IR 05000277/1986022
| ML20207S265 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/09/1987 |
| From: | Collins S, Keller R, Kolonauski L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207S252 | List: |
| References | |
| 50-277-86-22OL, 50-278-86-23OL, NUDOCS 8703190214 | |
| Download: ML20207S265 (147) | |
Text
{{#Wiki_filter:. U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NOS. 86-22(0L) and 86-23(OL) FACILITY DOCKET NOS.
50-277 and 50-278 FACILITY LICENSE NOS.
DPR-44 and DPR-56 l LICENSEE: Philadelphia Electric Company 2301 Market Street
Philadelphia, Pennsylvania 19101 l FACILITY: Peach Bottom Atomic Power Company EXAMINATION DATES: December 8-10, 1986 hbW (d{{/h(ULLA( 1/1 i[87 / CHIEF EXAMINER: Ly@ Kolonauski, Reactor Engineer (Examiner) Date eL f#/.M p t 9/g 7 REVIEWED BY: ._ . M1 er, C tef, rojects Section 1C Da(e ' - _obert g h d (Collins, Deputy Director S[d87 famuelJ.([hMj __ Date APPROVED BY: 01 vision of Reactor Projects SUMMARY: Operator licensing examinations were administered to four (4) Senior Reactor Operator candidates, three (3) Reactor Operator candidates, and one (1) Instructor Certification candidate during the week of December 8,1986. One candidate failed the SRO written examination. All remaining candidates passed the written examinations. All candidates passed the operating examinations.
ggKf50bb77 D ho PDM
. REPORT DETAILS TYPE OF EXAMS: Initial Replacement _ X Requalification EXAM RESULTS: l R0 l 3R0 l Inst. Cert l l Pass / Fail l Pass / Fail l Pass / Fail l l
I I ' I I I I I l Written Exam l 3/0
3/1
1/0 l I I I I I I I I I I l Oral Exam l 3/0 l 4/0
1/0 I I I I I I I I I I I l0verall l 3/0
3/1
1/0 l l
I I I 1.
CHIEF EXAMINER AT SITE: Lynn Kolonauski, NRC 2.
OTHER EXAMINERS: David Lange, NRC 3.
Summary of generic strengths or deficiencies noted on oral exams: Selected SR0 candidates indicated a high level of proficiency in their use of the Emergency Operating Procedures and the Emergency Plan.
4.
Summary of generic strengths or deficiencies noted from grading of written exams: A weakness was identified in Categories 1 and 5, " Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics".
5.
Cortments on availability of, and candidate familiarization with plant reference material in the control room: Plant reference material was readily available, and most candidates were adequately familiar with it. One candidate, however, was particularly weak in the use of Piping and Instrumentation diagram. Personnel Present at Exit Interview: NRC Personnel Lynn Kolonauski, Reactor Engineer Examiner David Lange, Lead (BWR) Reactor Engineer Examiner ' Thomas Johnson, Senior Resident Inspector Facility Personnel Drew Smith, Superintendent of Operations Sheryl Wookey, Training Coordinator, PBAPS Steve Roberts, Operations Engineer R. S. Fleischmann, PBAPS Plant Manager 7.
Summary of NRC Comments made at exit interview: The generic strengths and weaknesses noted in paragraphs 3 and 4 were discussed.
The personnel of the PBAPS Training and Operations departments were cooperative throughout the examination period, and the examiners experienced few plant access delays.
8.
Summary of facility comments and commitments made at exit interview: The facility acknowledged the NRC comments noted in paragraphs 3 and 4.
The facility personnel thanked the examiners for their cooperation throughout the examination week.
9.
CHANGES MADE TO WRITTEN EXAM DURING EXAMINATION REVIEW: All comments about the written examinations were resolved during the exam review. The examiners returned to the regional office with no unresolved comments. The following list represents significant changes made to the examinations.
Question No.
Change Justification 1.05/5.04 a.2 Answer changed to Original answer " decreases".
incorrect.
2.01 Changed answer to: Utilizing Figure 1
- 1, 2 breakers open to answer question
- 11, 22 breakers close lead candidates to
- 12, 21 breakers remain give only the open because their response of Fig. I control switches are equipment.
in pull-to-loc. - --- - ... -.. .-
<- i , Question No.
Change Justification
2.03 a.
Included A0-4807.
Additional infor- . (for Unit 2 only) mation supplied
in answer.
by facility.
2.06/6.08 c.
Included 140 psig Additional infor- ,
(downstream).to mation required j protect piping.
for full credit.
2.10 b 3 " Decrease".
A decrease in cooling flow may occur because actual plant re- ! sponse is sluggish.
The plant reference material did not
contain this infor-mation.
, , 3.05/6.03 b.
Accepted also Acceptable Note 20 of Trip alternate answer.
i Procedures.
! 3.10 b.
Graded according to Inaccurate term candidate's interpre-(RFP Servo Motor) tation of " servo motor".
used in question.
' 3.11 a Credit given for a CS leak detection
l description of a CS is only for breaks line break outside outside of the of the shroud, shroud.
,
3.12 b.
Question deleted.
PBAPS does not use a reverse pull sheet.
4.07 Included Unit 2 Additional infor-l Offgas Main Absorber mation supplied by
Bed Radiation Detector, facility.
5.11 Changed answer to Auto flow control
reflect use of 102.5% not used at PBAPS.
Kf curve.
d ' 8.04 a.
Changed TS 3.5.F.3.b Correct answer.
to TS 3.5.F.4.b.
i 8.04 a., b.
TS 3.5.F.3.a also Correct alternate l accepted.
answer.
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- - , Question No.
Change Justification
8.04 c.
TS 3.5.F.1 added to Additional infor-answer key.
mation required for full credit based on facility review.
8.07 TS 3.9.A.1 added to Additional infor-answer key.
mation provided by facility.
8.08 Accepted also Note 20 Correct alternate of Trip procedure, answer.
8.09 a.
Added to key: Original answer ' 1. MCPR > 1.07 incomplete.
2. CTP < 25% 8.11 Changed answer to The question TS 3.8.c.4 (b,c,e) addressed the Main which addresses Stack Rad Monitors the Main Stack Rad as opposed to the Monitors.
Main Stack HIGH Rad Monitors.
Attachments: 1.
Written Examination and Answer Key (RO) 2.
Written Examination and Answer Key (SRO) > , . r , ] , t-e ,--_---..,_..._,-...r-__ . -. -..., -,.. .m... . _ _, _ - _ _. ,.e.
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V Arrnchment / . l hb ' U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _PEAgH_ggTTQM_g&3 __ REACTOR TYPE: _gWR-@E4_________________ DATE ADMINISTERED: _ggflg[gg____ _ EXAMINER: _LANGE _Q.______ _ ___ t CANDIDATE: _ _ __________________ INglgggligNg_lg_ggNDiggIE1 Use separate paper for the answers.
Write answers on one side only.
Stcple question sheet on top of the answer sheets.
Points for each qu;stion are indicated in parentheses after the question.
The passing grrde 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 __YBLQE_ _IQIgL ___@gggE___ _ygLQE__ ______________ggIEgggY_____________ _23299__ _23199 ___________ ________ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _26199__ _26199 ___________ ________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 26.S~ 25."l6 _M__ M ___________ ________ 3.
INSTRUMENTS AND CONTROLS _22259__ _22199 ___________ ________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 99.0 199 g7__ ___________ Grade Totals Final g ggjpg All work done on this examination is my own.
I have neither given nor received aid.
< ___________________________________ Candidate's Signature . k \\
._ . - - _ _ _ _ _ - .-. --
. ' .. . NRC RULES AND GUIDELINES FDR 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 penalties.
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.
Use black ink or dark pencil gnly 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 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 three lines 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, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTIDN AND DD NDT LEAVE ANY ANSWER BLANK.
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 examination has been completed.
! ! . -. . . ,-.__ __., _ _ _ _ _ _ _ _ _ _. _ _ - _ _ _ _ _ _ _. _ _ _. _ _. _. _ _ _,. _..,, _ _ _. _
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. . 18. When you complete your examination, you shall: c.
Assemble your examination as follows: (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 examination 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.
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. 1 __P81NCIELEg_gE_NgCLE8B POWER _ PLANT _gPERATIgN PAGE
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IHEBMggYN@dlCgi_HE@l_lB9NSEEB_6Ng_E(U1g_ElgW ' W.O( ) DUESTION 1.01 (2.00) While an I/C technician calibrates the reactor high pressure switches, En error is made and the plant scrams on an erroneous high pressure signal at 0700 after 3000 hours at full power. Because the cause of the scram has been determined, and in order to maintain power commitments to the grid, the Operations Manager orders an immediate reactor startup.
What are your considerations regarding poisons in the core if the (1.0) a.
reactor is brought critical during the afternoon of the same day? . b.
Will control rod density change over the next 24 hours after the (1.0) startup? Briefly explain your answer.
W.02.s QUESTION 1.02 (2.00) a.
Most condensers are designed with excess condensing capability; (1.0) that is, the condensed liquid leaves the condenser hotwell several degrees below the saturation temperature. How would plant effici-ency be affected (increase, decrease, or not affected).if the temperature of the circulating water was greatly DECREASED? Explain your answer, b.
If the main condenser was absolutely air tight, would there be (1.0) any need for the air ejectors? Explain why.
QUESTION 1.03 (1.00) Four minutes following a reactor scram, the reactor power indication (1.0) is on 80 on Range 4 of the IRMs. What will power be one minute later? W.03 ) DUESTION 1.04 (1.50) PBAPS. Unit 3 is taken critical during startup, and a steady-state period is established. After the point of adding heat (POAH), the reactor period lengthens to infinity, and the reactor operator notes that the moderator temperature has changed from 255 deg F to 270 deg F.
a. What reactivity coefficients turned reactor power? LIST them in (1.0) order from the largest effect to the least effect.
! b.
How much positive reactivity was added to establish a stable (0.5) l positive period after criticality was obtained? t (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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1.
_ PRINCIPLES _DF NUCLEAR _ POWER _ PLANT OPERATIONt ~ THERDODYNADICS _ HEAT TRANSFER AND FLUID FLOW
(S.04 ) QUESTION 1.05 (2.00) a.
Indicate how available NPSH (net positive suction head) will (1.0) change (i e., increase, decrease, or no change) in the following instances.
1.
CRD pumps-system flow significantly decreased 2.
RWCU pumps-reactor power decreased from 100% to 20% 3.
Recirculation pumps-RPV level is decreased 4.
Recirc pumps-reactor pressure is increased l ' h.
Explain what happens within a pump when the "available" NPSH (1.0) drops below the value of the " required" NPSH. Give three (3) possible adverse effects on the pump.
, (T.09) QUESTION 1.06 (2.00) During calibration of the level switches that initiate the High Pressure Cool ant Injection (HPCI) system, an initiation signal is inadvertantly introduced at 100% power. For the following parameters, state the INITIAL change (if any) AND give the cause f or such a response.
Assume that the FWCS is in three element control.
a.
Total steam flow indicated (0.67) b.
Total Feedwater flow indicated (0.67) c.
APRM indication (0.67) ! l (9.CC)
QUESTION 1.07 (1.S0) a. During a reactor startup, what three indications / items are used (0.75) by the operator to determine when criticality is achieved? l ' b.
Fill in the blank: The FIRST rods in a new rod group have (0.75) ____ _ (higher, lower, or the same) rod worths than/as the LAST rods in that group. BRIEFLY EXPLAIN your answer.
! QUESTION 1.08 (1.00) l What is the purpose for using burnable poison in the core at PBAPS? (1.0)
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- t , INE8DgDyN@dlCS _UE@l_IB8NSEE8_8ND_E(glD_ELQM - t
DUESTION 1.09 (2.00) Explain what happens to the following parameters when power is lost to both Recirc MG set drive motor breakers. Briefly explain why.
a. Critical Power (CP) (0.67) b.
Critical Power Ratio (CPR) (0.67) c.
MCPR limi t (0.67) (5.07 ) QUESTION 1.10 (3.00) e.
Describe why spray cooling of the RPV is desirable when recovering (2.0) from the blowdown cooling made of core cooling.
I b.
Describe the core cooling mechanisms available to cool the core (1.0) when absolutely no ECCS systems are available for core cooling.
(7 0J) QUESTION 1.11 (2.00) Step LQ-2 of Procedure T-117, " Level / Power Control", states: If: Power i s above 3% or cannot be determined and torus temperature is above 110 deg F and an SRV is open or Drywell pressure i s above 2 psig then lower RPV level by terminating and preventing injection into the RPV except Baron injection and CRD until either: Power is below 3% or RPV level reaches -172" (TAF) or all SRVs remain shut and Drywell pressure is below 2 psig.
a. Explain how the reduction in RPV level reduces power.
(1.0)
b.
Assume that an ATWS has occurred at full power and with a 100% (1.0) rod pattern. What adverse effect could be expected if the RPV level was NOT lowered, and SLC injection was the only attempt made to lower power? DOESTION 1.12 (1.50) Procedure GP-2, " Normal Plant Startup", includes the following precaution: "Do not operate the turbine with the LP wheel bore temperatures less than 120 deg F when the turbine is running at speeds above 100 rpm."
a. EXPLAIN the basis for this precaution.
(1.0) b.
What action is equired if the LP turbine temperatures drop ( 0. 5 ) below 120 deg F? , l (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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c.
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IMEBdggyN851CS _NE81_IB8NSEEB_ONg_E(ylg_ElgN
(9.01) DUESTION 1.13 (2.50) The reactor is operating at full power when the RFP Master Controller (2.5) malfunctions, resulting in a total loss of feedwater. A reactor scram is expected to occur within seconds.
During,this short period, is reactor POWER level expected to INCREASE or DECREASE? Give TWO (2) reasons for your choice, including WHY each causes a change in reactor power.
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2 __P(@NT_DgglgN_INCLUplNg_g@FETY AND_EMERggNCY_@YgTEMg PAGE
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QUESTION 2.01 (3.00) Using attached Figure 1 " Main and Auxiliary Power Distribution", (3.0) cs a reference, answer the following question.
List the sequence of events that occur automatically during a Unit 2 fast transfer.
Limit your answer to equipment response within the Main and Auxiliary Power Distribution System.
QUESTION 2.02 (2.50) After two months of operation at full power, a scram occurs on PBAPS Unit 2 because of a personnel error. The Shutdown Cooling Mode of . RHR is unavailable.
a.
According to TRIP Procedure T-115, " Alternate Shutdown Cooling", (1.5) give the Alternate SDC flow path AND describe how core decay heat is ultimately removed.
b.
Step AK-14 of T-115 states: (1.0) If cooldown rate exceeds 100 deg F/hr, reduce core spray or RHR flow into the RPV until the cooldown rate is below 100 deg F/hr or RPV pressure decreases below 50 psig, whichever occurs first.
Why is the operator directed to stop the flow reduction if RPV pressure decreases below 50 psig? QUESTION 2.03 (3.00) e. State the equipment response within the HPCI system upon receipt (1.0) of an auto isolation signal.
b.
Describe how the HPCI isolation signal is reset. Include the (1.0) system permissives-required.
c.
Assume that HPCI has automatically initiated. Is it possible to (1.0) isolate the HPCI system with an auto initiation signal present? If so, EXPLAIN how this is accomplished.
. (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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PLANT _DE@I@N_INCLUDIN@_@AFETY_AND_EMER@ENCY SY@ TEM @ PAGE
- (C.061 QUESTION 2.04 (2.00) The following questions concern manual starting of the Emergency Diesel Generators.
s. Explain why a diesel generator should not be run unloaded.
(1.0) b.
Procedure S.B.4.A, " Manual Start of Diesels", states that, (1.0) "At least one minute must elapse between a diesel shutdown or trip and a diesel restart."
If an attempt WAS made to restart the diesel before the one minute time lapse, what would be the status of the diesel? QUESTION 2.05 (2.00) a.
State the source of control power to the 4.16 kV breakers.
(0.5) b.
For the following conditions, state whether the Emergency Diesel Generators WILL or WILL NOT supply the Emergency 4 kV bus.
1.
Bus voltage drops to 25% of normal; power is available to (O.5) the alternate feeder.
2.
Drywell pressure is >2 psig; and bus voltage drops to 65% (0.5) of normal.
c.
If a 4.16 kV bus is transferred to its alternate feeder due (0.5) to an undervoltage condition, will the bus automatically transfer back to its normal feeder if voltage is restored? ( 6. 0 5 ) ' QUESTION 2.06 (2.50) c. Briefly explain WHY it is important to closely monitor RWCU (1.0) system flowrate when operating in the blowdown mode.
b.
State the purpose of the RWCU Blowdown made and state when (0.5) this mode is used.
c.
State the automatic closure signal for RWCU CV-55 (the Clean Up (1.0) Drain Header Control Valve) while operating in the blowdown mode and explain its basis.
QUESTION 2.07 (2.00) List eight (B) of the Recirc Drive Motor Breaker start permissives.
(2.0) Include setpoints, if applicable.
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~ (6.0$) QUESTION 2.08 (2.50) Regarding the Standby Gas Treatment System: e. What are three (3) of the four conditions which will auto (1.5) initiate the system? Setpoints are required.
b. List the response of the SBGT fans and dampers upon receipt of (1.0) an auto 4:r!:tirr signal.
(Hlhh6FV1 , QUESTION 2.09 (2.50) R2garding the Residual Heat Removal (RHR) System while operating in the Shutdown Cooling (SDC) Mode: a.
State why i t is necessary to shutdown the recirc pump and close (1.0) its discharge valve in the RHR loop selected for SDC.
b.
State whether the following actions WILL or WILL NOT occur if (1.0) RPV prassure were to increase tc 90 psig while in the SDC Mode: 1.
Shutdown Cooling Suction valves auto close.
2.
All running HPSW pumps trip.
3.
RHR pump suction valve auto closes.
4.
Rx vessel head spray isolation valve auto closes c.
If RPV pressure were then to drop to 70 psig, would the RHR pumps (0.5) dedicated to Shutdown Cooling auto restart? QUESTION 2.10 (2.00) c.
It becomes necessary to increase the drive pressure to a certain (0.5) CRD HCU. As the RO assigned to this task, would you expect to OPEN or CLOSE the drive water pressure control valve? b.
Explain how your action in part "a" above would change the (1.5) following rates. (i e., increase, decrease, or no change) 1.
Scram valve charging flow , 2.
CRD total syutem flow 3.
Cooling flow (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) , - - - - - - - - - - - - - -, - _.., _ _ - _,. _ _ _ -,. -,
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, QUESTZON 2.11 (2,00) e. Describe the purpose of t.be RBCCW head tank.
gg,o3 b.
If power is lost to both TBCCW pumps, briefly describe the auto-(1.0) matic actions that occur AND list the loads to which cooling is provided.
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QUESTION 3.01 (1.50) e.
If the IRMs are indicating 20 on Range 4 and an operator (0.75) down ranged to Range 3, what trips if any, would occur? Why? b.
With the mode switch in STARTUP, and IRM "C" reading 11 on (0.75) Range 7, what trip (s) i f any, would occur if IRM "C" was down ranged to Range 6? Why? QUESTION 3.02 (2.50) List four (4) rod blocks associated with the IRM's.
Include setpoints cnd AUTOMATIC bypasses for each, as applicable.
QUESTION 3.03 (1.50) c. List the four (4) automatic actions that occur when HIGH radiation (1.0) is detected in the Control Room Ventilation Monitoring System.
b.
What is the equipment response when both radiation detectors reach (0.5) the HIGH-HIGH setpoint? QUESTION 3.04 (2.00) e. List five (5) conditions which will cause a Recirc MG set scoop (1.0) tube lock-up.
b.
What conditions must be met before the lock-up can be reset? (1.0) QUESTION 3.05 (3.00) c. Assuming valid initiation signals exist, what two (2) conditions (1.0) will close the ADS valves once blowdown has commenced? ASSUME NO OPERATOR ACTION AND THAT ADS REMAINS FULLY OPERABLE.
b.
During blowdown the operator depresses the ADS "A" timer reset (1.0) button. Describe the response of the ADS system.
c.
According to station operating procedures, ADS may be manually (1.0) reset (depressing the ADS reset buttons) only when what two (2) conditions are met? (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) . - - - - -.- -. c ,. - _ _ - - -. - =. - - -,,. - -. .. - - _ _ - --
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(6.04) QUESTION 3.06 (3.00) c. Assume that a complete loss of Drywell Chilled Water has occur-(0.5) red while at full power. What three (3) heat loads are affected? b.
While Unit 2 is at full power, the #2 13 kV bus is lost.
(1.5) Explain the automatic actions that should occur in order to maintain cooling to the Drywell.
c.
List the four (4) non-essential loads that are isolated from (1.0) RBCCW during a loss of power situation.
QUESTION 3.07 (3.00) e.
Unit 3 is operating at full power. APRM channels B and D have (1.5) failed high. An auxiliary plant operator in training shifts the RPS "A" power supply to its alternate power supply. What would be the result? (Confine your answer to RPS system response.)
b.
State the three protection functions (with setpoints and time (1.5) delays, if applicable) that trip the RPS power supply breakers.
((o. 6% QUESTION 3.08 (3.00) Answer the following questions using the attached EHC figure (Figure 2).
ASSUME NO OPERATOR ACTIONS.
c. Unit 2 is operating at 100% power and two bypass valves open.
(1.0) Describe how the en-aeae=tr --^* ellrd by D C will respond.
~{t1?$ifti Contral volW6 b.
Describe the effect both on the plant and within the EHC system (1.0)
if a Stator Cooling trip occurs at 100% reactor power.
c.
Explain what will happen to the EHC system and to the plant, (1.0) if the maximum combined flow limit is turned down to 50%
while the plant is operating at 100% power.
l ( 4.07 ) DUESTION 3.09 (1.50) c. List the conditions that must be met in order to open the (1.0) Containment Spray valves (MO-26,-31) after a LOCA.
b.
What is the purpose of the Containment Spray Override keylock (0.5) switch? l (8888* CATEGORY 03 CONTINUED ON NEXT PAGE
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- QUESTION 3.10 (2.00) Assume that the Feedwater Level Control System is operating in three element control using RPV water level detector "A".
Reactor power is et 1007., and all Feed pumps are in automatic.
For each of the failures given below, state HOW reactor water level will initially respond (ie., increase, decrease, or remains constant) and briefly explain WHY. Limit your explanation to the response of the Feed Wnter and FWCS components.
c.
Channel "A" RPV water level detector fails low (1.0) b.
Loss of RFP lube oil to the "A" pump servo motor (1.0) (C. t o ) QUESTION 3.11 (1.50) a. Explain HOW the Core Spray line break detection system would (0.73) sense AND indicate a break inside the core shroud.
b.-Assume that the Core Spray system has automatically initiated (0.75) on a 2 psig Drywell pressure signal. In anticipation of an increase in RPV pressure, an operator manually overrides and closes the Core Spray inboard injection valve (MO-12A(B)). Will this valve reopen if RPV pressure then drops below 450 psig? If NOT, explain the action (s) that the operator wculd have to take in order to open the valve.
(s.III QUESTION 3.12 (2.00) Figure 3 contains a section from a rod sequence instruction sheet.
(2.0) For the following situations, state ALL of the responses by the Rod Worth Minimizer (RWM). That is: withdraw error, withdraw block, insert error, insert block, select error, or a combination of these.
NOTE: CONSIDER "a" AND "b" SEPARATELY.
a. Refer to Figure J. Give the RWM response (s) if the rod marked "*" is pulled to position 16 during a plant startup.
b.
Refer to Figure J. Give the RWM response (s) if the rods marked "+" are inserted to position 20 during a plant shutdown. Rod 26-43 is then selected.
(***** END OF CATEGORY 03
- )
,
. - . 9 __PBgCEgygEg_;_Ng85@bt_@pNQ@@@L _Edg@@ENgy_@NQ PAGE
-
60919Lg@lC@L_CgNIBgL ' QUESTION 4.01 (2.50) a. Given that a specific component is marked with a deficiency tag, (1.5) respond to the following True/ False statements regarding that component. Justify any FALSE answers.
1.
A Maintenance Request Form (MRF) has yet to be issued f or this component.
2.
A " block" has not been issued for this component. That is, the component may still be operated.
b.
Explain how an operator can verify that a posted " Operator Aid" (1.0) is authorized and current.
( 7 04) QUESTION 4.02 (3.00) e. State the PBAPS whole body limits for radiation exposure received (0.5) per quarter, both with and without an NRC Form 4.
b.
State the three (3) entry requirements that one must meet prior (1.5) to entrance to a contaminated area, c.
Briefly describe the process f or the request and documentation of (1.0) an ALARA review for a specific job.
M42. ) QUESTION 4.03 (2.00) Answer the following True/ False questions in accordance with A-40, (2.0) " Working Hour Restrictions" and A-7, " Shift Operations". If FALSE.
EXPLAIN WHY.
c.
It is permissible for a Control Operator, who has just completed his normal 0700-1500 shift, to take the place of the 1500-2300 Control Operator if necessary.
b.
It is permissible for a Control Room Operator to work twelve (12) hours a day from Monday through Sunday, as long as he is off on the following Monday.
c.
The Control Room Supervisor may leave the Main Control Room without a complete turnover to a qualified replacement, as long as he plans to return within 30 minutes.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
- .-. . . --- - - . . . . _ - - .
. 4 __E89EEE98EE_!_d90EObt_OEU90DObt_ggggggggy_AND PAGE
- 80919LQgIC@L_CgNIBgL (102)
QUESTIDN 4.04 (2.00) ' c. According to ON-107, " Loss of CRD Regulating Function", if (1.0) both CRD pumps are off, and reactor pressure is below 550 psig, and three (3) or more CRD accumulator low pressure annunciators alarm, the operator is directed to scram and enter T-100.
Why is a reactor scram required in this condition? b. According to S.4.2.D, " Shutdown of the CRD Hydraulic System", (1.0) the CRDH system should remain in service when the reactor is shutdown. Explain why.
(?.03) QUESTION 4.05 (3.00) List the entry conditions with appropriate setpoints for: e. T-101 RPV Control (1.5) b.
T-102 Containment Control (1.5) , > !
( 7. 04 ) QUESTION 4.06 (3.00) a.
List the seven (7) immediate operator actions to be taken prior (2.0) to leaving the main control room should a control room evacuation be iacessary.
b. Once at the Remote Shutdown Panel, procedure SE-1 directs the (1.0) , ! operator to close RHR shutdown cooling valves MO-10-17 and [ MO-10-18. What is the basis behind this step? i !
QUESTION 4.07 (1.50) A sume that high radiation is detected in the Off Gas system.
(1.5)
According to ON-103, "Off Gas Stack High Radiation", what three (3) main control room indications should be checked for indications of ' j high radiation? l (***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
, . . -.,, .. - . -.. - - .. -..- --.. -.- -..-.---- --.,.- ---.-- ,-.- -..---.- -.. -. -.. - - - - -. - - - -
. . , .. 9:__P89gEDuggS_ _NgBd661_OBNgBd661_Edg8ggNgY_AND PAGE
- 88D196991g86_ggNIBQL QUESTION 4.08 (2.00)
Per the immediate actions of OT-113, " Loss of Stator Cooling", the operator is directed to verify that two (2) separate automatic actions have occurred. State these automatic actions AND explain the purpose behind each.
QUESTION 4.09 (3.00) PBAPS Procedure S.3.6.B, " Initiation of the Standby Liquid Control System", gives three (3) cases in which the licensed operator SHALL inject SLC if shift supervision is not available to grant permission.
a. STATE these three cases.
(2.0) b.
List the automatic at occur when an SLC pump is started from (1.0) the main control room.
MWF)v3 QUESTION 4.10 (1.50) a. What are the immediate operator actions as required by 0T-114, (1.0) "Inadvertant Opening of a Relief Valve"? b.
What signal initiates the " Safety Relief Valve Open" Alarm? (0.5) (***** END OF CATEGORY 04
- )
(************* END OF EXAMINATION ***************) .
. 19F QT.m 1 T ~ - utarv - . 1.
PRINCIPLES _OF NUCLEAR POgdER PLANT OPERATION PAGE
- T!gRMgpVNM ICg2_ HEAT TRMgFER_ M p_ FLUID _FLOgd
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
09.01) ANSWER 1.01 (2.00) a.
Xenon will peak at its highest reactivity approximately 10 hours (1.0) after the scram.
The worth of the rods will be relatively lower than what they would be worth on a xenon free startup.
Thus the reactor will go critical at a lower rod density than before.
b.
Because of the added positive reactivity due to Xenon decay (1.0) and burnup, the control rods will need to be inserted after startup in order to control power. Thus rod density will increase.
' REFERENCE LOT 1510, LO #3,6 pages 5-7 K/A Group II Reactor Theory, 292006, Fission Product Poisons K.1.07, Xe following a scram 3.2/3.2 K.1.14, Operator compensation for the change in Xe with time 3.1/3.2
(COL 1 i ANSWER 1.02 (2.00) 'pi" Plant ef f iciency would[ decrease} (0.25) because the heat rejected (1.0) a.
to the circulating water must be added to the feedwater by the reactor (0.75).
b.
Air ejectors would still be needed (0.25) because air in-leakage (1.0) is not the only sour ce of noncondensibles in the main condenser.
Other NC include radiolytic 02 and H2, and fission product gases.
I REFERENCE i a.
LOT 1250 Plant Efficiency, LO #2 (1.0) K/A Group II Thermo, 293004; K.1.1.2: Discuss subcooling 2.9/3.1 , l b.
LOT 500 Main Condenser Air Removal, LO #1 (1.0) K/A group I Components, 291006; K.1.18: Reasons for NC gas removal ! ! 2.8/2.9 . I
. li__E81NCIELEg_gE_NyCLE@B_EQWE8_E(@NI_QEEB@llgN PAGE
- t IHEBUQQYNOMICgt_HE@l_IBONSEEg_@NQ_ELylQ_ELQW - ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-LANGE, D.
