Shift Supervisor, direct the operator to take?" The Shift Supervisor would not direct the operator to enter T-100 if a scram occurs but, rather, would enter T-100 and direct the operators accordingly. Therefore, this step should not be counted as part of the answer to part (b).
In addition, the SR0 candidate may enter OT-104, Unexplained Reactivity Insertion, in response to the power increase which the reactor may undergo when level oscillates so this answer should be accepted.
REF: OT-100, pg. 1: 0T-104, pg. 1; OT-110, pg. 1 NRC Resolution: Practically accepted (a typographical error occurred in preparation, omitting - sign in 5 inches and leaving it as +5 inches. Answer Key is modified to an oscillation of + 5 inches. Entry to OT-104 is not warranted by the condition, but no credit will be removed if it is mentioned.
Also, for the question as asked entry into OT-100 is not warranted because no low level is expected.
7.07(c). Answer key typographical error. Answer should read ON-107 vice ON-105, for "Control Rod Drive System Problems".
REF: Off Normal Procedure Index
,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ . _
Attachment 4 6 l NRC Resolution: Accepted (Changed Answer Key) 7.07(d). Acceptable answer is entry into ON-109, "Total Loss of the SRM, IRM, or APRM Systems". This is because Technical specifications require the SRM Upscale and Inoperative TRIPS be operable until the IRMs are on range 8 or higher in order to meet LCO 3.3.6 "Control Rod Block Instrumentation".
REF: Technical Specifications 3.3.6, Table 3.3.6-1, : Table 3.3.6-2 NRC Resolution: Accepted (Changed Answer Key) 7.09(b). Alternate acceptable answer is 3. In T-112, steps EB-7, EB-8 and EB-9 ask for 5 ADS valves or ADS /SRV-valves. However, in step EB-10, it asks the operator "Are at least 3 ADS /SRVs open". For this reason either 3 or 5 are acceptable answers.
REF: T-112, Emergency Blowdown NRC Resolution: Accepted (Changed Answer Key) Category 8: Administrative Procedures, Conditions, and Limitations 8.01.1. Answer key is incorrect. This is a safety limit violation in accordance with Technical Specifications for Thermal Power, Low Pressure or Low Flow.
REF: Technical Specification 2.1.1 NRC Resolution: Accepted (Changed Answer Key) 8.03(a). Alternate correct answer should include the portion of A-7 dealing with job responsibilities of the Control Supervisor (section 5.1.2.2) and Control Room Operators (section 5.1.2.4). Although the wording is different than section 5.2.8 of A-7 dealing with Initiation of a Scram or Shutdown, the intent is the same and the SR0s should be given credit.
REF: A-7, pgs. 9, 10, 11, 17 NRC Resolution: Accepted (Exact wording not required) 8.03(b)(1). Alternate acceptable answers should be his designated alternate or Plant Manager. Due to the reorganization within PECo, the title will enange to Plant Manager vice Station Superintendent.
A-7 states Station Superintendent or his designated alternate.
REF: A-7, 5.2.12
Attachment 4 7 NRC Resolution: Accepted (Changed Answer Key) 8.03(b)(2). Alternate acceptable answer should be Shift Supervision. They are Senior Licensed Operators.
REF: A-7, 5.2.12 NRC Resolution: Accepted (Changed Answer Key) 8.05(a). Answer key states that permission to perform an ST may be delegated.
In accordance with A-43, the may not be delegated. Shift Supervis-ion's permission to perform an ST must be given as well as having the responsible C0/ACO's permission. This permission can not be delegated.
REF: A-43, p. 2, part 2.4 N_RC Resolution: Rejected (Step 2.4 of A-43 states Shift Supervision and or the ACO or C0 shall be responsible for giving permission to start an ST. Further, Paragraph 5.5.3 again states that the AC0/C0 can give this permission. If it is the policy at Limerick not to delegate this responsibility, the Proce-dure A-43 needs revision to remove the flexibility currently allowed.
8.06(c). An acceptable alternate answer is (COL) in that an independent veri-fication of a permit does not have to be done if the equipment check of f list (COL) is going to be performed. This is in accordance with procedure A-41 sections 5.4.1 and 5.4.4.
