IR 05000293/1986015
| ML20206N143 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/17/1986 |
| From: | Crescenzo F, Howe A, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206N138 | List: |
| References | |
| 50-293-86-15-OL, NUDOCS 8608260104 | |
| Download: ML20206N143 (145) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT COMBINED EXAMINATION AND INSPECTION REPORT NO. 86-15 0/L FACILITY DOCKET NO.
50-293 FACILITY LICENSE NO. DPR-35 LICENSEE: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 FACILITY: Pilgrim Nuclear Power Station EXAMINATION DATES: May 5, 1986 to May 8, 1986 CHIEF EXAMINER:
b.58.,b 7-/7-86 A. Howe - Reactor Engineer (Examiner)
Date r
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$6 REVIEWED BY:
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F.CresQz,Re r Engineer (Examiner)
Date REVIEWED BY:
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Robert M. Keller, Chief, Projects D~ ate Sect on 1 APPROVED BY:
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Harri B. Kfh er, Chief, Projects
/Dath/
Branch No. 1
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SUMMARY: Operator License examinations were conducted at Pilgrim Nuclear Power Station during the week of May 5,1986.
Examinations were administered to five reactor operator candidates and two senior reactor operator candidates. All candidates passed the examinations with the exception of one reactor operator candidate who failed the written portion of the examination.
An inspection of the Operator License Replacement and Requalification training program was also conducted.
86o8260104 e60814 PDR ADOCK 0500
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REPORT DETAILS TYPE OF EXAMS: Replacement-EXAM RESULTS:
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CHIEF EXAMINER AT SITE: Allen Howe (USNRC Region I)
OTHER EXAMINERS: Gordon Robinson (USNRC Consultant)
Jeff Sherman (USNRC Contractor)
Frank Crescenzo (USNRC Region I)
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Summary of generic strengths or deficiencies noted on oral exams:
Strengths:
Candidates were familiar with content, use, and location of procedures in the control room. Also, candidates were familiar with plant equipment and its location.
Weaknesses:
1.
Candidates demonstrated an unfamiliarity with local RCIC operation in that they failed to energize the Topaz inverter as a part of the startup procedure.
2.
During discussions of plant transients, the reactor operator candidates failed to recognize entry condi-tions for the E0P's.
2.
Summary of generic strengths or deficiencies noted from grading of written exams:
RO - Strengths:
Candidates demonstrated good knowledge of plant design, emergency systems, and safety systems.
Also, candidates demonstrated an overall knowledge of instruments and controls.
RO - Weaknesses:
Candidates had difficulty applying thermodynamic principles to various problems. Although examples could be given, the concept of the definition of pump runout was poor. Also, candidates were weak in the causes of rod blocks, especially in the REFUEL mode.
3.
General comments on the written examination process.
a.
On the R0 examination, question 4.02 was deleted and replaced with a new question while the examination was in progress. The change was needed because the procedure that the question was based on was revised and the candidates were trained on the new revision. The old revision was used to write the question because the facility erroneously sent the old revision to the preparer of the R0 examina-tion. This error was determined after the examiners had arrived on site and the decision was made at that time to replace the question, b.
During the written examination review process it was determined that
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several changes to the facility training material were needed. The facility agreed to incorporate the changes prior to the next sche-duled license examination.
4.
Summary of Facility commitments made to the NRC As a part of the exit meeting, the facility made a commitment to review four (4) procedures to determine the need for changes and to make changes as appropriate. As a part of this commitment the facility was allowed two
(2) weeks to determine the scope of the project and notify the NRC of pro-jected completion dates. At the time of writing this report, the facil-ity has provided to the NRC projected completion dates for only two (2) of the four (4) procedures.
This is an incomplete commitment which will be followed up by the Resident Inspector.
Details on the two (2) procedures which the facility committed tb review are located in attachment 4.
5.
Personnel Present at Exit Interview:
NRC Persor.nel Allen Howe, USNRC Region I Frank Crescenzo, USNRC Region I Martin McBride, USNRC Senior Resident Inspector NRC Contractor Personnel Jeff Sherman (EG&G)
Facility Personnel A. Pedersen, Nuclear Operations Manager L. Beckwater, Nuclear Operations Department, Compliance E. Ziemianski, Nuclear Management Services Manager S. Hudson, Operations Section Manager R. Cook, BECO Training H. Balfour, BECO Training L. Oxsen, Vice President Nuclear Operations T. Sullivan, BECO Training P. Mastrangelo, Chief Operations Engineer D. Sanford, Training Manager 6.
Summary of NRC comments on replacement examinations made at exit interview:
The plant was found to be well maintained and clean. The operations staff, especially those in the control room were cooperative. Access onto the site was efficient.
The training staff was well organized and was well prepared for the exam review.
Generic strengths and weaknesses noted during the operating exams were discussed.
The facility made a commitment to review four (4) procedures as noted in a previous section of this repor The schedule for the upcoming NRC inspection of the requalification pro-gram was also discussed.
7.
Results of Training Program Inspection An inspection of the Pilgrim Licensed Operator Training Program was con-ducted in conjunction with the examination period. Both the replacement and requalification training programs were inspected although the level of inspection was greater for the requalification program due to the unavail-ability of replacement training program participants.
The requalification program inspection consisted of facility record file review along with interviews of currently licensed individuals.
No attempt was made to evaluate actual classroom instruction other than by interviews with participants.
The results of the records inspection indicate that the facility's system for ensuring compliance with documentation requirements of 10 CFR 55 App A is adequate. Two previous year examinations for licensed personnel were found to be missing from the files, however, when brought to the facility's attention, they were located promptly. A review of lecture attendance records indicated that overall attendance was good.
It was noted that two individuals were delinquent in their attendance at requalification lectures. The Pilgrim training manual allows for excused absence from requalification lectures if proper justification is made by the plant manager based on previous year results, experience, etc. No such docu-mentation existed for these two individuals. This was discussed with the training department management and the following were offered:
a)
The training department was aware of the problem, b)
One of the individuals had intended not to renew his license.
c)
The other individual had been involved in other work and was now receiving specialized on-shift requalification training.
It was suggested to the facility that a more rigid procedure for documen-tation/ justification of requalification absence be developed.
The results of the interviews revealed no significant requalification training program deficiencies.
Several of the operators expressed con-cerns that the program was not flexible enough to allow training in the areas of individual deficiencies. This was discussed with training department personnel and it appears that an effective program based on the operator's individual needs/ desires does in fact exis..
J The replacement training program inspection consisted of discussions with training personnel who will be involved in the planning and execution of the two upcoming reactor operator classes.
During interviews with the licensed staff, it was noted that several of the licensed reactor opera-tors are currently, or will be, attempting to leave the licensed position for various reasons.
This attrition is of concern because of the loss of valuable experience prior to licensing new replacement operators. Addi-tionally, the training department currently maintains a " Milestone" policy whereby license candidates are permanently dropped from the program if certain examination grades are below 80%.
Because of these factors and potential failures during the NRC exams it is probable that licensed staffing problems may persist for some time. As noted in a previous Inspection Report (50-293/86-06), the licensed staff is lacking sufficient numbers of reactor operators.
The facility is fully aware of this problem and is pursuing a realistic schedule for licensing 20 additional reactor operators by January 1988.
i Summary: The requalification training program meets the minimum require-ments of 10 CFR 55. The replacement training program appears to be evolving into a more aggressiva and effective program for providing needed licenses. However, pitfalls due to attrition or failure, could cause the program to fall short of its goal of providing a substantial increase in the number of licensed reactor oparators.
Attachments:
1.
Written Examination (s) and Answer Key (s) (SR0/RO)
2.
Facility Comments on Written Examinations made after Exam Review
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NRC Resolution of comments on written examinations 4.
Details on Procedures Facility committed to review
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U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION
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Facility:
Pilgrim Reactor Type:
BWR Date Administered:
May 1986 Examiner:
Gordon E.
Robinson Candidate:
Nf A S TE _2_
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
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% of Category
% of Candidate's Category Value Total Score Value Category
25 1.
Principles of Nuclear Power Plant Operation, Thermo-dynamics, Heat Transfer and Fluid Flow
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Plant Design Including Safety and Emergency Systems
25 3.
Instruments and Controls
25 4.
Procedures - Normal, Abnormal, Emergency, and Radiological Control
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100 TOTALS Final Grade
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All work done on this examination is my own.
I have neither given nor received aid.
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Candidate's signature
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW (25)
Question 1.01 (2.0)
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You have been informed that the Shutdown Margin (S'DM) for your reactor is 4%
You have an Indication of 200 cps on the SRM Instrumentation.
Control rods are withdrawn and your countrate on your SRM's increases to 1000 cps.
What is your new Shutdown Margin?
(2.0)
Question 1.02 (2.0)
Steam quality and void fraction are both terms used to express relative amounts of a liquid or a vapor in a closed space.
a)
Briefly explain the difference between steam quality and void fraction.
(1.0)
b) As the steam quality increases, does the resistance to flow increase or decrease? Briefly explain your answer.
(1.0)
Question 1.03 (1.5)
Will the Doppler Coef ficient become more negative, less negative, or remain about the same for a decrease in void fraction from 60% to 40%7 Briefly explain your answer.
(1.5)
Question 1.04 (2.5)
Indicate whether the following statements concerning fission l
product poison behavior are TRUE or FALSE for your reactor.
IF, FALSE, change the statement so that it is correct.
NOTE:
If any part of the statement is not true, mark the statement FALSE.
a)
Equilibrium xenon concentration is directly proportional to power level.
(0. 5)
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' Equilibrium samarium concentration is the same for all b)
power levels.
(0.5)
c)
Both xenon and samarium concentrations increase immediately af ter a reactor shutdown from high powers.
(assume equilibrium had been reached before shutdown)
(0.5)
d)
There is an increase in the equilibrium xenon concentration as the core ages because power production from plutonium-239 increases.
(0.5)
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e)
Xenon-135 decays wi th a hal f-li fe of about 9 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while samarium-149 is stable.
(0. 5)
Question 1.05 (2.0)
The subcooling of condensate exiting the condenser is termed condensate depression.
- a)
Briefly explain why condensate depression is necessary?
(1.0)
b)
Briefly explain why excessive condensate depression should be avoided.
(1.0)
CATEGORY CONTINUED ON NEXT PAGE
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o Question 1.06 (3.0)
Your reactor is operating at full power and a feedwater controller malfunction results l'n a loss of feedwater flow.
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A reactor scram will occur due to low reactor water level.
Prior to the scram is the reactor power expected to increase, decrease, or remain constant? Give two reasons for your
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answer.
(3.0)
Question 1.07 (2.5)
Control rod shadowing can increase, decrease, either increase or decrease, or has no effect on rod worth.
Select one of the above.
Briefly Justify your answer.
(2.5)
Question 1.08 (3.0)
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Match one item from Column B and one item from Column C with each limiting parameter in Column A.
Column A Column B Column C Limi ting Parameter item Measured Limiting Condition a)
LHGR 1)
To tal f uel i)
1% plastic strain bundle power on cladding (1,0)
b) APLHGR 2)
Local fuel pin II)
boiling transition power in node (1.0)
c)
CPR 3)
average fuel lii)
clad temperature pin power in of 2200*F (1.0)
node Question 1.09 (2.5)
Consider a centrifugal pump a) At constant speed a reduction in total head will cause an increase, decrease or no change in capacity.
(0.5)
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What is the relationship between mass flow rate at the inlet and outlet of the pump?
(0.5)
c)
Doubling the speed of the pump will increase the horsepower by a factor of two, four, se s i x,7 o< e tt*/
(0.5)
d)
Briefly define pump runout and Indicate how i t can occur.
(1.0)
Question 1.10 (1.0)
Your reactor has just reached criticality af ter a refueling outage and is placed on a hundred (100) second period.
Thereaf ter no rod movenent or recirculation flow changes occur. Af ter some time has elapsed, you find that the power increases by a factor of 10 in 300 sec9nds.
Have you reached the heating range? Show all work.
(1.0)
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Question 1.11 (3.0)
Will the following cause the reactivity of your reactor to i
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Increase or decrease? Briefly explain each answer.
a)
Loss of a feedwater heater (1.0)
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Sudden increase in reactor pressure (prior to a resctor scram)
(1.0)
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Build-up of corrosion products (CRUD) on the fuel pins.
(1,0)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25)
' Question 2.01 (3 0)
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Consider the High Pressure Coolant injection System (HPCI)
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a)
If a HPCI leak occurs, the primary containment isolation and control (PCIS) system will isolate the system from
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the containment and trip the turbine.
List three different measured parameters that would cause the above actions to (1.5)
occur.
b) What is the purpose of the nitrogen purge line to the HPCI turbine exhaust 1 (1.0)
c)
When the alternate shutdown switch is placed in LOCAL the turbine cannot be operated from the control room but it does not affect the automatic operations of the turbine.
[TRUE or l'ALSE7]
(0 5)
a Question 2.02 (2.0)
a)
The Reactor Building Closed Cooling Water System (RBCCW)
provides cooling water to the MG set room area recirculating coolers. What other three coolers in the Reactor Recirculating system are cooled by RBCCW7 (1.0)
b) Will Indicated core flow read higher or lower than actual core flow if one reci rculation pump trips during power operati on? Briefly explain your answer.
(1.0)
Question 2.03 (3.0)
a)
Identify the below listed components by matching them with the appropriate letter from Figure 1.
Drive Piston (0.2)
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Scram outlet (0.2)
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Cooling Water Orifice (0.2)
Index Tube (0.2)
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Collet Fingers (0.2)
b)
Using Figure 1 as a guide, describ.e the flow pach through the drive mechanism on an insert signal.
(2.0)
Question 2.04 (2.5)
a)
In addition to the diesel generator being up to speed and voltage and the 10 second timer completed, list four condi tions that must be satis fied,before automat ic closure of the DG 4160 v output breaker occurs.
(1.6)
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b)
Can the diesel engine be started If power is lost to the diesel auxiliary panels (103A/104A)? Briefly
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(0.9)
explain your answer.
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Question 2.05 (3.0)
Approximately five percent of the steam flow from the reactor is
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used for auxiliary steam loads.
List three separate auxiliary systems or equipment that receive this steam and give the consequences, if any, should the auxiliary steam supply be shut off. Assume the.
(3 0)
reactor was operating at full power.
Question 2.06 (3.0, A core spray line breaks inside the shroud, a) Will the break cause an alarm in the control room? Briefly (1.5)
explain.
b)
How will the break affect core spray performance for (1.5)
that loop?
Question 2.07 (2.5)
Consider the standby Gas Treatment System (SGTS)
a)
uist tour initiation signals that automatically start (2.0)
the SGTS.
Include setpoints.
(0.5)
b) Why is a deluge spray system provided for the SGTS?
Question 2.08 (3 0)
Consider the Reactor Water Cleanup (RWCU) System.
a) Match the following temperature locations or limitations in Column A with the correct temperature listed in Column B.
(1.0)
COLUMN A COLUMN 8 1. Exiting high pressure reactor water Below 130*F from the regenerative heat exchanger 140*F
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to the non-regenerative heat IS0*F exchanger 460*F 2. Exiting reactor water from the non-regenerative heat exchanger 515'F to the RWCU reci rculation pumps 3. Exiting puri fied reactor water from the regenerative heat exchanger to the feedwater line 4. RWCU recirculation pump trip on high temperature CATEGORY CONTINUED ON NEXT PAGE s
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Question 2.08 b and c
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b)
in addition to the various high temperature trips, give three conditions that will cause a Group 6 isolation of the RWCU Sys tem.
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(1.5)
c)
The discharge isolation valve to the [ liquid radwaste
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system in the reject line automatically closes on high pressure upstream of the discharge valve (TRUE or FALSE?)
(0.5)
Question 2.09 (3.0)
Consider the Residual Heat Removc! (RHR) System.
a)
How does the LPCI selectica ;ogic determine which recirculation loop is brcken?
(assume both recirculation pumps are running)
(1.0)
b)
Assuming both recirculation pumps are running and that the break is located in the A loop, Indicate the automatic actions that occur to insure that water goes into the reactor via loop B rather than loop A.
(2.0)
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INSTRUMENTATION AND CONTROLS (25)
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Question 3.01 (3 0)
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Consider the Source Range Monitors (SRM's)
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a) Other than when a SRM is manually bypassed, name two conditions which will cause the " Upscale high rod block" to be automatically bypassed.
(1.5)
b)
The gas pressure in the fission chamber is reduced from fourteen and a half (14.5) atmospheres to two (2)
atmospheres because of a pin hole leak. Would the countrate increase, decrease, or remain the same because of the loss of gas pressure? Briefly explain your answer.
(1.5)
Quest ion 3.02 (2 5)
For the Narrow Range Yarways a)
State its range referenced to instrument zero.
(0.5)
b)
How far above the top of active fuel (TAF) is instrument zero?
(0.5)
c)
UniIke the Wide Range Yarways, the Narrow Range Yarways cannot be depended upon to indicate correct water level if conditions exist that could cause reference leg fg7..
(TRUE or FALSE 7)
(0.5)
d)
Narrow Range Yarways are temperature-compensated to improve level measuring accuracy (TRUE or FALSE?)
(0.5)
e)
Narrow Range Yarways are used to trip the HPCI Turbine on high water level as well as initiate the HPCI system on low low water level (TRUE or FALSE?)
(0. 5)
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Question 3.03 (2.5)
For each of the lettered conditions given below, indicate '
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which of the following WILL AUT0l1ATICALLY occur:
(if more than one acticn occurs, state the most severe action, i.e.,
hal f-scram is more severe than a rod block).
i)
scram ii)
ha l f-scram iii)
rod block iv)
no action a)
Two out of four of the Turbine Stop Valves close while operating at 50% power.
(0.5)
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b)
Three out of four of the MSIV outboard valves fail shut while operating at full power.
(0.5)
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Question 3.03 continued
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c)
One Main Steam Line High Radiation Monitor fails high.
(0.5)
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Du r i ng s ta r t'-up, A an~d H I RM 's fa i t downs ca l e.
(0.5)
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e)
During startup, A and D SRM's fall upscale while the reactor is s'ubcritical.
(Shortin'g links have been installed)
(0.5)
Question 3 04 (2.0)
Consider the' Air Ejector Off-Gas Radiation Monitoring System a)
State the type of detectors used and the principal type of radiation detected.
(0.8)
b)
What automatic actions could be initiated by this system
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over a 15 minute time span if no operator action is taken and both channels fall downscale.
(include any time delays)
(1.2)
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Question 3,05 (2.5)
While operating at full power and in three element control, one of the four steam flow inputs to the feedwater level control system is lost.
If no operator action is taken, will the reactor vessel water level be greater, less, or remain the same after the initial transient effects settle out? Briefly explain your answer.
(2.5)
Question 3.06 (3.0)
J Consider Average Power Range Monitors (APRM's)
a)
Power is supplied to the APRM's by what system?
(0. 5)
b)
List the three rod blocks (include set points where appropriate) associated with the flow converter.
(1.5)
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c)
Briefly explain why recirculation loop flow rather than total core flow is used in the flow converter.
(1.0)
Question 3 07 (2.0)
Loss of essential instrument air occurs while operating at full power.
a)
The reactor automatically scrams on low air pressure of 65 psig.
(TRUE or FALSE?)
(0.5)
b)
Indicate whether the following valves fail open, closed, or "as is."
i)
Feedwater regulating. valves (0.5)
li)
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lii)
The steam jet air ejector steam supply valve (0.5)
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Question 3.08 (2.5)
a)
The Recirculation Flow Control System uses two speed limiters (No. I and No. 2) to limit the recirculation pump speed under certain conditions.
For each speed limiter, give the condition
'that will initiate the limit and the percent of rated pump speed allowed.
(1.5)
b)
iheredscooptubelock-uplight Indicates that a lock-up is present. No Signal Failure Alarm occurs.
Indicate two other possible causes of the lock-up.
(1.0)
Question 3 09 (2.0)
With the mode switch in refuel, and during refuel operations, what four conditions (other than those initiated by the neutron monitoring system) prevent control rod withdrawal?
(2.0)
Question 3 10 (1.0)
For the 4160 V emergency service busses, list the normal power source and the three other available power sources.
List them in the order that they would energize the busses should the normal power source fall.
(1.0)
Question 3.11 (2.0)
Consider the Rod Block Monitor (RBM)
a)
Indicate two operational conditions that cause the Rod Block Monitor to be automatically bypassed.
(Assume bypass joy stick is centered).
(1.0)
b)
Indicate which of the followina condition (s) give a rod block when operating at 50% of full power.
(1.0)
1)
Failure to Null
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Less than 80% of assigned inputs operable lii)
RBM channel output drops to 5%
iv)
Accidentally selecting more than one rod at one time END OF CATEGORY
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25)
Question 4.01 (2.0)
You_are operating at full power and receive the following alarms:
PANEL C-7 ALARM and
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DRYWELL AIR COOLER HIGH DRAIN FLOW ALARM in accordance with Procedure 2.4.14, LEAKS INSIDE PRIMARY CONTAINMENT, list four (4) possible symptoms (other than alarms) that you would expect to occur.
(2.0)
Question 4.02 (3.0)
While operating at full power, M the Reactor Building Closed Cooling Wa er (RBCCW) System is lost due to the failure of all pumps in -
oops List the four immediate Operator Action steps that are given in Procedure M, Loss of M RBCCW m (3.0)
5; 3, V h!I question 4.03 (3.0)
According to Procedure E0P-01, RPV Control, Level and Pressure a)
Indicate the four entry conditions (include setpoints)
(2.0)
b)
Give the two separate condi tions (when confi rmed by two independent indications) which allow the operator to secure or place a CSCS in manual mode.
(1.0)
Question 4.04 (1.5)
in accordance with Procedure 6.1-012, Access to High Radiation Areas a)
What three access control conditions are required if the area is between 100 mrem /hr and 1000 mrem /hr?
(1.0)
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b)
What additional controls on access must be followed if the area is between 1000 mrem /hr and 10,000 mrem /hr?
(0.5)
Question 4.05 (2.0)
The operator must monitor the recirculation pumps speed when they are in manual individually a)
Wh a t is the allowable speed mismatch i)
if power is greater than 80%
(0.25)
ii)
if power is 80% or less?
(0.25)
b)
What are the two undes i rable operational condi tions that the limi ts in part "a" are designed to prevent?
(1.5)
CATEGORY CONTINUED ON NEXT PAGE
,
i
.
..
.
Question 4.06 (3.0)
Consider the Loss of Essential D.C. Bus D4 (Procedure 5.3.11)
a)
Indicate which of the below listed equipment is not
_
available due to loss of power (1.0)
1)
Core Spray Pump A II)
RHR Pump C III)
HPCI Iv)
,
v)
Diesel Generator A b)
If the bus cannot be re-energized, what are the immediate Operator Actions requi red?
(2.0)
-Question 4.07 (2.5)
When it is necessary to shutdown the reactor from outside the control room due to the inhabi tability of the control room (Procedure 2.4.143)
it is the Control Room Operator's job to scram the reactor.
a)
Give two methods by which this can be done.
(1.0)
b)
Indicate which method you would try first and justify your answer.
(0.75)
c)
How does the operator verify that the reactor is shutdown?
(0.75)
!
',
question 4.08 (2.0)
l
!
The PNPS Tagging Procedure (Procedure 1.4.5) provides instructions for the use of protective tags.
l
!
a)
What is the meaning of a white tag with a l
green border on a reactor recirculation pump?
(0. 5)
,
l b) What is the meaning of a white tag on 345 kV l
system equipment?
(0.5)
l l
c) Who authorizes issuance and pla' ement of protective c
I tags?
(0.5)
d)
What two approvals must be obtained to temporarily clear selected red tags?
(0. 5)
.
CATEGORY CONTINUED ON NEXT PAGE l
!
-
. - _, _ - - -
-.
-
.
..
.
Question 4.09 (2.5)
a) ~In accordance with Procedure E0P-02 RPV Control Power, Indicate the three entry conditions.
.
(1.5)
-
-
b)
List the two automatic actions that occur upon initiation of an ATWS trip as described in Procedure 2.2.126,
.
A.T.W.S.
(1.0)
.
Question 4.10 (2.0)
a)
Briefly explain why a failure of a jet pump is a serious problem from a safety standpoint.
(1.0)
b)
in addition to sudden unexplained changes in recirculation flow and Jet pump flow, indicate two other indications of a jet pump failure as given in Procedure 2.4.23, Jet Pump Flow Failure.
(1.0)
Question 4.11 (1.5)
-
,
Consider the immediate Operator Actions for a Rod Drif t (Procedure 2.4.3)
a)
What condition requires that the operator scram the reactor?
(0,5)
b)
If the rod is drifting in, what is the proper operator action?
(1.0)
'
4 END OF EXAM
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Y t + 1/2 at w = mg 5 "
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A=Aeg E = mc KE = 1/2 my a = (Vf - V,)/t x = en2/t1/2 = 0.693/t1/2 PE = m9h geff = [(tin)(t))
t h
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b 6E = 931 am I = I e "*
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SUR = 26.06/T Q = UAe T HVL = -0.693/u
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SCR = S/(1 - K,7f)
SUR = 28pf t* + (s p)T CR = S/(1 - K,ffx)
x
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CR (1 - K,ff)) = CR (2-eff2}
j T = (t*/a) + [(8 - p)/fo)
T = 1/(p - s)
M = 1/(1 - K,ff) = CR /CR,
'
g T = (8 - p)/(Ap)
M = (1 - K,77,)/(1 - K,7f j)
= (K,ff-1)/K,77 = d,ff/K SOM = (
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= -((t*/(T K,ff)] + [I,ff (1 + AT))
T = 0.1 seconds-I
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p
.
P = (r+V,)/(3 x 1010)
Idj=1d22 I = oN I d)
.= 1 d j
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R/hr = (0.5 CE)/d (meters)
-
.
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm.
1 curie = 3.7 x 1010dps 1 gal. = 3.78 liters 1 kg = 2.21 lbm.
I ft3 = 7.48 gal.
1 hp = 2.54 x 103 Stu/hr
-
Density = 62.4 lbm/ft3 1 rrw = 3.41 x 106 Stu/hr
Density = 1 gm/c.d lin = 2.54 cm Heat of vaporization = 970 Btu /lom
- F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm.
- C = 5/9 ("F-32)-
,
I Atm = 14.7 psi = 29.9 in. Hg.
.
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.
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-
G. E. ROBINSON 5/86 ANSWER SHEET NUTb PILGRIM
,
_
REACTOR OPERATOR EXAM l.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW (25)
1.01 Ref: G.
E. Reactor Theory, pg. 1-36 and pg. 3-10 Enabling obj 1.5 Chapt. 3, Enabling obJ 5.3, Chapt. 1
-
(2.0)
SDM = l-Keff; CRI (1-Keff) = CR2 (1-Kef f 2)
(200 cps)
(.04)
(1000 cps) SM2
=
.008 0.8%
SM2
=
=
1.02 Ref. G. E. Heat Transfer and Fluid Flow Manual, pgs. 8-26 and 8-46.
Enabling obj.
6.2 and 9.l_, Chapt. 8 (1.0)
a)
Steam quality is the ratio of the mass of steam to the total mass of fluid while void fraction is the ratio of the volume of steam to the total volume of the fluid.