ANSWER 1.03 (1.00) Period following a scram is -80 seconds.
(0.25) P = Po e**t/T (0.75) P = 80 e**(60/-80) P = 37.8 on Range 4 REFERENCE Rx Theory Handout, Sections 24-25/ LOT 1430 LO#4 (1.0) K/A Rx Theory Group II, K.1.08: Power, period egn, solve for changes 2.7 292003 (S.03 ) ANSWER 1.04 (1.50) a.
Mod temp coeff (1.0) Fuel temp coeff Void coeff b.
Delta Tmod = 15 deg F (0.5) Assume no real contribution from void or fuel coefficient.
-1 x 10**-4( k/k)/deg F x 15 deg F = 1.5 x 10-3 k/k added = 0.0015 k/k REFERENCE Rx Theory Handout, Sections 26-30 (1.5) LOT 1440, LO #3,5 K/A Rx Theory Group I, 292004; K.1.08 Compare relative effects of coeff 3.3/3.3 (Sd34 ) ANSWER 1.05 (2.00) a.1.
increases (0.25) 2. jdtreases dfCN" (0.25) 3.
decreases (0.25) 4.
increases (0.25) b.
Without adequate NPSH available, the pump would cavitate. The (1.0) formation of vapor bubbles and their subsequent collapse would cause noise, excessive pump vibration and pump internal wear.
REFERENCE PBAPS Fluids Handout, Section 4; LOT 1290 LO #4 K/A Components, 291004 Pumps-K.1.14 Relation between flow and suction head 2.5/2.5 (1.0)
l'g__PRINCIPLEg_gF_NyCLEAR_PgWER_ PLANT _gPgRATIgN PAGE
- i IHERMODYNAMICg1_ HEAT _TRANgFER_AND_FLyID_FLgW
- ANSWERS -- PEACH BOTTOM 2&3 ~ ~ ' " ' '^o i nNr#. D.
? A.-6gagi K.1.06 Need for NPSH 3.3/3.3 (1.0) M ObtQ LS or) ANSWER 1.06 (2.00) j a.
Total indicated steam flow would remain approximately constant (0.17), because of the small size of the HPCI steam supply line in comparison to the 4 Main Steam Lines (0.5).
b.
Total Feedwater flow would decrease (0.17) because of the level increase caused by HPCI (0.5).
c.
Power would increase (0.17) because of the increased inlet subcooling (0.5).
REFERENCE LOT 1610 Temp Transients (2.0) K/A Abnormal Evolution Group II 295008, High Reactor Water Level AK 2.05 HPCI Interrelation 3.8/3.9 (9.06/ ANSWER 1.07 (1.50) a.
1.
positive stable period (0.25 each) 2.
constantly increasing count rate 3.
no rod motion b.
The first rods in a group have HIGHER rod worths than those (0.75) later in the group (0.25) because the local flux surrounding them was increased by the withdrawal of adjacent rods in the previous groups (0.5).
REFERENCE LOT 1530, PAGE 10 K/A Group I Rx Theory, 292008 Rx Op Physics a.
LO #3, K.1.08 (4.1) Power, period changes close to crit.
(0.75) b.
LO #1.B, K.1.03 (4.1) Explain op charac close to crit.
(0.75) ANSWER 1.08 (1.00) PBAPS uses burnable poisons because they allow adding more fuel to (1.0) account for fuel depletion over $!: ore life without adding more control rods.
. l l , .. -... , ,
}' 1.__PRINCIPLEg_QF NUCLEAR _PQWER_ PLANT _QPERATIgN PAGE
- t THERMQDYNAMICS _ HEAT TRANQFER_ANQ_FLUIQ_FLQW , t ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-LANGE, D.
REFERENCE
LOT 20, LO #6, pages 4,5 (1.0) K/A Group II Theory, 292007 Burnable Poisons K.1.01 2.9/3.1~ ' ANSWER 1.09 (2.00) ! a.
CP decreases (0.17). Channel quality would be closer to the critical quality because the coolant spends more time in the channel and is able to pick up more energy (0.5).
b.
CPR increases (0.17). Even though the CP decreases, the actual power decreases more, so CPR decreases (0.5).
c.
MCPR limit incroases (0.17) because a RFC failure is more severe at low flows (0.5).
REFERENCE LOT 1370, LO #1,2 pages 3-5/ LOT 1380 / LOT 1640, LO #1 K/,A Group I Thermodynamics, 293009, Core Thermal Limits K.1.23 Changes in Core Flow-Effect on CP 2.8/3.2 (C.07/ ANSWER 1.10 (3.00) a.
Spray cooling is preferred to recover water level following (2.0) blowdown cooling because the core temperature is elevated during steam cooling to provide the delta T so steam can carry away core heat. Core Spray spargers can safely lower core temperatures with a reduced possibility of cprq cladding Q TQ ' p ud. -b u s c' w h 6U--s ' amage.
, b.
1.
Alternate injection systems / flow paths are used as directed (O.5) by the level restoration procedure. (T-111)
' 2.
The core can be steam cooled when RPV level reaches the core (0.5) midplane. (T-113 Blowdown cooling) REFERENCE a.
LOT 1560, LO #12 (2.0) K/A Group I Emer Evol;295031 Lo Lvl, EK3.03 Spray Cooling 4.0/4.3 b.
LOT 1560, LO #11 (1.0) K/A Group I Emer Evol;295031, EA2.04 Adequate Core Cooling 4.6/4.8 , ! ! ,
. 1:__ERINglPLEg_gF NUGLEAR_PQWER_ PLANT _gPERATigN PAGE
' t IBEggggyN@diggt_ME@T TBQNSEEB_QNQ_ELg1Q_ElgW - ANSWERS -- PEACH BOTTOM 2L3-86/12/OB-LANGE, D.
(6 01) ANSWER 1.11 (2.00) a.
Power is reduced by decreasing RPV level because natural circ-(1.0) ulation is inhibited. As level is decreased, the actual flow area across the top of the shroud and through the moisture separators is decreased causing a restriction in core flow.
b.
If water level is not reduced in order to control power, it is (1.0) likely that primary containment will fail because the time required for SLC to shutdown the reactor exceeds the time that PC can absorb the energy that is beis.g produced.
( 4 4@ Eb, k
REFERENCE T-117 Bases, LOT 1560, LO #1,6 a.
K/A Emer Evol Group I, 295037 ATWS, EK.1.02 4.1/4.4 (1.0) RPV water level effects on power b.
K/A Emer Evol Group I, 295037 ATWS, EK.1.03 4.2/4.4 (1.0) Baron effects on power ANSWER 1.12 (1.50) a.
Sufficient prewarming is necessary to improve the LP wheel's (1.0) f raci tare toughness, thus reducing the chances of wheel burst.
The metal is much more brittle at lower temperatures.
b.
Trip the turbine.
(0.5) REFERENCE LOT 560, page 50/ LOT 1070, LO #2,3 (0.5) K/A Group II Systems, 245000 Mn T/G Aux (1.0) K.5.02 Turbine operation, limitations 2.8/3.1 ..
_ _ - . 1 __E81NglELEg_QE_NQCLE98_EgWEB_ELONI_gEEB911gN PAGE
i IUEB599yN951GS2_UE91_lBONSEEB_9NQ_ELQlg_ELgW - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
LS.04) ANSWER 1.13 (2.50) DECREASE (0,5) REASONS: 1.
Immediately, the loss of feedwater flow causes a (1.0) h ~1n-moderatur: _Z hi which introduces negative reactivity into the core.
2.
When total feedwater flow drops below 20%, the Recirc (1.0) pumps auto runback to 30% (or if RPV level drops to (17" and an individual feed pump flow drops to < 20%, the Recirc pumps auto runback to 60%). The reduction in core flow causes an Lacr.eese in weiding which adds negative reactivity to the core.
REFERENCE , l LOT 1630, 1640 LO #3 l LOT 40 LO #4 K/A System 259001 FW, K3.12 KN of effect on Rx power if FW malf 3.8/3.9 , _. _ _ - _ - -
.-. _ _ __ _ . 2___ PL @NI_ Dggl@N_ I NCLyDINg_@9ggly_9N9_g DEB @gNCy_ @y@lgdy PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
DK-ANSWER 2.01 (3.00) it l hk b 2 b 8N \\ 500 kV reaker t 1p (l4(irt iVLp ) (3.0) Generato field exciter breaker trip h/22 ctord Onh KIMN Aux bus be ker trips-1 and 2 g g Turbine trip via EHC Fast transf'er elay energized . Energize Aux re ys 386 X1, X2, X3 4 M2l bOV1 N j Annunc.iator Mq M y-e 4 (cd h Data log arL W VM SIKk- , / REFERENCE LOT 640, LO #6, Auto actions during fast transfer (1.5) K/A Group II Systems, 262001, AC Elec K.6.03 KN of Gen trip on AC Dist, 3.5./3.7 i K/A PW Gen, A.1.07, use of elec drawings, 3.0/3.7 (1.5) . ' ANSWER 2.02 (2.50) a.
The reactor is flooded up to the main steam lines. The water is (1.5) then allowed to flow down to the relief valves to the torus.
The heat energy will be removed from the torus via torus cooling.
b.
If RPV pressure were to drop below 50 psig, the SRVs may close, (1.0) and the cooling flow path will be lost.
REFERENCE T-115 Bases, LOT 1560, LO #13 (1.5) K/A Group III Abn Evol, 295021, Loss of SDC A.K.3.05 Alternate flow path 3.6/3.8 K/A Group II Plant systems, 205000, RHR (SDC) (1.0) K.3.01 Loss of Rx pressure 3.3/3.3 , k l . - _ - _ _ _- _ ._ _- .-. . .- - _ _ _ -_ _.-
. 2.__ PLANI _pggl@N_1NCLyplNG__ SAFETY _ANp_EMERggNCY SYSTgMg PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 2.03 (3.00) a.
HPCI turbine trip (1.0) l Steam supply valves close (MO-15, 16) Tr(bytt bO' i RT ' Suppression pool suction isolation valves close V2 o,ty y AD-137 and 138 (exhaust line drain valves) close M 2 CL k) b. All isolations, except the 75N (100?) steam supply isolation must (1.0) be manually reset with the TWO isolation logic pushbuttons.
Operator must ensure that the control switches for the AO valves which closed upon isolation are in the closed poultion.
137,138 Exhaust line drains 4807 Steam line warmup bypass (Unit 2 only) c.
The only way to isolate HPCI with an auto initiation signal (1.0) , present is to press the manual isolation PB on control room panel C04B.
REFERENCE LOT 340, HPCI LO #5,6 K/A Group I Plant Systems, 206000 HPCI A.3.09 Response to isolation 4.2/4.1 (1.0) K.4.02 KN of system isolation interlocks 3.9/4.0 (1.0) K/A Components, 291003 Controllers K.1.03 Operation of a valve (1.0) controller, including a seal-in feature 3.3/3.4 (6.06) ANSWER 2.04 (2.00) a.
Operating an unloaded diesel increases the air blower temperatures (1.0) to the maximum operating value, and may result in blower damage.
b.
The E D/G will turn over, but the fuel racks will not open. A (1.0) failure to start trip will result.
REFERENCE PBAPS System Procedure S.B.4.A K/A Group I Systems, 264000, E D/G l K.1.06 KN of E DG starting system 3.2/3.2 (1.0) A.2.03 AB to predict impact if E DG is run unloaded 3.4/3.4 (1.0) i , t i . - - _ - .. _ _ _ _ _ _ _ - - - _ _. _ _. _ _ - -- - - ~, _. _. -,. -. -,., - _ _ -. _. .
. 2.
PL@NI_DE@l@N_lNCLUDIN@_@@EETY QND_EgEB@ENCy_@y@ led @ PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 2.05 (2.00) a.
125 Vdc station batteries (0.5) b.
1.
WILL NOT (0.5) 2.
WILL NOT (0.5) c.
NO (0.5) REFERENCE ' LOT 660, LO #4,5 K,A Groug II Plamt Systems, 262001 AC Elec Dst K.1.01 Relation w/E DG 3.8/4.3 (0.5) K.1.02 Relation w/DC dist 3.3/3.6 (0.5) K.4.03 Intlk w/ auto bus transfer 3.1/3.4 (1.0) (G.oS } ANSWER 2.06 (2.50) \\1 o e a.
Because the regenerative HX are bypassed, blowdown flow must be (1.0) limited to the capacity of the NRHX in order to prevent over-heating of the demineralizer bed.
b.
Blowdown is used during startup or hot standby operations to (0.5) reduce reactor water inventory.
Also @ Mp A M c.
In order to prevent draining of the system, CV-55 closes on low (1.0) pressure sensed upstream (5 psig). This avoids, in particular, an isolation condition while dumping to the main condenser.
Y Y Y REFERENCE LOT 110, LO N6,7 K/A Components, 291007, Demin K.1.06 Reason for temp limits 2.7/2.7 (0.5) Group II Systems, 204000 RWCU K.1.09 KN of relation b/t RWCU and RPV level 3.2/3.3 (2.0) __ -. __. _ _ _. _
. 2:__ PLANT DESIGN _ INCLUDING _gAFETY AND_ EMERGENCY _ SYSTEMS PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 2.07 (2.00) Discharge Valve shut (8 required, 2.0) -Suction Valve open Drive Motor Breaker Fully Inserted-Generator field breaker open-Lube Oil Pressure >20 psig One MG set cooling fan running Generator lockout relays reset HMG set lockout relays reset-MG set supply air isolationgates fully withdrawn REFERENCE LOT 30, LO #5 K/A Group II systems, 202001 Recirc System K.4.10 Pump start permissives 3.3/3.4 (fa.01) ANSWER 2.08 (2.50) a.1.
reactor water level 0" (any 3 at 0.5 each) 2.
D/W press 2 psig 3.
Rx Bldg exhaust 16 mr/hr 4.
refuel floor exhaust 16 mr/hr b.
The auto start signal from Unit 2 will start the "A" fan and (1.0) t open its inlet and outlet dampers. The inlet and outlet dampers to both the "A" and "B" filter trains will open. (If fan "A" fails to start, "B" will start.)
REFERENCE LOT 210, LO # 2 K/A Group I 261000 SBGT A.2.05 Predict impact of fan trip 3.0/3.1 (1.0) K.4.01 KN of auto initiation 3.7/3.8 (2.0) _ _ _. _ _
.
PLANT _ DESIGN INCLUDING __ SAFETY AND_ EMERGENCY _@Y@TEMS PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 2.09 (2.50) a.
To prevent pump bearing damage due to reverse rotation.
(1.0) b.
1.
WILL (0.25 each) 2.
WILL NOT 3.
WILL NOT 4.
WILL c.
NO (0.5) REFERENCE LOT 370 RHR, LO #6 K/A Group II Systems 205000 SDC (1.5) K.4.02 KN of Hi press isol 3.7/3.8 Components 291004 Pumps, K.1.13 Centrifugal pump oper 2.6/2.7 (1.0) ANSWER 2.10 (2.00) a.
CLOSE (0.5) b.
1.
No change (0.5) 2.
No change (0.5) 3.
";;
- ....y c dec,ust (0.5)
REFERENCE LOT 70, LO #2 K/A Group I 201001 CRDH A.1.01 AB to predict changes in CRD drive water header pressure 3.1/2.9 ANSWER 2.11 (2.00) l a.
It provides adequate NPSH to the RBCCW pumps and provides a surge (1.0) volume.
b.
After a 40 second time delay, the RBCCW system back-up to TBCCW (1.0) system air operated valves will open to provide cooling to the CRD pump lube oil coolers and air compressors.
REFERENCE LOT 460, LO #3,4 K/A E/A Ab Evol (Group II) 295018 Loss of Component Cooling Water A.
K.2.01 System loads 3.3/3.4 (0.5) K.3.07 Cross connect w/ backup systems 3.1/3.2 (0.5) , -. -,---- -- . -. ,,, _. - .__
_._, ... _ _. -. _ _ .. - _._- - _- -. . _._ _. _ -... _..... - _. _ _ _ _.... _. -
- -.-
2o N @ T DEgl W_INCLL(DING _ SAFETY AND_ EMERGENCY SYSTEMS PAGE
' ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
, .
b l
,
, .
! t ' ) . t I , ) Y , I I !
e' f t a
t-e i f i i I t i
h +
..., -, - - -,,, -.,,, -,.,,.,, _. - _ _, - -., _ .--,_,y _ ,, _-_,m., _-,m__---,___., -.., _ _.,, - - +
. 3.
INSTRUMENTS AND CONTROLS PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 3.01 (1.50) a. None (0.25). The IRMs would indicate 20 on Range 3;bO-40 scale](0.5). b.
An IRM upscale trip (rod block) would occur (0.25), because the IRM would read 110[on the 0-125 scale] (0.5). REFERENCE LOT 250, LO #3,5,7 K/A Group I Components, 291002 Sensors, detectors (1.5) K.1.20 Neutron monitoring units 3.2/3.3 ANSWER 3.02 (2.50) BLOCK SETPOINT BYPASS )
__
Upscale 108/125 Mode switch in RUN Downscale 2.57. scale Downscale on range 1 (and MS in RUN) Inop N/A N/A Manual only (and MS in RUN) Detector not full in N/A Mode switch in RUN (4 Blocks at 0.25 each, 2 Setpoints at 0.25 each, 3 Bypasses at 0.33 each) REFERENCE LOT 250, RBM, LO #4 K/A Group I Systems, 215003 IRMs, K.4.01 Rod Blocks associated with IRMs 3.7/3.7 ANSWER 3.03 (1.50) a.
1.
Fresh air supply, A/C supply, and return air fans trip.
(1.0) 2.
Dampers to the high efficiency filter open 3.
Selected emergency supply fan starts 4. g 2 isolation damper shuts g g g,g,g Gmtitt roCw iIin LOT . fans trip and associated dampers shut.
(0.5) b.
All
REFERENCE LOT 450, LO #2, PAGE B K/A Group II Systems, 290003, Cntrl Rm HVAC (1.5) K.1.01 Interrelation w/ Rad Mon 3.4/3.5 . - - .. -.. - - - -. - - -, - -. - -_ .. _ -, -. - - - -.. _ _. -., _ - _ .. - - -. - - . - .. -
8-IN@lByNENI@_AND_CONIBQL@ PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 3.04 (2.00) a.
Signal power failure (0.2 each) Hi oil temp, >210 deg F Low oil pressure, <30 psig for 15 sec 13 KV power unavailable Loss of speed signal Loss of DC power to scoop tube lack relay b.
Initiating signal cleared (i e., lockout reset) (1.0) (Balance actual generato speed to speed demand),(then press reset) REFERENCE LOT 40, LO #6,7 pages 10-12 K/A Group I Systems, 202002 Recirc FC (2.0) K: SG #12, Verify alarm setpoints, take actions in DNs 3.5/3.3 ANSWER 3.05 (3.00) e SRO a.
Rx pressure <50 psi (0.5) Trip of the RHR nd CS pumps (0.5) opL b.
Blowdown will continue. (If both the "A" and the "B" were reset, (1.0) blowdown would be interrupted.)
c.
Level > -130" (0.5) Makeup to vessel available hk
- * *
(0.5) REFERENCE LOT 330, LO #1,4, pages 5-7 K/A Group I Systems, 218000 ADS (3.0) K.4.03 KN of ADS Logic control 3.8/3.9 i ! l t l l l
- - , - _., _ - _ _ - -.. - -. _ _. _. _. _ _ _ _ _. _ _ -. _ _ _. _.,. __ _ _, _.
. 3.
INSTRUMENT @_AND_CgNTRgl@ PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
( G.04) ANSWER 3.06 (3.00) get-9tT6 a.
DW Area cooling units, Recire pump motor coolers, DWEDS cooler (0.5) b.
The motor operated transfer valves and the air operated isolation (1.5) valves will open in order to isolate DW chilled water and line up RBCCW to both the "A" and "B" headers. The non-essential loads would be isolated by an air operated valve within the RBCCW system.
c.
Non-regenerative heat exchangers (1.0) RWCU pump coolers Instrument Nitrogen coolers Sample coolers REFERENCE LOT 150, LO #2,3,4,5,6 K/A Group I Systems, 223001, Prim Cont and Aux K.6.01 KN of DW Cooling 3.6/3.8 (2.5) K.2.10 KN of Pwr supplies to DW Cooling Units 2.7/2.9 (0,5) ANSWER 3.07 (3.00) a.
APRM "B" and "D" would insert a half scram on RPS Channel "B".
(0.5) Because the RPS power switch cannot be switched to its alternate (0.5) feed without a half scram in RPS "A" occurring, a full scram would occur.
(0.5) b.
Over voltage 131 (+2v) (0.5) Underfrequency
(+.2Hz )(6 sec TD - Cab " M6 NU) (0.5) Undervoltage 113 (+2v) (4 sec TD ong,, ((At qqid ) (0.5) REFERENCE LOT 300, LO #4,8,6 /page 8, K/A Group I Systems 215005 APRM effect on RPS A.1.02 3.9/4.0 (1.5) 212000 RPS affected by RPS supply power malf A.2.02 3.7/3.9 (1.5) t
. 3.
_INSTRUMENTg_AND_CONTRgLS PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-LANGE, D.
(6.057 ANSWER 3.08 (3.00) Two BPVs cause about 6% flow.kReactor pressure will decrease (1.0) a.
theexcesssteamflow)Loweryeactorpressurewill because of cause the CVs to close to approximately(94% flow.)
% - b.
Stator Cooling trip will cause an EHC load set runback to about (1.0) 23% (of Gen output). the Recirc pumps will trip (A: 1s TD, B: 10s) (and power will decrease to about 55%) As the load set decreases, the M reasing reactor pressure signal will cause the BPV to go full open.(The CVs will be at about 23% (of Generator Output amps)) c.
Max Comb Flow at 50% limits total CV and BPV opening to 50%. (1.0) With power at 100%, as CVs close, reactor pressure increases, voids collapse and a high flux (or high pressure) scram will occur.
REFERENCE LOT 590, LO #4 K/A Group I Systems, 241000 Rx/ Turbine Press Reg K.6.10 BPV failure effect on EHC 3.6/3.7 (1.0) K.6.16 Stator water cooling failure 2.9/3.1 (1.0) A.1.15 Max Comb Flow Limit-predict changes 3.1/3.1 (1.0) bb OT) ANSWER 3.09 (1.50) a.
DW pressure greater than 1 psig (1.0) 2/3 core coverage LOCA signal present jgeh b. (sport M hrnit, hermal) Permissive switch in manual j b.
It allows opening of containment spray valves by bypassing the (0.5) requirement for 2/3 core coverage anu a LOCA signal.
REFERENCE i LOT 370, LO #3 l K/A Group II Systems, 226001 Containment Spray (1.5) K.1.08 KN of interrelation w/ RPV instr 3.2/3.4 K.1.13 KN of interrelation w/ DW instr 3.1/3.2
, . ~. _ _ _ -__,- - __.
. _.
-.. _ _,. _. _ , _. -. _
. 3:__INglgggENJg_9Np_CpNTRpl@ PAGE
, ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 3.10 (2.00) a.
Level increases (0.25) due to level mismatch requesting more water. The feed pumps will increase speed in order to provide it. (0.75) b.
No change in level (0.25). The servo will lock up the FW turbine control valve as is (0.75).
L gg,g g _ n REFERENCE 2. (.ctre( (Rh ed' WM 5O, LOT 550, LO #14 K/A Group I Systems, 259002 RPV Level Control KN of effects of malf on FWCS K.6.05 RPV level input 3.5/3.5 K.l.05 FW system 3.6/3.7 (6.l0 ) ~ ANSWER 3.11 (1.50) symyr ytt f 5M , a.
The low leg of the leak' detection system would now read the (0.75) outer annulus pressurs. The high leg would still read the top of the core plate pressure. The delta P between these would then _m ' ,.d d. The control room delta P annunciator alarms at 15 psid.
igpd % m pu d V'MY ' I t t S p$id b.
NO (0.25), it will not reopen unless the outboard injection (0.75) valve is closed first (0.5).
REFERENCE LOT 350, LO #B, pages 7-9 K/A Group I Systems, 209000 Low Pres Core Spray K.4.04 KN of line break detection 3.0/3.2 (0.75) A.4.03 AB to manipulate / monitor CS injec v1vs 3.7/3.6 (0.75) ANSWER 3 12 ) .,s s a.
WD error, select error, WD block +9ee9-estM) s , _ t______ m.
.n a, ....% .. _ _. m i. u_ m_ n..,._m.
m__.,. __i____ _____ __ ____ ____m u w. mm. , __ __ _. _.
REFERENCE % (3/p/34 LOT 90- LO #2 K/A Group II Systems, 201006 RWM A.2.04 AB to predict effect on RWM for out of seg rod 3.1/3.5 (2.0)
. __ PROCEDURES _ _NgRMAL _ABNgRMAL _ EMERGENCY ANg PAGE
4_,
1 R@gigLggIC@L_CgNTROL . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 4.01 (2.50) a.1 FALSE (0.25), a deficiency tag indicates that a MRF HAS been (1.0) initiated (0.75).
a.2 TRUE (0.5) b.
Utilize the information in the Operator Aid Log Book.
(1.0) REFERENCE LOT 1570, LO #39, 3i (2.5) A-26 " Corrective Maintenance", A-95 " Operator Aids" K/A PW Generics, K.1.02 Kn of Tagging Procedures 3.9 (7.01) ANSWER 4.02 (3.00) a.
w/ 2500 mrem /qtr, w/o 1000 mrem /qtr (0.5) b.
All dosimetry devices (0.5 each) Signed on RWP (if one is required) Anti-C clothing as specified by sign c. Fill out a Review Request Formfor stamp to MRF or RWP request f orm) ( 1. 0) An ALARA group member will review the request.
(Ilhe ALARA group findings will be returned to the originator.)
(Request and evaluation will be stored in history files.)
REFERENCE LOT 1730, LO #3-Exposure Limits; LO #1-Entry req.
(2.0) K/A PW Gen, K.1.03 KN of 10 CFR 20 and facility radcon 3.3/3.0 LOT 1770, LO #2 (1.0) K/A PW Gen, K.1.04 KN of ALARA pgm 3.3/3.6 (F.02) ANSWER 4.03 (2.00) a.
TRUE (0.5) b.
FALSE (0.25), no individual may be allowed to work more than 72 (1.0) hours within any 7 day period (0.75).
c.
TRUE (as long as an SLO remains in the MCR) (0.5) REFERENCE A-7 " Shift Operations", A-40 " Working Hour Restrictions" (2.0) LOT 1570, LO #2b, 3h
. -,, -,. - - - --, -- -.,.., - - - , ,, ,,, - -,. - - , --- - - - - -
. PROCEDURES - NORMAL _ABNQRMAL _ EMERGENCY _AND PAGE
4.
t t 8091gLQQlCOL_CgN189L - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
K/A PW GEN, A.1.03 Use procedures related to shift activities and staffing. 2.7/3.7 (7.02 / ANSWER 4.04 (2.00) a.
With reactor pressure below 550 psig, ther accumulators are (1.0) required to assure adequate scram speeds. The scram is inserted before all accumulators depressurize.
b.
CRDH provides a continuous flow of water to the CRD units to (1.0) prevent crud buildup.
M isr coo ((rt) REFERENCE LOT 70, CRD Hydraulics, LO #2, OT 107 Bases (1.0) K/A Group II Abn Evol, 295002, Loss of CRD pumps A.K.1.01 KN of Rx pressure vs. rod insertion capability 3.3/3.4 A.K.3.01 KN of reason for Rx scram due to CRD pump loss 3.7/3.9 S.4.2.D, "SD of CRDH" (1.0) K/A Group I Systems, 201002 CRDH K.3.03 Loss ad CRDH effect on CRD Mechanism 3.1/3.2 CI03) ANSWER 4.05 (3.00) a.
RPV level below -48" or unknown (1.5) DW Pressure >2 psig Group I Isolation Scram condition and Rx power > 3% or unknown b.
Torus temperature > 95 deg F (1.5) Torus level outside of 14.6' to 14.9' DW pressure > 2 psig DW Temp > 145 deg F ! REFERENCE LOT 1560, LO #9, pages 6-7 K/A Group I E/A Evol, 295010, High DW Press (3.0) SG #11, Recognize entry conditions 4.2/4.5 __ - . _ _, - -. _ _. _. -. - - - - - - _. - _ _. - - - - _ _ _ _ . - - . ... _ _. _ __ _ _
. h___PRQQEDyggS - NQRM9L _@@NQRM@L _gMERQENQY_@ND PAGE
t t 80D19L991G06_G98189L . ANSWERS -- PEACH t#3TTOM 2&3-86/12/08-LANGE, D.
Cl.04 ) ANSWER 4.06 (3.00) a.
Runback Recirc flow to minimum (0.29 each) Transfer house loads Manually scram and execute procedure T-100 concurrently Place DW instrument air in service Close the eight MSIVs Establish torus cooling Obtain the Emergency SD Panel key b.
To prevent a potential LOCA due to high reactor pressure in the (1.0) low pressure shutdown cooling line.
REFERENCE SE-1 and SE-1 Dases K/A Group II E/A Evol 295016 CR Abandonment SG #10 Perform immed actions w/o refer 3.8/3.6 (2.0) (A)A2.03 Interpret Rx press as applied to CR Aban 4.3/4.4 (1.0)
ANSWER 4.07 (1.50) Air ejector radiation monitors (0.5 eal Main Steam Line radiation monitors 34t(4( Areas from which the SBGT is taking suction also - K2, - N m d id flpe r N d M d REFERENCE ON-103 K/A Group II E/A Evol, 295017 High Offsite Release Rate (1.5) (A)K 2.03 KN of interrelation w/OG 3.3/3.5 , _ _, _ _ _.. _ _ _.. _ _ _ , _.. - _ _ _. , _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _
E
~ 4- - PRQQEDQRES_-_NQRMAL _ABNQRMAl _EMERGENgY_ANQ PAGE
t t 8891gLgGig@L_ggNIBQL - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 4.08 (2.00) a.
