IR 05000272/1986027
| ML20212R283 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/29/1986 |
| From: | Collins S, Keller R, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212R252 | List: |
| References | |
| 50-272-86-27OL, 50-311-86-27OL, NUDOCS 8702020577 | |
| Download: ML20212R283 (97) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NOS. 86-27(0L) 86-27(0L) FACILITY DOCKET NOS. 50-272 and 50-311 FACILITY LICENSE NOS. DPR-70 and DPR-72 LICENSEE: Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 FACILITY: Salem Nuclear Generating Station EXAMINATION DATES: November 5-7, 1986 CHIEF EXAMINER: / /9bc 4[.5( rris, e %r Engireer (Examiner) Date REVIEWED BY: f J/2gW
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R. M. Keller, Chief', Project Section 1C Date APPROVED BY: /b b Mi I/XI% s. J. Collins,~ Deputy Director, Division Date of Reactor Projects SUMMARY: Four Senior Reactor Operator (SRO) candidates were examined during this period; three of the SRO candidates received their license. One of the candidates' failed the written examination hbhU$ DOC [2 V
. M REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: l SR0 l l Pass / Fail l l l IWritten Exam l 3/1 l l l l l Oral Exam l 3/0 l l l l l Simulator Examl 3/0 l l l 1 l0verall l 3/1 l l l l CHIEF EXAMINER AT SITE: B. S. Norris, USNRC Summary of generic deficiencies noted from grading of written exams: Ques. N .02 Candidates could not determine at which time in-core life (BOL or EOL) various accidents are more sever .05 Candidates had difficulty in calculating the resultant Start-up Rate while control rods are still in motio .03 Given a set of conditions, with respect to the Steam Dump System, candidates could not determine the method of cooldown of the Reactor Coolant Syste .09 Candidates were not familiar with the operation of the Rod Position Indicating search coi .0 Candidates were not aware of the basis for the pressure and temperature limits on the RHR syste .08 Candidates did not know why the 13Kv Disconnects should not be opened on a station blackou .07 Candidates were not familiar with the requirements for a one hour notification per 10 CFR 5 .
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; 3 Personnel Present at Exit Meeting:
NRC Personnel B. S. Norris - Reactor Engineer (Examiner) %"it , K. H. Gibson - Resident Inspector, Salem Facility Personnel R. E. Schaffer - Assistant Manager, Operations Training R. Best - Training Supervisor, Simulator K. Moore - Training Supervisor, Operations Training J. Gueller - Operations Manager, Salem
- Summary of Comments made at Exit Meeting: The Examiner noted that the upgrade efforts on the simulator were not yet complete and, as such, certain malfunctions could not be used (such as overriding of board controls). The facility stated that this class of candidates was chosen by their performance and ability rather than seniorit . During the Simulator portion of the examination, an evaluation was made of the operator's understanding of the Emergency Operating Procedures. The simulator scenarios included (1) Loss of Coolant Accident with a Blackout, (2) Loss of Feed with no Auxiliary Feedwater Storage Tank, and (3) Faulted Steam Generator with a possible PTS condition. Results of the evaluation are:
t The operators were familiar with their responsibilities and with the required actions during the emergencies, both individually and as a tea The procedures were performed by the minimum staff identified in the , Technical Specification " The operators did not physically interfere with each other while per-forming the E0Ps.
l The operators did not duplicate actions unless required by procedure (i .e. , independent verification).
i Where transitions were necessary from one E0P to another E0P or to
- ~ other procedures, precautions were taken to ensure that all necessary steps, prerequisites and initial conditions were met or completed.
l The operators were knowledgeable about where to enter and exit the other procedures.
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- Facility comments on the written examination are listed in Attachment 2, with the NRC resolution detailed in Attachment Attachments: Written Examination and Answer Key (SRO) Facility Comments on Written Examination made after Exam Review NRC Resolution of Facility's Comments
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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ATTACHMEtn' 1 FACILITY: _@ALEM_1h2 REACTOR TYPE: _PWR-WEC4 DATE ADMINISTERED _@6/11/05 EXAMINER: _NgRRI@z_B. CANDIDATE: _ INSIgyCI1gNS_Ig_CONQ1Q8Ig1 Una separate paper for the answer Write answers on one side onl Stcple question sheet on top of the answer sheet Points for each gaantion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grac,e of at 1 cast 80%. Examination papers will be picked up six (6) hou:-s after tha examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY __y@byg_ _IDI66 ___gGOBE _Y@ bye __ _ CATEGORY 23,5) _ggr:4H> _29._QQ _ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 24.fV
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_ ;zs9__ _253 99 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _25 00___ _ _ 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL z z . .s~o _,gixx__ _25399 _ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 95.So tet>r4Hk__ __ _____ Totals Final Grade All work done on this examination is my ow I have neither given nor received ai Candidate's Signature _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ . _ _ _ __ __ __ . _ _
, NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application cnd could result in more severe penaltie ~ R:stroom trips are to be limited and only one candidate at a time may 1 cave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Uze black ink or dark pencil gely to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the Gxaminatio . Fill in the date on the cover sheet of the examination (if necessary) . Uze only the paper provided for answer . Print your name in the upper right-hand corner of the first page of gach ecction of the answer shee Consecutively number each answer sheet, write "End of Category __" as cppropriate, start each category on a Dgw page, write only gg one sidg of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least thrgg lines between each answe . S;parate answer sheets from pad and place finished answer sheets face down on your desk or table.
' 12. Ure -abbreviations only if they are commonly used in facility litgtatutg.
< 13. The point value f or each question is indicated in parentheses af ter the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Pcrtial credit may be given. Therefore, ANSWER A'.?_ PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the gxamingt onl . 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 examinatio This must be done after the examination has bcen complete ,
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10. When you comp 1cto your cxeminstion, you chn11: l i
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_c. Assemble your examination as follows: l (1) Exam questions on to l (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe b. Turn in your copy of the examination and all pages used to answer the examination question c. Turn in all scrap paper and the balance of the paper that you did not use for answering the question d. Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke ;
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Dz__IHggRY QE_NQGLEAR PQL4ER PLANT QPERATION 2 _ELylpH2 AND PAGE 2 It!EB59DYNBUIGS
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~ QUESTION 5.01 (3.00) For each of the parameters listed below, provide the desired indication or trending that would be expected for NATURAL CIRCULATION cooling and what might result if the parameter was not trending as expecte Th Subcooling Steam Generator Pressure Steam Generator Level Pressurizer Level DUESTION 5.02 (3.00) - Indicate at which time in core lif e (BOL or EOL) the following accidents cro more sever Juntify your answe Main steam line break Total loss of coolant flow Rod withdrawal accident from low in the source range prior to any significant reactor coolant temperature increas DUESTION 5.03 (2.50) Th3 reactor is at 15% power, equilibrium Xenon and Samarium, boron concentration is 1200 ppm, BOL, Control Bank C rods are at 200 steps.
, Using Figures 5.1 to 5.7: What will be the final boron concentration if power is raised to 100%, ) ARD, steady state conditions? State which figure you use as applicabl ! State all assumptions and show all work.
QUESTION 5.04 (2.25) Figure 5.8 is a sketch of power vs. time; prior to time "O", the reactor hed been shutdown for 30 day Using Figures 5.9 & 5.10 as references, ekstch the aproximate xenon concentration associated with this power hictory. Use the same form provided.
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5 THEQRY DE_ NUCLEAR _PQWER PLANT OPERATIONi FLUIDS 2 AND PAGE 3 IHERMQDYNAMICE
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QUESTION 5.05 (2.75) A reactor at BOL is critical at 10E(-9) amps. Rods are withdrawn at 30 ctops per minute for 10 seconds in bank sequenc What is the SUR immediately before and after the rod motion stops? Assume differential rod worth is 20 pcm per step, effective delayed neutron fraction is 0.006 and the effective precursor decay constant is 0.1 sec- (1.75) How and why would the SUR change, if at all, if the initial conditions were at EOL? No calculations are necessar (1.00) QUESTION 5.06 (2.25) Tha reactor is in Mode 2, all shutdown banks are f ull out, and the control brnks are withdrawn to 180 steps on Bank Using the below information and Figure 5.11, determine the critical baron concentration by a 1/M plo Show all wor (1.75) Baron Concentration Count Rate 1500 ppm 100 cps 1400 120 1300 145 1200 185 1100 255 1000 435 State TWO assumptions that must be made for a 1/M plo (0.50) QUESTION 5.07 (3.00) Th3 reactor is operating at 20% power when the breaker for 21 RCP trips ope Assume no reactor trip, no operator action, and control rods are in manua Describe the change in the below parameters (increase, d: crease, or remains the same) for Loop 21 and Loop 22. Driefly explain your answer. Calculations are not necessar RCS loop flow Loop Tavg Steam generator pressure (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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32__JHggRY_QF_ NUCLEAR POWER PLANT OPERAT19N 2 _FLUIDB2_AND PAGE 4 IUgBMggyNAMICS
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QUESTION 5.08 (2.50) What is the subcooling margin of the plant if the following conditions exist: (1.00) That = 565 F Tavg = 549 F Tcold = 533 F Ppzr = 21B5 psig Psg = B50 psig If power is raised from 50% to 100%, how will the subcooling margin change (increase, decrease, or remain the same)? Justif (0.75) Which one of the following conditions would result in the smallest subcooling margi Assume RCS pressure is the same f or all case Justif (0.75) Controlled natural circulation cooldown immediately f ollowing a reactor trip from loss of flo . Continued operation at 5% powe . All RCPs are operating at normal no-load temperature after an extended shutdow QUESTION 5.09 (2.25) If reactor power is increased from 50% to 100%, how will DIFFERENTIAL
' rod worth change (increase, decrease, or remain the same) for the following conditions? Justify your answer. Consider each case separatel Rod position and boron concentration are held constant, temperature is allowed to decreas Boron is held constant, rods are withdrawn from 150 steps to maintain temperature constan Rod position is constant, baron concentratirn is diluted to maintain temperature constan QUES ON 5 O (1.50) g Th3 p1 has been at full power for several days. During a controlled rcducti in power, the Reactor Coolant System pressure increases drcsti al y for a small incremental load decrease. Pressurizer level and cpray are ormal, and heaters are operating properly. WHAT is the problem, HOW ould y u verify the problem, and WHAT should be done to correct it?
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6t- PLANT _gYETEMS_DEgIQN _GQNTROL t _AND_LNSTRQMENTATION t PAGE 5
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ObESTION 6.01 (2.00) During a Loss of Coolant Accident, all automatic saf ety injection cyctems functioned properly, pressurizer level stabilized, and RCS prcssure stabilized at 1700 psig. What is the approximate leak rate? Juatify your answer and state your assumptions.