(0.25)
6.)
inc reases (0.75)
as steam quality increases the density decreases thus increasing the flow rate (velocity).
The higher the flow rate (velocity) the greater the resistance to flow.
1.03 Re f:
G. E. Reactor Theory, pgs. 4-39, Enabling obj 6.3, Chapt. 4 (0.5)
less negative (1.0) - An increase in moderator density decreases the slowing down
.
length and lifetime of a neutron and probability of resonance absorption (Note: any one of the above is acceptable)
1.04 Ref:
G. E. Reactor Theory Manual, pgs. 6-8, 6-15, 6-11 and 6-13 Enabling obj 2.1, 2.5, 2.6, 3.5, 3.6, Chapt. 6 (0.5)
a)
False Not directly proportional (0.5)
b) True'
.
(0,5)
c)
True (0.5)
d)
False Pu-239 causes a decrease in equilibrium xenon
,
(0. 5)
e) True 1.05 Re f:
G. E. Heat Transfer and Fluid panual, pg. 7-45 (1.0)
a)
To prevent cavitation in the condensate pumps (or to provide adequate condensate pump suction head (1.0)
b)
Condensate depression reduces plant ef ficiency therefore excessive condensate depression results in unnecessary loss of plant efficiency.
- _ _ _
.
1.06 Ref:
G. E. Reactor Theory Manual, pg. 4-8 Enabling obj 1.5, 3.6, Chapt. 4
_
(0.5)
decrease in power (2 of 3 required)
(1.25 pts. each)
(2.5)
1.
Immediately the loss of feedwater flow causes a decrease in moderator subcooling which introduces negative reactivity because of a negative moderator coefficient.
2.
When feedwater flow drops below 20%, the recirculation pumps will auto runback to 28%. The decrease in core flow causes an increase in voiding thus a reduction in power.
Decreasing level in the downcomer will reduce the available head for core circulation and results in decreased core flow and thus reactor power will decrease.
1.07 Ref:
G. E. Reactor Theory Manual, pg. 5-16 Enabling obj 2.5, Chapt. 5 (0.5)
increase or decrease rod worth (1.0)
if several rods are quite close together, the presence of rods in the reactor will depress the flux and lessen the worth of the other rods in the immediate vicinity.
(1.0)
it is also possible for a rod or group of rods to be inserted into a reactor thus suppressing their local flux, while increasing the flux in another region of the core and thereby increasing the rod worth positioned where the flux was increased. or co,e couphg of Cual uQs occ uv win e n on 4'* t ve ds o ve ed o m e% ckso n end s ee so M dra w n /
- e red a re4 * A s,e// west.8 4 714 neutva. 88w 1.08 Re f:e ar d'.*I.8 "He*a*t T ran s fe r an d F l u i d F l ow Man ua l, pgs. 9-16, 9-19 y
and 9-33 Enabling obj.
3.1, 3.6, 4.1, 4. 2, 5.1 and 5. 3, Chapt. 9 (Note: 0.5 pts for each match)
'
COLUMN A COLUMN B COLUMN C (1.0)
a) LHGR 2) local fuel pin i) 1% plastic strair power in node on cladding (1.0) b) APLHCR 3) average fuel pin lii) clad temp. of 2200*F power in node (1.0)
c) CPR 1) total fuel bundle ji) boiling transition power 1.09 Ref:
G. E. Heat Transfer and Fluid Flow Manual, pgs. 6-95, 6-13, 6-96 and 6-109 Enabling obj 10.12, 4. 3, 10.14, 10.15, Chapt. 6 (0.5) a)
increase
.
(0.5)
b)
mass flow rate in equals mass flow rate out (0.5) c)
.sie e&4/
(0.5) d)
pump runout is when pump flow 'is greater than design flow
.
- - -,,,-,
,_
.
_
-.
.
- - -
.
-
(0.5)
the excessive flow usually occurs due to low pressure downstream of the pump.
(will also accept various examples i.e., pipe break, too much flow demand, etc.)
,
.
.
-
I 1.10 Ref.
G. E. Reactor Theory Manual, pgs. 3-15 and pg. 7-10 Enabling obj 3 3, Chapt. 3; 3.3, Chapt. 7
"
,
-
_
T 00 sec P(o) e
in 10 (1.0)
P(t)
=
=
T
'
130.2 seconds T =
=
2.3 period has become longer, therefore heat range has been reached 1.11 Re f:
G. E. Reactor Theory Manual, pgs. 4-9, 4-19 and 4-26
~
Enabling obj 1.5, 3.6 and 6.3, Chapt. 4 (0. 25) a)
reactivity increases (0.75)
due to cooler water entering reactor there is an increase in water density (shorter slowing down and diffusion lengths)
less neutron leakage (moderator temp coefficient)
'
(0.25) b)
reactivity increases (0.75)
due to collapse of voids, increase in water density, less neutron leakage (Void Coef ficient)
(0. 25)
c)
reactivity decreases (0. 75)
due to higher fuel temperatures more parasitic absorptions (doppler effect) occur in the resonance region (Doppler Coef ficient)
.
.
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25)
2.01 Ref:
Sys. Ref. Text 0215, HPCI, pgs. 3-1 and 45-1 Enabling obj 2 and 4, Sys. Ref. Text 0215 (0.5)
a) High Steam flow (0.5)
high area temperature (0. 5)
low steam line pressure (1.0)
b)
prevents condensation from building up in the exhaust line estwfs drawe 9 and causing water hammer upon system initiationwe 4 s.d. ptG Aubenef er W esa.w.*o eph==,4 i o n g.4 drmed es e s es y s,sek. /s 4,eae44 n, n,,4,.*g, (0.5)
c)
FALSE
,5 2.02 Ref: Sys. Ref. Text 0134, Reactor Recirculation System, pgs. 29-! and 32-1 Enabling obj 5 and 17, Sys. Ref. Text 0134 (O.33)
a)
M.G. Set oii coolers of,,
ye,,r, p..f g,y.,// s,,/w (0. 33)
pump motor lub oil cooler y,h: emy fico e (0.34)
recirculation pump seal coc,lers (0.25)
b)
Higher (0.75)
because 20 jet pump flows are summed even though flow is reversed in 10 of them.
2.03 Ref: Sys. Ref. Text 0183, Control Rod and CRDM, Figs. 2 and 20 Enabling obj 3 and 4, Sys. Ref. Text 0183 (0.2)
a) 0 Drive Piston (0.2)
R Scram outlet (0.2)
F Cooling Water Orifice (0.2),
D Index Tube
-
(0.2)
M Collet fingers (2'. 0 )
b)
See enclosed Figure 20 2.04 Ref; Sys. Ref. Text 0213, DG, pgs. 27-1 and 47-1 Enabling obj 2 and 4, Sys. Ref. Text 0213 (0.4)
a)
DG lockout relay reset (0.4)
Start-up transformer 4160v breaker open (0.4)
Auxiliary transformer 4160v breaker open (0.4)
Shutdown transformer 4160v breaker open (0. 3)
b) Yes (0.6)
The diesel engine can'bc started provided the air receivers are charged prior to'the loss ov LL3tw y fteecslsor c$ d vods, h veCr Ct'ba Y h Y ^
4W $)e be e's l5
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- lea REACTOR PRESSURE L=
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120 121 DIRECTIONAL
' k CONTROL V ALVES X
) EXHAUST HEADER CONTROL VALVES
INSERT
-
123,121 123 JL 122 WITHDR AW AL
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FIGURE 20 RE..
.
2.05 Ref:
Sys. Ref. Text 0138, Main Steam System, MSIV's and RV's, pgs. 4-1 and 27-1
_
Enabling obj 9, Sys. Ref. Text 0138-(1.0)
a)
turbine steam seal system (loss of steam to regulator)
no major concern (1.0)
b)
off gas ejectors; condenser vacuum will decrease (1.0)
c) augmented off gas system, loss of steam to superheat gases removed by air ejectors for efficient operation of the he
$4 Wpessse
/* ss & d* U *' ' N'*
2.06 Ref:
Sys. Ref. Text 0220, Core Spray, pgs. 10-1 and 11-1 Enabling obj 8, Sys. Ref. Text 0220 (0.5)
a)
NO (1.0)
if a core spray line breaks inside, the differential pressure indicating switch (dPIS) will detect reactor pressure inside the shroud as usual, therefore no abnormal differential pressure will be indicated.
(1.5)
b) The core spray loop can perform a flooding function but its spray will not provide full core spray coverage.
2.07 Re f:
Sys. Ref. Text 0234, SGTS, pgs. 9-1 and 6.-I Enabling obj 2 and 3, Sys. Ref. Text 0234 (0.5)
a)
reactor low water level, + 9 inches (0.5)
high drywell pressure, greater or equal to 2.5 psig (0.5)
high radiation level in refueling floor exhaust ducts, 16 mR/hr (100 mR/hr tech. specs.)
(0.5)
Simultaneous downscale from all four refueling floor duct
'
ev 'g,{,
- ".N"bMeele + e e,t our cl nnel o<.d j opscole Os v/h) ru N *$-
e
,
(0.5)
b)
to prevent fires from occurring in the charcoal filter beds NOTE: The trip condition is worth 0.3 pts. and setpoint 0.2 pts.)
2.08 Ref:
Sys. Ref. Text 0103, RWCU System, pgs. 4-1, 4-2, 18-1, 19-1 and 21-1 Enabling obj 5, 6, 8 and 9, Sys. Ref. Text 0103 (0.25)
a)
1.
190*F Ca e "P4 8 50 *I
4'* T )
(0.25)
2.
below 130*F (0.25)
3. @v 4'of (0.25)
4.
140*F (0.5)
b)
Standby Liquid Control' Actuatign (0.5)
Reactor Low Water Level (+ 9 inches)
(0.5)
RWCU inlet high flow (300% of rated flow)
(0,5)
c) TeUE (5 /.re
'
.
.
-
&
2.09 Ref:
Sys. Ref. Text 0221, RHR, pg. 3.1 Enabling obj 6 and 7, Sys. Ref. Text _0221 (1.0)
a)
compares the pressure in the five riser pumps in one loop with the pressure in the corresponding risers in the
--
other recirculation loop. The undamaged loop will have the higher pressure.
(0.5)
b)
Loop B recirculation discharge valve is closed (0.5)
Recirculation pump B motor is tripped ( Need ne+ tas/w/c 'M'8 'd mM sol (0.5)
Loop A injection valves are sealed (for 10 min.)
(0.5)
Loop B injection valves open (when reactor pressure is less than 400 psig)
'
.
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,
.
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, _ _
_ _ _ _ _. _ _ _ _ _
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- _ _ _
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_
.
.
--
.
INSTRUMENTATION AND CONTROLS (25)
3.01 Ref:
Sys. Ref. Text, 0170, SRM pgs. 12-1 and 3-1
-
Enabling obj 2, 4 and 6, Sys. Ref. Text 0170
_
(0.75)
a)
Mode switch is in run
'
(0.75)
associated IRM. range switches are at range 8 or above (0.5)
b)
decrease (1.0)
The argon fillgas is ionized by the high energy fission fragments.
Lower gas pressure would produce less ionization and fewer pulses that would pass through the PHD 3.02 Ref:
Sys. Ref. Text 0177, Nuclear Boiler Instrumentation, pgs. 6-1, Fig. 7, 2-1, 8-1, Sys. Ref. Text 0177 Enabling obj.
2, Sys. Ref. Text 0177 Note:
Part C is not specifically covered by an enabling obj but is closely related to enabling obj 5 and is information the R.O. should know (0.5)
a)
+ 50 in to - 50 in (0.5)
b)
127 inches (0.5)
c)
FALSE (0.5)
d) True (0.5)
e) True 3.03 Ref: Sys. Ref. Text, 0191, RPS, pgs.13-1, Figs. 9,10 and 11 Sys. Ref. Text, 0170, SRM, pg. 12-1 Enabling obj 3, 4 and 9, Sys. Ref. Text 0191
,
(0.5)
a)
ii ha l f-scram av Me 4=4ww (0.5)
b)
i scram (0.5) ' c)
half-scram
'
(0.5)
d) lii rod block (0.5) e)
lii rod block 3.04 _Re f : Sys. Ref. Text 0211, Process Radiation Moni toring, pgs. 7-1 and 8-1 Enabling obj 3 and 5, Sys. Ref. Text 0211 (0.4)
a)
ion chamber detectors (0.4)
primarily sensitive to gamma rays (0.4)
b) once the timer times out (usually set at 13 minutes)
(0.4)
off gas hold up drain valve closes (AO-3750)
(0.4)
of f gas outlet isolation valve closes (AO-3751)
,
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-
-
,---.
-.
-
.
.
"
.
.
3.05 Ref:
Sys. Ref. Text 0178, Feedwater Level Control System, pg. 20-1 Enabling obj_ 11-2, Sys. Ref. Text 0178 (0.5)
decrease
_
(2.0)
Feedwater flow will appsar to be greater than steam flow, therefore a permanent mismatch between indicated steam flow and actual feedwater exists.' This mismatch produces a voltage differential (negative) which is balanced by a voltage differential produced by desired water level minus actual water level (positive).
3.06 Ref:
Sys. Ref. Text 0169, APRM, pgs. 17-1, 10-1 and 5-1 Enabling obj 2, 3 and 4, Sys. Ref. Text 0169 (0.5)
a)
RPS (0.5)
b)
10% flow mismatch between channels (0.5)
flow exceeds 110%
(0.5)
either module is unplugged f ("
,'
- [*
p ie.
.
(1.0)
c)
core flow is the summation of all the jet pump flows which has large fluctuations and is therefore noisy. This noise would cause a number of spurious scrams.
Reci rcula t ion loop flow is stable which greatly reduces the chance of spurious sc rams.
3.07 Ref:
Sys. Ref. Text 0103, Compressed Air System, 14-1 and 15-1 Enabling obj 3, Sys. Ref. Text 0103 (0.5)
a)
FALSE (0,5)
b)
i)
"as is" (0. 5)
ii) open (0.5).
iii)
closed
'
3.0.8 Ref:
Sys. Ref. Text 0135, Recirculation Flow Control System, pgs. 2-1 and 10-1 Enabling obj 4, 5 and 7, Sys. Ref. Text 0135 (0.75)
a)
Speed Limi ter 1 Recire. discharge valve less than 90% open or Feedwater flow is less than 20% of rated Maximum pump speed is 28% of ra.ted (0.75)
Speed Limiter 2 All three RFP's are not operating and reactor water level is 19 inches or less Recirc. pump speed is limi ted to less than 65%
(1.0)
b)
(any two)
(0.5 pts, each)
M-G set drive motor bus undervoltage (80% of rated voltage)
on bus A-3 or A-4
.
Lube oil high-high temp. (210*F)
Lube oil low pressure (six second delay)
(30 psig)
.
.
.
3.09 Ref: Sys. Ref. Text 0236, Refueling, pg. 28-1 Enabling obj
Sys. Ref. Text 0236 (0 5)
fuel grapple is not fully up and refueling platform is over the core.
(0.5)
any hoist is loaded and the refueling platform is over the core
,
M*
/ '"#
(0.5)
one rod is withdrawn and a second rod is selected
"*O (0.5)
the service platform hoist is loaded g
4.me #re 3.10 Re f:
Sys. Re f. Text 0103, 4160V, pg. 2-1 4,tg m,.= k / $8'#/
Enabling obj 2 and 4, Sys. Ref. Text 0103 c.ek,g n e de n e..It &4.
o.w a* * *
(1.0)
Unit auxiliary transformer (normal)
tart-up transformer (preferred)
3 kV shutdown transformer (emergency)
Diesel Generators (standby)
(Note:
0.2 pts. for each source and 0.2 pts, for correct order)
3.11 Ref:
Sys. Ref. Text 0176, Rod Block Moni tor, pgs. 10-1 and 21-1 Enabling obj 3, Sys. Ref. Text 0176 (0.5)
a)
If the reference APRM is less than 30% *
Weehem,,, A so l s
(0.5)
If an edge rod is selected (0.25)
b)
i)
Rod block e.. ' d be <* / E l** E 'I
'h P
(0.25)
li)
no rod block av less Ha~ ** l*
'S N " 5* #* *
...)
no rod block (0.25)
iii (0.25)
iv)
rod block (Note:
the answers expected are (i) and (iv)
.
.
l
s
'
.
.
.
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL (25)
_
4.01 Ref:
Procedures 2.4.14, Leaks Inside the Primary Containment, p.
2^.
--
(0.5)
a)
Excessive Sump Pump Operation due to increased leakage
-
to drywell equipment drain sump or drywell floor drain sump (0.5)
b) Abrupt change in drywell humidity (0.5)
c)
Significant drywell pressure (0.5)
d)
Significant drywell temperature 4.02 Ref: Procedure 2.4.42, Loss of A or B RBCCW Loops, p. 2. M sn
- M'4o/
(0.75) a) Open the manual section and discharge the valv on the east wall of "A" loop RBCCW area.
(0.75) b)
Close one 3" makeup valve from one the loop surge tanks (to minimize back surgin d overflowing of tanks.
(0.75) c)
Start all pumps in operating loop.
(0.75) d) Maintain the t exchanger outlet temperature 80*F.
Adjust r sor power i f necessary to maintain temperature.
4.03 Ref:
Procedure E0P-01, RPV Control, Level and Pressure, pgs. 2 and 21 (0.5)
a)
RPV water level below 9 inches or (0.5)
Drywell pressure above 2.5 psig or (0.5)
An isolation which requires or initiates reactor
-
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(0.5)
RPV pressure above 1085 psig (0.5)
b) Misoperation in automatic mode is confirmed.
(0.5)
Adequate core cooling is assured.
4.04 Ref:
Procedure 6.1-012, Access to High Radiation Areas, pgs. I and 2 (0.34)
a)
area shall be barricaded (0. 33)
conspico,usly posted as a High Radiation Area (0.33)
entrance controlled by RWP (0.5)
b)
locked doors shall be provided
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I. SYMPTOMS
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A.
Increasing temperatures in components cooled by the RECCW System.
B.
Loss of all RBCCW System pressure.
II. AUTOMATIC ACTIONS None III. IMMEDIATE OPERATOR ACTION A.
Reactor Initially in Operation g g3po ggt 1.
, Attempt to restore the RBCCW System.
If neither loop can be restored within 5 minutes, proceed with the following steps:
n 0.2 2.
Scram the reactor.
3.
Tie the two loops together and reduce the system heat lead as much as possible by shutting down non-essectial equipment.
4.
Let the pressure slowly decay, using the main condenser as a heat sink.
,
B.
Reactor Initially in Shutdown Status 1.
Attempt to restore the RBCCW System.
If neither loop can be restored within 10-15 minutes, proceed with the following steps:
2.
Insure that all centrol rods are inserted, and trip the CRD
- '
pump in operation.
-
3.
Tie both loops together and reduce the system heat load as much as possible by shutting down acn-essential equipoent.
.
4.
Let the pressure slowly decay, using the main condenser as a heat sink.
IV. SUBSEQUENT OPERATOR ACTION A.
Refer to Procedure 5.7.1.1, " Emergency Categories and Associated Emergency Action Levels", to determine if an emergency category re-quires implementation.
b B.
Notify the Station Manager or designated alternate. Notify the NRC within one hour. At that time, establish and maintain an open continuous communications channel with the NRC.
C.
Begin a slow controlled depressuritation of the reacter utilizing main condenser as a heat sink.
If. the MSIV's close, reset the trip
,
'g and reopen the valves.
.
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Re(ereitCe PAP.5 J'toutbusE
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4.05 Ref:
Sys. Ref. Text 0135 Recirc. Flow Control System, pg. 14-1 Enabling obj 8, Sys. Ref. Text 0135 (0.25)
a) above 80% reactor power - 10 percent or less (0.25)
-
80% or less - 15 percent or less (0.75)
b)
to prevent excessive vibration associated with the jet pump risers (0.75)
for LOCA's it is possible for the LPCI loop select logic to select the wrong loop for injection with the recirculation loops operating at large differential flows 4.06 Ref: Procedure 5.3.11, Loss of Essential D.C. Bus, pg. 2 and 3 (1.0)
a)
i)
Core Spray Pump A 11)
RHR Pump C iv)
RCIC v)
DG A Note: HPCI is still available c
(0.4)
b) Trip feedwater pump A and A manually (0.4)
Then scram the reactor (0.4)
Manually trip the main turbine (0.4)
Manually trip all breakers on affected buses (after turbine trips and load centers are de-energized)
(0.4)
Take appropriate action for partial loss of Service Water and RBCCW systems.
4.07 Ref: Procedure 2.4.143, Shutdown from Outside Control Room due to Fire in C.S.R. or Tornado Alert or inhabitability of Control Room, pgs. 7 and 8 (0.5)
a)
Open breakers to the APRM's at the RPS power panels
.
(0.5)
Trip the power supply to the RPS (cens'4m * # * * ** * * )
(0.25)
b) APRM Method (0.5)
Trioping power supply infers loss of power and causes t
l undesirable events (such as early closure of MSIV's)
to take place.
(0.75)
c) observe that scram valves are open at the hydraulic control modules.
l 4.08 Ref:
Procedures 1.4.5, PNPS Tagging Procedure, pgs. 5, 6 and 8 (0.5)
a) This is a testing tag which indicates that a piece of equipment is being tested or is being held for testing (0. 5)
b) no not operate (0.5)
c) on-watch Watch Engineer or Operating Supervisor l
(0.5)
d)
implementing Supervisor / worker abd Watch Engineer l
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4.09 Ref:
E0P-02, RPV Control Power, pg. 2 Ref:
Procedure 2.2.126, A.T.W.S., pg. 2
-
(O'. 5)
a) A condition exists which requires a reactor scram and reactor power is above 3%
(0.5)
or power cannot be determined (0.5)
or all control rods are not inserted past position 04 (0.5)
b) Trip both reactor recirculation pumps (field breakers of the pump motors)
(0.5)
Energize two independent backu'p scram solenoid valves ( A g2 3 0 '~ -a
~
(thus discharging the SPVAH)
4.10 Re f:
Procedure 2.4.23, Set Pump Flow Failure, pg. 1 (1.0)
a)
it could preclude the capability of maintaining 2/3 core coverage during a LOCA ey mereeses eFF*c4we-8ofeiddeau% em = = loch.
('.5)
b)
Recirculation System delta P deviation O
(determined by OPER 09 Test 32)
(0.5)
Sudden change in core AP without accompanying change in Recirculation pump speed 4.11 Ref: Procedure 2.4.3, Rod Drift, pg. 2 (0.5)
a)
If two rods in a nine rod array start drifting in (1.0)
b)
check that the cooling water pressure is not too high then give the drive a notch out signal
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U.
S.. NUCLEAR REGULATORY COHHISSION
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SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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FACILITY:
-PILGRIH
_
_________________________
REACTOR TYPE:
BWR-GE3
_________________________
DATE ADMINISTERED: 86/05/06
_________________________
EXAMINER:
HOWE, A.
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 grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
________ ______
___________
________ ___________________________________
__'S 00'_1____
__25 00
________ 5.
THEORY OF NUCLEAR POWER PLANT
_1__
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OPERATION, FLUIDS, AND THERMODYNAMICS
_
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1_
___________
________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
,
AND INSTRUMENTATION-5.00 25.00
________ ______
___________
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_
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________ 8.
ADMINISTRATIVE PROCEDURES,
___________
CONDITIONS. AND LIMITATIONS 100 00 100.00 TOTALS
,
________ ______
___________
________
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FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid.
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APPLICANT'S SIGNATURE
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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QUESTION 5.01 (2.50)
a. When comparing the individual BETA's from thermal fission of U-235, Pu-239 and fast fission of U-238, which BETA is largest?
(0.5)
6.
From BOL to EOL, does the core average beta INCREASE, DECREASE or REMAIN THE SAME?
EXPLAIN your answer.
(1.0)
c.
For equivalent positive reactivity additions to a critical reactor, will the period be longer at EOL or BOL? WHY?
(1.0)
OUESTION 5.02 (2.75)
Assume that the reactor is being started u!> with the bulk coolant temperature being less than the saturation temperature. Suddenly several control rods malfunction and the reactor begins to increase in power level on a short period.
a.
Of the Void, Doppler and Moderator Temperature coefficients which would come into effect first, second and third to lesson the rate of pow.er increase?
(0.75)
b.
EXPLAIN your choices of part a.
1.
Assume the operator takes no action.
2. Include a discussion of fuel time constants in your answer.
3.
Assume a scram does not occur.
(2.0)
GUESTION 5.d3 (3.00)
.
Following.an automatic initiation of LPCI at a reactor pressure of 350 psig, reactor pressure decreases to 100 psig. For each of the parameters listed below, determine any change (i.e.
increase, decrease, or remain the same). BRIEFLY EXPLAIN why the parameter changes or c'emains the same.
a.
LPCI injection flow E1.03 b.
LPCI pump discharge head ( assume constant NPSH )
[1.0J c.
LPCI pump power requirements
[1.03 (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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QUESTION 5.04 (2.00)
Of the statements listed below, choose the statement (s) that is (are)
true with respect to suberitical multiplication.
c. As K-eff approaches unity, a larger change in neutron level occurs for a given chan3e in K-eff.
[0.53 b.
If source neutrons are present in a reactor that is just critical (K-eff=1), count rate will increase at an exponential rate.
CO.53 c. When K-eff=0.95 and the neutron source = 100 n/sec.,
CR'will equal 1500.
00.53 d.
The closer K-eff is to unity, the longer it takes for neutron level to reach equlibrium for a given change in k-eff.
[0.53 GUESTION 5.05 (2.75)
Regarding core thermal limits:
a.
The process computer output, CMFLPD, is used to monitor Which core thermal limit?
E0.503 b. Which core thermal limit ensures peak cladding temperature will not exceed 2200 deg.F following a LOCA?
CO.753 c. What is the failure mechanisim and the limiting condition for LHGR?
[1.03
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c. What potential problems would exist in a fuel bundle if it were operated in excess of the MCPR limit?
CO.53 OUESTION 5.06 (2.00)
You are on shift, reactor power is at 95%, control rod manipulations are in progress per Reactor Engineer instructions.
The reactor operator (RO) reports to you that that the LPRM readings on the four rod display are errone'ous since indicated power decreased when a rod was withdrawn from notch 36 to 40.
A check of the LPRM's in question show no malfunction. What is the cause of the power decrease? Explain your answer.
(2.00)
(***** CATEGORY 05 CONTINUED 0,N NEXT PAGE
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OUESTION 5.07 (1.50)
The reactor is operating at 75% power when the EPR system power is lost. How would the following parameters INITIALLY change and why?
A.
Reactor pressure (0.50)
B. Core flow (0.50)
C.
Reactor power (0.50)
QUESTION 5.08 (1.00)
The reactor scrams after operation at high power for a long time.
Station management has determined that the plant shall be cooled down as quickly as possible. What are three factors that will contribute to a change in the SDh during and after cooldown?
(1.0)
GUESTION 5.09 (2.00)
a.
Consider two control rods. Both rods are at notch position 16.
