IR 05000272/1985024
| ML20137Q507 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/14/1986 |
| From: | Dante Johnson, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137Q489 | List: |
| References | |
| 50-272-85-24, 50-311-85-27, NUDOCS 8602060255 | |
| Download: ML20137Q507 (200) | |
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A U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 85-24(0L) and 85-27 (OL)
FACILITY DOCKET NO. 50-272 and 50-311 FACILITY LICENSE NO. OPR-70 and 75 LICENSEE: Public Service Electric and Gas Co.
P. O. Box 236 Hancock's Bridge, New Jersey 08038 FACILITY: Salem 1 and 2 EXAMINATION DATES: November 19-26,' 1985 CHIEF EXAMINER: M Donald F. JohnsonY ead Reactor Engineer
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Date (Examiner)
REVIEWED BY: k)
Robert M. FelTer, Chief Projects Section 1C
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Date APPROVED BY: .
Hatry B. $1(ter, Chhf,
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Projects Bhinc* No. I SUMMARY: Seven ?,eactor Operator (RO) and eight Senior Reactor Operator (SRO)
candidates were examined during this period; four R0 and six SRO candidates received their licenses. Three of the RO's failed the written exam, two of those three also failed the simulator and oral exams; one SRO failed the written exam and one SR0 failed the simulator and oral exams.
i 8602060255 860123 PDR ADOCK 05000272 G PDR
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F REPORT DETAILS TYPE OF EXAMS: Replacement-EXAM RESULTS:
l RO l SR0 l l Pass / Fail l Pass / Fail I l l l l l 1 l l Written Exam I 4/3 l 7/1 l l 1 I I I I I I l Oral Exam I 5/2 1 7/1 l 1 1 I I l I l l l Simulator Exam l 5/2 1 7/1 l l l l 1 1 I I I l Overall l 4/3 l 6/2 l l 1 I I 1. CHIEF EXAMINER AT SITE: D. (NRC)
2. OTHER EXAMINERS: R. M. Keller (NRC)
D. G. Ruscitto (NRC)
N. F. Dudley (NRC)
B. S. Norris (NRC)
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1. Summary of generic deficiencies noted on simulator / oral exams:
i a. Candidates were unable to perform a simple heat balance on a heat exchanger, nor were they able to locate' all of the necessary
- parameters when prompted.
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b. Candidates were not aware of the sources of radiation for the RCS both shutdown and at 100% power; they were not aware of the relative
magnitude for radiation levels on. spent fuel; they were not aware of the potential problem associated with the incore nuclear detectors.
c. Several deficiencies relate to the Candidates use of the E0Ps:
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-(1.) When transitioning to a new procedure, the Desk Operator (RO)
i would repeat the same verification steps and wait for a response
- from the Board Operator (RO) rather than realize that it had
. just been verified and would not change (Example: " Veri fy l reactor tripped")
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(2.) When transitioning to a new procedure, the Desk Operator did not always ensure that the Shift Supervisor (SRO) verified the
! kickout parameter.
(3.) When obviously in the wrong procedure, the Shift Supervisors l were slow to stop and rediagnose the situation.
l d. During the simulator examinations, when equipment was reported as i not functioning properly, the Control Room personnel would simply
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accept the report and not question why the equipment was malfunction-ing. .
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2. Summary of generic deficiencies noted from grading of written exams:
a. SRO Exam
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- Questior, 5.05.a - Candidates did not know the reasons for the differences in requirements for performance of an Inverse Count ;
Rate Ratio (ICRR) calculation for rod withdrawal and boron l
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- Question 7.09 - A0P-ELEC procedures do not include all of the Technical Specification sections w11ch are applicable to the failure of electrical buses. (Examples: AOP-ELEC-VIB-C, r
Appendix 1 and A0P-ELEC-125v-C, Appendix 1 are incomplete).
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Question 8.07 - Candidates were unable to identify the proper type of tagout to be used for maintenance activities.
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- Question 8.11 - Candidates were unable to properly classify.
events when provided with the E-Plan Classification Guide.
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b. R0 Exam
Question 1.08 - Candidates were unable to determine the system flow.and operating pressure when running centrifugal pumps in parallel with a positive displacement pump.
Question 1.13 - Candidates were unable to calculate the new steady state values-for Tavg and S/G pressure after closure of one Main Steam Isolation Valve.
Question 3.0.6.a - Candidates were unable to list the Control Room indications for the failure of-the high voltage power supply to a power range detector.
Question 4.03 - Candidates were unfamiliar with the length of time an On-the-Spot change is valid for; they were also unfamiliar with the approv_al requirements.
3. Personnel present at Exit Interview:
NRC Personnel D. F. Johnson, Lead Reactor Engineer (Examiner)
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B. S. Norris, Reactor Engineer (Examiner)
T. J. Kenny, Senior Resident Inspector Facility Personnel D. Hanson - Manager, Nuclear Training A~. Thompson - General Manager, Nuclear Services R. Schaeffer - Assistant Manager, Operations Training J. Gueller - Operations Manager, Salem J. K. Ll.oyd - Principal Training Supervisor P. J. Landers - Principal Training Supervisor J. M. Zupko, Jr. - General Manager, Salem Operations 4. Summary of NRC Comments made at exit interview:
a. Preliminary results on oral / simulator exams:
One Senior Reactor Operator - marginal on oral One Reactor Operator - marginal on oral and simulator One Reactor Operator - failed oral and simulator b. Salem has only two shutdown initial conditions (ICs) for the simulator, neither of which was functional for this exam. This must be corrected prior to the next scheduled examination.
c. Generic weaknesses (details are contained within the report)
(1) sources of radiation (2) simple heat balance calculations (3) slow to rediagnose when obviously following wrong emergency procedure.
5. Summary of_ facility comments and commitments made at exit interview; a. Facility agreed to correct the problem of no shutdown ICs prior to the next scheduled requalification training (February,1986)
6. Changes made to Written Exam during examination review:
a. SR0 Exam See Attachments 3 and 4 b. R0 Exam (all changes to answer key)
1.01.b changed to:
" Core exit TCs - stable or decreasing RCS hot leg temperature - stable or decreasing S/G pressure - stable or decreasing RCS cold leg temperature - saturation temperature for S/G pressure 1.12.d(b) Changed to "Feedwater heater outlet" 2.09.g Changed to "FC (MS 169 and 171)"
3.10.b. Deleted " Emergency stop pushbutton" Added " Backup differential" Attachments:
1. Written Examination and Answer Key (RO)
2. Written Examination and Answer Key (SRO)
3. Facility Comments on SR0 Written Exam 4. NRC Resolution of Facility Comments on SR0 Exam
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111 R STER U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATORjLICENSE EXAMINATION FACILITY: S ALEM .1&2.t3
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REACTOR TYPE: PWR-WEC4
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DATE ADMINISTERED: 35/11/19'i
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EXAMINER: NORRIS, B. S.
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APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
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Uss separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each question are indicated i,n parentheses after the question. The passing Srede requires at-least 70% in each cate3ory and a final grade of at leest 80%. Examination papees will be picked up six (6) hours after tha examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
________ ______ ___________ ________ ___________________________________
25.00 25 0
________ ___1_0 _ ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER
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PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00
________ __..___ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25 '5 0
___1_00____'__1_0 _ ___________ ________ 3. INSTRUMENTS AND CONTROLS 25.00 25 0
________ ___I_0 _ ___________ ________ 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS
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FINAL GRADE _________________%
All work done on this examination is my own. I have nelther Givcn nor received aid.
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
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QUESTION 1.01 (2.50)
c. Describe how Thermal Driving Head (TDH) causes natural circulation. [0.503 b. What indications are used to determine if natural circulation is occuring? [1.003 c. While performing a natural circulation cooldown without RVLIS, pressurizer' level rapidly increases. What is the MOST probable cause? [1.003 QUESTION 1.02. (2.50)
The reactor is at 15% power, equilibrium Xenon and Samarium, boron concentration is 1200 ppm, BOL, Control Bank C rods are at 200 steps.
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Using Figures 1.1 to 1.8 What will be the final baron concentration if power is raised to 100%,
ARD, steady state conditions? State which figure you use as applicable.
State all assumptions and show all work.
QUESTION 1.03 ( .75)
Moderator temperature coefficient becomes MORE negative from BOL to EOL PRIMARILY because of (choose ONE):
a. Decrease in number of thermal neutrons available for absorption in the moderator.
b. Increase in resonance escape probability per degree change in moderator temperature.
c. Decrease in thermal utilization factor per degree change in moderator temperature.
d. Increase in rod worth due to fuel burnout.
QUESTION 1.04 ( .75)
Deppler coefficient becomes MORE negative from BOL to EOL because off (choose ONE)
o. Increase in effective fuel temperature.
b. Production of Pu-240.
c. Clad creep and fuel pellet swell.
d. Overlappin3 of resonant peaks.
(**x** CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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9 a 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
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____________________________________________ GUESTION 1.05 ( .75) Control rod worth is GREATEST at (choose ONE): a. Low boron concentration.
b. High boron concentr,ation.
c. Lou moderator temperature, d. High moderator temperature.
QUESTION 1 06 (1.00) Differential boron worth becomes MORE negative over core life because there is _____________ competition amons boron atoms for thermal (more/less) ngutrons, therefore each boron atom is worth _____________.
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GUESTION 1.07 (1.00) a. What is the MAJOR reason for ensuring sufficient Net Positive Suction Head (NPSH)? b. Does the required NPSH increaser decreaser or remain the same if the speed of the pump increases? QUESTION 1.08 (2.50) Refer to Figure 1.91 If all three pumps are running, what will be the total system flow (spm) and operating pressure (psis)? Show your work.
QUESTION 1.09 (2.50) c. List the production and removal reactions of Xenon and Samarium. 01.503 b. Provide TWO reasons for Xenon contributing nore negative reactivity at full power than does Samarium. E1.003 (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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____________________________________________ QUESTION 1.10 (2.00) Figure 1.10 is a sketch of Xenon concentration vs. time.
Using Figures 1.11 & 1.12 as reference, sketch the approximate power history that would be associated with this Xenon concentration plot. . U e the same form provided._ Assume that all power changes are made as STEP chan3es.
GUESTION 1.11 (3.25) The reactor is critical at 2x10E-9 amps; 150 seconds later power is observed to be 7x10E-8 amps.
a. What is the Startup Rate (SUR)? CO.753 b. Given a delayed neutron fraction of 6x10E-3 and an average delayed neutron life of 12.7 seconds, how much reactivity was added for the SUR in part a? [1.503 c. What would be the power level if the SUR in part "a' continued into the power range? Explain. A NUMERICAL ANSWER IS NOT REQUIRED. E1.003
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GUESTION 1.12 (3.00) Consider each of the following sets of conditions separately and determine whether each is subcooled, saturated, or superheated and by how much' a. 2235 psis & 610 F E0.503 b. 1100 psia & 435 F E0.503 c. 25.55 in Hg vac. & 128.7 F CO.503 d. At what three locations in the plant would you expect to find the conditions described in parts a, b, & c above? [1,503 (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
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____________________________________________ DUESTION 1.13 (2.50) The plant is at a steady state power level of 33% with the below initial conditions when one of the Steam Generator Main Steam Isolation Valves goes shut due to the Train B dump valve failing.
Calculate the final steady state values for the listed parameters.
. Assume no operator action, no reactor trip, turbine controls in automatic, and rod control in manual. State all assumptions and show all work.
Initial conditions: Tav3 = 555 F Tstm = 538 F Core Delta T = 22 F a. Turbine power __ b. Tavs in the affected loop c. Steam Generator pressure in the affected loop d. Tavs in the non-affected loops o. Steam Generator pressure in the non-affected loops
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 _______________________________________________________ QUESTION 2.01 (3.00) 1. sPW c. What are three reasons for havins Rod Insertion Limits? Cov960 b. With respect to the Rod Insertion Limits, " ..the steamline break
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accident imposes the hishest shutdown marsin requirement."
Explain why this is a true statement. 01.00] c. Considerins each of the following sets of conditions separately, g , ,o which will make the Steamline Break accident worse? [4v460 BOL or EOL Reactor shutdown or at 100% power Tavs at 350 F or at 547 F QUESTION 2.02 (3.00) a. When must the plant be shifted from the cold-les recirculation mode to the hot-les recirculation mode? CO.503 b. Why must the plant shift to the hot-les recirculation mode? [0.50] c. Using Figure 2.1 and the provided hi-liter, show ALL flowpaths when in the hot-les recirculation mode. ,
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QUESTION 2.03 (2.00) Aproximately 4% of the reactor coolant flow bypasses the core and thus is not available for heat removal.
List the four flow paths which bypass the core.
QUESTION 2.04 (2.00) e. What is POPS and what is its purpose? [1.00] b. How is POPS activated? [0.503 hP. What is the setpoint for POPS? CO.503 l (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx) ! .
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 _______________________________________________________ QUESTION 2.05 (1.75) o. What limitations are placed on plant operations if one or more Steam Generator code safety valves is declared inoperable? WHY? [1.003 b. How many code safeties can be declared inoperable and still be allowed power operation? E0.253 c. Why are the code safeties considered "
...the only reliable means of heat removal *? CO.503 OUESTION 2.06 (1.50)
a. List the normal, alternate, and emer3ency sources of water available to the auxiliary feed system. [1.003 b. What is the power supply for the motor driven auxiliary feedwater pumps? [0.503 OUESTION 2.07 (4.50) a. Complete the drawins of the containment spray system using Figure 2.2 Include all major components, valves (valve numbers not required), sources of water, and detectors. [3.53 b. What would be the affect on the containment atmosphere if the Spray Addition Tank contents were NOT injected? E0.53 c. What auto signals will initiate containment spray? Include setpoints and logic. [0.53 (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 i ------------------------------------------------------- QUESTION 2.08 (3.00) Answer the followin3 questions regarding the Steam Dump electro pneumatic centrol system with either (A) always true, (B) sometime true and sometime false, or (C) always false. Assume no malfunctions exist unless specified.
If answer (B) is chosen, briefly explain why.
a. The control air which acts on the steam dump valve diaphragm passes through the positioner and three solenoid valves. [1.003 b. The logic signal which tells the steam dump valve when to open and.how much to open passes throu3h the I/P, converter. E1.003 c. When in the steam pressure mode of operation above the low-low Tave setpoint, a simultaneous failure of both turbine impulse pressure channels (low) will cause steam dump actuation. E1.003 QUESTION 2.09 (1.75)
- How do the control valves in the below systems fail on a loss of Control Air?
(i.e. fail open, fail close, or no affect) Consider each separately.
a. CVCS letdown flow b. RCP seal flow c. RHR . d. AFW alternate water sources e. Pressurizer power operated relief valves f. Reactor Coolant Drain Tank 9 Main Steam Isolation Valves j QUESTION 2.10 (2.50) Unit 2 at Salem has three emergency diesel generators, each supplying its own 4160 volt vital bus. List 5 loads on Bus 2A and 5 loads on Bus 2C. (NOTE: two pumps for the same system are considered one < load)
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3. INSTRUMENTS AND CONTROLS PAGE 9 - ____________________________ " QUESTION 3.01 (3.00) List the interlocks and/or automatic control features, if any, essociated with the below listed valves (refer to Figure 3.1): e. 2CV277 e. 2CV18 b. 2CV4 f. 2CV35
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c. 2CV7 s. 2SJ2 d. 2CV21 h. 2CV71 QUESTION 3.02 (2.50) List the conditions that will automatically initiate the following ESF signals. Include setpoints and logic as apropriate.
a. Safety Injection ('S') 6. Phase B Containment Isolation ('B?) < GUESTION 3.03 (3.00) a. Explain how and why indicated pressuri=er level will change due
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to a leakin3 reference les on a pressurizer level transmitter. [1'.003
= b. Explain how and why indicated pressurizer level wiLJ change due a .-
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to a steam leak inside containment. [1.003 c. The pressurizer master level control channel provides inputs to three CONTROL functions. -List two of them. E1.003 .
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____________________________ QUESTION 3.04 (1.00) What is the significance of the followins two alarms? o. Rod Control Urgent Failure b. Rod Control Non-UrSent Failure QUESTION 3.05 (2.50) o. Will the Overtemperature Delta T and the Overpower Delta T setpoints increase, decrease, or not change for each of the below conditions? Consider each separately. [1.503 1' . Tav3 increases 2. Pressure increases 3. N41 upper detector fails high b. Justify your answers for part a.2 above. E1.003
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QUESTION 3.06 (3.50) The plant is operating at 100% load and you are the Board Operator: a. What indications would YOU have if the high voltage power supply was lost to ONE of the power ranse detectors (both upper and lower)? C1.803 b. What actions must be taken to continue power operation? [1.003 c. With one channel out-of-service already, may surveillance be performed on a second channel while at power? Explain. [0.703 (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) _ .. ._- -
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3. INSTRUMENTS AND CONTROLS PAGE 11 ____________________________ QUESTION 3.07 (2.00) For the below Process Radiation Monitoring Systems, list; purpose protection on alarm, if any type of detector used (Geiser-Mueller, scintillation, ion chamber) c. Steam Generator Blowdown Liquid Monitors b. Letdown Line Monitors GUESTION 3.08 (2.00) Th2 reactor is at 75% power, CBD is at 176 steps. For each of the below conditions, determine rod direction and speed.
Consider each case separately a. Bank Selector Switch (BSS) in Auto, Tav3 = 564 F, Tref = 565 F b. BSS in Auto, Tavs = 564 F, Tref = 560 F c. BSS in Manual, RAISE pushbutton depressed, PT-505 fails low d. BSS in Manual, RAISE pushbutton depressed, PR-NI41 upper fails high QUESTION 3.09 (3.00) Salem's Tavs prostem is a compromise between two extremes; that is, a constant Tavs program and a constant steam pressure Tstm) program.
This compromise minimizes three disadvantages of the constant programs - what are these three disadvantases?
(NOTE: one of the disadvantages is due to the constant Tavs program, ths other two are due to the constant steam pressure program)
GUESTION 3.10 (2.50) c. What two automatic signals will start the diesel senerators? [1.003 b. What will trip the diesel senerators when they are operating in the emergency mode? [1.50] ,
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t 4. PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 12
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____________________ QUESTION- 4.01 (2.25) List the nine major IMMEDIATE actions required by E0P-Trip-1
" Reactor Trip or Safety Injection'
OUESTION~ 4.02 (3.00) In order to maintain the plant at 100% power, work must be performed inside the containment in a radiation field of 850 mrem /hr samma, 30 mrad /hr thermal neutron, 20 mrad /hr fast neutron, and 200 mrad /hr beta. The caintenance man selected is 29 years old and has a life time exposure through last quarter of 53 REM on his NRC Form 4; additionally, he has accumulated 0.5 REM this quarter.
a. Assuming the man wears the appropriate protective clothing, how long
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may he work in this area without exceeding his 10CFR limits? Show all work. [2.003 b. During a declared emergency, this individual volunteers to enter a high radiation area and perform work necessary to prevent further effluent release. In accordance with the Station Procedures, what is his maximum allowed whole body exposure? E0.503 c. Whose authorization is needed in part b. [0.503 OUESTION 4.03 (1.50) Answer the following with respect to AP-3, " Document Control Program *: I a. How lon3 is an On-The-Spot change valid for? E0.503 l l b. Who must approve an On-The-Spot change prior to implementation? E0.503 ( c. Who must approve the On-The-Spot change after implementation? ! When must this aprovel be accomplished by? E0.503 l l r l (xxxxx CATEGORY 04' CONTINUED ON NEXT PAGE xxxxx) ( , L
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____________________ QUESTION 4.04 (3.50) a. List the bases for the below statement out of IOP-3, ' Hot Standby to Minimum Load *? [2.503
'Within fifteen minutes prior to criticality, verify RCS Tavs to be greater than or equal to 541 F' -
b. If, during power operation, Tavs drops below 541 F, what action (s) must be taken as per the Technical S P ecifications? [1.003 OUESTION 4.05 (2.40) Answer the followins in accordance with IOP-2, ' Cold Shutdown to Hot Standby *! a. Below 250 F, at least one RHR pump or one RCP must be in operation.
Why? (List 2 reasons) , E1.203 b. RHR must be isolated from the RCS before temperature reaches 350F or pressure reaches 375 psi 3 Why? (List a reason for each) E1.203
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QUESTION 4.06 (2.00) As a certified reactor operator, assigned to a shift as an extra person, you are required to perform the second verification of a valve alisnment on the Auxi1~iary Feedwater Pump 421.
a. May both the first and second verification be performed to3 ether?t0.503 b. How is the position of a manual valve verified? [1.503 (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) __.
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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14
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____________________ QUESTION 4.07 (3.00) In accordance with EI-4.10, ' Control Room Evacuation", part of the icnediate actions is to station personnel at various locations within the plant to operate equipment or to monitor indications. For the followin,s equipment, state WHERE personnel must so:
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c. Pressurizer heater control b. Auxiliary feedwater tank level indication c. Charging pump start /stop control d. Steam senerator level indication e. Main steam stop valve control GUESTION 4.08 (2.40) Indicate at.what power level each of the below hapenst a. Enter. Mode 1 (power operation) b. P-7 permissive c. P-2 permissAve .- d. Intermediate ranse rod stop 1" -- e. Power ranse lo ranse hi level trip f. P-8 permissive-GUESTION 4.09 (2.20) a. What is the basis for the limit on Axial Flux Difference (AFD) per the Technical Specifications? E1.00] b. Given the below information and assumins that it is 0800, 19 Nov 85, at what time (include date) will it be permissible to raise power greater than 50%? State all assumptions and show all work. .
[1.20]
Time Time (leave (return Date band) band) Power ____________________________________________________________ 19 Nov 85 0716 0800 45% 19 Nov 85 0325 0359 65% 18 Nov 85 1802 1830 80%
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R5656[66I65[~C6 TR6[~~~~~~~~~~~~~~~~~~~~~~~~ ____________________ QUESTION 4.10 (2.75) Answer the following in accordance with OP-II-3.3.8 ' Rapid Boration" ao What^are the preferred and alternate sources of water? E0.503 b. Flow is to be greater than ________ spm. [0.753 co If only one boric acid pump is available, how lon3 will you have to rapid borate for the following: [1.503 Show your work and state your assumptions (1) two rods stuck fully out (2) an unexpected cooldown of 50 F while shutdown (xxxxx END OF CATEGORY 04 xxxxx)
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Equations p= I + 0 Q = MAh' tK,ff 1 + At Q = mcpAT 26 p SUR = t* + (6 p) t Q = UAat C3 (1-K3 ) = C2 (1-K2 ) P = Po 10 sur (t) hl=K 1r = 3.14 P = Po et /t b e = 2.72 SUR = 26.06 CR = .,... t 1-K,ff
.- , = 1* Bp _ p 1 _ CR y
- - p pl p-S M CR
y = h-hf -
= 0.693 t* = 10 5 seconds h th fg )(= s-sf s
fg l Conversions
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I curie = 3.7 x 10" dps 1 kg = 2.2 lbs 1 gal = 3.78 liters I gm/cm3 = 62.4 lbs/ft3 1 in = 2.54 c W 1 ft3 = 7.48 gal-1 yr = 2.15 x 10' see 1 gal = 8.3453 lbm 1 MW = 3.41 x 10' BTU /HR 1 in Hg = 1.037 pf:
= 0. *t 9 s pv a.
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--- iREER55VsARfCs- REAi iEAssFEE As5 FEUi5 FE5E
____________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
. ANSWER 1.01 (2.50) a. Flow is caused by a pressure difference E0.25] due to fluids at different elevations E0.25]. [0.50] b. RCS elta T or = 11 Loa' Delta T CO.3 each] ,/' Core exit the mocoupl (or W temp.) onstant or decto sin 3 Steam Generat pressu e const nt or creasin- consist nt wi h
'
RCS emp c. Voiding in the RCS (durins depressurination) [1.003 REFERENCE Student Notebook, Chapter 1, Reactor Coolant Systemt p3s 17-20 E0P-Trip-5, para 3.11 - Caution ____________ ____,_____________________________________________..______ KRA 000017EK1.01/IF 4.4
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ANSWER 1.02 (2.50) Reactivity due to power increase (Fis 1.1) 250 - 1450=-g200pcm [0.50]
"
Reactivity due to rod withdrawal (Fis 1.2) 825 - 0 = +825 pcm [0.50] Reactivity due to Equilibrium Xenon Change (Fi3 1.3) 1400 - 3180 = -1780 pcm CO.50] Boron worth (Fis 1.7)
=~ -8.2 pcm/ ppm E0.503 Total: -1200pem + 825pem -1780pcm = -8.2 pcm/ ppm (1200 ppm - BC) -2155 pcm'= -8.2 pcm/ ppm (1200 ppm - BC)
BC = (9840 pcm - 2155 pcm)/(8.2 pcm/ ppm) BC = 937.19 ppm CO.503 REFERENCE Rsactor Theory, pg 152 Reactor En3ineerins Manual, Part 1 (Fisures) r __-_--_-______________________________________________-_______-______ KRA 001010K5.21/IF 3.4 s.ot , b carg. wf rg 1 s A4/c. se desea no,y C o- ss c ~ <-4 ) R el h * f icy -
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17
~ ~ ~~~~TEER 66Y 555C57~EEST TRd 5FER 5U6~EEU56~fLUU
____________________________________________ ANSWERS -- SALEM 182 -85/11/19-NORRIS, B. S.
o ANSHER 1.03 ( .75) s C E0.753 REFERENCE . Student Notebook, Chapter,46, Transient Analysis, pss 6-9 Reactor. Theory'_P3s 150-152 _______________ _____________________________________________________ KRA 001000K5.26/IF 3.3 ANSWER 1.04 ( .75) B E0.753 REFERENCE Student Notebook, Chapter 46, Transient Analysis, pgs 10-11 Reactor Theory, pgs 169-170 _____________________________________________________________________ KRA 001000K5.48/IF 3.3 ANSWER- 1.05 ( .75) D ., E0.753 REFERENCE Reactor Theory, ps 203
._____________________________________________________________________
KRA 001000K5.02/IF 2.9 ANSWER 1.06 (1.00) Less E0.503 Hore E0.503
REFERENCE Reactor Theory P3 191
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! KRA 001000K5.30/IF 2.9 I
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 18
--- isEER55isAsiCE- sEsi iEAnsFEE AR5 FEUi5 FE5s
____________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 1.07 (1.00) a. To prevent cavitation of.the pump. [0.503 b. Increase [0.503 REFERENCE G:neral Physics HT/T & FF Fundamentals, pss 319-320 _____________________________________________________________________ KRA Pumps (ps A-9)/IF 3.4 ANSWER 1.08 (2.50) Rafer to Figure 1.9: ,. Pump 41 wi11.rhave zero flow since the system is operating at l a pressure greater than its shutoff headi total flow is found by adding pumps 42 & #3. ~. E1.003 ! ' Flow rate = 68 (+or- 3) spm E0.753 Operating pressure = 86 (+or- 3) psis E0.753 REFERENCE Gzneral Physics HT/T & FF, pgs'324 - 332 , _____________________________________________________________________ KRA Pumps (ps A-9)/IF 2.4
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19
--- isEEs557RERICs- sEAi iEAnsFEE As5 FEUi5 FE5E
_________________________.__________________ ANSWERS -- SALEM 182 -85/11/19-NORRIS, B. S.
l l ' ANSWER 1.09 (2.50) a. Production: I-135 -> Xe-135 CO.25 each] Xe directly from fission Pm-149 -> Sm-149 Removal: 'Xe-135 + n -> Xe-136 Xe-135 -> Cs-135 Sm-149 + n -> Sm-F46/4'O ( b. 1. Hisher fission yield for Xenon and its precursors than for - Samarium's precursors. __ E0.503 2. The microscopic cross-section of absorption is greater for Xenon. [0.503 (Xe 2.7x10E6 barns; Sm 5600 barns)
'
REFERENCE Reactor Theory, pas 219-233 Nuclear Reactor Engineering, Glasstone & Sesonske, Tables A.2 & A.3 _____________________________________________________________________ KRA 001000K5.33/IF 3.2 001000K5.38/IF 3.5 ANSWER 1.10 (2.00) See attached Figure 1.10 REFERENCE Student Notebook, Chapter 46, Transient Analysis, pgs TA-12 & TA-19 Reactor Theory, pas Rx-Th-TP-38.5, & 39.2-39.5 _____________________________________________________________________ KRA 001000K5.13/IF 3.7
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1. PRINCIPLES 0F NUCLEAR POWER PLANT OPERATION, PAGE 20
--- iREER557nARiEs- sEAi iEAREFEE AR5 FEUi5 FE5s
_--___________-_-_-_-____-----_-____________ ANSWERS -- SALEM 182 -85/11/19-NORRIS, B. S.