QUESTION 6.02 (3.00) Lict the interlocks and/or automatic control features, if any, cOcociated with the below listed valves (refer to Figure 6.1): CV277 (0.40) CV18 (0.20) CV4 (1.20) CV35 (0.2O) CV7 (0.20) SJ2 (0.40) CV21 (0.20) CV71 (0.2O) QUESTION 6.03 (3.00) For each case below explain the resulting method of reactor coolant system tcmperature control and indicate the approximate final RCS Tav Ascume all systems normal except as stated and no operator actio Consider each case separately.jp % c & gg gg g J/96sm by JfsN The normal steam pressure setpoint is reduced by 85 psi while in Hot Standby awaiting reactor startu (1.00) The Train A steam dump interlock pushbutton for 'OFF' is depressed while at 5% reactor power awaiting turbine startu (1.00) Power is 100%, pressure setpoint is raised to 1200 psig and Main Steam pressure control is selected at the same time a reactor trip occur (1.00) QUESTION 6.04 (3.00) Using Figure 6.2 as a reference, list the normal, alternate, and cmcrgency power supplies for the 2D Vital InstrumentBus.[<Tfure 4, 2 /3 WWf n- et fere-ace, du ng to mgk /fue /ag a M. 7k 17se m Jes:,r evr & Stjeere ad h/4by y'm -Mitr e e. r e 6 po.r/ We prwk .sy/44 Q
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DEESTION 6.05 43.4G+(g,p) For each of the below radiation .nonitoring channels, state: What type of detector is used (Geiger Mueller Tube, Scintillation Tube, Ion Chamber, or Proportional Counter) Any control functions or interlocks If there are any differences between Unit 1 and Unit 2, state what the difference i Control Room Intake Duct (1R1B & 2R1B) Mrr % "_-- E!==.t '!*1^^ $ w eg / .h m qcsnekm , Component Cooling (1R17A & 2H17A) Plant Vent (1R41B & 2R41B) Fuel Handling Building (1R5 & 2R5) QUESTION 6.06 (2.25) Would the below events cause the INDICATED power to increase, decrease, or remain the same assuming that the actual power remains unchanged at 75% bcfore and after the event. All control systems are in automati Juctify your answer. Consider each event separatel Cooldown of the reactor coolant Core aging
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QUESTION 6.07 (3.00) Diccuss how auctioneered high Tavg is used for CONTROL in the below cyctem Include in your discussion any setpoints or bands of controls Rod control system Steam dump control system Pressurizer level control system Feedwater control system (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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bz__ PLANT _SygTgMS DggIgN _ggNTRQL - 2 _AND_ 2 INSTRUMENTATION PAGE 7
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ObESTION 6.08 (2.10) Lict the automatic start conditions for the Motor Driven and Turbine Driven Auxiliary Feedwater Pump Include the logic where appropriat DUESTION 6.09 (1.90) Assume that the reactor is shutdown and that power has been lost to the Rod Position Indicating system. The search coil for the first rod to be teated is energized and the readings indicate no rod couplin Whnt TWO conditions could cause this reading? QUESTION 6.10 (2.25) What THREF combinations of containment spray and containment fans ensures , o cufficient heat removal capability during a DBA7 l l f J e !
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QUEST 1DN 7.01 (2.50) 1 A maintenance man is working inside the containment while the reactor in at power. He is working in a radiation field of 850 mrem /hr gamma and 30 mrad /hr thermal and fast neutro Tha maintenance man is 28 years old and has a lifetime exposure through Icet quarter of 48 Rem on his NRC Form 43 additionally, he has cccumulated 1.0 Rem so far this quarte How long can the man work in the area before he exceeds his 10CFR2O limits? Show all work and state all assumption (1.50) During a declared emergency, this individual volunteers to enter a high radiation area and perform work necessary to prevent further effluent releas In accordance with the Station Procedures, what is his maximum allowed whole body exposure? (0.50) In accordance with the Radiation Protection Program Manual (RPPM), whose authorization is required in part b7 (0.50) DUESTION 7.02 (1.50) When reducing the concentration of fission gasses in the RCS per II-3.3.11 "VCT Degassification", VCT pressure is not permitted to fall , below WHAT pressure? (0.50) What are TWO reasons for the minimum pressure in part a? (1.00)
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QUESTIDN 7.03 (3.00) Unit 1 is operating at 60% power and station air pressure is decreasin A2cuming no operator action and all controls are in automatica What action will occur to maintain air pressure? Include setpoint.
i (0.50) In accordance with EI I-4.18, the reactor is to be manually tripped if air pressure falls below what value? (0.50) The following actions are taken after the reactor is trippe Give the reason for each steps (2.00) Stop 22 and 24 RCP . Maintain steam generator levels using 21 and 22 AFW pump , l If it becomes necessary to reduce RCS pressure, cycle one ' PORV until the desired pressure is reached.
't DUESTION 7.04 (2.40) Antwer the following in accordance with IDP-2, " Cold Shutdown to Hot Stcndby": , Below 250 F, at least one RHR pump or one RCP must be in operatio List TWO reasons why this requirement exist (1.2O) RHR must be isolated from the RCS before temperature reaches 350F or pressure reaches 375 psig. List ONE reason for each paramete (1.2O) QUESTION ' 7.05 (3.00) Tha below listed actions are from the Continuous Action Summary for EDP-LOCA-1, " Loss of Reactor Coolant". State what condition (s) must be octisfied for each action to be initiate LOCA-3 (Transfer to Cold Leg Recirculation) Isolate Faulted SG SGTR-1 (Steam Generator Tube Rupture) Open Miniflow Isolate accumulator (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) .
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QUESTION 7.06 (2.00) ) Tha following questions concern refueling operations, answer in
, cccordance with the Technical Specifications: What is the basis f or the requirement that the reactor be subcritical at least 100 hours prior to movement of irradiated fuel assemblies? (1.00) When may the operating RHR loop be secured? (1.00)
QUESTION 7.07 (3.50) Tha below conditions exist on the plant. Classify each of the Critical SOfety Functions using Figures 7.1-7.10 and list them in order from highest ,
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priority to lowest priorit . Pressures Containment 24.0 psig, RCS 2150 psig, Pressurizer 2150 psig, Steam Generators 1050/1050/500/1050 psig Levels: Containment sump 80%, Pressurizer 23%, RVLIS 40%, R44 33 R/hr, Steam Generators 15/10/0/16%NR Temperatures: Pressurizer 645F, Thermocouples 750F(all), RTDs 600F(all), Hot Leg 547F Nuclear Instrumentation: PR 0%, IR 3E(-8) amps, SUR IR O.1dpm, SR Ocps, . SUR SR Odpm ! Flows RCS Ogpm QUESTION 7.08 (1.60) i What is the reason for the below caution from EI I-4.9 " Blackout": j "Do not open 13KV Disconnects 1T50 and 2T50."
l QUESTION 7.09 (1.50) l l P;r EI I-4.16, " Radiation Incident", one of the immediate actions for .' cn alarm on channel 2R1A, Control Room, is to check the response of the olcrming channel. How is this accomplished? ,
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Zs.__E8QGEDWBES_ _ NORMALt _ARNQBMels_ EMERGENCY ANQ PAGE 11 80Q196991ge(_CQNIBQL
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QUESTION 7.10 (3.00) Answer the following in accordance with the Emergency Operating Procedures: Using Figure 7.11 and the highlighter provided, indicate the flowpath(s) correlating to the following set of condition Additionally, indicate tha position of all the valves that can be controlled from the Control Room: RWST level = 1.0 foot Containment pressure = 25 psig RCS pressure = 1200 psig Time since safety injection actuated = 15 hours QUESTION 7.11 (1.00) Antwer the following in accordance with III-13.3.2 " Steam Generator Blowdown-Normal Operation": What action must be taken if circulating water flow is lost to both sides of No. 22 Condenser? (0.50) Why must the GB1, Steam Generator drain valve, be closed for normal operation? (0.50)
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00ESTIDN 8.01 (2.50) Antwer the following in accordance with Salem Technical Specifications: Define HOT SHUTDOW (1.00) Which THREE parameters must you observe in order to verify that Reactor Core Safety Limits are being adhered to? (1.50) DUESTION 8.02 (3.00) Tha plant is operating at 100% power, all control systems are in automatic, cnd all equipment is operable except as stated below. Using Section 3.0 of th3 Salem 2 Technical Specifications provided, state what Limiting Condition of Operations (LCO) is violated (NOTE: reference to the page number & paragraph number is sufficient; if there is more than one action etctement for a system, be specific as to which paragraph). If the situation is not a violation of an LCO, justify by reference to page number cnd paragraph number. Consider each situation separately: #23 AFW pump failed its latest surveillanc #22 SI pump circuit breaker is racked out for replacemen #23 Charging pump motor is being replace #2A Emergency Diesel Generator fuel transfer pump has seized bearing Assume that all four of the above situations occur consecutivel . QUESTION 8.03 (2.00) An wer the following in accordance with AP-1, Administrative Procedure Progra Should the Senior Shift Supervisor approve the change? Justify your answer. Consider each part separately An On-the-Spot Change to delete an inspection hold point in a surveillance procedure, the change is applicable to Units 1 & The justification f or the change is adequat (1.00) An On-the-Spot Change to delete one of the approval signatures on the Start Approval Form from AD-16 " Post Reactor Trip /SI Review" (1.00)
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QUESTIDN 8.04 (2.00)
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Antwer the f ollowing questions in accordance with OD-15, Use of Operations Department Procedures: Under what conditions is it permissible to deviate from Technical Specification requirements? (1.00) Whose permission is required to deviate from Technical Specifications?
(0.50)
! What notification must be made if there is a deviation from a Technical Specification requirement? (0.50)
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QUESTION 8.05 (2.50) [ What type of safety or jumper / lifted lead tags, if any, are required for occh of the llow ng situations? 4 Leads nee to e lifted in accordance with an installation procedure or a new input to the annuniator pane A blind fla e is to be installed in the CVCS system while a relief valve is be g calibrate Work is to conducted on the Reactor Coolant Pump motor which will requi e ogging the pum ' Work is b ng one on the ventilation system which requires making damper ad ustm nts with no flow and then checking that the required flow has een stablishe Followin an ou ge it is discovered that the hand switches for the spr y valves have been incorrectly wired so that each switch operate the opp ite spray valve. Work is to be done to correct the wir ng proble . t > QUESTION 8.06 (2.00) It is discovered today, 5 November day shift (0800 - 1600), that a monthly curveillance due last Monday, 29 October mid shift (0000 - 0800), was not performed. This surveillance has been performed on time for the past cix month Is this item considered operable? Explain your answe (***** CATEGORY OO CONTINUED ON NEXT PAGE *****) _,. _ _ _ _ .__ - _ . _ _ _ _ _ _ _ _ _
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QUESTION 8.07 (2.00) Stcte whether each of the below events requires a DNE HOUR notification pcr 10CFR5 Consider each separatel The plant is in a condition NOT covered by operating and emergency procedures.
t The loss of the offsite notification syste A valid automatic initiation of the Reactor Protection System.