Rod A is located near the center of the core and rod B is located at the core edge. The reactor scrams after operating at high power for a long time. A hot startup was performed and power reached 30% about ten hours after the scram. To add the most reactivity at this time with a one notch withdrawl, WHICH rod would you choose and WHY?
(1.00)
'
b. Would a fully inserted control rod have greater differential
'
worth if it was next to a fully withdrawn control rod or next to a fully inserted control rod.? Explain your answer.
NOTE: Assume average core flux is constant.
(1.00)
!
OUESTION 5.1'O (2.50)
E0P-1 RPV Control, Level and Power discusses the effects of accident conditions on reactor vessel level indication. Under what two accident conditions would level indication be unreliable?
(Explain how indicated level is affected (higher / lower) and WHY?)
E2.003 (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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QUESTION 5.11 (3.00)
,
Indicate whether the following will INCREASE or DECREASE reactivity during operation AND briefly EXPLAIN why.
a.
Moderator temperature increases while below sattuation temperature.
(.75)
b. Fuel temperature increases.
(.75)
c.
Loss of a feedwater heater.
(.75)
d.
A sudden reduction in reactor primary system pressure.
(.75),
b
.
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
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QUESTION 6.01 (1.50)
How can a diesel generator be shutdown eith an emergency _
signal present? (List at least four (4) methods)
(1.50)
GUESTION 6.02 (3.00)
The main generator is at 640 MWE when condensor vaccum starts decreasing rapidly. What alarms and trips will occur as a DIRECT result of this decreasing condensor vaccum and WHAT is the purpose of each alarm / trip?
(3.0)
(NOTE: SETPOINTS ARE REQUIRED)
GUESTION 6.03 (3.00)
a.
The Standby Liquid Control System has a maximum injection time designed into it.
WHAT is the maximum injection time and WHY is the time estabilished?
(1.25)
b.
What indications are available to you in the control room to verify proper SLC system injection?
(Seven (7) required)
(1.75)
OUESTION 6.04 (2.50)
The reactor is at 35% power when Vital Services 120/240 V Power Supply Panel Y-2 becomes and remains de-energized. WHAT are five (5)
MAJOR components, systems, or subsystems that are lost which could SIGNIFICANTLY REDUCE your ability to SAFELY OPERATE the plant from the CONTROL ROOH?
[2.5]
.
.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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PLANT-SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
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QUESTION 6.05
,50)
Consider a complete loss of Essential Instrument Air with the plant operating at full power and with no operator cetion and answer the fo11owin3 1 a. What would be the effect on the following components:(Note: limit your answer to effects caused in relation to instrument air only)
1.
CRD Hydraulic scram valves (0.33)
2. SJAE steam supply valve (0.33)
3.
Main Feedwater Regulating valves (0.33)
et Hould an automatic scram signal be generated? If not, why? If so, from what automatic signal?
(0.75)
er 0:1;f instrument air has been restored and all RPS signals ste cleared, what action is necessary to repressurize the SPVAH and close the scram valves?
(0.75)
GUESTION 6.06 (3.00)
With regard to the Automatic Depressurization System (ADS):
a.
What FIVE indications, other than annunciators, are available in the control room to determine if a Safety Relief Valve has lifted following a main steam line isolation from power?
(1.25)
b.
The reactor is operating at 100% power when a small b.reak occurs inside containment. HPCI and RCIC fail to start automatically and manually. All signals are
'
valid for ADS actuation except for the timer not timed out.
1.
The operator pushes the High Drywell Pressure Reset pushbutton (drywell pressure is 2.8 psig).
Will the ADS timer reset? EXPLAIN.
(0.75)
2.
RCIC is now started and water level is raised i
to -39 inches. What effect does this have on ADS initiation if the timer has NOT timed out? If the timer has timed out? Briefly EXPLAIN both.
(1.0)
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
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QUESTION 6.07 (2.00)
For each of the IRM (Intermediate Range Monitoring) range changes listed below provide the following:
1.
The indicated level on the new range and 2.
Any automatic actions initiated as a result of the indicated level on the new range.
a.
Switching from range setting 5, reading 25, up to rance setting 7.
Reactor mode switch in Startup.
(1.0)
b.
Switching from range 6, reading 39, down to range 5. Reactor mode switch in startup.
(1.0)
OUESTION 6.08 (3.00)
Regarding the Core Spray System i
a.
Why is there an interlock between the inboard and the out-board isolation valves allowing only one of the two valves in each loop to be opened with reactor pressure greater than 400 psig?
(1.0)
b.
Assume an operator inadvertantly goes to 'CLOSE' on the inboard isolation valve control switch with an auto initiation signal present. Reactor pressure is at 675 psig and rapidly decreasing. At 350 psis the operator realizes
<
his mistake and releases the switch. Will the inboard isolation valve AUTO OPEN with the auto initiating signal
'
still present (YES or NO)? Briefly JUSTIFY your choice.
(1.0)
c.
The Core Spray Break Detection alarm has just actuated.
Assuming a valid alarm, give the possible areas where
-
the piping could break (or leak) and cause the alarm.
NOTE: Limit your answer to CS piping NOT instrument lines. Also consider only one CS subsystem and when describing areas give boundaries such as valves or i-other major components.
(1.0)
.
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QUESTION 6.09 (3.00)
The Feedwater Level Control System is being operated in 3-element control using reactor level detector channel A".
Reactor power is
'
at 85%, steady state. For each of the instrument or control signal failures listed below, INDICATE-HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase, decrease, or remain constant) and BRIEFLY EXPLAIN WHY in terms of what is happening in the Level Control System icmediately following the failure.
NOTE: Your answers should include the effects on Reactor Level, Steam Flow / Feed Flow mismatch, feedwater valve position. Consider each failure seperately.
a.
'B'
Feedwater Line Flow signal FAILS HIGH.
(1.0)
b. Channel
'A'
Main Steam Line PRESSURE detector signal fails low.
(1.0)
c.
The electrical signal to the
"B" Feedwater Regulating Valve is lost.
(1.0)
GUESTION 6.10 (1.50)
Concerning the Refuel System
,
a.
For the following indications in the Operators Cab Left
'
Control Panel' give the condition and setpoint which will energize the indicating light:
1.
Grapple full down position 2. Grapple hoist jam
'3 Grapple slack cable (0.33 each)
.
b.
The refueling interlocks sense various control rod positions and refuel equipment status. What is the design intent for the refuel interlocks? (Answer in terms of overall purpose of system)
CO.5]
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QUESTION 7.01 (2.50)
You have received the HIGH PRESSURE IN THE OFF-GAS LINE and HIGH TEMPERATURE IN THE OFF-GAS LINE alarms and suspect an
-
explosion in the Off-Gas system.
_
a.
What immediate actions must be taken?
(1.0)
b.
What would you check-in the control room to determine if radiation was being released to the environment?
(0.5)
c. What action should be taken to prevent a second off sas 1 system explosion?
(1.0)
GUESTION 7.02 (3.00)
The reactor is operating at 95% power when a complete loss of all service water occurs.
a.
Other than annunciaters, what are two symptoms that all service water has been lost? Note: Your answer should be stated in terms of two DIFFERENT TYPES OF PARAMETERS which are observable in the control room. BE SPECIFIC.
(1.0)
6. Attempts to restore service water have failed and ten (10)
seconds have elapsed. What are four (4) immediate actions you would perform.
(1.0)
c. The procedure allows only ten (10) seconds to restore the service water pumps before taking the other immediate actions. WHY is such IMMEDIATE action necessary?
(1.0)
.
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QUESTION 7.03 (3.00)
Concerning operations of recirc pumps:
a.
Procedure 2.1.~9 REACTOR RECIRCULATION PUMP OPERATION, contains a caution which precludes starting an idle recirc pump unless certain delta temperature requirements are met.
What are these requirements and what are the specific reasons for these requirements?
Note: Reactor is critical and one pump is running.
(1.50)
b.. Procedure 2.2.84 REACTOR RECIRCULATION SYSTEH, has a caution
' pertaining to limits on recire M-G set starts.
1.
What are the conditions and start limits?
(1.00)
2.
Why are these limits imposed?
(0.50)
GUESTION 7.04 (3.00)
Regarding primary coolant leakage into the primary containment a.
What are the limits for coolant leakage?
(1.0)
6.
Other than alarms, what are three symptoms of excessive leakage?
(1.0)
c.
Procedure 2.4.14, LEAKS INSIDE THE PRIMARY CONTAINHENT, directs you to enter E0P-04 PRIMARY CONTAINHENT CONTROL, TEMPERATURE if either of two specific criteria are met.
What are these criteria?
(1.0)
.
GUESTION 7.05 (2.00)
What are the four HAJOR OBJECTIVES of E0P-1, RPV CONTROL?
[2.03 (x**** CATEGORY 07 CONTINUED ON NEXT PAGE
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QUESTION 7.06 (1.00)
As a prerequisite to core alterations, procedure 4.3, FUEL HANDLING, requires SRH's to be operable and survie11ance testing of SRM's to be performed per the Technical Specifications.
_
a.
What is the minimum number of operable SRM's required during core alterations? Include any location requirements.
(1.0)
QUESTION 7.07 (2.00)
a.
In.accordance with procedure 2.1.1 "Startup From Shutdown *,
under what conditions shall secondary containment integrity be maintained?
(1.0)
b.
You are allowed a certain period of time after placing the reactor in the RUN mode before the primary containment atmosphere oxygen limits have to be maintained. What is the oxygen limit and how long de you have to establish conditions within the limit?
(1.0)
buESTION 7.08 (2.50)
You are in the control room and observe the conditions given below.
If the condition is normal, so statei if the condition is not normal, state any procedures you would immediately enter. Answer each item seperately. Assume no other conditions, and all givens valid.
a.
RPV level at -5 inches, reactor shutdown CO.53 b. Power at,16%, STARTUP MODE.
[0.53 c.
Drywell temperature at 135 F.
CO.53 d.
Suppression pool level on LI-1001-604A is 115 in.
CO.53 e.
Drywell pressure is 2.6 psis, reactor shutdown CO.53 (***** CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
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QUESTION 7.09 (2.50)
Given the following parameters:
Drywell temperature = 180 F 15 psig Drywell pressure
=
100 F Guppression pool temperature
=
12.5 psis Suppression pool pressure
=
Refer to the sttached Drywell Spray Initi_ation Pressure Limit Curve f rom E0P-5 and answer the following.
a. Can drywell spray be initiated?
E0.53 b.l:Can suppression pool spray be initiated?
[0.53 c.
What could be the possible consequence and explain how this consequence could develop when existing conditions are in an area of the attached E0P-5-curve where drywell spray
'
is prohibited and drywell sprays were initiated?
[1.53 OUESTION 7.10 (1.50)
Determine whether the following statements concerning
'A'
priority Radiation Work Permits (RWP) are TRUE OR FALSE.
a.
The Watch Engineer may designate an RWP as
'A'
priority to ensure safe plant operations.
(0.5)
6. The Watch Engineer shall specify the survey and survie11ance requirements on an
'A'
priority RWP.
(0.5)
c.
As a minimum the ALARA Engineer must approve an
'A'
priority RWP prior to issue.
(0.5)
,
.
QUESTION 7.11 (2.00)
During an ATWS with vessel level unknown, E0P-8, RPV Power Control By Level, directs you to control power as close as practible to, but above, a' specific minimum level. What is this minimum power level and why is this limit established (three reasons) ?
(2.0)
!
(*****
END OF CATEGORY 07
- )
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
=
'
GUESTION 8.01 (2.00)
_
According to 10 CFR 50, the reporting requirement of ONE HOUR for the following conditions is applicable. Determine if this statement is TRUE or FALSE for each condition given below.
A..The plant is in a condition not covered by operatin3 and emergency procedures.
(0.5)
!
B.
The loss of the offsite notification system.
(0.5)
C.
A valid automatic initiation of the Reactor Protection System (RPS).
(0.5)
D.
Aishutdown was commenced because the plant was in violation of'the Technical Specifications.
(0.5)
GUESTION 8.02 (3.00)
Consider the following los entries 900 am Power at 100%. Drywell to suppression chamber dp=1.19 psid.
1000 am Su.veillance test on drywell to suppression chamber vaccum breakers started 1011 am Drywell to supp. chamber dp=1.10 psid 345 pm Vaccum breaker surv. test of 1000 am entry complete drywell to supp. chamber dp=1.11 Psid 425 pm Core spray surv. test complete 1031 pm Drywell to supp. chamber dp= 1.12 psid Are conditions satisfactory? Explain your answer using the attached TECHNICAL SPECIFICATIONS and describe any actions you would take.
Fully reference any applicable sections used to develop your answer.
(3.0)
.
GUESTION 8.03 (3.00)
A.
What three indications are available to determine whether or not a jet pump has FAILED?
(1.5)
B.
If a jetpump is found to be inoperable, tech specs require the reactor to be placed in cold shutdown. What is the basis for this requirement?
(1.5)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
QUESTION 8.04 (2.50)
_
-
You are in the control room as the Watch Engineer on the backshift; power is at 100%. A large leak occurs in the
' common header" steam piping at the point it connects to the
'D'
The reactor is scrammed and all MSIV control switches are placed in close. At this point you determine that the
'A'
inboard and outboard MSIV's are STUCK open.
a.
Refer to the Emergency Action Level chart attached and determine the EAL, if any, and justify your decision.
[2.03 b.
Answer TRUE or FALSE: You can upgrade the event as warranted J
bht NRC concurrence is required to downgrade the event.
[0.53
GUESTION 8.05 (2.25)
,
Concerning tagging procedure 1.4.5 c.
Selected Red Tags may be temporarily removed for what purpose?
E0.753 b.
What does a white tag with a green border mean when attached to a 345kv component?
CO.53 c.
You have determined the continued operation of a pump would result in the introduction metal pieces into the feedwater system and damage the feedwater pumps. The maintenance supervisor says he will not be ready to repair the pump for three days.
What type of tag would you select to tag the equipment and j
explain why?
C1.03
.
l GUESTION 8.06 (2.00)
A temporary change implemented by SRO may be made provided the l
change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license on the unit-affected.
1.
WHO is considered " Plant Management Staff'?
(1.0)
2. What two other requirements.are necessary to implement a temporary change?
'
(1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16-
__________________________________________________________
.
QUESTIO.4 8.07 (3.00)
During refuel operations on the 12:00am to 8:00am shift, you are informed by the I & C supervisor that the
'B'
SGTS train failed to meet the differential pressure operability requirements on the
~
filters during surveillance testing. Using the attached copy of the TECHNICAL SPECIFICATIONS determine:
1.
Can refuel operations continue?
2.
If not why not? If so under what conditions?
Fully reference all applicable sections used to develop your answer.
C3.03
QUESTION 8.08 (3.00)
During 100% power operations on the 4:00pm to 12:00am shift, you are informed that LIS-72C is reading greater than +50 inche,s (offscale high).
Using the attached instrument list and copy of the TECHNICAL SPECIFICATIONS determine what actions must be taken?
Fully reference all applicable sections used to develop your answer.
E3.003 QUESTION 8.09 (3.00)
The reactor is soberitical and cold (less than 212 F). The mode switch is in REFUEL and control rod worth testing is in progress.
Control rod,26-31 is fully withdrawn; all other rods are inserted.
' Using the attached copy of the TECHNICAL SPECIFICATIONS answer the following questions. FULLY REFERENCE ALL APPLICABLE SECTIONS USED IN D$TERhINING YOUR ANSWER.
a.
In addition to the IRM high flux channels, what four (4) RPS trip functions channels need to be operable?
[1.253 b.
During an audit of maintenance procedures it is discovered that the IRh high flux trips on channels A, B,
C, and H were improperly calibrated such that the trip setting is higher than the Tech Spec limit. What actions per Tech Specs are required?
[1.753 (*****
CATEGORY 00 CONTINUED ON NEXT PAGE
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
_
_
OUESTION 8.10 (1.25)
Regarding procedure 1.3.6 Adherence to Technical specifications You are on watch during a weekend and in a situation where you may be in less than verbatum compliance with the Tech Specs. What five (5) actions must be performed to clairify the Technical Specifications?
(include all documentation)
(1.25)
.
(***** END OF CATEGORY 08 xxxxx)
.
(***xxx******* END OF EXAMINATION xxxxx**********)
.
,
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l 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
--_----_---_-_
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 5.01 (2.50)
a. U-258 (0.5)
b.
Decrease (0.25) As Pu-239 production increases (0.25), and U-235 decreases (0.25) the core average will decrease due to Pu-239's Beta being so much smaller (0.25).
(1.0)
.c.
80L(0.25) As BETA decreases the contribition to reactor period also decreases (from transient period equation).
Since BETA is larger at BOL the contribution to the neutron cycle from delayed neutrons is greater and period is longer.(0.75) (1.0)
REFERENCE G.E.
Reactor Theory, ch.
3, pg. 3-30, 3-36.
ANSWER 5.02 (2.75)
(
a.
Doppler, Moderator Temperature, Void (.75)
b.
As the rods are withdrawn, power level increases. But the additional heat generated is not immediately transported to the coolant. The fuel time constant of 8 or 9 seconds slows the rate that the heat generated in the fuel is conducted into the coolant.(0.5) It takes about 3 time constants or 30 seconds for the total increase in heat generated to be transported to the coolant. So the fuel temperature rises first, causing the doppler to be the first effect.(0.5) The next effect would be the moderator temperature, as the coolant is heated to saturation.(0.5) Finally comes the effects of voids, as the heated generated in the fuel boils the water flowing
'
thrnugh the core.(0.5)
REFERENCE Rx Theory, section 26 G.E.
R euW Theory, CL. 4,
G.E. Ncal TL WJFC' AM '**'o FL*% CA 4 9.WL
,
%
.
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.
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 5.03 (3.00)
a.
increase [0.253, as the pressure of the system decreases the flow increases due to the centrifugal pump head / flow characteristic. [0.753 b. decrease EO.253, as the pressure of the system decreases the operating point on the pump characteristic curve is shifted to a lower pump discharge pressure. [0.753 c.
increase E0.253, from the pump characteristic curve as'the flow (capacity) increases the power requirements also increase. [0.753 REFERENCE G.E.
Heat Transfer and Fluid Flow, Ch. 6 pg. 6-95 & 6-96.
ANSWER 5.04 (2.00)
a.ed.
(0.5 given for each correct choice, i.e.
full credit given for above, but if any incorrect answers are given,-0.5 will be given)
REFERENCE G.E.
Reactor Theory, ch. 3 pg.
3-8, 3-9, 3-12.
ANSWER 5.0,5 (2.75)
a. LHGR E0.503 b.
APLHGR.or HAPLHGR E0.753 c.
FM-fuel clad cracking due to differential expansion of the pellet and the cladding E0.53 LC-pin power is limited to prevent greater than or equal to 1% pl~astic strain on the cladding.
CO.53 d.
Transition boiling occurs which can result in clad cracking due to loss of adequate cooling.
CO.53 REFERENCE G.E. Heat Transfer and Fluid Flow, ch.9, pg. 9-15, 9-18, 9-19
.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 5.06 (2.00)
Roverse power effect E0.5]. Caused when withdrawl of'a shallow rod raises power low in the bundle-[0.503 which causes an increase in the void content in the upper _ region of the bundle that adds greater neg. reactivity in the upper part of the bundle than pos. reactivity added by the rod E.4503. Thus total power in the bundle decreases EO.5].
,,
,es ts:Z:J REFERENCE G.E. Reactor Theory, ch.5, pg. 5-25.
ANSWER 5.07 (1.50)
A.
Increases (0.25)due to the control valves going shut in response to the MPR becoming the controllin3 signal. (0.25)
B.
Increases (0.25)due to the reduction in the void content of the two phase mixture in the core. (0.25)
C.
Increases (0.25)due to the collapse of voids from the higher pressure which adds positive reactivity.
(0.25)
REFERENCE PNPS LP-MHC, Hechanical Hydraulic Control System, pg. MHC-10-1 G.E.
Reactor Theory, ch.
4, pg. 4-24 G.E.
Heat Transfer and Fluid Flow, ch. O, pg. 8-41.
ANSWER 5.08 (1.00)
1.
fission product poisons ( may break down to Sm a Xe)
[0.333 2.
moderator temperature coefficient E0.33]
3.
fuel temperature coefficient E0.33]
(Other factors will be considered if justified)
REFERENCE G.E. Reactor Theory, ch.7, pg.
7-6.
.
--aw-t
.
i
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
_____ _ _
______ ______________________________________
~
______________
ANSWERS -- PILCRIM'
-86/05/06-HOWE, A.
ANSWER 5.09 (2.00)
a.
Rod B (0.25) Upon scram recovery, fission product poisons cause a severe flux depression in what was the highest power producing region of the core. This results in a higher relative flux in regions of low poison concentration. These shifts in the flux distribution increase the worth of peripherial rods and decrease the worth of those in the center of the core.CO.75T b.
The withdrawn rodCO.25] Flux is higher in this area, thus rod worth is greater.[0.753
<
REFERENCE G.E. Reactor Theory, ch.5 pg. 8,9,18. ch. 6 pg. 12 ANSWER 5.10 (2.50)
1.
Reference les flashing CO.25] produces indicated level higher than actual [0.253. Develops under conditions of heated reference less undergoing rapid depressurization below 500 psis causing a lower reference les thus a lower dp and higher indicated level.[0.753 2.
Elevated cold reference les temperaturesEO.253 produces indicated level higher than actualEO.253. Develops under conditions of elevated
,'
drywell temperatures causing a lower density in the reference les thus a lower dp and higher indicated level.[0.753
,
REFERENCE
,
.
pg.35,36 EOP-1 ANSWER 5.11 (3.00)
-
a.
Adds negative reactivity [0.253 due to the increase in neutron leakage. Moderator temperature coefficient. CO.503
,
b.
Adds negative reactivity CO.25] due to the increase in neutron capture in the fuel - Doppler coefficient. [0.50]
c.
Adds positive reactivity Co.25] due to the decrease in neutron leakage - Moderator temperature coefficient. [0.50]
d.
Adds negative reactivity CO.25] due to the increase in neutron 1.
leakage - Void coefficient. [0.50
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
____ggggg7g ggg______________________________________
______________
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
.
REFERENCE G.E. Reactor Theory ch.
4, pg.
8, 9,
16, 24, 34, 35, 37.
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
.
ANSWER 6.01 (1.50)
- local manual trip (keylock at entrance to D/G room)
n manual trip i.- cc..u s!
- (keylock switch to S/D C-103/C-104)
n engine overspeed (1070 rpm)
_-m
,__,__.1 i_:_
5 int I h?> NYNclIN on D/Cr en3:n e { $*?f'
- * *
O'
'
t (0.375 each)
REFERENCE PNPS Diesel Generator LP, pg. DGS-33-1.
SLO-7 ANSWER 6.02 (3.00)
L.
condensor low vaccum alarm, 26' HG; warns operator of problem.
2.
reactor scram, 23' HGi anticipates scram which would be a result of the turbine trip.
3.
Iow vaccum turbine trip ($1 vaccum trip), 20' HGi protect T/G and reduce heat input to the condensor.
4.
lo-lo vaccum bypass valve trip (42 vaccum trip),
7'
HGi protect condensor from overpressure condition.
(0.25 for each setpoint, 0.50 for each reason)
- CAF FOR-REASONS ****
REFERENCE PHPS Main Condensor and Air Removal LP, pg. MCV-13-1, MCV-14-1; SLO-3
.
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
ANSWERS -- PILGRIM
- -86/05/06-HOWE, A.
ANSWER 6.03 (3.00)
a. Maximum time * 125 minutes CO.53 fast enough to overcome any positive reactivity added due to cooldown following the xenon poison peak.[0.753 (1.25)
b.
1.
loss of squib valve continuity annunciater 2. squib valve ready light out on selected system 3. selected pump motor running light (red) is lit 4. RWCU isolates 5. pump disch. press. > Rx press 6 '. R:' power decreasing 7.'SLC tank level decreasing (0.25 each, total of 1.75)
REFERENCE PNPS Standby Liquid Level Control Lesson Plan pg. SLC-4-1,SLC-11-1 SLO-7, SLO-9.
ANSWER 6.04 (2.50)
1.
Feedwater Valve controls 2.
Recirc MG-set controls 3.
Recorders on pnl. 905 lost 4.
Process Computer is lost 5.
Rx vessel level and pressure i ndicators on pnl. 905 lost 4.
RPIS indication is lost 7.
SJAE (stack gas) isolation valves close 8.
Rod select relays and rod withdrawl permissive lost
-
(5 required at 0.5 each)
(2.5)
REFERENCE PNPS 120/240 VAC LP, pg. 120 V-30-1. Proc.5.3.6 Loss of Vital AC, pg.
5.3.6-2.
SLO-4.
.
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
.
-86/05/06-HOWE, A.
ANSWERS -
PILGRIM
-
^
.
ANSWER 6.05 (2.50)
s.
1.
scram valves.would ope.1 under spring pressure and control rods would be inserted (0.33)
2.
SJAE steam supply valve closes (0.33)
t,3, g gy7xc,g;g 3.
feedwater regulatin3 valves fail as i s (0.33)
couu u g g
s b.
YesCO.253 t c i r d u r+
y te: =:11 ir.itirt cr cute cera: N y,,w'
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si y, 1 E2t 55 ;m s ir th: :1r h==ri=% CO.503
~
2 c.
Reset by placing the scram discharge volume isolation switch 3 "4j in. ISOLATE which will allow air to reposition the air dump valve.CO.753 REFERENCE PNPS INSTRUMENT AIR LP, PG. AIR-14-1, AIR-15-1. SLO-7 PNPS RPS LP, pg.
6.
SLO-3 s
ANSWER 6.06 (3.00)
a.
1.
acoustic monitior indicating lights on panel c-171 2. R/V temperature recorder on the back panel would increase 3. water level should be dropping - 1 ft/ min 4.
torus level would be oscillating 5.
torus temperature would increase 6.
drywell pressure would increase (5 required 0.25 each)
b.
1.no(.25), As long as there is a valid high drywell pressure signal the high drywell pressure contact will not open and the ADS timer not reset.(.5)
(.75)
2.
If the timer has not timed out, the timer will deenergize and the ADS will not initiate.(.5) If the timer has timed out, then ADS has initiated and will continue to operate even with level restored.(.5)
(1.0)
~
REFERENCE PNPS ADS LP, pg. ADS-9-1 to ADS-11-1. SLO-5.
PNPS Proc. 2.4.29, Stuck Open Safety Relief Valve, pg. 2.4.29-2.
,
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
_
ANSWERS -- PILGRIM-
-86/05/06-HOWE, A.
.
.
_
ANSWER 6.07 (2.00)
a.
2.5 on range 7.
No automatic action (downscale at 2%).
(1.0)
'
b.
39 on range 5.-
IRH hi3h rod block and IRH high-high half scram.
(1.0)
REFERENCE PNPS LP, IRM, pg. IRM-12-1, IRM-13-1 Figure 8.
SLO-4.
I ANSWER 6.08 (3.00)
a.
Allows for valve testing during operations while protecting the system low pressure from high pressure primary coolant.
(1.0)
b.
No. The manual close signal overides the the auto open signal until the original automatic initiating signal is cleared.
(1.0)
c.