REF: A-41, pg. 12, parts 5.4 1 and 5.4.4 NRC Resolution: Accepted (Changed Answer Key) 8.10. Candidate answers should be graded based upon what interpretative Tech Spec logic they used. The answer key takes the conservative l i approach by placing the "A" channel in the tripped condition. How-ever, a strict interpretation of Technical Specifications would lead the candidate to come to the conclusion that no actions are required since Tech Spec Table 3.3.1-1 does not require the IRM neutron flux high trip to be OPERABLE in OPCON 1, and table notation (b) states that the function is automatically bypassed when the mode switch is in RUN and the associated APRM is not downscale (given in the question).
REF: Technical specification 3.3.1 NRC Resolution: Comment not accepted. The APRM nownscale Scram function of APRM C would not be operable with the companions IRM inoperable. No change to answer key.
- _ . _. . _ _ l -
, ,
I
' Attachment 4 8-Question Number: 1:01 d Facility l Comment: - Answer can be either TRUE or FALSE. The phrase "as a resultlof" is' vague, and-can be' interpreted two ways by the candidate.
1) Delayed neutrons are produced upon decay of delayed neutron precursors, and not from the fission of U-235 and U-238!directly, therefore the answer is FALSE.
2) Delayed neutrons are even+ually produced after fission of U-235 and U-238, therefore the answer is:TRUE. ' NRC Resolution: Disagree with comment. Delayed neutrons can only result if a fission has occurred. The delayed neutrca precursors. occur through. beta minus decay ~of the fission products produced f rom fission therefore the neutrons result only from fission occurring. Answer key was not changed.
Question Number: 1.02 d Facility Comment: Answer should be SDM INCREASES. Increasing PU-240, an absorber of epithermal neutrons, reduces keff in the core, which results in an increased SDM.
NRC Resolution: Agree with comment. Answer key is modified to accept SOM Increases as the ' correct answer.
Question Number: 1.05 a.
Facility Comment: Additional correct answer: INCREASING TEMPERATURE OF RECIRC PUMP SUCTION.
During startup, Recirc Pump Suction Temperature is monitored, and is considered to be Reactor Water Temperature for an operating Recirc pump.
Reactor water temperature is recorded and monitored to the heating range per GP-2. ' NRC Resolution: Agree with comment. Increasing Recirc Pump suction temperature. Will be accepted as an alternate answer.
l
!
I J
'l o
Attachment 4 9
Question Number: 1.05 b Facility Comment: Interval 2 and Interval 3 are both correct answers.
The period for Interval 2 is 90.11 sec.
The period for Interval 3 is 80.03 sec.
Both are greater than 80 seconds, and either could be an interval in which the heating range was entered.
NRC Resolution: Disagree with comment. The answer for Interval 3,80.03 sec, is beyond the significant digits that are used when referring to reactor period. Also the question indicated that only ONE answer was correct and the 10 second change in period in Interval 2 is the most correct indicator of entry into the heating range.
Question Number: 1.08 Facility Comment: Cand1Jates may discuss how the reactivity coefficients change when feedwater temperature is reduced, as well as how the coefficients affect the transient, and should not be penalized for this. ' NRC Resolution: Agree with comment, however candidates will be penalized for incorrect additional information. The candidates were informed of this during the examination briefing.
Question Number: 1.08 c Facility Comment FSAR analysis assumes a decrease in Feedwater temperature of 100 F. The isolation of a feedwater train would result in less of a change in temperature than 100 degrees F, and with appropriate operator action, a scram on high flux can be avoided. Operator action per OT-104 "Unexplained Reactivity Insertion" would result in voids decreasing followed by voids increasing.
NRC Resolution: Partially agree with comment. Partial credit will be given for discussions concerning reducing recirculation flow.
- _ _ - _ _ _ _ _ _ _ _ _ _ . _____ _ - - .
Attachment 4 10 Question Number: 1.09 Facility Comment: The question is similar to SR0 5.08. The wording is awkward and extremely confusing. The question is not based on a Learning Objective.
NRC Resolution: Disagree with comment. Learning objective LOT-1300 #2 indicates that the candidate shall ce able to calculate a reactor heat balance. If the candidate can calculate a reactor heat balance then the candidate should be able to analyze the situations in the question to determine if a correct heat balance was performed.
Question Number: 1.10 a.