Verify that both recirc pumps trip (0.25). If a turbine runback occurs without a reduction in reactor power, high reactor pressure will result in a scram. To avoid this, the loss of stator coolant logic will trip both recirc pumps to reduce power. (0.75) b.
Verify turbine generator runback (0.25). This occurs to reduce the heat produced by the stator windings to an acceptable level.
If this does not occur, the turbine will trip. To avoid the turbine trip (and the scram) the turbine generator runs back. (0.75) REFERENCE OT-113 Bases K/A Group II Plant systems, 245000 Main T/G Aux K.3.03 KN of effect of Mn T/G Aux malf on Rx press 3.9/4.0 (1.0) K.3.04 KN of effect of Mn T/G Aux malf on Rx power 3.9/4.0 (1.0) ANSWER 4.09 (3.00) a.
When the Control Rod System is incapable of shutting down the reactor, SLC shall be injected immediately ift 1.
reactor power starts to increase as indicated by both nuclear (0.67) instrumentation and steam production, or 2.
if five or more adjacent control rods or 30 or more control rods (0.67) cannot be inserted past 06 and reactor level cannot be maintained, ! 3.
or if five or more adjacent control rods or 30 or more control (0.67) rods cannot be inserted past 06 and suppression pool temperature reaches 110 deg F.
r l b.
Pump start (0.33) RWCU isolation (0.33) Squibs fire (0.33) l REFERENCE i S.3.6 SLC LOT 310, LO #6,12 l K/A Emerg Evol, 295037, ATWS (2.0) l A.1.04, AB to operate SLC 4.5/4.5 i K/A Group I Systems 211000 SLC (1.0) K.4.07 KN of RWCU interlock 3.8/3.9 K.4.OB KN of system initiation upon op of entrl switch 4.2/4.2 l l l ,
. b__PBgggggggg_;_Nggd8g1_9pNggd@L_gdgBggNgy_ANg PAGE
a B89196991996_G9 BIB 96 - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-LANGE, D.
ANSWER 4.10 (1.50) a.
1.
Place both loops of Torus Cooling in service.
(0.5) 2.
If Torus temp exceeds 95 deg F, enter procedure T-102 and (0.5) execute it concurrently.
b.
A signal is sent from the acoustic monitoring system when flow ( 0. U ) is sensed in the SRV ta11 pipe.
REFERENCE OT-114, OT-114 Bases .
' . ' TEST CROSS REFERENCE PAGE
. DOESTION VALUE REFERENCE
01.01 2.00 BANOOOO401 01.02 2.00 BANOOOO460 01.03 1.00 BANOOOO550 01.04 1.50 BANOOOO551 01.05 2.00 BANOOOO554 01.06 2.00 BANOOOO563 01.07 1.50 BANOOOO576 01.08 1.00 BANOOOO577 01.09 2.00 BANOOOO579 01.10 3.00 BANOOOO581 01.11 2.00 BANOOOO589 01.12 1.50 BANOOOO594 01.13 2.50 BANOOOO618
24.00 02.01 3.00 BANOOOO570 O2.02 2.50 BANOOOO590 O2.03 3.00 UANOOOO602 O2.04 2.00 BANOOOO604 02.05 2.00 UANOOOO606 O2.06 2.50 DANOOOO608 02.07 2.00 BANOOOO609 O2.08 2.50 BANOOOO611 02.09 2.50 BANOOOO612 02.10 2.00 DANOOOO614 02.11 2.00 DANOOOO615
26.00 03.01 1.50 BANOOOO317 03.02 2.50 BANOOOO571 03.03 1.50 BANOOOO591 03.04 2.00 BANOOOO592 03.05 3.00 BANOOOO595 03.06 3.00 BANOOOO596 03.07 3.00 BANOOOO601 03.08 3.00 BANOOOO603 03.09 1.50 BANOOOO607 03.10 2.00 BANOOOO613 03.11 1.50 BANOOOO616 03.12 2.00 BANOOOO617
26.50 04.01 2.50 BANOOOO503 04.02 3.00 BANOOOO504 04.03 2.00 BANOOOO500 04.04 2.00 BANOOOO590 04.05 3.00 BANOOOO593
I
.,_ _.___ ._ _ ~ - _ _ _ __
, - . TEST CROSS REFERENCE PAGE
. DUESTIDN VALUE REFERENCE ________ ______ __________ 04.06 3.00 BANOOOO598 04.07 1.50 BANOOOO599 04.08 2.00 BANOOOO600 04.09 3.00 BANOOOO605 04.10 1.50 BANOOOO638 ______ 23.50 ______ =__ 100.00 ,
if{.}}Q [c GL E NATIW15 = c = es s o V,t o %st8 $- E = aca .. (y. y,)fg f E=%ma A=M PE = agh A = mR8 V = Ve + at A = Age"AI f W = v AP A = In 2/tg = 0.693/tg AE = 931 As [(t ) (t )3 ' g b
- '" " ut > + g n
- a. ge, aT , g I = Ice "IX & = UA aT I = Io10NL &=iah TYL = 1.3/p $uR(t) HVL = -0.693/p P = Po10 SCR = S/(1 - K,g) t/t P = Poe CRx = S/(1 - K,g X) SA = 26.06/t CRs (1-K,g,) = CR (1-K,g )
m = 1/(1-K,g) = CR /CRo
26.06P g, A + (p-P)X SD/= (1 - K,g)/K,g , i = (A*/P) + (p-P)/XP A K,g i = A/(P-p) P = (K,g - 1)/K,g = /K,g i = (p - P)/(AP + f) P - (A*/tK,g) + [peff/(1 + AT)) P=I(1+AT) P = (I$v)/3E(10) / A* =~10-8 seconds t = 12.7 seconds X = 0.L seconds-1 'C=f('F-32) 'F=l*C+32 R/hr + 0.5CE/d2 (meters) I curie = 3.7E(10) dps R/hr = 6CE/d2 (feet)
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _ PEACH _ggTTQN_2h3 _______ REACTOR TYPE: _gWR-GE4_ ____ ___ DATE ADMINISTERED._g6/12fgg_ ___________ EXAMINER: _KgLgNAugKI _L._.
_ _ _ _
CANDIDATE: ,_________ ____ IN@l@gCllgNS_Ig_C@ND1981El Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade 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.
X OF CATEGORY % OF CANDIDATE'S CATEGORY __VALUE_ _TQTAL SCORE VALUE CATEGORY . _22199 _ 22199 ________ 5.
THEORY OF NUCLEAR POWER PLANT ___________ OPERATION, FLUIDS, AND THERMODYNAMICS 24 04 24 2.
M __ M ___________ ________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _29199__ _29199 ________ 7.
PROCEDURES - NORMAL, ABNORMAL, _ _ _ _ _ _ _.. _ _ _ EMERGENCY AND RADIOLOGICAL CONTROL 22199__ _22199 ________ 8.
ADMINISTRATIVE PROCEDURES, ___________ CONDITIONS, AND LIMITATIONS P10 19ergg__ Totals ___________ Final Grade g gggg g All work done on this examination is my own.
I have neither given nor received aid.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = = - _ _ = - - - - _ _ _ Candidate's Signature
__ o ' . . . 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 penalties.
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.
Use black ink or dark pencil gnly 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.
6.
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 three lines 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.
l 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
( 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
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 examination has been completed.
t ._.-_.,-,-._y.,__,___._,,. , -, _ _ _, _ _ _ _, _ _.,. _ _. _, _. -, ._._,m___....._ , _ _ _. _.. _. _,.. _ _ _ _.,, _ __ _ _.... _
a e . 5.
THEORY OF NUCLEAR __ POWER _ PLANT OPERATION _FWIggi__AND PAGE
1 - ItjER!$9DLN@t!ICS (i.0\\ ) QUESTION 5.01 (2.00) While an I/C technician calibrates the reactor high pressure switches, an error is made and the plant scrams on an erroneous high pressure signal at 0700 after 3000 hours at full power. Because the cause of the scram has been determined, and in order to maintain power commitments to the grid, the Operations Manager orders an immediate reactor startup.
a.
What are your considerations regarding poisons in the core if the (1.0) reactor is brought critical during the afternoon of the same day? b.
Will control rod density change over the next 24 hours after the (1.0) startup? Briefly explain your answer.
(l.02 ) QUESTION 5.02 (2.00) ' a.
Most condensers are designed with excess condensing capability; (1.0) that is, the condensed liquid leaves the condenser hotwell several degrees below the saturation temperature. How would plant effici-
ency be affected (increase, decrease, or not affected) if the temperature of the circulating water was greatly DECREASED?
Explain your answer.
I b.
If the main condenser was absolutely air tight, would there be (1.0) any need for the air ejectors? Explain why.
(104) QUESTION 5.03 (1.50) , PBAPS Unit 3 is taken critical during startup, and a steady-state period is established. After the point of adding heat (PDAH), the reactor period lengthens to infinity, and the reacter operator notes l that the moderator temperature has changed from 255 deg F to 270 deg F.
' a.
What reactivity coefficients turned reactor power? LIST them in (1.0) order from the largest effect to the least effect.
b.
How much positive reactivity was added to establish a stable (0.5) positive period after criticality was obtained? !
l ! I I l (*ssss CATEGORY 05 CONTINUED ON NEXT PAGE *****) l i ' -- -. -. - - - - - - - - . - - -. - .- -. - - - - -. - - -- -
. . . . ' - 18. When you complete your examination, you shall: a.
Assemble your exqmination as follows: (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 examination 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.
I _, _ _ - -. - -. _ _.. _
. 5.
THEDRY OF NUCLEAR POWER PLANT OPERATION _FLUIpp1_AND PAGE
2 - I_HE_RDppY!MDICQ (1 05) DUESTION 5.04 (2.00) a.
Indicate how available NPSH (net positive suction head) will (1.0) change (ie., increase, decrease, or no change) in the following instances.
1.
CRD pumps-system flow significantly decreased 2.
RWCU pumps-reactor power decreased from 100% to 20% 3.
Recirculation pumps-RPV level is decreased 4.
Recirc pumps-reactor pressure is increased b.
Explain what happens within a pump when the "available" NPSH (1.0) drops below the value of the " required" NPSH. Give three (3) possible adverse effects on the pump.
(l.oc ) QUESTION 5.05 (2.00) During calibration of the level switches that initiate the High Pressure Coolant Injection (HPCI) system, an initiation signal is inadvertantly introduced at 100% power. For the following parameters, state the INITIAL change (if any) AND give the cause for such a response.
Assume that the FWCS is in three element control.
a.
Total steam flow indicated (0.67) b.
Total Feedwater flow indicated (0.67) c.
APRM indication (0.67) (l.07/ QUESTION 5.06 (1.50) ' a.
During a reactar startup, what three indications / items are used (0.75) by the operator to determine when criticality is achieved? b.
Fill in the blank: The FIRST rods in a new rod group have (0.75) ___________ (higher, lower, or the same) rod worths than/as the LAST rods in that group. BRIEFLY EXPLAIN your answer.
(l.101 QUESTION 5.07 (3.00) a.
Describe why spray cooling of the RPV is desirable when recovering (2.0) from the blowdown cooling made of core cooling.
b.
Describe the core cooling mechanisms available to cool the core (1.0) when absolutely no ECCS systems are available for core cooling.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
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. 52_IMDBY_QE_NyCLE98_EggB_P(9NI_9EEB911g&_ELylph_999 PAGE O > I M B599YN001CS t - QUESTION 5.08 (2.00) Step LG-2 of Procedure T-117, " Level / Power Control", states: { If Power is above 3% or cannot be determined and torus temperature is above 110 deg F and an SRV is open or Drywell pressure is above 2 psig then lower RPV level by terminating and preventing injection into the RPV except Boron injection and CRD until either Power is below 3% or RPV level reaches -172" (TAF) or all SRVs remain shut and Drywell pressure is below 2 psig.
a.
Explain how the reduction in RPV level reduces power.
(1.0) b. Assume that an ATWS has occurred at full power and with a 100% (1.0) rod pattern. What adverse effect could be expected if the RPV level was NOT lowered, and SLC injection was the only attempt made to lower power? (l.13/ ' QUESTION 5.09 (2.50) The reactor is operating at full power when the RFP Master Controller (2.5) malfunctions, resulting in a total loss of feedwater. A reactor scram is expected to occur within seconds.
During this short period, is reactor POWER level expected to INCREASE or DECREASE 7 Give TWO (2) reasons for your choice, including WHY each causes a change in reactor power.
QUESTION 5.10 (2.00) a.
What phenomenon is prevented by the RHR stayfill system, and (1.0) what might result if the RHR system is operated with the stayfill system out of service? b.
If the RHR stayfill system had been isolated for maintenance, (1.0) and had since been returned to service, what two (2) observations could you make in order to verify that the RHR system was properly filled? (ass ** CATEGORY 05 CONTINUED ON NEXT PAGE **sse) , , . .
_ _. . 5___IM 96L 9E_ N9CL E96_ Egg B_ P(9NI_QEEBAllg5_E(919h _ AND PAGE S IM B599YN051CS - QUESTION 5.11 (2.50) Given the following two (2) conditions and using attached Figure 1, (2.5) determine which condition is operating MORE closely to its MCPR limit.
Show ALL work and state ALL assumptions.
Condition 1 Condition 2 Rx dome pressure = 920 psig Rx dome pressu e = 980 psig Core Flow = 50 % Core flow = 75 X Rx power = 1660 MW Rx power = 2490 MW P-1 MCPR = 1.60 P-1 NCPR = 1.50 NOTE: Assume that the MCPR limit as given by TS Table 3.5.K.2 is 1.27.
,
($$$$$ END OF CATEGORY 05 88888) , - - -. ,, -,-~, ...,-,---_-,.,-.,v,, .. -. - -.-- -. _ _.. - -. -..,, ., _ - - - _. ~. _ -, _ ,,n-- - -.., - -,,,
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. ( 5.011 QUESTION 6.01 (1.50) a.
If the IRMs are indicating 20 on Range 4 and an operator (0.75) down ranged to Range 3, what trips if any, would occur? Why? b. With the mode switch in STARTUP, and IRM "C" reading 11 on (0.75) Range 7, what trip (s) if any, would occur if IRM "C" was down ranged to Range 67 Why? (1.03 ) GUESTIDN 6.02 (2.50) List f our (4) rod blocks associated with the IRM's.
Include setpoints and AUTDMATIC bypasses for each, as applicable.
( 5.65 ) QUESTION 6.03 (3.00) a.
Assuming valid initiation signals exist, what two (2) conditions (1.0) will close the ADS valves once blowdown has commenced? ASSUME NO DPERATOR ACTION AND THAT ADS REMAINS FULLY OPERABLE.
b.
During blowdown the operator depresses the ADS "A" timer reset (1.0) button. Describe the response of the ADS system.
c. According to station operating procedures, ADS may be manually (1.0) reset (depressing the ADS reset buttons) only when what two (2) conditions are met? ( 7.M) QUESTION 6.04 (3.00) a. Assume that a completo loss of Drywell Chilled Water has occur-(0.5)
red while at full power. What threw (3) heat loads are affected? b.
While Unit 2 is at full power, the #2 13 kV bus in lost.
(1.5) Explain the automatic actions that should occur in order to maintain cooling to the Drywell.
c. List the four (4) non-ensential loads that are isolated from (1.0) RBCCW during a loss of power situation.
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! (48888 CATEGORY 06 CONTINUED DN NEXT PAGE 88888) l -
. 6I__ELON1_SISIEDS_ DESIGN _CQN18Q(t_@ND_lNSl@UDENI@llgN PAGE
t . (I.01) DOESTIDN 6.05 (3.00) Answer the following questions using the attached EHC figure (Figure 2).
ASSUME NO OPERATOR ACTIONS.
a.
Unit 2 is operating at 100% power and two bypass valves open.
(1.0) Describe how the ::- -, --.... i = m-..L.
- W 1, C'" aill respond.
TCNt b.
Describe the effect both on the plant and within the EHC system (1.0) if a Stator Cooling trip occurs at 100X reactor power.
c. Explain what will happen to the EHC system and to the plant, (1.0) if the maximum combined flow limit is turned down to 50% while the plant is operating at 100X power.
(L LA ) DUESTION 6.06 (2.00) The following questions concern manual starting of the Emergency Diesel Generators.
a.
Explain why a diesel generator should not be run unloaded.
(1.0) b. Procedure S.O.4.A, " Manual Start of Diesels", states that, (1.0) "At least one minute must elapse between a diesel shutdown or trip and a diesel restart."
If an attempt WAS made to restart the diesel before the one minute time lapse, what would be the status of the diesel? (3.07) DUESTION 6.07 (1.50) a.
List the conditions that must be met in order to open the (1.0) Containment Spray valves (MO-26,-52) after a 'OCA- %mg b. What in the purpose of the# Containment Spray Override"keylock (0.5) switch? (2.C4 ) DUESTION 6.00 (2.50) a.
Briefly explain WHY it is important to closely monitor RWCU (1.0) system flowrate when operating in the blowdown mode.
b.
State the purpone of the RWCU Hlowdown mode and state when (O.D) this move is used.
. c. State the automatic closure signal for RWCU CV-SS (the Clean Up (1.0) Drain Header Control Valve) while operating in the blowdown mode , and explain its basis.
s (88888 CATEGOHY 06 CONTINUED ON NEXT PAGE 88888) , - - -.. _,. _. _. - _ _ _ _ _ _ _ _.. _ _ _ _ _ _.. . -r .-, _ _ _ _ - _. _ _ _. _ _ _. _ _ _ - _ _ _ _ _ . _ _ __ _. _ _ _. _ _.,,.,
. 6 __PL@NI_SygIEdg_pEglgN _CgNIB9L _ONQ_JNSIBUDENIBI19N PAGE
i
. (2.01) DOESTION 6.09 (2.50) Regarding the Standby Gas Treatment System: a.
What are three (3) of the four conditions which will auto (1.5) initiate the system? Setpoints are required.
b.
List the response of the SBGT fans and dampers upon receipt of (1.0) an auto hem 4eMen signal. MIM(/h (Lil) DUESTION 6.10 (1.50) a.
Explain HOW the Core Spray line break detectici system would (0.75) sense AND indicate a break inside the core shroud.
b.
Assume that the Core Spray system has automatically initiated (0.75) , on a 2 psig Drywell pressure signal. In anticipation of an increase in RPV pressure, an operator manually overrides and closes the Core Spray inboard injection valve (MO-12A(B)). Will this valve reopen if RPV pressure then drops below 450 psig? If NOT, explain the action (s) that the operator would have to take in order to open the valve.
(LIA) Ml.0 QUESTION 6.11 l.0 Figure 3 contains a section from a rod sequence instruction sheet.
4270) For the following situations, state ALL of the responses by the Rod Worth Minimizer (RWM). That ist withdraw error, withdraw block, insert error, insert block, select error, or a combination of these.
NOTE: CONSIDER "a" AND "b" SEPARATELY.
a.
Refer to Figure Give the RWM response (s) if the rod marked "*" . is pulled to position 16 during a plant startup.
Refer to Figure Give the RWM response (s) if the rods marked "+" . . are inserted to position 20 during a plant shutdown. Rod 26-43 is then selected.
pcWr/d (***** END OF CATEGORY 06
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.
PROCE'DURES - NORMAL _ABNgRMAL _ EMERGENCY _ANg PAGE
1
89D19LQQICOL_CQNIBgl - (,4.02. ) QUESTION 7.01 (2.50) a.
State the PBAPS whole body limits for radiation exposure received (0.5) per quarter, both with and without an NRC Form 4.
l.O b.
State the three (3) entry requirements that one must meet prior JJ<tI) to entrance to a contaminated area.
c.
Briefly describe the process for the request and documentation of (1.0) an ALARA review for a specific job.
(4.o t ) QUESTION 7.02 (2.00) a.
According to ON-107, " Loss of CRD Regulating Function", if (1.0) both CRD pumps are off, and reactor pressure is below 550 psig, and three (3) or more CRD accumulator low pressure annunciators alarm, the operator is directed to scram and enter T-100.
Why is a reactor scram required in this condition? b.
According to S.4.2.D, " Shutdown of the CRD Hydraulic System", (1.01 the CRDH system should remain in service when the reactor is shutdown. Explain why.
b1.0T I QUESTION 7.03 (3.00) List the entry conditions with appropriate setpoints for: a.
T-101 RPV Control (1.5) b.
T-102 Containment Control (1.5) ($66) DOESTION 7.04 (3.00) a.
List the seven (7) immediate operator actions to be taken prior (2.0) to leaving the main control room should a control room evacuation be necessary.
b.
Once at the Remote Shutdown Panel, procedure SE-1 directs the (1.0) operator to close RHR shutdown cooling valves MD-10-17 and MO-10-18. What is the basis behind this step? l (***** CATEGORY 07 CONTINUED ON NEXT PAGE
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. Z.
PRQGEDURES - NQRMAl _AgNQRMAl _EMERQENQY_AND PAGE
t t R@ Dig (QQ1G@(_GONTROL . DUESTION 7.05 (1.50) a.
How is the term " Locked High Radiation Area" defined in the (0.5) PBAPS Tech Spec 6.13.1? b.
Assume that you were not issued a key to a specific high radiation (1.0) area. Explain how you could obtain one.
QUESTION 7.06 (1.50) While executing PBAPS procedure S 4.2.L, " Maximum CRD Flow to the (1.5) Reactor Vessel Under Emergency Conditions", you notice that the full core display shows a green background with no "00" position indication for several control rods.
Does this indicate a failure within the Rod Positon Indication System (RPIS)? EXPLAIN why.
QUESTION 7.07 (2.50) a.
What two (2) immediate actions should you take as the SRO (0.75) stationed on the the Refuel Floor if a spent fuel bundle is dropped during fuel handling operations? b.
As the Refuel Floor SLO, can you authorize the continuance of (0.75) of fuel handling operations? If not, who can? c.
Assume that the radiation level indicated for the refueling (1.0) floor exhaust duct rises to the high alarm point. List the automatic actions that should occur AND the immediate operator actions that should be taken.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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7 __PROCg_DL.lRES - NORM % i
- 89Dl%ggIC%_CgNIBgl DUESTION 7.08 (3.00) a.
Classify the following event according to Appendix 1 of EP-101, (1.0) " Classification of Emergencies". (Figure 4) A Main Steam Line High Temperature Alarm is received. While verifying the isolation, an RO reports that BOTH the inboard outboard MSIVs for Main Steam Line "C" have failed open.
b.
Answer the following TRUE or FALSE questions concerning the (2.0) Emergency Response Plan. If FALSE, explain why.
, 1.
Implementation of EP-101 (Classification of the Emergency) does not constitute implementation of the Emergency Plan.
2.
The Emergency Director may de-escalate the emergency classification of an event without concurrence from the Site Emergency Coordinator.
3.
The Emergency Director is authorized to decide when the plant can enter the recovery phase after an accident.
QUESTION 7.09 (3.00) The following questions concern DW/P of T-102, " Containment Control" : NOTE - Attached Figure 5 contains DW/P.
i a.
Refer to DW/P-4. Give two (2) reasons for not using SBGTS to (1.5) reduce DW pressure if DW temperature exceeds 212 deg F? j b.
Step DW/P-10 refers to the DW Spray Initiation Pressure Limit (1.5) Curve. Explain why you are allowed to spray the drywell DNLY if you are in the safe region of this curve.
QUESTION 7.10 (1.50) l Step PSR-7 of T-99, " Post Scram Recovery", directs the operator (1.5) to execute T-101 if any control rod is not inserted past "O6".
This is true even if no entry condition for T-101 exists. EXPLAIN.
l (***** CATEGORY 07 CONTINUED ON NCXT PAGE
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. 7 _ PROGEDURES - NOR % _AgNQRMAl _EMERgENG LAND PAGE
- t 809196991G06_GQNI896 - DUESTION 7.11 (1.50) The following questions concern S.3.5.B, " Manual Operation of RCIC" following a Group I isolation.
a.
What is the reason behind the caution: (1.0) "DO NOT open MO-27 until immediately before starting the turbine."
Figure 6 (RCIC), is attached for your reference.
b.
What is the maximum allowable torus temperature during RCIC (0.5) operation following a scram from power and an RPV isolation? (***** END OF CATEGORY 07
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. O ARMINISIB811ME_P8QCEQUBES _CQNQlliQNS _QNQ_blM11@11QNS PAGE
t t . QUESTION 8.01 (2.00) A reactor startup is in progress. All APRMs are reading on scale.
(2.0) The front panel RO is just about to place the mode switch into RUN when IRMs "B" and "D" fail. The Shift Supervisor wants to take the mode switch to RUN anyway. His reasoning is that the IRMs are not required in the RUN mode.
Using the attached Tech Specs, decide what action is required. State whether or not you agree with the supervisor, and explain your answer.
- BE SURE TO REFERENCE ALL TECH SPECS THAT YOU USE ******
i ab ) QUESTION 8.02 (2.00) Answer the following True/ False questions in accordance with A-40, (2.0) " Working Hour Restrictions" and A-7, " Shift Operations". If FALSE.
EXPLAIN WHY.
a.
It is permissible for a Control Operator, who has just completed his normal 0700-1500 shift, to take the place of the 1500-2300 Control Operator if necessary.
b.
It is permissible for a Control Room Operator to work twelve (12) hours a day from Monday through Sunday, as long as he is off on the following Monday.
c.
The Control Room Supervisor may le ave the Main Control Room without a complete turnover to a qualified replacement, as long as he plans 0 return within 30 minutes.
QUESTION 8.03 (2.00) What three requirements must be adhered to when making a temporary (2.0) change to a procedure, as stated in 6.8.3 of the PBAPS Tech Specs? i (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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- -. . O A M IPi_ISTRATIVE_PROCE_DURES2_C M DITIONg2_AND L_IM_ITATIONS PAGE
. QUESTION 8.04 (3.00) Using the attached PDAPS Technical Specifications, answer the following questions. Be sure to reference ALL Tech Specs that you use to develop your answer.
Unit 2 is in a refueling outage and fuel movement is taking place.
It is determined that the Emergency Diesel Generator associated with Core Spray Loop "A" injection valves (E D/G "A") is inoperable.
i a.
What actions, if any, apply to the Core Spray System? (1.0) b.
Can refueling continue? WHY? (1.0) c.
What actions, if any, would apply to the Core Spray system if (1.0) the associated Emergency Diesel Generator became inoperable during a reactor startup at 3*/. power? QUESTION B.05 (3.00) a.
Assume that the current plant condition requires a one (1) hour (1.0) report to the NRC. Briefly explain how you would accomplish this, b.
From the events listed below, identify those which require a (2.0) one hour notification to the NRC as given by 10 CFR 50.72.
1.
The TS limit for the RPV cooldown rate is intentionally violated while executing procedure T-112.
2.
The Reactor Protection System automatically initiates on a valid high Drywell pressure signal.
3.
An orderly plant shutdown is initiated in order to comply with the PBAPS Technical Specifications.
4.
A fire breaks out in the Technical Support Center.
. (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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. O.__@DdlNI@IB@llyE_P8gGEDUBEgt_GQNDlIlgN@t_@ND_LIMITATIQN@ PAGE 1@ . QUESTION 8.06 (3.00) a.
Who (by title) assumes the role of Site Emergency Coordinator (0.4) during the initial phase of emergency repsonse at PBAPS? b.
Name three alternates who may assume this role.
(0.6) c. Assume that a Site Emergency has been declared at PBAPS.
(2.0) LIST ALL of the Emergency Response Facilities outside of the Main Control Room that will be activated AND briefly describe the function fulfilled by each facility.
QUESTION 8.07 (3.00) Unit 3 is in Startup after a refueling outage. Prior to rod with-(3.0) drawal, you are informed that the E13, 4 kV Emergency Bus, Sequential Loading Relay has failed a calibration check and must , be replaced.
' Does this impact the startup? Explain your answer fully.
Use the attached Tech Specs and reference all Tech Specs that you use to develop your answer.
QUESTION 8.08 (2.00) a.
According to A-7, " Shift Operations", WHEN may Operations (1.0) personnel override the automatic actions of engineered safety features? b.
State the " area of responsibility" (i e., the assigned location) (1.0) for the Control Room Supervisor as given in A-7.
QUESTION B.09 (3.00) a.
List the three (3) Fuel Cladding Integrity Safety Limits.
(1.0) Include applicable setpoints and plant conditions.
b.
What actions must be taken in the event that a Safety Limit is (2.0) exceeded? (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) , . - - - - - - -- -
. @- ADMIN'ISTRATIVE_PRgGEQUREgi_GDNDITIgNgi_ANg_LIMITATIgNg PAGE
. QUESTION 8.10 (2.00) a.
Give the minimum shift complement of NRC licensed personnel at (1.5) PBAPS as stated in A-7, " Shift Operations".
Include the assigned duty station of each.
b.
Name the licensed personnel, in addition to those listed for (0.5) "a", who must be present during Core Alterations AND describe the area of responsibility.
QUESTION 8.11 (2.00) Assume that both of the PBAPS Units are in the RUN mode when the (2.0) Unit 2 Uninterruptable AC Power is lost. As a result, the Main Stack Radiation Monitoring System is lost.
What actions must be taken in order to comply with the PBAPS Tech Specs? Address ONLY the Stack Rad Monitoring failure.