CS line between the outside of the core shroud to ch?ckvalve in drywell (assuming no seat leakage; if seatleakage assumed, j
up to the inboard isolation valve)
(1.0)
REFERENCE PNPS LP Core Spray, pg. CS-11-1, CS-14-1, CS-15-1. SLO-2, 8.
9
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l 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
.
-
.
ANSWER 6.09 (3.00)
a.
'B'
feedwater line flow signal fails high and causes an increase in the total feed flow signal. A feed flow / steam flow mismatch is generated which causes feedwater control valves to close. Since steam flow is now greater than actual feed flow, reactor vessel level decreases.
(1.0)
b.
( Channel
'A'
steam pressure detector fails low and produces a lower density compensation signal to the steam flow proportional amplifier.) This lower signal results in a lower steam flow output to the steam / feed flow comparator, which results in a
'
fee'd flow > steam flow mismatch, and causes feedwater control
!
valves to close. Reactor vessel level decreases since actual steam flow is greater than feed flow.
(1 0)
c.
Loss of signal to a feedwater valve causes the valve to receive a lockup trip. The
'A'
FRV will continue to control level, there-fore reactor level will remain the same.
(1.0)
(.25 for response,.75 for reason)
!
REFERENCE PNF'G RVLC LP, pg. FWLC-5-1, FWLC-7-1, FWLC-8-1, FWLC-11-1.5LO-4, 11.
t i
ANSWER 6.10 (1.50)
,
a.
1.
Grapple 56 ft. below the refueling platform tracks.
2. Grapple hoist load greater than 1200 lbs.
3.
Grappl.e hoist less than 50 lbs.
(0.33 each)
b.
Provide redundant methods of preventing inadvertant criticality even after procedure violations.
(0.5)
REFERENCE i
PNPS LP REFUEL SYSTEM, pg. RF-7-1, RF-27-1. SLO-2, 3.
,
t
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,
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ 5656LUU5UAL C6UTR6L
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AN_SWERS -- PILGRIN-86/05/06-HOWE, A.
ANSWER 7.01 (2.50)
c. Monitor alarms and instruraentation and determine the type of system malfunction that has occured.
[0.53 If the vapor valves have closed, reduce reactor power and scram the reactor.
[0.53 b. reactor building process radiation monitor E0.53 c. If isolation valve is open purge the off gas system through the main stack vent.
E1.03 REFERENCE PNPS Proc. 2.4.55 Augmented Off-gas Explosions, pg.2 ANSWER 7.02 (3.00)
increasing RBCCW nd TBCCW temps.
a.
'
m loss of service water pressure
- loss of service water flou (0.5 each)
b.
1.
scram 2.
Trip the turbine 3.
isolate cleanup system 4.
shutdoun the recire pumps 5. after recovering reactor level to normal (after scram)
shutdown all feedwater and all but one cond. pump 6. place iuel pool cooling at maxiraum flow to for a
'
-
temporary heat sink'to RBCCW 7. reduce RDCCW to one pump per loop and TDCCW to one pump 8. use RCIC to feed reactor as needed (four required at 0.25 each)
I c.
(Inability of RBCCW and TBCCW to provide adequate cooling of ncrmal' heat loads thus vital equipment will overheat)
Quick action needed in anticipation of the loss of vital reactor plant equipment as a result of overheating [1.03 REFERENCE PNPS Proc. 5.3.3 LOSS OF ALL SERVICE WATER, pg.
2, 4.
.
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.
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
_
RA5i5t65iEEL E5UTR6L~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
-
AN_SWERS -- PILGRId-86/05/06-HOWE, A.
ANSWER 7.03 (3.00)
s. mex 50 des. F delta T'
between loops CO.253
'
cold water from loop can cause power transient and possible.
scram on hi finx E0.503 max 145 deg. F delta
'T'
between vessel dome and bottom head CO.253 cold water shock could cause damage to CRD stub tubes CO.503 b.
1.
Under normal starting the motor may be started twice successivly from ambient temperature CO.25] or once from
' rated motor temp.[0.25] Time between restarts at rated temp.
is 15 min. running or 45 min. idle.CO.503 2. Eact h-G set start stresses the drive motor windings both thermally and mechanically. [0.53 REFERENCE PHPSrPROC. 2.1 9 RECIRC PUMP OPS., pg ?; PROC. 2.2.84 RECIRC SYSTEM,pg 15.
ANSWER 7.04 (3.00)
a.
unidentified - 5 spm max.
[0.503 total of unidentified + identified - 25 opm max.
[0.503
-
(admin limits of 2.2.77 considered if specifically stated)
b.
x excessive sump pump operation of either drywell equipment
drain sump or drywell floor drain sump
- abrupt. change in drywell humidity
'
significant change in drywell temp or pressure (two separate ans)
E0.33 each3
,
c.
1.
drywell temp.> 194 F at > 40 ft el.
2.
drywell temp. > 150 F at < 40 ft el.
[0.50 each]
l REFERENCE PNPS proc. 2.4.14 LEAKS INSIDE THE PRIMARY CONTAINHENT, pg 2;
'
proc. 2.2.77 DRYWELL LEAK DLTECTION SYSTEMS, pg 7,
.
.
t
!
,
y,,,.. ~. - -,. -,,-
-.,--.,,--e
,
- -
,.,,,
,..--..m--y,.-_,,.,c-,-..-_._m,
-r_
.
_.
.
.
~
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
R5656L66 6AL 66 TREL
~
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
____________________
AN,SWERS -- PILGRIM-86/05/06-H0HE, A.
ANSWER 7.05
'(2.00)
1.
restore and maintain RPV level above TAF and below +48*.
2.
control RPV pressure and cooldown the RPV to cold conditions 3.-maintain core cooling ( to prevent excessive clad heatvP and oxidation)
P ace reactor in a safe stable condition l
4.
(0.50 each)
REFERENCE
PNPS E0P-1 RPV CONTROL, pg 2 ANSWER 7.06 (1.00)
a.
Two operable SRM's reqd. One in quadrant where fuel being loaded one in an adjacent quadrant.
(1.0)
REFERENCE PNPS T/S 3.10 8.
ANSWER 7.07 (2.00)
a.
Whenever the reactor is critical or when the Rx water temgerature is above 212 degrees and the head vent closed. (1.0)
b. p%EO.51,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (0.5)
4 vo or r/5 39.
4 om,4.
REFERENCE
PNPS Procedure 2.1.1 Rev. 46 pg. 4
,
!
ANSWER 7.08 (2.50)
a.
E0P-1 b.
c.
normal d.
EOP-6 e.
E0P-1, E0P-5 (0.5 each a-d, 0.25 each e)
,
f
.
f G
-
'
4 - _ _.., _,. _.
. _ _ _.. _ _
,
-
, _,
,
. _., _ _ _, *.
m.,.
,_,.., _..,,,,, _ _ __
-
.-
- -.
,
.
-
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~
~~---
R 655EUU55IL C6UTR5L
~~~~
____________________
_ ANSWERS -- PILGRIM-86/05/06-H0HE, A.
REFERENCE PNPS E0P's ANSWER 7.09 (2.50)
a. no CO.53 i
b.
yes
[0.53 c.
In this condition there could be insufficient noncondensable gases in the drywell and the initiation and continuation of drywell spray could create negative pressures greater than the capacity of the vaccum breakersEO.753 resulting in potential destructive containment negative pressures.[0.753 REFERENCE PNPS E0P-5, pg. 11,15, 30.
,
ANSWER 7.10 (1.50)
a.
true b.
false c.
false (0 5 each)
REFERENCE PNPS Procedure 6.1-022, pg.
3,5,6.
ANSWER 7.11 (2.00)
- 8% min power
.
- Ensures adequate baron mixing in the core
'n Ensures RPV level above TAF e
- Hold as close to 8% to minimize supp. pool heatup i
(0.5 each)
REFERENCE PNPS E0P-8, pg. 27, 28.
.
,
4 w-y e
v>-
-y----nn m
r..-s-a
.---u
.-,,----------m.----
-. -. - - - - - - +
, = -
n
-..-..,.-n~.
- -.-- -
.
-
._
,.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
________________-_________________________________________
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 8.01 (2.00)
A.
true C.
false B.
true D.
true (.5 each)
(2.0)
REFERENCE 10 CFR 50.72 ANSWER 8.02 (3.00)
Conditions are not satisfactory.[0.503 Section 3.7.k 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit for test'was not fo11 owed.[0.53 Section 3.7.1 to recover dp in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not fo11 owed.[0.53 Per 3.7.1 must initiate an orderly shutdown and be in a cold condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.[1.503 REFERENCE PNPS T/S 3.7.A i,j,kel ANSWER 8.03 (3.00)
A.
1.
Recire pump flow 2.
Total core flow 3.
Diffuser-to-lower Plenum dp (3 0 0.50 ea)
(1.5)
B.
In the case of DBA LOCA, the blowdown area is increased and (0.75)
'
i the capability for reflooding the core is reduced.
(0.75)
'
REFERENCE PNPS Tech Spec 3.6.E.1, BASES 3.6.E.,LP Reactor Recire System pg 1.
SLO-11
.
ANSWER 8.04 (2.50)
a.
Site Emergency [1.003 Main Steam line rupture outside primary containment that is not isolable.[1.003 b.
FALSE [0.53 REFERENCE PNPS Procedure 5 7.1 pg.4, 5.7.1.1 pg.3, EAL Chart
!
,
,,--,--,,n--,,w--
,_- _--- -
8.
' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 8.05 (2.25)
a.
To verify that a particular maintenance activity has been satisfactorily completed, when.it is impractical to completely clear the maintenance request.
[0.753 b.
hold CO.53 c.
Orange N.W.E tas E0.53 to protect equip. from damage when there is no accompanmying maintenance request.[0.53 REFERENCE PNPS. Proc.1.4.5 pg.
6, 7,
8.
ANSWER 8.06 (2.00)
1.
Any Boston Edison Management (non-union) Person permenantly assigned to Pilarim.
(1.0)
,
2.
i. The intent of-the original procedure is not changed.
ii. The change is documented and reviewed by ORC and approved with in 7 days.
(1.0)
REFERENCE PNPS procedure 1.3.4 pg 10.
ANSWER 8.07 (3.00)
Yes operations can continueEO.53 Per 3.7.8.e.
two trains shall be operable dur,ing fuel handling except as allowed by 3.7.B.1.c.
[0.53 Per 3.7.B.L.c.
fuel handling operations can continue with one train
' inoperable for the next 7 days provided that within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and daily thereafter all active components of the other SBGT train shall be demonstrated operable.[2.03 REFERENCE PNPS T/S 3.7.B.1
.
%
.
!
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- PILGRIM-86/05/06-HOWE, A.
ANSWER 8.08 (3.00)
(LIS-72C is inoperative thus) action of T/S table 3.2.B note (1) [0.53 applies.The CSCS equipment is required to be operable per sect 3.5 and minimum number of trip systems requirement is not met [1.03 thus LIS-72C shall be repaired a
or the reactor shall be placed in cold shutdown within 24 hrs.[1.53 REFERENCE
.
PNPS T/S 3.2.8 Table 3.2.8.
ANSWER 8.09 (3.00)
a.
1.
mode switch in shutdown
!
2. manual scram 3.
scram disch vol. hi level
'
4.
APRM 15% high flux scram (0.25 each)
I Per(note 7) of T/S table 3.1 1.
(0.25)
b.
Per table 3.1.1 note 1 since both channels have less than the minimum required number of operable IRH trips E0.753 insert all rods and complete insertion uithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required per note 1.A.
[1.03 REFERENCE PNPS T/S 3.1 table 3.1.1 ANSWER 8.1,0 (1.25)
- Call the Nuclear Operations Manager a Discuss.the situation
- Obtain verbal concurrence a Document conversation on a std. telephone call record
- Issue a failure and malfunction report (5 reqd. at 0.25 each)
REFERENCE l
PNPS procedure 1.3.6 pg. 2
.
$
s
+.. - - +
c.
~ - -,,.
.--
.---ms,
,, _ _ _, -
-.-t---
--.
2._
u
-_ _ - ~.
ww
- -, - - - -
%,._em.
,w-.4_
w
-.-
-,r,-.
- - -
-w v,,,----,,,--
-
. - -.
.
.- - -
--
-.
..
.
TEST CROSS REFERENCE PAGE
'
GUESTION VALUE REFERENCE
________. ______
__________
05.01 2.50 AXA0000062
,
05.02 2.75 AXA0000219 05.03 3.00 AXA0000220 05.04 2.00 AXA0000221 05 05 2.75 AXA0000222 05.06 2.00 AXA0000223
.
05.07 1.50 AXA0000224 05.08 1.00 AXA0000225 05.09 2.00
.AXA0000259
,
05.10 2.50 AXA0000260 05.11 3.00 AXA0000261
______
25.00
06.01 1.50 AXA0000226
'
i 06.02 3.00 AXA0000227 06.03 3.00 AXA0000228 06.04 2.50 AXA0000229 06.05 2.50 AXA0000230 06.06 3.00 AXA0000231 06.07 2.00 AXA0000232 06.08 3.00 AXA0000233 06.09 3.00 AXA0000234 06.10 1.50 AXA0000235
______
]
25.00
,,
07.01 2.50 AXA0000236
'
07.02 3.00 AXA0000237 07.03 3.0,0 AXA0000238 07.04 3.00 AXA0000239
!
'07.05 2.00 AXA0000240 l
07.06 i.00 AXA0000241 07.07 2.00 AXA0000244
,
07.00 2.50 AXA0000245 07.09 2.50 AXA0000246
'
07.10 1.50 AXA0000257 07.11 2.00 AXA0000258
______
25.00 00.01 2.00 AXA0000083
,
00.02 3.00 AXA0000247 I
00.03 3.00 AXA0000248 00.04 2.50 AXA0000249 08.05 2.25 AXA0000250
!
08.06 2.00 AXA0000251 08.07 3.00 AXA0000252
!
08.00 3.00 AXA0000254 00.09 3.00 AXAOOOO255
,
__ ___
-,. - _,. _ _ _ _ _ _,
_. _ _ _ _ _ _ _, _ _... _ _,,. _,.., _,.. _ _ _ _.. _ _ _,, _, _ _,,, _ _.. _ _ _,. -,. _.. _
.. _ _
.
_
.
TEST CROSS REFERENCE PAGE
OUESTION VALUE REFERENCE
________
______
___---_---
08.10 1.25 AXA0000256
______
25.00
______
______
100.00
\\
.
6
%
_ _ _
.. -
a.,
,,
.s.
...
-
....
_..
.
-. -
.
.
'
EQUATION SHEET
,
.
,
_
f = ma v = s/t Cycle efficiency = (Ne:wcet
~ 2
.
out)/(Energy in)
-
w = mg s = V t + 1/2 at o
E = mc
KE = 1/2 mv a=(vf - v,)/t A = xN A = A,e'**
PE = mgh V
V + at w =e/t x = an2/t
=
0.692/t f
o
=
1/2 1/2 MPSH = P,- P,g 1/2'
I III}3 t
"
j
'
1/2 b
b I*I I
1/2 b
.
maoAV
~
AE = 931 am I = I e-Ex Q = Ecat
.
Q = UAah I = I e~"*
n I'= I, 10-*/IY'
Pwr = W ah f
TVL = 1.3/u-sur(t)
-
P = P,10 Hvt = -0.693/u p = p e /T t
o SUR = 26.06/T SCR = S/(1 - K,ff)
CA, = S/(1 - K,ff,)
SUR = 25 /t= + (a - c)T CR (1 - K,7fj) = CR (1 - keff2}
j
T=(t=/o)+((a-c)ho)
M = 1/(1 - K,ff) = CR /CR j
T = t/{s - a)
-
M = (1 - K,ff,)/(1 - K,f,3;
-
T = (a - o)/(:o)
SOM = (1 - KdfII* d'
'
,
= ( K s. s. -l ) /K,s. s. = AK, e. s. /Kef t' = 10~* seconds o
e
.
.
1 = 0.1 seconds '
-
o = ((t=/(r Kdf!2 ' E*eff/II'*II)
1e3 3 = I e>
2=2[22 P = (tev)/(3 x 1010)
Id d
j
t = oN R/hr = (0.5 CE)/c (meters)
j NPSH = Static head - h3-P R/hr = 6 CE/d2 (feeti
!
Wa:er Parameters Miscellaneous Conversions I gal. = 8.345 lbc.
I curie = 3.7 x ICICcos 1gaj.=3.78 liters 1 zg = 2.21 1:m,
1 ft- = 7.48 ga.
,i no = 2.5: x 104 S tw nr
.
Density = 52.4 lo../ f t '
I mw = 2.41 x 106 8tu/hr Dens::y = 1 ge/cmJ
!in = 2.54 cm Heat of vaport:ation = 970 Stu/;be
=F = g/i=C + 32
- eat of fusien = 14' 5tt./1Dm
=C = 5/9 '=F-32}
I atm = 14. 7 ;;si = 2 9. i in, ng.
'
-
._
.. -
_
_
._
_
-
. _..
. _ - _
(
.g
~.
-
-
.
...
.
.
E
,
Table 1.
Saturated Steam: Temperature Table Abs Press
~ ~Ipecific Volume Enthalpy Entropy Temp I.b per Sal.
Sat.
Sat.
Sat.
Sat.
Sat.
Temp rain Sq In liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr a4 t
p vg vrg vg hg h gg hg sg sig s
t
'
g
'
37 8 0 08859 0 0160??
33047 3304 7 0 0179 1075.5 1075.5 0.0000 2.1873 2.1873 32.8
'
34 8 0 09600 0 016021 30619 3061.9 1.996 1074.4 1016.4 0.0041 2.1762 2.1802 34.0
36 0 0 10395 0 016020 2839 0 2839 0 4 008 1073.2 1077.2 0.0081 2.1651 2.1732 36.0 31 0 011749 0016019 7614 1 7634?
6 018 1072.1 1078.1 0.0122 2.1541 2.1663,
38.0 48 8 112163 0 016019 2445 8 2445 8 8 027 1071.0 1079.0 0.0162 2.1432 2.1594 40.9 47 0 0 13143 0 016019 2277 4 / 2272.4 10 035 1069 8 1079.9 0.0202 2.1325 2.1527 42.0
.
,
44 0 0 14192 0 016019 2112 8 2112 8 12.041 1068.7 1080.7 0.0242 2.1217 2.1459 44.0 i
l 46 e 0 15314 0 niS020 19657 19657 14.047 1067.6 1081.6 0.0282 2.1111 2.1393 45.0 48 0 0 16514 0 01K071 IR.in 0 '
1810 0 16 051 1066 4 1082.5 0.0321 2.1006 2.1327 48.8 58 0 0 17796 0 016023 1704 8 1704 8 18 054 1065.3 1083.4 0.0361 2.0901 2.1262 50.0 57 0 0 19165 0 016024 15892 1589 2 20 057 1064.2 1084.2 0.0400 2.0798 2.!!97 52.0 54 0 0 20625 0 016026 1482 4 1482 4 22.058 1063.1 1085.1 0.0439 2.0695 2.1134 54.0
'
51 8 0 22183 0 016028 1383 6 1383 6 24 059 1061.9 1086.0 0.0478 2.0593 2.1070 58.8 58 0 0 23843 0016031 1797 7 1292.2 26 060 1060.8 1086.9 0.0516 2.0491 2.1008 58.8 i
' IO O
~
025611 0 016033 1207.6 1207 6 28 060 1059.7 1087.7 0.0555 2.0391 2.0946 68.4
!
$7 0 0 27494 0 016036 1129 2 1129.2 30 059 1058.5 1088.6 0.0593 2.0291 2.0885 52.8 E4 0 0 29497 0 016039 1056 5 1056 5 32.058 1057.4 1089.5 0.0632 2.0192 2.0824 64.8 II O O31626 0016043 989 0 989.1 34 056 1056.3 1090.4 0.0670 2.0094 2.0764 68.0 l
El 8 0 33889 0 016046 976 5 926 5 36.054 1055.2 1091.2 0.0708 1.9996 2.0704 84.8
,
if I O36292 0 016050 868 3 868.4 38.052 1054.0 1092.1 0.0745 1.9900 2.0645 14.0 12 8 0 38844 0 016054 814 3 814.3 40 049 1052.9 1093.0 0.0783 1.9804 2.0587 72.5 l
14 0 0 41550 0 016058 164 1 764 1 42.046 1051.8 1093.8 0.0821 1.9708 2.0529 74.0
,
75 0 0 44420 0 016063 717 4 717.4 44 043 1050.7 1094.7 0.0858 1.%I4,f.0472 75.0 78 0 0 47461 0016n67 611 R 673 9 46 040 1049.5 1095.6 0.0895 1.9520 2.0415 18.0 II 8 050683 0 016072 633 3 633 3 48 037 1048.4 1096.4 0.0932 1.9426 2.0959 80.5 87 0 0 54093 0 016077 595 5 595 5 50.033 10473 1097.3 0.0969 1.9334 2.0303 82.5 14 0 0 57702 0 016082 560 3 560 3 52 029 1046.1 10982 0.1006 1.9242 2.0248 B4.8
,
III O61518 0 016087 227 5 527 5 54.026 1045.0 1099.0 0.1043 1.9151 2.0193 86.5
.
III O65551 0n16093 496 8 496 8 56 022 1043.9 1099.9 0.1079 1.9060 2.0139 88.0 et 0 0 69813 0 016099 4681 4681 58 018 1042.7 1100.8 0.1115 1.8970 2.0086 30.0 17 I O14313 0 016105 4413 441 3 60 014 104I.6 1101 6 0.1152 1.8881 2.0033 92.0 94 0 0 79062 0 016111 416 3 416 3 62 010 1040.5 1102 5 0.1188 1.8792 1.9980 94.0
>
98 8 0 84072 0 016117 392 8 392.9 64 006 1039 3 1103.3 0 1224 1.8704 1.9928 96.9
98 8 O R9156 00lr17.1 370 9 370 9 66 003 10382 11042 0 1260 1.8617 1.9876 98 0
-
.
_ _ _ _.
__
_ _
_.-_
.
.
P Ahs paets Spmlic Vnhrme Enthalpy Entropy
'
.
Temp lb per Sal Sal.
Sal.
Sal.
Sal.
Sal.
Temp I ala Sq in Iirpshi Ivap Vapor liquid Ivap Vapor liquid Evan Vapor Fahr I
p vi vig vg hl h lg h
sg sig s
t g
g i38 I O94924 0016130 350 4 350 4 67.999 1037.1 1105.1 0.1295 1.8530 1.9825 100.0 102 8 100789 0016137 331 1 331 1 69 995 1035.9 1105.9 0.1331 1.8444 1.9775 102.0 134 8 106965 0 016!44 313 1 313 1 11.992 1034.8 1106.8 0.1366 1.8358 1.9725 IKO 136 3 11347 0 016151 296 16 29618 73.99 1033.6 1107.6 0.1402 1.8273 1.9675 100.0 108 8
!?030 0 016158 780 28 78030 75.98 1032.5 1108.5 0.1437 1.8188 1.9626 Int 118 I I2750 0 016165 26537 265 39 77.98 1031.4 1109.3 0.1472 1.8105 I.9577 '
118.0 112 e 13505 0 016173 251 37 25138 79 98 1030.2 1110.2 01507 1.8021 1.9528 112.0 114 I4799 0 016180 238 21 23822 81.97 1029I 1111.0 0.1542 13938 I.9480 114.0 IIII 1 5133 0 016188 225 84 22585 83.97 1027.9 1111.9 0.1577 1.1856 1.9433 118.0 III o 16009 0 016196 714 70 f 214 21 85.97 1026.8 1112.7 0.1611 1.7774 1.9386 110.0 111 I.
I6927 0 016204 203 25 203 26 87.97 1025 6 1113.6 0.1646 1.7693 1.9339 120.0 122 8 1 7891 0 016213 192 94 19295 89.96 1024.5 1114.4 0.1680 13613 1.9293 122.0 124 s 18901 0 016721 18323 183 24 91 %
1023.3
!!!5.3 0.1715 13533 1.9247 124.0 III e I9959 0 016229 174 08 174 09 93.%
1022.2 1116.1 0.1749 1.7453 1.9202 120.0 III e 71068 0 016718 16545 16547 95.96 1021.0 1117.0 0 1783 13374 1.9157 128.0 131 0 2 2230 0 016247 15732 157.33 97.%
1019.8 1117.8 0.1817 1.7295 1.9112 130.0 132 8 23445 0 016256 149 64 14966 99 95 10183 1118.6 0.1851 13217 1.9068 132.0 134 8 2 4717 0 016265 14240 142.41 101.95 1017.5 1119.5 0.1884 13140 1.9024 134.0 138 3
-
2 6047 0 016214 135 55 13557 103 95 1016 4 1120.3 0.1918 13063 1.8980 138.0 Ils a 27419 0 016784 179 09 129 11 105.95 1015.2
!!21.1 0.1951 1.6986 1.8937 138.0 148 8 28892 0 016293 12298 123 00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 148.8 142 0 3 0411 0016303 II7 21 111 22 109.95 1012.9 1122.8-0.2018 1.6534 1.8852 142.0 1440 3 1997 0 016312 Ill 14 111 16 111.95 10113 1123.6 0.2051 1.6759 1.8810 144.0 141 0 3 3653 0 016322 10658 106 59 113.95 1010.5 1124.5 0.2004 1.6684 1.8769 I40.0 14s e 3 5181 (1016317 10168 10130 115.95 1009.3 1125.3 0.2117 1.6610 1.8727 148.0 151 8 37181 0 016343 9705 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686'
1M0
,'
1528 3 9065 0 016353 9266 92 68 119.95 1007.0 1126.9 0.2183 1.6463.l.8646 152.t 154 8 4 1025 0 016363 8850 8852 121.95 1005.8 11273 02216 1.6390 1.8606 1541 tilI 4 3068 0 016374 84 56 84 57 123.95 1004.6 1128.6 0.2248 1.6318 1.8566 Int
,
158 8 45197 0 016184 80 87 R0 83 125 %
1003.4 1129.4 0.2281 1.6245 1.8526 158.0 i
186 8 4 7414 0 016395 7727 1729 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 100.0 152 8 4 9722 0 016406 13 90 73 92 129 96 1001.0 1131.0 0.2345 1.6103 1.8448 182.0 154I 52124 0 016417 70 70 70 72 131.96 999.8 1131.8 0 2377 1.6032 1.8409 104.0 Ill 8 5 4623 0 016428 6767 6768 133 97 998.6 1132.6 0.2409 1.5%1 1.8371 1Kt
-
18II 5 1723 0 01644u 64 78 64 80 13597 997.4 1133.4 0.2441 1.5892 1.8333 1K0 IIB I 5 9926 0016451 6204 62 06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170.0
.
112 I 6 2136 001R463 59 43 5945 139.98 995.0 1135 0 0.2505 1.5753 I.8258 172.0 1748 6 5656 0 016474 56 95 56 97 141.98 993.8 1135.8 0.2537 1.5684 1.8221 174.0 tilI 6 8690 0 016486 54 59 54 61 143.99 992.6 1136.6 02568 1.5616 1.8184 178.0 lie s 7 iR40 0016498 5735 5736 14599 991.4 1137.4 0.2600 1.5548 1.3147 175.0
~
'
r.