Facility Comment: Additional correct answers are: REACTOR POWER - Class discussion included graphing NPSH vs Reactor Power for Recirc pumps. Reactor Power is a parameter available to the operator on the Process Computer.
RECIRC SUCTION TEMPERATURE - Saturation temperature is directly related to NPSH, and is a parameter available to the operator on the Process Computer.
RECIRC PUMP SPEED - Friction loss is affected by pump speed and is directly related to NPSH. It available on a meter on the C601 panel in the main control room.
NRC Resolution: Disagree with comment. The question asked for factors that CONTRIBUTED to the NPSH. Reactor power does not contribute to NPSH, rather the increase in feedwater flow does. Recirc Suction Temperature is a result of feedwater temperature and feedwater flow. The facility's reference included with comment did not differentiate between REQUIRED and AVAILABLE NPSH. Recirc pump speed is considered in determining REQUIRED NPSH but has negligible effect on AVAILABLE NPSH. Answer key was not changed.
Attachment 4 11 Question Number: 1.11 a Facility Comment: "Bundle" and "Assembly" are used interchangeably, and both should be accepted.
NRC Resolution: Agree with comment.
Question Number: 1.12 a Facility Comment: Additional correct answer: MAPLHGR/APLHGR (LCO) Although LOT-1410 uses MAPLHGR (LCO), Tech Spec 3.2.1 defines the LCO for APLHGR, not PAPLHGR.
NRC Resolution: Answer key is modified to accept PAPLHGR/APLHGR (LCO) as correct answer. The answer key is modified to NOT accept MAPLHGR/MAPLHGR (LCO) as a correct answer. The answer was changed because the reference provided by the facility's comments is more accurate or disagrees with the reference material provided for use by the examiner in developing the written examination. The originally referenced material should be corrected or destroyed.
Question Number: 2.02 c Facility Comment: Both WOULD and WOULD NOT are correct answers, depending on candidate's assumptions.
1) If C RHRSW is providing cooling water to the B RHR heat exchanger, it will trip on a LOCA signal, since it is powered by the Unit 1 0-12 Bus, and WOULD NOT is the correct answer.
2) If D RHRSW is providing cooling water to the B RHR head exchanger, it will not trip on a LOCA signal, since it is powered by the Unit 2 0-22 Bus, and WOULD is the correct answer.
NRC Resolution: Delete part c of the question. Value for the question has been decreased from 3.0 points to 2.5 points.
Attachment 4 12 Question Number: 2.04 a Facility Comment: Correct answer is 118 seconds. MOD 86-189 changed this time delay from 58 seconds to 118 seconds. This is referenced in 836.1.b, section 8.2 NOTE: "116 to 120 seconds after manual initiation a 10 minute timer is enabled."
NRC Resolution: Answer key is modified to accept 118 seconds as the correct answer. The answer was changed because the reference provided by the facility's comments is more accurate or disagrees with the reference material provided for use by the examiner in developing the written examination. The originally referenced material should be corrected or destroyed.
Question Number: 2.04 c Facility Comment: Additional correct answers: 1) Decreasing tank level 2) Decreasing reactor power 3) Continuity Lights extinguish NRC Resolution: Agree with corument. Answer key is modified to accept decreasing tank level and decreasing reactor power as the additional correct answers. Continuity lights extinguish will be accepted in lieu of squib valves firing.
Question Number: 2.05 a 2 Facility Comment: Correct answer is 455.
NRC Resolution: Agree with comment. Answer key is modified to accept 455 as the correct answer. This was a typographical error on the answer key.
l l
-
Attachment 4 13 Question Number: 2.05 a 3 Facility Comment: Question should be deleted. Check valve will open when discharge pressure is greater than reactor pressure. Discharge pressure is not required knowledge for RO.
NRC Resolution: Disagree with comment. The check valve will open when reactor pressure is less than shutoff head of the pump. Learning objective 5.b of LOT-0350 requires that the R0 know the shut off head of the pump. Answer key was not changed. Partial credit was given for understanding the concept of where the valve will open.
Question Number: 2.05 b Design pressure should be deleted from this question. Design flow is required for R0 knowledge. Design pressure for rated flow is not required knowledge for RO.
NRC Resolution: Agree with comment. Full credit will be given for design flow.