1 A shutdown was commenced because the plant was in violation of the Technical Specification QUESTION 8.08 (3.00)
You are the Senior Shift Supervisor on the midshift. Pressurizer level cnd pressure are decreasing rapidly and cause a low pressurizer pressure , rocctor trip and safety injectio Pressurizer level continues to drop with the ECCS pumps injecting. Pressure on 23 Steam Generator is higher thcn the others, and an Aux Operator reports that a safety valve is jammed op er. . Containment pressure is norma Using Figure 8.1, classify the even Justify your answe (2.50) a TRUE or FALSE: You can upgrade the event as conditions warrant, but to downgrade the event you must have NRC concurrenc (0.50) . QUESTION 8.09 (2.50) ' Mark Figure 8.2 to designate the areas that require shelter and/or cvccuation for the below condition Wind 30 mph ESE l Containment pressure 18 psig Letdown line monitor (2R31) showing increasing rad levels Pressurizer level decreasing l
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-_.____.,..,_m._,, _ _ , _ , . - _ _ . _ . _ _ ~ _ _ _ _ . _ _ _ , _ _ . _ _ . . . _ _ . _ _ _ _ _ _ . . . . _ _ . _ _ _ , , , , , _ , _ - , - _ - . , , _ - _ _ , . _ , . _ _ _ . _ _ , . . ADMINJgIBATIVE_PBQQEDUBES2_QQNDlllDNH2_BNQ_LIMITATJgNE PAGE 15 l *
i l ObESTION 8.10 (3.50) In accordance with the Technical Specifications: What is the minimum temperature for criticality? (0.50) What action must be taken in Mode 1 if the temperature is less than the minimum? (1,00) List FOUR basis for the minimum temperature for criticalit (2.00)
(***** END OF CATEGORY OB *****) (************* END OF EXAMINATION ***************) - - - - - . . - - - .. - _ - - - - - _ - - . -.-. . - - - - . . .
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HEAT SINK STATUS TREE
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l START l
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SE NR GREATER THAM ASM IN AT LEAST ONE INTACT SG vES l No
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* TOTAL PLOW CAPASILITY TO INTACT SGs GREATER THAN 22E94 lb/hr vES l No
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ALL SEs LESS THAN 1125 PSIG .
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ALL SE NR . . LEVELS LESS . THAN 67M
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VES l No g g l ..........
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ALL SGS . LESS THAN . - 1979 PSIE . . . ( . . . I vES l NO . .
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.LL Ss . . . .
NR LEVELS . . . . snEATER THAN . . . . ISM . . . .
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l : * * * lsnEENl lvELLou lvELLoul lvELLoMl lvELLOMl l RED l l l
~FthuM Salem Unit 2 y
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SHUTDOWN MARGIN STATUS TREE
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Isaseml [vettow] Ismeem] lvettoul _
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FIGuRL Salem Unit 2
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- THERMAL SHOCK STATUS TREE
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[ START l
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RCs COOLDO6SI GREATER TMAN -
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198 .EG EN LesT Se Mi vEs l NO
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RCs press / TEMP POINT TO
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THE RIGHT OF LIMIT A ves I =O .
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vesi =O
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C Lt. L RCs LEGS
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GREATER . .- . ALL RCs
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lEREENl lvELLOMl l PURPLE l l red ll GREEN l l GREEN] lveLLOM] l PURPLE l l l r FIGUAL Salem Unit 2 7. 6
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. -- * THERMAL Sif0CK LIMIT A CURVE PTS PLANT OPERATIONAL UMITS CURVE p me
_ , , p _ 22 F 20SO peng 200o-L M IT A T1 72 Soo-
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1000-300-ma r ; mo r 27e r 250 350 350 400 o S'O 150 150 250 TEMPERATURE F fthuAl Salem Unit 2 ?.7 _ - - __ __ _
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. TABLE A NORMAL CONTAINMENT /~* (less than 4 PSIG) ! ' (less than IE05 R/hr) p.: l l
PT403 TSAT PT403 TSAT PT403 TSAT
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PT405 C;"r RTD PT405 CET RTD PT405 CET RTD 2400 656 646 1600 589 590 800 511 503 2375 655 645 1575 596 587 775 507 499 2350 653 643 1550 594 585 750 504 496 2325 652 642 1525 592 583 725 499 492 2300 650 640 1500 590 581 700 495 488 2275 649 639 1475 588 579 675 491 483 2250 647 637 1450 586 577 650 487 479 2225 646 636 1425 583 574 625 483 475 2200 644 634 1400 581 572 600 477 470 2175 642 632 1375 579 570 575 473 466 2150 641 631 1350 576 567 550 468 461 2125 639 629 1325 574 565 525 462 455 2100 637 627 1300 571 562 500 457 450
- 2075 636 626 1275 569 560 475 452 445 2050 634 624 1250 566 557 450 446 440
,- 2025 632 622 1225 564 555 425 440 433 2000 630 620 1200 560 552 400 434 428 1975 629 619 1175 558 550 375 427 421 1950 627 617 1150 555 547 350 420 414 1925 625 615 1125 552 544 325 412 407 1900 623 613 1100 549 54 *
1875 621 611 1075 547 539 275 396 391 1850 618 609 1050 544 535 250 387 382 1825 617 608 1025 541 532 225 376 372 1800 615 606 1000 538 529 200 364 360 1795 613 601 $75 535 526 175 351 348 1750 611 602 950 532 523 150 336 332 1725 609 600 925 529 520 125 318 315
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1700 607 598 900 524 516 100 295 293 1675 605 596 875 521 513 75 265 263 1650 603 594 850 518 510 50 211 210 1625 601 592 825 514 506 25 211 209
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NOTES: Subcooling = TSAT from Table - RCS Tem . Use Core Exit TC's for RCS Tem Wide Range RTDs may be used l only if Core Exit TCs inoperabl F/4"# C Salem Unit 2
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TAELE B
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ADVERSE CONTAINMENT (4 PSIG and greater)
' *- (IE05 R/hr and greater)
I p.) * PT403 TSAT PT403 TSAT PTIO3 TSAT PT405 CET RTD PT405 CET RTD PT405 CET RTD
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2400 643 638 1600 580 576 800 474 471 1575 577 573 775 468 466 S 2350 640 635 1550 575 571 750 464 462
< 2325 639 634 1525 572 569 725 458 456 2300 637 632 1500 570 566 700 452 450 2275 636 631 1475 568 564 675 446 445 2250 633 629 1450 565 562 650 440 439 2225 632 628 1425 562 558 625 434 432 2200 630 625 1400 559 556 600 427 425 2175 628 623 1375 557 554 575 420 418 2150 627 622 1350 553 550 550 412 410 2125 625 620 1325 551 548 525 404 402 2100 622 618 1300 548 544 500 395 393 2075 621 617 1275 545 542 475 385 384 2050 619 614 1250 542 538 450 375 374 2025 617 612 1225 539 536 425 363 362 , 2000 615 610 1200 535 532 400 350 349 1 1975 .613 609 1175 533 530 375 334 333 '
1950 611 607 1150 529 526 350 316 315 1925 609 605 1125 526 523 325 293 292 1900 607 602 1100 522 519 300 259 259 1875 604 600 1075 519 516 275 212 211 1850 602 598 1050 516 513 250 212 211 1825 601 597 1025 512 509 225 212 211 l 1800 599 594 1000 508 505 200 211 211 1775 596 592 975 504 501 175 211 211 1750 594 590 950 500 498 150 211 210 1725 592 588 925 496 494 125 211 210 1700 590 586 900 491 489 100 211 210 , 1675 587 583 875 487 485 75 211 210 ' 1650 585 581 850 483 481 50 211 210 1625 583 579 825 478 476 25 211 209 NOTES: Subcooling = TSAT from Table - RCS Tem Use Core Exit TC's for RCS Tenip. Wide Range RTDs may be used
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2.
- only if Core Exit TCs inoperable.
, , N"## Salem Unit 2
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. TABLE B ADVERSE CONTAINMENT .. (4 PSIG and greater)
PT403 TSAT PT403 TSAT PT403 TSAT PT405 PT405 PT405 2400 645 1600 582 800 477 2375 643 1575 579 775 473 2350 642 1500 577 750 467 2325 640 1525 575 725 462 2300 639 1500 572 700 457 2275 636 1475 569 675 451 2250 635 1450 567 650 445 2225 633 1425 564 625 439 2200 631 1400 561 600 433 2175 630 1375 559 575 425 2150 628 1350 556 550 419 2125 626 1325 553 525 411 < 2100 624 1300 550 500 402 2075 622 1275 547 475 393 2050 620 1250 545 450 383 2025 619 1225 541 425 373 '
^ . 2000 617 1200 539 400 361 1975 614 1175 535 375 _. 347 1950 612 1150 531 350 331 1925 610 1125 529 325 312 1900 609 1100 525 300 287 .1875 607 1075 522 275 249 1850 -
604 1050 518 250 212 1825 602 1025 514 225 212
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1800 600 1000 511 200 212 1775 598 975 507 175 212 1750 595 950 503 150 212 1725 593 925 499 125 212 . ' 1700 591 900 495 100 212 1675 589 875 491 75 212 l 1650 586 850 486 50 212 1625 584 825 482 25 212
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NOTES: I Subcooling = TSAT from Table - RCS Tem ' Use Core Exit TC's for RCS Temp. Wide Range RTDs may be used
; only if Core Exit TCs inoperabl R4dl Salem Unit 2 pg - . _ - - . . - - - _ _ . . _ - - _ _ -. . - - - . - -. - -- -.
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e EMERGENCY CORE. COOLING SYSTEM ONE-LINE DIACRAM CCNTA!74Gff SuhF fD_G l4L'S I am u s am I 2ttn sH1 SJ" * 8 k o; _ E fN2 n 2181H4
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CVECT CLASSIFICATICU SECTIC3 1 e
. Primary Leakage
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Ini tia ting Event / Notification / Condition Emergency Action Level Reporting NOTE Refer to Section 2 for Primary to Secondary Leak Primary leak exceeding 1. Exceeding time limits of ATT 1 T/S limits resulting in T/S LCO 3.4.6.2 (Unit 1) or UE unit shutdow T/S LCO 3.4.7.2 (Unit 2) a) PRESSURE BOUNDARY resulting in the initiation leakage of a unit S/D b) 1 gpa UNIDENTIFIED
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leakage c) 10 gpm IDENTIFIED leakage d) 40 gpm CONTROLLED leakage e) RCS PRESS ISOL VALVES
- (See T/S)
an g Primary leak > 50 1. One chg pap cannot maintain ATT 2 f
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gym level A
*** For initiating condition 'C'*** ***Ref er to Section 5 prior to classification *** .
C. Known LOCA greater than 1. Low /decre,asing PZR level with .ATT 3
. total makeup capacity maximum charging flow SAE i PZR safety /PORV fail- 1. PZR press <2200 psig & POPS ATT 1 l
' ure to ressat not armed, or PZR press <375 UE psig & POPS armed; AND , ' 2. PORV/ safety valve tailpipe hi temp, or PRT temp, press, or level increasing , ' Pipe c ra ck s in stagnant 1. Cracks in weld areas of safety ATT 6 borated water systems related piping (as reported by (50.72b-1H Engineering or ISI/MIET)
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IE Inf orma tien Notice - one hour
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SGS 1 of 1 F/4"d' 8/ Rev. 4 to faja
EVENT CLASSIPICATION SECTION ; Primary to Secondary Leakage Initiating Event / Notification / Condition Emergency Action Le ve l Re por ting Primary to secondary 1. Exceeding T/S LCO 3.4 ATT 1 leak (Unit 1) or 3.4.7.2 (Unit UE a) I gpm TOTAL thru 2): all S/Gs See Tech Spec Action b) 500 gpd thru any Statements 1 S/G S/G tube rupture 1. Rx trip /SI on LO PZR ATT 2 (several hundred gpa) PRESS; AED A 2. R15 or any R19 alarm; AND 3. SSS/EDO judgement that a S/G tube rupture is in - progress (several hundred 9Pa) Steam break with 1. a) STMLINE ISOL SI on HI ATT 2 primary to secondary STM Flow with LO Tave or A leakage LO STM PRESS; or b) SI on STMLINE HI DIFF PRESS; AND 2. R15 or any R19 alarm Single S/G tube 1. RX trip /SI on LO PZR PRESS; ATT 2 failure with loss 2. R15 or any R19 alarms AND A of offsite power 3. PZR press recovers > S/G press; AND 4. Of f site power loss SGS 1 of 2 Re .