M.:
~
-
.
.
,.
I Abs Press Speofic Volume Enthalpy Entropy Temp lb per Sat Sat Sat.
Sal.
Sat.
Sal.
Temp
'
fahr Sq In liqmrt Evap Vapor Liquid Evan Vapor Liquid Evan Vapor Fahr I
p vi vrg vg he h gg h
5g sig sg i
g 1838 7.5110 0 016510 5021 5022 14800 900 2 1138 2 0.2631 1.5480 1.8111 180.4
,
182 8 7850
' 0 016522 48 172 18 189 150 01 989 0 1139 0 0.2662 1.5413 1.8075 182.8
'
184 8 8 203 0016534 46 232 46 249 152 01 937.8 1139.8 0.2694 1.5346 1.8040 184.0 188 8 8 568 0 016547 44 383 44 400 154 02 986 5 11405 02725 15279 I.8004 186.8 ise s 8 911 0 015 % 9 42A71 42 638 156 03 985 3 1141.3 0.2756 1.5213 1.7959 100.0 198 8 9 340 0 016572 40 94 /
40 957 158.04 984.1 1142.1 02787 1.5148 1.7934 198.8 192 e 9 747 0016585 39 337 39 354 16005 982.8 1142.9 02818 1.5082 1.7500 192.8 134 s 10 168 0 016598 37 808 37 824 162 05 981.6 11433 0.2848 1.5017 13865 194.0
~
198 8 10 605 0016611 36 348 36 364 164 06 980.4 1144 4 0.2879 1.4952 13831 196.0 lle O 11 058 0 016674 34 954 34 970 166 08 979.1 1145.2 0.2910 1.4888 1.7798 198.8 2?s s 11 526 0 01K637 33 622 33 639 168 09 977.9 1146.0 0 2940 1.4824 13764 200.1 i
l 204 8 12 512 0 016664 31 135 31 151 172.11 975 4 1147.5 03001 1.4697 1.7698 284.0 208 8 13 568 0 016691 28 862 28 878 176 I4 972.8 1149 0 0.3061 1.4571 1.7632 288.0 212 O 14 696 0 016719 26 182 26 199 18017 970.3 1150 5 0.3121 14447 13568 212.0
' 216 0 -
15 901 0 016747 74 R18 24 894 184 20 9678 1152.0 0.3181 1.4323 13505 216.0
- r "
228 O 17 186 0 016775 23 131 23 148 188 23 965.2 1153.4 0.3241 14201 1.7442'
220.0
'
224 8 18 556 0 016805 21 529 21545 192.27 962.6 1154.9 0.3300 1.4081 13380 224.0 225 0 20 015 0 016834 20 056 20 013 196.31 960 0 1156 3 0.3359 1.3961 13320 220.0 2328 21 567 0 016864 IS 701 18718 200 35 957.4
!!57.8 0 3417 1.3842 13260 232.0 til e 23 216 001ER95 17 454 17 471 204.40 954 8 1159.2 0.3476 13725 13201 235.0
240 8 24 968 0 016926 18 304 16 321 208 45 952.1 1160 6 0.3533 1.3609 13142 240.0 2440 26 826 0016958 15 243 15 260 212.50 949.5 1162.0 0.3591 1.3494 13085 244.0 288 0 2819G 0 016990 14 264 14 281 216.56 946 8 1163.4 0.3649 1.3379A IJ028 248.9 j
252 0 30 883 0 017012 13 358 13 375 220 62 944.1 11643 03706 1.3266 16972 252.0 256 0 31 091 0 011055 17 570 12 538 224 69 941.4 1166.1 0 3163 1.3154 1 6917 254 0 200 8 35 427 0 017089 11 745 11 762 228 76 938 6 1167.4 0.3819 1.3043 16862 260.0 284 0 37 894 0 017123 11025 11 042 232.83 935 9 11683 0.3876 1.2933 16808 264.8 Ile s 40 500 0 017157 10 358 10 375 236 91 933.1 1170 0 0.3932 1.2823 1.6755 268.0 l
212 I 43 249 0 017193 9738 9 755 24099 930 3 1871 3 0.3987 1.2715 1.6702 272.0 fit I 46 141 0 017228 9 167 9 180 245 08 927.5 1172 5 04043 1.2601 16650 275 I t
200 8 49 200 0 017264 8 627 8 644 249.17 924.6 1873 8 04098 1.250!
1.6599 200.0 284 I 52 414 0111730 8 1780 8 1453 253 3 9213 1175 0 0 4I54 12395 1.6548 284.8 281 9 55 795 0 01734 16634 7 6807 257.4 918 8 1176 2 04208 1.2290 1 6498 288.9 212 8 59 350 0 01138 12301 7 2475 261.5 915 9 1177.4 0 4263 1.2186 1.6449 292.0 P
7ts t 63 084-0 01741 6R759 6 8433 265 6 913.0 1178 6 04317 1.2082 1 6400 296.0
.
-
u,
_-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - - _ _ _
- _ -. -_ _
_ _ -
-
-
.
i
.
P cn
~-~ Abs Press Sper.ific Volume Enthalpy Entropy
,
Temp tb per Sal Sal.
Sat.
Sal.
Sat.
Sal.
Temp i
Ialis Sqin liquit!
Ivap Vapor liquid Evan Vapor liquid Evap Vapor l'ahr
,
h s
i V
h l h it g
sg sgg g
I p
Vg Vit t
300 0 61 005 0 01145 6 4483 6 4658 2697 9100 1179.7 0 4372 1.1979 1.6351 30s.e 184 3 71119 001749 6 0955 6 1130 273 8 907.0 1180.9 0 4426 1.1877 16393 384.8 30s e 75 433 0 01153 5 1655 5 7830 278 0 904 0 1182.0 0 4479 1.1776 1.6256 300.0 312 0 79 953 0 01157 5 4566 5 4142 2821 901.0 1183.1 04533 1.1676 16209 312.8
.
31s s 84 688 0 01761 5 1611 51849 286 3 897.9 1184.1 0 4586 1.1576 1.616 31E 8 329 8 89643 0 01166 4 8961 4 9138 290 4 894.8 1185 2 04640 1.1477 1.6116 320.0 M4I 94 826 001110 4 6418 4 6595 294 6 891.6 1186.2 04692 1.1378 1.6071 324.0 328 e 100 245 0 01714 4 4030 4 4208 2983 888 5 1181.2 0.4745 1.1280 1.6025 32s e r
f 332 I 105 907 0 01179 4 1788 4 1966 302 9 885 3 1188.2 0.4798 1.1183 1.5981 332.8 336 a 111 870 0 01183 3 96Al 3 9859 307.1 882.1 1189.1 0.4850 1.1086 1.5936 336.8
.
348 e 117.992 0 01781 3 1699 3 7878 311 3 878 8 1190.1 0.4902 1.0990 1.5892 348.8
l 344s 124430 0 01192 3 5834 3 6013 315 5 815.5 1191.0 0 4954 1.0894 1.5849 344.0 34s s 131.142 0 01197 34018 3 4258 319 7 872.2 1191.1 0.5006 1.0799 1.5806 348 8
,
352 3 138 138 001801 3 2423 3 2603 323.9 868.9 1192.7 0.5058 1.0705 1.5163 352.0 nos 145 424 0 01806 30R63 31044 328.1 865 5 1193.6 0 5110 1.0611 1.5721 35E.8
'
344 I 153 010 0 01811 2 9392 2 9573 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360.8 3s4 s
- 160 903 001816 2 8002 28184 336 5 858.6
!!95.2 0 5212 1.0424 1.5637 364.8 354 8 169113 0 01821 2 6691 2 6813 3408 8551 1195.9 0 5263 1.0332 1.5595 368.8 372 8 177 648 0 01826 2 5451 2 5633 345 0 851.6 1196 7 0.5314 1.0240 1.5554 372.0
ils I 186 517 0 01811 24219 2 4462 349 3 848.I 1197.4 0.5365 1.0148 1.5513 376.8 388 8 195 729 001836 2 31TO 2 3353 353 6 844.5 1198.0 0.5416 1.0057 1.5473 380.0 304 8 205 294 0 01842 2 2120 2 2304 357.9 840 8 11983 0.5466 0.9%6 1.5432 384.0 388 I 215 220 0 01847 2.1126 2 1311 362.2 837.2 1199.3 0.5516 0.9876 1.5392 388.0 192 0 225 516 001853 2 0184 2 0369 366.5 833.4 1199.9 0.5567 0 9786 1.5352 392.0 394 8 236 193 0 01858 1.979!
l9477 370 8 8293 1200.4 0.5617 0 9696 Is5313 396 8 480 I 247 259 0 01964 18444 I8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400.0 484 0 258 725 0 01870 11640 11827 379 4 822 0 1201.5 0 5717 0 9518 1.5234 404.0 488 8 270 600 0 01815 1 6817 1 1064 383.8 818 2 1201.9 0 5766 0 9429 1.5195 488.8 412 0 282 894 001881 16152 16340 388.1 814.2 1202.4 0.5816 0.9341 1.5157 412.8 418 8 295 617 oOIR87 I5461 15651 392 5 810.2 1202.8 0 5866 0.9253 1.5118 416 8 428 8 308180 0 01894 14808 14997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 429 e 4248 322 391 0 01900 14184 14314 401.3 802.2 1203.5 0 5964 0.9077 1.5042 424.0 421 I 336463 0 01906 13591 13782 405 7 798 0 12033 0 6014 0 8990 1.5004 4288 432 8 351 00 0 01913 130266 132119 410 I 793 9 1204.0 0 6063 0.8903 1.4966 432.8 43E I 366 03 0 01919 124RH1 126806 414 6 7893 1204.2 0.6112 0 8816 I.4928 436.I 440 8 381 54 0 01976 I19761 121687 4190 785 4 1204.4 0.6161 0 8729 1.4890 448.8 4448 39756 0 01933 1.14874 1.16806 423 5 181.1 1204 6 0 6210 0 8643 1 4853 444.8
-
444 8 414 09 001940 110712 112152 428 0 716 7 12043 0 6259 0 8557 1.4815 448.8
,'s 431 14 nniq47 Ins 7r4 I nFFil
- m 777 1 MnA 9 A C16e n es te ea11e
/.h
"
,
\\
.
..
.
to I
Ahs Prett Sperif er Volume Enthalpy Entropy Temp lb per Sal Sal Sal.
Sat.
Sal.
Sat.
Temp iabr Sq I, I irpo rt Fvan Vapor liquid Evap Vapor liquid Evap Vapor Fahr I
p vg v tg vg he h ig h
sg 5,g sg t
g 4ED I 466 87 0 01961 0 97463 0 99424 441.5 763.2 1204.8 0.6405 0 8299 1.4704 440.0 454 5 485 56 0 01969 0 93588 0 95557 446.1 758 6 1204.7 0 6454 0 8213 1.4667 464.8 458 3 504 83 0 01916 0 89885 0 91862 4501 754.0 1204.6 0.6502 0 8127 1.4629 468.8 417 8 524 67 0019R4 0 86345 0 88329 455 2 749.3 1204.5 0.6551 0 8042 1.4592 472.0 415 0 545ll 0 nt%7 087954 0R4950 459.9 744 5 1204.3 0 6599 03956 1.4555 475.8 413 8 SCC 15 0 02000 0 19116 / 0 81717 464.5 739 6 1204.1 0.6648 01871 1.4518 400.0 414 3 58181 0 02009 01661)
018622 4691 7343 1203.8 0.6696 01785 1.4481 484.8
'
4I10 610 10 0 01011 0 13641 015658 473 8 729 7 1203.5 0.6745 03700 1.4444 400.0 417 0 633 03 0 01026 0 10191. 0/?P20 478 5 724 6 1203.1 0 6793 0 7614 1.4407 492.0 495 8 6% El n n7014 0 GROM 010100 4R3 2 719 5 1202.7 06842 03528 1.4370 496.9 5:38 680 86 0 02043 0 65448 0 67492 487.9 714.3 1202.2 0.6890 0 7443 1.4333 500.0 504 0 705 18 0 0705)
O f,7938 0 64991 4923 709.0 1201.7 0.6939 01357 1.42 %
504.9 500 0 73140 0 02062 0 60530 0 62592 497.5 703 7 1201.1 0.6987 0.7271 1.4258 548.9 512 0 15732 0 02012 05Mll8 060289 502.3 6982 1200.5 03036 01185 14221 512.0
,
515 0 744 76 0 07081 0 % 991 0 58019 5071 6923 1199.8 03085 01099 1.4183 516.8
'579 8
~
812 53 0 02091 0 53864 0 55956 512.0 687.0 1199 0 01133 03013 1.4146 529.9 574 0 841 04 0 02102 0 51814 0 53916 516.9 681.3 1198.2 03182 0.6926 1.4108 524.0 571 8 R70 31 0 02112 0 49843 0 51955 521 8 675.5 1197.3 07231 0.6839 I.4070 528.9 512 8 900 34 0 02123 0 41947 0 50010 526 8 669 6 11 %.4 0 7280 0.6752 1.4032 532.8 516 0 931.17 c07134 046173 0 48757 5313 663.6 1195.4 01329 0 6665 1.3993 536.8 541 8 962 79 0 0?!46 (144367 0 46513 536 8 657.5 1194.3 0 7378 0 6577 1.3954 540.0 544 0 995 22 0 02157 0 47617 0 44834 54I 8 651.3 1193.1 03427 0.6489 1.3915 544.0 541 0 1028 49 0 02169 0 41048 0 43217 546 9 645 0 1191.9 0 7476 0.6400 3.3876 548.8 5520 1062 59 0 02182 0 39419 0 41660 552 0 638.5 1190.6 0 7525 0.6311 1.3837 552.0 556 8 1097 55 0 07194 011966 040160 5572 632.0 1189 2 0 7575 0 6222 13797 556.0 553 8 1133 38 0 02207 0 36507 0 38714 562.4 625 3 1187.7 01625 0.6132 1.3757 564 8 584 0 1170 10 0 07721 0 35099 0 31320 567.6 618 5 1186.1 03674 0 6041 1.3716 564.0 5EI 3 1207 72 0 02235 0 33141 0 35915 572 9 611.5 1184 5 03725 0.5950 1.3675 548.0 572O 1746 26 0 07249 0 37429 0 34678 578 3 604.5 11821 01775 0.5859 1.3634 572.0 578 0 1785 71 0 077A4 031167 0 33476 5831 597.2 1180.9 03825 0.5766 1.3592 578.0 Sit O 1326 17 0 02279 0 29937 0 32216 589.1 589.9 1I79.0 03876 0.5673 1.3550 500.8
584 I 1367 7 0 02295 0 28153 0.31048 594 6 582.4 1176.9 0.7927 0.5580 1.3507 584.8
',
i
!st O 1410 0 0 02311 021608 0 29919 6001 5741 1174 8 03978 0.5485 1.3464 588.0 Y
512.0 1453 3 0 02328 0 26199 0 28827 6051 566 8 1172.5 0 8030 0.5390 1.3420 592.8
"
516 8 1491.8 0 07345 0 ?s425 0 ???70 611.4 558.8 1170.2 0 8082 0.5293 1.3375 596 0
.
.
%
%
,
?
m
- r ' '
!
Ahs Psets Specifsc Volume Enthalpy Entropy
.
femp lb pes Sal.
Sal.
Sal.
Sal.
Sat.
Temp Fahs Sqin liquid Fvan vapor Li vid Evap Vapor Liquid Evap Vapor Fahr I
P
'l
'Ig_.
't I
h rg h
sr sig s
I g
t Bes t 15432 0 02364 024384 0 26747 617.1 550.6 1167.7 0.8134 0.5196 1.3330 500.8 set 8 15897 0 02382 0 23314 0 25757 622.9 542.2 1165.1 0.8187 0.5097 1.3284 544.8 g
I:I 8 16373 0 02402 0 27194 0 24796 628 8 533 6 1162.4 0 8240 0.4997 1.3238 500.0 1<
112 8 16861 0 02477 0 21442 0 23865 634 8 524.7 1159.5 0.8294 0.48 %
1.3190 812.0 518 8 1735 9 0 07444 0 70516 0 779E0 640 8 515.6 1156.4 0.8348 0.4794 1.314I til.O 0 1961 ! 0 22081 646.9 506.3 1153.2 0.8403 0.4689 1.3092 029.8 E20 g.
I786 9 0 02466 524 s 18390 0 02489 0 18737 0 21226 653.1 406 6 1149.8 0.8458 0 4583 1.3041 824.0
,
828 8 1892 4 0 02514 0178R0 0 20394 659.5 486.7 1146.1 0.8514 0.4474 1.2988 528.9
'
832 8 19170 0 02539 017044 0 19583 665.9 476.4 1142.2 0.4571 0.4364 1.2934 532.8 E3s I 20078 0 075E6 0 16726 0 18792 672.4 465.7 1138.1 0.8628 0.4251 1.2879 536.8
$48 I 2059 9 0 02595 0 15427 0.18021 679.1 454.6 1133.7 0.8686 0.4134 1.2821 648.0 844 8 2118 3 0 02625 0 14644 017269 685 9 443.1 1129.0 0.8746 0.4015 1.2761 644.0 848 8 21781 0 02657 0 13876 0 16534 692.9 431.1 1124.0 0.8806 0 3893 1.2699 648.9 852 0 2239 2 0 02691 0 13124 0.15816 700 0 418.7 1118.1 0.8868 0.3767 1.2634 652.0
,
858 I -
23011 0 0277R 017387 0 15115 707.4 405 7 1113.1 0.8931 0.3637 1.2567 656.8
'
Ist 8 2365.7 0 02768 011663 0 14431 714.9 392.1 1107.0 0.8995 0.3502 1.2498 000.0 884 8 2431.1 0 02811 0 10941 0 13757 722.9 377.7 1100 6 0.9064 0.3361 1.2425 684,8 lil 8 24981 0 02858 010229 0 13087 731.5 362.1 1093.5 0.9137 0.3210 1.2347 668.0'
';
812 8 2566 6 0 02911 0 09514 0 12424 740.2 345.7 1085.9 0.9212 0.3054 1.2266 872.8 sis 8 2636 8 0 02970 0 08799 011769 749 2 328.5 1077.6 0.9287 0.2892 1.2179 876.0 IIII 27085 0 03037 0 08080 0.11117 758.5 310.1 1068.5 0.9365 0.2720 1.2006 080.0 114 8 2782.1 0 03114 0 01349 0 10463 768.2 290.2 1058.4 0.9447 0.2537' 1.1984 884.0 les s 28574 0 03204 0 06595 0 09799 778.8 268.2 1047.0 0.9535 0.2337 l.1872 888.0 i
182 8 2934 5 0 03313 0 05797 0 09110 790 5 243.1 1033.6 0.9634 0.2110 1.1744 502.8 III I 3013 4 0 03455 0 04916 0 08371 804 4 212.8 1017.2 0.9149 0.1841 1.1591 608.0 Tet I 3094.3 0 03662 0 03857 0 07519 822.4 172.7 995.2 0.9901 0.1490 1.1390 780.0 182 8 3135 5 0 03824 0 03173 0 06997 835.0 144.7 979.7 1.0006 0.1246 1.1252 782.8 184 8 3171.2 0 04108 0 02192 0 06300 854.2 102.0 956.2 1.0169 0.0876 1.1046 704.0 785 8 31983 0 04427 0 01304 0 05730 873.0 61.4 934.4 1.0329 0.0527 1.0856 705.0 135 47'
32082 0 05078 0 00000 0 05078 906.0 0.0 906.0 1.0612 0.0000 1.0612 795.47'
l
.
i
' Critical lemnerature-3 a
"%
,
-
.
o
.
r Table 2:
Saturated Steam: Pressure Tabk Specific Volume Enthalpy Entropy Abs Press Temp Sal
-
Sat.
Sat.
Sat.
Sat.
Sai.
. Abs Press.
I.blSo in Fahr liquir!
Fvan Vapor ti uid Evap Vapor Liquid Evap Vapor Lb/Sq in.
.
p t
v, v,,
g g
hg n
sg 5,
s p
v
g g
g
---
.
- _ _ _.
..._._
,
,
881165 32 018 0 016022 3302 4 3302 4 0 0003 1075 5 1075.5 0 0000 2.1872 2.1872 8.08885 8 25 59 323 0 016032 1235 5 12355 27 382 10601 1087.4 0 0542 2.0425 2.0967 8.25 ISI 19586 0 016071 641 5 641.5 47.623 1048 6 1096 3 0.0925 1.9446 2.0370 0.54 II 10114 0016135 33159 333 60 6973 1036I 1105.8 0.1326 I.8455 1.9781 1.8 5I 16224 0016407 13 515 13 532 13020 1000.9 1131.1 0 2349 1.6094 1.8443 S0 18 8 193 21 0016592 38404 38 420 161.26 982.1 1143.3 0.2836 1.5043 1.7879 10 8 14 Ell 212 00 0 0I6719 26182
/26 193 180.17 970.3 1150.5 0 3121 1.4447 I.7568 14.598 15 8
?!3 03 nnir.726 26 214 26290 181.21
%97 1150.9 0 3137 1.4415 1.7552 15.0 28 8 22795 0016834 20 010 20 087 19627 9601 1156.3 0.3358 1.3%2 I.7320 29.8 33 e 25034 0 017009 13 1766 13 1436 218 9 945 2 1164.1 0.3682 1.3313 1.6995 30.9 43 8 26725 0 011151 10 4794 10 4965 2361 933 6 1169.8 0.3921 1.2844 1.6765 48.0 18 s 281 02 0 017274 8 4967 8 5140 2502 923 9 1174.1 0 4112 1.2474 1.6586 50.0 la 3 292 71 0 017383 71562 71736 2622 915.4 1177.6 0.4273 1.2167 1.6440 50 8 13 3 302 93 0017482 6 1875 6 2050 272.7 907.8 1180.6 0.4411 1.1905 1.6316 78.0
!
sa t 312 04 0 017573 5 4536 5 4111 282.1 900 9 1183.1 0.4534 1.1675 1.6208 88.8 33 8 320 28 0 017659 4 8719 4 8953 290 7 894 6 1185.3 0.4643 1.1470 1.6113 N.0
'
~
32782 0 017740 4 4133 4 4310 298.5 888 6 1187.2 0.4743 1.1284 1.6027 100.9 letI lit S 334 79 0 01782 4 0306 4 0484 305.8 883.1 1188.9 0.4834 1.1115 1.5950 119.0 128 8 341 27 0 01789 3 7091 3 7275 312.6 877 8 1190.4 0 4919 1.0 % 0 1.5879 129 0 138 8 34733 0 01796 3 4364 3 4544 319 0 872 8 1191.7 0.4998 1.0815 1.5813 130.0 143 I 353 04 001803 3 2010 3 2190 325 0 8680 1193.0 0.5071 1.0681 1.5752 148.8 I5O O 358 43 001809 29958 30139 330 6 863 4 1194.1 0 5141 1.0554 1.5695 150.I 153 8 363 55 001815 28155 2 8336 336I 8590 1195.1 0.5206 1.0435 1.5641 168.5 118 3 36842 0 01821 2 6556 2 6138 3412 854.8 1196.0 0 5269 1.0322 1.5591 178.I 183 I 37308 0 01827 2 5129 2 5312 346.2 8507 1196.9 0.5328 1.0215 # l.5543 180.0 133 8 377 51 0nIR11 23847 2 4030 350 9 8467 1197.6 0.5384 I.0113 1.5498
.190 0 238 I 38180 0 01839 22689 2 2873 355 5 842 8 1198 3 0.5438 1.0016 1.54 54 200.8 213 s 385 91 001844 216373 2 18217 359 9 839.1 1199 0 0.5490 0.9923 1.5413 flg O 223 3 38988 001850 2 06179 202629 364 2 835 4 11S9 6 0 5540 0.9834 1.5374 229 0 238 I 393 70 0 01855 197991 199846 368 3 831.8 1200.1 0 5588 0 9748 1.5336 238 8 243 8 39739 001860 189909 I91169 372.3 828 4 1200.6 0.5634 0.9665 1.5299 248.8 253 3 400 97 001865 182452 184317 376 1 825 0 1201.1 0.5679 0.9585 1.5264 254 0 Ils t 404 44 0 01870 115548 117418 379 9 8216 1201.5 0.5722 0.9508 1.5230 258.8 2III 40780 0 01875 169131 131013 383 6 818 3 1201.9 0 5764 0 9433 1.5197 270.8
'
2-IIII 41107 0 01580 163169
!E5049 387.1 815 1 1202.3 0.5805 0 9361 1.5166 288.8
,
la 258 8 414 25 0ntRA5-157591 159482 390 6 812.0 1202.6 0 5844 0 9291 1.5135 290.0 338 8 41735 0 01889 1.52384 154274 394 0 808.9 1202.9 05882 0.9223 1.5105 300.0 353 8 43113 0 01912 I20642 1.32554 409 8 194.2 1204.0 0 6059 0 8909'
l.4968 358.8 ac3 8 444 60 Onl934 1.14162 1.16095 424.2 780 4 1204.6 0.6217 0.8630 1.4847 40.
.
Specific Volume Enthalpy Entropy
_..
Abs fress Temp Sal Sal.
Sat.
Sat.
Sat.
Sat.
Abs Press.. >
P t h/Sel In F alir lirluial tvan vapoi liquid Evan Vapor liquid Evap Vapor tblSq In.
.
7; p
i v
V V
h, hg h
s, sgg g
p s
s y
I g
g 4538 G 23 001954 1 01224 1 03179 437.3 7675 1204.8 0 6360 0.8378 1.4738 454.8 5:38 46701 0 01975 0 90767 0 92162 449.5 155.1 1204.7 0 6490 0 8148 1.4639 500.0 553 5 47694 001931 0 8/183 0 84177 460 9 743 3 1204.3 0 6611 0 1936 1.4547 558 0 E:3 3 45520 0 01013 0 14967 0 76975 4713 732 0 1203.7 06723 01738 14461 set c 153 3 49483 0 07032 0 68811 0 10843 481.9 720.9 1202.8 0.6828 03552 14381 558.0 133 3 50308 on7050 063505 0 65556 49I 6 110 2 1201.8 06928 03377 1.4304 788.8 753 8 510 84 0 07069 9 5h30 0 60949 500.9 699 8 1200 7 03022 03210 1.4237 750 0 I:38 518 21 0 0/087 0 54809 0 56896 509 8 689 6 1199 4 03111 0 7051 1 4163 809 9 850 8 525 24 0 02105 0 51197 0 53302 518 4 679 5 1198 0 03197 0.6899 1 40 %
850 0 5:38 53195 0 07173 0 47968 0 50091 5267 669 7 1196 4 03279 0.6753 1.4032 900.0 153 O 53339 0 07141 0 45064 0 47205 534 7 660 0 11943 0.7358 0 6612 1.3970 958.8 10:3 3 544 58 0 021 %
042436 0/4596 542 6 650 4 1192.9 07434 0 6416 13910 1000 8 1853 3 55053 0 02177 0 40047 0 42224 5501 640 9 1191.0 03507 06344 1.3851 1859 8
'
1103 3 55628 0 02195 0 37853 0 40058 557.5 631.5 1189.1 0 7578 0.6216 1.3794 1108.8 1155 8 561 82 0 02214 0 35859 0 38013 564 8 622 2 1187.0 0.7647 0 6091 1 3738 1150.0
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Title / Location Setpoint/ Initiating Device ATWS Div. 2 tripped
+117f + 15 psig pressure Panel 905 left C-13 PT-122B, D*
I-49" level I
LT-1208, D Reactor vessel hi pressure scram 1085 psig Panel 905 right A-1 PS-263-55A, B, C, D Reactor high pressure
+1040 psi Panel 905 left A-11 WR reactor pressure recorder 640-27 (panel 905)
Cond. vac/ main steam iso.