Use the attached Technical Specifications AND be sure to reference ALL Tech Specs that you use to develop your answer.
i l l (***** END OF CATEGORY 08
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(************* END OF EXAMINATION
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, -,
- h g8Ta &T?l ' " '5._.;THEgRY_gF_ NUCLEAR _PggqR_ PLANT OPERATION _FLUIggi_ANg - PAGE .17
THERMODYNAMICS . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
(l.0l j ANSWER 5.01 (2.00) a.
Xenon will peak at its highest reactivity approximately 10 hours (1.0) after the scram.
The worth of the rods will be relatively lower than what they would be worth on a xenon free startup.
Thus the reactor will go critical at a lower rod density than before.
b.
Because of the added positive reactivity due to Xenon decay (1.0) and burnup, the control rods will need to be inserted after $ startup in order to control power. Thus rod density will increase.
REFERENCE LOT 1510, LO #3,6 pages 5-7 K/A Group II Reactor Theory, 292006, Fission Product Poisons K.1.07, Xe following a scram 3.2/3.2 - K.1.14, Operator compensation for the change in Xe with time 3.1/3.2 \\ (1.02 i ANSWER 5.02 (2.00) a.
Plant efficiency would decrease (O.25) because the heat rejected (1.0) to the circulating water must be added to the feedwater by the reactor (0.75).
b.
Air ejectors would still be needed (0.25) because air in-leakage (1.0) is not the only source of noncondensibles in the main condenser.
Other NC include radiolytic 02 and H2, and fission product gases.
REFERENCE a.
LOT 1250 Plant Efficiency, LO #2 (1.0) K/A Group 11 Thermo, 293004; K.1.1.2: Discuss subcooling 2.9/3.1 b.
LOT 500 Main Condenser Air Removal, LO #1 (1.0) K/A group I Components, 291006; K.1.18: Reasons for NC gas removal 2.8/2.9 -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _.
. 5 _;THEDRY OF NUCLEAR POWER PLANT OPERATION _ELUIDS _OND PAGE
.
t z - It!ERt!ODYt@t!ICS ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
i ((.04) ANSWER 5.03 (1.50) a.
Mod temp coeff (1.0) Fuel temp coeff Void coeff b.
Delta Tmod = 15 deg F (0.5) Assume no real contribution from void or fuel coefficient.
-1 x 10**-4( k/k)/deg F x 15 deg F = 1.5 x 10-3 k/k added = 0.0015 k/k REFERENCE Rx Theory Handout, Sections 26-30 (1.5) LOT 1440, LO P3,5 K/A Rx Theory Group I, 292004; K.1.08 Compare relative effects of coeff , 3.3/3.3 . (t.09) ANSWER 5.04 (2.00) a.1.
increases (0.25) 2. Mcreases dRCrcQ44 (0.25) 3.
decreases (0.25) 4.
increases (0.25) b.
Without adequate NPSH available, the pump would cavitate. The (1.0) formation of vapor bubbles and their subsequent collapse would cause noise, excessive pump vibration and pump internal wear.
REFERENCE PBAPS Fluids Handout, Section 4; LOT 1290 LO #4 K/A Components, 291004 Pumps-K.1.14 Relation between flow and suction head 2.5/2.5 (1.0) K.1.06 Need for NPSH 3.3/3.3 (1.0) . , ,. - - -....... _ _.. - - - -. - -. - - - - - -. , .., -- -
.. _ _ _ _ _ ___ . _ _ _ .
- 5 _;IHEQRY OF_NUGLEAR_PgWER_ PLANT _QPERATIgN _ FLUID @t_AND PAGE
t THERMDYNAMIgg . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
( (.06 ) , ANSWER 5.05 (2.00) l a.
Total indicated steam flow would remain approximately constant (0.17), because of the small size of the HPCI steam supply line in comparison to the 4 Main Steam Lines (0.5).
b.
Total Feedwater flow would decrease (0.17) because of the level increase caused by HPCI (0.5).
c.
Power would increase (0.17) because of the increased inlet subcooling (0.5).
REFERENCE LOT 1610 Temp Transients (2.0) K/A Abnormal Evolution Group II 295008, High Reactor Water Level AK 2.05 HPCI Interrelation 3.8/3.9 U.07) ANSWER 5.06 (1.50) a.
1.
positive stable period (0.25 each) ' 2.
constantly increasing count rate 3.
no rod motion b.
The first rods in a group have HIGHER rod worths than those (0.75) ! later in the group (0.25) because the local flux surrounding them was increased by the withdrawal of adjacent rods in the previous groups (0.5).
REFERENCE LOT 1530, PAGE 10 K/A Group I Rx Theory, 292008 Rx Op Physics a.
LO #3, K.l.08 (4.1) Power, period changes close to crit.
(0.75) b.
LO #1.B, K.1.03 (4.1) Explain op charac close to crit.
(0.75) ' . - - - - _ _ _ _ _ _ _
. 5 _; THEORY OF_._ NUCLEAR _ POWER _ PLANT OPEBA110N _ELUIDS _AND PAGE
1
TLiERMODYNAMICS . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
( t. l 0 ) ANSWER 5.07 (3.00) a.
Spray cooling is preferred to recover water level following (2.0) blowdown cooling because the core temperature is elevated during steam cooling to provide the delta T so steam can carry away core heat. Core Spray spargers can safely lower core temperatures with a reduj(d possibility of core cladding damage.
c MailidgD b tr p pftKAdAArt." %% injection systems /ykhns b.
1.
Alternate flow paths are used as directed (0.5) by the level restoration procedure. (T-111) 2.
The core can be steam cooled when RPV level reaches the core (0.5) midplane. (T-113 Blowdown cooling) REFERENCE a.
LOT 1560, LO #12 (2.0) K/A Group I Emer Evol;295031 La Lvl, EK3.03 Spray Cooling 4.0/4.3 b.
LOT 1560, LO #11 (1.0) K/A Group I Emer Evol;295031, EA2.04 Adequate Core Cooling 4.6/4.8 ANSWER 5.08 (2.00) a.
Power is reduced by decreasing RPV level because natural circ-(1.0) ulation is inhibited. As level is decreased, the actual flow area across the top of the shroud and through the moisture separators is decreased causing a restriction in core flow.
b.
If water level is not reduced in order to control power, it is (1.0) likely that primary contaf7 ment will fail because the time ' required for SLC to shutdown the reactor exceeds the time that PC can absorb the energy that is being produced.
Cet(dgy g4(egg y ng (gggg:(g m, REFERENCE T-117 Bases, LOT 1560, LO #1,6 a.
K/A Emer Evol Group I, 295037 ATWS, EK.1.02 4.1/4.4 (1.0) RPV water level effects on power b.
K/A Emer Evol Group I, 295037 ATWS, EK.1.03 4.2/4.4 (1.0) Baron effects on power _ -. ~ , -. - _ _ - -.-
. ,' 5._; THEORY OF NUCLEAR _ POWER PLANT _OPERATIONt_ FLUIDS _AND PAGE
z It!ERtjgDYNAtjICS - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
(l.t 3 ) ANSWEP 5.09 (2.50) DECREASE (0.5) REASONS: 1.
Immediately, the loss of feedwater flow causes a (1.0) decrease in moderator subcooling which introduces negative reactivity into the core.
2.
When total feedwater flow drops below 20%, the Recirc (1.0) pumps auto runback to 30% (or if RPV level drops to <17" and an individual feed pump flow drops to < 20%, the Recirc pumps auto runback to 60%). The reduction in core flow causes an increase in voiding which adds negative reactivity to the core.
REFE;<ENCE LOT 1630, 1640 LO #3 LOT 40 LO #4 K/A System 259001 FW, K3.12 KN of effect on Rx power if FW malf 3.8/3.9 ANSWER 5.10 (2.00) a.
Water hammer (0.5) upon pump start could cause damage to the pipes and other components within the system (0.5).
b.
1.
System pressure (local indication) (1.0) 2.
Water flow from the high point vents (d.g;SA) 3.
LPCI line accumulator level alarm cleared in Control Room 2 %%iHC REFERENCE LOT 360, TS 3.5.A Bases K/A 203000 RHR/LPCI A.2.17 KN of keep fill malf effect on LPCI 3.3/3.5 293006 Fluids K.1.05 KN of Op implic of water hammer 3.2/3.7 _ . _ _
. ,'5 _;THEOR'Y OF NUCLEAR POWER PLANT OPERATION _FLijIDgt_AND PAGE
t ItgRt!ODYN@t!IC@ . ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
ANSWER 5.11 (2.50) Condition 1 Condition 2 (1.5) g, From Figure, Kf g,[t t,f.( Kf = ,o t.2.7 = MCPR Limit = ( 1. 27 ) (.1.diMT) = 1.d55 MCPR Limit = ( 1. 27) ( =M dMCPR = 1. 60 - 3 5"i = O<O3 dMCPR = 1.50 - 1.<4T = OrO9 L 4L o.t9 t.27 0. 2.3 Condition 1 is closer.
(1.0) REFERENCE TS Bases 4.5.L/ TS 4.5.K,L K/A 293009 Core Thermal Limits K.1.19 Explain Basis of CPR LCO 3.6 K.1.27 Flow biasing w/ MCPR 3.3 __
. _ . 6 _;PL8MI_SYSIE5S_DESIGNt_CQNIBgL _QND_INSIBUMENI@IlQN PAGE
t . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
(1.ct) ANSWER 6.01 (1.50) a.
None (0.25). The IRMs would indicate 20 on Range 3; O-40 scale (0.5).
b.
An IRM upscale trip (rod block) would occur (0.25), because the IRM would read 110 on the 0-125 scale (0.5).
REFERENCE LOT 250, LO #3,5,7 K/A Group I Components, 291002 Sensors, detectors (1.5) K.1.20 Neutron monitoring units 3.2/3.3 (3.021 ANSWER 6.02 (2.50) BLOCK SETPOINT BYPASS
_
, Upscale 108/125 Mode switch in RUN Downscale 2.5% scale Downscale on range 1 (and MS in RUN) Inop N/A N/A - Manual only 1) (and MS in RUN) ' Detector not full in N/A Mode switch in RUN (4 Blocks at O.25 each, 2 Setpoints at O.25 each, 3 Bypasses at O.33 each) REFERENCE LOT 250, RBM, LO #4 K/A Group I Systems, 215003 IRMs, K.4.01 Rod Blocks associated with IRMs 3.7/3.7 (3.09) 4, [) - ANSWER 6.03 (3.00) c, ' pressure <50lpsig - (0.5) a.
Rx Trip of the'RHW}and CS pumps (O.5) i CE b.
Blowdown will continue. (If both the "A" and the "B" were reset, (1.0) blowdown would be interrupted.)
Skt l ,c.
Level > -130" 5. 3. (0. F (O.5) l _ kh3 Makeup to vessel available glen (O.5) l Conft*>d m(50 Wraktb, dO.
- M d i 4 d. h M y I?tt M REFERENCE l
LOT 330, LO #1,4, pages 5-7 K/A Group I Systems, 218000 ADS (3.0) _- _ __ - _. -
-_ - . 6._; PLANI _gygIgMg_DEg1GN _CgNIBgl _AND_1NgIBUDENTATIgN PAGE
2 i . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
K.4.03 KN of ADS Logic control 3.8/3.9 (3.06 > ANSWER 6.04 (3.00) a.
DW Area cooling units, Recirc pump motor coolers, DWEDS cooler (0.5) b.
The motor operated transfer valves and the air operated isolation (1.5) valves will open in order to isolate DW chilled water and line up RBCCW to both the "A" and "B" headers. The non-essential loads would be isolated by an air operated valve within the RBCCW system.
c.
Non-regenerative heat exchangers (1.0) RWCU pump coolers Instrument Nitrogen coolers Sample coolers REFERENCE LOT 150, LO #2,3,4,5,6 K/A Group I Systems, 223001, Prim Cont and Aux K.6.01 KN of DW Cooling 3.6/3.8 (2.5) K.2.10 KN of Pwr supplies to DW Cooling Units 2.7/2.9 (0.5) (5 0D ANSWER 6.05 (3.00) Two BPVs cause about 6% flow.(Reactor pressure will decrease (1.0) a. because of the excess steam flow.) Lower reactor pressure will cause the CVs to close to approximatelyl94% flow.)
b.
Stator Cooling trip will cause an EHC load set runback to about (1.0) 23% (of Gen output). the Recirc pumps will trip (A: 1s TD, B: 10s) (and power will decrease to about 55%) As the load set decreases, the kNtreasing reactor pressure signal will cause the BPV to go full open. [The CVs will be at about 23% (of Generator Output amps).] l c.
Max Comb Flow at 50% limits total CV and BPV opening to 50%. (1.0) , With power at 100%, as CVs close, reactor pressure increases, l voids collapse and a high flux (or high pressure) scram will occur.
REFERENCE LOT 590, LO #4 K/A Group I Systems, 241000 Rx/ Turbine Press Reg K.6.10 BPV failure effect on EHC 3.6/3.7 (1.0) K.6.16 Stator water cooling failure 2.9/3.1 (1.0) A.1.15 Max Comb Flow Limit predict changes 3.1/3.1 (1.0) ___________ _
. 6 _;BL@NI_@Y@IEd@_DE@lGN _CONIBOL _@ND_lN@lBUDENTATION PAGE
t t . ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
(2.04 ) ANSWER 6.06 (2.00) a.
Operating an unloaded diesel increases the air blower temperatures (1.0) to the maximum operating value, and may result in blower damage.
OP - CA40n dtydsits b.
The E D/G will turn over, but the fuel racks will not open. A (1.0) failure to start trip will result.
REFERENCE PBAPS System Procedure S.B.4.A K/A Group I Systems, 264000, E D/G K.1.06 KN of E DG starting system 3.2/3.2 (1.0) A.2.03 AB to predict impact if E DG is run unloaded 3.4/3.4 (1.0) (7.0S/ ANSWER 6.07 (1.50) a.
DW pressure greater than 1 psig (1.0) 2/3 core coverage LOCA signal present (SWM * S%%MhhM) Permissive switch in manual b.
It allows opening of containment spray valves by bypassing the (0.5) requirement for 2/3 core coverage and a LOCA signal.
REFERENCE LOT 370, LO #3 K/A Group Il Systems, 226001 Containment Spray (1.5) K.1.08 KN of interrelation w/ RPV instr 3.2/3.4 K.1.13 KN of interrelation w/ DW instr 3.1/3.2 i ! ' i . _ -, _ - _ . _.
- -,
. 6- ! PLANT SYSTEMS DESIGN, CONTROL _AND INSTRUMENTATION PAGE
3 i
ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
(L.04 ) ANSWER 6.08 (2.50) iS Because the regenerative HX pre" bypassed [ blowdown flow must be (1.0) a.
limited to the capacity of the NRHX in order to prevent over-heating of the demineralizer bed.
b.
Blowdown is used during startup or hot standby operations to (0.5) reduce reactor water invent %e a Denain Of ({ NI,' c.
In order to prevent draining of the system, CV-55 closes on low (1.0) pressure sensed upstream (5 psig). This avoids, in particular, an isolation condition while dumping to the main condenser.
REFERENCE LOT 110, LO #6,7 K/A Components, 291007, Demin K.1.06 Reason for temp limits 2.7/2.7 (0.5) Group II Systems, 204000 RWCU K.1.09 KN of relation b/t RWCU and RPV level 3.2/3.3 (2.0) (2.0 5 ) ANSWER 6.09 (2.50) a.1.
reactor water level 0" (any 3 at 0.5 each) 2.
D/W press 2 psig 3.
Rx Bldg exhaust 16 mr/hr 4.
refuel floor exhaust 16 mr/hr b.
The auto start signal from Unit 2 will start the "A" fan and (1.0) open its inlet and outlet dampers. The inlet and outlet dampers to both the "A" and "B" filter trains will open. (If fan "A" fails to start, "B" will start.)
REFERENCE LOT 210, LO # 2 K/A Group I 261000 SBGT A.2.05 Predict impact of fan trip 3.0/3.1 (1.0) K.4.01 KN of auto initiation 3.7/3.8 (2.0) ,
. h_ZPL@NI'_@ySIEt!S_ DESIGN,,_CQNIBgL _QND_lNSIBUt!ENI@IlgN t PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
Y WYYWOf b@ 04(d.1 of-catsQ (3.lt )
E d N et10tc.
gI ANSWER 6.10 (1.50) a.
The low leg of the lea etection system would now read the (0.75) outer annulus pressure. The high leg would still read the top of the core plate pressure. The delta P between these would then-i_~ __2. The control room delta P annunciator alarms at y psid.
C olly M j M d F8 M F l T(4 ff(d b.
NO (0.25), it will not reopen unless the outboard injection (0.75) valve is closed first (0.5).
REFERENCE LOT 350, LO #8, pages 7-9 K/A Group I Systems, 209000 Low Pres Core Spray K.4.04 KN of line break detection 3.0/3.2 (0.75) A.4.03 AB to manipulate / monitor CS injec v1vs 3.7/3.6 (0.75) (5 12 )
ANSWER 6.11 J,3ve07 f.II a.
WD error, select error, WD block 66.2R each) h_ ~( i n ~ 4- -- r,,,,_,, m;mmm, ,, us u u, ucacus m.
. u.
u REFERENCE LOT 90- LO #2 K/A Group II Systems, 201006 RWM A.2.04 AB to predict effect on RWM for out of seg rod 3.1/3.5 (2.0)
, . 7, _'PRQQEguRES _NQRM %2 ABNQRM%2_EMERGENgY__AND PAGE
. _ !!a919!=9919%_99NIRgl . ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
(4.01) ANSWER 7.01 (2.50) a.
w/ 2500 mrem /qtr, w/o 1000 mrem /qtr (0.5) b.
All dosimetry devices par $ each) Signed on RWP (i f one is required) C.16 Anti-C clothing as specified by sign c.
Fill out a Review Request Form or stamp to MRF or RWP request form.(1.0) An ALARA group member will review the request.
The ALARA group findings will be returned to the originator.
(Request and evaluation will be stored in history files.)
REFERENCE LOT 1730, LO #3-Exposure Limits; LO #1-Entry req.
(2.0) K/A PW Gen, K.1.03 KN of 10 CFR 20 and facility radcon 3.3/3.8 , LOT 1770, LO #2 (1.0) K/A PW Gen, K.1.04 KN of ALARA pgm 3.3/3.6 l (4.04) ANSWER 7.02 (2.00) a.
With reactor pressure below 550 psig, the accumulators are (1.0) required to assure adequate scram speeds. The scram is inserted before all accumulators depressurize.
b.
CRDH provides a continuous flow of water to the CRD units to (1.0) prevent crud buildup.
REFERENCE LOT 70, CRD Hydraulics, LO #2, OT 107 Bases (1.0) K/A Group II Abn Evol, 295002, Loss of CRD pumps A.K.1.01 KN of Rx pressure vs. rod insertion capability 3.3/3.4 A.K.3.01 KN of reason for Rx scram due to CRD pump loss 3.7/3.9 S.4.2.D, "SD of CRDH" (1.0) K/A Group I Systems, 201002 CRDH K.3.03 Loss od CRDH effect on CRD Mechanism 3.1/3.2
J , - . _ _
. 7 _;PROCEDUREg_ _NgRMAL _ADNgRMAL _ EMERGENCY _ANg PAGE _29
1 RADIOLOG_ICAL_CQ M RgL . ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
i G44s7 ANSWER 7.03 (3.00) a.
RPV level below -48" or unknown (1.5) DW Pressure >2 psig Group I Isolation Scram condition and Rx power > 37. or unknown b.
Torus temperature > 95 deg F (1.5) Torus level outside of 14.6' to 14.9' DW pressure > 2 psig DW Temp > 145 deg F REFERENCE LOT 1560, LO #9, pages 6-7 K/A Group I E/A Evol, 295010, High DW Press (3.0) SG Mll, Recognize entry conditions 4.2/4.5 (4% ) ANSWER 7.04 (3.00) a.
Runback Recirc flow to minimum (0.29 each) Transfer house loads Manually scram and execute procedure T-100 concurrently Place DW instrument air in service Close the eight MSIVs Establish torus cooling Obtain the Emergency SD Panel key b.
To prevent a potential LOCA due to high reactor pressure in the (1.0) low pressure shutdown cooling line.
REFERENCE SE-1 and SE-1 Bases K/A Group II E/A Evol 295016 CR Abandonment SG #10 Perform immed actions w/o refer 3.8/3.6 (2.0) (A)A2.03 Interpret Rx press as applied to CR Aban 4.3/4.4 (1.0) . - - , - - - - -
. -_ -__ . 7.
f_P R Q C E D U R E S - N Q R M A l _ A B N Q R M A l _ E M E R G E N C L A N D PAGE
t t RQQ1g(QQ1CQL_CONTRQL ' ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
i l ANSWER 7.05 (1.50) a.
>1000 mrem /hr (0.5) b.
Go to the HP Office (Turbine Bldg, 116' elev); a controlled (1.0) key cabinet is located there. A key would be issued by a " responsible person" and the issuance recorded in the key log.
REFERENCE LOT 1570, SRO LO #5m, A-84 K/A PW Gen, K.1.03 KN of Fac Rad Con req 3.8 (1.5) ANSWER 7.06 (1.50) NO (0.5), this indication should be considered normal for this procedure because of the high flows through the cooling water header.
The rods have overtravelled beyond full in (1.0), REFERENCE S.4.2.L, LOT 80 K/A System 201001 CRDH, A.1.10 AB to predict changes asso" w/CRDH Cooling Flow 2.8/2.6 ANSWER 7.07 (2.50) a.
1.
Terminate all fuel handling (0.75) 2.
Evacuate the Refuel Floor b.
No (0.25), authorization to continue fuel handling must be (0.75) obtained from the Plant Superintendent (O.5).
c.
Automatic: Control Room Alarms-Hi Rad Refuel (0.5) Group III Isolation Immediate: Evacuate the Rx Bldg (O.5) Verify the Group III Isolation REFERENCE PBAPS Procedures FH-6C and E-5 K/A AB Evol 295023 Refuel Accidents AK3.01 KN of Refuel Floor Evac 3.6/4.3 AK2.03 KN of Rad Mon Equip 3.4/3.6 SG 1 KN of notif of plant personnel 3.3/4.2 i _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. Z:_;PBgCggURES_- NORMAL _ABNgRMAL _gMERggNCY AND PAGE
2
BBD19LQgICgL_CgNIBQL - ' ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
ANSWER 7.08 (3.00) a.
Unusual Event (1.0) b.
1.
TRUE (O.5) 2.
TRUE (0.5) 3.
FALSE (0.5), the concurrence of the SEC, ED, and the ESO is (1.0) required (if all have been activated) (0.5).
REFERENCE LOT 1520, SRO LO #2 EP-101 K/A PW Gen A.1.16 AB to use E Plan 2.9/4.7 ANSWER 7.09 (3.00) a.
Avoid excessive water vapor flow through the SBGT charcoal filters (0.75) DW temp >212 F indicates a steam break as the cause of high DW (0.75) pressure b.
The unsafe area of the curve represents a PC deficient in non-1 (1.5) condensibles. If DW sprays were initiated while in this unsafe condition, negative pressures could be achieved in the Primary Containment, possibly beyond the capability of the Rx Bldg to PC vacuum breakers.
REFERENCE T-102 Bases, LOT 1560 LO ANSWER 7.10 (1.50) If some rods are not inserted past 06, the reactor could go critical during cooldown (0.75). Therefore, control rod insertion is required; T-101 directs control rod insertion (0.75).
REFERENCE T-99 Bases LOT 1560 Trip Procedures K/A AB Evol 295015 Incomplete Scram AX1.02 KN of cooldwon effects on Rx power 3.9/4.1 _ - _--. - - . -. - . . --- --.- - .
. 7 _~ MggEQURES_- NgRMAl _A_BNgRMAL _E_MERgENCY AND PAGE
i
B0919Lgglg@L_ggNIBQL - ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
ANSWER 7.11 (1.50) a.
To avoid excessive draining of the Condensate Storage Tank to (1.0) the Torus b.
120 deg F (0.5) REFERENCE S.3.5.B i .
! l .
_ _ _ _. _ _ _
_ ._.
. . _. _. C _;@QMINigIB@IlyE_P8QgEQUBEgt_qgNQ111QNgt_@ND_LIMITATIQNS PAGE
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
ANSWER 8.01 (2.00) Per TS Table 3.1.1, a minimum of three channels per trip system is (2.0) required, or the proper action must be taken.
A half scram should be taken in RPS Channel B.
If this is not possible, control rod insertion must be initiated within 4 hours.
--OR-- , If very close to placing the mode switch into RUN, the half scram may be taken as required by TS. Once in RUN, the IRMs are not required and the half scram may be removed.
REFERENCE TS Table 3.1.1 LOT 1840, SRO LO M4 (4 05/ ANSWER 8.02 (2.00) a.
TRUE (0.5) b.
FALSE (0.25), no individual may be allowed to work more than 72 (1.0) hours within any 7 day period (0.75).
c.
TRUE (as long as an SLO remains in the MCR) ( 0. 5 ) REFERENCE A-7 " Shift Operations", A-40 " Working Hour Restrictions" (2.0) LOT 1570, LO #2b, 3h K/A PW GEN, A.1.03 Use procedures related to shift activities and staffing. 2.7/3.7 ANSWER 8.03 (2.00) a.
The intent of the original procedure is not altered.
(0.67) b.
The change is approved by at least two members of the plant (0.67) management staff, at least one of whom holds an SRO license on the affected unit, The change is documented and reviewed by PORC and the Manager-(0.67) c.
' Nuclear Plant within 14 days of implementation.
REFERENCE TS 6.8.3 LOT 1570, LO #1 A-3 l
. - -. _ - . _ _ _ _ _ - - - ' G.20DMINISIRBIIVE-PROCEDURES _CQNDITIQN@t_AND_LIMITATlQN@ PAGE
t . ANSWERS -- PEACH BOTTOM 2&3 - -86/12/08-KOLONAUSKI, L.
ANSWER B.04 (3.00) a.
3.OD does not apply due to cold shutdown conditions (pg 34b) (1.0) CS loop A, by definition, is inoperable.
However, TS do not require it to be operable in this condition because3.5.F.4.b is met.
-- Refueling dg -- No ootential of draining RPV b. YES (0.25). 3.5.F.4.b is met. (0.75) (1.0)
- er A.for Jod. 0K 3.F f i k 3.5.A.2appliesforSTARTUP, 3.OD does not apply.
( 1. O ) c.
s... e. i r. is: REFERENCE PBAPS TS LOT 1840, SRO LO #4 ANSWER 8.05 (3.00) a.
Use the ENS (red phone) to inform the NRC Operations Duty Officer (1.0)
b.
1-2-3-4 (2.0) REFERENCE LOT 1570, SRO LO #4, Sf K/A PW Gen, A1.05 Reports 3.8 (3.0) , i ANSWER B.06 (3.00) a.
Superintendent-Nuclear Generation Division (0.4) Mk ma.h ind @ _.. Superintendent-Nuclear Servic(o.2 cx4' inqu/ b.
es (O.
Station Superintendent-LGS (O ) Station Management-PBAPS (or designated station senior engineer) (.2) c.
OSC (O.25) supports Operations personnel (O.25) AUXOSC (0.25) contains Rad Pro personnel (0.25) EOF (0.25) activities coordinated between on-site and off-site response groups (0.25) TSC (0.25) Tech engineers assess plant conditions and make recommendations to the MCR personnel (0.25).
i
0 -__________._._,,m.. -,. _ _. _ _. _ _ _, _ _,, _, _., - _ -. _ _ _ _ - -, _,- - _,. - ... _... -. _ _ _. , _ _ _ _ _, , _.. _. _ .
. ' 8 _1AQdidl@lB@llyE_P8QgEgygE@t_GQNQlligNgt_@NQ_ LIMITATIONS PAGE
7 . ANSWERS -- PEACH BOTTOM 2&3-86/12/OB-KOLONAUSKI, L.
REFERENCE LOT 1520, SRO LO M2,4 K/A PW Gen A.1.16 AB to use E Plan 4.7 ANSWER B.07 (3.00) Sequential Loading Relay Table 3.2.B (page 71), Notes (page 72) (1.0) TS 3.2.B (page 57) requires that the instrumentation of the table be operable when the equipment is required to be operable.
With this relay inop, the automatic start of the E13 loads cannot (1.0) be assured.
fr By TS 3.5.A-ECCS required for startup. Startup may not commente.
(1.0)
- Also h TI 3 9. A.I
- ik reactr thay ed lot vnadA UW REFERENCE PBAPS TS 3.2, 3.5 LOT 1840, SRO LO #4 l l ANSWER 8.08 (2.00)
a.
If continued operation of the ESF will result in an unsafe (1,0) } condition, as specified in the applicable procedure.
! b.
The CRS must be in the main section of the control room in sight (1.0) of, or in hearing range of, the Control Room Operator, or must be in hearing range of the CR annunciators.
REFERENCE }u LOT 1570 LO M2a A-7 itGr K/A PW Gen A1.11 AB to direct personnel in CR 3.3/4.3
ANSWER 8.09 (3.00) TRO p c ( c > l.07 ~ a.
1.
Rx pressure GE 800 psia AND CF GE 10% of rated co,333 2.
Rx pressure LE 800 psia DR CF LT 10% of rated AYRM (CTVy2r/l(0.33) 3.
Rx S/D with irradiated fuel in vessel, RPV Ivl 7160 (0.33) -UW) b.
1.
The unit shall be shutdown.
(0.5) 2.
The violation will be reported to the NRC, the Superintendent (0.5) Nuclear generation Division, NRB committee.
3.
A SL violation report shall be prepared and reviewed by PORC.
(0.5) 4.
This report will be submitted witnim 10 days to the NRC, (0,5) the NRB committee, and the Superintendent Nuclear Gen Division.
.
.