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, EVEDT CLASSIFICATIO3 SECTIOC 2 Primary to Seconda ry Leakage (com'd ) ..
Initiating Event / Notification / Condition Emergency Action Level Reporting
*** For Initiatiaq condition "N" *** ,
ee* Refer to Section 5 prior to classification *** Steam break with 1. a) STMLINE ISOL SI on HI ATT 3
> 50 gpm primary to STM FLOW with LO Tave SAE secondary leak and or LO STM PRESS; or indicated fuel b) SI on STMLINE HI DIFF damage PRESS; AND 2. R15 or any R19 alarms AND 3. 1R31C (2R31) in HI ALARM (offscale); AND 4. a) RCS a na ly si s shows failed fuel increase >14/30 min, total > 54, or I-131 equiv >300 uCI/cc OR b) Survey of main s team lines indicates high activity (SSS/EDO judgement) being ;
released to atmosphere OR c) One or.more of the f ollowing: R46A > 4.57 uCI/cc R46B > 4.57 uCi/cc R46C > 4.57 uCi/cc R46D > 4.57 uCi/cc
*** For initiating condition *F* ***
*** Refer to Section 5 prior to classification ***
' S/G tube rupture 1. Rx Trip /SI on LO PZR PRESS: ATT 3 (several hundred gpa) AND SAE with loss of o f f si te 2. R15 or any R19 alara; AND power 3. PZR press does not recover >
S/G press; AND 4. Of f site power loss
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, l SGS 2 of 2 Re . . ._ -
. EVE 5T CLASSIFICATION SECTION 3
- '. Secondary Leakage Initiating Event / Notification / Condition Emergency Action Le ve l Reporting A. Secondary line break 1. HI STM FLOW or abnormal ATT 1 causing a rapid increases in feed flow to UE uncontrolled secondary 1 or 2 S/Gs; AND depressurization 2. LO S/G press, LO TAVE, or decreasing RCS press S/G safety /PORV failure 1. Visual / audible indication ATT 1 to ressat~ at vent stacks after S/G UE press < setpoint; OR 2. Excessive f aed or s team flow for affected S/G C, Steam break with primary 1. a) STMLINE ISOL/SI on ATT 2 to secondary leakage HI STM FLOW with LO TAVE A or LO STM PRESS; OR b) SI on STMLINE HI DIFF PRESS; AND 2. R15 or any R19 alarm
*** For initiating condition "D" *** *** Refer to Section 5 prior to classification *** Steam break with >50 gpa 1. a) STMLINE ISOL/SI on HI ~STM ATT 3 primary to secondary FLOW with LO TAVE or LO SAE leakage and indicated STM PRESS; OR fuel damage b) SI on STMLINE HI DIFF PRESS; AND ,
2. R15 or any R19 alarm
* R31C (2R31) in HI ALARM (offscale); AND 4. a) RCS analy sis shows failed fuel increase >14/30 min, total >54, or I-131 equiv >300 uCI/cc *
OR , b) Survey of main s te am lines in- i dicates high activity (SSS/EDO l j udgement) bein released to a t- ,
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mosphere) OR c) One or more of the following:
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R46A >, 4.57 uCi/cc R46B > 4.57 uCi/cc R46C I 4.57 uCi/cc R46D 1 4.57 uCi/cc SGS 1 of 1 Re . . - . -- . .
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EVE 5T CLASSIPICATION SECTION 4
. Fuel Damage / Degraded Core .
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. Initia ting Event No ti f ica ti or -
Condition Emergency Action Level Repor tin c A. Fuel damage R31C (2R31) alarm; AND ATT 1 a) RCS sample shows f ailed UE fuel increase of
> 0.1%/30 sing OR b) RCS sample shows activity >
limits of T/S LCO 3.4.8 (U/1) or 3.4.9 (U/2) NOTE When SSS/EDO judgement indicates probable failed fuel, completion of RCS analysis should not restrict / delay classification of the emergency
*** For initiating condition "B" *** *** Ref er to section 5 prior to classification * ** Severe loss of fuel RCS sample shows I-131 equiv ATT 2 c ladding > 300 uci/cc; OR A R31C (2R31) HI ALARM (off-scale).and RCS sample shows failed fuel increase of > 14/30 -
min or total'> 54; OR R44 > 200 R/hr C. Fuel damage accident R5, R9, or R29 and R41B or ATT 2 with release to FHB R41C alarm; A or containment OR 2/4 of R2, R7, R10A , or R10B alarm and R44 1 1 R/hr; AND Fuel handling problem with possible fuel damage RCP seizure leading 1. RCP motor current spike to ATT 2 to fuel failure locied rotor value then zero, A or nudden flow decrease in affacted loop, or RCP vib alarm and loose parts monitor alarm; AND Letdown monitor 1R31C (2R31) alarm; AND Analysis shows RCS activity > limits of T/S LCO 3.4.8 (U/1) or 3.4.9 (U/2) SGS 1 of 2 Rev. 4
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__ ._ _ __ EVENT CLASSIFICATIOD SECTION 4
< si' l l ~ .. Damage / Degraded Core (con'd)
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Initiating Event / Notificatior ; Condition Emergency Action Level Reporting i Major damage to spent R5, R9, or R29 alarms ATT 3 fuel in containment OR SAE or Fuel handling 2/4 of R2, R7, R10A, or R10B building alarm, and R44 1 1 R/hr;
*
AED Confirmed fuel damage or l'oss of water level to below fuel
*** For initiating condition "F" *** , * ** Ref er to Section 5 prior to classification * * *
F. Degraded core with Containment high range monitor ATT 3 possible loss of R44 alarm (1 1000 R/hr) SAE Coolable geometry OR Possible loss of coolable geometry: 1.a) >,5 core exit T/Cs > 12 00 * F , OR b) ),2 WR hot leg RTDs > 70*F AND _ 2.a) RCS De lta ' T rapidly increasing (THOT
- increasing), or decreasing to zero (THOT=TCOLD).
AND Degraded Co re RCS analysis shows failed fuels a) Increase > 1% over 30 min, OR b) Total failed fuel > f 5% OR c) I-131 equiv > 300 , uCi/cc
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I SGS 2 of 2 Rev.
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3t0*:04 ') FicStos Fredact Boundary Failures fts following esnditions indicate failure of a fission predset boundar Aay tue baundary failuros, with the pac 01b111ty of a third, represente a CEssaA& ol BRER$5BCr condition and Attaebeest S oust be implemented immedi-
, noattor for these conditione after any single boundary failur I' .itiating Event M
Coudition Emergency Action Level Notification / Reporting DSSRADED CORB sots when ass /500 judgement in-dicates probable failed fuel, , completion of RCs analysis ) i should not restrict / delay i Baergency Classificatio ' 1 RCs Analysis shows failed fuels a) ineresee > 14 / 30 min, or b) total > St. or c) I-131 ogeiv > 300 uCi/cc OR 4 Engineering analysie indicates that a degraded cor;
- condition existe (as presented to the Emergency j Coordinator by Engineering support staff). OE OR 3. > 5 c o r e e mi t 7,'Ce RCs delta T rapidly increasing .
I 1200 F, or >2 WR That > 700F and (Thot increasing), or decreasing OR to sero ( Th o t = Tcold) 4 a44 > 200 R/hr n
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Less OF PRIE&BT COO &&BT
. . PSR level decreasing with seximum charging flow OR 2. 2/4 coat prese >4.0 poig and accumulator diacharge (modes 1, 2, 3)
OS 3. 2nadequate RCs subcooling (F250 or manual """"' calculations modes 1, 2, 3) ATT Da 2/3 5 03 """" gg C. R4 4 > 20 R/hr and 2/4 of R2, R7, ~ R104, """"" or RIOS alara OR S. 2/S CFCU drainage cont sump level alarse or 2/4 coat and >s1'3= (70 s) (no ladication prese >4.0 poig of in-coat sta break) ! l C. COWT&I555BT FAILEBE Contaissent 52 *****"*F8810" #4' OR /4 coat prese >47 peig and increasing i OR j 3. so coat spray capability containment press with <5 CFCOs available, and >23.5 peig and or 1 cent spray train with increasing OR
<3 CFCOs available OR Unisolable steam break outside containment with indications of primary-to-secondary leakage On 5. Coat penetration isolation valve (s) or cont ,
failure (any breach in containment as indicated by the significant rise in airborne activity in the area of concern or any significant and immediate decrease in containment pressure when containment pressure is abnormally high). , Scs 1 of 1 nov. 3 l -- -_ _ EVEMT CLASSIFICATION SECTION 6
. Radiological Releases - s ..
Note: Action levels listed re for valid RMS channel indic ion The validity of the dication should be confirmed b sample analysis or other me s as necessar Initiating Event / Notification / Condition Emergency Action Level Reporting A. Liquid release that Valid RMS indication of one ATT 1 or more of the following: UE exceeds T/S limits for > 15 min 1R18 - > 5.0 2E5 cpu 2R18 - off scale high 1 R19 - > 2.05E4 cpu 2R19 - ), 6.13E4 cpm 1 R13 -
), 2.95E3 cpm 2R13 -
), 2. 20E3 cpm 2R37 -
), 1.63E4 cpm
.