600 psi Yalve closure scram bypassed PS-263-51A, B, C, D Panel 905 right C-14 mode switch in other than RUN 4. Interlocks and Trips Interlock or Trip Functions
f Reactor feed pump At +60" reactor water level, signals from j
high water level trip alarm module 640-44A(B) trip the reactor feed pumps. This helps preclude turbine
! l damage from water carryover.
'
Main turbine stop valve At 48" in the reactor vessel, signals f rom trip on vessel high level LITS-59A and B trip the main turbine stop i
valves shut. This protects the main
turbine from damage by moisture carryover.
j RCIC turbine trip on AT +48" in the reactor vessel, signals from vessel high level LIS-72A*and B trip the RCIC turbine. This protects the turbine f rom damage by moisture carryover.
.
. IIPCI turbine trip on At 48". in the reactor vessel, signals f rom
!.
vessel high level LIS-72A and B trip the HPCI turbine.
This l
protects the turbine from damage by moisture carryover.
i
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CSCS initiation on At -49" in.-the reactor vessel, signals f rom
,
vessel low-low level LIS-72A, B, C, D initiate the core standby cooling systems.
This includes starting the standby diesel generator and the core spray, RHR (LPCI mode), RCIC, HPCI and ADS systems.
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BOSTON
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EDISON
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NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure No. 5.7.1.1 EMERGENCY CATEGORIES AND ASSOCIATED EMERGENCY ACTION LEVELS
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t List of Effective Pages 5. 7.1.1 -1 5. 7.1.1 -2 5.7.1.1-3
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Attachments 5.7.1.1A-1
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5.7.1.10-1 5.7.1.10-2 5. 7.1.1 C-1 5.7.1.1C-2 5.7.1.10-1 5.7.1.10-2 5.7.1.1E-1 5.7.1.1E-2
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5.7.1.1F-1
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Approved d@Mw/
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Date
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s 5.7.1.1-1 Rev. 4
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I.
PURPOSE The purpose of this procedure is to define each emergency action level
,
for declaring an emergency in each category.
II.
ACTION A.
Declaration of Emergency by the Watch Engineer
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1.
If emergency conditions are suspected, refer to Attachment A through E to determine if an EAL has been reached or exceeded and if so, the emergency class.
2.
After taking all necessary immediate response actions, declare the emergency and proceed to the next step.
3.
Follow the instructions given in the proper procedure for that emergency class declared.
Personnel Emergency - Procedure No. 5.6.1
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Unusual Event - Procedure No. 5.7.1.2
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Alert - Procedure No. 5.7.1.3
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q Site Area Emergency - Procedure No. 5.7.1.4
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General Emergency - Procedure No. 5.7.1.5
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4.
As conditions-change, the emergency classification may have to be upgraded or downgraded as stated in Attachments A through E.
5.
Request the Security Supervisor to call-in additional operations personnel as necessary. NOTE: Make every practicable ef fort to minimize personnel radiation exposure.
Do not exceed normal radiation exposure restrictions without the specific approval of the Emergency Director (Watch
Engineer). The urgenry and potential risk of the emergency
.
situation will determine the degree of precautions to be taken (i.e., surveys, respiratory protection, protective clothing).
III.
RESPONSIBILITIES It is the responsibility of the Watch Engineer to ensure any emergency situation is promptly identified and properly classified and the appropriate class procedure is implemented.
It is also his responsibility to initially function as the Emergency Director until properly relieved of this duty.
It is the responsibility of the Station Manager to relieve the Watch Engineer as Emergency Director as snon as possible af ter being notified of an emergency.
It is also his responsibility to ensure the proper response, assessment and corrective action is implemented in accordance
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with the appropriate procedures.
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5.7.1.1-2 Rev. 4
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IV.
DISCUSSION Emergency situations are classified to cove'r the entire spectrum of
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possible radiological and non-radiological emergencies that may be a
encountered at PNPS. The classifications are:
1.
Personnel Emergency 2.
Unusual Event
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,
3.
Alert
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4.
Site Area Emergency
5.
General Emergency Accidents may be initially classified in a particular category and later changed to another classification if conditions warrant.
e Each of the emergency classes are characterized by Emergency Action Levels (EAL's). Except for the Personnel Emergency Class, the EAL's consist of specific sets of plant parameters (i.e., instrument indications, system status, etc.) to be used to initiate emergency class designation, notifications, and mobilization of emergency organizations.
These EAL's are summarized in Attachment F. "EAL CHART." A controlled copy of the EAL CHART is located in the following Emergency Response
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Facilities: Control Room. Technical Support Center (TSC), Emergency Operations Facility (EOF), and the Recovery Center.
If an EAL is I
revised, Document Control must be notified in order to ensure that the controlled copies of the EAL CHART are appropriately modified.
The on-duty Watch Engineer, initially acting as the Emergency Director has the ultimate responsibility to initially classify and declare emergency conditions based on the EAL's. He may also declare an emergency condition based on any event that may affect the safe operation of the plant or the health and safety of plant personnel and the general public.
- V.
ATTACHMENTS A.
Personnel Emergency - Emergency Action Levels 8.
Unusual Event Emergency Action Levels
,
C.
Alert Emergency Action Levels
D.
Site Area Emergency Action Levels -
E.
General [mergency Action Levels F.
EAL CHART t
5. 7.1.1 -3 Rev. 4
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ATTACHMENT A - Personnel Emergency - Emeraency Action levels
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1.
Onsite emergency medical treatment required for any individual with E without evidence of internal or external contamination.
2.
Of fsite emergency medical treatment required for any individual without evidence of internal or external contamination,
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5.7.1.1A-1 Rev.,4
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ATTACHMENT B - Unusual Event Emergency Action levels
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1.
Release rate of airborne or liquid radioactive ef fluent technical specification for 15 minutes or more as indicated by the Main Stack Reactor Building Vent or Liquid Waste Discharge Radioactive Effluent Monitor.
2.
Greater than 500,000 uti/sec at air ejector for
,;
15 minutes or more or an increase of
.
100,000 uCi/sec within a 30 minute period using the most recently calculated conversion constant f rom mr/hr to uti/sec 3A 3.
Greater than 20 uti/mi of total iodine in reactor water (confirmed).
4.
Irradiated fuel in vessel, reactor coolant temperature greater than 212*F and any of the following conditions:
a.
Reactor coolant system unidentified leakage l
greater than 5 GPM and total leakage greater than 25 GPM when averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
period as indicated by drywell sump flow integrators.
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b.
Reactor coolant pressure in excess of technical specification safety limits.
c.
Failure of a safety / relief valve to properly close after reduction of applicable pressure as indicated by thermocouple or acoustic monitors.
d.
Primary containment isolation valves inoperable requiring shutdown according to technical specifications.
e.
Loss of ECCS or fire protection systems requiring plant shutdown according to technical specifications.
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f.
Loss of all onsite AC power capability.
g.
Loss of indication and annunciation on any safety related system requiring plant shutdown attording to technical specifications.
h.
A fire onsite that is not controlled within 20 minutes after fire fighting efforts have begun.
5.7.1.1B-1 Rev. 4
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I ATTACHMENT B - Unusual Event Emergency Action Levels (Continued)
5.
An earthquake that causes observed station damage.
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6.
Onsite airplane crash.
- 7.
Unplanned explosion onsite.
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8.
Notification of release of toxic gases within 1 mile of site boundary.
9.
Confirmation of a tornado touching ground onsite.
10.
A hurricane with sustained ( > 15 minutes) wind speed greater than 90 mph at the site.
11.
Tran'sportation of an injured person of fsite for emergency medical treatment with evidence of internal or external contamination.
12.
Attempted entry onsite with evidence of intent to sabatage.
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5.7.1.18-2 Rev. 4
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ATTACHMENT C - Alert Emeroency Action Levels 1.
Release rate of airborne or liquid radioactive effluents in excess of ten times technical specifications for 15 minutes or more as indicated by the Main Stack, Reactor Building Vent or Liquid
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Waste Discharge Radioactive Effluent Monitor.
2.
Greater than 5 Ci/sec at air ejector for 15 minutes
,;
or more using the most recently calculated
.
conversion constant from ar/hr to uti/sec.
3.
Greater than 200 uti/mi of total iodine in, reactor. water (confirmed).
4.
Irradiated fuel in vessel, reactor coolant
'
temperature greater than 212 degrees F and any of the following conditions:
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a.
Reactor coolant system total leakage in
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excess of 50 gpm average over one hour as indicated by the drywell sump flow
,
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integrators.
b.
Loss of all offsite power coincident with loss of both emergency diesel generators.
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c.
Loss or potential loss of habitability of Control Room as witnessed by physical indicators (eg. smoke, fire),
,
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d.
Any natural phenomena that could potentially impair ECCS capability.
i e.
Failure of the reactor protection system to properly shutdown the reactor on valid trip signals after immediate and reasonable
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corrective actions.
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f.
Loss of all onsite D.C. power (24,125 and 250 VDC) for 15 minutes or less (greater than 15 minutes is a Site Area Emergency.
g.
A fire within any process building that has the potential to affect safety systems within one hour or less if not controlled.
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h.
Loss of al1 control room annunciators for 15 minutes or more.
5. 7.1.1 C-1 Rev. 4
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ATTACHMENT C - Alert Emergency Action Levels (Continued)
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5.
Airplane crash onsite that causes substan-tial observed damage to process buildings or switchyards.
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6.
An earthquake that causes substantial observed process building damage.
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7.
Direct radiation levels observed in the plant
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which increase by a factor of 1000 over a period of I hour or less. (Excluding the results of controlled processes such as in-core neutron detector withdrawal into the TIP room.)
8.
An ongoing security compromise requiring assistance by an offsite security force.
9.
Complete loss of safety related systems required by Technical Specifications necessary to maintain the reactor in a cold shutdown condition.
10.
Any other plant condition which in the judgment of the Watch Engineer warrants increased awareness by offsite agencies.
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5.7.1.1C-2 Rev. 4 i
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ATTACHMENT 0 - Site Area Emergency Action Levels 1.
Irradiated fuel in vessel, reactor coolant
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temperature greater than 212 degrees F and
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any of the following conditions:
i a.
Reactor vessel water level below the top of core (9 inches on 903 panel) and decreasing.
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j b.
Main steam line rupture outside primary
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containment that is not isolatable.
c.
Loss of both of fsite power and onsite AC power for more than 15 minutes, ie. loss of Line 342 and 355 and shutdown transformer and both diesel generators.
d.
Loss of all onsite DC power (24,124 and 250 VDC)
for 15 minutes or more.
e.
Failure of the reactor to shutdown on manual
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initiation within 15 minutes with MSIV's open (with MSIV's closed this is a General Emergency)
f.
Loss of all annunciators coincident with any g
Alert Emergency Action Level.
.
g.
Evacuation of the main Control Room without establishing the ability within 15 minutes to begin shutdown procedures.
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2.
An airborne radioactive effluent release which results in a calcuated dose rate at any offsite location greater than 500 mrem /hr for 2 minutes to the whole body or 2500 mrem /hr to the thyroid under adverse meterology as indicated by a level of 5 x 105 cps on the Main
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Stack or Reactor Building Vent Radioactive Effluent Monitor.
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3.
An airborne radioactive effluent release which results in a calculated projected dose at any offsite location in excess of 1 Rem whole body or 5 Rem thyroid using actual neteorology.
4.
Drywell high range monitor reading in excess of 105 R/hr (equivalent to 1 Rem whole body at the site boundary in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />).
,
5.
Results of environmental measurem.nts that indicate projected doses in excess of the EPA Protective Action Guides of 1 Rem whole body and/or 5 Rem thyroid for the expected duration of the event.
6.
Loss of physical control of the main control room by
,
the Boston Edison Operators.
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5.7.1.10-1 Rev. 4
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(Continued)
l ATTACHMENT 0-SiteAreaEmergencyActionLevels
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7.
High alarm on at least two refueling floor process radiation monitors for 10 minutes or more caused by
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an actual release of radioactive material from a fuel bundle (s).
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$. 7.1.10-2 Rev 4
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ATTACHMENT E - General Emergency Action Levels
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1.
A sustained (one hour) airborne radioactive effluent release which results in a calculated
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dose rate at any location offsite in excess of 1 rem /hr whole body and/or 5 rem /hr thyroid under
-
actual meteor ological conditions.
2.
Drywell high range monitor reading in excess of
,J 106 R/hr (equivalent to 5 Rem whole body at the
.
site boundary in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />)
3.
Results of environmental measurements that indicate projected doses in excess of 5 Rem whole body and 25 Rem thyroid.
4.
The occurrence of any two of the three events listed below ( A, B and C) with the potential for the third event to occur as indicated by water level, pressure or radiation level trends while there is irradiated
'
fuel in the vessel and reactor coolant temperature is in excess of 212 degrees F.
a.
Loss of Fuel Cladding as defined below:
k (first fission product barrier)
1.
Greater than 200 uCi/mi of I-131 (confirmed) in
.
reactor water or, 2.
Greater than 5 C1/sec (corrected for a 30 minute decay) at the air ejector using the most recently calculated conversion constant from mr/hr to uti/sec or, 3.
High radiation level trip on main stream line monitors (2 out of 4) caused by fuel failure or, 4.
Reactor vessel water level below the top of the core ($ inches indicated on 903 panel) for 15
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minutes or more.
b.
Loss of Reactor Coolant Pressure Boundary as indicated below:
(second fission product barrier)
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1.
Drywell atmosphere monitor reading in excess of 100 R/hr or,
2.
Drywell pressure greater than.10 PSIG and increasing
c.
Loss of Primary Containment as defined below : (third fission product barrier)
1.
Failure of any inboard and outboard isolation valve 5.7.1.1E-1 Rev. 4
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ATTACHMENT E - General Emergency Action tevels (Continued)
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in series to close when an isolation signal is present or, 2.
Drywell pressure in excess of 56 PSIG and increasing or,
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3.
Radiation level on Main Stack high range monitor in excess of 5 R/hr.
,
5.
Loss of physical control of the Reactor Building by Boston Edisgn operations or security personnel.
6.
Failure of the reactor to shutdown on manual initiation within 15 minutes with MSIV's closed.
7.
Irradiated fuel in vessel, reactor coolant temperature
-
greater than 212 degress F and any of the following conditions:
a.
Reactor vessel water level below 2/3 core coverage (-48 inches indicated on 903 panel) and decreasing or,
.
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b.
Any combination of events or failure that are likely to result in the loss of any of the following functions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown:
1)
Ability to reliably maintain water over the reactor core
_
ii) Ability to reliably remove decay heat f rom the reactor core
~
iii) Ability to reliably expell decay heat to the ultimate heat sink.
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j 5.7.1.lE-2 Rev. 4
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ATTACHMENT F EAL CHART i
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1.2 EI.A070E C00: ANT SYSTD! INTICRI7T 2.2
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12.M:72N C05'D272CNS FOR 0?I?>.730N 5'. WIIttAN t RIQUmMD:7
3.1 R.IACTOR PROTICTION SYST m 4.1
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3.2 FR07Ec77vI INsntMI.NIA720N 4.2
4.3
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3.3 EIACTIVITY CONTROL
A.
Reactivity Limitations A
3.
Cc:trel Rods
81 C.
Scra: Insertion Times C
D.
Cer. trol Red Accu =ulators D
St.
E.
Rea:tivity A=c: alias E
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F.
Alternate Requireze::s
' 85 C.
Scra= Dis charge Volume C
3.4 5:AN IY LIQUID CCK3D:. SYs;IY.
4.4
-
A.
Fer=al Syste= Availability A
3.
Operatie vith Incperatie Cc=pene::s
96 C.
Sodiu: Fe:taberate Selution C
97 D.
Alternate Requirements
,
103 CO?.' AND CONIAlhMT.NT CDCLIN 575 Ix5 4.5
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I 3.5
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A.
Cere Spray and 170I Subsysta=s A
103 106 3.
Ceetain=ent Coeling subsystes
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C 107 C.
3?:I subsystaz lo!
D
.
D.
RCIC Subsystsc 109 E.
Aute:atic Depressurinatien System E
F.
Minimu= ten Fressure Coeling syntam and
. Diesel Camerator Availability F
110 C
111 C. (Leista!)
. R.
Maintamasce of Filled Discharge Pipe I
112
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4.6 123 3.6 FR7.A.RT SYSTD'. SOLYSAET A.
Therr.a1 and Pressuris.ation Lie.itations A
123
i~ 12!.
3.
Coelant Chenistry 125 C
C.
Coelant 1ankage
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126
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D.
Saf ety and Relief Talvas 127 E
.
E.
Jet Furps 127 F
7.
Jet Fu=p Flow Mismatch'
127
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C
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C.
Structural Integrity 1273 Righ Energy' Fipin- '.r)utside Containment) I 137s c
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Shock Suppres. ors (Snubbsm)
i Acendent No. 65
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Sureeillance PaQQ No.
3.7 CONTAINMENT SYSTEMS 4.7 152 A.
152
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B.
Standby Gas Treatment System B
158 C.
159
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D.
Primary Containment Isolation Valves D
160 3.8 RADI0 ACTIVE EFFLUENTS 4.8 177 A.
Licuid Effluents Concentration A
177 B.
Radioactive Liquid Effluent
177 Instrumentation
.
C.
Liquid Radwaste Treatment.
C 178 D.
Gaseous Effluents Dose Rate D
179 E.
Radioactive Gaseous Effluent E
180 Instrumentation F.
Gaseous Effluent Treatment F
181 G.
182 H.
Mechanical Vacuum Pump H
183 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 194, A.
Auri11ary Electrical Equipment A
194 B.
Operation with Inoperable Equipment B
195 3.10 CORE ALTERNATIONS 4.10 202
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A.
Refueling Interlocks A
202 g
B.
Core Monitoring B
202 C.
Spent fuel Pool Water Level C
203 3.11 REACTOR FUEL ASSEMBLY 4.11 205A A.
Average Planar Linear Heat
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Generation Rate (APLHGR)
A 205A B.
Linear Heat Generation Rate (LHGR)
B 205A-1 C.
Minimum Critical Power Ratio (MCPR)
C 2055 l
s 3.12 FIRE PROTECTION 4.12 206 A.
Fire Detection Instrumentation A
206
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B.
Fire-Suppression Water System B
206 C.
Spray and/or Sprinkler Systems C
206c D.
Halon System D
206d E.
Fire Hose Stations E
206e F.
206e
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G.
Dry Chemical Systems G
206e-1 H.
Yard Hydrants and Exterior
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Hose Houses H
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Page No.
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4.0 MISCELLANEOUS RADIDACTIVE MATERIALS SOURCES 206k
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4.1 Sealed Source Contamination 206L 4.2 Surveillance Requirements 206k
4.3 Reports 2061
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4.4 Records Retention :
2061 5.0 MAJCR DESIGN FEATURES 206m 5.1 Site Features 206m 5.2 Reactor 206m 5.3 Reactor Vessel 206m 5.4 Containment 205m
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5.5 Fuel Storage 207 5.6 Seismic Design
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207 6.0 ADMINISTRATIVE CONTROLS 208 6.1 Responsibility 208 6.2 Organization 208 6.3 Facility Staff Qualifications 208
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6.4 Training
- 208a
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6.5 Review and Audit 212 6.6 Reportable Event Action 216 6.7 Safety Limit Violation 216 6.8 Procedures 216
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6.9 Reporting Requirements 217 6.10 Record Retention 224
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6.11 Radiation Protection. Program 226 6.12 (Deleted)
6.13 High Radiation Area 226 6.14 Fire Protection Program 227 6.15 Environmental Qualification 228 Operational Objectives
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56rveillance 7.0 RADIOLOGICAL ENVIRONMENTAL M3NITORING PROGRAM 8.0 229 7.1 Monitoring Program 8.1 229 L
7.2 Dose - Liquids 8.2 232 7.3 Dose - Noble Gases 8.3 233 7.4 Dose - Iodine-131. Iodine-133, 8.4 234 Radioactive Material in Particulate form, and Tritium 7.5 Total Dose 8.5 234 l
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4.1 SRVIH.13.NCE RECCIRD'EN*S_
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1D'IT!.N CONCITICS TCR CPE?.CION
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RIACTCK PRr*EC'* ION SYSTD'
3.. - RIAC OR FRCTECTICN SYSTD'.
Aeplicabiliev:
A-elicability:
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Applies te the surveillance of App 1'es t6 the instru=e=tation the instru=e=tatic: and associ-ced associated devices which ated devices which initiate re-initiate a reactor scram.
.
actor scra=.
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Obdective:
Cb$ective:
To specify the type a:d frequency To assure the operability of the of surveilla:ce to be applied to reactor protection system.
the protection instru=e=tation.
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Specification:
Seecification:
Instrume tatien syste=s shall A.
The setpoints, r.ini=c: nu her of be fu ctic: ally tested and
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trip syste=s, a:d W1=== nu=ber calibrated as i='icated in of instru=e=t chs:sels that an.tst Tables 4.1.1 and 4.1.2 re-
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be operable for each pesition of spectively.
the reactor mode switch shall be on given 1: Table 3.1.1.
The B.
Daily during rezete: power system response ti=es from the operation, the maxi =us frac-opening of the sensor contact up tic: cf 1'-4ti=g pcver de=sity
'
to and including the ope:ing of shall be checked a:d the scra=
{
the trip actuator ce= tacts shall a:d A??.". Red Elec.k settings not exceed 50 milli-seconds.
given by equations in Specification 2.1.A.1 and 2.1.3 shall be calculated if r-v'- -
fraction of limiti:E power density.ezceeds the fractics of rated power.
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TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minim *m Number Modes in Hhich Function J
Operable Inst.
Trip Function Trip level Setting Must Be_ Operable Action
Channels per Refuel (7) Startup/ Hot Run Trip (I)'Svstem Standby
I Mode Switch in Shutdown X
X X
A
X X
A IRM
$ 20/125 of full scale X
X (Si A
3 High Flux
Inoperative X
X (5)
A APRM
High Flux (14) (15)
(17)
(17)
X A or 8
Inoperative X
X(9)
X A or B
High Flux (157.)
~>2.5 Indicated on Scale (11)
(11)
X(12)
A or B
Downscale 1157. of Design Power X
X (16)
A or B 1 085 psig X(10)
X-X
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A,
2 High Reactor Pressure i
High Drywell Pressure 12.5 psig X(8)
X(8)
X A
Reactor low Hater Level
>9 In. Indicated Level X
X X
A
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High Water level in Scram Discharge Tank 139 Gallons X(2)
X X
A
Turbine Condenser low Vacuum
>23 In. Hg Vacuum X(3)
X(3)
X A or C
$ X Normal Full Power
2 Main Steam Line High Radiation Background (18)
X X
X(18)
A or C
Main Steam Line Isolation C
Valve Closure 1101. Valve Closure X(3)(6)
X(3)(6)
X(6)
A or C ZO
Turb. Cont. Valve Fast
->150 psig Control Oil
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O Closure Pressure at O 2 Acceleration Relay X(4)
X(4)
X(4)
A or D g
'k 1 07. Valve Closure X(4)
X(4)
X(4)
A or D
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4 Turbine Stop Valve-Closure OF AeRM high riu scram setpoint <(.65w. 55)
rRe Two recirc. pump operation r1 Amendment No. 86 MFLPD
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KO7ES TOR TABLE 3.1.1
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1.
There shall be two operab'le or tripped trip systens for each function,
.
If the minimum number of operabic instru=ent channels per trip.ny. tem s
cannot be met for both trip systems, the appropriate actions listed below shall be taken.
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A.
Initiate insertion of'operabic rods and complete insertion of at:
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operab1'c rods, within four hours.
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B.
Reduce power 1cvel 'to IrJ! range and place mode switch in the
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startup/ hot standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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C.
Reduce turbine load ar.d close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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D.
Reduce power to less than 4S*/. of design.
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Pernissible to bypass, with control rod block, for reactor protection system'rcset in refuel and shutdown positier.s of the reactor r.ede switch.
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Permissibic to bypass when reactor pressure is < 600 psig.
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Permi.csihic to 5>ypass when turbine first stage pressure is less then 305 psig.
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5.
IrJi's are bypassed when lir?.:-;':r crc onseale cr.d the rc..cter e.de suitch yg is in the run position.
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6.
The design permits closure of any tra lines without a scram being initiated.
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L' hen the reactor is suber$tical, fuel is,in the reactor vessel
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and the reactor water tcuperature is it.cc than 212*T, only the
- following trip functions need to be operab.le:
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A.. !! ode cwitch in shutdown
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11anual scram
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C.
Iligh flur. IRH
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Scram discharge volu.se high Icvel E.
ArrJ: (15%) high flux r. cram l8
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Not required to be operabic when primary containment integrity is not required.
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9.
Not requirca while pcrforming low pouc'r. physics tests at atr.iospheric l
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pressure during or after refueling at pnect Icvels not to exceed 5
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NOTES FOR TABLE 3.1.1 (CONT'D)
10.
Not required to be operable when the reactor pressure vessel head is not bolted _to the vessel.
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11.
The APRM downstale trip function is only active when the reactor mode switch is in run.
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12.
The APRM downstale trip is automatically bypassed when the IRM instrumentation is operable and not high.
13.
An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there,is less than 50% of the normal complement of LPRM's to an APRM.
,
14.
W is percent of drive flow' required to' produce a rated core flow of 69 Mlb/hr.
Trip level setting in percent of design power (1998 Mrit).
15.
See Section 2.1.A.I.
16.
The APRM (15%) high flux scram is bypassed when in the run mode.
17.
The APRM flow biased high flux scram is bypassed when in the refuel or
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startup/ hot standby modes.
L 18. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection test with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the test.
The background radiation level and associated trip setpoints may be adjusted
.I during the test based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.
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Amendment No. 86
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UNCONTROLLED CO Lif:iTI*:1 CC:"DITIO*: l'OR 01 r -IATIOTJ StrmT1htfJ:t'}: UT'J'.!IREP.r!!T
m 32 FeECTnT I::SEi'*U:TATION 4.2 JUoTvrTnT Ii!r:7t':T::TATm!:
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Atpli cstili ty:
Annliectility: - t
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Ar;11c: to the plant instr.u-Applic: to the surveillence centation which initiates'and requirc ent of the ir.:tru-controls a protective fur.ctior.
ten'ation t*.nt initiates nna
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controls protective functicn.
Objective:
Objective:
To assure the operatility of To specify the type and frc-
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prote:tive instrumentation.
. quency of survcillar.:e to te
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applied to protcetive instru-cer.tation.