ANSWER 1.11 (3.25) a. SUR(t) P = Po10 E0.25} SUR(2.5) 7x10E-8 = (2x10E-9)10 CO.253 SUR(2.5) 35 = 10 1.544 = 2.5(SUR) SUR = 0.618 dpm E0.253 b. Beta-eff = 0.006 - [0.253 Lambda-eff = 1/12.7 = 0.08 E0.253
.T =-26.06/SUR E0.253 = 2 6 . 0 6 / 0,. 618,, = 42.17 see CO.253 T= (Beta - rho)/ rho (lambda) ..
A, pf." E0.253 42.17 = (0.006 - rho)/ rho (0.08) rho = 0.006/4.37 = 0.00137 = 137 pcm Yg, f/g E0.253 c. At power levels above POAH E0.253,'6opplerEO.253 and MTCEO.253 will turn power such that the power will increase to a level where the reactivity added by the SUR is equal to the negative reactivity added due to Doppler and HTCEO.253.
REFERENCE Student Notebook, Chapter 46,-Transient Analysis, pgs 42-44 Reactor Theory, pss 134-135
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21
--- isiss55isAsi5s- REAi isissFEE Es5 FEUi5 FE5s
____________________________________________ ANSWERS -- SALEM 182- -85/11/19-NORRIS, B. S.
ANSHER 1.12 (3.00) a. Tsat at 2250 psia = 652.7 F 652.7 - 610 = 42.7 F E0.253 subcooled E0.253 b. Tsat at 1100 poia = 556.28 F 556.28 - 435 = 121.28 F E0.253 subcooled [0.25] c. 25.55 in Hs (0.49Ff b=12.55 psia 14.7 - 12.55 = 2.15 psi ~a E0.253 xif use Table 1* Tsat at 2.15 psia = 128.7 F saturated [0.253 ORuif use Table 2: Tsat at 2.15 psia = 119.1 F 128.7 - 119.1 = 9.6 F E0.103 c superheated [0.153 d. (a) RCS hot les or core exit p,eg%/cc /,c /f r E0.503 (b) u_ igg 7 22 e.c condenser Scotcr outlet (c) Condenser (shell side)
, w //,f-(,yg ,j j, j) CO.503
[0.503 REFERENCE Student Notebook, Chapter 1, Reactor Coolant System, pg 35 Student Notebook, Chapter _34, Condensate & Feed System, ps 55 Gsneral Physics HT/T & FF, pss 83-88
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_____________________________________________________________________ KRA 001000K5.56/IF 4.2-0, S 'D r Ca. , ,
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22
--- isEER557RKsiEs- REAi iEAREFEE As5 FEUi5 FE5s
____________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 1.13 (2.50) a. Turbine power
^ * g ,,,7, g ,, g [ 3, n c e
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b. Tav3 (affected) ad 7. r,. 4 final: goes to Thot of 566 F C0.50] c S/G pressure (affected) initial: Psat at 538 F = 947 psia final: Psat at 566 F = 1189 psia E0.50] d. Tavs (non-affected) - Delta T must increase by 1/3 => final Delta T = 29.3F final: Tavs = Th - (Delta T)/2 = 566 - 29.3/2 = 551.33 F [0.503 e. S/G pressure'(non-affected)
(Tavs - Tstm) must increase by 1/3 => final Tavs - Tstm = 22.67 F therefore, Tstm = 551.33 - 22.67 = 528.66 F final: Pstm at 528.66 F = 901.3 psia E0.501 REFERENCE Student Notebook, Chapter 1, Reactor Coolant System, pg 35 Student Notebook, Chapter 5, Steam Generator, pg 33 Student Notebook, Chapter 22, Rod Control, pg RS-3
_____________________________________________________________________ K8A 035010K1.09/IF 3.8
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 _______________________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 2.01 (3.00) o. 5.3 e. To ensure adequate shutdown margin [4,44 each] To minimize the reactivity effect of a rod ejection accident To minimize radial flux peaking factors k .M r. poor, altd I.'ae.*/4) b. Due to the large value of positive reactivity inserted by MTC during the resultin3 uncontrolled RCS cooldown. E1.00] c. EOL E-Gre6 each3 Shutdown a J.5 547 F REFERENCE Student Notebook, Chapter 23, Rod Position Indication, pss 20-21 _____________________________________________________________________ K8A 001000k5.08/IF 3.9 ANSWER 2.02 (3.00) a. 1. When cold-les recirc has been in operation for 22.5 hours [0.25] 2. When boron concentration decreases m.~ " - CO.253
~~ ~
b. (r- To quench the steam bubble in the vessel head E0.253 2. To flush the boron off the fuel rods and back into solution E0.253
'
c. See Figure 2.1 E2.00] REFERENCE Student Notebook, Chapter 10, Emergency Core Cooling System, pas 53-64 _____________________________________________________________________ KRA 006030K4.03/IF 3.4 006030A1.03/IF 3.6 ANSWER 2.03 (2.00) 1. Nozzel bypass flow E0.50 each] 2. Control rod & thimble bypass flows 3. Baffle wall bypass flow 4. Head coolins bypass flow REFERENCE . Student Notebook, Chapter 3, Reactor Vessel & Internals, pg 19 ____-________________________________________________________________
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- s 2.
kPLANTDESIGNINCLUDINGSAFETYANDEMERGENCYSYSTEMS PAGE 24 ____.__________________________________________________ ANSJERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
IMA-002000K6.13/IF 2.3 ANSAER 2.04 (2.00) s , , i "e . Pressurizer Overpressure Protection System E0.503 provided to prevent brittle fracture of the RCS at low pressure to.503 b. The system must be manually activated by'the operator E0.503
$P. 375 psi 3 (+/- 10 psis) E0.503 REFERENCE Student Notebook, Chapter 25, Pressurizer Pressure & Level Control, .PSs 14-16,,
_____________________________________________________________________ KSA 002000K4.10/IF 4.2 010000K4.03/IF 3.8
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ANSWER 2.05 (1.75) a .- Plant power is restricted [0.503 due to the reduced heat removal capability of the secondary E0.503 b. 3 per senerator ,,
[0. 2 53 ,,
c. the code safeties required no outside motive force for actuation [0.503 REFERENCE Student Notebook, Chapter 5, Steam Generator System, PS 16 Student Notebook, Chapter 33, Main Steam, pss 11-12 ______________________________ ______________________________________ KRA 035010SGtS/IF 3.1
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ANSWER 2.06 (1.50) c. normal - auxiliary feed water storase tank E0.25 each3 alternate - demineralized water storage tank emergency - fresh water / fire ~ protection storage tanks
- service water b. MDAFWP #21 - 4160 vital bus 2A [0.25 each3 .MDAFWP #22 - 4160 vital bus 2B REFERENCE Student Notebook, Chapter 11, Auxiliary Feed System, pgs 15, 51, AF-5 Student Notebook, Chapter 40, Electrical Distribution, pg 70
_____________________________________________________________________
'KSA 061000K1.07/IF 3.6 061000K2.02 & 2.03/IF 3.7 061000K4.05/IF 4.5 (
ANSWER- 2.07 (4.50) e. See attached Figure 2.2 E3.503 b. The iodine removal' process would not be as effective due to the lack of.the sodium-hydroxide ( NaOH raises pH from 9.5 to 11.0) [0.503 c. High-High containment pressure E0.303 23.5 psis E0.103 2/4 sensors E0.103 REFERENCE Student Notebook, Chapter 12, Containment & Containment Spray, PSs 1(#9), 13, 20-21, & CS-7 _____________________________________________________________________ KRA 026020A1.01/IF 3.1 026020SGt9/IF 3.6 '^
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ANSWERS -- SALEM-1&2 -85/11/19-NORRIS, B. S.
ANSWER 2.08 (3.00) a. B (when energized, the positioner is bypassed and supply air pops-open the trip-open valve) [1.00] ' b. D trip open logic bypasses the I/P converter C1.00] c. C (turbine impulse pressure is not an input in the Psta mode) [1.00] REFERENCE Student Notebook, Chapter 26, Steam Dump System, pss 11 & SD-2 _____________________________________________________________________ KRA 041020K4.17/IF 3.7 041020K6.03/IF 2.7 ANSWER 2.09 (1.75) c. FC (CV2 lt CV277) CO.25 each] b. F0 (CV104) c. F0 (RH18) ~ d. FC (AF52) e. FC (PR1 & 2)' f. FC (WL12) ' '~ ..~ s. Ne .TTmmt ft-
'
alcctro hyd,ovliu) fort $ I64 4 s9h REFERENCE Student Notebook, Chapter 6, CVCS, ps 8 4, RCP' PS 22 8, RHR, pg 16 11, AFW' PS 18
, 25, Pr Pres & Lvl Cntr1, pg 12 29, Rad Liq Waste, ps WL-2 33, Mn Stm, pg 16
______-______________________-__________________-_---__________---__- KRA 078000K3.02/IF 3.4
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 _______________________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 2.10 (2.50)
(five required from each Bus) E0.25 each3 Bus 2A Bus 2C
______ ______ Auxiliary Feedwater Pump (#21) Containment Spray Pump (#22) Containment Spray Pump (421) Service Water Pumps (#25 & 26) Service Water Pumps (#21 & 22) Safety Injection Pump (#22)
-Residual Heat Removal Pump (#21) Component Coolins Pump (#23)
Safety Injection Pump (#21) Centrifugal Chargins Pump (#22) Component Cooling Pump (421) 460 volt Vital Bus 2C 460 volt Vital Bus 2A 230 volt Vital Bus 2C 230 volt Vital Bus 2A REFERENCE Student Notebook, Chapter 42, Electrical Distribution, ps 70 _______________________________________u_____________________________ KRA 064000K3.03/IF 3.6
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ANSWER 3.01 (3.00) a. -not able to open unless P=r level > 17% E0.20 each]
-closes if Pzr level < 17%
b. -not able to open unless at least one charging pump is running-not able to open unless P:t level > 17%
-not able to open unless 2CV2 AND 2CV277 open- -closes if Pzt level < 17% -closes if Phase A isolation-closes if 2CV2 OR 2CV277 closes c. -closes on Phase A isolation d.. -diverts letdown flow around the demineralizers on'high temp (120 F)
out of the letdown heat exchanger e. -throttles to maintain pressure (350 psis) in the letdown heat exchanger
.f. -modulates letdown flow to maintain level in the VCT . ~~
3 -Opens on-low level in the VCT-opens on an 'S' signal h. -none REFERENCE Student Notebook, Chapter'6, Chemical & Volume Control System, pgs 8-19 , _____________________________________________________________________ KRA 0040x0K4.xx/IF 3.0
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l ANSWERS -- SALEM 182 -85/11/19-NORRIS, B. S.
ANSWER 3.02 (2.50) a. 1. Low Pressurizer Pressure E0.203
< 1765 psis E0.203 2/3 sensors E0.103 2. High Containment Pressure [0.203 > 4.0 psis E0.203 2/3 sensors E0.103 3. e Differential Pressure E0.203 High > 100Steam 1ipYe remaining S/Gs psi $14 E0.203 2/3 sensors in one S/G E0.103 4. High Steamline Flow (coincident with) [0.103 LowLow Tavs OR- [0.053 Low Steamline Pressure , E0.053 Y' Pft Steamflog) 2/4 Tavgi 2/X Steam pres. sensors E0.103 # / Steamflow > 40% when at 0 - 20% load 8" 40 - 110% when at 20 - 100% load CO.103 diabiD Tavs < 543 F E0.053 ' Steam Pressure < 500 psis E0.05]
b. High High Containment Pressure E0.203 23.5 psis CO.203 2/4 sensors E0.103 REFERENCE Student Notebook, Chapter 10,. Emergency. Core Cooling System, pas 8-9
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ANSWER 3.03 (3.00) a. Per level will indicate higher than actual [0.503 because the delta P across the bellows will be decreasin3 [0.503 b. Pzr level will indicate higher than actual E0.503 due to the steam heatin3 the reference leg which causes the
bellows to-open which appears as a loss of fluid E0.503 c. CVCS char i3 n3 flow Eany two .t 0.50 each3 Back/up heaters Level alarms REFERENCE Student Notebook, Chapter 25, Pressurizer Pressure & Level Control, pas 25-26 _____________________________________________________________________ KRA 011000K1.xx/IF 3.7 000028EA2.12/IF 3.1 ANSWER 3.04 (1.00) a. operational problems exist in the logic and/or power cabinets E0.303 it affects the system ability to' hold or move control rods E0.203 b. problems within the auxiliary power supply to the logic and/or power cabinets ~ E0.303 no immediate effect [0.203
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REFERENCE .. _- Student Notebook, Chapter 22, Rod Control System, pg 47 _____________________________________________________________________ KRA 001000K6.04/IF 2.4
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ANSWER 3.05 (2.50) a. OTDT OPDT E0.25 each] Decrease Decrease Increase No change Decrease No change b. OTDT - as pressure increases, you are further away from DNB E0.503 OPDT pressure has no effect on power REFERENCE Student Notebook, Chapter 24, RCS Temperature Indication System, pas 9-13 _____________________________________________________________________ KRA 012000K4.02/IF 3.9 ANSWER 3.06 (3.50) a. 'PR Loss of Detector Voltage * alarm (window D-7) [0.30 each]
' Upper (& Lower) Section Deviation above 50%" alarm (windows D-37/45) ' Power Range Channel Deviaiton" bistable (window D-39) ' Power Ranse High Neutron Flux Rate (Nes)* bistable (unidow D-38)
Indication to zero (meter & recording pen) Power-on. lamp out on the Excore NI Cabinet e/ 23 . rs b. Bypase/ the failed channeV for control E 0'. &O3 and protection EO . 593 c.
m16tn one hur CAs.Q Yes E0.20] T.S. allows one additional channel to be bypassed for up to two hours for surveillance testin3 [0.503 REFERENCE
"tudent Notebook, Chapter 20, Excore Nuclear Instrumentation, pss 17-24 Escrsency Instruction I-4.19, pss 6-7 Technical Specification 3/4.3.1, pg 3/4 3-5
_____________________________________________________________________ KRA 015000K3.01/IF 3.9 015000A2.01/IF 3.5 I l
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ANSWER 3.07 (2.00) a. Purpose primary to secondary leak detection CO.403 protection - B/D isolation valve will colse on the alarming S/G CO.403 detector - scintillation CO.203 b. purpose - detection of failed fuel CO.403 protection - none CO.403 detector - UNIT 1 - scintillation CO.10] UNIT 2 - ion chamber CO.103 REFERENCE Student Notebook, Chapter 17, Radiation Monitoring System, pas 23-24 _____________________________________________________________________ KRA 068000A4.04/IF 3.8
'073000K4.01/IF 4.0 073000K4.02/IF 3.3
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ANSWER 3.08 (2.00) a. No motion (in the deadband) CO.50 each3 b. IN at 40 spm (acceptable for speed if answer middle of ramp) c. DUT at 48 spa (blocked in auto only) d. No motion (overpower rod stop) REFERENCE Student Notebook, Chapter 22, Rod Control, pss 25-29, 69-71, .RS-10 & RS-11 _____________________________________________________________________ KaA 001000K4.02/IF 3.8
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__________________________ _
-ANSWERS -- SALEM 1&2 -85/11/19-NORRISr B.-S.
ANSWER 3.09 (3.00) 1. Constant Tavs program causes poor turbine performance E1.003 (due to the low secondary steam pressure at high power levels) 2. Constant steam pressure program requires a very larse pressurizerE1.003 (due to the wide ranse that Tavs must cover) 3. Constant steam pressure program also causes Tavs to approach E1.003 limits (that could cause bulk boilins or excessive fuel centerline temperatures) REFERENCE Student Notebook, Chapter 22, Rod Control, pas 6-7 _____________________________________________________________________ KSA 002000K5.11/IF 4.0 ANSWER 3.10 (2.50) a. safety injection signal [3.503 loss of all offsite power to the 4160 volt vital buses (84 94ko" r) [0.503 b. senerator differential "" CO.40] engine overspeed E0.403 engine low low oil pressure ^ E0.403-c ;; r 3 c c.c y :tcp , > h b :.: t t e . i l u u o l i Br.% d.*[[*N M* / * E0.303 REFERENCE Student Notebook, Chapter 41, Diesel Generatore pas 36 & 50 _____________________________________________________________________.
KSA 064000K4.02/IF 3.9 064000A3.01/IF 4.1
. - . , .
'
. .
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 _____-_________-____----______--__-___-______--_ RADIOLOGICAL CONTROL __________-_-------_ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 4.01 (2.25) 1. Manually trip the reactor E0.25 each, all 9 required] 2. Check all turbine stop valves closed 3. Check total AFW flow > 440,000 lbm/hr 4. Confirm reactor trip 5. Check at least two 4Kv vital buses energized 6. Check if SI required 7. Check for SEC loadins for enersized vital buses 8. Check safeguards valves positioned properly 9. Check containment pressure < 23.5 psis REFERENCE . , . E0P-Trip-1, Reactor Trip or Safety Injection, P3s 2-5 __________---___--_-________-------__--____________________--_______- KRA 000007EA2.02/IF 453 ANSWER 4.02 (3
.. ,.,. 00)
c. 5(N-18) = 55 REM E0.25] Total lifetime to date = 53 + 0.5 = 53.5 rem Total lifetime available = 55 - 53.5 = 1.5 rem E0.253 Total this quarter available = 3 - 0.5 = 2.5 rem E0.253 Lifetime is more restrictive than quarterly limit Samma 850 mrem /hr = 850 mrem /hr E0.253 th neut 30 mrad /hr x 30F = 90 mrem /hr E0.253 f neut 20 mrad /hr x 100F = 200 mrem /hr E0.253 4 beta negligible d've to protective clothing E0.253 ____-_________--______--__________________-__ total =1140 mrem /hr total dose ratn 1.5 rem /(1.14 rem /hr) = 1.32 hr = 79 min E0.253 b. 25 REM whole body one time exposure CO.503 c. General Mana3er, Salem Operations E0.503 REFERENCE 10 CFR 20 AP No. 24, Radiological Protection Program, pss 5 & 14 _________________-__-----____--_-______---___----__-___--__-_-------- KRA Plant Wide Generic 415/IF 3.4 y ,-f mk a,,4 ha o( *hy4- cagy ? eJc.d./un s.siW e4sj e no such: A Ad eva//Ar Y /ff * o7td m&ki =/ /340 mecan/A r- 64/
"'N /*fS% = /.Ibe = dimin t. nrm /4,
_
. * - . . . ' .
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
- ~~~~~~~~~~~~~~~~~~~~~~~~ --- E 5i5E55iEAL c5sTR5L
_-_-------____-_____ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 4.03 (1.50) e. Permanent changes - valid until incorporated into next review E0.253 Temporary changes - valid up to the effective date on the chanSe notice E0.253 b. Supervisor in charge on the work E0.25] Senior Shift Supervisor or Shift Supervisor on duty [0.253 c. Same level as original procedure E0.25] within 14' days CO.253 REFERENCE '/' AP No. 3, Document Control Program, pgs 3-4'
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KAA Plant Wide Generic # 21/IF 3.8 ANSWER 4.04 (3.50) a.1. MTC is within the analyzed temperatura ranse E0.50 each] 2. Protective instrumentation is within r.ormal operating vanse 3. P-12 (Tavs > 543 F) above its setpoint 4. Pressurizer is operable 5. Reactor vessel is above minimum temperature (RT-NDT) > b.1. Restore Tavs > 541 within 15 minutes, or E0.503 2. Be in hot standby witnin next 15 minutes [0.503' REFERENCE IOP-3, Hot Standby to Minimum Load, ps 6 TS' P3s 3/4 1-6 & B3/4 1-2
---------------------------------------------------------------------
KRA 001000K5.15/IF 3.4 001000K5.16/IF 3.4 f
. -#
. ' . . .
-
.
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
- ~~~~R A5i5L65isst C5UTR5L'~~~~~~~~~~~~~~~~~~~~~~~
____________________ ANSWERS -- SALEM 1&2 -85/11/19-NORRIS, B. S.
ANSWER 4.05 (2.40) a. coolins /)4yf dI4b7 E0.603 reliable temperature indication E0.603 kwetret 4 ire h 4:<= * M b. temperature - heat removal capability of RHR HX E0.603 pressure - design pressure of RHR system (when combined with [0.60] discharse pressure of RHR pump) REFERENCE IOP-2, Cold Shutdown to Hot Standby, ps 2 Student Notebook, Cha 8 ,- R e s i d u a l H e a t R e m o v a l , pgs 5-6 171_ JL y4Lf-JL _ _ _ _ _ _ _ _ _______________________________________________ _ _p t e r KRA 000002EK3.02/IF 3.03 ANSWER 4.06 (2.00) - a. No E0.50] b. verified closed - attempt to close [0.50 each] verified open - move in closed direction, then reopen verified throttled - fully close, then open required number of turns REFERENCE OD-7, Valve Operations & Systems Alignment, pg 2 _____________________________________________________________________ KRA Plant Wide Generic /IF 3.7 ANSWER 4.07 (3.00) n. Panels 2GP & 2EP - Pressurizer Heaters E0.60 each]
(electrical penetration - elev 78'-)
b. Panel 379 - Auxiliary Feedwater Storase Tank Panel (outside auxiliary building - elev 100') c. Panel'213 - Hot Shutdown Panel (auxiliary building - elev 04') d. Panel 213 - Hot Shutdown Panel (auxiliary buildin3 - elev 84') o. Panel 687 - MSSV Local Control Station (c.;}-th N a ;;:.:th penetration areas - elev te f onor 100')
%c/ Lehu ,,af My~h 0
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1 _ .. m .A ,
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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37 ________________________________________________ RADIOLOGICAL CONTROL ____________________ ANSWERS -- SALEM 182 -85/11/19-NORRIS, B. S.
REFERENCE EI-4.10, Control Room Evacuation, pss TBL 1-1 to TBL 1-3 _____________________________________________________________________ KRA 000068EA1.12/IF 4.4 ANSWER 4.08 (2.40) e. 5% [0.40 each3 b. 10% c. 15% . d. 20% e. 25% f. 36%
' 50% '*' 2%)
1- r-icr to 5^%
'
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'
2. ? 05% / 2%)
, '9^X REFERENCE IOP-4, Power Operation, pas 3 & TBL 1-1 to 1-2
_____________________________________________________________________ KRA System Wide Generic #12/IF 3.5 ANSWER 4.09 (2.20) o. Assures that F0(2) E0.503 (upper bound envelope of 2.32 times the normalized axial peakins factor) is not exceeded during either normal [0.253 or in the event of xenon [0.253 redistribution followins power changes.
b. 19 Nov (0800 - 0716) x (0.5) = 22 penalty minutes E0.253 19 Nov (0359 - 0325) x (1.0) = 34 penalty minutes [0.253 60 - 22 - 34 = 4 penalty minutes available CO.253 1830 - 4 = 1826 on 18 Nov 85 started current 60 minutes [0.253 Can so greater than 50% anytime after 1826 on 19 Nov'85 CO.203 ,. .
-~ - , .
. . ., . .
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 38
- ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R 65bl65iEst E5 TRUL
_____________-___-__ ANSHERS -- SALEM 152 -85/11/19-NORRIS, B. S.
REFERENCE TS, pgs B 3/4 2-1 & 2-2 _____--_-_______----______------_____-_______---_______--_____-----__ KRA 001000A3.03/IF 3.6 ANSHER 4.10 (2.75) a. BAST CO.253 RHST E0.25] b. 70 spm CO.75] c. with one pump it takes 1 minute to inject 200 pcm E0.50]
(1) (2 rods)x(1500 pcm/ rod)x(1 min /200 pcm) = 15 min [0.50] (2) (50 F)x(20 pcm/1 F)x(1 min /200 pcm) = 5 min E0.503 REFERENCE OP-II-3.3.8, Rapid Boration, pgs 1-3
______________________________________-______________________________ KRA 000024EA2.05/IF 3.3 000024EK1.01/IF 3.4 000029EK3.11/IF 4.2 m xum.
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7n /) STER C o P bl U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR; LICENSE EXAMINATION FACILITY: SALEM 1&2j ___________________-____ REACTOR TYPE: PWR-WEC4 _____________________-__- DATE ADMINISTERED: 85/11/19 , __________J_____________.
EXAMINER: DUDLEY _________________________ APPLICANT * _________________________ INSTRUCTIONS TO APPLICANT: __________________________ 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 Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up sin (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY ________ ______ _______-___ _-______ ___________________________________ 25.00 25.00 ________ ______ ___________ _------- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 PLANT SYSTEMS DESIGN, CONTROL, ________ ______ ___________ ________ 6.
AND INSTRUMENTATION
e66 00 25.00 PROCEDURES - NORMAL, ABNORMAL, ________ ______ ___________ ________ 7.
EMERCENCY AND RADIOLOGICAL CONTROL 25.00 25.00 ADMINISTRATIVE PROCEDURES, ________ ______ ___________ ________ 8.
CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS ________ ______ ___________ ________ FINAL GRADE _________________% All work done on this examination is my own. I have neither 3ivan nor received aid.
~~~~~~~~~~~~~~ 5PPLIC5UT I5~55GU5TURE
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.'5 . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 . ____7 g ggggg gjgg______________________________________
;
______________ QUESTION 5.01 (2.00) a. List two causes of waterhammer. (0.8) b. Give two examples of how waterhammer can be minimited. (1.2) GUESTION 5.02 (2.50) The Unit i reactor is operating at 50% power, BOL, when a steam dump fails open. Assume rods are in manual, no operator action is taken, and no reactor trip occurs. Explain HOW~and WHY reactor power and Tave will change.
QUESTION -5.03 (2.00) a. As fuel temperature increases the reasonance absorption peaks for U-238 become lower in height and the bands broaden but the-area under the curve remains theoretically constant. Why then is heatup of the fuel a negative reactivity effect? b. Does the doppler COEFFICIENT become more or less negative as fuel temperature increases? EXPLAIN.
+ QUESTION 5.04 (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 expected.
a. Th b. Subcooling c. Steam Generator Pressure d. Steam Generator Level e. Pressurizer Level (***** CATEGORY 05 CONTINUED ON NEXT PAGE mauxx)
>.
. . '
. 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
---- -------------------------------------- --------_--_-_
GUESTION 5.05 (2.50) c. Explain why an ICRR is required for withdrawal of shutdown banks if count rate triples and is required during boron dilution when the countrate only doubles? (1.0) b. When is an ICRR required during withdrawal of control banks? (1.5) OUESTION 5.06 (3.00) a. What will happen to the current drawn by a reactor coolant pump as the system is heated from 200 des F to 564 des F? Explain your ansuer.
b. What would happen to the current drawn by a reactor coolant pump if voids form in the reactor core during accident conditions? Explain your answer.
QUESTION 5.07 (3.00) Indicate at which time in core life (BOL or EOL) the following cecidents are more severe or result in a longer time spent a t' higher power. Justify your answer.
a. Main steam line break b. Total loss of coolanj flow c. Rod withdrawal accident from low in the source range prior to any significant reactor coolant temperature increase.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE ****x)
. . * . . ' .
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
----- -------------------------------------- --_-----------
00ESTION 5.08 (3.00) After operating for several weeks at 100% power, a reactor at BOL is shutdown for maintenance. An Estimated Critical Position (ECP) is calculated for a startup to be conducted 12 hours after the shutdown. EXPLAIN why the Actual Critical Position (ACP) is HIGHER than, LOWER than, or the SAME as the ECP if the following occur (consider each independently) a. The Steam Dump Pressure Controller is changed to 900 psia after the ECP calculation, b. The startup is delayed 4 hours.
c. The rod worth curves used in the ECP apply for HFP not HZP.
QUESTION 5.09 (4.00) Assume Unit I has just tripped after operating at 75% power for three days. The control rods were at 154 steps on Bank D, the baron concentration remains constant during the event at 1200 ppa,
temperataure is bein3 controlled at 547 F, and the core ts near the beginning of cycle. Using the 3r,aphs provided answer the following questions, state all additional assumptions and show all calculations.
a. What is the margin by which the plant is shutdown. (1.5) b. What is the difference between the margin by which the plant is shut down and shutdown margin as defined in Technical Specifications? (1.0) c. How much makeup water most be added if the reactor is_to be restarted 8 hours after the trip? Assume rods will be at at 154 steps on Bank D. (1.5)
(mmmmm END OF CATEGORY 05 *****)
. . * , . ~ . .
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . QUESTION 6.01 (1.50) Answer the followin3 questions concerning the Reactor Coolant Pumps a. List two design features which protect the RPCCW system from a failure / rupture in the thermal barrier cooling coil? (1.0) b. Explain why VCT pressure is important to the operation of the RCP seals. (0.5) DUESTION 6.02 (2.50) The reactor is operating at a steady state 25% power, all control systems are in automatic. Turbine load is increased to 100% and the steam pressure detector for 91 S/G sticks at the 25% value. Explain the signal processing including all steps of the resulting transient and ending at the final stable conditions assuming no operator actions.