AND Confirmed a na ly sis of liquid waste effluent indicating discharge exceeding T/S limit Gaseous release that Valid RMS Indication of one ATT 1 or more of the following: UE exceeds T/S limits for > 60 min R41B -
),2.30E3 cpm /s -
OR _
>,1.38E5 cpm increase over 60 min R41C -
), 1.28E5 cpm for
>60 min R45B - ) , 8. 0 3 E-3 uCi /cc for
>,60 min OR Gaseous release that 1. Valid RMS indication of one ATT 2 -
exceeds 10 times T/S or more of the following: A limits for > 15 min R41B - > 2.24E4 cpa/m OR 3.36ES eps increase over 15 min R43 -
> 2.20E-1 mr/Hr R45B -
> 8.03E-2 uCi/cc for 15 min
J
,S 1 of 3 Rev. 1 s
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EVENT C A5 S I F IC ATION SECTIOD 6 (con':
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O. Radiological teleases ,.' Initiating Event / Notificati: Condi tion Emergency Action Level Reportir Dose Rate at Minimum 1.a) Field team measures whole ATT 3 Exclusion Area (MEA) body dose rates greater SAE equivalent to 500 mR/hr than 50 ares /hr for 30 min WB or 2500 mR/hr thyroid or greater than 500 aren/hr for > 2 min (MEA is for 3 min a,t the MEA defined as any monitoring OR location greater than b) Field team measures, at the 0.79 miles away from the MEA, Thyroid dose rates af f ected unit (equiv I-131 c on ce ntra tions ) greater than: 250 mren/hr (1.0E-7 uCi/ce) for 30 min or 2500 aren/hr 1.0E-6 uCi/cc for 3 mins OR Calculation of 500 arem/hr dose rate WB or 2500 aren/hr thyroid dose rate for > 2 min based on EP IV-111, EP IV-113 or EP IV-114 and actual meteorology; OR When methods 1 and 2 above are not available based on valid gaseous e f fluent monitors:
'R41B _> 3.41ES cpa /mi OR R45D - 1 5.56E1 cpm /ar OR R45B - > 2.05E-1 uCi/cc; OR R43 - 1 5.50E-1 ar/ Hrs OR One or more of the following:
R46A - D > 4.57EO uCi/cc and releas's path is via safety release R46A - D > 8.12EO aci/cc and release path is via PORV (MS 10s) R46A - D > 7.32E1 uCi/cc and release path is via AFW pump exhaust Dose rate at Minimun Field survey team measures ATT 5 Exclusion Area (MEA) at the MEA 1 R/hr W.B. or GE equivalent to 1 R/h r 5 rad /hr Thyroid (I-131 WB or 5 rad /hr thyroid concentration of 2.0 E-6 (MEA is defined for the uCi/cc); purpose of fie ld moni- OR toring teams as any distance greater than , 0.79 miles away from the affected unit) SGS 2 of 3 Rev. 1
. . . _ _ _ _ _ - . - . , . . , _ _ _ _ . _ , . . EMF C-AfSZP'OA?EiB
_ d2C?;O5 ~b' ~cin : G. Radiological Re leases
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Initiating Evsn t/ Notifica Condition Emergency Action Level Re po r ti r 7 . Calculation of 1R/hr WB or h* 5 r ad /h r thyroid using EP IV-111, EP IV-113 or EP IV-114 and actual meteorology; OR When methods 1 and 2 above are not available based on valid gaseous effluent monitors: R41B >,6.82E5 cps /m OR R45D >, 7.12E1 cps /n R45B >,4.10E-1 uCi/cc OR' R45C > 4 10E- 1 uCi/cc OR R43 >, 1.00E0 mr/Hr OR One or more of the following:
} R46A -D > 9.10E0 uCi/cc and release path is via safety ' '
reliefs OR R46A - D > 1.62E1 uCi/cc and release path is via PORV (MS 10s) OR R 4 6A - D > 146E1 uCi/cc and release path is via AFW pump exhaus E. Loss or -thef t of licensed Judgement of SSS/ED ATT 35 material in such quanti- (50.72b-1: ties and under such cir-cuestances that it appears to the licensee that a substantial hazard may result to persons in un-restricted area (Refer to 10C FR 20.402) G. Loss, theft or diversion Any loss, theft or diversion ATT 1 of any special nuclear of any special nuclear material UE material known or suspected (eg. fuel elements or incore detectors see 10C FR 70.52) either in transit to or from this faci-lity or a t this facility H. Abnormal degradation Excluding normal valve ATT 13 of systems designed to packing or gasket leakage OTHER , contain radioactive Examples material (not fuel a) through-wall leak in WHUT clad, RCS, or cont)
.. Radiological s abotage Judgement of SSS/EDO ATT 1 (attempted or substantiated) UE SGS . 3 of 3 Rev. 1 \ . _ _ _ . ___ _ _
_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _
%
f
, .
'
FXX3 PREDE1 ERMINED PKHECTIVE ACTION RB00PMEMRTION . CASE PREDE31mMINED PAR REMARKS Case A Core Degradation (1) with potential Evacuate all sectors 0-2.0 mi.le 'lhese Predetermined PAR's - for IDCA (2) and failure of shall be used until the Containment Boundary (3) Technical Support Center (TSC) and Emergerg Case B Operations Facility (BOF) are operational. Following i Core Degradation (1) and IDCA Evacuate all sectors 0-2.0 mile activation of the TSC and (2) with no imediate potential Evacuate downwind sector and EOF these predetermined for Contairinent Boundary Failure (3) adjacent sectors 2.0-5.0 mile PAR's may be used as ; Shelter all other sectors 2.0- guidance for making actual ! mile Protective Action Recomenda- i tions if PAR based upon I EP IV-108 is not availabl NC71E:
'
= ""
If meterological conditions (i.e., wind, rain, snow, etc) do not allow for a timely evacuation, than Case D is preferred to Case , ,_
= '*
Case C 's \
'
Core Degradation (1) and IDCA (2) Evacuate all sectors 0-5.0 mile /
'
with Failure of Contairment Evacuate downwind sector and ' Boundary (3) as judged by Senior adjacent sectors 5.0-10.0 mile = Shift Supervisor /EDO j
'
_ -
.
Case D - to M Core Degradation (1) and IDCA (2) Shelter all sectors 0-5.0 mile ' '" with iminent Failure of Containment Shelter downwind sector and , Roundary (3) as judged by Senior adjacent sectors 5.0-10.0 mile Shift Supervisor /ED AGuAL _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . --
E THEgRy_QE_Nyggg@R_PgWER PLANT _OPERATIgN2_FLUIp22_AND PAGE 16 IMEBd9DYN8 digs . 'O ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. ~ ANSWER 5.01 (3.00) Th stable or decreasing (0.30 each) Loss of natural circulation flow F subcooling oR ff,fg ,, r7t'* N + Voiding in core W hat leg which woul interrupt flow SG pressure tracking Tc saturation pressure SG not removing heat f /h a ur w *'s-r / c<3,& SG 1evel in Narrow Range Ap-g -t % ,g y,, SG no longer available as heat sink ' b e,r.s + wi f5L n %/
, _,,;,,,y,- # PZR level 50% J, gjfc ,,7 y.s. c w /u Voiding of hot leg which would inteMupt flow REFERENCE General Physics, HTFF, pgs 356-357 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
K&A OO2OOOK5.14/IF OOOO15EK1.01/IF ANSWER 5.02 (3.00) EOL (0.25)
,u MTCthe from is more drop in negative coolantand thus imparts greater positive reactivity temperature (0.25;
,
Doppler Only Power Coefficient (DOPC) is more negative and thus imparts more reactivity from the power increase ( ,
,,
g; 5v EOL (0.25) Closer to DNBR (0.25) DOPC is more negative and thus imparts more positive reactivity as power drops which tends to hold power up p0.5-)
,
,
#
/3 nr "y I'V ' a .s / vcY hr/pdy to 1' sten pwe< rn /rca /y - ,
Srt:-rf' i Ic ;, s k i r.; CUn grcatar i G . 21, DOPC is less negative and thus imparts less negative reactivity after the fuel temperature begins to increase (0.50) REFEREb.CE Student Notebook, Ch 47, Accident Analysis, pgs 28, 47 & 56
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~s~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OOOO40EK1.05/IF I
L.__IMEQBY_QE_NQQ(E68_EQWEg_E(@NI_ OPERATION t _ELylDS t_ANQ PAGE 17 ISEBdQDYNAMIGS
. '
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. S.
OO2OOOA2.03/IF OOOOO1EK1.02/IF ANSWER 5.03 (2.50) Racctivity due to power increase [ Fig 5.13 225 - 1450 = -1225 pcm (0.50) R cctivity due to rod withdrawal CFig 5.23 900 - O = +900 pcm (0.50) Racctivity due to Equilibrium Xenon Change [ Fig 5.33 1250 - 2750 = -1500 pcm (0.50) Coron worth CFig 5.63
=' -8.1 pcm/ ppm (0.50)
Total: -1225pcm + 900pcm -1500 pcm = -8.1 pcm/ ppm (12OOppm - BC)
-1925 pcm = -8.1 pcm/ ppm (12OOppm - BC)
BC = (9720 pcm - 1825 pcm)/(8.1 pcm/ ppm) BC = 974.69 ppm (0.50) REFERENCE Racctor Theory, pg 152 Racctor Engineering Manual, Part 1 (Figures)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO1010K5.21/IF ANSWER 5.04 (2.25) Sea attached Figure REFERENCE Student Notebook, Chapter 46, Transient Analysis, pgs TA-12 & TA-19 R2cctor Theory, pgs Rx-Th-TP-38.5, & 39.2-3 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ K&A OO1000K5.13/IF _ _ _ __ __ _ _ _ __ --__ .
Et__IBE98Y_QE_NQG(g@B_EgWE8_E(@NI_QEE8@IlQNt _ELUlQS t_@ND PAGE 18 IBEBdQDYN@dlGS
dNSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. l l
*
ANSWER 5.05 (2.75) Rho = (10 sec) (1 min /60 sec) (20 pcm/ step) (30 steps / min)
= 100 pcm (0.25)
Rho dot = (20 pcm/ step) (30 steps / min) (1 min /60 sec)
= 10 pcm/sec (0.25)
During rod motion: SUR = 26(lambda (rho) + rho dot)/(beta eff - rho) (0.25)
= 26(0.1(1E-3) + 1E-4)/(6E-3 - 1E-3) (0.25) = 26(2E-4)(5E-3) = 1.04dpm (0.25)
After rod motion: SUR = 26(lambda (rho) ) / beta ef f - rho) (0.25)
= 26 ( 1E-4) (5E-3) = 0.52 dpm (0.25) %nr 'O.20; At EOL, beta effective is less (o. 5cjk O. iO)
which gives a higher SUR and shorter time to increase power. {p,59pdO. '.O! REFERENCE Rarctor Theory Handout, pgs 309-310
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO1000K5.03/IF OO1000A1.06/IF A1.02/IF ANSWER 5.06 (2.25) /M = CRO 'O.25) ___ CR1 Starting graph at 1/M = 1.00 (see figure 5.11) (0.50) Correctly plotting given points (0 A')'O.50) Correct answer 850 +or- 25 ppm (0.50) Source strength is constant (0.25 each) Detector efficiency is constant REFERENCE R2rctor Theory, pgs 253-257, Figure TP-4 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA OO4000A4.02/IF OO4000A4.03/IF OO4010A3.04/IF _ THEQRY_OF NUQLEAR_PQWER_ PLANT QPERATiQNt _ELQLQSt_8NQ PAGE 19 IHERdQQYN8dlGS ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. S.
j e ANSWER 5.07 (3.00) Flow in Loop 21 will drop to zero then increase slightly in the reverse direction due to backflo (0.50) Flow in Loop 22 will increase slightly due to the lower head loss through the cor (0.50) Tavg in Loop 21 will decrease to Tcold due to the backflow from the cold leg plenum and an eventual zero steaming rat (0.50) Tavg in Loop 22 will increase due to the decreased flow through the cor (0.50) Ste2r gene ater preccu e in Leep 21 Mill decre2ce due te caturation d er Tccid in th2t Ice 'O.50) Ete r gener2ter preccure in Lecp 22 Mill decrc2cc clightly due to the inc-eared stea- lead and cer-ecpending d ep ir p essure-
-EC - U^ 'Ocit: t? 'O.50) All % ,b dst .t k 5 $ y. ~ N f REFERENCE e y,eru //g 'o-W// g _,, f,,r f/ ,,, ww / j,f*! ,,ecicefaJe ' l4 f, , u_, ,,, % ; tu[&p)&
Student Notebook, Ch 47, Accident Analysis, pgs 41-43
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OOOO17EK1.04/IF EK2.07/IF ANSWER 5.08 (2.50) Tsat for 2200 psia (2185 psig) (0.20)
= 649.45 F (0.40)
SCM = Tsat - That = 649.45 - 565 = B4.45 F (0.40) SCM will decrease (0.25) That will increase as power increases (0.50) Natural Circulation Cooldown (0.25)
'
Core delta-T during NC cooldown will approach full load delta-t (0.50) REFERENCE Steam Tables Student Notebook, Ch 22, Rod Control, pg RS-3 Student Notebook, Ch 1, RCS, pg 35
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KLA OO1000K5.56/IF OO2OOOK5.09/IF j
;
1 THEQSX_QE_NQCLE88_EQWEB_ELONI_QEE8611QNt _ELQ1QQt_6ND PAGE 20 ISEBdQDXN6dlCQ
- . *
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. e
.