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Spe:'ficatiens:
Specificttions:
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A.
lYi er.- Contain ent Isolation A.
Pri crv Contain. ent Iselatien Pur.e ticr.s Itt:tions L' ten pri.rary ocntainment inte -
Instrumentation shall be .:n:-
'rity is required, the lititinc tionally tested cni calitr:tci condition: of operation for as indicated ir. Table 4.2. A.
)
the ir.ctru entation that ini-
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tiates prir.ary containment Systen Icgic shall be fun: tion-isoittion are given in Table ally tested as indicated in !Lile 3.2.A.
4.2.A.
B.
Core ar.d Containment Coolinq 5.
Core end Centtir er.t Cc:lin-Sys e-s - Initiation & Control Eyc t er.: - Initittior..L Con.rcl The lititinc conditions for Instrur:entation shr.11 're fun:-
operstion for the instrumentn-tionally tested, calibrat:i and tion that initiates or con-checked as indicated in Tatic trols the core and containner.t 4.2.B.
coolin; systems are given in Tatle 3.2.B.
This instruren-Systee locic shall be function-
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tation ust be operable when ally tested as ir.dicated in the syste.(r) it initintes or Table 4.2.E.
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control: are rcquired to be operable as specified in Sec-tion 3.5
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LIP 3 TING CONDIT10N FOR OPERATION SURVETLLANCE RE0JIREMENT C.
Control Rod Block Ac?uation C.
Control Rod Block Attua?lon
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The limiting conditions of l.
Instrumentation shall be operation for the instrumentation functionally tested, calibrated that initiates control rod block and checked as indicated in Tatie are given in T,able 3.2.C.
l 4.2.C.
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2.
The minimum number of operable Syster logic shall be
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instrument channels specified in functionally tested as indicated Table 3.2.C for the Rod Block in Table 4.2.C.
Monitor may be reduced by one 1.n one of the trip systems for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in
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any thirty day period.
D.
Radiation Monitoring Systers -
D.
Radiation Moaitoring Systers -
Isolation & Initiation Functions Isolation & Initiation Fur.ctions 1.
Reactor Building Isolation and 1.
Reactor Building Isciation and
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Control System and Stans:y Gas Centrol System anc Stans:y Gas Treatment System Treatment System The limiting conditions for Instrumentation shall be operation are given in Table functionally tested, calibrated 3.2.D.
and checked as indicated in Table 4.2.D.
't System logic shall be functionally tested as indicated in Table 4.2.D.
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Amendment No. M 89
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UNCONTROLLED COPY
SURVEILLANCE RE001REE NT LIMITING CO c1 TION FOR OPERATION E.
Dryuell Leak Detection
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Dryeell teat Detection The limiting conditions of
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Instrunentation shall be o;:eration for the instrumentation functionally tested, calit-ated and Checked as indicated in Tatie that monitors drywell leak 4.2.E.
detection are given in Table
.
3.2.E.
F.
Surveillante Inforcation Readouts F.
Surveillance Inforr.ation Feat:uts
The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and Checked as indicated in Table 4.2.F.
surveillance infctmation readout.s are given in Table 3.2.F.
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Amendment No. 89
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UNCONTROLLED COPY
LIMITING CONDITIONS FOR OPERATION SUR'.'EILLANCE REQ'JIREMENTS G.
Recirculation Pump Trip / Alter,
G.
Recirculation Pcmp Trip /
nate Rod Insertion Initiation Alternate Rod Insertion
,
This system is only required Surveilianceforinstru-when the reactor mode switch
centation which initiates.
,
is in the RUN mode.
Recirculation Pur.p Trip and Alternate Rod Insertion The recirculation pump trip shall be specified in Table system causes a pump trip and 4.2-G.
the alternate rod insertion system provides for initiating ~
control rod insertion on a
-
signal of high reactor pressure or low-low reactor water level when the mode select switch is in the RUN mode.
The limiting conditions for operation for the instrumenta-tion are listed in Table 3.2-G.
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Amend ent No. fi, 62
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TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument
Channe l s. Per,J_r,l p,,Sys t ert. (1)
Trly function Tripl evel Setting Remarks
Reactor low-low Hater at or above -49 In.
1.
In conjunction with low Level.
Indicated level (4)
Reactor Pressure, initiates Core ',piay and LPCI.
2.
In conjunction with High
-
Drywell Pressure. 120 second time deley and LPCI or Core Spray pump interlock initiates Auto Blow'down (ADS). ~
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3.
Initiates HPCl; RCIC.
.-
Initiates starting of Diesel Generators.
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~~Eeactor High Hater Level
<+48" indicated Trips HPCI and RCIC turbines.
level I
' ' -Reactor Low Level
>307" above vessel ~~~
Prevents inadvertent operation (Inside shroud)
zero (approximately of containment spray during 2/3 rore height)
accident conditto..
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2 Containment High Pressure i < p < 2 psig Prevents inadvestent operation CE:
of containment spiay during 22:
accident condition.
O (I) CZ)
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.
PHPS
.
TABLE 3.2.5 (Contd)
INSTRtfMENTATION THAT INITIATES OR CONTROLS Tile CORE AND CONTAlfetENT COOLING S
.
Minimum # of Operable Instrument Channels Per Trip System (1)
Trip Function Trip Level Setting
'
Remarks
High Drywell Pressure
$2.5 psig 1.
In tlates core Spray; LPC
,
llPCI.
2.
In conjunction with Low-Low Reactor Water Level.
- 120 second time delay and LPCI or Core Spray pump running, initiates Auto
'
.'
Blowdo'wn '(ADS).
3.
Initiates starting of Diesel Generators.
.
I Reactor Low Pressure 400 psig + 25 Permissive for opening core Spi and LPCI Admission ' valves.
.
1 Reactor dow Pressure 1110 psig In conjunction with PCIS signa:
E,
$
permits closure of RilR (LPCI)
N injection valves.
v.
- 1 Reactor Low Pressure 400 psig j; 25 In conjunction with Low-Low
.
Reactor Water Level initiates g,
Core Spray and Li-CI.
.
O C) O
-
(2) -g Reactor Low Pressure 900 psig j; 25 Prevents actuation of LPCI i
.T Z break detection circuit.
4 O
-
r-.
r-
.
FT1
,
O
-
,
____ ___ _.
..
_ _
_
_
,,
.
s
-
,
?N?S i
TABLE 3.2.b (Cont'd)
IKSTRUMENTATION 'DIAT INITIA_W.s C.t CON'i.ROLS Trid CORE AND CONTAINMENT COOLING SYSTEMS
.
Kinimum i o'
.
!
Operable Instrumant l
Channels Per Trip System (1)
Trip 7 unction Trip Level Setting Remarks
.
g
Core Spray Pump Start In conjunction with loss of i
Timer 0 < t < 1 sec.
power initiates sequentiel
LPCI Pump Start Timer 4 < t < 6 sec.
starting of CSCS pumps
LPCI Pump Start Timer 9 < t < 11 sec.
!
Auto Blowdown Timer
> 90, <120 sec.
In conjunction with Low Low
!
4'
,
~
Reactc-Water Level, High Dr)
well P. essure and LPCI or Col
.
Spray Pump. running interlock,
,
)
initiates Auto Blowdown.
I I
RilR (LPCI) Pump Discharge 150 + 10 psig Defers ADS actuation pending i
Pressure Interlock firmation of Low Prensure con cooling system operation.
(1
Core Spray Pump D14 charge 150 + 10 psig or Core Spray Pump running ir i
Pressure' Interlock lock.)
i t
!
l
Emergency Bus Voltage 20-25% of rated 1.
Permits closure of the I g
Relay voltage remets Generator to an unloaded at less than 50%
emergency bus.
,
2.
Permits starting of CSC!
4 kV motors.
CZ C>
-
i C
.O)
OZ4
'D
,4.~.3
\\
1
-
r-r-
,
,
!
(2)
- ~
x-
.
.c +!'
. ;. !
.
r.
e
. >
.
.
.
PNpS TABLE 3.2.5 (Cont'd)
.
,
INSTRlHENTATION TilAT INITIATES Olt CONTROLS Tile CORE AND CONTA1HHENT C00I.1M SYSTEMS Hinimum I of Operable Instrument g
Channels Per Trip System (1)
Trip Function Trip level Setting Remarks
Startup Transformer OV with 1.1 Sec 1.
Trips Startup Trans'former lone of Voltage Time Delay to Emergency Bus Breaker. :j 3094V with 18 Sec 2.
Locks out automatic closure Time Delay of Startup Transformer to
,
Emergency Bus.
3.
Initiates starting of
' ' Diesel Cc~nFrators 'In conjun-tion with loss of
.
auxillary transformer.
4.
Preventh.stmultaneousstart-
-
ing of CSCS components.
'
5.
Starts load shedding logic g-for Die.sel Operation in con-e
-
junction witti (a) Low Low es E.
Reactor Water Level and Lnd E
Reactor Pressure or (h) liikh N
tirywell pressure or (c) Core x
Standby Cooling System com-ponents in service in con-Junction with Auxiliary e.
"C::
Transformer breaker open.
- 22:
'
C7)
'
( ) (2)
1.
These trip setroints define the ranp.c of trip acttings
"{
selected from,the appropriate relay curve
-< o
.
-
-
.-
,
FT1
.
(37
,.
.
_
___
_m
-
--
--
x
.
.
e
^-
Illll
,
,
'?!I ld!
I l
.
.
.
.
'
ritrS
.
!.
1HSTRIRIENTATION TIIAT 1MITIATf:S 04 CONTROL.S TilC COME AND CONTAliNitNT C TABLE 3.2.B (Cont'd)
I
,
.
.
Plinimum I of Operable Instrument Channele Per Trly System (I)
Trip runction Trip I4 vel Setting Remarks
"
Startup Transformer 3743Y + 22 with,
l.
Trles Startup Transforse'
,
I Degraeled Voltage Relays 9.2 + 0.5 sec.
r to Emergency Bus areaker.
time'deley'
.
"' '
2.
Lncks eist entomatic closus 1,2,3 & 4
,
of Startup Transformer to
127A-f>04 Emergency Dus.
Rus A6
.
g y y g4 3.
Initiaten starting of
,,
-
-
Dicr.el Cenerators in
conju'nction witti lone of
eum111ary transformer.
.
i
Prevente simultaneous starl
,
Ing af CSCS congeonents.
.
.
5.
. Starts load slied.I!ng Ingic for Diesel Operation in con g.
Junction wille (n) Low Imv
.
.
-
Reacsor Water Level and Low Ps Heactor Pressure or (b) IIIc
'
g--
drywell pressure or (c) Car
,
Standby Cooling System coe-32.'
"
ponents In service in con-(q) 9
,
..
,
?
O z(-) 2
. Transformer breeher open.
.
Junct ion witle, Aus t11ery 9--i
.
.
o..
D --
' dC )3
-
.- C
r--
e, r~~
c>
Frl
"
'
(Cl
.
.
t
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__
__
~
.
-
TABLE 3.2.B (Cont'd)
INSTRilMENTATION T11AT INITIATES OR CONTROLS Tile CORE AND CONTAIN Minimus # of Operable Instrument
_ Channels Per Trip System (1)
.
Trip Function Trip Level Setting Remarks
,
RilR (LPCI) Trip System
NA bus power monitor Honitors availability of power to logic systems.
- Core Spray Trip System NA bus power monitor Honitors availability of power to logic systems.
ADS Trip System bus NA power monitor Monitors availability of power to logic system's~and val'ves.
IIPCI Trip System bus
tu power monitor Honitors availability of power to logic systems.
'
,
f RCIC Trip System bus NA power monitor Monitors availability of power
-
to logic systems.
Recirculation P mp A d/p
<2 psid Operates RilR (LPCI) break de-
,
,
tection logic which directs U
Recirculation Pump B d/p
<2 paid cooling water into unbroken E
recirculation loop.
II Recirculation Jet Pump 0.5<p<1.5 psid
$
Riser d/p A>B n
fh
Core Spray Sparger t
Reactor Pressure Vessel-1( 1.5) psid Alarm to detect core spray g
d/p sparger pipe break.
,
w
<
O
. OO.Zi
.'O D 4 O
.
-
rF a
O
,
O
.
.
.
.
...
...
.
.
.
.
_. _.
C
~
.
.
e
.
PNIS
,
TABLE 3 2.B (Cont'd)
It:STP.'.;!T.TATION THAT INITIATES OR C0:m'0LS THE C0!1E A*TD C0!.7AINKENT CCOLIiG SYSTD'S
,.
'
Pini=un ! of Operable Instru.9ent
-
'
Chanc.els Per Trip System (1)
Trip Function Trip Level Setting Ferath
Condensate Storce,e Tank 2 18" above tank ero Provides interlock to !!PCI pump
,
Lov I,evel suction valves.
- Supp essior. Chamber d l'll" below torus zero
-
High Level
-
,
1 BCIC Turbine s'tcas Line s 30C'y, of rated stess flov
'
(2)
High Flow
.
.
.
a,
RCIC 'Nrbine Compartment 6170'F (2)
[
.
- 'all
!
a
.
,2 Torus Cavity Exhaust Duet $150*F (2)
j l
-
BCICVe.1 vesta 51onArea
$200'F (2)
,
'
k*all
-
.'
s
,
Is'(5)
RCIC Steca Line Lo-Press 100> P3 50 psig (2)
f
HPCI 'ibrbine Stcan Line g 300$ of rated flov (3)
.
'*
High Flow
{
HitI Turbine Compartment
$17b'F (3)
Exhaust Ltcts
,
b~
Toi as Cavity Exhuirt. Ibet 190 - 200 F (3)
.
HPCI/?'? '!alve Station 6170'F (3)
-
C O
.2'
r
'
O O
Area Extaust Duet
-
a Oq
.'U Z 15 (5)
!! pct St.eas Line Low Ivess.
100 > P 350 psig (3)
-< O
-
r-F rri
-
O
.
s'
-
.
v
.
.
I
'
1:0TES FOR TA1C 3.2.B
'
D
~
s
-
1.
Whenever any CSCS subsystem is required by Section 3 5 to be operabic,
'
there shall be two (f;ote 5)' operab'c trip syster.s.
If the first col-u en conr.ot be cet for one of the trip r.ysten:s, that sys tem shall be repaired or the reaeter shall be placed in the Cold Shutdown Condition
,
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after this ; trip syster. is c:cdc or found to 1c inoperabic.
.
.2.
Close isolation valves in RCIC subsystem.
_
Close isolation valves in !!PCI st.brystec.
la. Instruz.cnt set point corresponds to 70.5" of ac'tive fuel.
-
RCIC and ErcI have only one trip system for these sensors.
.
.
e
.
e
.
.
..
-
L
.
'I
-
.
.
-
O
-
.
.
)
.
.
.
.
.
.
.
/k V
Pr
.
UNCONTROLLED COP.Y.
'
,
,
y v~
u
- ~ -
.-
'
.
'
.:
.
.
.o PMpS
.
TADLE 3.2.8.1
,
,
INSTRifHr.NTATION T1IAT HONITORS FNF.RCF.NCY Btf5 VOt TACE
,
'
.
.
.
-
Minimise f af i
Operable Instrument Channels Per Trip System function _
Setting Remarks
'
,
rmerr.cncy 4160V Runen AS 305612% with Alerts Operator'
!
& A6 per.raded Voltar.e
-
9.2 a 0.1 1 to possible degraded Annunciation Relaya (1)
Second Time Delay voltage conditions 177A-A5
Bus A5 (
I&2
..
,
127A-A6 g
Bus A6 1&2
'
.
,
-
.
.
-
.
-
.,
,
.
E"
.;
.
?
'
.
k
-
)
O,*
.
!
a C
(1) In the event that the alarm system is determined Inoperable, conenence logging safety h
related bus Voltage every 1/2 hour untti such time as the alann is restored to operable g
'
Oz status.
_O
. U ~y
'
7J
.-< o
r-
.
F Fr1
-
O
.
,,
iS'372 *C C0"r:73C?:5 Tt9 077 47IcN s'.';tv7.32.tp::r pV2r'>TJ.7
t.
3,5 'co tr Ao eti::All.vr'.7 CN1.1Ni
,
k.$ c:nr R:n CT 71.Itr?.t.7 cD '. 3-
-
_ sis:u s sv:1ns A--i f e et!1t tv A r11:stilltv
'
A;;11es to the c;::stier.e.1 status of A;;1ies to the Surveillta:e Fe uire :..t:
the ccre e.nf s ;;ressica 7:21 cc:lird cf the c:re and su;;re: sic: Pc:1 coeli:q s ut sys t e=.s.
subsy:ter.s which are r. uired when the corresp..iir.; limitir,i Ccnf.itica for op-tration is in effect.
Ctfective
-
-
f:t$e:tive
.
.
To assure the c;eratility cf t'.e c:re 7o verify the c;erability cf the core a.i a_-1 : 7;ressien pool coolin; sutsyste=s su;;ressien pe:1 coelin; suts; ster.: u.ter snier sal canditie.s ter V1.ich this all ceniitic.s fer which this co: ling ca-co: ling ce;stility is as essentiel re-
- tbility is an essential res;
- nse to sta-s;ct.se te station ate: realities, tien ain:::.t.11 ties.
-
,
5 effleatien 5 eeffic:tien
.
A.
C:.e frrev r.no IJ:7 Setsvrie-s A.
ce-e ?--r _e '. W T _f 4r"'*--
.
1.
5:th ecre s; rey ratsy:t'e:s shr.11 1.
C:re 3; ray Subsyste: 7esti.=g.
- te c; err.tle yh::ever irra!.intei 7tes P.-eev e r.ev (
fuel is is the vessel a:1 y-ict t: re::t:r sti,.rt.:p frs: a Celi la. Sir;1sted Onte/07: stir.g
Cor.iitica, except as s;::ifici
,3;;,_,,,gg, cygg, in 3 5 A.2 belev.
p g;.,:,,g g,3 g,3g,
,
b.
Pa:; 0;erability on=e/==.th
-
.
.
c.
.%ter Operated C:ce/ce:th and
Talve operatility 0::e/ cycle fro =
the Alterr.ste Stutdevn Fanel
.
,
d.
Pe=p flev rate
.
Each pu=p shall deliver at le.ast l
3600 sps against
-
a systen head
'
.
correspondict<to a
,
res: tor vesstl-
.
.
pressure of 104 psig.
,
,
(
.
,
.
s.
Cete Spray Bender
,
A y Instru=entation
.>
A emd:est Ec. fl, 62
.
UNCONTROLLED COP.Y.
.
-
, _ -,,.
~
---m-,
-m
, - - -,,
e m-,
-
---n.,--.m-
-. -
m-,
- -, _. - - -
-
.
.
.
..
_
.
1 : ISO CC CITIONS TOR C?I7ATICH
-
5"TVI!LLANCE ISt'IT." INT
-
3.5. A Cere 5:rsy amd LPCI Suisysters 4.5.A Ccre 5; ray ard I?:1 Satsyste=2~
(c c=t' d)
~(cc::'d)
.
.
,.
'
check Dr.:e/ day
,
,,
,
Calit:ste On= e/3 ::-.22
-
.
Test 0=:e/3 : ::Es 2.
T:c a:d af:e th e da t e tha t c:e 2.
ne: 1: is deterzimed that c:e c::e of the cere spray suisyste=s is a;;ay suisys:e= is is:;eratie,
=a d e or f e =r.d t o b e i=: y e r abl e
.
the c;erable ec:e s;:ay sulsys:en, for tr.y rease=, ec=tinued reactor the L?CI subsyst a= ar.d the dies t1
-
c; era:ic: is per:issible duri 3 ge: erat::s shall be de===stra:ed to the sue:e edi:g s eve: days, pro-be c;erah3e i=:ediately. The eye:-
vided that derimg' such s eve: days alle cere spray sutsyste: shall be
,
a11'attive ce:pe e=ts of the other Ze==:strated to be c;e:atle daily
cere s; ay subsyste: a:d active the:erfter.
.
ec:ye:e::s of the L?:I suisyste=
,
-
s:d the diesel ge=e: ate:s a:e op-eratla.
3.
The 17:I Suisyste=s shall be cie:-
3.
L? I Suisys:c: Te sti:3 shall be as
alle whenever 1::adiated fuel is felipus:
-
is the rea:te; vessel, a:d ;;ic:
.
a.
Si=ulated Aute=2-0 :e/ Opera:ing O!
to rea:te startup f:c= a C)16 C4:ditien, ex:ept as spe:ified tic Actuatic: Test Cycle
~
1: 3.5. A. 4, 3. 5. A. 5 :d 3. 5.T.5.
' -
,
'
b.
Pu=p 0;eral:iliry 0=:e/ :sth
.
-
-
.
c.
Ik:c: Op erat ed on=e/Y.:::h a:d i
valve c;eraitlity Dn=e/ cycle f:ce the A1:ernate
.
Shutd: : pa:e1
-
d.
Fu=y Tlev 0::e/3 :::.:hs Three LPCI pu=7s shall delive
-
14,400 g;: a ga tr.s: a sys:e=
head ec::espendi:3 to a vessel pressure of 20 psis
,
.
.
.
-
.
.
.
-
.
(
Ase = d=e:t No.
i. 62
.
.
D
.
E UNCONTROLLED
~
COPY.
.
_ _ _.
..
.-
.
l l
,
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIEEMENT 3.5.A Cere Spray and LPCI Subsys'ters 4.5.A Core Spray and LPCI Subsysters (Cont'd)
(Cont'd)
.
4.
Tro= and af ter the date that one 4.
When it is detercined that one of of the RHR (LPCI) pumps is made the RHR (LPCI) pumps is inoperable or found to be inoperable for any at a time when it is required to reason, continued reactor opera-be operable, the containment cooling tion is permissible only during subsyste=, the retaining active com-
,
the succeeding thirty days pro-ponents of the LPCI Subsystem, both vided that during such thirty days core spray systems and the diesel the containment cooling subsystem, generators shall be demonstrated to
the remaining active components of
be operable immediately and the op-the LPCI Subsystem, and all active erable LPCI pumps daily thereafter.
components of both core spray sub-systems and the diesel generators are operable.
5.
When it la determined that the
-
5.
From and after the date that the LPCI subsyste-is inoperable, both LPCI subsystem is made or found to core spray subsystems, the contain-be inoperable for any reason, cent cooling subsyste: and the continued reactor operation is per-diesel generators required for missible only during the succeed-operation of such co:ponents if ing seven days unless it is sooner no external source of power were made operable, provided that during available shall be de=onstrated (
such seven days the containment to be operable immediately and cooling subsystem (including 2 LPCI daily thereafter.
pumps) and active creponents of both core spray subsyste=s, and the diesel generator s required for operation of such components if no external source of power were avail-able shall be operable.
6.
If the requirements of 3.5. A can-not be met, an orderly shutdown
'l of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
.
i
/
.
UNCONTROLLED
,
i
' COPY.
_
.
.
,v
.
.
LIpITING CONDITION FOR OPERATIONS SURVEILLANCE REQUIREMENT
_
- 3.5.B Containment Cooling Subsystem 4. 5. B Cont ai nmelnt Coolint_ Subsyste:
1.
Except es specified in 3.5.B.2, 1.
Containment cooling Subsyste Testing
.
3.5.B.3, and 3.5.T.3 below, shall be as follows:
both containment cooling subsystem loops shall be operable whenever Item Trequency irradiated fuel is in the reactor vessel and reactor coolant temper-a.
Pu=p & Valve Operability Once/3 conths ature is greater than 212 F',
and and Once/ cycle prior to reactor startup from a from the Cold Condition.
Alternate
~
Shutdown Station b.
Pump Capacity Test After pu=p Each RECCW pu=p shall r.aintenance deliver 1700 gp: at and every l-70 ft. TDH.
Each
-
3 months
~
SSWS pump shall de-liver 2700 gp: at 55 f t.
TDH.
!
c.
Air test on drywell Once/5 years and torus headers and 4*
nozzles
.
- Conditional relief granted from this LCO for the period October 31, 1980 through November 7, 1980.
Amendment No. 44
-
(
.
L
,
~
'
UNCONTROLLED
'
'
COPY-lt
-
"
.
,
,
i LIMITING CONDITION ;0R OPERATION SL'RVE11.1.ANCE PIQUIREMENT
- 3.5.B Containment Coc.11ng Subsystem 4.5.B Containnent Cooling Subsystee (Cont d)
(Cont'd)
_
2.
From and af ter the date that one 2.
When one contain.ent cooling subsystem containment cooling subsystem loop loop becomes inoperable, the operable is made or found to be inoperable subsystem loop and its associated
.
for any reason, continued, reactor diesel generator shall be demonstrated operation is permissible only during to be operable ir.rediately and the the succeeding seven days'unless operable containment cooling subsystem such subsystem loop is sooner made loop daily thereafter.
operable, provided that the other containment cooling subsystem loop,
.
including its associated diesel generator, is operable.
3.
If the requirements of 3.5.B can-not be met, an orderly shutdown
'.
shall be initiated and the reac-
-
tor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
HPCI Subsystem C.
HPCI Subsystem 1.
The HPC1 Subsystem shall be oper-1.
HPCI Subsystem testing shall be per-s I
able whenever there is irradiated formed as follows:
fuel in the reactor vessel, reactor pressure is greater than 104 psis, s.
Simulated Auto-Once/operatin and prior to reactor startup from matic Actuation cycle a Cold Condition, except as speci-Test fied in 3.5.C.2 and 3.5.C.3 below.
b.' Pu=p Operability Once/ month and Once/ cycle from the Alternate Shutdown Station c.
Motor Operated Once/ month Valve Operability and Once/ cycle
.from the Alternate Shutdown Station d.
Flow Rate at Once/3 months (
- Conditional relief granted from this 1000 psig LCO for the period October 31, 1980
- b through November 7, 1980.
e.
Flow Rate at Once/operatin 150 psig cycle Amendment No. 44
,
UNCONTROLLED COP.Y.
107
- _. _ _ _ _ _
_
_ _ _ _ _ _
3.5.c NPCI sudsystem (cont'd)
'4.5.c NPCI rubsyste= (cont'd)
-
e The EPCI p=p shall deliver at least L250 g;= fcr a sys.
-
'.
te: head correspondi 6 to a
,
rea:ter pressure of 10CX) to 150 psis.
-
'2.
Tro: and after the date that the 2.
When it is deter =ined that the HPCI
>
EPCI Subsyste= is esde or found to Subsyste: is in:perable the E IC, the
.
de inoperable for any reason, con-1PCI subsyste=, both core spray sub-
.
tinued, reactor operation is per-syste:s, and the AOS subsyste actua.
cissible er.ly during the succeed-tion logic shall be de:onstrated to ing seven days unless such subsys-be operable i=:ediately.
The ROIC tes is sooner made operable, pro-syste= and ADS subs >tte= logic shall vidin5 that during such seYen be de=o strated to de operatie eatly
..
days all active co:ponents of the thereafter.
ADS subsyste=, the RCIC syste=,
the 1PCI subsyste= and both core spray subsyste=s are operable.
'
,
.
3.
If the require =ents of 3.5.C can-
-
not be cet, an orderly shutdown shall be initiated and the reac-
.
tor pressure shall be reduced to or below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
.