QUESTION 6.03 (2.50) Answer the following questions: a. What limits are the following RCS trips designed to protect against? 1) OTDelta T 2) OPDelta T 3) Low pressuriner pressure 4) Hi pressurizer prppsvre [0.4 each] b. For a given RCS temperature, how does the low Pressurizer pressure trip serve to limit the range of OTDelta T setpoint? [0.93 OUESTION 6.04 (2.00) For each of the following Unit 1 area radiation monitors indicate what interlocks, if any, are actuated at the high alarm set points.
a. Control room (1R1A) b. Fuel storage area (1R9) c. Containment (1R21: 1R44A, D) d. Fuel handling crane (1RfDh,B)
(***** CATEGORY 06 CONTTNUED ON NEXT PAGE *****) .
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; 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 ,.
.
QUESTION 6.05 (2.00) Indicate on the attached figure the expected electrical lineup i for each of the following conditions.
a. No. 2 Main Generator Synchronized
- b. Both Main Generators off the line c. Blackout i
!, QUESTION 6.06 (3.00) a. What three pieces of equipment are protected by D.C. oil pumps during a loss of power accident? (1.0) b. What five support systems are needed for cooldown after a loss of power accident? A n W A. aras Ae f(t Att AMioidoc, (2.0) ! , 00ESTION - 6.07 (2.00)
;
What would be the approximate leak rate if a Loss of Coolant Accident occurred, all automatic safety injection systems functioned properly, pressurizer level stabalized, and RCS pressure stabolized i at 1700 psis? Justify your answer.
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QUESTION 4.08 (3.00) l
: Answer the following questions concerning the Power Range NI's:
' o. What happens when N41 is selected on the power mismatch bypass switch? List 3 systems which are affected.
b. List two reasons why the detector current comparator alarm , function is automatically defeated when power is below 50%. ; I
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.
2 (2 (3) (3) 2T60 1T60 ;
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 , ______________________________________________________ GUESTION 6.09 (3.50) What are the inputs or conditons which would cause actuation of the following Safeguard Equipment Control (SEC) signals. Provide set points and coincidence were applicable.
a. Mode li Accident only (2.7) b. Mode 28 Dlackout only (0.8) DUESTION 6.10 (3.00)
'
c. In what positions of-the Rod Control System Dank Selector Switch is ' Dank Overlap * in service? (0.5) b. At what Bank B position should Bank C rods begin to be withdrawn during sequence operation? (0.5) c. DRIEFLY EXPLAIN why the Rod Control System Startup Pushbutton is not used when recovering from a dropped control rod at power. (0.5) d. Indicate the direction of rod motion (IN, OUT OR NONE) if the following instrument failures occur with RCCS in AUTOMATIC control at 50% power.
1) Loop 3 Teold input to Tave fails LOW.
2) Turbine impulse pressure fails LOW.
3) Power range input channel fails high instantanicusly. (1.5)
(***** END OF CATECORY 06 *****)
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, 7. PROCEDURES - NORMAL, ABNORNAL, EMERGENCY AND PAGE 8
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QUESTION 7.01 (1.50) After criticality is achieved, loop 1 Tave is noticed to be less than the minimum required temperature for critical operations. All other temperature channels are above this limit. What actions, if any, should be taken? . QUESTION 7.02 ( .60) If two rods do not fully insert on a reactor trip, how long should the reactor operator be directed to rapid borate?
(Choose the most correct answer.)
A. Until RCS boron level increases by 150 ppm D. Until Keff is 0.95 C. Until directed otherwise by Shift Supervisor D. Until 16 minutes have passed 00EST-10lf-7T03 62T00M What-TWO-condi41 ons-might-resul t-i f-21-RCP-and_-24-RCP-were- the 4 enLy-r-eac tor-. coolant _. pumps -opsr.at.ing dur ing_couldown belou_350_I.?,, 00ESTION 7 04 (2.00) What are FOUR plant conditions which place the plant on a RED PATH and requires the operator to utilize the status tree? QUESTION 7.05 (2.40) a. Explain why RCS pressure may decrease and stabilize at a now equilibrium value when Safeguards Pumps are stopped.
b. Explain why RCS subcooling requirements increase as the number of running Safeguards Pumps are decreased.
(m**** CATEGORY 07 CONTINUED ON NEXT PAGE unsus) _ .
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 9
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____________________ GUESTION 7.06 (2.00) o. When must Reactor Coolant Pump 21 be secured if all component cooling water is lost to the pump while Unit 2 is operating at 80% power. (1.5) b. What is the maximum amount of time a reactor coolant pump can be operated if both seal injection and component cooling water are lost and no temperature limits are exceeded? (0.5) GUESTION 7.07 (2.00) o. What is the definition of a partial loss of Reactor Coolant? (0.8) b. At what point during a partial loss of Reactor Coolant accident should Safety Injection be initiated and the ' Safety Injection Initiation * procedure, EI-4.0, be entered? (1.2) GUESTION 7.00 (3.00) An entry into the containment is required while at 100% power and will result in an ettimated whole body dose of 120 mrem.
Tha following four candidates are equally qualified to perform tho task. Which candidates may be allowed to perform the task in accordance with administrative procedures? .. Explain your reasons for accepting or rejecting each candidate.
No waivers can be obtained.
CANDIDATE 1 2 3 4 SEX male male female male AGE 20 30 24 27 HK/ EXPOSURE 100 mrem 30 mrem Omrem 150mrom OT/EXp0SURE 1900 mrem 000mrom 20mrom ? ACCUM LIFE EXp0SURE 9900 mrem 4000 mrem 2200mrom ? REMARKS none None 3 months History pregnant unavailable (***mm CATEGORY 07 CONTINUED ON NEXT PAGE ===**)
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. 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10
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Rd555L55iEAt C5sTR5t ____________________ 00ESTION 7.09 (3.00) If Unit 2 is at 100% power, what Technical Specifications should be consulted for immediate (one hour) action statements for each of the following! a. Loss of 2C 115 V Vital Instrument Dus (THREE REQUIRED) (1.5) b. Loss of 2C 125 VDC Bus (ONE REQUIRED) (0 5) c. Loss of 2C 4 KV Vital Bus (TWO REQUIRED) (1.0)
.
QUESTION 7 10 (3.50) For each of the following situations indicate whether the reactor should be tripped, shutdown, operated at reduced power, or operated at 100% power. Also indicate what deteriorated conditions, if any, would require an immediate reactor trip. Assume plant is at 100% power and consider each situation seperately.
a. Turbine output is 90% when reactor power is 100% due to a steam leak downstream of the MSIV.
b. Circulating, water flow to the condenser is reduced due to pump failures. Vacuum has decreased to 23 inches.
c. Pressoriner heaters are deenergized and cannot be energized from their normal power supply.
d. Control air pressure drops to 80 psig.
e. Both seals of the %e Mer,cyr-eseepa. hatch are found to be leaking excessively. (gurent 00ESTION 7.11 (3.00) In accordance with E0P-TRIP-1, what steps should be taken if a roactor trip is not confirmed?
(***** END OF CATEGORY 07 mus**)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 11
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QUESTION 8.01 (1.50) During a backshift work is being done to replace and reposition supports for the discharge piping of an HPSI pump. The workers cannot find the supports specified in the work package and have determined that it is impossible to install the supports at the required locations.
The workers have located some pipe hangers in the shop and want permission to install them as close as possible to the specified locations. What actions, if any, should the shift supervisor take? Support your answer.
i GUESTION 0.02 (2.50) What two people must the Senior Shift Supervisor contact or no;ify fo11owin3 a reactor trip and what type of notifications or requests should be made? l OUESTION 8.03 (2.00)
- Following watch relief, may the SNSS order the oncoming SRO (NSS)
from Unit 1 to relieve the off-going SRO (NSS) from Unit 2 if the rolief for.the Unit 2 SRO is expected to arrive within half en hour. Justify your decision assuming both plants are at power and the STA's are unlicensed.
! OUESTION 0 04 (2.00) What four actions or reports must be completed prior to allowing Unit 1 to return to unrestricted operations following a loss nf fcedwater transient which resulted in indicated RCS pressure of 2750 psig.
QUESTION 0.05 (2.00) What are four of the five basis for the requirement to maintain RCS temperature above 541 F during critical operations?
(assus CATEGORY 00 CONTINUED ON NEXT PAGE sussa)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 12
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GUESTION 8.06 (2.00) e. What person or organization is responsible for scheduling Technical Specification surveillance items? (0.5) b. What extensions, if any, are allowed for completing Technical Specification surveillance items beyond the specified time interval? (1.5) GUESTION 0.07 (2.50) What type of safety ok jumper / lifted lead tags, if any, are required for each of the following situations? a. Leads need to be lifted in accordance with an installation procedure for a new input to the annoniator panel.
b. A blind flange is to be installed in the CVCS system while a relief I valve is being calibrated.
l c. Work is to be conducted on the Reactor Coolant Pump motor which will require jogging the pump.
I d. Work is being done on the ventilation system which requires making damper adjustments with no flow and then checking that the required flow has been established.
l l e. Following an outage it is discovered that the hand switches for the spray valves have been incorrectly wired so that each switch operates the opposite spray valve. Work is to be done to correct the wiring problem.
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(***** CATEGORY 00 CONTINUED ON NEXT PACE *****)
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. 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13 _________________________________________________________ . QUESTION 8.08 (2.50) Should each of the following requests be allowed to be conducted? Justify your decision. Unit 1 is in mode 6 and Unit 2 is at 100% power. Consider each request separately.
a. A request to conduct an approved modification on a safety nuclear instrument channel which will equire deenergizing the detector. Over Power Delta T channel C is t r i pped . t/ f.'I T2 b. A request to commence the procedure for replacement of the reactor vessel head while boHy pressurizer code safety valves are inoperable.
an L c. A request to replace the governor on Unit 2 Emergency Diesel Generator number 2.
d. A maintenance request to replace the s e a l t, on both air lock doors on Unit 1.
e. A maintenance request to replace the HEPA filters in the Auxiliary Building ventilation exhaust system. u /. / T 1 00ESTION 0.09 (2.00) Technical Specification 3.4.5 c.tates 'two power operated rettef volves (PORV's) and their associated block valves shall be operable'. For the following situations state what actions are required to be taken within an hour if operations at power are to continue? < a. One or more PORV's inoperable b. One or more block valves inoperable (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
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. 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14
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GUESTION 8.10 (3.00) The plant is operating at 55% power and the latest leak rate data shows: 6.2 spm - Total leakage 3.1 spo - Leakage due to a leaking charging pump relief valve 1.2 spa - Leakage into the RC Drain Tank 1.5 spa - Leakage through a RHR discharge valve 0.8 3pm - Total primary to secondary leakage 4.2 spa - Leakage past RCP seals
,
a. What actions, if any, are required in Unit 17 Justify your answer. (1.0) b. What acticns, if any, are required in Unit 2? Justify your answer. (2.0) DUESTION 8.11 (3 00) Classify each of the following occurances using the Event Classification sheets provided. Identify what factor was used to determine the final classification.
~ a. A smal'l plane crashes on the river bank across from the plant.
The fuel tanks of the plane have ruptured and the radioactive medical isotopes the plane was carrying cannot be located.
b. While working with the polar crane in the containment during Mode 5, the brake on the hook fails, dropping the hook onto a piping run. The sample line from the pressurizer liquid space ruptures and area monitors R410 and R41C increase readings to 1.7E3 cpm.
c. An accident in the containment results in an injury to a worker.
The worker is in contaminated anti-C clothing and has broken his leg. Due to the accident the worker receives 2 Rem of radiation Previously he had accumulated 500 mrom for the quarter and year.
d Following operations at 100% power for six months a loss of coolant accident occurs which initiates containment spray. During the transient five relief valves on the steam generator lift and two fail to reseat.
(umans END OF CATEGORY 00 sassa)
(mmmmumassamma END OF EXAMINATION muassmasassmans)
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19 ______________________________________________________ ANSWERS -- SALEM 182 -05/11/19-DVDLEY ANSWER 6.01 (1.50) c. A downstream valve automatically closes if hi-flow is sensed. E&r&1 l A.,n up
; e . .. . s t . ,.. r m e a.
r~s i che.7k valve
, a ra , u . . prevents
,,,w backflow [OTSE
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5,y L :. e , . s ]
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b. VCT pressure is kept greater than 15 psig to ensure 82 seal is not robbed of a trickle cooling flow. CO.53 l REFERENCE I SNGS Student Notes Chap. 4, RCP, p 20, 25 l ' ANSWER 6.02 (2.50) l l As power increases, steam flow detector delta-P increases CO.43 and ! steam pressure decreases CO.43.
l Since steam flow uses a squre root extractor corrected for steam pressure
- and steam pressure is failed to 2 4,% , indicated flow will be higher thah actual flow. [0.53 '
Flow signal will open the FWRV. [0.43 As level increases, the level signal will close FWRV. 00.43 Eventually, the level error will cancelf the flow error and steam flow will equal feed flow at a higher level [0.43 REFERENCE
- SNGS Student Notes, Chap 27, SG Water Level Control, p 21, 22, 26 i
ANSWER 6.03 (2.50) c. 1) DND 2) excessive fuel power and temperatures 3) OND 4) protects RCS integrity from overpressure [0.4 each] b. Since RCS pressure is an input to OT0 elta T and lower pressure lowers the setpoint, the low pressurizer pressure reactor trip places a lower limit on OTDelta T range at a given RCS temperature. [0.93 I REFERENCE SNGS Student Notebook Chap. 20, RPG, p 12 l l
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*6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 . ----------------------------___-__-_-_________________
ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 6.04 (2.00) c. Control room ventilation isolation CO.53 y b. Euel 5:rdl ins wee-e+erm-E0rH4- (4 ' M " " FHB exhaust ventilation shifts No. 22 filter unit *CO.43 c. None CO.53 i d. E2: 3:ncy ::r ,i ;; light re ;3 y Prevents crane hoist up operations CO.40 REFERENCE System Operations Procedures, IV-11.3.1, Table 1 i SNGS Student Notebook Chap. 17, Radiation Monitoring System, p 26-20 ANSWER 6.05 (2.00)
*
x -ctCSED r 0 - OPW .
'1 LINEUP CIRCUIT BREAKER NUMBERS
" 9-10 2-10 500KV . 500KV 1-8
.
1-5 '5 .6 2-6 '2-8 1-9
-
(2) (2) (3) (3), 2T60 1T60 gi) (1) .
-
_ a {. o o :x x x- X 'X ~X X- X ,
' .. ! '
i, ! .o o x. x x o 0 > X X i o x b o o 0 0*2dgy;9)<1y 7 '
.
c-lo l REFERENCE System Operating Procedure, IV-1.3.1 i ANSWER 6.06 (3.00)
- o. SG feed pumps CO.41
- Turbine generator CO.31(flirMeren swr)
l (r y (Emergency seal oil pumpf CO.31 l b. RHR system ' CCW system
Primary makeup system l AFST SG makeup * i Charginn and letdown C O r41sa..+3d . Wr M T- to . g "^y 3 O C'1 rd. ] ,
,
REFERENCE j E0P - LOPA-1, p 2, 14
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. 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21
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ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 6.07 (2.00) Leakage would be the sum of high .and-ir+tenaediatec head pumps discharge flou rate. CO.63 ,,._,,. High head flow' 2X ~400 spm = 800 spa CO.63 Intenmediate--head ' V ^~ e sp: - 850
- h s,p : r0. R se s t r u~ a. 4 u r m*' r ~4 g:m CC 61 LEAK RATE = 1450mgpm CO.23 375 REFERENCE SNGS Student Notebook Chap , ECCS, Table 1, Fig.
ANSWER 6.00 (3.00)
a. Input from PR channel N41 is defeated to the auctioneering units [ 0 . 6'] of the Rod Control System C O .*}] , Axial Flux Distribution Monitoring System CO.33,.and-SGWLG-System-Car 3 M t b. 1) Below 50% power, heat generation is sufficiently low to assure safety 00.753 2) Since the setpoint is 1.02 times the average detector output, at low powers even slight dif ferences would cause the alarm CO.753 REFERENCE SNGS Student Notebook Chap. 20, Encore NI, p 24, 25 ANSWER 6.09 (3 50) a. Low PZR pressure CO.33 1765 psis CO.13 2/3 CO.13 High Containment pressure CO.33 4 psis CO.13 2/3 CO.13 High Steamline Dif. pressure CO.33 100 psid CO.13 1 to 2 CO.13 High Steam flow CO.13 40-110% powerCO.13 with low-low Tava or low Steam line MANJ AL CC.J] press. [0.33 o
'
b. UV on 4 HV buses CO.33 70% for 3.5 sec. CO.23 2/3 CO.11 90% for 10.5 sec. [0.23 REFERENCE SNGS Student Notebook Chap 10, ECCS, p 0, 9 SNGS Student Notebook Chap 13, SEC System, p 3, 35 L
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. 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22 ______________________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 6.10 (3.00) a. - AUTOMATIC E0.253
- MANUAL E0.253 b. E0;53 456%treps-oe r\\ si cm ca -54%-wi4hp>eful--of-bank-P ams 0 EL >l c. The pushbutton will remero all rods (only the affected group rods require reset). [0.53 d. 1) NONE 2) IN 3) NONE'EO.5 each3 tN REFERENCE SGNS Student Notebook, Chap 22, RCS, p 20-21, 26, 27, 31, RS-15
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. 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
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____________________ ANSWERS -- SALEM 1A2 -85/11/19-DUDLEY ANSWER 7.01 (1.50) Restore Tave within limits within 15 minutes by using rods, boron, or steam demand. E1.03-If unable to restore temperature be in hot standby within the next 15 minutes. [0.53 REFERENCE IOP-3, p2 TS, p 3/4 1-6 ANSWER 7.02 ( .60) D. Until 16 minutes have passed REFERENCE E0P - LOPA-2, p 13 DELEfC AMGME 7.03 '2.00) Un e v e n-R GS-t e m p e r- a tswe s-a mo n g-t h e-l o o p u C 1-. 03_ Candi-t ions--in_-the-SG'-s-leading-to-inadver. tent SI-cr e a ted by SG-differencial-pressure-.-01.03 REFERENCE IOP-6, p 12 ANSWER 7.04 (2.00) Nuclear power > 5% after reactor trip Core exit Te > 1200 F re Core exit Te >,g 700 F and RVLIS full range < ET and no RCPs running NR SG 1evel < 14% and feed flow < 15-spr u a7 "-/A Te decrease > 100 F in 60 minutes and T @ r6-F-'1/ e' h-f r o 4. ~w CTMT press > F7 psig
[any 4 9 0.5 each]
REFERENCE Wastinghouse Owner Group Emergency Response Guidelines, Executive Summaryl Generic Issue Foldout PaSe Items, Fig 1
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
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____________________ ANSWERS -- SALEM 1R2 -85/11/19-DVDLEY ANSWER 7.05 (2.40) a. Flow is reduced.
-Head loss due to friction in the pipes decreases.[0.63 TherefoN discharge head of pump decreases. [0.63 6Ufu rad As: .'*rsw- ! ~ . ~' f~ *.~ * t . - n : o,; o- z., w a .;r . . b. Since-less--enersy- i s being-removed-f r om-the-cor e-greater 4 subcooLing-must be-maintained-to-ensure-no boiling' ir. thea '
.copnt ,23 s I r c wcinn rt N n>sruer t~'" A'~ -
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*M i ~ ~ < il C * ' 'd'
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a;u caw ic ; ,, . . , ; ; c . . .~ r. t. r ' - 'JAd *w 'ic: r u..- REFERENCE ;s re=_ .f -Av,r . e . ri 4 7 E0P - FRHS-1, p 16 Gsneral Physics, Thermodynamics and Fluid Flow Fundamentals, p 302 ANSWER 7.06 (2.00) a. Within 5 minutes. [0.53 or a high motor bearing temperature of 175 F. [0.53 Pump radial bearing temperature reaches 210 F. CO.53 b. 1 minute C0.53 REFERENCE System Operating Instructions, II-1.3.1, p 3, 4 ANSWER 7.'07 (2.00) a. A small_RCS break which can be compensated for by the available charging pumps. [0.P3 b. If pressurizer pressure and/or level continues to decrease CO.73 after starting additional charging pumps [0.23, reducing letdown flow E0.23 and energizing pressurizer heaters. CO.13 REFERENCE I-4.17, p 10
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M_- - 7-
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25
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ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 7.08 (3.00) risco gci '-yp t 1. No [0.353 because he limit." [0.43 2. Yes E0.353 exposure would wouldbeexceed 5(N-19)ph less than Eh mren/qt [0.43 3. Yes E0.353 because she would not exceed limit of 500 aren/ge CO.43 4. No.[0.353 because he wsvis exceed 200+ mrem /qt-[0.43 I t rl5 900 f d4 fvM REFERENCE Salem AD-24, p 13 ANSWER 7.09 (3.00) a. Reactor Trip Instrumentation [0.53 PORV Operability E0.53 A.C. sElectrical u> n: a acera : S
: .ys t em E0.53 iSf .
b. RCS Loop-Startup and Operation E0.53 croir n.c. >cJam s . r c.m A t c. PORV Operability CO.53 A.C. Electrical System E0.53
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REFERENCE AOP-ELEC-VIB-C, p APPX 1-1 AOP-ELEC-125V-C, p APPX 1-1 AOP-ELEC-4KV-C, p APPX 1-1 ANSWER 7.10 (3.50) a. Shutdown reactor. [0.43 No requirement for a reactor trip E0.33 b. Reduce power to maintain vacuum. [0.43 If turbine vibration reaches 7 mils trip turbine. [0.33 c. Shutdown reactor. [0.43 If PZR pressure decreases below 1865 E0.33 d. Continue to operate. [0.43 If pressure reaches 65 psis trip reactor. [0.33 o. Shutdown reactor. [0.73 St a ten't s. * :: scan ti,,,v :e tw:s u rirr REFERENCE I-4.26 Secondary plant leak, p1 I-4.13 Loss of circulating water / loss of condenser vacuum, p2 I-4.24 Pressurizer pressure control malfunction, p 10, 11 I-4.18 Loss of control air, p2 I-4.23 Loss of containment integrity, p2 . __
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. 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
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____________________ ANSWERS -- SALEM 1&2 -85/11/19-DUDLEY ANSWER 7.11 (3.00) Trip reactor with second trip handle. [0.53 Open both reactor trip breakers. CO.53 Open 2E6D and 2G60. [0 53 Dispatch operator to open Reactor Trip breakers and trip Rod MG set. [0.73 Manually insert control rods. [0.53-Go to E0P (FRSM-1) response to nuclear power seneration. [0.33
. REFERENCE E0P - TRIP-1, p3 J
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8. ADMINISTRATIVE' PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 27 _____________________________________________________.____ ANSWERS - SALEM 1&2 -85/11/19-DUDLEY ANSWER 8.01 (1.50) l Ini-ti a te-a-R ev i+ i on-R eq u e s t- F o r m e- AD-1- A-1-a nd-f o r w a r d-t o-t h e-S e n i o r e. i Op e ra t i on s-T e chn i c i a l- S uper v i s o r-4SOTsh-EO . 6 M Job-ca nn o t-b e-h a nd l e d-a s-O n-t h e-S p o t-C h anse se-CO +916 W.y dLLad W4T REFERENCE AD-1, p 1, 8 oninics facon cantat w.xn r%ric e-
,g ,.,.,.g ,,,.,,, . , ., , g.7 , y , ,, , ,, , , f g. s. , ., ,7 ANSWER 8.02 (2.50)
Controls Engineer [0.53 request assistance for review of sequence of events print out CO.73 Operations Manaser E0.53 notification of the event CO.43 and subsequent findings of the review report. [0.43 STA i 1 . ' c . . * n . d: ' e s , rMc
.
REFERENCE AD-16r Post Reactor Trip / S'afety Injection Review r p 1-2 ANSWER 8.03 (2.00) No. [0.73 The off soins Unit 1 NSS mus't stay to meb the T.S. manning requirement. CO.83 Action statements cannot be entered for operatin3 convience. [0.53 , ' REFERENCE Salem T.S., p 6-4, 6-5
. ANSWER 8.04 (2.00)
r Unit must be placed in hot standby in one hour. CO.73 NRC is informed within one hour. [0 .' AJ y Safety L'imit Violation Report shall be prepared. [0.33 ,- Report is submitted to NRSC and Upper Management (within 14 days) [0. &3
, REFERENCE Salem 1 T.S. p 6-12a w eia
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28 __________________________________________________________ ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY-ANSWER 8.05 (2.00) HTC is within normal analyzed ranse.
Protective instrumentation is within its normal operatin3 ranse.
P-12 is'above its setpoint The prssuri=er is capable of beins opeable with a steam bubble The reactor vessel is above minimum RT NDT Eany 4 0 0.5 each] REFERENCE Salem T.S., p 83/4 1-2 ANSWER 8.06 (2.00) a. (CAF) Operations department. E0.53 b. Extension of up to 25% of interval E0.753 Three consecutive intervals shall not exceed 3.25 times the normal interval. E0.753 - REFERENCE solem TS, p 3/4 0.2 ANSWER 8.07 (2.50) a. None b.
Jumper / lifted lead, a g.,q g;,. ,y y c. Red blocking tas a,t:_ gg t ,.fne d. Wi$heeblocking tas arrn
._11cv Dereissive tas i r T r., Am r 4, w. s c e. Whi-tc ca:;tien +994 ixvc E0.5 each]
REFERENCE Salem AD-13, p 1, 2 Salen AD-15, p 1-3
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS .PAGE 29 __________________________________________________________ ANSWERS -- SALEM 182 -85/11/19-DUDLEY ANSWER' 8.08 (2.50) me't /41.
.. hs Ng B wne ed CfM' a. Ne6[0.23 :::,11 r e ,e l t-in-te-i --on f2 / art-OFDT , c. EO . 33 b. No CO.23 cannot enter another Mode with reliance on action statement E0.33 c. No- 00 23 de not-intent 4ena44y-enteset-iorstateamet-Joe _.
maintenace; E O : 3 3 : Pa lC il e * " ' ' " N" " W '~ " ~f "'" ' " ' ' ' , l,5),[,-}"'
,
d. Yes E0.23 if only one seal is replaced at a time and other' door (> 31 is kept shut. CO.33 c. Yes E0.23 if one train at a time is taken out of service. [0.33 REFERENCE Salem TS, p 3/4 3-5 3/4 4-4, 3/4 0-2 3/4 8-1 3/4 9-4 3/4 7-22 ANSWER 8.09 (2.00) a. Either restore the PORV(s) to operable status E 0'. 5 3 or close the associated block valve (s) and remove power from the block valve E0.53.
b. Either restore the block valves to operable status EO.51 or close the block valves and remove power from the block valves CO.53.
REFERENCE Unit 2: .TS, p 3/4 4-8 ANSWER 8.10 (3.00) a. No action required since no leakage limits exceeded. E1.01 b. Close 2 manual valves or deactivate auto valves to isolate HP from LP part o'f system E0.73 or be in cold shutdown E0.51 (hot standby in six hours or cold shutdown in 12 hours.)
1 spa leakage from.RCS has been exceeded. E0.83 i REFERENCE Salem 1 T.S., p 3/4 4-15 Salem 2 T.S., p 3/4 4-17 l
- . - . . 4
-
.
, - > ,
., 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 30 __________________________________________________________ ANSWERS -- SALEM.182 -85/11/19-DUDLEY
'
ANSWER 8.11 (3.00) a. No report E0.53 does not effect plant ops, was not initiated by plant, and does not effect local t. ravel patterns. E0.253 b. Unusual event E0.53 due to leak rate C O . 253 c.: rt TT.14 c . : sis o d e ic r'u'# "# ' c. Unusual event E0.53 due to transport of contaminated individual. [0.253 d. Site Emergency E0.53 LOCA greater than makeup capacity. [0.253 REFERENCE' Event Classification / Notification /Reportins soider Sec 1, p 1 Sec 7, p 1 Sec 19, p 1
. ' __
- - - -_ - - - . - , .- . - . , - . , . . . .
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 15
---- --------------------------------------
_------------- ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 5.01 (2.00) a. Valve operation, opening or closing Pump starting or stopping . Oscillation of auto control valves [awy A O. O'/rAn]. b. Slowly opening.of valves between voided and full systems Proper venting of components Adequate level on tanks in systems where the tanks provide supply or surge function Proper use of steam traps and vents Pro ALLber %' sequencing i TIM s rcaof valvesri, U r % L 5 W !}L pl d f[[ Wpressurized systems
*
{uW J (f) Pleac,.]