ANSWER 5.09 (2.25) Decrease (0.25 for direction, 0.50 for justification) Lower temperature ,gygy r9 mer the density thus reducing the number of neutrons available for capture by the rod Decrease There is less flux available for interaction as the rods move towards the top of the core Increase Decrease baron concentration increases the number of neutrons available for interaction with the rod REFERENCE R2rctor Theory Handout, pg 209 and Figure TP-3 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ KLA OO1010K5.04/IF ANSWER 5.10 (1.50) Tha problem is a result of excessive noncondensable gases in the vapor cpece of the pressurize (This is causing the spray to be ineffective end is negating the effects of the saturated system.) (0.50) To verify the problem, sample the steam spac (0.50) To correct the problem, vent the pressurize (0.50) REFERENCE HTFF Fundamentals, pg 274 Student Notebook, Ch 32, Primary Chemistry, pg 34
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO2OOOK4.07/IF 3.5
OO4010K5.03/IF ___- t__ PLANT _ SYSTEMS _ DESIGNt _CONTROLt_AND__ INSTRUMENTATION PAGE 21 . ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, D. ,
.
& ANSWER 6.01 (2.00) Lockage would be the sum of the high head pumps discharge and positive dicplacement pump: (0.60) High head flow 2 X '400 gpm = 800 gpm (0.60) Positive displacement pump = 75 gpm (0.60) LEAK RATE = 875 gpm (0.20)
[ NOTE: acceptable value for high head flow is 300-500 gpm each pump 3 REFERENCE Student Notebook, Ch 10, ECCS, pg 53, Table 2 (pg 86) & Fig EC-3 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO6000K5.06/IF OO6050A1.01/IF ANSWER 6.02 (3.00) not able to open unless Pzr level > 17% (0.20 each)
-closes if Pzr level < 17% not able to open unless at least one charging pump is running-not able to open unless Pzr level > 17% -not able to open unless 2CV2 AND 2CV277 open-closes if Pzr level < 17% -closes if Phase A isolation-closes if 2CV2 OR 2CV277 closes closes on Phase A isolation diverts letdown flow around the demineralizers on high temp (120 F)
out of the letdown heat exchanger throttles to maintain pressure (350 psig) in the letdown heat exchanger modulates letdown flow to maintain level in the VCT opens on low level in the VCT-opens on an "S" signal none
-- . - _,,-. PLANI _@YgTEME_DE@IGN t _CQNIRQLu_AND INSTRQMENTATIQN PAGE 22
-
,
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. S.
' REFERENCE Student Notebook, Ch 6, CVLG, pgs 8-19
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO40xOK4.xx/IF ANSWER 6.03 (3.00) oue s The norma,1 steam pressure setpoint of 6955 psig maintains Tavg at "A77 , a decrease in the setpoint to SMe n psig would cause the dumps to open and cool Tavg to '543 F where the P-12 interlock would close all steam dump (1.00) Secondary pressure would rise to the setpoint of the 4+eet secondary dtmospheric relief valves which would maintain pressure at +O79 psig
3" primary temperature 551 (1.00) In pressure control mode, the output of the turbine trip and load rejection controllers is blocke Temperature would be controlled at 347 corresponding to the setpoints of MS1 (1.00)
ff/ REFERENCE Steam Tables Student Notebook, Ch 26, Steam Dump System, pgs 5, 15, & SD-12 Student Notebook, Ch 33, Main Steam System, pg 13
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K&A 04102OK3.02/IF ANSWER 6.04 (3.00) Normal: 2B 4160 vac Vital Bus (0.60 each) 2B 125 vde Battery 2B 230 vac Vital Bus Alternate: 2A 230 vac Vital Bus Emnrgency: 2A 230 vac Vital Bus REFERENCE Student Notebook, Ch 42, Electrical Distribution, Figures ED-1 to ED-12
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
KLA 062OOOK4.06/IF .
. .-
6 t__E(@NI_SYSIEMS_ DESIGNt_CQNIROkt_AND INSTRQMENI61[QN PAGE 23 - ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. . s ANSWER 6.05 ' 2. 50 57. cm) Unit 1 CP tub , Unit 2 Scintillation tube (0.25) Isolates control room ventilation (0.25) CP tub: Removed 40.237 C;;;;u; acnte ditch 2rge'/k3veclerer 'O.25) Scintillation tube (0.25) Closes the surge tank vent valve (0.25) Scintillation tube (0.25) Unit i am actica, Unit 2 Containment vent isolation (0.25) GM tube (0.25) Fuel building evacuation alarm & Charcoal filtration initiation (0.25)
.
REFERENCE Student Notebook, Ch 17, Radiation Monitoring System, pgs 37-47
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K&A 072OOOK4.xx/IF OOOK6.01/IF OOOK4.01/IF OOBOOOA2.04/IF OOOO61EK2.01/IF OOOO61EA1.01/IF ANSWER 6.06 (2.25) Decrease (0.25 for direction, 0.50 for justification) Detectors see less leakage resulting from decreased Tavg Increase Due to the shifting of the neutron flux toward the perimeter of the Core Slight increase Due to the decrease in baron absorption in the downtomer region REFERENCE Student Notebook, Ch 20, Excore Nuclear Instrumentation, pgs 17-21
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K&A 015000A1.06/IF A1.07/IF A1.08/IF i
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I
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6= PLANT _@YETEM@_DE@l@N t _CQNTRQL t _AND_lNSTRUMENTATION PAGE 24 -
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ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. a b ANSWER 6.07 (3.00) Compared with Tref to develop rod speed and direction (0.50) R 72- l +++ (0.25) O ! + D : + l + S 8- l +++++++ P l + E O--l -
: :-+++++++++++++-i--- --- l--
E + ! D -8- +++++++ l
+ l + l + l 8-72+++ l l l 1 : : : : : : : -5 -4 -3 -2 -1 0 1 2 3 4 5 Tref - Tavg Compared with Tref to generate deviation signal to control the modulation of the steam dump valves (0.50)
Tavg - Tref = 5F - 18F --> Dumps O - 100% open (0.25) Tavg is used for level reference (0.50) ro.o7o Tavg 547F - 571F --> Level 22.3% '6 i% (0.25) Low Tavg signal coincident with Reactor trip causes partial feedwater isolation - BF-19s & BF-40s close (0.50) Tavg = 554F (0.25) REFERENCE Student Notebook, Ch 24, RCS Temperature Indicating System, pg 14 Ch 22, Rod Control System, pg RS-11 Ch 25, Pzr Pres & Level Control Systems, pg PP-11 Ch 26, Steam Dump System, pg SD-5
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K&A OO1000A1.01/IF A1.04/IF K4.11/IF A3.06/IF l
N
-- - . - . .
6 t__E6@NI_EYEIEMS_DEgl@Nt _CONIRQLt_@ND INSIRQMENIAllQN PAGE 25 - ANSWERS -- SALEM 1&2
, -86/11/05-NORRIS, B. b ANSWER 6.08 (2.10)
MDAFWP: Safety injection signal (0.30 each) Low Low level 1/4 SG Both feed pumps trip TDAFWP: Low Low level 2/4 SG Undervoltage - 1/2 taken twice, 4Kv group busses Loss 125 vdc Loss of control air REFERENCE Student Notebook, Ch 11, Auxiliary Feedwater System, pg 51
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K&A 061000K4.02/IF ANSWER 6.09 (1.90) Rod fully inserted (0.95 each) Rod ejected REFERENCE Student Notebook, Ch 23, Rod Position Indicating System, pg 6
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A 014000K6.01/IF A2.02/IF ANSWER 6.10 (2.25) Minimum of five fan units (0.75 each) Both containment spray loops Three fans and one spray loop REFERENCE Student Notebook, Ch 12, Containment & Containment Systems, pg 15
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K&A O26000K3.01/IF A4.04/IF . . _ _ - - . _ - . _ . . _ . . _ - _ .
7.____PRQQEQUBES -_NQRMALu_ARNQRMAlt _EME8QENCY ANQ PAGE 26 BAQ196991CA(_CQN18Q( -
. *
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. b ANSWER 7.01 (2.50) (N-18) = 50 REM (0.30) Total lifetime to date = 48 + 1 = 49 REM Total lifetime available = 50 - 49 = 1 REM (0.30) Total this quarter available = 3 - 1 = 2 REM (0.30) Lifetime is more restrictive than quarterly limit 0.85 REM /HR gamma + (.03 RAD /HR) (10 QF) neutron =1.15 REM /HR dose rate (0.30) 1.0 REM /1.15 REM /HR = 0.87 HRS = 52 MIN (0.30)
[O.2 for using the conservative quality factor] REM whole body one time exposure (0.50) Vice President - Nuclear (0.50)
REFERENCE 10 CFR 20 Rrdiation Protection Program Manual, pgs 6-7 AP-24, Radiological Protection Program, pg 14
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KLA Plant Wide Generic #15/IF NOTE TO SALEM TRAINING STAFF: _ _ There is an inconsistency within your manuals as to who may
. authorize an exposure of 25 Rem ANSWER 7.02 (1.50) psig (0.50) To ensure seal backpressure and charging pump NPSH do not fall below minimu (1.00)
REFERENCE 11-3.3.11, pg 2 Student Notebook, Ch 6, CVCS, pg 16
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KLA OO4000A3.4/IF I
!
_ _ _ _ _ _ _ _ - . _ _ l
7s PRQCEQQRE@ NQRM6Lu_6BNQRM6Lu_ EMERGENCY ANQ PAGE 27 860196QGlC6L_CQNIBQL
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ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. '
.