- Or. *_ ADM N I STRAT I VE_PRgCEDUREgz_CQND I T I gNgi_ AND_L I M I TAT I QNS PAGE 36 l
. ANSWERS -- PEACH BOTTOM 2&3-86/12/08-KOLONAUSKI, L.
REFERENCE LOT 1820, LO #1,5 ANSWER 8.10 (2.00) a.
One SLO (Shift Sup) (1.5) , One SLO (Shift Sup or Staff SLO) stationed in CR complex One LO for each unit shall be in control room at all times One LO in CR in addition to above b.
One SLO to directly supervise Core Alterations, with no concur-(0.5) rent duties.
REFERENCE LOT 1570 SRO LO #3b K/A PW Gen A1.03 Shift staffing 4.2/4.2 jny 2 ic, 2 ii,2.1 2 3 V.c t - Sa w.pk shy b 3. V 4 4 m, c,t.'pm nu n y u - 24 N s ANSWER O.11 (2.00) [ k f, o-($(7 t&C$t) b e 3.2F Action 7 a,gE requires initiation of alternate stack (2.0) mo 'toring within 72 hours, and that the system be restored within s v (7) days or a report be submitted to the NRC within ten days.
REFERENCE PBAPS TS 3.0, ON-112 LOT 720 SRO LO #1 K/A Systems 262002 UPS K1.15 KN of UPS Loss on effect of stack monitors 3.0 L oycgs Lesw4.r(yyl(co1. RMn Itn.dt Q N gym da% Rdn Jtuk if ad. - w tAA.r % cbyJ.. . . ..,. _ _ _ _ - - - . ... _ -.. -, -, - - -.
O . TEST CROSS REFERENCE PAGE
. QUESTION VALUE REFERENCE --
- --- (%6 twRr - sgo (1/8(4 @k:- [Wh 05.01 2.00 BANOOOO401 05.02 2.00 BANOOOO460 05.03 1.50 BANOOOO551 05.04 2.00 BANOOOO554 05.05 2.00 BANOOOO563 05.06 1.50 BANOOOO576 05.07 3.00 BANOOOO581 05.08 2.00 BANOOOO589 05.09 2.50 BANOOOO618 05.10 2.00 BANOOOO619 05.11 2.50 BANOOOO623 -_ 23.00 06.01 1.50 BANOOOO317 06.02 2.50 BANOOOO571 06.03 3.00 BANOOOO595 06.04 3.00 BANOOOO596 06.05 3.00 BANOOOO603 06.06 2.00 BANOOOO604 06.07 1.50 BANOOOO607 06.08 2.50 BANOOOO608 06.09 2.50 BANOOOO611 06.10 1.50 BANOOOO616 06.11 2.00 BANOOOO617
25.00 07.01 2.50 BANOOOO584 07.02 2.00 BANOOOO590 07.03 3.00 BANOOOO593 07.04 3.00 BANOOOO598 07.05 1.50 BANOOOO625 07.06 1.50 BANOOOO631 07.07 2.50 BANOOOO632 07.08 3.00 BANOOOO633 07.09 3.00 BANOOOO634 07.10 1.50 BANOOOO635 07.11 1.50 BANOOOO637 _- 25.00 08.01 2.00 BANOOOO582 08.02 2.00 BANOOOO588 08.03 2.00 BANOOOO621 08.04 3.00 BANOOOO622 08.05 3.00 BANOOOO624 08.06 3.00 BANOOOO626 08.07 3.00 BANOOOO627 08.08 2.00 BANOOOO628 i ?
.- ,9* .* TEST CROSS REFERENCE PAGE
. QUESTION VALUE REFERENCE =-- 08.09 3.00 BANOOOO629 08.10 2.00 BANOOOO630 08.11 2.00 BANOOOO636 = 27.00 100.00 . .. _ _ _ _ _, - _.. _ _ _.. ._ _ -_.
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_ __-__ _ _ _ _ _ _ _ _ . hN . ' ~ - APPDOIX EP-101-1 . Page 1 of 13, Rev.15 I .I I I l 6 NERAL CDNDITIONS l l l l l l l LNLELAL EVENT l AMRP l l (Initiate F-102) l (Initiate @-103) l I I I I i i l 1) Situaticn threatens normal level l 1) A situaticn which does or could l l of plant safety.
I regresent a substantial degrada-l l tien in the level of plant safety.
I I I I I oR I I I I ' l l 2) A sitaticn Which warrants the usel l l of additional perscnnel for l I l accident assessment and in-plant l I l respcmse.
l I I I I I OR l l l l l l 3) A situaticn where a release of I l l radioactive material warrants l I l off-site respcnse or mcnitoring, l I l but does not require protective l l l acticns.
l I I I I I I I I I I SrIE DERENCY l CEEPAL DERENCY l l (Initiate EP-104) I (Initiate EP-105) l I I I l l I l 1) A situaticn where the level of i 1) Substantial core damage and loss l l safety has or could be degraded I of, or hidh potential for, loss ofI l to the point of losing a plant I primary ccntainment.
I l functicn needed to protect the l I I public.
I OR l I .I I l OR I 2) A ccnditicn that warrants use of l l l additional perscnnel for accident l 2) A ccnditicn that warrants use of I assessnent, in-plant respmse, and I alditicnal perscnnel for accident I off-site emergency recptnse to aid I assessnent, in-plant respcnse, I in the implescentaticn of protec-l and off-site emergency response.
I tive acticns.
I I I I I oR I oR I I I I I 3) A sittaticn where a significant l 3) Iarge amounts of radicactive ! I release of radioactive material l material has been or could be re-I k l has occurred or could take place. I leased in a short pericd of time. I i I i i l I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _
_ _ _ _ _ _ , . APPENDIX EP-101-1 . Page 2 of 13, Rev. 15 . k I I I-UNPIAMED SHUrDOW I I I I i l ,
(NIEIAL EVENT I AIERT l
l l l l l l l LNPIAMED SHUrDOW l SCRAM WIM TRIPIE ID IEVEL l l l l l Refer to A-31 for notificaticns. Ib l 1) Scram alarm and l I not initiate IP-102.
I 2) Double low level alarm (-48") and l l l 3) Triple lw level alarm (-130") andl l l 4) Increase in ocntainment pressure I , I to greater than 1 psig but less l l but less than 2 psig cm PR-2/3508. l I I I I I SmAM WIE SMIL IEAK I I I I l l 1) Scram alarm and l I I 2) Double low level alarm (-48") and l
l l 3) Triple low level alarm (-130") and l l 4) Ccntainment high pressure alarm l I (2 psig) and l I l 5) Ocntairnent pressure 2 psig or l l l greater cm PR-2/3508.
l I I I I I I I I I I i l l SITE EERGENCY l GENERAL EERGNCY l l (Initiate EP-104) l (Initiate EP-105) I I I I
I i i ! I SCRAM WIE IiXA l SOAM WIE MXA & NO ECCS l l l l l 1) Scram alarm arx1 l 1) Scram alarm and l l 2) Double low level alarm (-48") arx5 1 2) Double low level alarm (-48") and l l 3) Triple law level alarm (-130") andl 3) Triple low level alarm (-130") andl I 4) Contairment high pressure alarm i 4) Active fuel range level indicatTnl l (2 psig) and I shows less than -226" cn l I 5) Ctntainnant pressure 10 psig or l Il-2/3-2-3-91A, B and I l greater cn PR-2/3508.
l 5) Failure to reset triple low level . l l alarm after 3 minutes and ' l l 6) Cantainment high pressure alarm l l l (2 psig) and l l l 7) Ocntainment pressure greater than l l 20 peig en PR-2/3508.
, I I I I l I I I I I I I I l l I t - ! I I I , - - - -. - - - - -. - . -. - - - - - . - - - - _ .. - _ _ _.
. APPENDIX EP-101-1 ' Page 3 of 13, Rev. 15 ( l i I PEIENNEL IElRY l l I I i l l INCEIAL EVENT I AIERP
l l l l l l l l l l l
1 I I I I I I I I I I I I I I I IIUlRY WITH EXCESS RADIXrION EXPOSLRE I l l CR CINTAMINRTION I I l l N/A l l Refer to A-31 for notificaticns. Do l l l not initiate EP-102.
I I I I I I I I I I I I I I I I I I I I I I I I I I I I i I i SrE DER 2NCY l CENERAL DERGFKY l l I I I I i I I i l l I l l I I I I I I I I I I I I I I i l I
I l l N/A I N/A I I I I I I I I I I I I I I I I I ( l i I ' I I I , I I I ' l .
_ _ _ __-_ APPENDIX EP-101-1 i Page 4 of 13, Rev. 15 f I I l PRIMRY CONTAINbENT l l l l l l l tNtstRL EVENT AIERP l l l l l l
NCN-ISCt.ABIE IEAl%GE I ILSS T PRIERY CONTADDENT INIEGRITY l
l l l 1) Primary contairunent leak rate is I 1) Reactor Building vent rad effluenti l greater tien 0.5 percent of I hi@ rad alarm and inability to voltme per 24 hrs at 49.1 psig or l maintain pressure greater than l 0.25 psig on narrow range l 2) N2 makeup system is not capable ofI PR-2/3508 or l maintaining pressure (not due to
- l I I lack of N2).
I 2) Torus room flood alarm with level l l l decrease in torus.
l I I l l RILtRE TO ISOIRTE PENErlVCTICN WIEN l l l ISOUTED BY A TRANSIENT l l' I I l 1) Incorrect valve positicn during l l Grotp I, II, or III isolaticn I alarms.
l I I I I I I I l l l l l l l
SrIE DER 2NCY l GENERAL EbERENCY I I l l l l
i LOSS OF PRIERY CONTAINENT INIEGRITY l l l l N/A l l 1) Erratic ocntainment pressure l l l fluctuaticns above alarm setpointsl l
of 1.5 psig, M isolaticm alares,1 ard L SITE EPERGNCY I 2) Group II and (Continued) l .I and ll POIENTIAL IAES T PRIERY OCNTADDENT l 3) Etainment dose rate greater than HIGi RADIXrION 4X10P3 R/hr cm RI4/9103A/C and RI4/9103B/D ard l 1) Ccmtainment high pressure alarm l 4) Reactor Bldg area hi@ temperature 1 (2.0 psig), and I l alarm, or Area Ra31aticn Mcnitors l 2) Scram, and l l cn PR-27J-18-55 ainormally hi@, I 3) Ccntainnent dose rate greater thanl and Reactor Bldg vent rad effluent l 4X10P3 R/hr cm RI4/9103A/C and I hi$ alarm, or main stack rad I RI4/9103B/D, and l l effluent on H 0-17-051 increasing l 4) Reactor Building area rad mcnitors I due to SGIS operaticn.
I alarming, and l l 5) Vent stack rad effluent mcnitor l l l alarm.
l ,
' ' l I l
l l . _ _ _ _ _. _ _ _ _ _
. . APPENDIX EP-101-1 . Page 5 of 13, Rev.15 l ( i l l l RADIOACTIVE REEASE l l I l l l l IN WIAL EVENT l AIERT l l l l l l
l IMTANTANE0W REEASE DCEEDING TEGI. l ACIIAL OR POIENTIAL REEASE 0.01 EM l l SPEG.
l MIME BODY OR 0.05 REM THYROID l l l l l 1) Liquid effluent release exceeding l 1) Uncontrollable release for more l l Tech. Spec. 3.8.B.1 l than 20 minutes from the l l 2) Gaseous effluent release exceeding ll a. Main stack greater than 1X10P5 l Tech. Spec. 3.8.C.1 which may be cps cn RR 0-17-051 or l ' I detected by the following l b. Rx. Bldg. Vent grenFer than l l a. If all releases are from cne l 6.10X10P5 cpn cn RR-2/3979 or l l release point.
12) Ccntinued particulate or icrline l 1. A spike en the main stack l release such that analysis of l greater than 1.8X10P4 cps en I particulate filter or charcnal l
RR-0-17-051 or l cartridge results in the followingl l 2. A spike cn tE Rx. Bldg. vent l estimated release rates: l l greater than 3X10P5 cpm cn I a. Main stack greater than l l RR-2/3979 or l 9.7 X 10P2 uCi/sec or l l b. If the relEse is from more l b. Rx. Bldg. vent grenEer than I than one release point i 1.1 X 10P2 tCi/sec or 1. A spike cn the main stack l l ~ i greater than 6X10P3 cps cn l l l RR-0-17-051 or l l l 2. A spike cn tE Rx. Bldg. vent I greater than 1X10P5 cpm cn l RR-2/3979.
l l l
l l SITE EERENCY l G:NERAL EERGNCY l l l l l l l l ACTUAL OR POIENTIAL REMASE 0.1 REM l ACTUAL OR POIENTIAL REEASE 1.0 REM l l MOE BCDY OR 0.5 REM THYIOID l MOLE BGTI OR 5.0 REM 'IHYlOID l l l
l 1) Unatntrollable rolese for more l 1) Unctntrollable release for more l l than 20 minutes f rom the l than 20 minutes from the l l a. Main stack greater than 2X10EB l a. Main stack greater than 2X10P9 l l cps cn E-7127 or l cps on RR-7127 or l l b. Rx. Bldg. vent greater than l b. Rx. Bldg. vent greater than l l 6X10P6 cpn cm RR-2/3W9 or l 6X10P7 cpn cn RR-2/3979 or l l 2) Ccntinued particulate or icrTfne l 2) Ccntinued particulate or io3Tne I release such that analysis of release such that analysis of l particulate filter or charcoal particulate filter or charcoal l cartridge results in the followingf cartridge results in the following l estimated release rates: I estimatai release rates: l l a. Main stack greater than I a. min stack greater than l
l 9.7X10P3 uCi/sec or l 9.7X10P4 uCi/sec or l I b. Rx. Bldg. Vent grEter than I b. Rx. Bldg. vent grEter than l , l l 1.1X10P3 tCi/sec.
I 1.1X10P4 uCi/sec l l h . - _, _ _, _ _ _ _ - _ _ _... _ _ _ . _ _., _ _. _ ... _... _ - _ _ _ _,... _. _... _ _,. , , -
. . APPENDIX EP-101.1 Page 6 of 13, Rev.15 i l l l FIRE I I I I i i I tmEIAL EVENT l ALERT l l l l l l FIRE IN PIUIECTED AREA IASTING 10 MIN. I FIRE MIIOi CDULD MKE AN ECCS IZOP l I OR PzJrc. Antr INITIAL anuPIS 'IO l l l EKrINGJISH IT 1) Fire alarm and verbal report from I it.
I 1) Alarm and verbal report f rom SSV. I ll l l l l l l l l l l l l l l l l l l l l l l l l l l l l l l
1 I I I I I I I I I I I I I I I I I I I SITE DERGENCY I GENERAL EERGENCY I I I I I I I I I I I FIRE MIIGI EKES AN ECCS INOP I FIRE MIIGI CAUSES DMGE TO PIANT I I I sysHNS SUFFICIDTI TO MAD TO OTHt:R I l 1) Fire alarm and verbal report f rom l CEtERAL DERGNCIES I I SSv.
I l l l 1) Fire alarm and verbal report from I .' l l SSV, and IDCA symptone, ECCS, or i l l cxntainnent fallure.
I l l l l l l
I I j l I I I I I I I I I I I I I I I I I I , t i I I I I I I I l . . . _.. _ _ _ _ _ _. _
M , APPDIDIX EP-101-1 Page 7 of 13, Rev. 15 . I I I I DvIaCNEVTAL -
I I I I l LNLELAL EVENT AIERT I I i i I I EARm00AKE l l l l l 1) An actual earthquke beycmd the i I Operating Basis Earthquake (OBE) l I I I I I ABNOREL PCND IEVEL I ,
I I I i 1) Ccnowingo Pond Invel cn i I l LI-2/3278A, B,C: l l l l l N/A l a) greater than 115 feet or l l I b) less than 98.5 feet wiUiout l l prior notificaticn by L.D.
I I I I I TORNADO I I I I I l 1) A tornado strikes the Power Block l l l with identifiab'e plant damage.
I I I I i l HURRICANE I I I I l l 1) Staticn is experiencing a hurri-l I l cane with wirris > 100 mph.
I i l l SITE DERGENCY I GENERAL DERGENCY I I I I I I I IARmOUAKE l l l l l 1) An earthqtake greater than Design l l l Basis Earthquake as detected cn I seismic instrtments.
I I I I I I I ABNCREL PCND IEWL l N/A I I I I l 1) Ccnewingo Pmd Imvel cn l l l LI-2/3278A,B,C exceeding the l l l following limits: l l l l l l a) greater than 116 feet or i l I b) less than 87 feet
I - I l l l
1 I I I I I ,
P . ' APPENDIX EP-101-1 Page 8 of 13, Rev. 15 - I I I I I4SS T POWER l l l l l mtStmL r. VENT l AIERP l I I i l i ILES T OFFSIIE m WSIIE POER l IIES T OFFSIIE AND WSIIE AC POWER I l l HR LESS THAN 15 MINUTES l l 1) 'Ibrbim generator trip with start-l 1) 'harbim generator trip with start-l up auxiliary transformer SU2 and l up auxiliary transformer SU2 and l SU3 unavailable for sevice for l SU3 unavailable for service and l l acre than 60 seccnds or-l l I l 2) Failure of all diesel generators l l 2) loss of voltage cn the four 4160 to energize their busses.
volt emergency busses or 480 volt ,l 1 cad centers stpplied from tie l LIES T ALL DC POWER FOR IESS THAN l l four 4160 volt emergency busses l 15 MIN.
I for more than 60 seccnds.
I 1) Imss than 105 volts cn the 2/3A,B, I l l C & D distributicn panels as indi-l-l l cated cn panels 2/3AD03, 2/3CD03, I l l 2/3BD03, 2/3DD03 aM l l l l l l 2) Imss than 21 volts cn the 24 volt l l I distributicn panels as indicata3 i l I cn panels 2/3AD28, 2/3CD28, I l l 2/3BD28, 2/3D028 and l I I I I l 3) Less of all alarms.
l l l l l SITE EERGENCY l TNERAL EERGEtCY l i I I IILES T OFFSIIE AND WSIIE AC POWER FWI l l 14NTR THAN 15 MINUIES I I I I l l 1) Turbine generator trip with SU2 & I l l SU3 unavailable for service for l N/A l l 1cnger than 15 minutes ard l I I I 1 2) Failure of all diesel generators l l , ' I to energize their busses for l l l 1cnger than 15 mirantes.
I SIIE DER 2NCY l l l (Ctntinued) l l Ims & ALL 125 VDC POER FCR LWGER l ILES T ALL 125 VDC POER FCR IG2R l l nmN 15 MINUIES I 'IHAN 15 MINUIES l l 1) Imss than 105V cn the 2/3A,B,C&D l distributicn panels as indicates l 2) Imss than 21V en the 24V distribu-cn panels 2/3AD03,2/3CD03,2/3 BID 3,I ticn panels as indimted cn panels l 2/3I003 for lcnger than 15 min andl 2/3AD28, 2/3CD28, 2/3BD28, 2/3Du28 . l I for 1cnger than 15 min. and I i l l l > l l 3) Ioss of all alarms for icnger thanl l l 15 minutes.
l ' I
y-APPENDIX EP-101-1 - Page 9 of 13, Rev.15 I I I SEXDNIAE CXNTAIl@ENT I I I I I i l twustRL EVENT l AIERP l I I I I l l lIIES T SEGEIAE QNTADDENT IME@ITY l l l l 1) Ioss of seccniary contair1 ment l l irtegrity for greater than 12 l hours.
I I I I I I -
I I I I N/A l l l l l l l l l l l l l l l l l l
l l l l l
I I I l l
I I I I I I I i l l SHE DERINCY I TNERAL DERGE:NCY I I I I I i l i I I I I I I I I I I I I I l l I I I I I I I I I N/A I N/A l l l l ' l I I I I I I I I I I I I I I I I I I I I I I I t i I I l l l l l
. , . - - -., - ..,. - ... _,, - -. _ - - - - - - -, _. - - ,__ ..., -
7.. . , ,_ . APPDIDIX EP-101-1 ., Page 10 of 13, Rev. 15 I i l INSTRDENT FAILlRE I I I I i l I tsu5tRL EVENT I AIERr I i l i l l SICNIFICANT LOSS OF ASSESSIENT OR I l l cxmouurcxrION CAPABILrrY IN WE MIN I l 1 CONT 10L IOOM I I 1) Cbmplete loss of all main ccntrol l l l zoom commmicaticn equipnent.
I I I I I ' I I I I I N/A l l l l l
1 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I l l SITE EMERGENCY I GENERAL DERGDCY I l l l l l l l l l l l l
I I I I I l l l l ' I I I l l l l l l l N/A I N/A I l l l l l l l
I l I I I I I I l i l I I I I I ' ( l I I I I I I a . - - - - - - ,,.. -... -., -. - -. _, _. -..,,,. - - -, -,., - -..,., -.. _, - ,.-, - -, _,., _, . -.. - - -, -.,. . _, _ _, - - -,. -,. -.- -
- _ _ _ - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . .., APPENDIX EP-101-1 Page 11 of 13, Rev. 15 f I l l stEL DURE l l l l l l l INUSIAL EVENT l ALERP l l l l l l i i rossIBrz ru:L oeRa l ru:L ouRm i I I I l 1) Air ejector discharge ral mcnitor i 1) Air ejector discharge rad monitor I l hi@ alarm and an increase of 500 l indicating greater than 2.5X10M l
mR/hr withiii TO minutes or a levell mR/hr en RR-2/3-17-152, or li of 2.5X10P3 mR/hr as indImted on l - I BR-2/3-17-152,or l 2) High coolant activity of 300 (Ci/ ' -- l cyn cbse equivalent I-131, and main! l 2) Hich reactor coolant activity as stehm line hi-hi radiaticn alarm I determined by sample analysis with resultant scram alarm, or I equal to or greater than 4 tCi/gm l l ~ l cbse equivalent I-131.
l 3) Spent fuel damage resulting in a l l refueling floor area radiaticn l I mcnitor alarm or a high radiaticn l . l l alarm cn refuel floor exhaust rad l l l nmitor, or l l I l -- l l 4) At 3 east 2 of the 4 ccntainment l l l rad mcnitors indimte levels l l greater than 4X10P2 F/hr cn RI4/ l l 9103 A/C & RI-8/9103 B/D.
I i SITE DERGENCY l GENERAL DERGENCY l i I I i i I l FUEL DMAGE I FUEL DMRT l l l l l 1) At least 2 of the 4 ccntainment l 1) When at least 2 of 4 cmtairunent I rad acnitors indimte levels I rai mcnitors indimte levels l greater than 4X10P3 R/hr cn I greater than 4X10M R/hr cn l RI-e/9103A/C and RI-8/9103B/D, or l RI-8/9103A/C and RI4/9103B/D, and I ~l ccmtairunent pressure exceeds 10 l l 2) Major damage to spent fuel in fuell psig cn PR 2/3508.
l pool or uncovering of spent fuel l I as cmfirmal by a fuel pool area I raiiaticn mcnitor alarm ands l
I I l a) Refuel floor exhaust radiaticn l l l nmitor hi@ alarm, or 'l
I b) Refuel floor area ra3Taticn
mcnitor alarm, or-- l l I l 3) Observed major damage to spent l l , l fuel.
l l l l l l l l _ _ _ _ _ _ _ _ _
r . .. APPENDIX EP-101-1 Page 12 of 13, Rev. 15 - I ' I I I mZAEs l I I I I l l tmstE EVENT AIL RT I I i l l MIEMTE HAZAES SPERE MZAES I I - l 1) Aircraft crash cn or near site as i 1) Aircraft crash at the facility or l I determined by Shif t Supervisicn or missile impacts into the Reactor I I Bldg., Diesel Generator Bldg., or I I 2) Significant explosicn cn or near I HPSW pump structure as determined I l site as determined by Shif t I by Shif t Supervisiat, or I l Supervisicn, or l l l l 2) Explosicn damage to facility l 3) 'Ibxic gas release cn or near site l affecting plant safety as deter-
I as determined by Shif t SupervisicnI mired by Shif t Supervisicn, or I I l 3) Chlorine gas detected in the l l centrol room.
I I I I I I I I I I I I I I I I I I I I I I I i l l SrE DER 2NCY I 2NERAL DERENCY l l l l l l l l l l l l l l l l l l l l l l l l l
I I I I I I N/A I N/A l l l l l l l l l
1 I I I I I I I I I I l I i l l I ( l I I I I I I I l ,
APPENDIX EP-101-1 ' Page 13 of 13, Rev. 15 . l l l CONraDL acoM EVACUNTICN I l i I i l I tNLELAL EVENT l AIERP l l I i l l I REWVE CINTROL NPABLISHED I l l l l l 1) Evacuaticn of main control room l l l anticipated or required and l l ccntrol established at remote l l sNtcbwn panels as determined by l Shif t Supervisicn.
I N/A I I I I
c , I I I I I l-
I I I I I I I I I I I I I I I I I I I I I I I I i l l SITE EERGEICI I GENERAL EERGEICY l l l l I i l l REMME ONrmL ICT ESTABLISHED l l l l l l 1) Evacuaticn of main control rotzn l l l and ccntrol of sNtcbwn systems l l l not established at remote shutdownl l I panels in 15 minutes as determinedl l l by Shif t Supervisicn.
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. . . - - - . .. . PBAPS _.. TABLE OF CONTENTS Page No.
, . 1.0 DEFINITIONS
LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1
1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2
' SURVEILLANCE LIMITING CONDITIONS POR OPERATION REQOIREMENTS , 3.0 APPLICABILITY 4.1
, 3.1 REACTOR PROTECTION SYSTEM 4.1
3.2 PROTECTIVE INSTRUMENTATION 4.2
,
- 3.3 REACTIVITY CONTROL 4.3
A.
Reactivity Limitations A
- B.
Control Rods B 101 C.
Scram Insertion Times C 103 l D.
Reactivity Anomalies D 105 , 3-4 STANDBY LIQUID CONTROL SYSTEM 4.4 115
tas . A.
Normal Operation . A_ 115 ir-3.
Operationtwith Inoperable Components D'- 116 , C.
Sodium Pentaborate Solution C 117
3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 124 , A.
Core Spray and LPCI Subsystems A 124 , [ B.
Containment Cooling Subsystem (NPSW) B 127 C.
EPCI Subsystem C 128 D.
RCIC Subsystem D 130 E.
Automatic Pressure Relief Subsystem E 131 F.
Minimum Low Pressure Cooling System F 132 l Diesel Generator Availability l G.
Maintenance of Filled Discharge Pipe G 133 l R.
Engineered Safeguards Compartments B ,133 Cooling and ventilations , I.
Average Planar LEGR I 133a t J.
Local LNGR J 133a - K.
Minimum Critical Power Ration (MCPR) K 133b l l l l Amendment No. 104 /108 _g, 2/7/85 ., . . -m., .- _. - -- _ - -.... _ _ _, _.. _ - _ -. - - - - - - - - - , _ ___...m,m , - ~.
, . . _. . . . -.. PBAPS TABLE OF CONTENTS (Cont'd) = ,: Paae . - SURVEILLANCE LIMITING CONDITIONS FOR OPERATION.
, , REQUIREMEN"5 3.6 PRIMARY SYSTEM BOUNDARY 4.6 143 A.
Thermal and Pressurization Limitations A 143 B.
Coolant Chemistry B 145 C.
Coolant Leakage C 146 D.
Safety and Relief Valves D 147 E.
Jet Pumps E 148 F.
Recirculation Pumps F 149 G.
Structural Integrity G 149 ' 3.7 CONTAINMENT SYSTEMS 4.7 165 A.
' Primary containment ' A 165 B.
Standby Gas Treatment System B 175 C.
Secondary Containment C 176 D.
Primary Containment Isolation valves D 177 3.8 RADIOACTIVE MATERIALS 4.8 203 . A.
General A 203 i R.
Liquid Effluents B 204 C.
Gaseous Effluents C 208 D.
40 CYR 190 D 216 E.
Radiological Environmental Monitoring-E 216a-2 F. ~ Solid Radioactive Waste ~ F 216a-5 G.
Mechanic ~al" Vacuum Pump G 216a-6 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 217 A.
Auxiliary Electrical Eculpment A 217 B.
Operation with Inoperable Equipment B 219 C.
Emergency Service Water System C 221 3.10 CORE 4.10 225 A.
Refueling Interlocks A 225 B.
Core Monitoring B 227
C.
Spent Fuel Pool Water Level C' 22Sa D.
Heavy Loads Over Spent Fuel D 228a 3.11 ADDITIONAL SAFETY RELATED PLANT CAPARILITIES 4.11 233 A.
Main Control Room ventilation A 233 B.
Alternate Beat Sink Facility B 234 C.
Emergency Shutdown Control Panel C 234 D.