REFERENCE Nuclear Energy Training, Thermodynamics, NET 4-2, pg. 2.1-4,5 ANSWER 5.02 (2.50) T ave decreases since more ener3y is being removed. [0.73 Rx Power increases due to the positive reactivity added through HTC TF i/ ' and_4op g e w co m Dci M r* WhtO s .. u.' .s no .m a r i v r:c.tci 1 -: i r [C Q Power stabilizes at a higher value. [0.53 Tave stabilizes at a lower value. E0.53 REFERENCE SNGS Student Notebook Chap 47, Accident Analysis, p 24 ANSWER 5.03 (2.00)
'
a. Due to decrease in self shielding of the fuel pellet [0.53 as the range of neutron energies in the absorption band increases CO.53.
b. Less negative C0.53 due to increased overl'apping of. resonance absorption bands E0.53 OR decreased self-shielding CO.53 REFERENCE SNGS Student Notebook Chap 46, Trainsient Analysis, p 10, TA-7 ,
.
l l
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. . .
% 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 16
---- --------------------------------------
g
-__-----------
ANSWERS -- SALEM 1&2 -85/11/19-DUDLEY ANSWER 5.04 (3.00)- a. Th stable or decreasing E0.33 Loss of matural circulation flow. [0.33 b. 10 F subcooling E0.33 Voiding in core or hot leg which would interrupt flow. [0.33 c. SG pressure tracking T pressure. [0.33 SG not removing heat. Y+ve. O.33 saturation d. SG level in Narrow RanSe EO.33 SG no longer available as heat sink. [0.33 e. PZR level 50% [0.33 Voiding of hot leg which would interrupt flow. CO.33 gi,-FC-T-i n e r-e a s e s-GO .33-b e c a u s e-t h e-o-l a d-h e a t--t r-arsf e n-ca p a c i4y-iw r e d uc e d .< 4incr ease-temper a tur e-due - to-lower-conduct-ivi ty-of-ce+ed ) EO. 43 <- REFERENCE General Physics; Heat Transfer Thermodynamics and Fluid Flow Fundamentals, p. 356-357.
ANSWER 5.05 (2.50)
,
it 50 fm vnt a.4 Shut down bank withdrawal occurs when the reactor is further te vtv .s N away from criticality then when boron dilution occurs. [0.53 onc na> Prevents- unexpected-c r-i ticali ty.-CO.53 Doubl-ing-the-count e
. rate-halves the-the-amount by--which-r-eactor is-shutdown +E0.534 j g,gg g cc,;g,.. ,
g,4;rica is b. ECP different from last criticality by 3reater than the tolerancesu,w er .,
[ca,3.y on the work sheet. E1.03 ( > (TC nm c..e s ,WC4 REFERENCE * #" #"
Rx Eng Man, Part 1, p 1 M A1 N>T # # 'A' u In ou t r d Part 4.1, p 1 Part 4.2, p 1 CAF
.
+ . . . . - .
% 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17
---- -------------------------------------- --------------
ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 5.06 (3.00) c...cr ; a. Powen decreases [0.53 The pump is pumping the same volumetric flow rate but the density of the water decreases CO.53 which causes a lower mass flow and less pump powee. [0.53.
caru;ENT v e d" b. T^c-er decreases. [0.53 The pump is pumpins against a lower discharge head due to the reduction of head' losses in the area of the void. [1.03 REFERENCE Heat Transfer-and Thermodynamics, Part A, Chap 2, See III, p 332 ANSWER 5.07 (3.00) a. EOL - MTC is more negative and thus imparts greater positive rea'ctivity.
.from the drop in coolant temperature. pddI3 Doppler onl coefficient (DOPC) is less negative and thus imparts.Fe@y Powerd reactivity from the power increase. CO.53
,
l
, _ .
b . &Ot' ECL
- 4TC-i s--l es s-nesatri ve-and-thus-impants-less-negative reacLi-vitye.
Zvon the-coolanL--heat-ve. E%51-L Cl.C E A' TC ()!.'Oh' [c il DOPC is more negative and thus imparts more positive reactivity as power drops which tends to hold power up. [0.53 c . fMRP - Beta-eff is less, making SUR greater. [0.53 00' DOPC is less negative and thus imparts less negative reactivity after the fuel temperature begins to increase. [0,53 REFERENCE SNGS Stu' dent Notebook Chap 47, Accident Analysis, p 28, 47, 56
% * > : f
. . . .
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18
---- -------------------------------------- -----__------- ' ANSWERS -- SALEM 1&2 -85/11/19-DVDLEY ANSWER 5.08 (3.00)
o ec ca ca u o toyegs a. An-ir. creased, pressure setting rc:;es'Tave. Assuming a negative MTC, this adds FG' nc?gt2vv reactivity so the ACP is HIC"ERv
- ~~
L. C Jc '\
(1.0)
b. A 4 hour delay adds positive reactivity since Xenon is decaying beyond its ECP value. The ACP is LOWER. (1.0) c. Rod worth increases with temperature, so the ilFP worth is Sreater than the HZP worth. This implies less positive reactivity has been added so the ACP is HIGHER. (1.0) REFERENCE SNGS Student Notebook Chap 46, p 6, 24, 38 ANSWER 5.09 (4.00) a. Power Defect (Fig. 2) +1125 pcm E0.43 IRW -(Fi3 4, 15) -3200 pcm (-3480 + 280) CO.63 SD rod worth (Fig. 16) -4470 pcm E0.43
------------
-6545 pcm E0.13 b. Differs by the worth of the most reactive rod being withdrawn. CO.63 From Fig. 17, 1290 pcm. [0.43 ed 1 0 7 "W - 6. iX '% 't. '/ % ~ ~/<
c. Xenon worth (Fig. 8) = -4600 + 3000
= -1600 pcm CO.63
> 1%en Drue.r ppm change =
Xenon worthf{ypjfferencial Boron Worth (Fig. 12)
= -1600 pcm */ -8.15 pcm/ ppm
= F96' epa 4 E0.63 5 3 rpn
~
Gallons of demin. (Fig. 101) F p Sal CO.33 REFERENCE Rx ENG Man, Part 3, Sec 3.7 (b), p 5, 6
.
~_2___n-- - - - - - - --J ,.a - . _ , _ . _ - - , - - --
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EQUIL]BRIUM XENON RERCTIVITY vs REACTOR POWER
.
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I MEUP CIRCUIT BREAKER NUMBERS E 500KV 1~5 5-6 2-6 -2-8 1-8 1-9 9-10 2-10 500KV.
(2) (2) (3) (3) 2T60 1T60
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(1) (1)
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P% EVENT CLASSIFICATION SECTION 1
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1. Primary Leakage Initiating Event / Notification / Condition Emergency Action Level Pecorting A. Primary leak 1. Exceeding action statements ATT 1 a) no PRESSURE BOUNDARY of T/S LCO 3.4.6.2 (Unit 1) UE leakage , or 3.4.7.2 (Unit 2) b) I gpm UNIDENTIFIED - leakage See Tech Spec c) 10 gpm IDENTIFIED Action Requirements leakage , d) 40 gpm CONTROLLED . I leakage , e) RCS PRESS ISOL VALVES (See T/S) B. Prima ry leak > 50 1. One chg pmp cannot maintatn ATT 2 gpm level A C. Known LOCA greater than 1. Low / decreasing P2R level with ATT 3 total' makeup., capacity .. maximum charging flew SAE
. *** Refer to Section 5 prior to classification *** ' ( O. P2R safety /PORV fail- 1. P2R press >2200 psig & POPS ATT 1 ure to reseat not a rmed , or P2R press (371 ~ 'U E , psig & POPS armed; AND 2. PORV/ safety valve tailpipe F1 . -
temp, or PRT temp, press, og
-
level increasing . E. Pipe cracks in stagnant 1. Cracks in weld areas of saf e ty ATT 6 , borated water systems related piping (as reported by (50.723-lHr)
. Engineering or ISI/MIET)
. IE Inf ormation Notice - one hour
*
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EVENT CLASSIFICATION SECTION 2
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2. Primary to Secondary Leakage Initiating Event / Notification / Cond ition emersency Action Level Recorting A. Primary to secondary 1. Exceeding action state- ATT I leak ments of T/S LCO 3.4.6.2 UE a) 1 gpm TOTAL thru (Unit 1) or'3.4.7.2 (Unit all S/Gs . 2): b) 500 gpd thru any See Tech Spec Action - 1 S/G Statements B. -S/G tube rupture 1. Rx trip /SI on LO P2R ATT 2 (several hundred gpm) PRESS: AND A
*
2. R15 or any R19 alarm: AND 3. P2R press does not recover
> S/G press C. Steam break with 1. a) STMLINE ISQL SI cn HI ATT 2 primary to secondary STM Plcw w '. t h LO Tave or A leakage LO STM PRE :S or b) SI on STMLINE HI DIFF PRESS; AND 2. R15 or any RI) a la rm
(' O. Single S/G tuce f ailure with ' loss 1. RX trip'/SI on LO P!R PRESS: 2. R15 or~any R19 alarm: AND ATT 2 A of offsite power'~ 3. P!R press recovers > S/G press; AND 4. Offsite pcwer loss . E. Ste5m break with 1. a) STMLINE IS3L SI on HI ATT 3
> 5 0 gpm p rima ry to .
STM FLOW with LO Tave SAE secondary leak and or LO STM 3RESS; or , indicated fuel b) SI on STML(NE HI DIFF damage. PRESS AND 2. R15 or any Rl) a la rm: AND 3. 1R31C (2R31) 1:a HI ALARM (offseale): AND 4. a) RCS analysis shows failed fuel increase >11/30 min, total >51, or I-131 equiv
>300 uCI/cc OR b) Survey of main steam lines . - indicates high activity (SSS/EDO juugement) being released to atmosphere OR c) One or more of the following:
R46A > 0.14 uCI/cc * R468 7 0.14 tCi/cc
*
R46C I 0.14 uCi/cc i I R47D 1 0.14 uCi/cc .
'
SGS * 1 of 2 Rev. 0
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*
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. , .,_, _ , . - - _ _ __- . . . , . ' . ..
EVENT CLASSIFICATION SECTION 2 2. Primary to Secondary Leakage (con'd)
*** Refer to Section 5 prior to classification * **
F. S/G tube rupture 1. Rx Trip /SI on LO P2R PRESS; ATT 3 (several hundred gpm) AND SAE with loss of of f site 2. R15 or any R19 alarm; AND power . 3. PZR press does not recover > S/G press; AND 4. Of f site power loss
* ** Re f er to Section 5 prior to classification * **
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Rev. 0 SGS
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EVENT CLASSIFICATION SECTION 3
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3. Secondary Leakage Initiating Event / Notification / Condition Emergency Action Level Reportine A. Secondary line break 1. HI STM FLOW or abnormal ATT 1 causing a rapid increases in feed flow *to UE uncontrolled seco.ndary 1 or 2 S/Gs: AND depressurization 2. LO S/G p res s , LO TAV E , or decreasing RCS press B. S/G safety /PORV failure 1. Visual / audible indication ATT 1 to reseat. at vent stacks after S/G UE
*
press < setpoint; OR 2. Excessive f eed or steam f lo'w 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 D. Steam break witt. > 50 gpm 1. a) STMLINE ISOL/SI on HI STM ATT 3 - primary to seconda ry FLOW with LO TAVE o r LO SAE leakage and ind cated STM PRESS; AND fuel damage b) SI on STMLINE HI Otif PRESS AND 2. R15 or any R19 alarm 3. 1R31C ( 2R31) in HI ALARt4 (offscale); AND 4. a) RCS analysis shows fatled fuel increase >11/30 min, total >53, or I-131 equiv >300 uCI/cc OR b) Survey of main steam lines in-dicates h.igh activity (SSS/EDO judgement) bein released to at-mosphere)
. OR c) One or more of the f ollow tng:
R46A > 0.14 uCi/cc R46B T 0.14 uCi/cc R46C I 0.14 uC1/cc R460 _I 0.14 uCi/cc
* ** Re f er to Sectic'n S prior to classification ***
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SGS 1 of 1 Rev. O
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EVENT CLASSIFICATION SECTION 4
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4. Fuel Damage / Degraded Core , Notification / Initiating Event Emergency Action Level Reporting Condition A. Fuel damage 1. IR31C (2R31) ala rm: AND ATT 1 a) RCS sample shows failed UE
'
fuel increase of
> 0.1t/30 min: OR b) RCS sample shows activity >
limits of T/S LCO 3.4.8 (U/1) or 3. 4.9 (U/2) 8. Severe loss of fuel 1. RCS sample shows I-131 equiv ATT 2
>300 uCi/cc; OR A cladding 2. 1R31C (2R31) HI ALARM (off-scale) and RCS sample shows failed fuel increase of > 16/30 min or total >5T C. Fuel damage accident 1. RS , R9, or R29 and R418 or .
ATT 2 R41C alarm; OR A with release to FHB or containment 2. 2/4 of R2, R7, R10A, or R10B
~
alarm and R21 > 1 R/hr: AND 3. Fuel handling problem with k possible -fuel damage
~
~
***
*** Refer to Section 5 prior to c1assification D. RCP seizure leading 1. RCP motor current spike to ATT 2 locked rotor value then :ero, A to ruel failure ,
or sudden flow decrease in . affected loop, o r RCP v id a la rm and loose parts monitor alarm: AND .! -
~ 2. Letdown monitor 1R31C (2R31)
alarm; AND
- 3. Analysis shows RCS activity >
limits of T/S LCO 3.4.8 (U/1)
. .,
or 3.4.9 (U/2) ' 1. RS, R9, or R2 9 a la rm: OR A'TT 3 E. Major damage to spent SAE fuel in containment 2. 2/4 of R2, R7, R10A, or R108 or Fuel handling . alarm e and R21 > 1 R/hr; AND - building 3. Confirmed f uel damage or loss of water level to'below fuel - level
.
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1 of 2 Fev. 0 SGS
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EVENT CLASSIFICATION SECTION 4
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4. Fuel Damage / Degraded Core (con'd)
.
Initiating Event / Notification / Cond i t ion Emergency ' Action Level Recorting F. Degraded core with 1. >5 core exit T/C >l200 F: OR ATT 3 possible loss of ' 2. Eoss of Main and Aux f eed, no SAE Coolable geometry ' S/G WR level, and no S/G press; - , OR 3. 1R31A (2R31) in HI ALARM (off-scale) and RCS analysis shows failed fuel increase >11/30 min, total > $t or I-313 eq'uiv > 300 uCi/cc: OR 4. > 2 WR hot leg RTDs > 700 F, and any R41 a la rm, containment sump level > 81'3", or 2/4 containment press >4.0 psig
* * * Refer to Section 5 prior to classification * * * . .
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sTsNT CLASSIFICATION asCTIon 5 1 Fiselon F1roduct Boundary Failurse the following conditions indicate f ailure of a fission product s oundary.
Any two boundary f ailures, with the possibility of a third, represents a causaAL anamGENCY condition and Attactement 5 must be implemented .
*
immediately. Monitor for these conditions af ter any single boundary failure. , A. am"**n** 0022
*
MOTE
.
When SSS/EDO judgement in-dicates procoole failed fuel, completion of RCS analysis shou.d not restrict / delay of emergency classification.
- 1. Letdewn monitor and RCS anaylsis snows failed fuel: 1R31A (2R31) in HI ALARM (offscale) al increase >11/30 min, or Di total >$t, or c I-131 equiv > 300 uCi/cc CR
.
OR 2. RCS Analysis shows failed fuel: a) increase > 14 / 30 min, or n) total > 54, or c) 1-131 equiv > 300 uC1/cc on 3. Engineering analysis indicates that a degraded core condition esists (as presented to the Emergency Coordinator by Engineering Suppokt Staff), ,
'
OR 4. >.5 core entt T/Cs RCS delta T rapidly increasing
> 1200 F, or >2 and (Thot increasing), or decreasing WR That > 7007 to zero (That * Teold)
8. LOSS OF FRIMARY COOLAarf, , . . 1. FIR level decreas'ing with maximum charging flow on 2. 2/4 cant press >4.0 psig and accumulator discharge (Modes 1, 2, 3) Ca 3. Inadequate RCS subcooling (P250 or manual 08 -,_,q 2/ 3[--- r ATT ealculation: Modes 1, 2, 3) 5 on CE 4. R 44 > 20 R/hr and 2/4 of R2, R7, RICA, or RIOS alarm . On 5. 2/5 CFCU drainage Cent sump level alarme or 2/4 cont and ~~~~
>81' 3" (no indication -
press >4.0 psig of in-cont sem bream)
*
.
C. CONTAIIBIENT FAILURE 1. Containment M 2conhntration>44 - os 2. 2/4 cent press >47 peig and increasing on containment press
- 3. No cent spray capa0111ty with (5 CFCUs availaDie, and >23.5 psig and gg i
or 1 cont spray train with increasing l
(3 CFCUs avellanle on '
4 Steam break, which canact be isolated, outside containment
'
with indications of prLaary-to-secondary leakage on . 5. Cont penetration isolation valve (s)'or cant hatch failure (any breach in containment as in-dicated by the significant rise in airborne activity gg in the area of concerns, __ __ _ . _ . ___ _ _ __. _ _
r
. . , . .
EVENT CLASSIFICATION SECTION 6 6 Radiological Releases . Note: Action levels listed are for valid RMS channel indications.
The validity of the indication should be confirmed by sample analysis or other means as necessary.
NotifLcatLon/ Initiating Event / Emergency Action Level Reporting <
,
Condition , A. Accidental, unplanned, 1. R41C > 1.5E3 cpm for > ATT 29 or uncontrolled 60 min; (50.72b - 4Hr) gaseous release, that Ojt . exceeds 2 times the. 2. R418 > 5.7E3 cpm increase
-
applicable concentrations in 60 min of the limits specified in Appendix B, Table II of Part 20, to un-restricted areas (averaged over 60 min) 8. Accidental, unplanned, 1. Any unplanned or uncontrolled ATT 29 or uncontrolled . liquid release outside the- (50.72b - 4Hr) liquid release controlled access area.
. C. Liquid release that 1. R18 alarm and no isolation ATT 7 UE exceeds T/S limits AND
, Confirmed analysis of Liquta
.
f or > 15 min waste ef fluent indicating discharge exceeding T/S l tm t t:5.
0 R, 2. Any R19 alarm and blowdown to 12(22) S/G blowdown tank -
.
D. Gaseous release that 1. R41C > 1.lES for > 60 min: ATT 7 UE exceeds T/S limits Ojt 2. R41B > 2.lE4 cpm increas in
.for-> 60 min 60 min:
1. R41C > 1.lE6' cpm for > 15 min: ATT 8 E. Gaseous release that A exceeds 10 times T/S OR limits for > 15 min 2. R418 > 5.3E4 cpm increase in 15 min: OR ' 3. R45B > 7.lE-3 uCi/,cc for > 15 min SGS 1 of 3 Rev. 0
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EVENT CLASSIFICATION SECTION 6 (con'd) 1.a) Field team measures'whole ATT 9 F. Dose Rate at Minimum SAE Exclusion Area (MEA) body dose rates greater equivalent to 500 mR/hr than' 50 mrem /hr for 30 min WB or 2500 mR/hr thyroid or greater than 500 mrem /hr
*
for > 2 min (MEA is for 3 min at the MEA defined as any monitoring OR
' location greater than b) Tield team measures, at the 0.79 miles away from the. - MEA, Thyroid dose rates af fected unit- (equiv I-131 concentrations)
greater than: 250 meem/hr (1.0E-7 uCi/cc) for 30 min or 2500 mrem /hr 1.0E-6 uCi/cc for 3 min: OR 2. Calculation of 500 mrem /hr dose rate WB or 2500 meem/hr enyrotd dose rate for > 2 min based on EP IV-lil, EP IV-113 or EP IV-114 and actual meteistology; OR ,
~.
3. When methods 1 snd ? above a re_ not
.
available based on valid gaseous .
- effluent monitors: .
R41C > 2. 3E6 cp i, OR R41B > 2.0E6 cpu, OR
- R458 > 0.14 uCi/cc, OR ~~ > 7.5 mR/ir, OR R43 One or more of the f ollow'Ing :
R46A > 0.14 uC i /cc R46B > 0.14 uCi/cc R46C > 0.14 uCi'cc
~
R460 > 0.14 uci'cc
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. SGS 2 of 3 Rev. 0
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EVENT CI.ASSIF7 CATION SECTION 6 (con'd)
,
G. Dose rate at Minimun 1. Fiald survey team measures ATT 5 Exclusion Area (MEA) at the MEA 1 R/h r W.B. or CE equivalent to 1R/hr 5 rad /hr Thyroid (I-131 WB or 5 rad /hr thyroid concentration of 2.0 E-6 (MEA is defined for the uCi/ce); purpose of field moni- OR , toring teams as any 2. Calculation of 1R/hr WB or distance greater than 5 rad /hr thyroid using 0.79 miles away from EP IV-111, EP IV-113 or the affected unit) EP IV-114 and actual meteorology; OR 3. When methods 1 and 2 above are not available based on valid gaseous effluent monitors: R/lC > 4.6 E6 cpm.,OR R41B > 4.0 E6 cpm, OR ' ~ R 5B > 0.28 uCi/ce, OR R.5C > 0.28 uCi/cc, OR
- R43 > 15.0 mR/hr, OR One or more of the following: ,
R46A > 0.28 uCi/cc R468 > 0.28 uCi/cc -- Ri6C > 0.28 uCi/cc R16C > 0.28 uCi/cc H. Loss or thef t of licensed 1. J2dgement of SSS/EDO. ATT 35 material in such quanti- (50.72t-lHr) ties and under such cir-cumstances that it appears to the licensee that a substantial hazard may result to persons in un-restricted areas.
( Re f er to 10CFR 20.402) I. Loss, theft or diversion 1. Any loss, theft or diversion ATT 12 of any special nuclear of any special nuclear material UE
- ~ material known or suspected (eg. fuel elements or incore detectpes see 10CFR 70.52) etther in transit to or from this faci- - lity or at this factJity J. Abnormal degradation 1. E'xcludi'ng normal valve ATT 13 of systems designed to paesing or gasket leakage OTHER contain radioactive Examples ma te ri.a l (not fuel a) through-wall L.eak in WHUT clad, RCS, 6r cont)
K. Radiological sabotage 1.' Judgement of SSS/EDO ATT 25 (attempted or substantiated) UE SGS . 3 of 3 Rev. 0
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EVENT CLASSIFICATION SECTION 7 7. - Radiological expcsure/ contamination Initiating Event / Notification / Cond i t ion Emergency Action Level Reporting A. Transport of a 1. As stated ATT 14 UE contaminated individual from site to offsite medical.
facility B. Increase in measured or l'. Measured or calculated dose ATT e calculated dose rate rates increased > 1000 times A ( mR/h r) by > 1000 times a) On installed or portacle
.(indication of severe monitors: OR degra'dation in control of b) Unit 1 increase of analog radioactive materi.ls) strip chart reading over 20 min: OR c) Unit 2 rad mon computer-trend increase C. Dose rate at Minimum 1. a) Field team measures whole ATT 9 body dose rates greater SAC Exclusion Area (MEA)
equi. valent.to 500 than 50 meem/hr for 30 min mR7hr WB or 2500 ma/hr or greater than 500 meem/hr WB or 2500 mR/hr. thyroid OR for > 2 min (MEA'is b) Yield team measures, at the defined as any monitor- MEA, Thyroid dose rates ing location greater (squiv I-131 concentrations) than 0.79 miles away greater than: from the affected anit 250 meem/hr (1.0E-7 uCi/cc for 1.0E uCi/cc for 3 min: OR 2. Ealculation of 500 meem/hr dose rate WB or 2500 mrem /hr thyrot,d dose rate for > 2 min based on EP IV-111, EP IV-ll3 or EP IV-ll4 .
. and actual meteorology:
OR 3. When methods 1 and 2 above are not
- '
available based on valid gaseous effluent monitors: * R41C > 2.3E6 cpm, OR R41B > 2.0E6 cpm, OR
- R45B > 0.14 uCi/cc, OR R43 > 7. 5 mR/h r, OR One or more of the following:
R4 6A > 0.14 uCi/cc R46B > 0.14 uCi/cc
,R46C > 0.14 uCi/cc R460 > 0.14 uCi/cc SGS . .
1 of'2 Rev. 0
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EVENT CLASSIFICATION SECTION 7 (con'd) D. Dose rate at Minimum 1. Field survey team measures ATT 5 Exclusion AREA (MEA) at the MEA 1 R/hr W.B. or GE equivalent to 1R/hr , 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 field moni- OR
. toring teams an any 2. falculation of 1R/hr WS or distance greater than 5 rad /hr Thyroid using 0.79 miles away from EP IV-lll, EP IV-113 or the af fected -unit) EP IV-Ll4 and actual meteorology; '
.
. OR 3. Wiien methods 1 and 2 above
- are not available based on
. valid gaseous effluent monitors:
< R41C > 4.6 E6 cpm, OR
- '# ',
R41B > 4.0 E6 cpm, OR R45B > 0.28 uCi/ce, OR R45C > 0.28 uCi/ce, OR R43 > 15.0 mR/hr ,
, One or more of the following: R46A > 0.28 uCi/cc e. R46B > 0.28 uCi/cc R46C > 0.28 uCi/cc R46D > 0.28 uCi/cc , E. Any incident involv- 1. Receipt.by an indivdual of: ATT 10 ing t yproduct, source, a) WB exposure > 5 r,em, or (50.725-4He or special nuclear b) skin exposure > 30 rem, or
- material causing any c) extremity' exposure > 75 of the listed results ' rem; OR l '
2. Loss of > one day of oper-ation; OR
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3. Damage to property > $2,000 F. Any event involving 1. Receipt by~an individual of: ATT 11 l byproduct, source or 'a ) WB exposure > 25 etm, OR (50.725-4Hr special nuclear material d) skin exposure > 150 rem, OR causing any of the c) extremity exposure 2 375 rem:
,
listed results CHL 2. Loss of > one working week of operation;
,
OR 3. Damage to property > S200,000 i G. Any personnel 1. Reported by HP: ATT 13 l overexposure a) > 10CFR 20.101 limits Other b) > 10C FK 20.10 3 limits c) > 10CFR 20.104 limits
*
SGS 2 of 2 , Rev. 0
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EVENT CI.ASSIFICATION SECTION 8 l 8. Nonradioactive leak / release Initiati'ng Event / Notification / Emergency Action Level Reporting Condition 1. Fan coil unit or SW piping ATT 6 A. Service water leak in (50.72b-lHR) cont. when cont. leak in containment intergrity is required . B. Unmonitored discharge 1. Observation of non-rad ATT 15 of' nonradioactive liquid waste basin overflowing Other
,
C. On-lani oil spill 1. Observation of on-site ATT 16 oil spill Other D. Oil spill into river 1. Observation of a spill ATT 16 from on-site into Delaware Other River; OR 2. Observation of oil stick
. on Delsware River
.
1. Observation of a release ATT 7 E. Toxic or flammable threatening plant personnel: UE gas; release that -
-'
threatens plant OR
'
personnel 2. Warning from offsite of a
. release that may travel onsite 1. Observation / measurement of ATT 8 F. Toxic or flammable gases exceeding flammability A gas release entering vital areas and/or toxicity limits after ,
entering CR or aux bldg ventilation syst'em 1. Toxic gases entering vital ATT 3 G. Entry of uncontrolled areas affecting safe opera- SAE topic gases ipto vital areas where lack of tion of the plant access to the area constitutes a safety problem.
1. Cracks in weld areas of Al s' 6 H. Pipe cracks in stagnant (50.72b-lHR) borated water systems safety relating piping
- (per Engineering or ISI/MIET)
.
a.
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SGS , 4 .
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EVENT CLASSIFICATION SECTIOrt 9 9. Fire Initiating Event / Notification / Condition - Emergency Action Level Reoortinc A. Fire lasting > 10 min 1. Observation of a fire ATT 17 that affects plant lasting > 10 min that UE operations, causes a mode change or power reduction, or that hampers site personnel in the perfor-mance of duties necessary
. .
for the safe operation of the plant.
1. Fire in an area potentially ATT 18
.B. Fire potentially A compromising the affecting a safety system:
function of one or a) containment more safety systems , _b ) control room / protection racks c) relay room
.
d) auxiliary building e). service water intake structure
. f) penetration areas g) fuel handling building h) switchgear, rooms (64' or '
84' el); OR 2. Fire that, in the judgement
* of the SSS/ECO, could affect a safety system .
C. Fire comp romising the l '. Fire in an area that has ' function of one or more affected a safety system: ATT 19 a) containment SAE safety sys'tems b) control room /p rotect ion
*
racks c) relay room d) auxiliary building e) penetration areas g) fuel handling buildtng h) switchgear rooms (64' or 84' el); OR
, '
2. Fire that, in the judgement of the SSS/EDO, has affected * a safety system D. Any major internal or 1. Any fire exceeding the design ATT 5 external events (eg. abilities of the Fire Protection GE fires, earthquakes, System and unable to terminate substantially beyond by additional Fir Company sup-
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design basis) which port such that in the judge-would cause massive ment of the SSS/EDO will re-common demage to plant sult in the Emergency Action
.
systems which result in- Levels of.Section 5, Sec' ion a General Emergency 6G, Section 7D, Section 150, Section 17D, Section 17E
_.