ANSWER 7.03 (3.00) Emergency Air Compressor starts (0.35) 85 psig (0.15) Trip the reactor at 65 psig (0.50) . Reduces heat input to the RCS (R vv . ed)A s L rcir.; 21 S 22 ir cervice givrc bette p erru r rrrtrel 'O.3 Use of 23AFWpump,kmayreducethepresurein21&23S/Genough to cause a steam line differential SI (O.67) Normal and auxiliary spray are not availabl (0.66) , REFERENCE Student Notebook, Ch 39, Control Air System, pg 8 EI I-4.18, Loss of Control Air, pgs 1-2 & 5
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KLA OOOO65EA1.04/IF OOOO65EA2.06/IF OOOO65EK3.08/IF ANSWER -7.04 (2.40) cooling (dLSy 2 d2 v 60 'O.iO) reliable tpmperature indication #"' 4) ' O . iO' be m s tre tt fdo Nm temperature - heat removal capability of RHR HX (0.60) pressure - design pressure of RHR system (when combined with discharge pressure of RHR pump) (0.60) REFERENCE , IOP-2, Cold Shutdown to Hot Standby, pg 2 f TI 44. Y./, 3 Student Notebook, Chapter 8, Residual Heat Removal, pgs 5-6
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KLA OOOO25EK3.02/IF OO5000K4.07/IF . _ _ - . . - - -
It PRQCgQQRES - NQRMALt_AQNQRMAlt_EMERQgNCY AND PAGE 28 88Q1060GIC86_CQNIBQL
* . l -
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. ANSWER 7.05 (3.00) RWST Low Level alarm (15.24 feet) (0.60 each) Any SG completely depressurized OR Depressurizing uncontrollably Uncontrolled SG 1evel increase RCS pressure greater than 2000 psig RCS pressure less than 1000 psig REFERENCE , EOP-LOCA-1, opposite pg 2 { p p- 7'* " -
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~W~_~f ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO6050SG#10/IF ANSWER 7.06 (2.00) Ensures sufficient time has elapsed to allow for decay of short lived fission product (1.00) One hour per an eight hour period while core alterations are in the vicinity of the reactor vessel hot leg (1.00) REFERENCE Technical Specifications, pgs 3/4 9-3, 3/4 9-8, & B3/4 9-1
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A 034000K4.Ox/IF OOOO25EK1.01/IF ANSWER 7.07 (3.50) Core Cooling - Red * (0.50 for each classification) Shutdown Margin - Purple (0.50 for correct order based Containment Environment - Purple on their classification) Hest Sink - Yellow Coolant Inventory - Yellow Th rmal Shock - Green
*(Subcooling: Tsat for 2150 = 627F Must use Adverse containment and Thermocouples 627 - 650 = -23F ==> NO SUBCOOLING)
' REFERENCE EOP-CFST-1, pgs D-1 & D-2
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_ . . . _ _
._
_ _ _ _ _ _ . _ . . _ . _ _ _ . - . _ _ _ , . _ _ PROgggyRES - NORMAL 2_9BNgBMAL2_EMEBQgNQY AND PAGE 29 - BBD196991G86_GQNIBQL
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ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. A K&A System Wide Generic #10/IF ANSWER 7.08 (1.60) Interlocks are such that the Gas Turbine Breakers and 13KV Breakers cunt be open (1.00) to reclose 1T50 and 2T5 (0.60) REFERENCE El I-4.9, pg 4
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OOOO56EK3.02/IF ANSWER 7.09 (1.50) Depress CHECK SOURCE pushbutton (0.25) and observe an increase in the reading (0.50) Again depress CHECK SOURCE pushbutton (0.25) and observe reading return to original value (0.50) REFERENCE El I-4.16, pg 2
~~~~~~~~s~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A 072COOA4.03/IF ANSWER 7.10 (3.00) Rnfer to Figure 7.11 for proper lineup: each valve position O.05 each x 50 valves (2.50) only Charging and SI pumps are injecti:ig to RCS (0.50) REFERENCE Student Notebook, Ch 10, Emergency Core Cooling System, pgs 58-61
& Figure EC-11 EOP-LOCA-1, pg 15 EOP-Trip-1, Table B ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A OO5000K4.08/IF OO5000K4.12/IF _ - _ _ _ _ - _
.-. .-. _ __
Zs__E8QQgQQBES - NQBM6(t_6HNQBM6(t_EMg8QENCY AND PAGE 30 880196991G96_GQNIBQL -
. * ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. &
s ANSWER 7.11 (1.00) Blowdown must be stoppe (0.50) To maintain the tube sheet free of settled solid (0.50) REFERENCE III-13.3.2, pg 2
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
KLA 035010K4.04/IF _ - . - . _ _ - _ _ . . _ . __ _ _- _ _ _ . _ . _ _ _ . . _ . _ _ , - ._ . - - _ - - - .
0 - AQt!1NIEIB@IlyE PRQQgQQRESu_QQNQlTlQNSu__ ANDllMLI@TlQNS PAGE 31 ,
,
ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. * G ANSWER 8.01 (2.50) Keff < O.99 (0.50 each) 200 F < Tavg < 350 F Rx (thermal) power (0.50 each) RCS pressure RCS temperature REFERENCE Tcchnical Specifications, pgs 1-8 & 2-1
-
K&A OO2-OOO-K5.18 / IF OO2-020-SGM5 / IF ANSWER 8.02 (3.00) Pg 3/4 7-5, para 3.7.1.2' Action 'a' (72 hours) (0.60 each) Pg 3/4 5-3, para 3.5.2 Action 'a' (72 hours) Pg 3/4 1-10, para 3.1.2.4 (NOT an LCO) Pg 3/4 8-1, para 3.8.1.1 Action 'a' (72 hours) Pg 3/4 0-1, para 3.0.5 (6 hours) REFERENCE Salem 2 TS, pgs as listed in answer
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A 061000SGN7/IF OO6000K2.01/IF OO6050SG#7/IF OO4020SG#7/IF OOOSGM7/IF . _ _ , _ ,-
Es__8QUINigTRATlyE_PJQQEQQRE@t_QQNQlTICJ$$ m_MQ_L LM I TAT I ONS PAGE 32 l ., ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. . . ANSWER 8.03 (2.00) No . (0.50) I An OTS Change is good for only one unit gr Ok a#Ch' * p '* 4 " p (0.50) No .
(0.50)
An OTS Change cannot be made to ads or ODs (0.50) REFERENCE AC-1, Change to Station Documents, pg 9
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A Plant Wide Generic # 22/IF ANSWER 8.04 (2.00) When an Emergency condition exists where there is a threat to the health and safety of the public (0.50) or the threat may be increased if the directions in the Tech Specs are expressly followe (0.50) Senior Shift Supervisor 3 drTd AOge- (0.50) NRC (0.50) REFERENCE , OD-15, Use of Operations Department Procedures, pg 4 K&A System Wide Generic #10/IF "- ANSWER 8.05 (2.50) , No (0.50 each Jum r/ lifted lead tag Re locking tag with temporary release Re b ocking tag with temporary release N ne REFERENCE AP-13, Temporary Jumpers, Pumps & Lifted Leads, pgs 1 & 2 AD-15, Safety Tagging Program, pgs 1-3
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A Plant Generic #14/IF _ _ - , _ __ -
Ba. - AQt!1MISIRGIlyg PRQGEQlREQu_GQUQLILQUE._6MQ_LLt1LIGILQMS PAGE 33 ANSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. ..
' .
ANSWER 8.06 (2.00) s Tha item is NOT operabl (0.50) Surveillance intervals for a monthly are not to exceed 7.75 days (257. of 31 days) (0.50) Thu surveillance is IMNdb days overdue (0.50) Fciture to meet surveillance requirements constitutes a failure to meet operability requirement (O.50) REFERENCE' TS, pg 3/40-2
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A System Wide Generic #5/IF System Wide Generic #7/IF ANSWER 8.07 (2.00) Yes (0.50 each) Yes No Yes REFERENCE 10CFR50.72
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A System Wide Generic #3/IF ANSWER 8.08 (3.00) General Emergency (1.00) Section 5: B.1 Pzr level decreasing with maximun charging flow (0.50) C.4 Unisolable steam break outside containment (0.50) with primary to secondary leak (0.50) ! False (0.50) REFERENCE Ev:nt Classification Guide, Section 5
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A Plant Wide Generic #36/IF 4.7 l ! ! _ __ _ _ _ _ _ _ _ _ _ _ _ . - .
Eo AQMIM1EIR_@Ilyg_ERQQE M E_@t_QQNDITIQN_Hs_@NQ_(IMITATIONS PAGE 33 A,NSWERS -- SALEM 1&2 -86/11/05-NORRIS, B. .
'
- ANSWER 8.09 (2.50) see Figure Core B determination (1.00) Determine which area is downwind (1.00) Properly completing figure (0.50) REFERENCE Event Classification Guide, Section 5, pg 4
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
K&A Plant Wide Generic #36/IF ANSWER B.10 (3.50) F (NO tolerance allowed) (0.50) With Tavg < 541F restore to within limits within 15 minutes (0.50) or be in Hot Standby within the next 15 minute (0.50) . MTC is within its analyzed temperature range (4 req'd @ O.50 each) Protective instrumentation is within its normal operating range P-12 inter 1cck is above its setpoint Pressurizer is capable of being in an DPERABLE status with a steam bubble Reactor pressure vessel is above its minimum RT-NDT temperature REFERENCE TS, pgs 3/4 1-6 & B3/4 1-2 K&A OO1000K5.15/IF OO1000K5.16/IF OO2OOOK5.18/IF OO2020SG#5/IF OO2020SG#8/IF i
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22E.4 lb/hr YES l NO
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ALL SGS . LESS TNaN . 1125 PSKG
. YES l MO e=J
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ALL SG NR LEVELS LESS i
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TNAM 67M YES l NO . .
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. . .
ALL'S ,T N N . .
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tE S1G .
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veS l NO . . .
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MR Le . . . GRE.TE. vets.NTN . . . .
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vES l NO
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. . . . .
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l GREEN lYELLOM]. lYELLOMl lYELLOMl lYELLOMl l RED l s
fthund Salem Unit 2 l l
.. . _ _ - _ _ _ - _ . .
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e o SHUTDCWN MARGIN STATUS TREE T
l START l I
. .R ewer Range Less Than SM YES l MO
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Im SUR 4 ZERO e em o NEGATZUE o VESl MO . O e Source e o Range 44444@ e EMERGIZED e e I YESl No e 1 e e e e e 6 e Source 4444499 4 4 Range SUR * * ZERO OR @ * MEGATZUE * IR SUR e MORE * VESl MO MEGATZUE *
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t Fo 4uK Salem Unit 2 3 l
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o e THERMAL SHOCK STATUS TREE l START l
$'N RCS COdLDOWN GREATER THAM .
Ace oEs Im LAST Se Mi YEs l mO
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THE RIGHT OF LIMIT A YES l MO C$lo EEEs
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~FI4aAL Salem Unit 2 7. 6 .