Question Number: 2.06 e.
Facility Comment: Suppression pool spray and full flow test valves also close. Candidates should not be penalized for including these in the answer.
NRC Resolution: Disagree with comment. These valves should not be open on a loop that is being operated in the Shutdown cooling mode as it would result in draining the vessel, therefore they would not be expected to reposition on the LPCI initiation. Points will be deducted if these are included as repositioning.
Attachment 4 14 Question Number: 2.08 c Facility Comment: Answer should be either NO SYSTEM or ESW, depending on candidate's interpretation of the question. ESW can be used to cool the RECW heat exchanger, and can provide cooling to the Recirc pump seal and motor oil coolers which are normally cooled by RECW. Depending on the assumptions made by the candidate, either NO SYSTEM or ESW is an appropriate answer.
NRC Resolution: Part c of the question will be deleted because question did not specify components that can be cooled. Point value for the question has been decreased from 2.5 points to 2.0 points.
Question Number: 2.09 a Facility Comment: Additional correct answers: OVERSPEED LOW CONDENSER VACUUM Each overspeed trip is independent of the other feedpumps, and could have caused one feedpump to trip.
Each RFP turbine exhausts to a different shell of the main condenser, independent of the other turbines, and could have caused one feedpump to trip.
NRC Resolution: Agree with comment. Answer key is modified to also accept Overspeed and Low Condenser Vacuum as correct answers.
Question Number: 3.01 a Facility Comment: Delete this question. The Recirc Pump Master Controller is not used at LBS, and the M/A Transfer Station is not used in auto. This question should be deleted since it does not apply to LGS.
NRC Resolution: Disagree with comment. The equipment is installed and training material contains information concerning its operation. The action described in the question could occur and the candidate should be aware of the consequences of the action.
Attachment 4 15 Question Number: 3.01 b Facility Comment: 75% flow is equivalent to approximately 60% pump speed. Either is an acceptable answer.
NRC Resolution: Agree with comment, however Figure 2 of LOT-0040 indicates that the limiter is based on 75% speed and should be corrected. The question requested speed, therefore the answer key is changed to 605 speed. Partial credit will be given for stating 75% core flow.
Question Number: 3.01 c Facility Comment: Delete this question. The Recirc Pump Master Controller is not used at LGS, and the M/A Transfer Station is not used in auto. This question should be deleted since it does not apply to LGS.
NRC Resolution: Agree with comment. Part c has been deleted. Point value for question decreased from 3.0 to 2.0 points.
NOTE: Part C was deleted because the possibility of this occuring, due to the fact that the Facility does not use the Master Controller is minimal due to procedural constraints, however the possibility of inadvertant operation of a switch on the individual recirc pump controller is significant enough to warrant the operator understanding the response of the system if this occurs.
Operators use buttons on the individual recirc pump controllers where the transfer switch referred to in part a is located.
Question Number: 3.02 b Facility Comment: Both HIGHER THAN ACTUAL and SAME AS ACTUAL are correct answers. Wide range is calibrated with no jet pump flow, and would read the same as actual assuming no jet pump flow. Assuming jet pump flow, wide range would read higher than actual.
NRC Resolution: Agree with comment. Answer key is modified to also accept HIGHER THAN ACTUAL.
, _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . __ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . - _ _ _ . . _ _ _ _ _ _
Attachment 4 16 Question Number: 3.03 a Facility Comment: Candidate should not be penalized for including LOW CONDENSER VACUUM 10.5 PSIA as an additional answer, since this could be bypassed or active in startup, depending on assumptions made. Candidate should not be penalized for including MANUAL, due to the wording of the question.
NRC Resolution: Partially agree with comment. Low Condenser Vacuum is not acceptable because , the stop valves, which bypass the scram signal, are not opened until after 2 ) Bypass Valves are open. This would not occur until pressure is greater than 920 psig. REF: 501.1.A and Gp-2.
Answer key is modified to also accept Manual as a correct answer.
Question Number: 3,04 d Facility Comment: Action 3 is not a correct answer. If the reactor is in startup mode and power reaches 15%, and scram signal is expected. A rod block should have been generated at 12% power. If rods can be withdrawn to 15%, the rod b'ock signal has failed, and a scram signal is now expected. Action 2 is the only correct answer.