Shock Suppressors, D 234a . , Amendment No.17, if 102,
102/104, (Updated March 18, 1985 ) . ... _ _ _ _ _
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. . . - . .. . PSAPS . . . - ' ~
LIST OF TART " ' ' . ZaEa Zina tasa . , . . 3.1.1 Reactor Protection system (scram)
' 2astrumentation Reguirement
' ' . 4.1.1 maatter Protection System Escram)
Zastruent Func*i-1 Testa
' 4.1. 2 manctor Protection system (scram)
2astrument calibratica'
. 3. 2. A Zhstrumentation That Zaitiates Primary
. .
containment Isolation , i 3. 2. 5 2nstrumentation That Initiates or coctrols
L the core, and containment cooling systems ' 3. 2. c Instrumentation That Zaitiates control
mod Blocks - 3.2.D Badiation Ilonitoring Systems That Zaitiata
and/or Isolates Systems - , i . 3.1 F tary=111==ce Zastrumentation
1- ' 3. 2. 0 2astrumentation That' 2mitiates Recircu2ation 79 . ... - --
Pump Tzip . 4. 2. A , stLaimum Test and calibratica Frequency
f.or PcIS - . . . . . .. . - . . t . . . . , >
Amendment No. J4.,M, 112/116 ,. 11/19/85 . eo e . ,wem -- - - wwww-w--mpy .py y,,,- - w-w-, --w------- - ---n
._ . PBAPS . LIST OF TABLES Table Title Pace
4.2.B Minimum Test and Calibration Frequency
for CSCS . . 4.2.c Minimum Test and Calibration Frecuency
for Control Rod Blocks Actuation 4.2.D Minimum Test and Calibration Frequency
for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frequency
for Drywell Leak Detection 4.2.F Minimum Test and Calibration Frequency
.for Surveillance Instrumentation 4.2.G ' Minimum Test and Calibration Frequency
for Recirculation Pump Trip 3.5.R.2 Operating Limit MCPR values for 133d Various Core Exposures 3.5.R.3 Operating Limit MCPR Values for Various Core Exposures 133e 4.6.1 In-Service Inspection Program for Peach 150 Bottom Units 2 and 3 3.7.1 , Primary Containment Isolation Valves 179 3.7.2 Testable. Penetrations With Double 184' O-Ring Seals e 3.7.3 Testable Penetrations with Testable 184 Bellows 3.7.4 Primary Containment Testable Isolation 185 Valves 4.8.1 Radioactive Liquid Waste Sampling and 210 Analysis 4.8.2 Radioactive Gaseous Waste Sampling and 211 Analysis 3.ll.D.1 Safety Related Shock Suppressors 234d 3.14.C.1 Fire Detectors 240m 3.15 Seismic Monitoring Instrumentation 240u 4.15 Seismic Monitoring Instrumentation 240v Surveillance Requirements Amencment No. 102/104-Vi~ (Updated March 18, 1985) m ::. _ _ _ _ _ _ _ _ _ _ - - - - - - --- '-
. - . . PEAPS - LIST OF TABLES - - -- , , Table Title Pace . . ,. . .- . ~ 4.3.3.c Maximus Values for Minimum Detectable 2IEd-4 - Levels of Activity . 3.14.C.1 Fire Detectors 240s ' 3.15 seismic Monitoring Instrumentat' ion 240u , . 3.15 seismic Monitoring Instrumentation Surveillance 240v Requirements . . . . . I . . . . .
- / - . . \\
. . ... .. . . . . . , . Amendment No. YU7, IDC,107 /111 vii 3/19/85 , ,, - . q . . . - . S . e e > _ _,. _ -.. - _ _, _ _ _,. ,-._g___.,-_-...n,_,_...,.__,,._ , _.,. -.,, - - - _ - - .. - -. _. _. _,,.. _.,. _, _ _. _ _. _. _ _,,.,. _ ,,., _ _, _ - -,, _ _,, _,,,, _ _. _ _, _ - _ ,
en e e e THIS PAGE LEPT BLANK INTENTIONALLY ) . - - . - . . -. . - - - - -. ..- - .... - . - - - - - .---. - - . - - --
.. . _ J '~" iLA.III - LIMITIbG CONDITIONS FOR OPfETINI APPLICAlilLITY i 3.0.C This specification delinu...... the action to be taken for circumstances not direct 1 Muvwed for in the acticn statements and.4.vse occuacnce would violate the intent of the specification.
For example, a specificatica may require.two outsystems to he operable and provides explicit action requirements if one suhaystem is incperatie.
Under the terms of specification 3.0.c, if both of the required suksystems are inoperable, and an actica requirement is not identified in the specificaticns, then the unit is to be in at least not . Shutdown within 6 hours 'and in cold shutdown within 36 ho urs.
. .[.0.D This specification delineates what additional ccnditions ,' must be satisfied to permit operation to ccatinuei consistent with the action statements for power sources, i when a normal or emergency power source is not cperable.
! It specifically prohibits operatica when one divisicn is -, incperatie because its normal or emergency power source is incperatie and a system, sutsystem, train, ccupenent or device in anctber division is inoperatie for ancther reascn.
The provisions of this specification permu. the action statements asscciated with individual systems, subsystems, trains, ccaponents or devices to he consistent with the acticn statements of the associated electrical power source.
It allows operation to be governed by the time limits cf the action statement associated with the Limitine. Condition for Cperation for the normal or emergen4 power source, not the individual actica statements for each system, subsystem, train, ccaponent or device that is determined to he inoperable solely because of the inoperability of its normal cr emergency scwer source.
For exasple, Specification 3.5.7.1 proyides for an cut-of-service time when one of the four diesel generators is nce operable.
If the definition of operable were applied withcut consideration of specification 3.0.D, all systems, subsystems, trains, components, and devices supplied by the anose.rahle emergency power source would alsc he incperatie.
This would dictate invcking the applicatie l acticn statements for each cf the applicatie I,1r.iting , Conditicas for operation.
Ecwever, the provisiens cf i specificatica 3.0.D permit the time limits f cr continued , operaticn to he consistent with the action stacament for the inoperatie emergency diesel, generator instead, prcvided the other specified conditions are satisfied.
. Mendnunt ho,71 /Go 3us-l , . .
-. . ".TSES - 7.IMITIFG CONDITIONS FOR 012lf.TJpft,AM n...;... Iranf'd).
, , In this I::ase*, this would ..W the corresponding . normal power sr.- .t must .....ii:1 c (as must be the . .. corpenents supp i e ed by th . normal pcwer scurce) and all redundant systems, subsy: items, trains, cosponents an,d , devices * in the other divisicn must he operable > or likewise satisfy Specification 3.0.D (i.e., be capable of performing their design functions and have an esorgency power scurce operatie). If these conditicns are not satisfied, shutdown is required in accordance with this specification.
- In the cold shutdown condition and ref uel mode, - Specificaticn 3.O.D is not applicable; and thus, the individual action statements for each applicatie Limiting conditicn for Cperation in these conditions must he adhered to.
. . r* .. >. > , . ,, ., . . . .
. . ( Amendx.ent No. 71 / Gb-34b-I ' -- .. . . - - - - -. - .
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r
. ... ... .... .. , -.. o . PBAPS
LIMITING CONDITIONS FOR 0?! RATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability: Applies to the instrumentation and Applies to the surveillance ossociated devices which initiate of the instrumentation and o. reactor scras.
associated devices which initiate reacto'r scraa.
Objective Objective , To assure the operability To specify the type and of the reactor protection frequency of surveillance system.
to be applied to the protection instrumentation.
ecification: Specifications a. When there is fuel in the vessel A.
Instrumentation systems the setpoint, minimus ausber shall be_,,f unctionally i of trip systems, and miniaua tested and calibrated number'of instrument' channels as indicated in Tables that must be operable for 4.1.1 and 4.1.2 . each position of the reactor respectively., j mode switch shall be as given ~ in Table 3.1.1.
i B.
The designed system response 5.
Daily, dur. 7, reactor power times from the opening of the operation, the maximum fractii: sensor contact up to and of limiting power density including the opening of the shall be checked and the t rip ac tuator contacts shall scram and AFRM rod block g not exceed 50 milliseconds.
settings given by equations i Otherwise, the affected trip in Specification 2.1.A.1 systen shall be placed in and 2.1.3 shall be calcu-the tripped condition, or laced if the maximum ' the action listed in Table fraction of the limiting 3.1.1 for the specific trip power density exceeds the function shall be taken.
fraction of rated power.
, i l i " Amendment No. J#, y X g, 76 /75
- l l
O
. 2840030300 - PBAPS _ . . LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2** One trip train * of the RPS 2** The following RPS alternate alternate power supply may power supply protective be in the bypassed or devices shall be functionally inoperative condition for tested at least once every six a period of 72 hours.
If this months and calibrated once condition cannot be satisfied, each refueling outage.
or if both trip trains are inoperative, the RPS bus shall be transferred to the RPS MG Acceptable set or de-energized within Device setting 30 minutes.
Undervoltage 113 t 2 Volts . Overvoltage 131 + 2 Volts Underfrequency 57 Hz t.2 Hz , . . I i
l l l l { ! ! A trip train consists of one breaker, one undervoltage relay, one
l overvoltage relay, one underfrequency relay, one time delay relay (MG set only), and the associated logic.
Effective upon installation of the protective trip devices.
. - 36a - Amendment No. 99/101 .r.r. 21, t on e.
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l ' . 2840030300 ' - . PBAPS . ... . LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS C. When it is determined that ' a channel has failed in the unsafe condit' ion, the other RPS channels that monitor the same variable shal'1 be functionally tested immediately before the trip - system containing the failure is tripped.
The trip system containing the unsafe failure may be placed in the untripped . condition during the period in which surveillance testing is saing performed on the othG RPS channels.
The. trip systeR may be in the untripped positi for no more that eight hours
. per functional trip period ! for this testing.
D.
Reactor Protection System D.
Reactor Protection System - Power Supply' Power Supply i 1** Reactor Protection System 1 **The following RPS' power devices shall b' protective supply (MG set)- Power Supplyr s functionally tested at least once every One trip train * per RPS MG set six months and calibrated may be in the bypassed or once each refueling outage.
inoperative condition for a . period of 72 hours.
If this . . Acceptable condition cannot be satisfied, or if both trip trains are Device setting inoperative, the RPS bus shall be transferred to th'e alternate Undervoltage 113 + 2 Volts source or de-energized within Overvoltage 131 + 2 Vol .30 minutes.
Underfrequency 57 Hz + .2 H: Underfrequency Time Delay 6 sec + 1 sec . . -36-Amendmant No. 15, 99/101 June '1, 1984 .
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ __ -_ - _-. __ - r ~ ~ . g T4 SLO O.l.1 me N PROTECTION BYSTEM (SCRAM) INSTRUMENTATICN REQUIRMENT - , , Minimuss geo.
Modes in Which leuerber of
k Cf Operable Penction Must De Instrument Instrument Trip Level Cpgrable Channels Action e.
- Channels Trip Function Setting a Provided (1) M. per Trip petual startup een Dy Design , Cystem (1) (7)
Mode Switch In X X A 1 Mode Switch A Shutdown ~ (4 Sections) ,ym
Manual scram X X X 2 Instrument A Channels t
IBM High Flux $120/125 of Full X X (5) e Instrument A Scale Caannels e
IpH Inoperative X X (5) 8 Instrument A g Chsanels
t
AyMM liigh Flux (. 66W e 54 -0.66&M) X 6 Instrument A or B FRP/MFLPD Channels (12) (13) ss r
APRM Inoperative (11) X X X 6 Instrument A or a Channels (10; 6 InstrusseInt A or B
APNM Downscale 22.5 Indicated - on scale Cnannels ' o
2 APNM High Flux $155 Power X X 6 Instrument A - ' in startup Channels [ - " '
High Reactor $1055 psig X (9) X X 4 Instrument A Pressure
Channal:s
High Drywell $2 Psig X (8) X (8) X 4 Instrument A Pressure Channels
Reactor Low 20 in. Indicated X X X 4 Instrument A water Leve1 Leve1 Charmels ,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ - _ _ - . .__ - __ ._.
_ .- __ . <, U"I" 2, g . '5[ Table 3.1.1 (Cont'd) ' REACTOR PROFECTION STSTEM (SCRAM) INSTRUDENFFATION REQUIRE 3ENT - ' g Mininom No.
Modes in Witch Neueser.
of operable Function Must be Instr ~
- U Instrument Trip Level Operable Channele* Action b Channels Trip Function Setting Provided. (1) , r Tri Refuel Startup Resn by Design , j ,- yotes 1) (7) ' i D
A
Migh teater tevel <50 callone X(2) X X 4 Instrument A '
in Scram Discharge Channele Instrument Volume ,
Turbine Condenser 523 in. Ng.
X 4 Instrument A or C l Low Vacuum Vacuum
Channels - I }
-
Main Steam Line 43 X Normal Full X X .X 4 InstrWiset Ay - Righ Radiation b r Background Channels-
m., a.d , !
Main Steam Line 410% Valve j X(6).
S IsistrGeet A, l l Isolation Valve E1osere Channels- ~ Closure !
t ' Cf2tf
j
Turbine control 500tPcS50 peig a X(4) 4 2d A or D ' ' Valve Fast Closure Control Oil Free-Channete(; oure Between Fast t
Closure Solenoid
and Dime Dump N' ' ' Valve U
g T
'.f) ; , .x. -
Turbine Stop <10% Valve X(4) S Inst t A or D
' Valve Closure Elosure Channel {; %4 ' ' - N,f. $ ' - --
._ - _ = =. . . . . - - . _ _ _ _ _ _ _ _ . Unit 2 . . PBAPS , ' NCFTES FOR TAB 1.E 3.1.1 7i~ Thereshallbetwooperab1k"[E'trippedtripsystemsforeach
"./.Q*Vy ' , 1.
funetton.
If the minimum number of operable sensor W annels for a trip system eennot be met, the affected tri tem shall be placed in the safe (tri N ) oogditten,'p '" , appropriate actions listed below shall he'taken.
' , - g e A.
Initiate insertion of operable r'ods and eemplete@ insertion of all operable reeeMrithin four bedsL T v.;r s - .a:= m~ B.
Reduce power level te) IRM range and place mode switch in ' the start up position within a bours.
C.
Reduce turbine load and eldes main steam line isolation valves within 8 hours..,3 j '4,,,,
- _,,, D.
Reduce power to less than 200 rated.
2.
Permissible to bypass, in refuel.and shutdown positions of j the reactor mode switch.
3.
nei.t.d.
I 4.
Bypassed when turbine first stage pressure is less than 220 psig or less than 30% of rated.
5.
IRM's are bypassed seen APRM's are onseale and the reactor mode switch is in the run position.
- . ., . 6.
The desien permits'elosure of any tuo lines without a scram
being initiated.
'& 9 r A.- ~ . 7.
When the reactor is suberitical and the reactor water temperature is lees than 212 degrees F, only the following trip functions need to be operable: A.
Mode switch in shutdown B.
Manual scram C.
Eigh flum IRM D.
Scran discharge instrument volume high level S.
Not required to be operable when primary containment integrity is not required.
9.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
Amendment No.gg117/121 3/14/86 -39-l ' I . - - - - -, ,,_---n,__.._,,. . -,,,.,. - -.,, _ _ _,,.,., _.,,.. _,,,.,,. -,,.. -, _,,,,, _,.. _ _ _.... _ _ _ _ _,, -,,.... -,,,.,,.,,,.,,, - -, _. -. _,. _... - -, _. _ - -. _ , _, _
l - . Undt 2 . rDAPS
- iOTES FOR TABLE 3.1.1 (Cont'd)
- - ,. 10.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high.
, 11.' An APRM will 'be considered opera'ble Jf..there are at least 2 r Jeve1 and.:at Jeast'.)6..LPRO.16 puts,pf W,he3gwa.Wiq LPRM inputa
!
- '
normal compi nt.
"?.. vi4; ,1. . 12.
This equation will.be used in the event 'of operation with a maximum fraction of limiting power density (MFLPD) gecater" than the fraction of cated power (FRP), where: . r.
~ _ FRP = fraction of rated thermal power (3293 MWt).
MFLPD = maximum fraction of limit;ing Mwer density where the limiting power densityTs*' ' ***-" ^"" 13.4 KW/ft for all 8x8 ',. fuel.
The ratio of FRP to MFLPD shall be set' equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
- W= Loop Recirculation flow in percent of design.
W is 100 for core flow of 102.5 million Ib/hr or greater.
,
AW= the difference'betwi.cn two loop and single loop effective rectreulation drive flow rate at the same , core flow.
During single loop operation, the - reduction in trip setting (-0.66 AW) is accomplished byecorrecting-the flow inputa eft the-flow biased High - -- Flux trip setting to preserve the original (two loop) relationship between APRM High Flux setpoint and recirculation drive flow <'or by adjusting the APRM Flux trip setting.* W = 0 for two loop operation.
, Trip level setting is in percent of rated power (3293 MWt).
13.
See Section 2.1.A.1.
, o Amendment Ho.
79-40- . . _ .-- -- .
_ - - - . . 13 I; r _ .. i "ill is !1 1.
& !!s g ~. r., , - . (. o
3I ~ 31* i
r 1r _i 1
- 3=1 n18 8.:jj.r.
l3 E - aul 181 -
- g..,.:5. j
, . :.s.n.
! -j u I::.eI = I a sgsialis MI s - 31. !i11 ml:ill I a . .:.:.: hi
. .: I*,, !
- l]r ll l
l-l - .
- - . . . . . l
. -
s .= 21.
- s a -
3 . . . I E ' ,
. . i =
- ~
i, E
3
es zl 3- . - - l 4:2 . .
- -
- - . 8.
. ll - -
- .
.I, ,,5 5ii i.. Il ".
- -
(- i,
. . i it - ei , I , - ge a .a
. - y_ s3
. -[, . 'J l
.!!
,i.
! a
sa
a
p :i s si si ii " al
.- t
i ip5 h: e ,,
' 233 s - . Amendment No.,Inf. 83 /82 - 65 - 9/82 . - _ _ _ _ _ _ - - - _ _ _. _ _ _ _ _ _ _
._ _ _ _ _ . . e T,ABLE 3.2.D (Cont'd) . - - INSTHUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINNENT COOLING SYSTENS Minimua No.
cf Operable Number of Instra-Instrument Trip Ebnction Trip Level Setting * ment Channels Pro Remarks ,
Chaznals Per vided by Design - ' . Trip Systemf1) '
Reactor Low 400-500 psig 4 Inst. Channels Permissive for opening
Pressure Core Spray and LPCI Admission valves.
Coincident with high dry well pressure, starts LPCI and Core Spray pumps.
Reactor Inw 200-250 psig* 4 Inst. Channels Permissive for closing Pressure
Recirculating Pump , Discharge Valve.
k
Reactor Low 505PS75 psig, 2 Inst. Channels In conjunction with PCI . e Pressure signal permits closure of RHR (LPCI) injection , , , valves.
, ~ e +
L - - . . h. '.
$
I A=ndment No.
68/67 (5/5/80) ' n , i .
_ , TABLE 3.2.B - , g INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT . COOLING SYSTEMS , E ' 2 Minimum No.
- eof Operable Number of Instru-Iestrument Trip Function Trip Level Setting ment Channels Pro-Remarks zPChannels Per vided by Design
. wTrip System (l) , . . ." w - !*
Core Spray Pump 6 +/- 1 sec 4 timers In conjunction with loss Start Timer 10 +/- 1 sec 4 timers of power initiates the _,g starting of CSCS pumps.
I
LPCI Pump Start Timer 5 +/- 1 sec 4 timers o (Two Pumps)
ADS Actuation Timer 90 </= t </= 120 2 timers In conjunction with seconds Low Reactor Water Level, High Drywell Pressure i and LPCI or Core Spray !) Pump running interlock, t initiates ADS.
ADS' Bypass Timer * 8 </= t </= 10 4 timers In conjunction with , minutes low reactor water level, bypasses high.drywell pressure initiation of ADS.
RHR (LPCI) Pump 50 +/- 10 pelg 4 channels Defers ADS actuation Discharge Pressure pending confirmation of.
Interlock Low Pressure Core Cooltng system operation . (LPCI Pump running interlock).
Core Spray Pump 185 +/- 10 psig 4 channels Defers ADS actuation l Discharge Pressure pending confirmation of ! Interlock - Low Pressure Core cooling ) system operation (Core Spray Pump ' running interlock).
- l-:f fective when modi fication associated with this amendment is omob t e.
..l , .. . _ _ _ _ _ _ _
TABLE 3.2.B (Cont'd) ~ INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.
of Operable . Number of Instru- .
- Instrument Trip Function Trip Level Setting ment Channels Pro-
'Chann21s Per vided by Design "## *
' Trip System (l) . ,
RHR (LPCI) Trip NA 2 Inst. Channels Monitors availability System bus power l monitor of power to. logic - .. i systems.
'l Core Spray Trip NA 2 Inst. Channels Monitors availability -
System bus power of power to logic ' monitor - systems.
ADS Trip System bus NA 3 Inst. Channels Monitors availability .
- power monitor of power to logic t
systems.
. i a
- HPCI Trip System bus NA 2 Inst. Channels Monitors availability
$ power monitor s of power to logic ' systems.
, l
RCIC Trip System bus NA 2 Inst. Channels Monitors availability ~
power monitor j of power to logic systems.
l l l . _ _ l- . - I k AM 47 october 10, 1978 -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
- s t
S TABLE 3.2.8 (CONTINUED) INSTRUNIBITATICIS TRIAT INITIATES OR CONTItOLS TflE CORE AND CONTAINNENT . COOLIIIG SYSTEMS ." ' RGilmun No.
' w3 of Operable Number of Instru- ~
! Instrument Trip Punction Trip Level Setting ment Channels Pro-Remarks l U Channels Per vided by Design Q Trip System , [
_ 1 (1) . M Core Spray Sparger 1 (plus or minus 2 Inst. Channels Alare to detect core l -
to Reactor Pressure 1.5) paid @ Vessel d/p spray sparger pipe { break.
(D l O 2 (1) Condensate Storage Greater than or 2 Inst. Channels Provides interlock tes l l , Tank I,ow Level equal to 5' above HPCI pump suction l
tank bottom , valves.
s 2 (U Suppression Chamber Less than or 2 Inst. Channels Transfers IIPCI pump I High Level equal to 5" above suction to ! torus midpoint suppression chamber.
2 (6) Condensate Storage Greater than or 2 Inst. Channels Transfer RCIC pump Tank Low Level equal to 5' above tank bottom suction to suppressia.se chamber.
. e e
.. . . .
.__ _ -. _ _ . . e . TABLE 1.2.B (COllTINUED) i l lilSTRUMEllTATIOll TilAT INITIATES OR CONTROLS Tile COHE AND COtlTAINHENT COOLIllC SYSTEMS , ' - I l Hinimun No.
Of Operable Number of Instru-Inntiument Trip Function Trip Level Setting ment Channels Pro-Remarks ' , Cimnnels Pe r vided by Design ! Trip System (1) . ] . g .
. _ _ _ _ _ _. l i
RCIC Turbine High 1450" H O(2) 2 Inst. Channels Flow ,2
RCIC Turbine lijgh 3 1 1 sebonds 2 Inst. Channels Plow Time Delay ,
2 RCIC Turbine Com-1200 deg. F (2) 4 Inst. ) ' o partment Wall s 116 Inst.
I
RCIC Steam Line 1200 deg. F 'al(2) 12 Inst. ] , Area Temp.
RCIC Steam Line 100) p> 50 psig (2) 4 Inst.
' Low Pressure
IIPCI Turbine Steam .1225".N O (3) 2 Inst. Channels .
Line lijgh Flow i i i llPCI Turbine High 3 1 1-seconds 2 Inst. Channels Flow Time Delay !
i ! t a . . %endment tio. 100/102 .1uly 7, l'"L1 - , - - . - -
_- _- - -, - - - _ _ _ - _ - - . .- . . . . . TABLE 3.2.B (CONTINUED) INSTRUMENTATION THAT INITIATES OIt CONTROLS TIIE CORE AND CONTAINHENT COOLING SYSTEMS Minimum No.
Cf Operable Number of Instrument Instrument Trip Function Trip Level Setting Channels Remarks . , Chhnnels Per l Trip System (1) Provided by
Design ' . 4(5) HPCI Steam Line 100>p>50 psig (3) 4 Inst.
Low Pressure
HPCI Turbine 1200 deg.F (3) 4 Inst.)
Compartment ) Temperature .) )
HPCI Steam Line <200 deg.F (3) 8 Inst.) 16 Inst.
' -
Area Temperature
- ) ) a ) - l Y
HPCI/RHR Valve $200 deg.F (3) 4 Inst.)
' Station Area < Temperature ~
LPCI Cross-Connect NA 1 Inst.. Initiates annun-Position , clation when valve ! is not closed.
1 per 4KV 4KV Emergency Bus ! Bus Undervoltage Relay, 251(+51)of Rated . Voltage 1. Trips all loaded ) (HGA) breakers.
, 2. Fast transfer l permissive.
3. Dead bus start of diesel.
I 1 per 4KV 4KV Emergency nus 951(+01.-104)of nun sequential Loading Rated Voltage Permits sequential Relay (SV) starting of - vital loads ~ .
I h m ro f...... f th 07 OO ,;} ii iqqq
_- _ _ _ _ _ _ _ _ . __ __.
- _ _ J . TABLE 3.2.B (CONTINUED)
INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.
Number of of operable Instrument Instrument Trip Function Trip Level, Setting Channels. Remarks Channels Per Provided by - vrlp System (1) Design , , . 2 per 4 KV Emergency 601(+51)of Rated 1. Trips emergency Bus Transformer Voltage. Test at transfer feed Undervoltage(IAV) zero volts in 1.8 to 4KV emer- . (Inverse time-seconds (+101).
gency bus.
! voltage)
' , , 2. Fast transfer permissive.
, - 2 per 4 KV Emergency Trans-901(+21) of Bus former Degraded raEed, voltage i ~ voltage (ITE) . (Instantaneous)
' 60 second 1. Trips emergency .,y (iSt) time delay.
transformer feed
to 4 KV enggency bus.
2. Fast transfer - permissive.
. , 6 second(+54) 1. Trips emergency time deTay.
transformer feed ,
to 4 KV emer- , . gency bus.
I 2.Past transfer permissive.
! , '4 ! 3. safety injec-i tion signal required.
I i, - ._ !
,.,,,....i,,,,,,,ri,, 07 / oo .il 11 109.1 ~ -
. . . ....... .... . . . . - - - - ---- --.. _ - _ _ . TABLE 3.2.8 (CONTINUED) ' INSTRUMENTATION TilAT INITIATES OR CONTROf.S Tile CORE AND CONTAINMENT COOLING SYSTEMS Ministm flo.
Number of of Operable i Instrument Icotrument Trip Function Trip Leyerl Setting Channels Pesarks Channels Per '
, Provided by
Trip System (1) Design , . 2 per 4 KV Emergency Trans-R7t(+51) of 1. Trips emergency l Eus former Degraded Rated Voltage.
transformer feed voltage (Inverse Tests at 2940 to 4 KV emer-time - voltage).
vol ts in 30 seconds gency bus.
(CV-6) (+101) 2. Fast trannfer . ~ permissive.
, . f
J t.r ' t . . e d e
h
, . m /li: enefe....n f tin 4 7 /o' e 3,,ii i1, 1 < m.1 , i
- . . . . ' . . . PBAPS . .. !
Notes for Table 3.2.5 - . r.. ' , ' 1.
Whenever any CSCS subsystem is required by Section 3.5 to be operable, there shall be two operable trip systems.
If the first column cannot be met for one of the trip systems, that trip system shall be placed in the tripped condition or the reactor shall be placed in the Cold shutdown condition within 24 hours.
. 2.
Close isolation valves in RCIC subsystem.
- . 3.
Close isolation valves in HPCI subsystem.
4.
Instrument set point corresponds to 378 inches above vessel zero.
5.
HPCI has only one trip system for these sensors.
- . 6.
With the number of OPERAMLE channels less than required by the Minimum OPERABLE Channels per Trip System rectuirement, place at least one inoperabl.a channel in the tripped condition within one hour or declare the RCIC system inoperable.
e . Anenhent No. )#,113-72-113/117 3/19/86 . .
, J -. . . .. - - . - . -. .. _ _ - _ _ - _ - - - - - -. _ _. _ - - _
.. ..... . .. .. -. -. - _. _ _ _ _ _ _ _ _ _ - -- f . TABTF. "I.~).C b, INSTRIJMENTATION THA INITIATP.S CONTROL ROD Bt,0CKS $ Minimum No.
Instrument Trip tevel Setting Number of Instrument Action R of Operable S Instrument Channels Provided " i Channels Per by Design !? Trip Sygtem
. ,
APRM Upscale (Flow < ( 0. fme#+ 4 7-0. 666w ) x 6 Inst. Channels (10) M Blased) FRP ~
- *
_MFI,PD~ (2) )
- M l
APRM Upscale (Startup G..., Mode) ~412% 6 Inst. Channels (10) , .
APRM Downscale >?.5 indicated on 6 Inst. Channeln (10) . e scale s 1 (7) Rod Block Monitor <(0.66w+41-0.66Aw)x 2 Inst. Channels (1) s (Flow Minsed) FRP ~ $- - MFI,PD (2) 1 (7) Rod Block Monitor 12.4 in'dtcated on 2 Inst. Channels (1) .Downscale scale 1,
IRM Downscale ('t) 12.5 indicated on R Inst. Channels (10) i' scale
IRM Detector not in (R) R Inst. Channels (10) Startion Posi tion
IRM Unscala (108 itidicated on R Inst. Channels (10) ' scale \\ 2 (5) SRM Detector not in (4) 4 Inst. Channels (1) Startup Position 7 (%)(6) SRM Upscale <10 counts /sec.
4 Inst. Channels (1)
Scram Discharne 42% qallons 1 Inst. Channel (9) l Instrisment Volume filqh I.evel . . . .. .
' ' . .
- ,'
' PDAPS ' , NOTES FOR TAIL $,3.2.C , For the startup and run positions of the Reactor Modo 1.
Selector Switch, there c. hall be two operable or tripped trip systems.for each function.
The SRM and IRH blocks need not be operable in "Run" modo, and the APRM and RBM rod blocks need not be operable in "Startup" mode.
If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily there'after; if this condition lasts If the longer than seven days, the system shall be tripped.
first column cannot be uet for both trip systems, tho - systems shall be tripped.
This equation will be uned in the event of operation with a ,2.
maximum fraction of limiting power density (HFLPD) greater than the fraction of rated power (FRP) whercs . . , FRP = fraction of rated thermal power (3293 NWt) MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for all '. 8x8 fuel.
. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of - 1.0, in which case the actual. operating value will be used.