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ERSE CONTAINMENT
'(4 PSIG and greater)
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PT403 TSAT PT403 TSAT PT403 TSAT PT405 CET RTD PT405 CET RTD PT405 CET t RTD 2400 643 638 1600 580 576 800 474 471 2375 642 637 1575 577 573 775 468 466 y l 2325 639 634 1525 572 569 725 458 456 l 2300 637 632 1500 570 566 700 452 450 2275 636 631 1475 568 564 675 446 445 l 2250 633 629 1450 565 562 650 440 439 i 2225 632 628 1425 562 558 625 434 432 2200 630 625 1400 559 556 600 427 425 623 1375 557 554 575 420 418 : 0 62 , 622 1350 553 550 550 412 410
-- _, 620 1325 551 548 525 404 402 2100 622 618 '1300 548 544 500 395 393 ,
2075 621 617 - 1275 545 542 475 385 384 i 2050 619 614 1250 542 538 450 375 374 2025 617 612 1225 539 536 425 363 362 2000 615 610 1200 535 532 400 350 349 1975 613 609 1175 533 530 375 334 333 1950 611 607 1150 529 526 350 216 315 1925 609 605 1125 526 523 325 293 292 1900 607 602 1100 522 519 300 259 259 1875 604 600 1075 519 516 275 212 211 1850 602 598 1050 516 513 250 212 211 1825 601 597 1025 512 509 225 212 211 599 594 1000 508 505 200 211 211 1800 1775 596 592 975 504 501 175 211 211 1750 594 590 950 500 498 150 211 210 1725 592 588 925 496 494 125 211 210 590 586 900 491 489 100 211 210 1700 587 583 875 487 485 75 211 210 1675 ! 585 581 850 483 481 50 211 210 1650 ' 1625 583 579 825 478 476 25 211 209 NOTES: Subcooling = TSAT from Table - RCS Tem . Use Core Exit TC's for RCS Tem Wide Range Ds may be used only if Core Exit TCs inoperabl P*"d# ' Salem Unit 2 7Y
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EMERGENCY CORE, COOLING SYSTEM ONE-LINE DIAGRAM fHt tet*5 nun..see c0NTA!PedDdT SLAP
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FXE PREDETERMINED PIUIECTIVE ACTIOi RBmP9tENDATION . CASE PREDETERMINED PAR RENARKS Case A Core Degradation (1) with potential Evacuate all sectors 0-2.0 mi.les. These Predetennined PAR's for IDCA (2) and failure of shall be used until the Containment Boundary (3) Technical Support Center
' - )s Case B~~
('ISC) and Emergency Operations Facility (BOF)
are operational. Following Core Degradation (1) and IDCA Evacuate all sectors 0-2.0 mile activation of the TSC and (2) with no imediate potential Evacuate downwind sector and BOF these predetennined for Containment Boundary Failure (3) adjacent sectors 2.0-5.0 mile PAR's may be used as Shelter all other sectors 2.0- guidance for making actual mile Protective Action Reconnenda-tions if PAR based upon EP IV-108 is not availabl N0fIE:
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If meterological conditions (i.e., wind, rain, snow, etc) do not allow for a timely evacuation, than Case D is preferred to Case , 4 Case C == # '"
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Core Degradation (1) and IDCA (2) Evacuate all sectors 0-5.0 mile with Failure of Containment Evacuate downwind sector and Boundary (3) as judged by Senior adjacent sectors 5.0-10.0 mile * * ' Shift Supervisor /EDO -{ J J
== J JJ Case D -
se a Core Degradation (1) and IDCA (2) Shelter ell sectors 0-5.0 mile with iminent Failure of Containment Shelter downwind sector and ""
,
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noundary (3) as judged by Senior adjacent sectors 5.0-10.0 mile Shift Supervisor /ED f/frWAC 21 _ __ _ __ _ _______ - __-_
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g ATTACHMENT 2 REVIEW COMMENTS TO SALEM SRO EXAM GIVEN NOV. 5, 1986
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5.01 Answer should allow for indication or tren Future use Trip 2. Not Arc materia .02 OK Reference also addresses MTC .. small MTC BO * RWA - Answer says BOL due to Beff being less; should be EOL due to Beff being les Second reason says due to DOPC being less negative, this is now a BOL effect, previous core it was an EOL effec FSAR assumes small value of doppler so BOL would be correct for dopple I l Ref. Ch. 47 AA, page 49, 50, 5 .03 OK l 5.04 Allow tolerance for accuranc .05 Equation sheet does not have an equation using rho dot (6) term in the SUR equation. Since these equations were provided to the students it should not be expected that they would add a term to come up with an equation using 4 to calculate a SUR immediately before rod motion stops.
t
5.06 OK Obscure at bes .07 OK, based upon parameter; not board indicatio Tave increase in 22 loop will only be short term steady state will remain same, l All steam pressures the same, no check valves, et .08- OK l 5.09 Decreases - agree. Reason due to density increasing i as temp. decreases; not density decreasing as temp.
I decrease .10 OK, delete .01 OK 6.02 Key wants both actions with and without interlock when in fact these are the same control or interlock functio Example is part A where key has two answers that appear to be redundan . 305-MISC:1
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6.03 OK With steam dumps off, S/G would pressurize to the MS-10's setpoint and maintain Tave at 5470 OK 6.04 Per the referenced Chapter 42, Electrical Distribution, page 35-36 discusses the Vital Instrument Busses and the power supplies. Page 35-36 lists only 3 power supplies, the normal 230 VAC to the inverter, the backup 125 VDC bus, and the 230 VAC emergency power source via the Solatron. Copy of reference attache .05 Unit 1 Scintillation detecto Remove OK Both units give Cntmt Isolatio .06 OK 6.07 OK OK Band of control for Tavg of 547 + 5710F causes a level change of 22.3% to 50% in the pressurize Allow some tolerance for the control bands and leve OK 6.08 OK 6.09 The search coil indicates if the rod is in the vicinity or no If not, then it is assumed to be fully inserted. If in the area, then it is assumed to be fully withdrawn. It can't be bot .10 OK 7.01 OK 7.02 OK
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7.03 Leaving 21 & 23 RCP's in Part A is not consistant with PORV question of Part It's confusing without procedure in han MISC:1 _ __ _ _ _
w ? t 7.04 It references a caution in IOP-2 but the more accepted reasons are from T.S. (heat removal & boron s +.ra ti fication ) . OK 7.05 A-D, O Doesn't exist as CA .06 OK 7.07 OK 7.08 OK 7.09 Source checking implies watching up scale & down scale question seems simple but answer wants specific .10 Conditions given made the event impossible to analyz No sump level was given. If cold leg recirc path was all that was ultimately required, why not ask for it? 7.11 Obscure procedur .01 OK 8.02 Consider assumptions of candidates in their answer On Part E, 2A pump is not equal to 2A ED .03 Change of holdpoint is a change of inten OK 8.04 OK 8.05 See attached sheet for comments the last time this question was aske (Question 8.07 from SRO Exam given Nov. 1985).
8.06 Disagree with answer. Oct. 29 through Nov. 5 is only 7 days, not 9 days. Still operable until 1800 on Nov. .07 Can't argue that they are listed in 10CFR50, but we spent considerable effort in developing Event Classification Guide to alleviate need to memorize all notification / classificatio .08 OK 8.09 OK 8.10 OK 305-MISC:1
* .q * Alarms and Trips Alarms and trips for the 460- and 230 volt systems are listed below. A trip is indicated by *. Bezel Alarm CLOSE for all infeed breakers i
i
- Overhead Annunciator
.i "460-230V BUS HOT SPOT" j 115 VOLT AC SYSTEM i
' The 115-volt system supplies power to the vital instrument busses, the miscellaneous AC distribution cabinets, the emergency lighting distribution ,
cabinets, the station computer, and the incore instrumentation distribution
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cabinets. All of the vital busses and distribution cabinets are provided with automatic transfer to an emergency power source, to ensure a regulated power
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(s supply, should the main power source fai ~ + Vital InstrumentSBusses -(Figure ED-8)
There are a total of four vital instrument busses. These busses normally receive power from their respective inverters. Emergency power sources are supplied, via autotransformers and Solatrons (voltage regulating trans-formers), from a 230-volt vital bus of a different channel. The inverters are supplied power from the AC/DC power supply cabinets. These cabinets are supplied power from the 230-volt vital busses. The 230-volt AC is transformed and rectified to a DC potential of approximately 125-VDC. Also contained in the AC/DC power supply cabinets are auctioneering circuits that compare the output of the rectifiers to a 125-VDC infeed.from.the 125-VDC. busses. The output of the rectifiers is usually higher, so the AC/DC power supply cabinets normally supply the inverters. Should the output voltage of the AC/DC power supply cabinets drop below the 125-VDC bus voltage, tne auctioneering circuit will shif t the source of power to the inverters over to the 125-VDC busse Rev 0 12/82 0941D:4
- .M.a c_mmsMUr apQ Ns'n 6 DY 'd W M Rt Eg - 48.07;a could need red blocking tag to lift lead (leads are lifted * de-energized)
If the leads were left in the circuit at completion of job (even though it was in accordance with a procedure) a lifted leads / jumper tag would be neede .b Red blocking tags to isolate system to remove relief valve, if system was to be placed back in-service, install lifted leads / jumper tag and remove red blocking tag .c Red blocking tag with a temporary release to jog pum .Could not jog pump with a yellow permissive tag since yellow permissive tags must be used in conjunction with red blocking tag .d Red blocking tag with a temporary release cy; non Worker blocking tags are only authorized to be used by T&D personnel (Transmission & Distribution) per AP-1 .e For a field switch - OK ? For a control room switch - No, notation is shift logs 8.0 Yes, there is no bistable for OPAT from PR NIS (set
.
to zero)
( .b Question could be answered (2) ways:
Yes, because lifting head on by itself does not constitute a mode change )Do not enter Mode 5 until last bolt on - head is tensioned) No, procedure to lift head also includes steps to tension studs so permission would be given to complete procedure not just a position.
i l Bottom line is, did the trainee know that you can't change modes in an action statement? I
.c Answer is totally wrong!
Should be ye If you check redundant equip. for operability first, and you feel tht work can be completed within action time frame. How else could you repair anythin .d OK
.e OK 8.0 OK .b OK 8.1 If the PRI -> SEC leak is thru a S/G leak (1) only (1)
S/G, then you will exceed the 500 gal / day Tech Spec limit I
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p., .
O Attachment 3 NRC RESOLUTION OF FACILITY'S COMMENTS Question Resolution 5.01 Accepted 5.0 Accepted 5.04 Accepted j 5.05 Will allow for a discussion of the SUR ; immediately before rod motion stop ! 5.0 Noted 5.0 Noted 5.0 Not accepted; based on Salem training material .0 Accepted 5.0 Accepted 5.10 Deleted 6.02 Considered in grading 6.03 Setpoints corrected 6.04 Not accepted; there are five different flow
. paths for electrical power to the vital instrument bu .05 Accepted 6.0 Accepted 6.09 Not accepted; an understanding of the search coil operation makes the assumption invali .0 Answer key modified 7.0 Answer key modified 7.0 Not accepted; reference EOP-Trip- .09 Considered in grading 7.10 Noted; purpose of the question was to determine if the candidates could analyze the given conditions and determine the correct flow paths, it was not required to determine the acciden .11 Noted 8.02 Considered in grading 8.0 Accepted 8.05 Deleted 8.06 Answer key corrected 8.07 Noted
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