NRC Resolution: Agree with comment. Answer key is modified to only accept Action 2 as the correct answer. Point value was changed from 0.25 for Action 2 to 0.5 point.
Question Number: 3.04 e Facility Comment: NO ACTION is also a correct answer. Since criticality can be reached at any time, a rod block will not always be generated in the condition.
NRC Resolution: Disagree with comment. "Approach to Criticality" is referred to in GP-2 at the point at which rod withdrawal commences. Criticality could not be reached with all rods in or SDM requirement could not be met. A rod block is generated because with all rods IRM's are downscale. The rod block is bypassed only if all IRM's are on range 1. Answer key was not changed.
- ___--_______ ._
I Attachment 4 17 Question Numt : 3.05 a During ex .i grading it was determined that the answer to part a should be 525 'ead of 420. The timers act sequentially on a low level signal w drywell pressure. Refer to figure LOT-0330-6. Answer key is modifi 7t 525 seconds as the correct answer.
Question h 3.06 Facility Comment: Numbers and calculation on LOT 0590-6 should be accepted in lieu of discussicrs. Candidate shculd not be penalized for continuing the transient e til the plant stabilizes.
NRC Resolution: Partially agree with comment. Instructions for the examination require that the answer be on answer sheets. Credit will be given for answer on diagram.
Candidate will not be penalized for continuing the transient if correct information is provided.
Question Number: 3.07 a Facility Cencent Standby Gas Treatment System will also start if aligned to the Refuel Floor Ventilation System.
NRC Resolution: Agree with commer.t. Answer key is modified to accept SBGT start if aligned to the refuel floor as an additional required answer. Comment could not be verifi de by reference material provided because it was illegible, however comm nt was verified in the Secondary Containment LP.
Question Number: 4.01 a Facility Comment: Directions in A-7 for an RO to shutdown the plant or scram are covered in two sections of A-7, using different wording. Section 5.1.2.4 is appropriate as section 5.2.8 as an answer.
NRC Resolution: Comment noted. The conaitions specified in 5.1.2.4 are conceptually l equivalent to 3 of the conditions specified in 5.2.8 and credit will be given , if the correct concer,t is specified, however section 5.2.8 gives additional I guidance to the operators which will be required in order to obtain full credit. Answer key wa s not changed.
_.
f Attachment 4 18 Question Number: 4.07 a Facility Comment: Answers on answer key are redundant: Additional correct answer: TO REMIND OPERATOR THAT THE MIN FLOW IS BLOC CLOSED.
This is because the basic purpose of the Blocking Procedures, i.e. assure safety of workers and avoid equipment damage.
. NRC Resolution: Partially agree with comment. Answer key is modified to accept prevent draining the vessel as the correct answer for full credit. "To remind operator that the min flow is blocked closed" will' not be accepted as an additional correct answer because the reference material supplied with the facility comments did not support the additional answer.
,
:4 ATTACHMENT 5 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Licensee Docket No.: 50-352 Facility License No.: .
NPF-39
Operating Test administered at: Limerick Simulator Operating Tests Given On: June 8 and 9, 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed: 1. In IC-20 at 75% power, the conditions are such that the APRM upscale alarms and rod block setpoints are exceeded and the recirculation flow is in a critical speed instability area.
2. Malfunction 72 (SJAE steam supply valve fails closed)-had no effect on condenser vacuum. The simulator operator suggested that this effect was due to a plant modification which added an automatic swap to an alternate steam supply, had been implemented in the software but the malfunction book-and panel hardware had not been updated.
3. Malfunction 44A (reactor pressure recorder XR-6233 failing upscale) did not produce a half-scram and other actions that were-listed in the malfunction book.
4. Malfunction 140E (stuck open safety relief valve) would cause the valve to open but allowed it to close when the operator manipulated the switch. It should have remained open regardless of operator actions.
5. Malfunction 110 (ATWS with failure of the RRCS, main turbine trip, and the main condenser available) reactor pressure remained at greater than 1190 psig regardless of power. Also, reactor level oscillated between 20 and 40 inches then dropped to -300 inches. An attempt to reproduce this transient caused similar effects.
... .. . . . ..
.. _ _ _ _ - _ _ _ _ - _ _ _ _ _ - _
}}