_ Loop Recirculat' ion flow-in percent' of desfgn., W is - W= - 100 for core flow of 102.5 million Ib/hr or greater.
Tr,ip level setting is in percent of rated power (3293 MWt).
45W is the difference between two loopland single loop effective recirculation drive flow rate at the same core f flow.
During single loop operation, the reduction in trip setting (-0.66A W) is accomplished by correcting the flow input of the flow biased Rod Block Monitor (RBM) to preserve the original (two loop) relationship between the RBM setpoint and recirculation drive flow, or by adjusting the RSM setting.
W = 0 for two loop operation.
3.
IRM downscale is bypassed when it is on its lowest range.
' 4.
This function is bypassed when the count rate is 2 100 cps.
-
. !
5.
One of the four SRM inputs may be bypassed.
. ! I 6.
This SRM function is bypassed when the IRM range switches - are on range 8 or above.
, . l 7.
The trip is bypassed when the reactor power is S 30S.. - l 8.
This function is bypassed when the mode switch is placed in I I Run.
Am2ndment !!o. 7s/79 "', ~ -M~ _. _ _ _ - - _ _ _ _, _ - _ _ _ _ _ _. . _ . - -
. _ _ _._ l . . PBAPS ( _.. NOTES FOR TA8LE 3.2.C (Cont.)
9.
If the number of operable channels is less than required by the minimue operable channels per trip function requirement, place the inoperable channel in the tripped condition within one hour.
This note is applicable in the "Run" mode, "Startup" mode and " Refuel" mode if more than one control rod is withdrawn.
10.
For the Startup (for IRM rod block) and the Run (for APRM rod ) block) positions of the Beactor Mode Selector Switch and with the number of OPERABLE channels: a.
One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the ino~perable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the . Dext hour.
b.
Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
. . Amendment tio. 88,91/93 2-10-84 - 74a -
_ _ _ _ _ _ _ _ _ _ _ _ _. _ . _ _ _ _ _ _ _ _ - - - - - _ - - _ _. _ ______________ - TABIE 3. 2.D
- ,
RADIATION MONITORING SYSTEMS' THAT INITIATE AND/OR ISOLATE SYSTE95 '
.i ' Min 1:::um No. of . ) Operable No. of Instressent l Instrument Channels Provided ' Action
Channels Trip Function Trip Level Setting by Design (2) . ' ,
Refuel Area Exhaust Monitor Upscale, (16 ar/hr 4 Inst. Channels A or B l
Reactor Rullding Area Upscale, <16 ar/hr 4 Inst. Channels B Exhaust Monitors . l - I
- NOTE 3 FOR TABIE 3. 2.D
'l 1. Whenever the systems are required to be operable, there shall be two operable or tripped instrtmient channels per trip system. I f this cannot be met, the indicated action shall be taken.
2. Action l A. Cease operation of the refueling equipment.
B. Isolate secondary containment and start the standby gas treatment system.
! '
,,, ' Amendment No. 102/104 December 31, 1984
.
. e . _ . . Table 3.2.E Deleted .. ~ . w . , I I l
Amendment No. 112/116 11-19-85-76-l l .. l l
- *
, ! I
__ _ _ _ - _ _ _ ._ t n, ! . TABLE 3.2.F SURVEILLANCE INSTRUDENTATION k Minimue No.
f" of Operable i Instrument Type } Instrument Indication Channels ,. F and Range Action * . i .N
Reactor Water Level y
Recorder 0-60" y (narrow range) (1) (2) (3) !
Indicator 0-60* - . .
Reactor Water Level
(wide range) Recorder -165" to +50" . (10) (11) ! h
Reactor Water Level ! U (fuel zone) Recorder -325" to 0" s (10) (11) . w
Reactor Pressure } g Recorder 0-1500 pelg ! D Indicator 0-1200 psig (1) (2) (3) t 5s Y
Drywell Fressure Recorder 0-70 poig (1) (2) (3)
Drywell Pressure (wide range) Recorder 0-225 poig (0) (9) I
Drywell Pressure , (subatmospheric range) Recorder 5-25 pela (0) (9)
! Drywell Temperatere Recorder 0-400 degrees F (1) (2) (3) , Indicator 0-400 degrees F i
Suppression Chamber Water Temperature Recorder 30-230 degrees F (1) (2) (3) (6) Indicator 30-230 degrees F l
f
Suppression Chamber Water , ! Recorder 0-2 f t.
Level (narrow range) Indicator 0-2 ft.
(1) (2) (3) I \\ -. - -
_ -_ __ ___ _ _ _ _ _____-____ __ _ _____ _ _ _ _ -. . .. _ ..-_ _ _ . TABLE 3.2.F (Cont'd) S - SURVEILLANCE INSTRUMENTATION I Minimum No.
- S, of Operable " Instruoent a.
Type [ Channels
Instrument Indication and Range , " Action U._.
- i
[ Suppression Chamber Mater Level (wide range) Recorder 1-21 ft.
(10) (11) k
ti; Control Rod Position D 28 Volt Indicating ) Lights $
) (1) (2) (3) (4) Neutron Monitoring I SRM, IRM, LPRM 0-1004 ) '
) Safety-Relief Valve
Poeition Indicetlon Acoustic or Therinocouple (5) l w D
Drywell Nigh Range Radiation Recorder Monitors 1-1E(+3) R/hr (7) -
Main Stack High Range RedIatlon Monitot Recorder 1. 45 (-2) to 1.4E(+4)uCl/cc (7)
Reactor Building moof " Vent Nigh Range Radiation Recorder 1. 4E (-2) to 1.4E(+4)uCl/cc (7) Monitor
Drywell Nydrogen Concentration Analyser " Analyser and Recorder and Monitor 0-20% volume (1) (2) (3)
- Notes for Table 3.2.F appear on pages 70 and 78a.
. k - -_m.
- -
-_ __ .. . . . P8APS E9I.!!.M_I6!EI_hhl ( 1) From and after the date that one of these parameters is reduced to one indication, continued operation is permissible during the succeeding thirty days unless such lastrumentation is sooner made operable.
2) From and after the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding seven days unless such instrumentation is s Ener made operable.
i ' 3) If the requirements of notes (1) and (2) cannot be met, an orderly shutdown shall be initiated and the reactor shall be ': in a cold condition within 24 hours.
4) These surveillance instruments are considered to be redundant to each other, 5) If this parameter is not indicated in the control roca, either restore at least one channel to operable status within thirty days or be in at least Bot Shutdown within the next 12 hours.
6) A suppression chamber water temperature instrument channel , j will be considered operable if there are at least ten (10) resistance temperature detector inputs operable and no two (2) adjacent resistance temperature detector inputs are . inoperable.
7) With the number of operable channels less than the minimum number of instrument channels shown in Table 3.2.F, initiate the preplanned alternate method of monitoring the appropriate parameter within 72 hours andt a) either restore the inoperable channel (s) to operable status within 7 days or the event, or b) prepare and submit a Special Report to the Commission within 10 working days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to , operable status.
8) With the number of operable channels less than the minimum number of instrumentation channels shown in Table 3.2.F, continued operation is permissible during the succeeding thirty days, provided both Drywell Pressure instruments (0-70 psig) are operable otherwise, restore the inoperable channel to operable status within 7 da Eat Shutdown within the next 12 hours.ys or be in at least . ~75-Anendnent No. M. H. H, 113 {3719/86)
__ .
__ _ _.. _ _ _ _ - .. _ _ _. _ _ _ _ -. _ _ _. _ _ __ _ _. _ _ _ , PBAPS NOTES FOR TABLE 3.2.F (Cont'd) ... , 9) If no channels are operable, continued operation is permissible during the succeeding 7 days, provided both Drywell Pressure Instruments (0-70 psig) are operable; otherwise, restore the inoperable channel (s) to operable status within 48 hours or be in at least 50t Shutdown within the neat 12 hours.
i 10) With the number of operable channels less than the minimum ! number of instrumentation channels shown in Table 3.2.F, . continued operation is permissible during the succeeding 30 ' days, provided both narrow range instruments monitoring the same variable are operable; otherwise, restore the inoprable channel to oprable status within 7 days or be in at least Eat shutdown within the nest 12 hours.
11) If no channels are operable, continued operation is permissible during the succeeding seven days, provided both . narrow range instruments monitoring the same variable are operable; otherwise, restore the inoperable channel (s) ' to - operable status'within 48 hours or be in at least Rot shutdown within the nest 12 hours.
. ~ ' , n.-. J, - f'C 4 .q.
.g
..y
! ~ ,,,e .n.
r, l i , I . - Amendnent rio, 113/117 (3/19/86) ,7,,, . .. - -. - -., - -. - _... _ _, - - _ -.. _ -. y ,,,--,mm-- _., _ - _ - _ _ - _. - .. - _. _ - _ _ _ _.,. - - -. - - - - - - ,_
-_ _.. _ _ __ . - . .
i
- -
\\ l PBAPS '
! I . d - - C ' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT ~ ' 3. 5 CORE AND CONTAINMENT COOLING 4.5 I. CORE AND CONTAINMENT l $YSTEMS ' COOLING SY$TEMS
, Applicability: Applicability: Applies to the operational Applies to the Surveil- , status of the core and sup-lance Requirements of the pression pool cooling sub-core and suppression pool systems.
cooling subsystems which are required when the . , corresponding Limiting - Condition for operation is in effect.
Objective: Obiective: To assure the operability of To verify the operability the core and suppression of the core and suppres-pool cooling subsystems sion pool cooling subsys-under all conditions for . tems under all conditions which this cooling capabi-for which this cooling lity is an essential re-capability is an essential - sponse to plant abnormali-response to station abner-i, ties.
-- malities..- . . .) kpecification: ' ! Specification: . - A.
Core Spray and LPCI A.
Core Scray and LPCI Subsystems Subsyste=s 1.
Two independent Core Spray 1.
Core Spray Subsystem Subsystems (CSS) shall be Testing.
operable with each subsystem Item Frecuency comprised of: a.
(Two 50%) capacity centrifugal (a) Simulated Once/ Opera-pumps.
Automatic ting Cycle Actuation b.
An operable flow path test.
capable of taking suction . from the suppression pool (b) Pump OnceAonth at)d transferring the water Operability
to the spray sparger in the reactor vessel.
(c) Motor Once/ month Operated valve Operability k. ' I . - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- .-. _ - -- - . - - - _ _ . . - PBAPS . . ,
LIMITING CONDITIONS , FOR OPERATION SURVEILLANCE REQUIREMENTS
3.5.A Core Spray s 4.5.A Core Spray & . LPCI Subsystem (cont'd)..,, LPCI Subsystem (cont'd) 'l . Both CSS shall be operable Item Frequency whenever irradiated fuel is in the vessel and prior (d) Pump Flow Rate Once/3 month's to reactor startup from a , Cold shutdown condition
- Each Pump in each loop shall
. except as specified in deliver at least 3125 gym l 3.5.A.2 and 3.5.F.3 below: against a system head corres-
ponding to a reactor vessel pressure of 105 psig.
,, (e) Core Spray Header
AP Instrumentation - . Check Once/ day
- l Calibrate Once/3 months
(f) Operability In accordance check to ensure with 4.5. A.2, that pumps will 4.5.A.4 and start and motor 4.$.A.5.
operated inject-ion valves will open.
.: ,
. - 2. From and after the date 2. When'it 'is determined that one [_ - ' y . that one of the core core spray subsystem is~inoper-l spray subsystems is able, the operable core spray ' made or found to'be subsystem an'd the LPCI subeys-i inoperable for any reas-tems shall be demonstrated ' > on, continued reactor to be operable in accordance operation is permiss-with 4.5.A.1(f) and 4.5. A.3 (e) ible only during the within 24 hours and at least succeeding seven days once per 72 hours thereafter > provided that during until the inoperable core such seven days all spray subsystem is restored ~ active components of to operable status.
the other core spray subsystem and active 3. LPCI Subsystem Testing shall
- ~
components of the LPCI be as follows: . subsystem are operable.
' ' '
- Until the required
- - modification is completed, the loop flow rate test at - . 6250 gym against a system . head corresponding to a . reactor vessel pressure of - ' 105 psig will be performed ,,' , to satisfy surveillance "; . , requirements.
\\ -125- '. . Amendment No. 87, February 18, 1983 , --+-r __ _ -,_- _ ____ .. -. -- -.- -.. . - _, . .- , -,, - -. - -,. - - -, -. _,. -,,., _ -. . - - _,, - - - - _ -., -., -, - -
. - -. _.. . . ' . . PBAP.S . . .
- LIMITING CONDITIONS
+ FOR OPERATION SURVEILLANCE REQUIREMENTS
. - . . Item Frequency ~' (a) Simulated Automatic Once/ operating i Actuation Test Cycle (b) Pump operability Once/1 month - . .. . . - . ' . . . - . ( . . . . s .. -125a- . % Amendigent No. 87, February 18,.1983 , i I
PBAPS ( , LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.B containment Cooling 4.5.B Containment Cooling Subsystem (cont'd.)
Subsystem (cont'd.)
3.
From and after the date that 3.
When it is determined that any 3 HPSW pumps are made or any 3 HPSW pumps are in-found to be inoperable for operable, the remaining any reason, continued reactor components of both contain-operation is permissible only ment cooling subsystems during the succeeding fifteen shall be demonstrated to be days unless such pumps are operable immediately and sooner made operable provided weekly thereafter.
all remaining components of the containment cooling sys-tem are operable.
4.
From and after the date that 4.
When 3 containment cooling 3 containment cooling subsys-subsystem loops become in-tem loops are made or found operable, the operable sub-to be inoperable for any rea-system loop and its associ-son, continued reactor opera-ated diesel-generator shall tion is permissible only dur-be demonstrated to be oper-ing the succeeding seven days able immediately and the unless such subsystem loop is operable containment cool-sooner made operable, provi-ing subsystem loop daily ( ded that all active compo-thereafter.
nents of the other contain-ment cooling subsystem loo.2, including its associated die-sel generators, are operable.
5.
If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition with-in 24 hours.
C.
HPCI Subsystem C.
HFCI Subsystem 1.
The HPCI Subsystem shall be 1.
HPCI Subsystem testing operable whenever there is shall be performed as fol-irradiated fuel in the reac-lows: tor vessel, reactor pressure is greated than 105 psig, and Item Frecuency prior to reactor startup from a Cold Condition, except as (a) Simulated Once/cpera-specified in 3.5.C.2 and Automatic ting cycle
3.5.C.3 below.
Actuation Test - AFRIL 1973-128- ___ .-.. . , t . PBAPS i , . , LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT
3. 5.C HPCI Subsystem (cont'd. ) 4.5.C NPCI Subsystem (cont'd.)
Item Frequency - (b) Pump Once/ month Operability (c) Motor Opera-Once/ month , ted Valve - Operability (d) Flow Rate at Once/3 months
- 1000 psig
. Steam Pressure ,
(e) Flow Rate at Once/ opera-150 psig ting cycle
Steam Pressure i The HPCI pump shall deli-
ver at least 5000 gym for ! a system head correspond-i ing to a reactor pressure j of 1000 to 150 psig.
2.
From and after the date that 2.
When it is determined that , the KPCI Subsystem is made or the HPCI Subsystem is in- , i' found to be inoperable for operable the RCIC, the LPCI
any reason, continued reactor subsystem, both core spray j operationiis permissible only subsystems, and the ADS during the succeeding seven subsystem actuation logic e
days unless such subsystem is shall be demonstrated to be sooner made operable, provi-operable immediately.
The ding that during such seven RCIC system and ADS subsys-days all active components of tem logic shall be demon-the ADS subsystem, the RCIC strated to be operable system, the LPCI subsystem daily thereafter.
and both core spray subsys-tems are operable.
3.
If the requirements of 3.5.C cannot be met, an orderly 7:Q* down shall be initiated atd the reactor shall be in a tsic Shutdown Condition within 24 hours.
! April 73 f AM 49 December 15, 1978 129 - - - - - - -. - - - - -- -- --
. . PSAPS
LIMITING CONDITIONS FOR OPERATION , SURVEILLANCE REQUIREMENTS 3.5.D Reactor Core Isolation 4.5.D S2*llES fRgIC sub-syste_a), Reactor Core Isolation 1. The RCIC sub-system shall be _ Cooling,jRCIC sub_-sysegg operable whenever there is 1. RCIC sub-system testing shall irradiated fuel in the reactor be performed as follows: . vessel, the reactor pressure is greater than 105 psig, and Item ' prior to reactor startup from Freg~uency ~~ a Cold Condition, except as (a) simulated once/ Operating specified in 3.5.D.2 below.
Automatic Cycle Actuation Test * i (b) Pump Once/ Month Operability (c) Motor Operated Once/ Month valve operability (d) Flow Rate at Once/3 Months approximately 1000 psig steam Pressure" l (e) Flow Rate at Once/ Operating approximately Cycle 150 psig , - - .o ,,c Steam pressere** a " - uc ,4 (f) verify a~uto ~ once/ ope rating *** , matic transfer Cycle from CS? to suppression pool on low Cs? water level 2. From and after the date that j the RCICS is made or found 2. When it is determined that to be inoperable for any reaso the RCIC sub-system is inop-continued reactor power opera n, etable, the EPCIS shall be i tion is permissible only during demonstrated to be operable the succeeding seven days immediately and weekly there-provided that during such after.
erven days the BPCIS is, i operable.
},%,$lhuj restart on 3. If the requirements of 3.5.D cannot be met, an orderly shut-
- The RCIC pump shall deliver at down shall be initiated and least 500 gym for a system head l
the reactor pressure shall corresponding to a reactor pressure ced to 105 psig within of 1000 to 150 psig.
- Effective at 1st refueling outage
after Cycle 7 reload.
Amendment tio. 199,113/117 (3/19/86) -130- -... . . - . .. . _.
. -
. - _ _ PBAPS LIMITING CONDITIONS FOR OPERATION ' SURVEILLANCE REQUIREMENT F i 3.5.E Automatic Depressurization 4.5.E Automatic Depressurization j System (ADS) System (ADS) 1.
The Automatic Depressuriza-1.
During each operating cycle tion Subsystem shall be oper-the following tests shall able whenever there is irra-be performed on the ADS: i diated fuel in the reactor vessel and the reactor pres-A simulated automatic actu-sure is greater than 105 psig ation test shall be per-
and prior to a startup from a formed prior to startup af-Cold condition, except as ter each refueling outage.
specified in 3.5.E.2 below.
2.
From and after the date that 2.
When it is determined that one valve in the automatic one valve of the ADS is in-depressurization subsystem is operable, the ADS subsystem made or found to be inoper-actuation logic for the able for any reason, conti-other ADS valves and 'the nued reactor operation is HPCI subsystem shall be permissible only during the demonstrated to be 9pe,rable succeeding seven days unless immediately and at least such valve is sooner made weekly thereafter.
operable, provided that dur-ing such seven days the HPCI ( subsystem is operable.
3.
If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated - - - - + and the reactor pressure shall be reduced to at least 105 psig within 24 hours.
! i l . ! I \\ J8" 29, 1976-131-( . .. .. . - . _ . _ _.. . __
. PgAPS LIMITING CONDITkONS FOR OPERATION SURVEILLANCE REQUIREMENTS
3.5.F Minimum Low pressure Coolina 4 5.F Minimus Low Pressure sad Diesel Generator Coolian and Diesel , Availability Generator Availability 1. During any period when one 1. When it'is determined that one diesel generator is inoper-diesel generator is inoperable.af able.. continued react or op e r-low pressure core cooling and scion is permissible only containment cooling subsystems during the succeeding seven shall be demons t ra ted t o be days unless such diesel gene-operable immediately and daily retor is sooner made operable, thereafter. In addition, the provided that all of the low operable diesel generatore pressure core and containment shall be demonstrated to be cooling subsystems and the operable imme di a t e ly and daily remaining diesel gene rators thereafter.
shall be operable.
If this requirement cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the Cold Shutdown Condition within 24 hours.
2. Any combination of inoperable > c omp onent s in the core and c ont ainme nt cooling systems shall not defeat the capabi-lity of the remaining oper-able components to fulfill the cooling functions.
When irradiated fuel is in the reactor vessel and the reactor is in the Cold Shutdown Condi-tion, both core spray systems, the LFCI and containment cooling subsystems may be inoperable, provided no work is being done which has the potential f or draining the reactor vessel.
~ 4.
During a refueling outage, fuel and LPRM removal and r ep la ce me n t may be perf ormed p rovided at least one of the following conditions below is satisfied: Amendment No. 65/6h-132- - March 26, 1980
PBAPS LIMITING CO'NDITIONS FOR OPERATION SUP,TEILLANCE REQUIREMENTS , _. 3.5.F.3 (Cont'd) 4.5.F.2 (C ont 'd ) a.
Both core spray systems and the LPCI system shall be operable escept that one core spray system or the LFCI system may be in-operable for a period of thirty days, or b.
The reactor vessel head is removed, the cavity is flooded, the spent f uel pool gates are removed, and the water level is maintained at least 21 feet over the t op of irradiated fuel assemblies seated in the spent f uel storage p ool racks and no work is being perf ormed which has the potential f or draining the reactor vessel.
, l I ( i f Amendment No.
65 /64-132a-March 26, 1960 - - - - -. ., _,, _ _ _., _ _ _ _... _ _ _. _, _ __ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ ._ . _.. . \\ . . PBAPS 'IMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMEMS . 3.8 Radioactive Materials 4.8 Radioactive Materials l Applicability Applicability Applies to the radioactive Applies to the periodic monit-offluents from the plant.
oring and recording of radio-active effluents.
Obioctive Obieetive To assure that radioactive To ascertain ~that radioactive material is not released to releases are as low as the environment in an uncon-reasonably achievable and trolled manner and to assure within allowable values.
that any material released is kept as low as reasonably cchievable and, in any event, is within the limits of 10 CFR 20.
Specificatien Speci ficatien A.
General A.
General ~ t is expected that releases Operating procedures shall ef radioactive material in be developed and used, and - cffluents will be keptr'speci-at small eauipment,which has been frections of-the limits installed to maintaln controli ~ fled in section 20.106 of 10 over radioactive materials CFR Part 20 and as further in gaseous and licuid effluents specified in these Technical produced during normal reactor Specifications. At the same operations, including expected time the licensee is permitted operational occurrences, shall the flexibility of operation, be maintained and used, to keep compatible with considerations levels of radioactive material of health and safety, to assure in ef fluents released to areas that the public is provided a at and beyond the SITE BOUNDARY dependable source of power as low as reasonably achievaM e.
oven under unusual operating conditicns which may tem-porarily result in releases higher than such small fractions, but still within the limits opecified in Specifications 3.8.B.1 and 3.8.C.1, and in Section 20.106 of 10 CFR Part 20. It is expected that in using this operational flex-ibility under unusual operating mnditions the licensee will 203 Amendment No. 102/104 December 31, 1984 _ _ _ _ _ _.
- . . . . , . PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMFNTS ) exert his best efforts to keep levels of radioactive asterial in effluents as , low as reasonably achievable.
. 3.8.8 Liould Radweste Effluents 4.8.5 Liquid Badweste Effluents 1.
The concentration of radice-la. Facility records shall be active material released to maintained of the radio-areas at and beyond the SITE active concentrations BOUNDARY (See Figure 3.8.1) and volume before dilution shall be limited to the con-of each batch of licuid centration specified in effluent released, and of 10 CFR 20 Appendix B, Table II, the average dilution flow Column 2 for radienuelides and length of time over other than noble gases and which each discharge 2x10" uCi/ml total activity occurred, concentration for all dis-Ib. Prior to release of each solved or entrained noble batch of liouid ef fluent, s gases. With the concentration a sample shall be taken of radioactive material re-from that batch and analyzed leased to areas at and beyond for the concentration of the SITE BOUNDARY exceeding each significant gamma energy these limits, without delay peak. The release rater decrease the release rate shall be based on the of radioactive materials circulating water flow and/or increase the dilution rate at the tied of flow rate to'rdstore 'the C discharge. ~ - concentration to within the 1e.~ Radioactive *11ould~ waste limi ts. ' sampling and activity analysis shall be performed in accordance with Table 4.8.1.
, , 2.
The dose or dose commit-2.
Cumulative dose contri-
ment to a MEMBER OF THE butions shall be determined l PUBLIC from radioactive in accordance with the materials in licuid ef fluent methodology and parameters releases from the two in the Offsite Dese reactors at the site to Calculational Manual (ODCM ) areas at and beyond the SITE at least once per month.
BOUNDARY (see Figure 3.8.1) shall be limited tot a. During any calendar cuarter to < 3.0 arem to the totaT body and to ,1 10.0 arem to any organ, and, h. During any calendar year ' Amendment No. 102/104 204 December.31, 1984 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
- , .. PBAPS LIMITING CONDITIOFS FOR OPERATION StJRVEILLANCE PFCt?IRTFTRTF to 4 6.0 mrea to the total body and to 4 20.0 area to any organ.
When the calculated dose from the release of radioactive materials in liquid effluents exceeds any of the above limits, prepare and submit to the commission within 21 - working days, pursuant to Specification 6.9.3, a ' Special Report which identifies the causes for exceeding the limits and corrective actions that have been taken to reduce the releases of radioactive materials in liquid ef fluents and proposed corrective actions to be i taken to assure that ' subsequent releases are - within the limits.
This Special Poport shall also include
(1) results of radio- ,
,...m.
logical analyses of the , - drinking water source and (2) the radiolooical impact on the potentially affected drinking water supplies with regard to 40 CFR 141, Safe Drinking Water Act. Reactor shut-down is not recuired.
3.
During release of radioactive 3a. The licuid radwaste ef-
wa stes, the following fluents radiation monitor conditions shall be met s shall be calibrated a. The minimum dilution every 12 months with a water reeuired to known radioactive source satisfy 3.8.B.1 shall positioned in a reproducible be met.
geometry with respect to the b. The gross activity sensor and every ouarter monitor and flow monitor by means of a source on the waste ef fluent check. Additionally, an line shall be operable instrument functional test except as speci fied in shall be performed every 205 Amendment No. 2% 102/104 December 31, 1984 _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ I
.. _ _. _ _ s s.
. ? L.' . * . _
. . , l . . ~ l PBAPS l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS + 3.8.3.3.d and 3.8.3.3.e, month and an instrument . below.
check shall be performed ' c.
The effluent control every day during release.
' " monitor shall be set in Functional test shall ' accordance with the demonstrate operability methodology and prameters of the radwaste discharge in the ODCM to alarm and automatic isolation valve, . automatically close the and control room an-waste discharge valve nunciation if any of the prior to exceeding the following conditions exist: i limits specified in 1.
Instrument indicates ' 3.8.3.1 above.
measured levels above - - d.
From and after the date the alarm / trip set-that the gross activity point.
t monitor on the waste 2.
Instrument indicates ' effluent line is made an INOP failure.
or found to be inoperable 3b.
The liquid effluent flow for any reason, effluent monitor shall be cali-i releases may continue brated every 12 months.
only if best efforts Additionally, an instru- . are taken to make such ment check shall be i monitor operable, performed every day ' I provided that prior during release.
i to initiating a - release: 1.
At least two independent.
samples of the tank's . m, contents are analyzed, and 2.
At least two technically - ' qualified members of the Facility Staff . independently verify ' l the release rate cal-culation and discharge
line valving.
e.
From and after the date that ,
the flow monitor on the vaste affluent line is made i or found to be inoperable
' .for any reason, effluent releases via this pthway may continue only if best efforts are taken to make such monitor operable, provided that the flow , rate is estimated at least once per 4 hours during actual releases.
, t Pump performance curves - I , . I Amendment flo, /), JP/, 115/119-206-12-10-85
- - . - - - .. ._. _ _ _,. _ _ _ _ _ _ - _ _ -.
__ . PRAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIP.EMENTS 2.
The ai r dos e' in " area s at 2.
Cumulative dose contributions and beyond the SITE for noble gases shall be , BOUNDARY (see Figure determined in accordance 3.8.1) due to noble gases with the methodology and in gaseous effluents released parameters in the ODCM from the two reactors at the at least once per month.
site shall be limited to the followings a. During any calendar - quarter for gamma radiation < 10 mrad.
. During any calendar quarter for beta rad iation s <,20 mrad, b. During any calendar year for gamma radiation: < 20 mrad.
During any calendar ( year for beta radiati on <,40 mrad.
When the calculated air dose from radioactive
noble gases in gaseous ef fluents exceeds any of . the above limits, prepare and submit to the Com-u.
mission within 21 working days, pursuant to Speci-fication 6.9.3, a Special Report which identifies the causes for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and proposed corrective actions to be taken to assure that sub-sequent releases will be within the above limits.
Reactor shutdown is not recuired.
3.
The dose to a MEMBER OF 3.
Cumulative dose contributiens THE PUBLIC f rom iodine-131, for iodine-131, iodine-133, iodine-133, tritium and tritium, and radionuclides from all radionuclides in particulate form with half in particulate form with lives greater than a days Amendment No.102 /10c 209 Decerter 31, 1984 Amendment No. 66 l
_ . . - -.- O . PBAPS ' LIMITING CONDITIONS FOR OPERATION SURVFILLANCE RFQUIREPEhiF ~ ' half-lives greater than shall be determined in ' i 8 days in gaseous ef fluents accordance with the J released from the two methodology and parameters reactors at the site to in the CDG at least once areas at and beyond the per month.
SITE BOUNDhRY (see Figure ) 3.8.1) shall bis limited to the following: n. During any calendar , quarter: 1 15 arem.
b. During any calendar ' year: 1 30 area.
When the calculated dose from the rel ease of iodine-131, fodino-133, tritium and radionuclides in particulate form, with half-lives greater than R days in gaseous ef fluents exceeds any of the above limits, prepare and submit to the Commission within 21 working days, - pursuant to specification ( 6.9.3, a Special Report which identifies the causes , for exceeding the limits % +" '
- i
, , and defines the corrective , actions that have been ! taken.and proposed corrective ' actions to assure that sub-sequent releases will be - - within the above limits.