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EVENT CLASSIFICATION SECTION 10 10. Earthquake / Severe Weather Initiating Event / Notification / Dmergency Action Level Recorting Condition A. Earthquake / seismic 1. OHA B43 (Unit 1 CR) alarms; event felt in-plant AND or detected on station 2. Seismic monitor is record-seismic instrumentation ing; AND * 3.. Seismic disturbance level: ATT 20 a) 10.02 g UE b) > OBE levels (1 0.1 g) ATT 21
. A c) > DBE levels (1 0.2 g) ATT 22 (call National Earthquake SAE
- Information Center 303-234-3994 for verifi-cation)
8. Floods 1. Tide level recorder shows: al > 97.5 feet ATT 7
.
~
UE
~ . b) > 99.0 feet ATT 2 ,
A
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c) -> 100.5 feet ATT 3 SAE C. Low water levels 1. Tide level recorder shows: ATT 7
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a) < 83.1 feet
-
UE b )- < 81.0 feet ATT 8
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A c) < 78.4 feet ATT 9
~
SAE 0. Hurricane or' unusual 1. As indicated by 33', 150', wind conditions or 300' elevation wind * speed channels on the *
* Meteorological Tower: .
a) sustained winds > 90 mph ATT 7 UE b) sustals.ed winds 195 mph ATT 8
, A c) sustained winds '~> 100 mph ATT 9 SAE I 1. Observed tornado funnel cloud ATT 7 E. Tornado on site within Minimum Exclusion Area UE (MEA)
1. Observed tornado funnel cloud ATJ 8 F. Tornado striking A facility . within Security Boundary
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1 of 2 Rev. 0 SGS
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EVENT CLASSIFICATION SECTION 10 10. Earthquake / Severe Weather (con'd) G. Tornado on site that 1. Tornado affecting: ATT 3 affects safety a) turbine building SAE structures b) service building c) auxiliary building d) containment
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e) service water intake struct-ture f) RWST, PWST, or AFWST g) fuel handling building H. Any major internal 1. Any severe weather conditions ATT S or external events causing damage to plant systems GE (e.g., fires, earth- that in the judgement of the quakes, substantially SSS/EDO will result in the ' beyond desing basis) Emergency Action Levels con-which could cause tained in Section 5, Section massive common, damages 6G or Section 70.
to plant systems which would result in a General Emergency ,
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EVENT CLASSIPICATION SECTION 11
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11. Site Bazards (explosions, crashes, etc.)
Init'iating Event / Notification / . Eme rc ency Action Level Re;orting t Condition ' 1. Crash within Minimum Exclusion ATT 7 A. Aircraft crash UE i occurring nearsite Area (MEA) > 1. Crash within Security Boundary ATT 8 B. Aircraft crash A occurring onsite . 1. Crash causing damage or fire. en ATT 19 C. Aircraft crash SAE ' affecting plant a) turbine building ' structures b) service building
.
c) auxilia ry building d) containment e) service water intake 8 strteture f) RWST, PWST, or AFWST ' g) fuel handling building D. Tu rbine rotating 1. Turbine trip ATT 7 UE component failure .
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E. Turbine. rotating 1. Turbinr trip AND
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ATT 2 2. Observaition of penetrations A component f ailure * causing casing pene- throug: outer castng i tration . F. Missile impact onsite 1. Missilar impact, from any sourte, ATT 8 j causing severe structural damage A
. .to any. building within the '
securit y Boundary G. Missle impact onsite 1. Missile impact, from any source, ATT 9 causing structural damage to: SAE . damaging a vital ' structure a) turbine building ' b) service building c) auxiliary building d) containment el service water intake s tructu re f) RWST, PWST, or AFWST g) fuel nandling building 1. Explosicn/ combustion or its ATT 17 H. Unplanned exple:isn UE affecting plant consequences within the operations security bou,ndary enat causes . a powe r reduction or a mode enange; ce that hampers site
' personnel in the performance of duties necessary for ene
. ! safe operation of the plant t '.. ,
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1 of 2 Rev. 0 SCE __ -- -_ _ . _ . _ ___ _- --
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EVENT CLASSIFICATION SECTION 11 (con'd) 8. Unplanned explosion 1. Explosion / combustion in one or ATT !! more of the following areas: A potentially comp romising the function of one or a) containment more safety systems or b) control room / protection normal operation of the racks plant c) relay room d) auxiliary building e) service water intake
- -
structure
- f) penetrucion areas g) fuel handling building * '
h) switchgear rooms (64' or 84' el
'that could have, in the judgement .
of the SSS/EDO, affected a safety
* -
system or affecting normal plant operation e ATT 19 J. Unplanned explosion 1. Explosion / combustion Ln one or more of the following areas: SAE compromising the function of one or al containment more safety systems b) control room / protection racks
.
c) relay room d) auxiliary building
.
e) service water intax1 structure f) penetratton areas g) fuel handling cutiding h) switchgear rooms (64' or 84' el) that has, in the judgement of
'the SSS/E00, affected a safety system 1. Any event such as an aircraft ATT $
X. Any major internal et crash, explosion or m,tssile GE external events (e.g. , fires, earthquakes, impact causing damage to plant substantially ceyond systems that in ene judgement design basis) which of the SSS/ECO will result in the General Emergency Levels would could result in contained in Section 5, Section a General Emergency.
6G, Section 70, Section 150, Section 170, Section 17E
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i EVENT CLASSIFICATION SECTION 12 12. Persennel Emergencies Initiating Event / . Notification / emercency Action Level Reporting Condition A. Transport of a 1. As stated ATT 14 UE contaminated individual from site to offsite medical facility B. Any serious' inju ry 1. As judged by SSS/ECO ATT 24 occurring onsite ocner C. Any fatality 1. As stated ATT 23 * occurring onsite (50.72D - 4Hr) O. Any personnel 1. Reported by HP: ATT 13 overexposure a) > 10 CFR 20.101 limits Otner b) > 10 CFR 20.103 limits
.
c) > 10 CFR 20.104 limits E. Any significant 1. Any individual' receives: ATT 10 personnel over- a) WB exposure > 5 rem (50.72D - 4Hr) exposure b) skin exposure > 30 rem
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c) extremity exposure > 75 rem F. Any serious personnel 1. Any individual rec'eives: ATT 11 overex;osure . 'a ) WB exposure > 25 rem (50.723 - 4Hr) b) skin exposure > 150 rem c) extremity exposure > 375 rem
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EVENT CLASSIFICATION SECTION 13 13. Security Events , Reference listings of Security Contingency Procedure Implementation of appropriate Security Contingency Procedure (SCP) - (see below).
. SCP 1 Loss or Degradation of Physical Security System ............. ' SCP 2 Loss of Security Computer Powe r. . SCP 3 Loss or Degradation of Commun-cation Systems .................. ,
'SCP 4 Loss or Degradation of Secur-ity Force ....................... 'SCP 5 Thceat Against the Station ...... * 'SCP 6 Discovery of Intruders or Attack ..........................
SCP 7 Internal Disturbance ............
'SCP 8 ' Hostage Situation ............... 'SCP 9 Fire, Explosion or Other Catastrophe ..................... 'SCP 10 Discovery of Sacotage Devices or
,( Evidence of Sabotage ............ t SCP 1) Civil Disturbance ... ........... .
*SCP 12 Security Alert ..................
'SCP 13 Tampe r Ala rm Annuncia t ion . . . . . . .
(if actual tampering has occured)
-
Initiating Event / . Notification / Condition cmergency Action Level Reportinq
;
A. Security Alert SCP 12 - Only ATT 25
.
UE i 8.. Substantiated SCP 1 And SCP 10, or ATT 15 threat, attempted SCF 2 And SCP 10, or UE entry or attempted SCP 3 And SCP 10, pr sabotage * S CP 4 - On ly , o r '
,
SCP 5 - Only , or , SCP 7 - Only , or SCP 8 - Only (Hostage Held
-
Offsite), or SCP 11 - Only, or SCP 11 And SCP 5
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C. Substantiated threat, SCP 3 And SCP 12, or ATT 26 attemped entry or SCP 8 And SCP 12, or A attempted sabotage SCP 9 And SCP 12, or with Security Alert SCP 11 And SCP 12 declared.
. *
Notify NRC - SGS .
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EVENT CLASSIFICATION SECTION 13 (con'd) 13. Security Events D. ongoing security
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SCP 6 - Only (Intruder un- ATT 26 compromise armed and non-violent) A
, - only, or SCP 5 - And SCP 10, or SCP 8 - Only (Hostage Held . - onsite/non-vital area), or SCP 9 - And SCP 10 ( Device, Fire in ,
non-vital a rea) or SCP 10- Only (Device in Vital Area) , E. Ongoing security SCP 6 - Only. (Intruder Armed or ATT 27 Violent in non-vital Areal SAE compromise involv-ing imminent loss SCP 9 - And SCP 10 (Device or Fire of physical control in vital areal of the plant SCP 10- (Device in Vital Area) F. Ongoing security ccm- SCP 6 - Only (Intruder Armed) ATT 28 or Violent in Follow- GE promise resulting in the loss of physical ing Vital Areast control of the plant 1. Control R6cm 2. Hot Shut down Panel SCP 9 - And SCP 10 (Device or \ fire witn loss of vital equipment) .in vital follow-ing vital areas:
-
1. Control Room 2. Hot' shut down panel
- G. Loss, theft or diversion 1. Any loss, theft or diversidn ATT*12 of any special nuclear of any special nuclear e.aterial UE material known or suspected (eg. fuel elements or incere detectors see 10CFR 70.52) eithe r in transit to or f rom this f acility or at this facility
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EVENT CLASSIFICATION SECTION 14
. 14. Technical Specification Items Initiating Event / Notification /
Condition Emecency Action Level Reportinc A. Any event or condi- 1. Shutdown initiated ATT 1 tion requiring unit UE shutdown to comply- .
*
with any T/S LCO ,
- .8. Primary leak 1. Exceeding action state- ATT 1 a) no PRESSURE BOUNDARY ments of T/S LCO 3.4.6.2 UE leakage ,(Unit 1) or 3.4.7.2 (Unit 2)
b) 1 gym UNIDENTIFIED leakage . See Tech spec , c) 10 gpm IDE.MTIFIED Acti*on Requirements . leakage d) 40 gpm CONTROLLED leakage e) RCS PRESS ISOL VALVES (See T/S) C. Complete loss of any 1. Entering Tech Spec 3.5.3 ATT 2 function needed for Action a or Action e ' A . cold shutdown D. Exceeding any T/S 1. Exceeding T/S LCO 2.1.1 or ATT 30 safety limit 2.1.2 (!0.723 - 1HR) E. Failure of any Rx Examples: ATT 2
. trip system to a) PIR press > 2385 psig A initiate or complete .without automatic Rx trip protective f unction (s) b) inability to insert- *
as required sufficient control rods to achieve required SCM
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F. Operation less . Examples: \TT 13 conservative than a) Shutdown required by an Other a T/S LCO or prohibited A/S not begun within by T/S specified time b) entering T/S 3.0.3 G. Any reactivity anomaly 1. Disagreement with predicted ATT 13
.
steady state reactivity Other
,
balance during power ops 1 1000 pcm (per ex Engr); OR , 2'. Ca lcula ted S DM < required: OR 3. SUR > 5.2 dpm OR ' 4. Unplanned positive reactivity insertion s 500 pcm
' N. Any unplanned 1. As stated . ATT 12 criticality -
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UE SGS - 1 of 4 Rev.'O
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i l EVENT CLASSIFICATION SECTION 14 (con'd) I. Any event or condition Examples: ATT 29 that alone could have a) clogged fuel line (50.72D - 4He prevented tne function resulting in no fuel of safety structures to emergency D/GS I or systems, such as: b) multiple instrument 1. Personnel error / drift resulting in losa procedure violation of protection function: 2. Equipment failure c) failure to restore a 3. Design, analysis, safety system to opera-construction, or bility f ollowing mainte-procedural deficiency nance or testing:
*
4. Loss of a single- d) improper procedure train system allowing incorrect valve .
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5. Functionally , lineup, resulting in redundant components functional loss.of could fail by the redundant ECCS sub-same machanism systems 6. Loss / degradation of an service or input system necessary for reliable or long-te rm
+-
operation of a safety system J. Abnormal degradation of 1. Excluding valve pacting or ATT 2 i fuel cladding, RCS gasket leakage within T/S A pressure boundary, limits or containment Examples: f unction or integrity al thru-wall f ailure c f RCS during operation piping / pressure boundary components:
*
b) welding / material defects
> allowed by applicable codes (per ISI/MIET):
c) containment leak rates (e.g.
Type 8 or Type C) > l l autnorized limits: ! d) loss of containment . sol
"alve functient * el loss of containment cooling capability f) loss of RCS relief and/or safety valve operability l
l K. Natural or man-made Examples: ATT 6 events that require a) threatened civil dist urbance (50.72b - 1 plant shutdown, safety sequiring shutdowns system operation, or b) damage caused by fire, tornado, other protective earthquake, etc.; actions IAW T/S c) entering T/S A/S 3.7.5.la 1U1), 3.7.5a (U2) for flood.nq (if shutdown required, see 14A also
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108) L. Errors discovered in 1. Per Enginee, ring notification ATT 29
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transient or accident (50.72b - 4
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*** 9 nt 4 Rev. O
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a.
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. EVENT CLASSIFFICATION SECTION 14 (con'd)
' Examples: ATT 13 M. Performance of any a) SI pumps f ail to deliver other system or component requiring corrective flow rate assumed in UFSAR; action to prevent b) safety related breaker fails operation less conserva- to trip on instantaneous tive than assumed in the overcurrent; UFSAR c) RCS thermal shock from .
inadvertent SI (per Engr} N. RPS or E'SF instrument 1. Excluding surveillance testing setpoints less or preventative maintenance ATT 13 Examples other conservative than required by T/S but a) 1/4 cent press channels fail not preventing the to actuate cent spray during fulfillment of the surveillance test functional requirements b) 1/4 POR press channels cause Rx of the affected system trip at 1850 pstg instead of 1865 psig during surveillance test o. Conditions leading to 1. Excluding surveillance testing ATT 13 or planned maintenance other operation in a degraded mode permitted by an " Example: : LCo , a) SI pump failed to start following systitm initiation, redundant
'
pump tested SAT; b) CCP inoper.able due to a faulty bearing, redundant pump tested SAT p. Inadequately . Example:: ATT 13 a) D/G . ripped on hi temp due tc otner implemented procedural inco rrect SW lineups to cooler, or administrative other SAT; controls which threaten to reduce EST or RPS . b) fail. ire to perform surveillance redundancy test.nq at required frequency 1. Excluding normal valve packing or ATT'13 O. Abnormal degradation other of systems designed to gasket leakage contain radioactive Examples material (not fuel a) through-wall leak in WHUT clad, RCS, or cont)
. .
3 of 4 Re v. 0 SGS .
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. . . EVE'1T CLASSIFICATION SECTION 14 (con'd)
R. Manual or auto- 1. ECCS actuation indicated: ATT 1 matic ECCS actua- a) ECCS trip setpoint reached UE tion with dis- or manual initiation; or charge to the b) Lit logic' lights for any vessel single or conicidence initiation RP4; AND , 2. Discharge to vessel is verified by control console indication (flow, valve positions, tank levels, etc.)
S. Inoperable 1. As reported by ISI/MIET. ATT 34 snubbers Other
*
T. Special Report Examples: required a) Positive MTC (T.S. 3.1.1.4 U1, ATT 15 3.1.1.3 U2) Other b) Seismic monitoring ATT 15 ins t rumenta t ion inoperable > 30 Other days (U1 T.S. 3.3.3.3) c) Meteorological monitoring ~ ATT 15 instrumentation inoperable > 7 Other days (U1 T.S. 3.3.3.4) d) Fire detection instrunentation ATT 15 inoperable i 14 days (T.S. Other < 3.3.3.6) f) RCS activity exceeds limits ATT t (T.S. 3.4.8 U1, 3.4.9 U2) (50.72c - 1Hr) f) POPS 6r RCS vent (s) used to ATT 15
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mitigate RCS pressure transient Other-(T.S. 3.4.9.3 U1, 3.4.10.3 U2) g) Abnormal degradation of ATT 6 containment structure (T.S. (50.72b - lHr) 3.6.1.6) h) ONE fire suppression water ATT 15 system inoperable > 7 days Other
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i) BOTH fire suppression' dater ATT 6 systems inoperable (T.S. (50.723-1He)
. 3.7.10.1)
j) Spray and/or sprinkler systems ATT 15 inoperable > 14 days (T.S. Other 3.7.10.2) k) Low pressure CO2 system ATT 15 . inoperable > 14 days (T.S. Other 3.7.10.3) 1)'rire hose stations inoperable > ATT 15
.4 days (T.S. 3.7.10.4) Other *
m) Fire barrier penetrations ATT 15 inoperable > 7 days (T.S. Other 3.7.11)
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EVENT CLASSIFICATION SECTION 15 15. Electrical / Power Failures
' Initiating Event / Notification /
Cond iton Dzercency Action Level Reporting A. Loss of offsite power / 1. Shutdown IAW T/S A/S ATT 1 3.8.1.la, 3.8.1.lb UE loss of onsite AC power capability , 3.8.1.lc, or 3.8.1.1d; OR 2. Loss of 500 kv, 13 kV, and 4 kV group buses: OR 3. Rx trip on loss of 4 kV . group buses: OR 4. Loss of 4 kV vital buses , with inability to energize *
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f' rem emergency D/Gs S. Loss of cffs'ite power / 1. Loss of 500 kV, 13 kV, and ATT 2 4 kV group buses; AND A loss of all onsite AC p owe r 2. Loss of 4 kV vital buses with inability to energize f rom emergency D/Gs C. Loss of c f f site power / 1. Loss of 500 kV, 13 kv, and ATT 3 4 kV group Ouses: AND SAE' loss of a11 onsite AC ( power f or >~ 15 min 2. Loss with of 4 kV vital buses inability to energize f rom emergency _D/Gs f or
> 15 min 1. Loss of 500 kV, 13 kV, and ATT 5 D. Loss of cffsite and GE onsite pcwer with 4 kV group buses; AND total loss of aux 2. Loss of 4 kV vital buses '
feed capability for with inability to energize several tours f rom emergency D/Gs for 2 hrs;.AND 3. Aux feed shows no flow.for
. > 2 hrs 1. Loss of all 125VDC buses: AND ATT 2 E. Loss of all onsite 2. Loss of all 28vDC buses A DC power *
1. Loss of all 125VDC buses: AND ATT 3 r. Loss of all onsite 2. Loss of all 28VDC buses: AND SAE DC power f or > 15 min 3. DC power cannot be restored , for > 15 min ,
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EVENT CLASSIFICATION SECTION 16 16. Loss of Annunciators / Control Room Evacuation Initiating Event / Notification / Condition Emergency Ac tion Level Reporting A. Loss of all OHAs for 1. As stated ATT 8
> 15 min due to an- A unknown cause B. Loss of all OHAs for -
1. As stated ATT 9
> 1 hr and plant SAE transient initiated or * * in progress C. Evacuation of Control 1. As stated ATT 8 Room anticipated or A required and control of *
S/D systems estaolished locally O. Evacuation of Control 1. As stated ATT 9 ~ SAE _,,, Room required and control of S/D systems s not established locally within 15 min .
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o . O . EVENT CLASSIFICATION SECTION 17 17. Loss or Failure of Engineered Sa f eguards Initiating Event / . Notification / Condition - cmergency Action Level Reporting A. Any problem with the 1. As judged by SSS/EDO ATT 31 Rx trip breakers Examples (50.723 - lHr a) bkr does not actuate as demanded B. Failure'of the RPS to , 1. peceipt of Rx protection logic ATT 2 initiate and complete input on RP4; AND A a trip which brings 2. Not all rod bottom lights lit
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the Rx subritical or NIs indicate Rx not subcritical C. Failure of the RPS to 1. Receipt of Rx protection logic' ATT 3 automatically , or by input on RP4; AND $AE operator action to 2. Not all rod Dottem lights lit manually, initiate and or NIs indicate Rx not complete a trip which suberitical; AND brings the Rx 3. No boration capabilities
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suberitical ' O. Transient requiring 1. Rx remains critical / returns to ATT $ operation of shutdown critical after trip;.AND GE systems with failure 2. No flow shown on SI/RHR systems
.to scram, resulting in or pumps not running with SI core damage or .
initiated additional failure of core cooling and makeup systems - E. Transient initiated by 1. Rx trip on lo S/G feed flow; AND ATT 3 2. WR S/G 1evels decreasing towards GE loss of feed and cond. offscale lo on all S/Gs; AND
'followed by failure of aux feed for extended 3. No aux feed flow shown or pumps not period; core damage running 2 min after required; AND possible.in several 4. Aux feed cannot be restored within hours ,
30 min - F. Complete loss of any 1. RHR system fails to attain / main- ATT 2 tain primary system (200F; OR A function needed for cold shutdown 2. IAW T/S A/S 3.5.3a or 3.5.3b G. Complete loss of any 1. Loss of main and aux feed: OR ATT 3 2. Loss of steam dumps and all S/G SAE function needed for hot shutdown Safeties and PORVs _
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EVENT CLASSIFICATION SECTION 18 la. operational Status Changes
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Initiating Event / Notification / emercency Action Level Reportino Condition 1. As judged by the SSS/EDO ATT 2 A. Any event or condition
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A during operation that results in the condi- * tion of the plant,' ' including principal - safety barriers, being seriously degraded
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Examples: ATT 6 8. Any event or condition ($3.72b - 1 H.: during operation that a) accumulation of voids that results in the plant could inhibit the abtitty beings to adequately remove neat 1) in an unanalyzed from the reactor core condition that b) voiding in instrument Lines, "~ significantly resulting in erroneous c'ompromises plant indication, causing the safetyr operator to misunderstand 2) in a condition out- the true condition of the side the design basist plant
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. 3) in a condition not covered by operating or emergency procedures 1) As judged by SSS/EDO ATT 29 C'. Any event, found while (t0.72b - 4H s hu tdow n, that had it been found during operation would have resulted in ths plant, including principal safety barriers, being seriously degraded or.being in an unanalyzed condition that significantly com-promises plant safety 1. Action required because no ATT 6 D. Any deviation from T/S action consistent with (50.72b - LH or licens,e condition in an emergency .when license and ./S can provide
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action is needed to adequate or equivalent protect the public protection , health.and safety (ac-tion must be approved at least by a licensed SRO) v
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i. - EVENT CI.ASSIFICATION SECTION 18 18. operational statbs Change (con'd) Initiating Event / ' Notification / Condition emergency Action Level Reportinc E. Event that results (or 1. ECCS automatic initiation ATT 1 should have resulted) in set point reduced or manual UE ECCS discharge as a initiation: OR ' result of a valid signal 2. Lit logic lights for inicia-(manual or automatic) tion on RP4: AND 3. Discharge to RCS is called for or verified Dy control console indication (flow,
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valve positions, tank levels, etc.1-F. Event that results in 1. As stated, except resulting ATT 29 manual or automatic from and part of a planned (50.723 - 4Hr actuation of Engineered test safety Features (EST), Examplest
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including the Reactor - a) SEC Mode operation Protection System (RPS) D) Rx trip c) SI (without water into RCS) ( d) Containment isolation G. Scheduled snutdown for 1. Any planned shutdown ATT 32 testing, maintenance Other refueling H. Derating' caused by 1. Upon official notification ATT 32 regulatory action from site madagement or NRC Other I. Any reactivity ancmaly 1. Disagreement with predicted ATT 13 steady state reactivity Otner balance during power ops 1 1000 pcmr OR 2. Calculated SCM < required OR 3. SUR 1 5.2 dpmr OR .
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4. Unplanned positive reactivity insertion > 500 pcm J. knyunplannedcriticality 1. As stated ATT 12
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K. Any event or condition 1. 31.wedown initiated ATT 1 requiring unit shutdown UE to comply with any T/S LCO , U SGS .? of 3 , Rev. 0
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EVENT CLASSIFICATION , SECTION 18 18. Operational Status Change (con'd) Initiating Event / .; Notification / Condition tmercency Action Level Reportinc L. Major loss of emergency 1. Loss of significant portion ATT 6 assessment capability, of control room indication (50.72b - 1H offsite response or plant monitors necessary . capability, or communi- , for accident assessment: OR cations capability 2. Loss of emergency communica ,
,tions, including ENS: CR 3. Loss of Puelic Prompt Notifi-cation System (sirens) -
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M. Excessive heatup/ 1. Exeeding limits of ATT 34 cooldown rates in T/S 3. 4. 9.1 o r 3. 4.9. 2 otner RCS or PIR N. Event re.aulting in 1. As stated - valve (s) - ATT 15 challenge to PCRV actuatt or should Other or safety valve have actuated
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O. Event resulting 1. As star.ed on actual ATT 34 in a safety valve discha.ge of safety otner discharge valves
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EVF.NT CLASSIFICATION SECTION 19 j 19. Public Interest Items
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Initiating Event / Notification / Cond i tion emeroency Action Level Reporting j A. Any plant conditions 1. As judged by the SSS/EDO ATT 7 that warrant increased UE awareness on the part of STATE / LOCAL . '
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authorities . 8. Any plant conditions 1. As judged by the SSS/ECO ATT 0 that warrant
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A precautionary activation of the TSC and placing EOF and other key emergency personnel on standby C. Any plant conditions 1. As judged by the SSS/ECO ATT 9 that warant SAC precautionary activation of the TSC and EOF * and/or notification to the qine~ral public
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i 0. Unusual movements of 1. As judged by SSS/ECO ATT 32 equipment or. personnel , Otner which may significantly i affect local traffic patterns
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, " E. Onsite events which 1. As judged by SSS/ECO ATT'32 involve alarms, sirens Otner or other sources of noise which may be . .,- heard offsite F. Transportation of hi 1. As judged by site management aft 32 or lo level radioactive Other ma*erial through LAC ,
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I G. Unus'4.11y large fish 1. As judged Dy SSS/EDO ATT 32 killa Other H. Protected *acuatic 1. As reported by Icythological ATT 33 species impinges on Associates or other site Other CW or sw intake screens personnel
' (ex. sea turtle, . sturgeon)
l *,_ t. Major loss of communi- 1. As judged by SSS/EDO ATT 6 cations capability (50.72D - 1Hr) s*
(off-site sirens or telephone system) .
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Public ServiC** Electric aruj Gn Corppany C:rbin A. McNeill, Jr. Pubhc Service Ehw toc andGasCompany PO Box 236 Hancoch Brnfge NJ 08038 buq 3N) 4800 vice Prewmet . Idui f t*.lf November 26, 1985
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Mr. Don Johnson Chief Examiner U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406
Dear Mr. Johnson:
SRO EXAM REVIEW Attached please-find our comments on the SRO written examination.
There are numerous comments on the. KEY - some of which are in complete disagreement with the key answer. There are four questions that could be interpreted by the candidate in more ways than
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the key reads: - ' 5.01.a The key lists evolutions which can cause waterhammer.
The question does not really ask for that.
5.07 The word "or" tends to allow an interpretation or answer which may not be baned on reactivity coefficients only.
6.01.b Anks a conceptual question rather than a minimum value
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or basis.
8.ll.b A correction (area to procenn monitors) was made for only one candidate.
Section 8 contains an overwhelming point value of questions which require "from memory interpretation" of Technical Specificiationn and various other procedures. This in in direct contrant with the actual and good practicon annociated with the SRO ponition.
, In the control Room and SSS Office, all reference materialn necessary l to make thene decisions are available. When faced with perplexing or questionable situations, the Sno in encouraged to nolicit qualifled advice. In no case in a decision baned on memory or annumed familiarity with a procedure. Questionn 8.07 and 8.08 alono contain 20% of the point valu f the nection and in mont canon represent what our reservations are with thin noction - referencen would be used to make a decision (real world), but on this examination, you are right or wrong.
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Mr. D. Johnson- -2- 11-26-85 This is not meant to negate or detract from the need for license holders to memorize those items that are required and necessary, and be familiar with approved procedures. Consideration should be given during the grading of this section to the fact that substantially more information is available (on-the-job) to the SRO for evaluating these situations. Otherwise, candidates could gcore <70% on this section and really not be unqualified for the job / position.
Our comments on the key are attached with their references (where appropriate). My training staff is available to discuss these issues with you, if you desire.