Reactor shutdown is not i reoui red. l 4.
During release of gaseous da. The reactor building wastes the following con-exhaust vent and main ditions shall be met to stack noble cas radiation avoid exceeding the monitors shall be cali-limits specified in brated every 17 months with 3.8.C.1: a known radioactive source ! a. The main off-gas stack positioned in a reproducible minimum dilution flow of geometry with respect to , 10,000 cfm shall be the sensor, and every ' i maintained.
cuarter by means of a h one reactor building functional test. The exhaust vent monitor channel functfonal test , l l 210 l Amendment No. 102 /104 December 31, 1984 l .. - - -.. -... _. _ -. _, _, _ _ _.. -.. _ -. - -. - -- ... -.... ... - .
-. _ _ _ __ -___ .
. PBAPS I LIMITING CONDITIONS PCR OPERATION SURVEILLANCE $tEQUIREMENTS k.
' and one main stack shall also demonstrate that noble gas monitor
control room alarm an-shall be operable and set to alarm in accordance nunciation occurs if any of the following conditions exist: , ' with the methodology 1.
Instrument indicates ' . and parameters in the measured levels ' ODCM. From and after the above the alara date that both reactor setpoint.
building exhaust vent 2.
Instrument indicates monitors or both main a downscale failure.
stack noble gas monitors Additionally, an instrument are made or found to be inoperable for any reason, check shall be performed every day.
, t effluent releases via (b.
The reactor building their respective pathway exhaust vent and the may continue provided a main stack flow rate
i least two independent monitors shall be grab samples are taken at least once per 8 hrs d' calibrated every 12 and these samples are months. Additionally, an instrument check shall analyzed for gross be performed every day.
activity within 24 4c.
hours, and at least two The reactor building exhaust vent and the main technically qualified stack iodine and particulate members of the facility sample flow rate monitors staff independently verify the release shall be calibrated every 12 months. Additionally, rate calculations.
an instrument check shall ! One reactor building exhaust vent iodine be performed every day for the reactor building , filter and one main exhaust vent sample flow ' stack iodine filter rate monitors, and every and one reactor build-week for the main stack l ing exhaust vent sample flow rate monitor.
particulate filter 4d.
The main stack sample and one main stack flow line Ri/Lo pressure particulate filter.with switches shall be their respective flow functionally tested every rate monitors shall be 6 months and calibrated operable. From and after every 18 months.
the date that all iodine filters or all particulate filters for either the reactor building exhaust vent monitor or the main stack monitor are made or . found to be inoperable for any reason, effluent releases via their ! respective pathway may i l l Acendment No. J77, 115/119 _211
12-10-85 . . . .. -. .. _ _ _,. _ _ - _--_ _. - -
_ _ _ _ _ - - , - _ -. . . PBAPS LIMITING CXMfDITIONS FOR OPERATION SURVEILLANCF RFOUIPEMENTS continue provided samples are continuously collected with auxiliary sampling . equipment for periods on the order of 7 days and ) ' analysed within 48 hours
af ter the end of the ~ ' sampling period.
d One reactor building exhaust vent flow rate monitor and one main ! stack flow rate monitor
bg shall be operable and set to alarm in accordance with the methodology and parameters in the ODOf.
From and after the date that both reactor building exhaust vent flow rate monitors or both main stack flow rate acnitors are made or found to be inoperab3 e for any reason, - effluent releases via ' their respective pathway i may. continue: provided I,
., the, flow, rate is estimated at,least once . ' per 4 hours.
v.- n , hwithlessthantheminimum ' number of. radioactive gas-ecus affluent.nenitoring instrumentation channels OPERABLE exert best efforts J to return the instruments to OPERABLE status within 30 days and if unsuccessful explain in the next Semi-annual Radioactive Effluent Release Peport why the in-operability was not cor-rectedin a timely manner.
i ! 5.
Gaseous effluents shall Sa. Doses due to gaseous ! be processed throuch ef fluent releases to the appropriate gaseous areas at and bevond ,
i waste treatment system the SITE BOUFDAPY as described below shall be projecte<* l prior to discharoe at least once per 212 i Amendment No. 102 /104 December 31, 1984 i - -. _ _ -. _ _ _. _ . _ _, _ _ _ _, _.... _ _ _ _. _ _ _. _ _. _.., _ _ _.... _ _ _ _ _ _. - m -.
. . ( . . PBAPS LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMFPPTF a. Gases from the Steam month in accordance Jet Air Ejector Dis-with the methodology charge shall be and parameters in princessed throuch the the COOT.
recombiner, holdup pipe, off-gas filter, and Sb. The appropriate gaseous off-gas stack.
radioactive waste system equipment as described b. Gases from the Mechanical in Specification 3.P.C.5 Vacuum Pump and Gland Steam shall be demonstrated Exhauster discharge operable every cuarter, shall be processed unless utilized to through the off-gas process gaseous waste stack.
during the previous 13 weeks, by analyzing c. R eactor, turbine, the gaseous waste radwaste, and recombiner processed through the building atmospheres appropriate eouipment shall be processed to determine that it through permanently meets the requirements of - or temporarily installed Specification 3.8.C.I.
equipment in the appropriate , tuilding ventilation system Sc. An air sample shall be and the Reactor Building obtained and analyzed Ventilation Fzhaust Stack, from allsbuildingfareas with the exception of the with an une~onitored - following unmonitored exhaust once per month, exhausts: 1.
Recirculation M-G Set and Reactor Building cooling Water eaufpment rooms. 2.
Control room utility and toilet rooms.
3.
Cahl, spread room.
4.
Eneroency switchgear rooms. 5.
125/250 VDC Battery rooms and the 250 VDC Battery rooms.
6.
Administration Building maintenance decontam-ination area.
With gaseous waste being discharged without treatment as recuired above, prepare and submi t to the Commission within 21 working days 213 Amendment No. 102/104 Decenter 31, 1984 ___ _____________ __ _ _
_- . - -- ..- . . . ,
. PRAPS
.. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS , pursuant to Specification e 6.9.3 a special Report which includes the following informations a.
Esplanation of Why gaseo'us
- '
redweste was being dis-charged without treatment, identification of any inoperable equipment or subsystems and the reason , for its inoperability.
b.
Action taken to restore the inoperable equipment to operable status.
. , c.
Summary description of ' action taken to prevent
a recurrence.
l Reactor shutdown is not required.
6.
The concentration of hydrogen 6a. An instrument check of the downstream of the recombiners operation of the hydrogen shall be limited to less monitors shall be performed than or equal to 24 by once per day.
. volume.
- ' a. With-the concentration 6b. The hydrogen monitors and of hyrogen downstream ' associated alarms downstream of ne*recombiner greater of the rooombiner.shall ' than 24 but less than or be calibrated once per equal to 44 by volume, month using standard gas ! restore the concentration containing 0-44 hydrogen, to within t.he limit within balance nitrogen or air , 48 hours.
by volume as specified , b. With the concentration of in the ODCM.
hydrogen downstream of the i recombiner greater than 44
by volume, an orderly reduction of power shall be
initiated within one hour to bring the hydrogen down-i stream of the recombiner
to less than or equal to 24 by volume.
c. Except as specified in
3.9.C.6.d, two hydrogen . monitors downstream of the recombiners shall be operable during power i .; operation.
i
Amendment No. 191, 115/119-214-12-10-85 , f -. -.. .. - -.. . . -
. l PBAPS ' l LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS d. With the numbiar of ' hyd' rogen monitors operal.1,e one less than required, operation may continue for up to 14 days provided grab samples are taken and analysed daily. With both hydrogen monitors inoperable operation may continue for up to 14 days provided grab samples are taken and analyzed every 4 hours during power operation.
7a. The radioactivity release , 7a. The radioactivity release rate of noble gases from the rate of noble gases from Steam Jet Air Ejector dis-the Steam Jet Air Ejector charge as determined by discharge shall be determined quantitative analysis of to be within limits at the identifiable gamma emitters following frequencies by shall not exceed 320,000 performing an isotopic uCi/see af ter.30 minutes analysis of a representa-decay. With the radio-tive sample of gases taken activity release rate of at the discharge of the noble gases from Steam Steam Jet Air Ejector.
- Jet Air Ejector discharge exceeding 320,000 uCi/sec 1.
At least once pgr month af ter 30 minutes decay unless the unit has restore the radioactivity - { been'out'of service for release : rate to Within * ' tho' entire month.
"" . its limit within 72 hrs.
2.
Within 4 hrs. following or be in hot standby an increase, if the within the next 12 hours.
off-gas monitors indicate an increase of greater than 50% in the steady state fission gas release after factoring out increases due to power changes. , 7b. One Steam Jet Air Ejector 7b. The Steam Jet Air Ejector radiation monitor shall radiation monitors shall be be operable during opera-calibrated every quarter tion of a main condenser and an instrtament check Steam Jet Air Ejector.
shall be performed cmce Upon loss of both steam per day. Additionally a Jet Air Ejector radiation functional test will be monitors, releases may performed every month. The continue via this pathway channel functional test for up to 72 hours provided shall also demonstrate that 215 Amendment No 102/104 December 31, 1984 . . _. _ _ _. _
__ _- __ _ _ _ _ . _ _. _ __ .
PBAPS LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ~ ( temporary monitor's are control room alarm an- ' used. Otherwise, be in at nunciation occurs if any of least HO'r STANDBY within the following conditions 12 hours.
exists i 1.
Instrument indicates measured levels above the alarm setpoint.
2.
Instrument indicates a downscale failure.
. Sa. Purging of the primary containment shall be through the Standby Gas Treatment System whenever primary containment integrity is l required as specified in i 3.7.A.2.
' b. Primary containment purging via the Reactor Building ventilation Exhaust System may be performed whenever primary containment
integrity is not required . as specified in 3.7.A.2.
l i '3.8.D 40 CFR 190 4.8.D 40 CFR 190 ' 1.
Thedoselord'osecNil-1.
Cumulative dose contribuI ~ ^ ' '( sent to.a* MEMBER OF THE tions from liquid'and gaseous PUBLIC from all uranium ef fluents shall be determined fuel cycle sources within in accordance with the 8 kilometers is limited, methodology and parameters to < 25 area to the-total in the ODO!. body or any organ (except 2.
Cumulative dose contribu-the thyroid which is limited tions from direct radiation to <75 arem) over the from the reactor units calendar year. With the and from radwaste storage calculated dose f rom the shall be determined in release of radioactive accordance with the method- - materials in liquid or ology and parameters in the ' gaseous effluents, exceeding ODCM.
twice the limits of speci-fications 3.8.B.2, 3.8.C.2, l or 3.8.C.3 calculations shall be made to determine whether the limits have been exceeded.
2.
The calculaticns should be 216 , , Amendment No. 102/104 December 31, 1984 i .. --- . - - - . _ . - -. -
-- . . PBAPS . _.
LIMITING CONDITIONS. FOR OPERATION SURVEILLANCE RECUIREMENTS made, including direct radiation contributions from the reactor units and from outside storage tanks to determins whether the limits have been exceeded. If such is the case, prepare and submit to the Commission, within 21 working days, pursuant to Specification 6.9.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and schedule for achieving conformance with the above limits.
This Special Report shall include an analysis that estimates the radiation exposure to a MEMBER OF THE PW LIC, including all . - effluent pathways and , direct radiation, including ( ,the releases covered by . this report, enlender' year,for.the - . It shall " ', also describe lenis of radiation and concen-trations of radioactive material involved and the cause of the exposure levels or concentrations.
If the estimated dose exceeds the above limits and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with 40 CFR 190. Submittal of the report is con-sidered a timely request and a variance is granted until staff action on the request is complete.
Amendment No. 102 /104-216a-I- . . December 31,'.1984 .. _ _ _ _ _ _ _ _ _ _ _ _ _ _.
_.-_ ... _. - _ . .- - _ _ _ . l PBAPS - . LIMITING 00NDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ! - - , I 3.8.E Radiological Environmental 4.8.E Radiological Environmental Monitoring Monitoring i i i 1.
All deviations from the 1.
The radiological environmental sampling schedule for the monitoring samples shall be i I
radiological environmental collected at the locations ' monitoring program, as and analyzed as specified i required by 4.8.E.1, shall in Table 4.8.3.a and the i be doctamented in the annual ODG. Deviations are permitted report.
from the required sampling a. When the radiological schedule if specimens are environments 1 monitoring unobtainable due to hazardous program is not conducted conditions, seasonal unavaila-j as des'cribed in the ODCM, bility, malfunction of auto-prepare and submit to natic sampling equipent or ! the Commission, in the other legitimate reasons. If
Annual Radiological equipment malfunction occurs, l Environmental Operating an ef fort shall be made ! Reports, a description to complete corrective i of the reasons for not action prior to the end conducting the program of the next sampling per,iod.
as required and the plans for preventing a.
The concentration of radio-a recurrence.
activity as a result of , b. When the level o'f plant offluents in an.
radioactivity as the environmental sampling I result of plant ef fluents medium shall be evaluated in an environmental en a quarterly Basis"
- sampling meditan at against the ecuation: one or more of the . locations specified concentration C1) + in the ODCM exceeds reporting level (1) the reporting levels of Table 4.8.3.b when concentration (2) +. 1 1.0 l averaged over any reporting level (2) i calendar quarter, ! prepare and submit All radionuclides used to the Commission by in this evaluation shall , the closing of the be averaged on a l month following the calendar cuarterly basis.
end of the affected calendar quarter, a l Special Report which ' includes an evaluation ! of any release conditions, l environmental factors or other aspects which i caused the reporting level of Table 4.8.3.b , ! to be exceeded. The l special Report shall also define the corrective - Amendment No. 102/104-216s-2-December 31,'1984 a.
, _ _ _. _ _. _ _ _. _ _.
-- . . __
PBAPS
.
LIMITINS CDNDITIONS FfiR OPERATION SURVEILLANCE REQUIREMENTS actions to be taken to reduct radioactive affluents so that the potential annual dose ' to a MEMBER OF THE PUBLIC is less than the calendar - year reporting level of Table 4.8.3.b. When more than one of the radio-nuclides in Table 4.8.3.b , sampling medium, this report shall be submitted
ift concentration (1? + reporting level (ID , concentration (23 +. 11.0 reporting level (2) When radimuclides other , I than those in Table 4.8.3.b , are detected and are the result of plant effluents, this. report shall be submitted if the potential . , annual dose to a MEMBER OF THE PUBLIC is equal to or ,
greater than the calendar , ! year limits of Specifica- ' tiens. The report is not i required if the measured level of radioactivity i was not the result of plant ef fluentst however, in such an event, the oondition shall be
reported and described in the Annual Radiological i Environmental Operating Report.
c. When milk samples become permanently unavailable f rom any of the sample locations listed in the ODCM, identi fy locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 21 working days. Specific locations from which samples are . Amendment No. 102/104-216a-3-Decembet,31,'1984 l -. - _ , _ _ _ _ _. _ .
' . . , PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS , , unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new locations for ob-taining replacement , samples in the next - Radioactive Does Assessment Report and include in the report revised figures and tables for the ODCM reflecting the new locations. ~2.
A land use census shall be 2a. The land use census shall conducted and shall identify be conducted every the location of the nearest 12 months by a door-milk animal in each of the to-door survey by 16 meteorological sectors consulting local within a distance of five agriculture authorities miles.
or by some other
appropriate means.
a. When a land use census (
' identifies a new loca- . tion which yields a . calculated dose or dose . commitment greater than ,, o -- - i the values currently being calculated in , . Speci fication 3.8.C.3, ' identify the new loca-tion in the next Radioactive Dose Assessment Report.
b. When a land use census identifies a location which yields a calculated dose or dos e commitment (via the same exposure pathway) at least 20% greater than a location from which samples are currently being obtained in accordance with Specification 3.8.E.1, add the new location to the radiological environmental monitoring program within 21 work-ing days. The indicator sampling location having-216a-4-Amendment No. 102/104 - Decader, 31,,.1984 __ _ _ _ _ - - _ _ .. _ _ _ _ _ _. . -. - _. _ _ _ _ __
_.
. _.. . - .. . PBAPS LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS the lowest calculated dose or dose commitment (via the same exposure ~ pathway) may be deleted from this monitoring program af ter October 31 of the year in which this land use census was conducted'. Identify the new location in the next Radio- ' active Dose Assessment Report and include in the report revised figures and tables for the ODOI reflecting the new locations.
3.
Analyses shall be performed 3a. A summary of the results on radioactive materials obtained as part of the supplied as part of the EPA Interlaboratory Comparison i Environmental Radioactivity Program shall be included
Intercomparison Studies Program, in the Annual Radiological
or another Interlaboratory Environmental Operating . Comparison Program that has Report pursuant to been approved by the Commission.
Specification 6.9.3.
a. With analyses not being
performed as required above report the corrective actions taken to prevent a recurrence in the ' Annual Radiological Environmental Operating - Report.
' 3.8.F Solid Radioactive Waste 4.8.F Solid Radioactive Waste 1.
The solid radwaste system 1.
The PCP shall be used to shall be used in accordance ensure meeting the burial
with a Process Control ground and shipping re-P rogram (PCP) to process cuirements prior to shipment wet radioactive wastes to of radioactive wastes fror.
meet shipping and burial the site, i ground reoui rements.
a. With the provimiens of the Process control Program not satisfied, suspend , shipments of defectively ' packaged solid radio-active waste f rom the si te. Reactor shutdown is not reouired.
Amendment No. 102/104-216a-5- ,pecember n,;'1984 _____-,--- _ __ _.. _ . . . . - --
__ _ . . . j I,IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.0 Mach'anical vacuum Pump 4.8.0 Mechanloal vacuum Pump , 1.
The mechanical vacuus pump At least once during each shall be capable of being operating cycle verify isolationofthemechanica isolated and secured on a automatic securing and signal of high radio-activity in the steam vacuus pump.
lines whenever the main steam isolation valves are open.
l 2.
If the limits of 3.8.0.1 are not met the vacuum pump" ' ^ ~ ~ h ~ ~ ~ ' ' ' ' l shall be isolated.* -
i l i
. - , ! . ' < . au -
l . ! ! I Amendment No. 102/104 216a-6 g becember 31, 1984 ' . , e ee . - - - - . - - - - - - . _. _. .. - _. -,_
. _ _ - _ - _ _ ..__ . -. _. -. - _ - - _ . . PBAPS ( . . TABLE 4.8.1 RADIOACTIVE LIQUID WASTE SAMPLING AND AFALifSIS Sample Lower Limit of Detection - Sample Type Sample Frecruency Sample Analysis (LLD)(1)(4)(5) -7 Waste Tank to Each Batch (2) Quantitative 5 x 10 uci/mi ' be released Analysis of Identifiable Gamma Emittera-6 I-131 1 x 10 uCi/a1
' -6 l P roportional Monthly ( 3 ) Fe-55 1 x 10 uci/ml Composite of-5 Batches Tritium 1 x 10 uCi/ml- -7 ( Gross Alpha 1 x 10 ,uci/ml
e % ,, - . ! ' . -e Proportional Monthly (3 ) Sr-89 5 x 10 uci/mi composite of-8 Batches Sr-90 5 x 10 uci/mi , I
One Batch Monthly dissolved noble 1 x 10 uCi/ml gases Notes - 1.
The Sample Lower Limit of Detection is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) , l limit for a particular measurement.
The values for the lower r limit of detection are based cm a 954 confidence level.
- 1
Amendment No. 102 /104 December 31, 1984 . k - _
. - _ _ - _ _ _ PBAPS . . , 2.
A batch release is the discharge of liquid wastes of a discrete vol ume.
Prior to sampling for analysis, each batch shall be isolated and thoroughly mixed to assure representative sampling.
3.
A composite sample is one in whidh the quantity of the sample is proportional to the quantity of liquid waste discharged and in Wich the method of sampling results in a sample representative of the liquids released.
4.
The principal gamma emitters for which the minimum detectable level specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, co-141, and co-144.
This list does not mean that only these nuclides are to be detected and reported.
Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the sample detectable limit for the analyses should not be reported as being present at the sample detectable limit level.
When unusual circumstances result in sample detectable limits ' higher than required, the reasons shall be documented in the Semi-Annual Effluent Report.
The values listed are believed to be attainable.
I I . . 5.
Certain mixtures of radionuclides may cause interference in the measurement of individual radionuclides at their detectable limit especially if other radionuclides are at much higher concentrations.
Under these ' circumstances use of known ratios of radicmuclides will be appropriate to calculate the levels of such radianuclides.
. Amendment No, 102/104 216b-2 December 31, 1984
. ..ii ,
__ ~ . . f PBAPE . . TABLE 4.9. 2 RADIOACTIVE GAFFOUS GSTE SAPPLIFG AFD AMALYSIS FF0F FAIF OFF-GAS STACK AFD RFACTOR BUII.DIPG VENT EXPAUST STACK l Sample Lower Limit of Sample Type Sample Frecuency Sample Analysis Detection (LLD)(3)(4) -4 . ' Grab Sample Monthly (2) Quantitative 1 x 10 uCi/cc(3) Analysis of Identifiabl e Gamma Emitters-6 Grab Famp3e Quarter 3y T ritium 1 x 10 uCi/cc-12 Charcoal reekly(3) 7-131 17 10 uCi/cc(?) Fil ters _ -10 '
Farticulate reek 3y(3) Cuantitative 1 x 10 uCi/cef?) Filters Analyris cf , Identifiable
Gamma Faitterr . ' ~ " ' -12
,< _.
., I-131 1 x 10 uCi/cef?) -11 i Pa rti culate Month 3y Gross Alpha 1 x 10 uCi/cc ! Filters (composite of weekly filters) ' -13 Monthly Fr-Po 1 x 30 uCi/cc Particulate - Filters-11 (cc=posite of Sr-90 1 x 10 uCi/cc weekly filterr) Feb3e Gas Continuously Noble cas -? Penitor Grosr f or p 1 x 10 uCi/cc (Main Stack) l Neble cas Ccnti nuousl y Foble Cas-f ' Ponitor GrosrJ or f 1 x Jn uCi/ec (Poef Ventr) 216c-1 Amendment No. 102 /104 December,31, 1984
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _. .. . l . PBAPS . . ,
Notes 1.
The sample Lower Limit of Detection is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
The values for the lower lirait of detection are based on a 954 confidence level.
2.
Sampling and analysis shall be performed following shutdown, startup or a thermal power change exceeding 15 percent of rated thermal power within one hour from a steady state condition unless (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3, and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
3.
Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing.
Sampling shall also be performed at least once per 24 hours for at least 3 days following each shutdown, startup or thermal power change exceeding 15 percent of rated thermal power in one hour and analyses shall i be completed within 48 hours of chanoing.
When samples. collected for 24 hours are analyzed, the corresponding LLD may be increased by a factor of 10.
This requirement does not apply if (1) . analysis has shown that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
4.
Certain mixtures of radionuclides may cause interference in the measurement of individual radionuclides at their detectable limit especially if other radionuclides are at much higher concentra tions. Under these circumstances use of known ratios of radimuclides will be appropriate to calculate the levels of such radi muclides. Nuclides which are below the sample detectable limit for the analyses should not be reported as being present at the sample detectable limit level. , 216c-2 Amendment No. 102/104 , Decemberi31, 198,4 % .
. . . - - - - _ _ - . . .._.
... - __-_____ .
PBAPS , , . . LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability: Applicability: , Applies to the auxiliary Applies to the periodic test-electrical power system.
ing requirements of the auxiliary electrical systems.
, Objectives Objective: To assure an adequate supply Verify the operability of the of electrical power for auxiliary electrical system.
operation of those systems required for safety.
f Specification: Specification . A.
Auxiliary Electrical Equipment A.
Auxiliary Electrical Equip- ! ment The reactor shall not be *ade critical unless all of tl.
- i following conditions are ' i satisfied: ' 1.
Both off-site sources and 1.
Diesel Generators the startup transformers and emergency transformers a. Each diesel generator shall , ' be m'nually started and loadedi ! are available and capable a of automatically supplying once each month to demonstrate-i i power to the 4kV emergency operational readiness.
The buses.
test shall continue for at.
l 1 east a one hour period at i 2.
The four diesel generators rated load.
l shall be operable and there I shall be a minimum of During the monthly generator i 104,000 gal. of diesel test the diesel generator l fuel on site.
starting air compressor shall be checked for operation and 3.
The 4kV emergency buses its ability to recharge air and the 480V emergency receivers.
The operation of load centers are energized., the diesel. fuel oil transfer pumps shall be demonstrated,~ , 4.
The four unit 125v batteries and the diesel starting time and their chargers shall be to reach rated voltage and l operable.
frequency shall be logged.
l I . e-217-APRIL 1973
i - - ~ - - , _y-. --w-- -. - - - v vry
- __ -_.
. _ _.. _ _ -. - - -. _ - -. _ _ . _. _ _ _ PBAPS ' LIMI'"ING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREME!CS '3.9.A 4.9.A (Cont'd), , . b. Once per operating cycle the condition under which the . . diesel generator is required will be simulated and a test ' conducted to demonstrate that i ' it will start and accept the emergency load within the specified time sequence.
The results shall be logged.
c. Once a month the quantity of .... , diesel fuel available chall be logged.
_, , d. Once a month a sample of diesel fuel shall be checked for quality.
The quality shall be within the acceptable limits specified in Table 1 of ASTM D975-68 and logged.
e. Each diesel generator shall be given an annual inspection in i accordance with instructions J i based on the manufacturer's recommendations.
. 2.bnitBatteries . a. Eve n week the specific ,
gravity, the voltage and ., temperature of the pilot cell a..fg{
- ,' 7" ana.everall battery voltage
,. , shall'bs measured and logged.
u , b. Every three months the i measurements shall be made of ,,.. voltage of each. cell to , nearest 0.1 Volt, specific gravity of each cell, and l temperature of every fifth cell.
These measurements shall be logged.
' c. The station batteries shall be l - l subjected to a performance p -- test every third refueling
- l outage and a service test during'the other refueling
outage.
In lieu of the performance. test every third , I refueling outage, any battery that shows " signs of degradation or has reached 851 of its service life" shall be
subjected to an annual performance test.
, Amendment No.'.107-218-11-9-84 ' n ,. + y, .3 . .. .. - - _ _ _ _ _. _ _ _ _ _ _ _ _ _ _., _ _ _ _ _,. _ _ _ _ _ _.. _,. _ _. _ _. _.. _. _. _,, _ _ _. . . -. . .. . ... . .
__ _.
. _ _. _ . PBAPS . . - SURVEILLANCE REQUIREMENTS . LIMITING CONDITIONS FOR OPERATION i - . 4.9.A (Cont *d) 3.9.A, The service test need not be performed on the refueling outage during Which the performance test was conducted.
The specific gravity and voltage of each ce,11 shall be determined after the discharge and logged.
- 3. Swing Buses a. Every two months the swing buses supplying wwer to the Low Pressure Coolant Injection System (LPCIS) valves shall be tested to assure that the transfer circuits operate as designed.
~ . .
.
. - . . Amendment No.
1037:0s-218a-11-7-84 .. - - - - - - . -- .... - -.. -. - -. -. -. -. . -,. -.-. . . - -.. -... .
? e LEFT BLANK INTENTIONALLY . G
w i ) - ,- i l l . I l ~. , . . ,N'
I l . PBAPS 1,:jlT[MG Corenf TIMas Fi i ni n ATin:4 suPV EI LI.,yac E_ F rnUI R EM ENTS
_ i . i. 9. D (M r.it ion with :na:a a a ts l e 4.9.D r.luiaw nt Whenever tJae reactor is in Run Mode or Startup Mode with the reactor not in a Cold Condition, the availability of electric powar shall be as specified Ln 3.9.A, except as follows: 1.
From and after the date incoming power is not available from or.e ctartup or emergency transformer, continued reactor operation is permissible to: ceven days.
During L:.A period, the f our diesel generators and associated emergency buses must be demons ra ed to be operable.
' 2.
From and after the date tha: g incoming power is not available from both c: art-up or emergency transformers, ' continued operation is permissicle, provided the four diesel generators and esccciated emergency buses are operable, all core and c or.t a inme n t coolinty cyctems are operable anc reactor power level is reduced to 25% of -ne design.
, l l ? I - l , -:19-Amendment No. 69/65 May 16, 1980 .. .- -- -.- - --. . _ _ - - -.
. PDAPS 1. :, a. I " : cr.::'i!, c:'O re. :' P g A T I O :: SURVEILLAt;CE idOUIKEMENTS 4.9.B 3. *>. b (Cuart ' d) 3. From asiel alter the date that one of the diesel generators ut associated emergency bus is made or found to be laivperabJe for any reasons, continued reactor operation is permissible in accordance with Specification 3. 5. F if Specification 3. 9. A.1 is satisfied.
From and af ter -he date that one of the diesel generators er associated emergency buses and either the emergency or startup transf ormer power socree are made or found to re anoperatla f or any reason, continued reactor operation is permissible tn accordance I with specification 3. 5. F, provided the other of f-site source, startup trancformer and emergency transformer are available and capable of - automatically supplying power to the uxv emergency buses.
L.
- r o.r. os.d after the date tna: ' . one of the 125 vclt batterf c y'. t e:.; ic made cr found to se inoperacle for any reason, rentinued reacter operation is permissible during the succeeding three days within +1ectrical safety
- ensiderations, provided repair work is ana :ated in
..e ..c c : expeditious manner l o re-urn the failed (
- .y.c n e s. : to an cperable
[ .- 1:+, and stec ::c2::en . 5.F s satisii=J.
. ~~0- '- A.mendment No. 69/63 May 16. lorn . }}