Sincerely, Attachment -
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SRO Licensing Exam Key Correcti.ons/ Comments Category 5 5.01.a Answer indicates operational evolutions which could cause waterhammer. However, question asks for causes of waterhammer which is: 2-phase flow (stm/ water air / water) thru an abrupt change in flow direction (open valve, start pump, pipe bend).
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.b Answer should also include:
Allow time for thermal equalization 5.02 Answer is correct except for last part of second line,
"Rx power increases due to the positive reactivity added through MTC and Doppler". As power increases, Doppler is adding negative reactivity to balance with positive reactivity due to MTC.
5.03.a OK
% .b OK 5.04.a-e Trainees provided natural circ. parameters as specified in the EOP's not the General Physics Heat Transfer Book
~ as referenced on ANSWER KEY (There are some differences such as value for Pzr level).
Also, Part c answer should say "S/G Pressure Tracking
]31 saturation pressure not Tang. (Tang is not a viable indication without RCP's in service).
Answer key has an extra answer marked d. which has no corresponding question.
5.05.a Answer key has (3) parts First - The comparison of rod movement vjt dilution Second - Why we do an ICCR Third - A statement about doubling count rate and having distance to criticality.
The question only asks to compare two different reactivity addition methods and explain why the limits for conducting an ICCR are different. "First" statement on ANSWER KEY does this. Other acceptable answers should be: S/D banks moving out could uncover a source range detector
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and cause count rate to increase rapidly so band is bigger.
~ Rate of reactivity addition ( K/k) for rods is different than boron dilution so different limits.
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"Second" answer was not asked for in question " third" answer is irrelevent to question.
.b Answer key states "... tolerances on the work sheet".
Should be:
>650 pcm due to rods , >1000 pcm due to xenon
! 500 pcm past ECP with no criticality I 5.06.a&b Both questions ask for current yet ANSWER KEY talks l about power (in part a the current would follow ppower with constant voltage so merely substitute current for power).
.b Answer indicated DECREASES, which implies it is steady at some lower value. With voids in system, the motor current would most likely fluctuate or increase and decrease.
Ref: Reference given on KEY does not really apply to a voided condition but to normal ops.
5.07.a Answer is correct except that for the present cycle cores at Salem, the doppler only power coefficient is more negative at EOL not less. (Doesn't matter to overall a answer since MTC is more dominant). **
.b Another acceptable answer.
AT EOL, we run closer t- IB (lower DNBR) due to flux shift to top of core r pressure at top of core, higher temp. at top e. Loss of flow causes.a shift in curve which woulo .ne the more limiting time at EOL.
.c Answer is incorrect for current Salem core cycle due to doppler only power coefficient being more negative at EOL.
5.08.a Answer is incorrect for Salem. Normal pressure setting for stm. dump at Salem is (1005-1035) by lowering setpoint to 900 psia you would dump more steam and cause tavg to decrease adding positive reactivity to the actual critical position (ACP) would be lower.
.b OK
.c OK 5.09.a Since question asked for "Beginning of Cycle" whihch is underfined and due to picking off numbers from a graph the ANSWER KEY should reflect a range of acceptable answers, not just one answer.
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- .b Since question indicated that Rx had already tripped with all rods in than the credit for the stuck out, most
.' reactive rod is not applicable. Answer should be tho' difference between answer in part a and Tech Spec limit (1.6% -K/k or 1600 pcm)
.c ECP also considers reactivity effects due to the following:
Samarium Power Defect (Since we were initially at 75% and now S/D, the power defect would add +p due
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to S/D. Going critical does not remove this +p addition so it must be taken into consideration.)
~6.01.a Additional answers: Design sizing of cc surge tak RMS monitors closing vent valve.
.b ANSWER KEY gives 15 psig valve, question didn't ask for this. Answer should be simply, " causing sufficient flow / cooling to-#2 seal".
6.02 The first part of answer (.4 pts) isn't necessary to answer question, only that indicated flow is higher than actual flow.
6.03.a Answers are OK, but the reference should be the Tech Specs not SNGS Notebook.
.b First part of answer is unnecessary. Only the second part is needed (limits range OT T must operate in due to low pzr pressure Rx trip.)
6.04.a Actually shifts the ventilation to " accident inside air" (which is still a control room ventilation isolation)
.b Should be Hepa plus charcoal. Also, must alarm name be mentioned since question only asks for any interlock? .c OK .d OK 6.05.a Questions ask for 500 KV BKR's yet ANSWER KEY is for 13 KV ring Bus BKR's.
Also, question only gives status of #2 generator (not
#1) so Unit 1 output BKR's may be shown as open or closed depending on whether trainee considered Unit 1 on or off'line.
.b OK
.c 1T60 and 2T60 ciruit switchers may be shown open or closed for a blackout condition. The only auto open signal
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for these is a groun'd fault on 13 KV Bus Section 1 or 4, so a blackout condition would leave them closed unless
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the operators manually opened them.
6.06.a May see these for components: SGFP Turbine (emergency oil pump) Generator (emergency seal oil pump)
.b Question asks for "cooldown after a loss of power accident."
The exam proctor stated that the vital busses were still
, to be considered energized (i.e., not a loss of all AC)
for this to occur the Emergency Diesel Generators must be the source of power to the vital Busses. With this in mind, additional support systems needed would be: Emergency Diesel Generators Service Water (to diesel generators) Atmosphere Steam Relief (MS-10 or main steam safeties) to allow you to cooldown to PNR initiation point of 350*F.
ANSWER KEY reference is LOPA 1 which is for loss of all AC (less than 2 vital Busses energized which is not the situation given in,the question.)
p ,. 6.07 Answer: H1 Head SI pump -flow rate is between 150 gpm - 550 gpm, shutoff to runout. For 1700 psig would be about 350-400 gpm.
350/400 x 2 = 700 to 800 gpm Intermediate head SI pumps have a shutoff head of about
.1550 psig and would not be injecting but running on recirc.
The positive displacement charging pump would remain running if already in that condition (no vital Bus component tripping for Mode 1 see actuation 50 could see additional
. 75 gpm)
l Answer should be in the 700-900 gpm range.
6.08.a Power range NI's are not an input to the SGWLC System at Salem.
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.b OK i 6.09.a Also manual SI actuation !
! .b OK
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l.10.a OK l b. Answer. 256 is for bank overlap unit total counter not Bank B step counter. Bank C starts to move out when Bank B is 'l i L'
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at 128 steps. Also, this does not correspond to 50% withdrawn.
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.c OK .d (1) OK (2) OK (3) Rods would step in for some time frame due to the powe mismatch circuit associated with auto rod control.
. 7.01 OK 7.02 OK 7.03 The pump combination listed on question is an acceptable combination to prevent the inadvertent SI created by 5/6 Diff. Press. The only correct answer is: " uneven RCS temps among the loops" 7.04 ANSWER KEY references the Westinghouse Owner Group Emergency Response Guidelines and not the Salem Specific E0P's so the parameters and setpoints are not entirely correct.
Should be: _ Nuclear Power >5% after Rx trip Core Exit Tc >1200 F
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Core Exit Tc >700 F and RVLIS-full range not >50% and no RCP's running and <10 F SWBC001 NR S6 Level <l5% and feed flow <22E4 Tc decrease >100 F in 60 min and to left of limit A on graph (not Tc<16 F) CTMT press >47 psig (not 17 psig) 7.05.a Break flow and injection flow, find new equalibrium at a new lower value.
.b As pumps are taken out of service, the effect on total flow (% flow change) is greater and if when a pump is taken out of service, the break f'sw is now > injection flow, the press. would decrease ar... subcooling would decrease.
l 7.06.a OK
.b OK 7.07.a OK l .b OK l
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7.08 ' i
' : . 1. Also, would. exceed 2000 mr/qtr which requires a dose extension.
i 2. ANSWER KEY should be 1000 mr/qtr not 2100 mr/qtr. l 3. Answer should show limit as 500 mr/ gestation period 4. Also, PBR our AP this individual would have already exceeded
the 100 mr/qtr limit which would have required some dose extension.
- 7.03.a
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Also. ESF instrumatation (not listed in the AOP) ! .b 2c Diesel generator is inoperable so verify other AC sources operable within (1) hr.
.c OK j 7.10.a OK l
.b OK i .c Shutdown reactor only (if 1865 psig auto Rx trip)
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.d OK ;
i .e Restore containment integrity within (1) hour 53; commence shutdown (per Teach Specs) 7.11 OK
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8.01 This is not a problem that can be corrected by a procedure
change. Correct answer should be: ;
Workers would not be allowed to install these supports ,
; since it deviates from original work package. Would
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need a system design change initiated with appropriate , QA, Enginneering, etc.,-review and approval.
' 8.02 Also PER " Event Classification Guide Attachment 29 Notifications"
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STA PRI/SEC Communicators i NRC i Operations Manager cg; Operations Engineer , Lower Alloways Creek (local municipality)
i 8.03 OK .- 8.'04 'Vice President Nuclear Notification within.24 hours (additional answer) 3_ 8.05 OK !~ 8.06.a OK i ! .b OK i i
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8.07.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 needed.
.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 tags.
.c Red blocking tag with a temporary release to jog pump.
Could not jog pump with a yellow permissive tag since yellow permissive tags must be used in conjunction with red blocking tags.
.d Red blocking tag with a temporary release or none.
Worker blocking tags are only authorized to be used by T&D personnel (Transmission & Distribution) per AP-15.
.e For a field switch - OK For a control room switch - No, notation is shift logs 8.08.a Yes, there is no bistable for OP T 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'~cEahge )Do not enter Mode 5 until last bolt on head is tensioned) No, procedure to lift head also includes steps to tension og studs so permission would be given to complete procedure not iust a position.
Bottcm line is, did the trainee know that you can't change modes in an action statement?
.c Answer is totally wrong!
Snould be yes. 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 anything.
.d OK
.e OK 8.09.a OK .b OK 8.10.a 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
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- and would take appropriate action
.' (.8 gmp)(60 min) = 48 gal /hr ! (48 gal /hr)(24 hrs) = 1152 gal / day .b OK I -8.ll.a OK i- .b Question incorrectly states area monitors R41B, C (These I
are process monitors). Exam proctor did correct this for (1) trainee who questioned it, however, no general
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correction statement was made to_ group, i The question doesn't provide sufficient information to be specific as to classifi' cation. In addition to ANSWER
.i KEY, other possible answers would be:
' l. Plant vent >1.5E3 ,
: , Section 6, Item A Attachment 29 notification ' ,
2. >50 gpm leakage I Section 1, Item B - Attachment 2
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. 3. Based on " area" RMS monitor Section 18, Item C
- Attachment 29 i.
! .c OK
.d OK
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ACTIONS COMMENTS / CONTINGENCY ACTIONS
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3.23 Is average RCS temp less than 554*F? YES N O - - - - -- - - - a. PERFORM step 3.23.1 when
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554*F.
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t b. GO TO step 3.24.
3.23.1 CLOSE:
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a. ll-14BF19. Feedwater Control Valve.
b. 11-14BF40. Feedwater Control Bypass, c. ll-14BF22. SG Inlet Stop/ Check.
3.24 Are both Reactor Trip breakers open? YES NO ---------
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a. DISPATCH Operator to
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open, THEN request the ,
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I&C Department jumper in
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a P-4 signal for that t train.
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3.25 CONTROL natural circulation with AFW flow and dumping steam.
3.26 MONITOR natural circulabion with the following trended parameters: 3.26.1 Core Exit TCs atable cg- decreasing.
3."26.2 RCS Hot Leg temp . stable or decreasing.
3.26.3 SG press stable or decreasing. l
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3.26.4 RCS Cold Leg temp 8 hit at saturation temp for SG press.
Salem Unit 1 12 Rev. O
M LNb Fail Paav 1
, QS- SECTf0N 1.7
. ESTIMATED CRITICAL POSITION WORKSHEET
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3.0 LIMITS ON CRITICAL ROD POSITION BANK C STEPS 3.1 INSERTION LIMIT (FIG.14):
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3.2 INTENDED R0D POSITION +500 PCM (ITEM 2.4.+500) - PCM ROD POSITION AT THIS WORTH (FtG. 4, HZP): BANK C STEPS
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BANK D STEPS 3.3 INTENDED R0D POSITION -500 PCM:
(2.4)-500 - PCM ROD POSITION AT THIS WORTH (FIG. 4, HZP): BANK D STEPS A DDT 4.0 REACTIV11Y CHANGES AND Sur.
PCM 4.1 CONTROL RODS (ITEM 1.5 - ITEM 2.4) = 4.2 POWER DEFECT: (AT POWER IN 1.3 AND BORON
+ PCM i CONCENTRATION IN 1.4) (FIGURE-2) = Cts. ygg,
- 4.3 XENON FIACTIVITY .
(A) XENON REACTIVITY AT TIME IN 1.2: (FtG. 6) (-) PCM HRS.
(B) ELAPSED TIME FROM 1.8 TO 2.2:
(C) XENON REACTIV!TY AT TIME IN 2.2:
(FtG. 8) -
(-) PCM PCM (D) REACTIVITY CHANGE (ITEM C - ITEM A) =
4.4 SAFARIUM REACTIVITY HRS.
(A) ELAPSED TIME 4.3 (B):
~
(B) REACTIVITY CHANGE (FIG. 10): (-) PCM
-
4.5 REACTIVITY CHANGE SUM: , _
+ + + " PCM (4.1) (4.2) (4.3D) (4.4B)
NOTE: IF THE ABSOLUTE MAGNITUDE OF ITEM 4.1 IS GREATER THAN 650 PCM - OR, IF THE ABSOLUTE MAGNITUDE OF ITEM 4.3D IS GREATER THAN 1000 PCM, THEN USE AN ICRR PLOT TO GUIDE CRITICAL APPROACH.
. . .
. W
~
JC05b IOP-3
.3 5.13.1 The Control Bank height at Criticality is below the Rod Insertion Limit for zero power. .
a. NOTIFY the Senior Shift Supervisor / Shift Supervisor. -
*
b. INSERT all Control Banks.
c. CALCULATE a Shutdown
'
Tech Spec Margin 3.1.1.1 d. RECALCULATE the ECP e. DETERMINE AND CORRECT the error 5.13.2 The Control Bank height at Criticality is 500 PCM below the ECP but above the Rod Inser,, tion limit.
no . .- a. NOTIFY the Senior Shift Supervisor /Shif t Supervisor.
- , b. INSERT all Control Banks.. , c. RECALCULATE the ECP d. DETERMINE AND CORRECT the error 5.13.3 Criticality has not been achieved with the Control Banks withdrawn to an equivalent of 500 PCM past the ECP.
a. NOTIFY the Senior Shift - Supervisor / Shift , nc
' , ' 73 i Q
Supe rvisor.
.0
l K b. INSERT the Control Banks OQ to the position for 500 PCM below criticality c. RECALCULATE the ECP Salem Unit 1 7 Rev. 5
IOP-3
. ) .
d. DETERMINE AND CORRECT Mark N/A the the error. If no error steps not was found, . required.
1. REQUEST confirming boron sample 2. NOTIFY Reactor
- Engineering . 3. FULLY INSERT all Control Banks 4. PULL rods for criticality using an ICCR plot 5.14 When the P-6 (Source Range Permissive)
light is energized, BLOCK the Source Range High Flux Trip by depressing both the BLOCK SOURCE RANGE "A" and
.
BLOCK SOURCE RANGE "B" pushbuttons
** ,
on the console.
. ; y -. - 5.14.1 VERIFY the Source Range Trains A&B TRIP BLOCKED light on RP4 and the SR Detector Voltage Trouble Overhead ' Annunciator alarm are energized.
5.14.2 When P-6 is exceeded and the Source Range Channels are blocked, the Second Pen on the NR-45 should
,
I be switched to the other Intermediate Range Channel.
5.15 LEVEL off the Reactor Neutron level at 1.0E-8 amps in the Intermediate Range.
5.15.1 VERIFY Critical Rod Position is not less than the limits o f Tech Spec 3.1.3.5 and Figure 14 of the Reactor Engineering , Manual within 4 hours of ~ achieving criticality.
l .+I i i ,.sO)ki@
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. . s.
-
2%^) * h'i* Salem Unit 1 8 Rev. 5
" M# b Core Parameter 9 Data Sheet Fh A u,- w-.h.
Dar. j h k,, Unit 1 Unit 2
'
Item / Parameter . Cycle 6 Cycle 3 1. Core Rating 3338 N t 3411 Nt 2. Full Length Control Bods 53 53
.
3. Control Rod Worth: a. Control Banks (BOL) 3475 pcm 4350 pcm b. Shutdown Banks (BOL) 4475 pcm 2650 pcm 7590 7000 4. Enrichments: (w/o) a. Region 3 3.12 b. Region 4 3.41 3.80
-
c. Region SA 2.80 d. Region SB 3.40 3.40 e. Region 6 3.40 f. Region 7 3.40 g. Region 8A 3.40 h. Region 8B 3.80 5. BPR's (12.5 w/o B2 03) 1660 Fresh 1664 Fresh 6. B eff (Effective Delayed Neutron Fraction) - a. BOL .006116 .006292 b. BOL .004994 .005014 300-18:23
/ ! / .
Unit 1 Unit 2 Item / Parameter Cycle 6 . Cycle 3 7. Xenon:
. a. 1004 (Equilibrium) -3175 pcm -2725 pcm b. 50% (Equilibrium) -2600 pcm -2275 pcm c. 1004 (Peak) -5700 pcm -5200 pcm d. 50% (Peak) -1425 pcm -3100 pcm 8. Samarium a. Equilibrium 588 pcm 588 pcm b. 1004 (Peak) 944 pcm 944 pcm 9. Doppler Only Power Coe f ficient $ a. HZP (BOL) -15.0 pcm/t -14.2 pcm/t b. HFP (BOL) - 9.8 pcm/t - 9.6 pcm/t Average (BOL) 2 -12 pcm/t 2 -12 pcm/t c. HZP (EOL) -22.5 pcm/t -24.2 pcm/t d. HFP (EOL) -9 pcm/t - 9.6 pcm/t Average (EOL) 2 -16 pcm/% 2 -17 pen /1 10. mderator Temperature Coefficient
a. BOL (ARD, EPP, 5710F/1000 PPM) -12.1 pcm/OF -17 pen /oF b. EOL (ARD, H'P,F 5710F/0 PPM) -32.1*pcm/oF -33* pen /oF
'
* Note: * Notes his value by h is value by Tech Spec is Tech Spec is allowed to be allowed to be q -38. PPM point -40. PPM point , k 300 PPM point 300 PPM point must not be must not be more NEG. than more NEG. than-29. -31.
300-18:23
_ . _ _
.
O Unit 1 Unit 2
~
Item / Parameter Cycle 6 Cycle 3 11. Power Coefficient
. a. BOL (HZP 1000 ppm) -17.6 pcm/t -21.5 pcm/t b. BOL (HFP 1000 ppm) -15.0 pcm/t -16.2 pcm/t Average BOL = -16 pcm/t e -19 pcm/t Power Coefficient (Cont. )
c. EOL (BZP, 0 ppm boron) -33.0 pcm/t -34.5 pcm/t d. EOL (HFP, 0 ppm boron) -22.5 pcm/t .-24.2 pcm/t O Average (EOL) = -28 pcm/t e -29 pcm/t
'
12. Doppler Only Power DEFECT a. BOL -1200 pcm -1150 pcm b. EOL -1310 pcm -1400 pcm 13. Power Defect a. BOL (900 ppm) -1625 pcm -1750 pcm b. EOL (0 ppm) -2350 pcm -2550 pcm 14. Differential Boro'n Worth a. BOL (800 ppm) = -8.4 pcm/ ppm u -8.45 pcm/ppe b. EOL (0 ppm) = -10.2 pcm u -10 pcm/ ppa
. '
300-18:23
- . - _
__ ___ _ ___ . i
Unit 1 Unit 2 Iten/ Parameter * Cycle 6 , Cycle 3 THUMB RULES
,
Most Reactive Stuck Rod Worth 1285 pcm 900 pcm Boration Factor 3 gal / ppm 3 gal / ppm Boron Worth 10 pcm/ ppm 10 pcm/ ppm
..
O .
..
.
_ Q 300-18:23
. _ - . _ .. . .-_ .- . . - - _ _ - .
. -
jgog ou _ III-2.3.4
. .
5.0 PROCQDpy 5.1 Operation During Plant Startup 5.1.1 VERIFY that the Co'ndenser Vacuum permissive light on 1RP4 is ON.
5.1.2 If Tavg block is bypas. sed (Tavg Bypass light ON) , DEPRESS the Train "A" and'"D" OFF-RESET Bypass
. Tavg pushbuttons to reset.
5.1.3 PLACE Steam Dump Interlock Train "A" and "B" controllers to ON.
5.1.4 If the Turbine Impulse Chamber Pressure permissive light is ON, DEPRESS the RESET LOAD REJECTION pushbutton. VERIFY that permissive light goes out.
5.1.5 ADJUST Main Steam Pressure setpoint to desired value. This' is 1005 psig for a no load Tavg of 547'F.
5.1.6 . VERIFY that all of the Steam Dump Valves are
, closed.
'
/
5.1.7 PLACE system in Main Steam Pressure Control Mode.
- 5.1.8
" ~ dP" VERIFY that the Block Cooldown and Block Non-Cooldown lights are ON. (This indicates that the valves are blocked) . As Tavg increases to
~
greater than 543*F, the Block Cooldown and Block Non-Cooldown lights will go OFF.
5.1.9 PLACE the Main Steam Pressure Controller in AUTO.
5.1.10 OBSERVE that the Steam Dump Valves open and MAINTAIN Main Steam Pressure as plant heatup to 547'F is completed.
5.2 Operation While At Power 5.2.1 VERIFY that the loss of load interlock is reset by observing that the Turbine Impulse Chamber Pressure permissive light on the status panel is OUT. * 5.2.2 If the loss of load interlock is not reset, PUSH the Reset Load Rejection pushbutton. VERIFY
-
reset by Turbine Impulse Chamber Pressure light going OUT.
{ Salem Unit 1 2 starlu 5 b M Rev. 4
W RX ENG MAN PART 1
. 6.oi-- C '
. i SECTION 1.7 ESTIMATED CRITICAL POSITION WORKSHEET 1.0 PREVIOUS CRITICAL CONDITIONS 'TROM CONTROL ROOM LOG) 1.1 DATE: 1.2 TIME: 1.3 % POWER: PPM 1.4 BORON CONCENTRATION: STEPS 1.5 CONTROL BANK POSITION: BANK C BANK D STEPS
' PCM 1.6 INTEGRAL ROD WORTH (FIG. 4)
MWD /MTU 1.7 CORE EXPOSURE (MWD /MTU) 1.8 iYPE OF SHUTDOWN (A) Rx TRIP AT DATE: IINE: DATE: IIME:
(a) ORDERLY SHUTDOWN
%/ MIN.
APPROXIMATE SHUTDOWN RATE:
.)
2.0 INTENDED CRITICAL CONDITIONS
~
2'.l~DATE: 2.2 TIME: (1) STERS 2.3 CONTROL BANK POSITION: BANK C BANK D STEPS PCM 2.4 INTEGRAL ROD WORTH (FIG. 4 HZP CURVE):
(1) THIS ECP MUST BE COMPLETED WITHIN FOUR (4) HOURS OF GOING CRITICAL, IAW TECHNICAL SPECIFICIATION 4.1.1.1.1(C)
U ENGR DEPr
.
DCT 7- m. .J. t _
> 04 THE.gp0T
' CNnNCEs NOT
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-
RX ENG PAN PART 1 SECTION 1.7 ESTIMATED CRITICAL POSITION WORKSHEET
'3 3.0 LIMITS ON CRITICAL ROD POSITION BANK C STEPS 3.1 INSERTION LIMIT (FIG.14):
3.2 INTENDED R0D POSITION +500 PCM (ITEM 2.4 +500) = - PCM RCD POSITION AT THIS WORTH (FIG. 4, HZP): BANK C STEPS
,
BANK D STEPS 3.3 INTENDED R0D POSITION -500 PCM:
(2.4)-500 - PCM ROD POSITION AT THIS WORTH (Fro. 4, HZP): BANK D STEPS
^DM 4.0 REACTIVITY CHANGES AND SUM PCM 4.1 CONTROL RODS (ITEM 1.6 - ITEM 2.4) -
4.2 POWEP DEFECT: (AT POWER IN 1.3 AND BORON
' + PCM '
CONCENTRATION IN 1.4) (FIGURE 2) = WL rat. spy
- 4.3 XENON PEACTIVITY .- [8"# (A) XENON REACTIVITY AT TIME IN 1.2: (Fro. 6) (-) PCM HRS.
98) ELAPSED TIME FROM 1.8 TO 2.2: l (C) XENON REACTIVITY AT TIME IN 2.2: l (Fro. 8) - (-) PCM l PCM 1 (D) REACTIVITY CHANGE (ITEM C - ITEM A) - 4.4 SAMARIUM REACTIVITY HRS.
! (A) ELAPSED TIME 4.3 (s):
(s) REACTIVITY CHANGE (Fro 10): (-) PCM
-
4.5 REACTIVITY CHANGE SUM: , _
+ + + = PCM (4.1) (4.2) (4.3D) (4.4s)
NOTE: IF THE AsSOLUTE MAGNITUDE OF ITEM 4.1 IS GREATER THAN 650 PCM
' OR, IF THE ABSOLUTE MAGNITUDE OF ITEM 4.3D IS GREATER THAN 1000 PCM, THEN USE An ICRR PLOT TO GUIDE CRITICAL APPROACH.
- ;
.---_ j '
LESSON NAME: COMPONENT COOLING WATER SYSTEM l
'
I INSTRUCTIONAL CONTENT:
)
b Ol e: 2.2 COMPONENTS , 2.2.1 Surge Tank There'is one Component Cooling Surge Tank provided in this system. The surge tank is a
, horizontally mounted carbon steel cylinder.
It is connected to the suction header of the cooling water pumps via two 4-inch lines.
The surge tank has a capacity of 2000 gallons with a normal water volume of 1000. gallons.
The tank is designed with an internal baffle to divide the tank into two separate volumes which provides redundancy for a passive failure during recirculation following a loss of coolant accident.
The Component Cooling Surge Tank capacity permits: a. The surge tank to accomode surges in system resulting from thermal expansion
.
and contraction.
, b. The normal air volume in the tank to
-
accomodate the amount of reactor coolant entering the component cooling loop following a rupture of a reactor coolant pump thermal barrier cooling coil for a period of three (3) minutes.
c. The normal water volume in the tank provides an intermediate source of makeup to the loop if a leak develops in the system. The water volume is separated into two parts by a baffle which protects against complete drainage of the tank in the case of a leak or failure in one " half" of the system.
The surge tank has a flanged connection at the top for additions of chemical corrosion inhibitor to the component cooling loop. For the purpose of homogenizing this chemical a one-inch recirculation line from the pump e is provided. The tank is normally discharg$o Veitfed 4he ABV exh. En dIsd via air-operated valve CC149,
. The ' relief valve CCl47 on the surge tank is 300-7:23 DATE: 03/01/85 Page 5 REV.: 0
-
LESSON NAME: COMPONENT COOLING WATER SYSTEM ' s INSTRUCTIONAL CONTENT:
}
The tank is vented to the d&1. dyN cf $he kOV EXlh via valve CC14,9. . The valve automatically closes upon a high radiation level.
Scintillation-type radiation detectors located downstream of each CCW heat exchanger sensing an activity of 0.5 decade above their
, minimum sensitivity closes CCl49 and actuates .the alarm "21(22) CC HEADER HIGH ACTIVITY" on the Surge Tank Vent Valve bezel. This vent valve cannot be opened until the radiation level returns to normal. Water withdrawn for radiation detection is recirculated back to the pump suction header through a 3/4"line.
The radiation detectors (R-17A and R-17B) are located in the Component Cooling Water Heat Exchanger Room, 84' elevation of the associated Auxiliary Bldg.
'The surge tank is designed to Section VIII of the ASME Boiler and Pressure Vessel Code.
The tank parameters are as follows:
) Number 1 Type Horizontal, with divider plate Design Pressure:
Internal, psig 100 External, psig 15 (vacuum breaker provided) dd/#6 Design temperature, OF 200 Normal operat,ing pressure, psig Atmospheric Total volume, gal. 2000 Normal water volume; gal. 1000 Material Carbon Steel 300-7:23 DATE: 03/01/85 Page 7 REV.: 0
- - _ - _ _
i
~ . . (,ao 3 a.. '
LIMITING SAFETY SYSTEM SETTINGS BASES
.
The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above 1.30 for control rod drop accidents. At hign power a single or multiple. rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.
Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protec-tion to the low setpoint trip of the Power Range, Neutron Flux The Source Range Channels will initiate a reactor trip at about 10counts ghannels.
per second unless manually blocked when P-6 becomes active. The Intermediate , Range Chanrals will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
Overtemoerature Delta T The Overtemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping l transit delays from the core to the temperature detectors (about 4 seconds), ' and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.
l I SALEM - UNIT 2 B 2-4
0C0.129 T LIMITING SAFETY SYSTEM SETTINGS
.
BASES Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2 and K3 inputs to the Overtemperature'dalta T channels and raising the P-8 setpoint to its 3 locp value. In this made of operation, the P-8 inter-lock and trip functions as a High Neutron Flux trip at the reduced power level.
l Overcower Delta T The Overpower delta T reactor trip provides assurance of fael integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in I density and heat capacity of water with temperature, and dynamic compensation for piping delays from the care to the loop temperature detectors. No credit
'
was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this ( specification to enhance the overall reliability of the Reactor Protection System.
Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection,,and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.
Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
SALEM - UNIT 2 8 2-5
- ..
- . -
OPERATING PROCEDURE II-17.3.2 COh"rROL ROOM VENTILATIGN OPERATION 1.0 PURPOSE . 1.1 This instruction describes the operation of the Control Area Air Conditioning System in the following modes of operation:
*
1.1.1. Normal 1.1.2 Accident Inside Air 1.1.3 Accident Outside Air 1.1.4 Fire Inside Control Area 1.1.5 Fire Outside Control Area . - 2.0 ' INITIAL CONDITIONS 2.1 The Control Room Ventilation System is aligned as follows: 2.1.1 TRIS lineup II-17.3.2 has been completed.
OR A Components Off Normal List for TRIS lineup S 2.1.2 l II-17.3.2 has been generated and all components listed have been evaluated for effects on normal system operation.
- ve 2.2 RMS channels 1RIA and 1RlB are in service IAW IV-11,3.1,
" Area Radiation Monitors-Normal Operation" and IV-11.3.3, " Process R'diation a Monitors- Normal Operation".
2.3 The normal Supply Fan " ROLL-O-MATIC" filters are selected to AUTO.
2.4 Heating water is available to the supply unit heating coils during cold weather operation.
2.5 Chilled water is available to the cooling coils on the normal and emergency supply units.
2.6 The' Fire Protection System is pressurized, and 11FP80 is open.
1 Rev. 8 Salem Unit 1
.
P 11-17.3.2
5.2.2 If fire is outside of the areas serviced by the Control Area Air Conditioning System, proceed as 1 follows: l
!
a. At 1RP2, DEPRESS the FIRE OUTSIDE CMC pushbutton module.
. b. OBSERVE the dampers listed on Table 1 shift to.the positions listed in OUTSIDE CONTROL AREA column.
. 5.2.3 Dampers ICA201 thru 1CA207 are normally open.
A fire in the proximity of any one of these dampers will cause them to close. These dampers must be reset by the fire brigade after the fire has been extinguished either at Panel 796-1, Control koom Aux. Equip Room el. 122', or locally at the damper itself.
5.2.4 If fire is in the Control Area Relay Room, the Halon System will be actuated. OBSERVE the dampers listed on Table 1 shift to the positions listed under Control Area Relay Room.
5.2.5 After the fire is extinguished, RETURN to NORMAL Control Area Air Conditioning. IAW Section 5.1.
5.3 Operation During Accident Conditions NOTE This mode is normally automatically initiated by Safety Injection or alarm of RMS channels IRlA or 1RlB for unit 1 or 2. (A signal on either unit isolates both units).
5.3.1 If both red lights on the "A" and "B" Control Area Isolation CMC switches have not illuminated, DEPRESS the ACCIDENT-INSIDE AIR CMC switch or lhP2.
NOTE , Control Room Intake Duct Isolation of either unit will isolate f both units Control Room Intake Ducts.
l ' 5.3.2 VERIFY that the dampers of Table 1 are in the positions listed in ACCIDENT-INSIDE AIR column.
Salem Unit 1 3 Rev. 8 .
, . .
IV-1.3.1 E g g _.
TABLE 1
\ 500KV BREAKER LINEUPS Lineup fl - No. 1 Main Generator Synchronized.
//
.-
{3kst.'/ ~
Lineup 62 - No. 2 Main Generator Synchronized. 9%
.
Lineup 93 - il and #2 Main Generators Synchronized.
Lineup 94 - Both Main Generators Off The Line.
, Lineup 15 - Blackout LINEUP CIRCUIT BREAKER NUMBERS NUMBER 1-5 5-6 2-6 2-8 1-8 1-9 9-10 2-10 500KV 500KV (1) (1) (2) (2) (3) (3) 2T60 1T60 1 X X X X X 0 0 X X X 2 O O X X X X X X X X 3 X X X X X- X 'X X X X 4 O O X X X O O X X "X i 0 0 0 0 0 0 0
. 5- 0 0 0 Notes Axr 7We Ow4n 1. Open when 5024 Line is not 'in service. A n o ,, g ,5 m
## #'"# ' " "
Open when 5021 Line is not inservice.
2.
(sa A "*** ~ 3. Open when 5037 Line is not in service.
>
. \ .'
Salem Unit 1, TABLE 1-1 Rev 5
._ LESSON NAME: Blackout Training; 500 kV Electrical System CONTENT / SKILLS
. , d. The neutral bushing on the high side of each of the three single phase transformers are connected directly to station ground.
01 2. Circuit Breakers (Nos. 1-5, 5-6, 2-6, 2-8, 1-8, 2-10, 9-10, 1-9) a. These breakers provide the automatic isolation of the appropriate sections of the 500 kV switchyard when a fault / malfunction is detected. These breakers also provide the capability to alter the alignment of the 500 kV switchyard from a remote location (the control
) room).
b. Ratings 1) Westinghouse - SF6 type 2) 3000 'mps a 3) 1800 kV c. More detailed information about these circuit breakers is available
in section IV. C of this document and in Attachment 1 of this handout.
3. Load Interruptng Switches (lT60, 2T60) a. These switches will automatically open when a ground fault is detected on 13 kV Bus Sections 1 or 01 4. Thus isolating the 13 kV Bus
'
Section from the 500 kV ring bus.
.
)
8320SAOP: 10 DATE: 8/29/83 vage o of 4o ROV. 4
' *
LESSON NAME: Blackout Training; 500 kV Electrical System CONTENT / SKILLS
. '
d. The neutral bushing on the high side of each of the three single phase transformers are connected directly to station ground.
01 2. Circuit Breakers (Nos. 1-5, 5-6, 2-6, 2-8, 1-8, 2-10, 9-10, 1-9) a. These breakers provide the automatic isolation of the appropriate sections of the 500 kV switchyard when a fault / malfunction is detected. These breakers also provide the capability to alter the alignment of the 500 kV switchyard from a remote location (the control
) room).
b. Ratings 1) Westinghouse - SF6 type 2) 3000 ' amps 3) 1800 kV c. More detailed information about these circuit breakers is available in section IV. C of this document 04 and in Attachment 1 of this handout.
3. Load Interruptng Switches (lT60, 2T60) a. These switches will automatically open when a ground fault is - detected on 13 kV Bus Sections 1 or 01 4. Thus isolating the 13 kV Bus ~ Section from the 500 kV ring bus.
l
' .
8320SAOP:10 DATE: 8/29/83 rage o of so REV. 4 . *
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.
-
. .
* I .
,
g
-
e& O
. -
6.08 ~ Lo&t c .
.
4 - , t
1
" ,
. hafQi>- . rods. The master cycler always transmits the zero count
"GO" pulse to slave cyclers servicing group 1 rods. The three-count "GO" pulses always go to the group 2 slave cyclers. In this way the group of a bank are staggered in movement as they are moved.
. Certainly, one question remaining concerns how the power cabinet knows which group of rods is to receive the current orders from the slave cycler since each slave can service
.several groups in a power cabinet. Remember that it was the bank overlap unit that determined which slaves were to receive "GO" pulses. Likewise, the BOU again helps out by determining the exact group in a power cabinet that will receive current orders. In the figure this signal has been labeled " Group Select" or " Multiplex" and is seen entering the power cabinets.
Bank Overlap Unit The major function of the bank overlap unit is to ensure that the control bank rods move with proper overlap and in sequence. Remember that bank overlap is only used when the Bank Selector is in AUTO or MAN, since the other positions are for. individual bank control. In the d.scussions of the slave and master cyclers, it has been showa that the bank overlap unit was responsible for directing the "GO" pulses and current orders so that the proper groups of rods would move. -
~
To accomplish its functions, the BOU has three major sections: the counter, the thumbwheel switches, and the multiplexer (or sequencer).
The counter section has two major in~ puts. It receive zero count and three count pulses from the output of the master cycler. It also receives IN and OUT signals from the supervisory circuit. By counting the total number of pulses, the counter knows how many steps the rods have taken. The IN/OUT information allows the unit to add pulses for outward rod movement and subtract for inward rod motion.
The objective of proper sequencing is to see that when the control rods are on the bottom, they move out, beginning with bank A, and are followed by B, C and D in order. They also must stop traveling when they reach the top of their travel (228 steps). Inward rod motion is in the opposite sequence. The overlap function decides when in the sequence the next bank will move. The plant uses 100 step overlap.
This means that control bank B will begin moving out when bank A reaches 128 steps withdrawn, bank C will begin moving when bank B reaches 128 steps withdrawn, and bank D will Page 26 _
. . .
b start moving when bank C is 128 steps withdrawn. Figure 16 shows the bank overlap feature for rods moving in their proper sequence. Notice that one axis is Total Step Count.
~ Bank overlap occurs at 128, 256, and 384 total steps. The other numbers represent the top oE each bank's travel in total steps.
The bank overlap counter is continually calculating the total step count based upon the zero and three count pulses.
The thumbwheel section monitors the counter and will both
' initiate the next bank's movement and stop a bank once it reaches 228 steps withdrawn. There are six thumbwheel switches, each with three adjustable thumbwheels used to establish the bank overlap and rod. travel points. Switches S1, S3, and S5 are for overlap of control banks B, C and D respectively. Switches S2, S4, and S6 are the rod withdrawal limits (228 steps for each bank) for control of banks A, B and C. The rod withdrawal limit for bank D is not covered by thumbwheel switch and will be discussed later.
' When the counter has reached the setpoint for a thumbwheel, this information is instantly sent to the multiplex section.
Two singals emerge form this section, which have already mentioned. One signal tells the master cycler which slave
: cycler should receive "GO" pulses, such as when bank movement should begin. Conversely, it would tell the master cycler to cease "GO" pulses to a slave cycler when that bank has reached its travel limit. The other signal from multiplex goes to the correct power cabinet to " multiplex",
or group select, the correct group within the cabinet.
As a result, if bank D rods are moving, the multiplex section of the BOU is directing the master cycler "GO" , pulses to slave cyclers 1BD and 2BD, while also sending.
signals to power cabinets 1BD and 2BD to make sure only bank D rods will get the firing orders from the slave cyclers.
I Realize that when an overlap condition exists, two banks are l moving at one time, meaning that the master cycler "GO" ! pulses are going to all four slave cyclers and, therefore, all four power cabinets are being used.
The bank overlap unit has a digital display and three , pushbuttons adjacent to the thumbwheels. The digital
display always shows the total step count. Under normal l conditions, this display should indicate a total count l somewhere between zero and 612 steps. The RESET button would reset the bank overlap counter back to O counts, which can be verified by observing the digital display. This is not normally done. If a reset was performed while banks are Page 27
.
d/Oct
# ' . -T The movable gripper latches now support the drive rod. The lift coil is now energized. This action results in the lift pole piece attracting and thus raising the lift armature, the movable gripper armature, the-movable gripper arms, the drive rod, and attached control rod up approximately 5/8 inch, or 1 step. At this point, the stationary coil is again energized with maximum current. The stationary gripper arms move in and lift the drive rod 1/16 inch. The stationary gripper arms assume control rod support from the movable gripper arms.
The movable gripper coil is deenergized, and the movable gripper arms move out. The lift coil is deenergized, and the movable gripper latch assembly is lowered to its normal position. Finally, the stationary coil current is reduces to its hold value and the drive rod lowers 1/16 inch. This action completes the withdrawal sequence that raises the control rod 5/8 inch (one step).
The sequence to raise (or lower) the drive rod one step is accomplished in 780 m sec. The maximum rod speed is 72 steps per minute (SPM) , or 45 inches per minute. By varying the time between steps, the control rod speed is varied between 72 and 8 SPM.
Reactor Control Unit (Figure 9) ! The reactor control unit develops the electrical control i signals to be sent to the Rod Control System to move the control rods when in automatic control. This circuit will not only determine the direction of rod travel but will also provide a rod speed signal. The majority of this circuit is located in a nonprotection section of the process control racks, with a small portion coming from power range Excore NIS. The figure provides a functional approach to circuit' operation.
Before covering the details of this circuit, it is worthwhile to take an overview of the Reactor Control Unit.
Keep in mind the overall objective of the Rod Control System when in the automatic mode of operation. The rods are positioned to keep Tavg on a program that is based upon plant power level. Having RCS average temperature increase with power helps to minimize the drop in secondary steam pressure. Turbine inpulse pressure is used as the input.to develop the Tavg program. This is an accurate indication of Page 12 ____. ._
I plant power level and normally power changes occur here - first with nuclear power following steam demand.
Auctioneered high RCS average temperature provides a feedback of the actual parameter b'eing controlled, thereby allowing fine control of T vga to its program. By comparing the program or reference temperature (created by turbine impulse pressure) with Tav a temperature error signal (Terror) .can be generated.g,The greater the difference between the two, the greater the transient has been; and an increasing signal can be used to produce a faster rod movement to help correct the situation.
Another part of the reactor control unit is the power mismatch circuit, which utilizes Ex' core NIS power range inputs and turbine impulse pressure. The objective is to provide a means for fast response to power transients. It would be easy to build a reactor control unit with a circuit just to look at the rate of change of turbine power and use it to tell the rods to move faster. However, to add overall stabilty to the system, turbine power is compared to NIS nuclear power to determine more closely the power mismatch between primary and secondary. This stability aspect can be seen more clearly after the entire reactor control-unit is discussed.
Thus there are two basic parts to the reactor control unit - the power mismatch circuit and the temperature error circuit. These two produce error signals which are combined into a signal called compensated temperature error. This signal determines rod speed and direction. A closer look at the individual components of the reactor control unit will provide better insight as to how automatic rod control is accomplished.
. A single turbine impulse pressure transmitter (PT-505) is used to represent secondary power levels between 0 and 120 percent. The reference temperature (Tr ef) for the program is developed by a special unit that takes the turbine power input and creates the program Tavq temperature band on the output. These values are currently 547 0 F and 5710F. The lag unit on the output provides some filtering but also puts a slight lag into signal response.
The actual RCS average temperature input comes from the autioneered high T avg unit. This signal is filtered (lag) and then modified (lead / lag) to improve response. This compensates for the actual delays which occur in sensing , reactor coolant temperature changes. The output is a signal representing a temperature between 5300F and 630 F.
.- Page 13 _
- pr
- !
Both the Ta vg and Tref signals then combine in a single summing unit. In mathematical terms the summer performs a T ref - T a vg operation. Depending oupon the input signals, the output of the summer could be positive, negative or zero. A positive signal translates to a need for rods to move out. The opposite movement is called for when a negative signal appears. This signal is called temperature error and is combined with the signal from the power mismatch circuit-to produce a final output.
'The two inputs to the power mismatch circuit represent nuclear power and turbine power in a range of 0 to 120 percent. A box labeled " Difference and Rate" in the figure first takes the difference between the two and then looks at the rate of change of the output. Mathematically, the difference' box performs a Qt - On Operation and the rate could be represented by d(Qe - Onl_ dt , or the rate of change of the difference. Realize, therefore, that an output will only be produced when there is a rate of change between nuclear power and turbine power.
Conversely, no output will appear if the inputs are not changing with respect to each other. Even if there is a difference between the two, if it is steady state, there will be no output from the difference and rate box. Realize also that this signal representing the rate of change of the power mismatch can be positive or negative. An increasing nuclear power with respect to turbine power would produce a-negative signal which translates to a desire for inward rod movement. Outward rod movement is desired when the output is positive, indicating that turbine power is increasing with respect to nuclear power.
The next circuit component, the non-linear gain unit, functionally performs two operations. As the name implies, it is a signal gain unit. When the' rate of change between nuclear power and turbine power is small, the signal input is small. Therefore, the gain provided for the signal is small. For larger inputs the gain is increased. The second function is subtle but important. It converts a rate of change'of power mismatch into an equvalent signal in degree farenheit (OF). Notice that the gains described are 0.30F/ percent and 1.50F/ percent. The conversion to temperature representation is much more useful to the reactor control unit since temperature is the other input to the final summer. The signal here can also be positive or negative as described for the difference and rate box.
Page 14
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,
1 . IOP-6
- *)'0]3 ,
It may be 5.15.6 CLEAR AND TAG the desirable to
-~~
Auxiliary Feedyater use the AFW Pumps Pumps to maintain
'
Steam Generator levels if so,
*
C&T may be deferred until pumps are
,'-
secured.
a. TAG the electrical power supplies for the motor driven pumps b. TRIP Pump and TAG the local reset lever for the steam driven pump '
.
REDUCE the number of If Chemical 5.15.7 addition is in running RCP's to two in'one of the following progress to the t'~ combinations RCS and more flow is a. 21 RCP and 24 RCP desirable this or step may be e' b. 22 RCP and 23 RCP marked N/A.
Operation of RCP's, other than specified, can result in uneven RCS temperatures among the loops and create conditions in
.
the Steam Generators leading to -- inadvertent i Safety Injections created by Steam Generator i different j pressure
- ( ' g s.
i Rev. 3 Salem Unit 2 . I
.-
l
-
' ECP-CFST-1 , 7M ~ FIGURE 1 SHUTDCWN MARGIN STATUS TREE
\ . 1 l
l START l
.
e
.
-3 OR MORE
, power Range Less Than S Y.
YES l MO e e e . IR SUR eeeeee e ZERO e OR e mEGArrvE e YESl MO . e e e Source + + + Range eee++e + + EMERGIZED e e e . YESl m0 e e e . e e e e e e e e source ******* * * * Range SUR e o e ZERO OR e e e MEGATIVE * * IR SUR e e i* YESl MO MORE e e MEGATIVE e e
* * THAM e e e e -e.2 DPM e e e e e e. vesl =0 + e.
. . * *
* * * * * * * * e e * e-e e e e e * * * * * * *
GREEM YELLOW , GREEN YELLOW PURPLE ERE7$ SAT FRSM-2 SAT FR5M-2 FRSM-1 FR5M-1
.
uhj RTri Uaud Salem Unit 2 4 Rev. O
.m e - --
ECP-CFST-1 FIGURE 2 CORE COOLING STATUS TREE b l START l
.
.
.
. 5 OR HORE CORE EMIT TC GREATER THAM 1289 DEGREES YE5 l MO
. .
.
.................&.................... .
.
.
.
. RCS e SU8 COOLING N M M *" O Gh*of ag * BY L.vCeOSCD
$
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. .............................. .
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. .
.
.
.
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.
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.
. YESl MO
. .
.
) . .................... .
. . .
. . .
. . .
. . .
. .
. . 5 OR ore *
. . . CORE EMIT TC
. . . GREATER THAM
. . 799 DEGREES
. . RoLzs
. . DYMAMIC RANGE YES l MO
. ER GREAT,or 5. x. .T H A M t . .
2 RC,
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2 x.
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oc 1 RCRC,
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. . . . .
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. . RoL1s RoL1s
. . FULL RANGE FULL RANGE
. . GREA1ER GREarER
. . . THAM 5 0 x. THAM 50x.
. . . .
. . . . . . . .
ves i O vEs i .O
. . . . . .
. . . .
. . .
. . . .
. . . . .
. . . .
" RED: GREEN YELLOH PURPL'E PURPLE }R ED3 YELLOW PURPLE FRCC-1 SAT FRCC-3 FRCC-2 FRCC-2 FRCC-1 FRCC-3 FRCC-2 MAQTED "
Salem Unit 2 5 Fev. O !
- pt ECP-CFST-1 l FZGUPE 3 HEAT SINE STATUS TREE l START l
. .
t
SG MR GREATER THAM 15 Y. IN AT LEAST ONE IMTACT SG YES l MO
. .
O 6 4 6 9 9 e e TOTAL FLON rw (Argja e CAPASILITY
*
*
TO IMTACT SGs GREATER THAM o f (a) m<
*
p,,,fa ,gp 22E94 lb/hr
* @ OsM A fsf YES l MO fgG , f r $ za e 4 ,
O e 4 4 4 4 9 9 4
ALL SGs e LESS THAM e 1125 PSIG e
+
YES l MO 4
.
4 4 4 e l
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**' e e ALL SG MR e e LEVELS LESS e e THAN 677. * *
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.
4 4 e e 444449444 4 * * O 4 e e e O O ALL SGs * * 4--- LESS THAM * * * 1970 PSIG * e *
* * * '
Ye5 l MO 4 .
.
.
.
4 4 4 4 4 4 e e 9 e 44444444494 e e e 4 9 e * * * 4 4 * e ALL SG e * * * MR LEVELS * 4 e o GREATER THAM 4 * e e 15X e * * *
* * 9 O ve5 l MO . .
e . 4 4 4 4 4 e e e 4 * * * 4 4 4 9 * # GREEN YELLOW YELLOW YELLOM YELLON $*. R E D SAT FRHS-5 FRHS-4 FRHS-3 FRHS-2 FRH5-1
!
Salem Unit 2 6 Rev. 0 , I ! i i
~ EOP-CFST-1 FIGURE 4 THERMAL SHOCK STATUS TREE
)
'
l5 TART l
.
.
pk RCS COOLDONM GREATER THAM 100 DEG IM f() LAST 60 Hin. f
, VES l MO l , . .
RCS PRE 55/ TEMP POINT TO
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.
.
.
THE RIGHT I OF LIMIT A i .
.
vE5 l wO ,
. . ALL RCS ....... . . .
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l g COLD LEG 5 GREATER THAM DEGREE 5 312 e I CSED EE25 vE5l MO TH M O I . . DEGREES vE5l wO
. . .
l .....
.
.
.
.......
.
.
. i .
. . . . RCS PRESS
+ . . . LESS THAM l . s7s PSIG
......... . .
.
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. . - .
.
.
.
vE5l =0
.. . . . . . I . .
l . . ALL RCS . ..... COLD tEG5 . ALL RCS
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: * t' 2hEAVEli' '"@ERe!5 .
i D$EREEU vE5l wo . . . . ' g g g l g vE5l NO g
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.
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. . . . .
. . . .
GREEN vELLOW FURPLE
~
6. RED-.'- l GREEN GREEN vELLOW PU[NE FRTS-1 FRTS-1 SAT SAT FRTS-2 FRTS-1 SAT FRTS-2 l i Sc2 (cc{ssov'l
&L cc.u n r '
cocL Deu-J #U~ A<-
'
TsJ & fMss f S W'S % Pu a . 5Mtss
,m_ mm
.
, MISTER .. e
*8!*
- EOP-CFST-1 FIGURE 4A THERMAL SHOCK LIMIT A CURVE
.
PTS PLANT OPERATIONAL UMITS CURVE s l .
* 3000 6'Y'
240 F 270 F f, 2540 pelg 2500-
~ , 215 F ' 2050 psig ' ,
2000-i
.g a
N 1500- UWIT A T1 T2 0 t E 1000-500-
-
o go go e0 e0 no do ao
, &
TE)APERATURE F salem Unit 2 8 so. 0 j
- - - - _ - - - _
n-. . I
'
ECF-CFST-1 FIGURE 5 CONTAINMENT ENVIRONMENT STATUS TREE
; '
l START l
.
O
+ .
CONTAIMMENT PRESSURE LESS THAM 47 PSIG
/[ % l MO
. .
O e CONTAIMMENT * e PRESS 44444 4 LESS THAM e 23.5 PSIG e e YES l MO + e . e e
* * *
e e e
* * * * * *
CONTAIMMENT e * * SUMP LESS 4e4e 4 e THAM 7 6 V. O e e t YES l MO + + e . . R-44 4 e e e e
*
RADIATION e e e e LESS THAM *****e O 4 e 35 R/ht' * 4
* * *
veS l No .
* 4 e e e e . * * e 6 * * * *
GREEN YELLOW kU R Pi.E PURPLE . REDS SAT FRCE-3 FRCE-2 FRCE-1 FRCE-1
.
M!, STER Salem Unit 2 9 Rev. O
-
- . '
6.10.2 Dsco lioits for fccolos of childbooring cgo: It is recommended by the National Council on Radiation Protection and Measurements and shall be the policy of the
["'/) \_ station that during the gestation period, the maximum dose equivalent to the fetus from occupational exposure of the '7,0T expectant mother shall expectant mother shall ,not exceed 0.5 rem (500 who millirem).
TI __ the RPE of her pregnancy.
notify her supervisor shall infors 6.10.3 Administrative dose limits:
- Administrative dose limits are established to ensure that regulatory and station limits are not exceeded during non-emergency conditions. The administrative limits shall be authorized by management and supervisory personnel ~1n defines blocks. Doses shall require higher management approval as as individual's accumulated exposure for the quarter approaches the regulatory limit.
The administrative dose limits that shall be used to control personnel radiation exposure during non-emergency conditions for personnel eighteen years of age and older are as follows NOTE If designated personnel are not available for approval, a Senior Shift Supervisor and the Shift Radiation Protection
Technician may authorize the exposure.
(v}100 mrem /Q +100 Current quarter exposure Technical Supervisor-RP with no of ficial documen-tation (may be increased after evaluation by
- Senior Supervisor-RP)
Automatic authorization 1000 mrem /Q +100 at the start of each quarter NRC Form 4 must be Senior Supervisor-RP and 2000 mrem /Q +100 complete and individual individual's Senior must be 19 years old. supervisor or equivalent 2500 mrem /Q No exposure extension Department head, 3000 mrem /Y is allowed for levels Radiation Protection Engineer and the General
,
4000 mrem /Y above 2500 mrem /Q for planned operations. Manager-Salem Operations
>5000 mrem /Y This level is considered Vice President-Nuclear an emergency exposure level and requires special authorization r ( ,) The Senior Supervisor-RP shall be responsible for establishing procedures to implement the administrative control limits.
AP-24 Paga 13 of 19 pages Rev. 7
- . . . _ -
.
. * "'],09cr.
.
[3sTRtyE*.~17!0N . 3/4.3.2 ENGINEERED SAFET/ FEATUPE ACTJAT'Tl SYSTE1 t!!STIU** EMIT!C LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Featurn Actuation System (E5FAS) instru-mentation channels and interlocks shown in Table 3.3-3 snall be OPERABLE with their trip setpoints set consistent with the values shewn in the Trio Set:oint eclumn of Table 3.3-4 and with RESFCRSE TIMES as shewn Table 3.3-5.
APPLICABILITY: As shown in Table 3.3-3.
ACTION: a. With an ESFAS instrumentation channel trip setpoint * os conserva-
. tive than, the value shcwn in the A11cwable Values dumn of Table 3.3 4. declare the channel inocerable and s -1v Lie applicable ACTION requirement of Table 3.3-3 until the channel is restartd to OPERA 8LE status with the trip setsoint adjusted consistent wita the Trip 5etsoint value.
, b. With an ESFAS instru. entation channel inocerable, take the ACTICM
- ,.
shewn in Table 3.3-3.
SURVEILLANCE RECUIRESENTS 4.3.2.1.1 Each E5FAS instru=entation channel snail te deconstratec OPERABLE by tne cerfor .ance of tne CHANNEL CHECX CHANNEL CALIBRAT*CN and CHANf:EL FUNCTICNAL TEST ocerations during tne :<CES and at ene frequencies snewn in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall te The demenstrated OPERA 8LE total interlock functton during tne automatic actuation logic test.
sna11 be demonstrated OPERA 8LE at least once ;er la menens during CHANNEL CALIBRATION testing of each channel affected by interlock cceration.
4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPCNSE TI5E of each ESFAS function shall be demonstrated to be within the limit at least once :er 13 mentas. Each test sna11 include at least one icgic train such esat both logic trains are tested at least once per 36 menens and one enannel per function sucn enat all enannels are tested at least once :er N times 18 mentns wnere N is tne total nuccer of redundant enannels in a s:ecific ESFAS functicn as snown in tne " Total No. of Channels" Column of Table 3.3-3.
SALE 1 - UNIT 2 3/4 3-14
- - - ._ _ _.
_ _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ - __ _ TABI 1-3 (C:ntinund) y
*
,
un-ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION f. i '4g HINIMUM g TOTAL NO. CilANNELS CtlANNELS APPLICABLE
FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE HODES ACTION C
$! Three Loops 1 Tavg/ operating l### Tavg in 1 Tavg in any 15 HI loop any operating two operating u loop loops OR, COINCIDENT WITH Steam Line Pressure- 1, 2, 3##
Low Four Loops I pressure / 1 pressure in 1 pressure in 14* Operating loop any 2 loops any 3 loops Three Loops I pressure / l### pressure I pressure in 15 4( , any 2 operating m. Operating operating loop in any operating u loop loops
U$ 5. TURBINE TRIP & FEEDWATER ISOLATION ,
-
e a. Steam Generator 3/ loop 2/ loop'in any 2/ loop in each 1,.2, 3 14* Water level-- operating loop operating loop liigh-liigh , 6. SAFECUARDS EQUIPMENT 3 2 3 1, 2, 3, 4 13 CONTROL SYSTEM (SEC) 7. UNDERVOLTAGE, VITAL BUS m
{'
5', s. Loss of Voltage 3 2 3 1, 2, 3 14* z P b. Sustained Degraded 3 2 3 1, 2, 3 14* Voltage w a
,
f
' .
e 8 g
_ _
- ?.oq 6 .' -
OCO243 3/4.8 ELECTRICAL POWER SYSTEMS . 3/4.8.1 A.C. SOURCES OprPATINy -
.
LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.Cl electrical power sources shall be OPERABLE: a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system (vital bus system), and b. Three separate and independent diesel generators with: 1. Separate day tanks containing a minimum volume of 130 gallons of fuel, and 2. A common fuel storage system consisting of two storage tanks, each containing a minimum volume of 20,000 gallons of fuel, and two fuel transfer pumps.* APPLICA8ILITY: MODES 1, 2, 3 and 4.
ACTION:
a. With either an offsite circuit or diesel generator of the above - required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.2 within one hour and at least once per 8 hours thereafter; restore at least two offsite circuits and three diesel generators to OPERA 8LE status within
/2 hours or be in at least HOT STANOBY within the next 6 hours and l
in COLD SHUTDOWN within the following 30 hours.
b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.2 within one hour and at least once per 8 hours thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN
^ One inoperaole fuel transfer pump is equivalent to one inoperable diesel generator.
' , l l SALEM - UNIT 2 3/4 8-1
-
I-4.24
*
y iY. LOC.
-' PART V PRESSURIZER HEATER FAILURE
'
2.0 INITIAL CONDITIONS 2.1 Overhead Annunciators 2.1.1 REACTOR COOLANT LOW PRESS (E-11)
.
2.1.2 PRESSURIZER LOW PRESS (E-19) 2.1.3 REACTOR COOLANT LOW PRESS HEATERS ON (E-27) 2.2 Reactor Coolant - high pressure alarm (Console) 2.2.1 High Pressure Deviation Alarm (Console) 3.0 IMMEDIATE ACTIONS 3.1 Automatic 3.1.1 As pressurizer pressure decreases, a reactor trip will occur at 1865 psig and a safety injection at 1765 psig. REFER to EI I-4.3 Reactor Trip or EI I-4.0 Safety Injection Initiation as appropriate.
)
. 3.2 Manual 3.2.1 ATTEMPT to energize the Control Group Heaters by depressing its ON pushbutton.
3.2.2 VERIFY all Pressurizer Heaters are in AUTO.
3.2.3 VERIFY Pressurizer Pressure Controller is in AUTO.
3.2.4 ATTEMPT to energize either or both Groups of Backup Heaters.
3.2.5 If pressure is high and pressurizer heaters are energized, ATTEMPT to de-energize the heaters.
4.0 SUBSEQUENT ACTIONS COMMENTS 4.1 Dispatch an operator to Local Tech Spec 3.4.4 Panel Elev 78 Penetration. Power operation VERIFY Local Breaker position, may continue up RESET if Tripped before swap to 72 hours, if to Emer. Sup. only one group of heaters has failed.
Salem Unit 1 10 Rev. 7
I-4.24
.\
4.2 If all pressurizer heaters The emergency are de-energized and cannot vital power supply be energized from their normal for pressurizer supply, PROCEED as follows: heaters are
-
designed to maintain pressure during a natural circulation cooldown. These
* heaters will not be sufficient to maintain pressure with Reactor Coolant Pumps running.
4.2.1 TAKE manual control of pressurizer level.
4.2.2 SLOWLY REDUCE reactor By maintaining or power IAW IOP-4 Power raising Operation and IOP-5 pressurizer level Hot Standby to Minimum during the power Load while attempting reduction, the to maintain pressure pressure decay
, ~
with pressurizer level. can be minimized.
4.2.3 When the reactor is This will stop shutdown and Tavg is Pressurizer Spray less than 541*F, STOP Valve bypass flow the No. 11 and 13 RCPs. which will reduce the rate of depressurization.
4.2.4 If the heaters cannot Tech Spec 3.4.4 be energized in 72 hours, l i PERFORM a normal cooldown IAW IOP-6 Hot Standby to Cold Shutdown.
a. Should it become necessary to initiate pressurizer spray to reduce pressure during the cooldown, START, No. 13 'RCP.
b. INITIATE Aux Spray if Delta T is less than
-
320*F via CVCS.
l Salem Unit 1 11 Rev. 7 .
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3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT . CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: H0 DES 1, 2, 3 and 4.
ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS . 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: a. At least once per 31 days by verifying that all penetrations * not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.1, and all equip, ment hatches are closed and sealed.
. b. By verifying that each containment air lock is OPERABLE per Speci fication 3.6.1.3.
- c. After each closing of a penetration subject to Type B testing, except containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at Pa (47 psig) and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and'C penetrations, the combined leakage rate is less'than or equal to 0.60 La.
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^Except vents, drains, test connections, etc. which are (1) one inch nominal pipe diameter or less, (2) located inside the containment, and (3) locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed at least once per 92 days.
SALEM - UNIT 2 3/4 6-1
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EVENT CLASSIFICATION SECTION 18 18. Operational Stat'us Change (con'd) Initiating Event / Notification / Emercency Action Level Recortine Condition 1. ECCS automatic initiation ATT 1 E. Event that results (or set point reduced or manual UE
*hould s have resulted) in initiation; OR ECCS discharge as a result of a valid signal 2. Lit logic lights for initia-tion on RP4; AND (manual or automatic) 3. Discharge to RCS is called for or verified by control console indication (flow, '
valve positions, tank levels, etc.)
1. As stated, except resulting ATT 29 F. Event that results in from and part of a pisnned (50.723 - e manual or automatic test actuation of Engineered Saf ety Features (EST), Examples: including the Reactor a) SEC Mode Operation Protection System (RPS) b) Rx trip c) SI (without water into RCS) i d) Contiinment isolat ion 1. Any pla'nned shutdown ATT 32 J. Scheduled shutdcwn for Other testing, maintenance 2 refueling 1. Upon official notification ATT 32 H. Derating' caused by fecm site management or NRC Otner
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regulatory action I. Any reactivity anomaly 1. Disagreement with predicted ATT 13 steady state reactivity Otner balance during power ops 1 1000 pcm; OR 2. Ca lculated SCM < required: OR 3. SUR 1 5.2 dpm; OR .
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4. Unplanned positive reactivity insertion > 500 pcm
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L. As stated ATT 12 J. Any unplanned criticality - UE i 31.utdown initiated ATT 1 K. Any event or condition 1 UE requiring unit shutdown i to comply with any T/S LCO
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ECG
' ATTACHMENT 29 ATT 29 Each step shall be initialed by the responsible individual when completed. .
I. Notifications 1.. Complete Section I of NRC Data Sheet (page 2 of this SSS, a t t ac hme n t ) . 2. Notify Operations Manager and confirm classification of SSS event.
L. Fry (Office - 4523; Car - (6091342-5103: Beeper - (8001612-4532; Home - (6091678-7634) or L. Catalfomo (Office - 4522: Beeper - (8001612-4534: 08. =@[> Home - (6091678-3176: Car - (302) 428-9084) fgg,y ggc IS 4Md/ notified hrs on time .date 3. If this attachment is being utilized as a result of a SSS reactor trip notify the General Manager - Salem Operations of the event.
J. Zupko (Office - 4500: Ca r - (6091342-5036: Beeper - [6091342-5803: Home - (6091468-5527)
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_gz7 L. Miller (Of fice - 4497: Car - (6091342-5077: , 69kto# Beeper - (8001612-4531; Home - (6091769-1727) 4. Notify LAC Dispatcher of event (direct line, or (6091 - SSS 935-7300).
notified at hrs on time date name 5. Notify Public Af f airs Manager - Nuclear or Alternate SSS with details of event: Public Affairs Manager - Nuclear (Contact One)
- - P. Silverio Office: 4699 Home: 829-1546 Beeper: 342-5804-W. Denman Office: 4480 Home: 935-6349 Deeper: 342-5849 - B. Go rma n Office: 4480 Home: 228-1089 Beeper: 342-5851 notified hrs on .
time date
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ECG ATT 29 6. Notify NRC Operations Center (ENS line, [202]951-0550, SSS or (301]427-4056), [301]427-4259, or (3011492-8893 of the event within 4 hours. Use NRC Data Sheet to record additional information provided to the NRC.
' notified at hrs on name time date 7. If this attachment is being utilized as a result of a SSS reactor trip notify the NRC Resident Inspector within 4 hours.
J. Linville - (Office - 4479; Home - (6091243-4998) R. Summe rs - ( Of fice * 4479; Home - [609l848-6741) .
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2 of 5 Rev. 1 SGS I ___
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ECG ATT 29 II. Reporting hC" 1. Ensure that an Incident Report is prepared in accordance
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SSS with AP-6.
2. Forward this attachment, along with the Incident Report sss and any supporting documentation, to the Senior Operations Technical Supervisor.
. 3. Review Incident Report and any other available relevant sots information for correct classification of event and corrective action taken.
4. Contact the LER Coordinator and request written followup SOTS ( required 30 days after event ). Provide this attach-ment and any other supporting documentation received from the SSS.
~ 5. Prepare Licensee Event Report.
LER LER number 6. Follow-up actions response requested from LER Response' request number 7. Prepare Licensee Event Report.
' LER Number 8. Return this attachment to the Operations Manager.
LER I
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SGS- . 3 of 5 Rev. 1
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W ADMINISTRATIVE PROCEDURES SALEM GENERATING STATION
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ADMINISTRATIVE PROCEDURE NO. 13 () TEMPORARY JUMPERS, PUMPS, AND LIFTED LEADS 1.0 PURPOSE . This procedure describes the program for controlling the use of temporary electrical or mechanical jumpers, pumps and the lifting of leads.
2.0 SCOPE This procedure addresses temporary electrical / mechanical jumpers and lifted leads that remain in any system upon completion of a procedure or active troubleshooting. These jumpers / lifted leads are not to be confuse, with jumpers / lifted leads utilized while performing: active troubleshooting, calibration, testing, maintenance, or an approved procedure which requires installation and subsequent removal of the temporary jumpers / lifted leads.
This procedure also addresses all temporary pumps which are installed in the controlled access area or installed outside the controlled access are to carry radioactive materials that are intended to be removed from the system and not be part of an approved design change. The potential of a temporary pump to transfer hazardous materials which are present due to a spill or other such accident shall be addressed in the review of all
/"') temporary pump requests, regardless of whether radioactive materials (m / are to be transported by the pumps or not.
3.0 REFERENCES 3.1 NRC Clarification of 10CFR50.59 Requirements for Jumpers / Lifted Leads and Procedures (AITS: F01004908) 3.2 .INPO Good Practice OP-202, Temporary Bypass, Jumper, and Lifted Lead Control 3.3 AP-8, Design Change, Test and Experiment Program 3.4 AP-9, Maintenance Program i 15 AP-11, Record Retention Program 4.0 DEFINITIONS 4.1 Hechanical Jumper: A piece of piping, hose, spoolpiece, blank or blind flange, or tubing which join two or more systems together or j bypasses a component (s) within a system, thus altering the systems design or configuration.
4.2 Electrical Jumper: A wiring connection used in a component or circuit which bypasses a component within an electrical O circuit, thus modifying the circuit design or configuration.
Page 1 of 5 Pages REV. 7 AP-13
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- 5.1.3 Indicatoc twitchtblo equipment within the work area which is properly isolated, but which is unsaf e to work on because of inadequate clearance to other energized C) . equipment (see table 1) .
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TABLE 1
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MINIMUM CLEARANCE TO ENERGIIED EQUIPMENT phase-to phase ' voltage (KV) minimum clearance 4-13 H + 2'0"
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26 H + 2'6" 69 H+3' 138 H +'3'4" 230 H+5' 345 H+7'
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500 H + 11' Where "H" is equal to the body height with arms fully extended.
NOTE Rubber gloves and sleeves shall be required to install barriers within 6 feet of energized equipment.
5.1.4 Identifies the person (s) for whom the equipment has been cleared and tagged.
5.2 Yellow Permissive Tag - serves the following purposes: 5.2.1 Distinctly marks electrical equipment that is safe for work and is used only in conjunction with red blocking tags on high voltage electrical equipment.
5.2.'2 Identifies the pehson for whom the equipment has been cleared and tagged.
5.2.3 Gives the person named on the tag the authority to work on, operate, or adjust the equipment.
5.2.4 Prohibits operation of the equipment by any other person unless permission is granted by the person named on the yellow tag.
, 5.3 White caution Tag - serves the following purposes: 5.3.1 Indicates, as part of a switching procedure, a position other than the normal operating position of electrical equipment.
5.3.2 Indicates an abnormal condition or limitation of the equipment. The tag shall identify the abnormal conoition and may include special instructions of a
,) temporary nature. These instructions aball be complied with under all conditions and circumstances.
AP-15 page 2 of 24 pages Rev. 3
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~ . 5.3.3 The white caution tag shall not be used on blocking points to isolate equipment f rom energy sources for (~') the protection of personnel or equipment.
v 5.4 Workers Blocking Tag - serves the following purposes: 5.4.1 Identifies the mech'anical or low voltage electrical blocking points between any circuit or equipment that
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is energized and the de-energized equipment upon which work is to be perf ormed.
. 5.4.2 Blocks and prohibits the operation of equipment by any individual other than the worker named on the
- tag or persons directed by him/her.
5.4.3 Indicates that the equipment is tagged at the request of the person in charge (named on the tag) and that the worker (named on the tag) may operate the equipment.
5.4.4 This tag cannot be applied to a component which has any other safety tag.
5.4.5 This tag is authorized for use only by Electric Transmission and D.istribution personnel as set forth in Section 8.19.
5.4.6 This tag is red with a yellow stripe. .
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6.0 DESCRIPTION AND USE OF TAGGING REQUESTS 6.1 All forms associated with the safety tagging system are contained in Operations Directive-8, " TRIS Tagging Operations". The following is a list and brief description
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of each form: , 6.1.1 Tagging Request Form (OD-8-A-1) - This form shall be used whenever red blocking tags or yellow permissive tags are needed for the protection of personnel or equipment.
6.1.2 Addendum to Tagging Request (OD-8-A-2) - This form shall be used in conjunction with the Tagging Request Form when specific blocking points are requested to be tagged.
6.1.3 Group Tagging Request Authorization Sheet (OD-8-A-3) This form shall be used only when a tagging request is made under group tagging rules.
p V AP-15 page 3 of 24 pages Rev. 3
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TABLE 2.2-1 (Continueit) . g- i r-9 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ' NOTATION (Continued) %'g h 2 \p i Note 2: Overpower AT 1 AT,[K4-K5 3 1+t3S T-K6 (T-T")-f2(AI)] wh'ere: AT, = IndicatedATatRATEDTilERMALPOWER
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T- = Average temperature, *F
= l T" Reference T,yg at RATED TilERMAL POWER 1577.9*F K = 1.080 .
K = 0.02/*F for increasing average temperature and 0 for decreasing S 7 average temperature ' e K = 0.00119/*F for T > T"; K6 = 0 for T 1 T"
138 jy 3
= The function generated by the rate lag controller for T,yg dynamic 3 compensation
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= Time constant utilized in the rate lag controller for T,yg r 3= 10 secs. ,
-1 l S = Laplace transform operator, Sec .
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f p(AI) = 0 for all Al Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than h
4 percent.
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.) SALEM GENERATING STATION OPERATIONS DEPARTMENT DOCUMENT APPROVAL COVER SHEET - ..
Title: REMOVING, RETUR?ING TO SERVICE AND LOSS OF PROTECTIVE SYSTEM CHANNEL
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No.: IV-10.3.1 Unit: 1 Rev.: 3
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Remarks: 3 pages of text, 10 tables, last OTSC - P-1 Revised to incorporate outout function - Axial riux l difference.
! i i Safety Related Review (Re f . AD-13) : S/R yes X no Author's Checklist Completed: yes X Author e Dated W i' I VSRO Mk _ Date 7-12-TY Ops. Eng. < Date 7-/7*7I
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SOTS + /4 efftTN _ Date Y Sps. Mgr. - W Date [W
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kd QA & L M Date 7 3o M SORC w - k NJ fh Date h l~Y Y General Manager * //77L h V / Ld Date
~( , * required for SPM documents only i + required for EOP validation acceptance only i ** required for safety related documents and fire \
protection documents Salem Unit 1/2 AD-1-B-1
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OPERATING PROCEDURE
.) IV-10.3.1 REMOVING, RETURNING TO SERVICE AND LOSS OF PROTECTIVE SYSTEM CHANNEL - ..
1.0 PURPOSE . 1.1 This procedure provides the instructions necessary for the following:
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1.1.1 Operation of plant with Loss of Protective System
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Channel.
1.1.2 Removing and Feturning to Service a Pfe. . ion System Channel.
2.0 INITIAL CONDITIONS 2.1 The failed or out of service Protective Channel is required for plant operation per Technical Specification for the current mode of plant operation.
3.0 PRECAUTIONS 3.1 A failed or out of service Protective Channel will require complying with an " Action Statement" per Technical Specifications.
3.2 When maintenance or testing is performed on a Protective Channel, verify that no other channel in the logic is energized which together with the subject channel being energized will cause the Logic Train to initiate an inadvertant automatic protection system actuation.
4.0 ATTACHMENTS LIST 4.1 Tables 4.1.1 Table 1 - Nuclear Instrumentation-Power Range 4.1.2 Table 2 - Reactor Coolant System-RCP Flow 4.1.3 Table 3 - Reactor Coolant System-Loop Temperature 4.1.4 Table 4 - Reactor Coolant System-Pressurizer 4.1.5 Table 5 - Steam Generator-No. 11
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4.1.6 Table 6 - Steam Generator-No. 12 4.1.7 Table 7 - Steam Generator-No. 13 Q Salem Unit 1 1 Rev. 3 _
_- IV-10.3.1
TABLE 1
] NUCLEAR INSTRUMENTATION POWER PANGE
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INOP BISTABLE PROTECTION OUTPUT DEV SWITCH RACK NO. FURCTION(s) REMARKS N-41 IBS-411C 2 OT Delta T, 2/4 Reactor Trip 1BS-411D 2 OT Delta T, 2/4 Auto. Turb. Runback
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and Block Auto and Manual Rod Withdrawal Place the NI Racks Switches: 1. Remove following No. 81 1. Upper Section Control switches Deviation Channel Power in the Defeat Fuses indicated 2. Lower Section position: Deviation Channel After De feat switches 3. Rod Stop Bypass on NI 4. Power Mismatch Rack 81 Bypass reposi-5. Comparator Channel tioned.
De feat Control 6. Axial Flux Test Pack 26 difference Position N-42 1BS-421C 6 OT Delta T, 2/4 Reactor Trip 1BS-421D 6 OT Delta T, 2/4 Auto.
Turb. Runback and Block Auto and Manual Rod Withdrawal Place the NI Rack Switches: 1. Remove following No. 81 1. Upper Section Control switches Deviation Channel Power in the Defeat Fuses indicated 2. Lower Section position: Deviation Channel After De feat switches 3. Rod Stop Bypass on NI 4. Power Mismatch Rack 81 Bypass reposi-5. Comparator Channel tioned.
Defeat
. Control 6. Axial Flux Test Rack 26 Difference Position i
Salem Unit 1 TBL l-1 Fev. 3
IV-10.3.1 '.
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INOP BISTABLE PROTECTION OUTPUT pEV SWITCH RACK NO. FUNCTION (s) REHARKS N-43 1BS-431C 13 OT Delta T, 2/4 Reactor Trip 1BS-431D 13 OT Delta T, 2/4 Auto. Turb. Runback and Block Auto and
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Manual Boi Withdrawal Place the NI Racks Switches: 1. Remove following No. 81 1. Upper Section Control switches Deviation Channel Power in the De feat Fuses indicated 2. Lower Section position: Deviation Channel After Defeat switches 3. Rod Stop Bypass on NI 4. Power Mismatch Rack 81 Bypass reposi-5. Comparator Channel tiened.
De feat Control 6. Axial Flux Test Rack 26 Difference Position i N-44 1BS-441C 15 OT Delta T, 2/4 Reactor' Trip 1BS-441D 15 OT Delta T, 2/4 Auto.
Turb. Runback and Block Auto and Manual Rod Withdrawal Place the NI Rack Switches: 1. Remove following No. 81 1. Upper Section Control switches Deviation Channel Power in the Defeat Fuses indicated 2. Lower Section position: Deviation Channel After De feat switches 3. Rod Stop Bypass on NI 4. Power Mismatch Rack 81 Bypass reposi-5. Comparator Channel tiened.
Defeat
. Control 6. Axial Flux Test Rack 26 Difference Position s. .
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Salem Unit 1 TBL 1-2 Rev. 3
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OEFINITIONS
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TABLE 1.1 OPERATIONAL MODES REACTIVITY i AVER' AGE COOLANT CONDITION, K,rf THERMAL POWER TEMPERAR!RE N00E '
> 5% 1 350*F 1. POWER OPERATION 1 0.99 2. STARTUP > 0.99 < 5% > 350*F 3. HOT STAN08Y < 0.99 0 1 350*F < 0.99 0 3 >
4. HOT SHUTDOWN
*F V
5. COLD SHUTDOWN < 0.99 0 1 200*F 6. 0 1 140*F REFUELING" 1 0.95
- Excluctng cecay heat.
" Fuel in the reactor vessel with the vessel head closure bolts less,.than fully.
tensioned or with the head removed.
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SALEM - UNIT 2 1-8 Amendment No. 28
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~ REACTOR COOLANT SYSTEM ,.
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OPERATIONAL LEAKAGE
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LIMITING CON 0! TION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:
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a. No PRESSURE BOUNDARY LEAKAGE, b. 1 GPM UNIDENTIFIED LEAKAGE, c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, V d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2230 * 20 psig.
f. 1 GPM leakage at a Reactor Coolant System pressure of 22.30 + 20 psig from any Reactor Coolant System Pressure Isolation Valve specified f in Table 3.4-1.
APPLICA8ILITY: MODES 1, 2, 3 and 4.
ACTION: -
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a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STAN08Y within 6 hours and in COLD SHUTDOWN within the following 30 hours.
b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
c. With any Reactor Coolant System Pressure Isolation Valve leakage greater .than the above Ifmit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by i use of at least two closed manual or deactivated automatic valves, or be in,at least HOT STANDBY within the next 6 hours and in COLD , SHUTDOWN within the following 30 hours, i SURVEILLANCE REQUIREMENTS _ 4.4.7.2 Reactor Coolant System lea'kages shall be demonstrated to be within I each of the above limits by; ' a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours. - b. Monitoring the containment sump inventory at least once per 12 hours.
SALEM - UNIT 2 3/4 4-17 *
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. EVENT CLASSIFICATION SECTION 1
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1. Primary Leakage Notification / Initiating Event / Pecorti.; Emercency Action Level Condition 1. Exceeding action statements ATT 1 A. Primary leak ut a) no PRESSURE BOUNDARY of T/S LCO 3.4.6.2 (Unit 1)
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leakage or 3.4.7.2 (Unit 2) b) I gpm UNIDENTIFIED See Tech Spec leakage c) 10 gpm IDENTIFIED Action Requirements leakage d) 40 gpm CONTROLLED leakage e) RCS PRESS ISOL VALVES (See T/S) 1. One chg pmp cannot maintain ATT 2 8. Primary leak > 50 A gpm level 1. Low / decreasing P2R level with ATT 3 C. Known LOCA greater than maximum charging flow SAE total makeup capacity
* **Re f er to Section 5 prior to classif ication* * * !
1. P2R press >2200 psig & POPS ATT 1 D. P2R safety /PORV fail- not armed, or P2R press <37! UE ure to reseat psig & POPS armed; AND 2. PORV/ safety valve tailpipe tt temp, or PRT temp, press, or level increasing 1. Cracks in weld areas of safe ty ATT 6 E. Pipe cracks in stagnant (50.720-1H: borated water systems related piping (as reported by Engineering or ISI/.MIET) . IE Inf orma tion Nctice - one hou r
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'e EVENT CLASSIFICATION SECTION 6 6. Radiological Releases .
Note: Action levels listed are for valid RMS channel indications.
The validity of the indication should be confirmed by sample analysis or other means as necessary. Notification / Emergency Action Level Reporting Initiating Event / ,
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Condition 1. R41C > 1.5E3 cpm for > ATT 29 A. Accidental, unplanned, (50.72b - 4Hr) 60 min; or uncontrolled OR
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gaseous release, that exceeds 2 times the 2. R418 > 5.7E3 cpm increase applicable concentrations in 60 min of the limits specified in Appendix B, Table II of Part 20, to un-restricted areas (averaged over 60 min) 1. Any unplanned or uncontrolled ATT 29 8. Accidental, unplanned, '50.72b - 4Hr or uncontrolled liquid release outside the liquid release controlled access area. . 1. R18 alarm and no isolation ATT 7
? C. Liquid release that AND UE exceeds T/S limits Con f i rmed a na ly s is o f itqui2 f or > 15 min waste effluent indicating discharge exceeding T/S l im i t:; .
2. Any R19 ala rm and blowdown to - 12(22) S/G blowdown tank ATT 7 D. Gaseous release that 1. R41C > 1.lES for > 60 min; UE OR exceeds T/S limits 2. R418 > 2.lE4 cpm increas in
.for > 60 min 60 min:
R41C > 1.lE6 cpm for > 15 mtn; ATT 8 E. Gaseous release that 1. A OR exceeds 10 times T/S RTIB > 5.3E4 cpm increase in limits for > 15 min 2.
15 min; OR 3. R45B > 7.1E-3 uCi/cc f or > 15 min Re v. 0 1 of 3 SGS ,
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EVENT CLASSIFICATION SECTION 18 18. Operational Status Changes
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Notification / Initiating Event / Reporting Condition Emergency Action Level 1. As judged by the SSS/EDO ATT 2 A. Any event or. condition A during operation that
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tion of the plant, including principal safety barriers, being seriously degraded Examples: ATT 6 B. Any event or condition (50.72b - during operation that a) accumulation of voids that . results in the plant could innibit the ability being: to adequately remove heat 1) in an unanalyzed from the reactor core b) voiding in instrument lines, condition that resulting in erroneous significantly compromises plant
~ indication, causing the safety; operator to misunderstand 2) in a condition out- the true condition of the side the design basis; plant ' 3) in a condition not covered by' operating or emergency procedures 1) As judged by SSS/EDO ATT 29 C '. Any event, found while (50.72b -
s hu tdow n , that had it been found during operation would have resulted in the plant, including principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly com-promises plant safety 1. Action required because no ATT 6 D. Any deviation from.T/S action consistent with (50.72b - or licens.e condition license and ./S can p rovide in an emergency when adequate or equivalent
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action is needed to protection protect the public health..and safety (ac-tion must be approved at least by a licensed SRO)
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1 of 3 Rev. O SGS
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-- ~ krMtl? men $ f ' ATTACHMENT 4 NRC Resolution of Facility Comments on SRO Exam-Question ~ Resolution 5.01.a Subjective comment - considered in grading 5.01.b Included 5.02 Included-5.04 Included 5.05.a Included 5.05.b Added for information 5.06.a Included 5.06.b Not changed 5.07 Included 5.08.a Included 5.09.a Subjective comment - considered in grading 5.09.b and c . Included 6.01.a Included 6.01.b Not changed 6.02 Not changed 6.03 Not changed "' 6.04.a Not changed 6.04.b Added for information 6.05.a Corrected 6.05.c Included 6.06.a Added fo~r information
.6,06.b Added " Service Water for MS-10" ~6.07 Inc'luded 6.08.a Included 6.09 Included 6.10.b Included 6.10.d(3)- Included 7.03 Deleted question 7.04 Included-7.05 Included 7.08 Included 7.09 Included ^7.10.c No_ changed 7.10e ~ Included '
8.01 Included 8.02 Accepted "STA, NRC, and Local"
'8.04' Not changed 8.07.a and b Not changed 8.07.c Included 8.07.d Included first part only 8.07.e' Changed answer to "None" 8.08.a and c Included 8.08.b Not Changed.
8.10.a Not. changed 8.11.b Included J }}