'l
.
3 5.D Feacter Core Isolatien coelin 4.5.D Feeeter core Isolatten ceoline (ECIC) Subsyste=
(FCIC) Subsyste:
1.
The ECIC Subsystem shall be oper-1.
ECIC,Subsyste= testing shall be per-
_
able whenever there is irradiated for ed as follows:
fuel in the rea: tor vessel, the re-actor pressure is Ereater than 104 a.
Si=ulated Auto-0::e/c;erating psig, and prior to reactor startup matic Actuation cycle from a Cold Conditien, axcept as Test specified in 3 5.D.2 below.
~
b.
Pt=p Operability onee/r.:r.th and once/ cycle fro:
'
the Alternate Shutdown Station c.
Motor Operated Once/c: nth and Valve Operability On=e/ cycle frc:
the Alternste
.
.:
Shutdown Station
-
g
.
' '
108
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A=endment No.
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UNCONTROLLED
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3.5.0 Reactor Core Isolation Cocline. I,. 5 3 neneter cere 7:else er.'t.....
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TifffCTSubsysten (Cont'd)
_(n:c) sub:y:te- (c:n;.e,) z
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d.
T;ev Pate at Cr.:e/3 e.:.;3:
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1CCO ;sig
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E.:w Pate at cr.:ef:ye.g,
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150 ;sig cycle
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- .e EcIC ;=p iha.11 deliver a
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Icas t !.00 r,;: fer a sy:te. h::,g
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c=rres;cr.ti..; to a rea:::r ;.res.
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sure of 1000 tr 150 ;sig
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2.
Tre: a.d af ter the d.ste that the 2.
W.e.n it is deter:.ir.ed that the R:,e
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RCICS is cade or fcund to be inop-subsyste= is irc;erable, the h?:::
eratle for any reason, centi ued -
sha.2.1 he de::r.st sted t be c;.erstle rea:ter ;:ver c;eratica is per=is.
ir. edfiately a.-d veek.ly thereafter.
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sible er.ly during the succeeding
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seven d.r.ys ;r:vided that during
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rach seven days the )i?:*S is c;er-
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atie.
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If the requirements of 3.5.D cannot
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be =e:, an cederly shu:devn shall
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be ini:ia:ed and the reactor pres-
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sure shall be reduced to c; belev (
104 psig withi: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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3 5.I Aute:stie re-esrei:stica L 5.E Aute:2 tie t:- esret:a:1:..
_Syste: (AOS)
Syste: (A 5)
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1.
The Auter.atic De;ressurizatien Sub-1.
Durinc each c;eratir cy:le the
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syste= shall be c;eratie whenever
'fcileving tests sha.11 te ;erfernet there is irradiated hel in the on the A:S:
react:r vessel a.nd the reteter pres.
su. e is greater than 104 ysig and a.
A sir.: lated autcastie :c:::.:1cn yrict to a 3*.artup fren a celd Can-test shall be ;:rfer:ed ;riep to
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ditica, ex:ept as s;ecified in
. startu; af ter es:.h re.^.;eling cut-3 3.I.2 below.
age.
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b.
Vi:h the rese:cr at pressure,
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each relief valve phall be man-
ually opened until a c=rres;cnding
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change is rese:cr pressure or sain
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turbine bypass valve pest:icts
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indicata that stes= is fics.:ne fr:a the valve
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c.
Perform a test frem the alternate A=ced:ent c.
shutdevn acel to verify that the
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relief va ve solenoids actuate.
- Test shall be perfomed after each
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refueling outage prior to startup.
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gwernc;rs menr~raccrvurw.=.cwn...
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3 5.E 1.uto.ctic L* rescurization u.5.E Au orntic Le:-er.u-1:atica
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Syste (Ats) (Cont'd)
,Syste: (CS )
(cont'd)
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2.
Tros and after the date that one 2.
vnen it is deter =ined that one sals
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vr.1ve in the auto:stic depressur-of the ADS is inoperatie the ADS ization subsyste= is c:tde or found subsystem actuation Icgic for the other AL3 valves end the.'!?CI ss.b-
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to be ir.opernble for any reason,-
systen shall be dec:nstr:ted to be continbed reactor operation is per-
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operabic i=nedictely end at 1 cast
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cissible c:1y during the succeeding veck.ly thercaf ter until the valve seven days ur.less such valve is is repaired.
l sooner c.ade operatie, provided tha.t drinC such seven, days the EpCI subsystem is operable.
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If the requirements of~3.5.E 'ca'n-
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not be met, an orderly shutdown shall be initiated and the reac-ter pressure shall be reduced to
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at least 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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3 5.T P.ini :.= Lo.i P essure Ceolinc_
4.5.T yini== ic. p esru_e Coolin:
and Diesel C-ene 3 tor Avtil-and Iicsel Genert.cr Ave.il-
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stility ability
1.
During any period when one diesel 1.
vnen it is desc h ed that one dies generator is inoperable, continued generator is inop.: rabic, a.111cu pr reactor operation is per=issible s=e core cooli:;r and cc:tsin=ar.t c@.g tubsh.s 2.-dl.1 te deso:-
only during the succeeding seven.
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days unless such diesel generator.
. strated to be cperable i==ciic. ely is sooner cade operable, provided e.nd daily therer.fter.
In adit-isn,
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that all of the lo.: pressure core and contain:cnt coolin; subsystc=s
.the opert.ble diesel c:nerator sh$
and the rer.aining diesel generator be de=onstre.ted to be opers.ble in-
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shall be operable.
If this re-mediately and daily thercarter e.-i
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quirement cannot be net, an order-the ir. operable diesel is repaired.
lyshutdownshallheinitiated
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and the reactor chall be placed
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in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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2.
Any co=binatien of inoperable coa-
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ponents in the core ar.d contr.in-eent cooling systems shall not de-feat the capability of the re=cin-
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inc operable co:ponents to fulfill the cooling functions.
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Amendme
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L2MITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
3.5.F Minimum Low Pressure Cooling
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and Diesel Generator Avail-ability (Cont'd)
3.
When irradiated fuel is in ihe re-
- actor vessel and the reactor is in the Cold Shutdown Condition, both core spray syste=s, the LPCI and containment cooling subsystems may be inoperable, provided no work is being done which has the potential for draining the reactor vessel.
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4.
During a Aefueling outage, for a period of 30 days, refueling oper-ation may continue provided that one core spray system or the LPCI system is operable or Specification 3.5.F.5 is met.
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5.
When irradiated fuel is in the
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reactor vessel and the reactor is in the Refueling Condition with the torus drained, a single g
control rod drive mechanism may
be removed, if both of the fol-(
loving conditions are satisfied:
a)
No work on the reactor ves-sel, in addition to CRD re-moval, will be performed which has the potential for exceed-
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ing"the maximu= leak rate from a single control blade seal if it became unseated.
j b)
i) the core spray systems are
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operable and aligned with a suction path from the conden-sate storage tanks.
ii) the condensate storage tanks shall contain at least 200,000 gal-lons of usable water and the refueling cavity and dryer /
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separator pool shall be flooded to at least elevation 114'-0".
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UNCONTROLLED 2 COP.Y.
Amendment No. 39 l
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SURVEILLANCE REQ'JIREF.ENT LIMITING CONDITION FOR OPERATION 3.5.H Maintenance of Tilled Dis-4.5.H Maintenance of Tilled Discharge charge Pipe Pipe
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Whenever core spray subsystecs, LPCI The following surveillance require ents subsystem, HPCI, or RCIC are required shall be adhered to to assure that the to,be operable, the discharge piping discharge piping of the core spray sub-fro: the pump discharge of these sys-syste s, LPCI subsystes, HPCI and RCIC te=s to the last block valve shall be are filled:
filled.
1.
Every month prior to the testing of the LPCI subsyste: and core spray subsyste=, the discharge piping of these syste=s shall be vented fro the high point and water flow ob-s e rv ed.
2.
Following any period where the LPCI subsyste= or core spray subsyste=s have not been required to be oper-
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able, the discharge piping of the inoperable syste: shall be vented fro = the high point prior to the return of the syste= to service.
3.
Whenever the HPCI or RCIC syste= is (
lined up to take suction fro the y
torus, the discharge piping of the
'l HPCI and RCIC shall be vented fro:
the high point of the syste: and water flow observed on a monthly basis.
4.
The. pressure switches which conitor the discharge lines to ensure that they are full shall be functienally tested every m: nth and calibrated
every three months.
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A=endment No. 39
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UNCONTROLLED COP.Y.'
LIMITING CONMTIOf,5 FOR OPERATION SURVE!LL ECE RE00?REMENTS 3.7 CONTAttWENT SYSTEMS 4.7 CONTAftW NT SYSTEMS Applicability:
A p r.l i c a b i l i t y :
Applies to the operating status of the primary and secondary containment systems.
Applies to the pritary and se:ondary containment integrity.
Objective:
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Otjective:
To assure the integrity of the primary To verify the integrity of tne p-imary and secondary containment systems.
Specification:
Specification:
A.
Primary Centainment A.
At any time that the nuclear system 1.
a. The suppression chamber water
is pressurized above atmospheric level and tem;erature shall be pressure or work is being done which checked once per day.
has the potential to crain the vessel, the pressure suppression pool water volume and temperature shall be b. Whenever there is in31 cation :#
relief valve c;e aticr. or testing ~
maintained within the following limits which adds heat to the suo-e* cept as specified in 3.7.A.2 and pression pool, the pool 3.7.A.3.
temperature shall be continually monitored and also observed a.d a.
Minimum water volume - 84,000 ft'
logged every 5 minutes until the heat addition is terminated.
b.
Maximum' water volume - 94.000 ft'
.j c. Whenever there is indication of c.
Manimum suppression pool bulk temp-relief valve operation witn tr.e erature during normal continuous bulk tem;;erature of the power operation shall be 180*F.
suppression pool reaching 160*F except as specified in 3.7.A.I.e.
or more and the primary coolant system pressure greater than 200 d.
Maximum suppression pool bulk temp-
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'p'sig, an external visual exami-erature during RCIC, HPCI or ADS nation of the su;oression cna :er operation shall be 590*F, except shall be conducted be' ore as specified in 3.7.A.I.e.
resuming power operation.
e.
In order to continue reactor power d. Whenever there is indication of operation, the suppression chamber relief valve operation with the pool bulk temperature must be local temperature of the sup-reduced to 1 0*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
pression pool T-quencher reaching
200*F or more, an external visual f.
If the suppression pool bulk temp-examination of the succression erature exceeds the limits of chamber shall be conducted before Specification 3.7 A.I.d RCIC, resuming power operation.
HPCI or ADS testing shall be terminated and suppression pool e. A visual inspection of the cooling shall be initiated.
Suppression chanber interior, including water line regions,
If the suppression pool bulk shall be made at each major temperature during reactor power refueling outage.
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operation exceeds 110*F, the
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reactor shall be scrammed.
m Amenement no. 83 UNCONTROLLED COP.
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..si 3snst,tte,,t ggjq,gotsig 3.7 CONTA!hvENT SYSTEMS (C nt'd)
4.7 CONTAINMENT SYSTEMS (Cont'd)
h. During reactor isolation f. The pressure differential l
conditions, the reactor betmeen the drywell and pressure vessel shall be
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suppression chamber shall be depressurized to less than recorded at least once each 200 psig at normal cool down shift when the differential rates if the pool bulk pressure is required.
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t.sm;erature reaches 12C'F.
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g. Suppression chamter water l
1. Differential pressure between level shall te recorde: at the drywell and suppression least once each shift = hen chamber shall be maintained at the differential pressure equal to or greater than 1.17 is required.
psid. except as specified in j and k.
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j. The differential pressure shall te established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of placing the reactor in the run mode following a shutdown.
The differential pressure may be reduced to less than 1.17
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psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.
k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4)
y hours for maintenance activities on the differential pressure control system and during requirea operability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression enamter vacuum breakers.
1. If the specifications of Item 1, above, cannot be met, and the differential pressure cannot be restored within the subsequent (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours, m. Suppression chamber water level shall be maintained between -6 to -3 inches on torus level instrument which corresponds to a downcomer submergence of 3.00 and 3.25 feet respectively.
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Amenement so. 83 UNCONTROLLED CO.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS Primary containment integrity 2.
Interrated Leak Rate Testing
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shall be maintained at all times when the reactor is critical a.
The prica"ry containment or when the reactor water tem-integrity shall be demon-
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strated by performing an perature is above 212 F and a
Integrated Primary Con-fuel is in the reactor vessel
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excep't vnile per iorming "open tainment Leak Test (IPCLT)
vessel" physics tests at power in accordance with either levels not to exceed 5 Mw(t).
Method A or Method E, as follows:
Method A Perform leak rate test prior
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t initial unit operation at the test pressure 45 psig, P (45), to obtain measured t
leak rate L= (45), or Method B
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Perform leak rate test prior to initial unit operation at the test pressure of 45 psig, P (45), and 23 psig, Pe (23),
t to obtain the measured leak rates, Le (45) and L (23),
g g
respectively.
3.
The suppression chamber can be drained if the conditions as specified in Sections 3.5.T.3 and 3.5.T.5 of this Technical
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Specification are adhered to.
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I, 152B Am:ndment No. 39
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4.7.A. prir.2ry Contain.ent (cont'd)
~ Subsequent leak rate teste shall be perforr.ed without preliminary Icak detection surveys or leak repairs itaediately prior to or durinC the test,'at an initial pre::ure of approxir.2tely 45 prig for Method A and:23 psic for Method B.
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Leak repdir=, if necessary to permit integrated leak rate testing, shall be preceded by locci leak rate teaturerents where ;ossible.
The 1cah reat difference, prior to and after repair when ccrrected to the aTpropriate test pressures of h5 prig for Methed A and 23 p:ic for Method B shall be added to the fint.1 integrated 1 cat rate results.
Closure of the contain=ent isolation valves for the purpore cf the test chall be acec plished by the reans provided for ncrtal cperatien of the valves.
The test duratien shall not be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> fer integrated leak rate censurcrents, but shall be e:dended to a sufflicier.t period of.
tire to verify, by reasuring the qu.ntity cf air re:uired to return-to the starting point (or other rethods of equivalen't sensititity),
the validity r.nd accuracy cf the leah rate results.
b. Acceptance Criteria fer IFOLT w
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Method A
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'l The taxi::: allc.sble leak rate (Lo) shall not execed 1.0 reight percent of the contained air at 45'psig per 24 hcurs.
The aller.:able operational leah rate, L o(L5), which shr.11 be cet t
prior to resumption of p ver creration fc11 cuing a test (either as ceasured or fo11cring repairs End retest) shall not enceed C.751.-.
r Methed B
'Ihe allovable test leak rate L (23) shall not exceed the lesser t
value established as follows:
L (23) = 1.0 xIIe(23)
t
(Ic@5)
L (23) / I (45)dE 1.0
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or
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L(23)=1.0(P(23)
1/2 t
t P (43)
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where P (23) and P (145) are measured units er absolute pressure.
t s
i The allevable operational leak rate, Lto(23), which shall be cet
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prior to recu nption of power operation following a test (either as measured or follo. tin; repairs and ratest) shall * avaaad n " t. 6 3 UNCONTROLLED COPY.
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4.7.A Prinnrv Conteinment (Cont'd)
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Corrcetive Action for IPCLT Methods A and B If leak repairs are necessary to meet the allowable opcrational Icak
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rate, the integrated 'eak rs', test need not be repeated provided local leakage measurer.>ents are conducted and the Icak rate differences prior to and after repairs..when corrected to the test pressure and deducted from the integrated leak rate measurenents, yield a leakage rate value not in excess o.f the allowable operational Icak rate.
d.
Frequenev for IPCLT Methods A and B After tie initial pre-eperational leakage rate test, two integrated leak rate tests shall be perforned at approximately equal intervals between the major shutdowns for inservice inspection conducted at
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ten-year intervals.
In addition, an integrated leakage rate test shall be performed at the end of the ten-year interval, which may coincide with the inservice inspection shutdown period.
c.
Local Leak Este Tests (LLRT)
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Metheds A and B (i
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(1) Primary containment testabic penetrations and isolation valves shall be tested at a pressure 2;45 psig, except for the main steam line isolation valves which shall be tested at a pressure 2 23 psig, each operating cycle.
Ecited double-gasketed seals
shall be tested whenever the seal is closed after being opened and at least once each operating cycle.
(2) Personnel air lock door seals shall be tested at a pressure 2; 10 psig each operating cycle.
f.
Acceptance Criteria and Corrective Actien for LLRT Method A If the total Icakage rates listed below are exceeded, repairs and re-tests shall be performed to correct the conditions.
(1) Double-gasketed seals 10% Lto(45)
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(2) All testable penetrations and isolation valves 60% Lp (45)
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(3) Any one penetration or isolation valve except main steam line
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isolation valves 5% Leo(45)
ft
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4,7.A Prinary Containment (Con'd)
b (4) Any-one main steam line isolation valve 11.5 scf/hr 023 psig.
Method B
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Ifthetotalleakagerateskistedbelowasadjustedtoatest pres-
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sure of 23 psig are exceeded, repairs and retests shall be pe formed to correct the condition.
(1) Double-gasketed seals 10%L (23)
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(2)
(a) Testable penetra,tions and isolation valves 60% Lp (23)
3,17 (b) Any one penetration or isolation valve except steam line isolation valves 5%L,(23)
g (c) Any one nain steam line isolation valve 11.5 scf/hr 023 psig I
Leak rates measured at 45 psig shall be adjusted to a test pressure
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of 23 psig according to:
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LLRT(23)adj =LLRT x L (23)
n meas L (45)
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g.
Continueus Leak Rate Monitor k' hen the primary containment is inerted, the containment shall be
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continuously nonitored for gross leaka;e by review of the inertir.g system makeup requirer.cnts.
This r.onitoring system may be taken out of service for caintenance but shall be returned to service as soon as practicable.
h.
Dryvell Surfaces
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The interior surfaces of the drywell and torus above water line shall be visually inspected each operating cycle for evidence of deteriora-
tion.
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3.7 Primary Containnent 4.7 _Priesry Containnent
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Pressure Sucoression Cha.mber -
3.
Pressure SuEeression Cha.ber -
lteactor Buildin:; Vacutn Breakers Ite actor Evildine vacuu. E eaters
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I Except as specified in 3.7.A.3.I a.
a.
The pressure suppression chamber -
below, two pressure suppression reactor building vacu n breakers
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che.ber - reactor building vacuum and associated instrv,entation breakers shall be operable at a'11 including set point shall be che:ked
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times when primary containment in-for proper operation every three
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tegrity as required. The setpoint months.
of the differential pressure instru-
'rentation which actuates the pres-
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sure suppression chamber - reactor building breakers shall be 0.5.psig.
b.
Frce and after the date that one of a
the pressure suppression chamber -
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reactor building vacuum breakers is made or found to be inoperable for any reason, reactor operation is pemissible only during the succeeding seven days unless such
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vacuum breaker is sooner made operable, provided that the repair procedure does not violate primary contaircent integrity.
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4.
Drvwell-Pressure Suopression 4.
_Drywell-Pressure Sucoression
'CF~ amber Va:uu Breakers Cha-ber va:uu-Breakers I
a.
When primary containment is a.
Periodic Operability Tests
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required, all drywell-pressure
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suppression chamber vacu p
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breakers shall be operable
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except during testing and as
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stated in Specifications
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3.7.A.4.b c and d below.
Drpell-pressure suppression cha,ber vacuum breakers shall be considered operable if:
(1) The valve is demonstrated to (1) Once each month each drywell-pressure open with the applied force of suppression chamber vacuum breaker the installed test actuator as shall be exercised and the operability indicated by the position of the valve and installed position switches and remote position indicators and alams verified.
indicating lights.
(2) The valve shall return by (2) A drywell to suppression cha.ber dif-gravity when released after ferential pressure decay rate test
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being opened by remote or shall be conducted at least every 3, ranual means, to within 3/32".
months.
of the fully closed position.
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(3) Neither of the two position (
alam systens which annunciate on Panel C-7 and Panel 905 when any vacue breaker opening exceeds 3/32. are in ai,am.
UNCONTROLLED
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Amenement so. se COP.Y.
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LIMIT!rG CONDITION FOR OPERATION SURVEILLAN:E..EDUIREVENTS 3.7 Primary Containment 4.7 Primary Containment
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b.
Any drywell-suppression chamber b.
During each refueling outage:
vacuum breaker may be non-fully closed as determined by the ao$ltion (1) Each vacuum breater sha'l te switches provided that the d ywell tested to determine trat tne
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to suppression charter differential disc c;en, freely to tre t:.:-
decay rate is demonstrated to be not and returns to the closed greater than 25% of the differential position by gravity witn r.c pressure decay rate for the maximum indication of binding.
allowable bypass area of 0.2ft'.
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c.
Reattor operation may continue pro-(2) Vacuu. breater Ecsition vided that no more than 2 of the dry-switches and installe: ala c well-pressure suppression chamber systems shall be calibrated and vacuum breakers are determined to be fun tionally tested.
inoperable provided that they are secured or known to be in the (3) At least 25% of the va:vum closed position.
breakers shall be visually
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ins;e:ted such that All sacuum d.
If a failure of one of the two breakers shall have been installed position alarm systems inspected following every fcurth occurs for one or more vacuum refueling outage.
If deficien-brea6ers, reactor cperation may cies are found, all va:uum continue provided that h breakers shall be visually differential pressure decay rate inspected ar.d deficien:ies test is initiated immediately and cc rected, performed every 15 days thereafter until the failure is ccrrected.
(4) A drywell to suppression cha-ter The test shall meet the leak rate test shall de ;? strate re:uirements of that the differential pressure Specific 6 tion 3.7.A.4.b.
de:ay rate does not e=:eed the rate which would occur thrc.;h
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a 1 inch orifice without the addition of air or nitr:;en.
5.
Orycen Concentration 5.
Orycen Concentration a.
The prirary containment atmosphere The pricary centain ent o.);e.
shall be reduced to less than concentration shall be measured 4% oxygen by volume with nitrogen and recorded at least twice gas during reactor power weekly, operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.S.b.
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aenement no. 87 N0m0MD COP.
LIMIT!?;G COND1710f; FOR OPERATION SURVEILL ANCE REOUIREf'E'.TS 3.7 Primary Containment 4.7 Primary Cca.tainment
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b.
Within the 24-hour period
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subsequent to piacing the reactor
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in the Run mode following a'
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shutdown, the containment atrosphere oxygen concentration
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.shall be redu:ed to less than l
4% by volume and maintained in this condition.
De-inerting ray commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.
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If the specifications of 3.7.A cannot be met, an orderly shutdoven shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Amenement no. 87 UNCONTROLLED COP to.....
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3.7. A PR!FARY COS7A7h?.rhi 4.7. A pFIP).RY C057AIhMh7
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7. contnineent Ateesphern Dilutirn 7. Centein=2nt Aters;here Dilutien a. Within the 24-hour period a. ne post-LOCA containt.ent after placing the reactor at : sphere dilution system in the Run Mode the post-shall be functionally LOCA Centaineent Atcos-tested once per operating phere Dilution System must cycle.
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be operable and capable of
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supplying nitroEen to the containcent fer atsosphere dilution. If this speci-
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fication cannot be met, the system must be restored to an operable condition within 30 days or the
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reactor cust be at least in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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b. Within the 24-hour period b. The level in the lic;uid N after placing the reactor stcrage tank shall be re y
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in the Run Mode, the corded weekly.
NitroEen Sterage Tank shall
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contain a minicum of 1500 gallons of liquid N ' II
this specification cannot be cet the r.inic:um volume vill be restored within 30 days or the reactor a:ust
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be in at least Het Shut-devn within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c. There are 2 H, analyzers c. The B analyzers shall y
available to Ierve the be tested for operability dryvell, once per centh and shall be calibrated once per 6 cenths.
With only 1 H, analyzer
crerable, reactor oper-ation is allowed for up to 7 days.
If the in-operable analyzer is not z.ade fully operable within d. Once per menth each c.anual or, 7 days, the reactor shall pcver operated valve in the be in at least Hot Shutdown CAD system flow path not within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
locked, sealed or othervise secured in position shall be
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With no B, analyzer oper-observed and recorded to be able, reattor operation in its correct positten; is allowed for up to 48 hcurs. If one of the in-
~operable analysers is net made fully operable within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be in at least Hot Shutdown
,j vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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Amencent No. 55 UNCONTROLLED'
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[' T-Cer.tr6TK:: ks F31tration systen
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Standby Gas Treatment Syrtem 1.
Standby Gas Treatment Systea a
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a. (1.) At least once every 18 Exte;t as specified in a:nths, it shall be g. a.
3.7.B.1.c below, both trains '
de.:nstrated that pressure
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of the standby gas treatment drop acress the cc.bined system and the diesel genera -
high efficiency filters ters required for cperation o.f and charc:a1 adscrier banks
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su:h trains shall be operable'
is less than 6 inches of
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at all tires when secondary water at 4000 cfm.
contain-ent integrity is required or the reactor shall (2.) At least once every 18 be shutd:wn.in 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.
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c:nths, de.:nstrate
that the inlet heaters b.
(1.) The results of the in-en each train are place cold DOP tests on c;erable and are ca;atle FEFA filters shall show of an output of at least
> SSt DDP removal. The 14 kk'.
Farform an 7esults of halogenated ir.strJ.ent functional
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hydro: art:n tests on test en the hu.idistats charcoal adsorber banks centro 11ing the heaters.
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shall sh:w > 95t halogenat-
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(3.) The tests and analysis of
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Specification 3.7.E.1.b.2
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(2.) The results of the shall be perfemed at least laterat:ry carbon saeple ente every 1E c:nths or analysis shall shew >951 falic.cing painting, fire rethyl iodide recevaT at cr chemical release in any
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a velccity within 10% of ventilation zene cc-.;nicat-g/
system dasign 0.5 to ins with the system while,
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1.5 e;/m3 inlet cethyl
[ the system is c;eratir.g that icdide cencentratien.
- c:uld centatinate the EEPA
>70t R.H. and >1903F.
fil'ters or cherccal adscrbers
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Frcs and after the date that (4.) At least cnce every 18 one train of the Standby Gas n:nths. autematic Treatrent System is made or initiation of each branch fcund to be inoperable for of the standby gas any reason, centinued treatment systa shall reacter operation or fuel be demonstrated. with handling is permissible only 5;ccificaticn 3.7.B.1.d
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during the succeeding seven satisfied.
days providing that within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and daily thereafter.
(5.)Eachtrainofthestandby
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all active com;:nents of the gas treat-ent syste shall
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other standby gas treateent te c;erated fer at least
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train shall be de::nstrated 15 minutes per c: nth.
to be operable.
(6.) The tests and analysis of d.
Tans shall operate within Specification 3.7.6.1.b.(2)
410% of 4000 cfm.
shall be perfor. ed after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system
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operation.
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- Conditional Relief granted from
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Amendnent No. J0'. M, 52, UNCONTROLLED COPY.
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Except as specified in b. (1.) In: lace ccid D00 testing 3.7.B.1.c. both trains of shall be perforced on tne the standby gas treatment HEFA filters after each
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syster shall be operable ces;1eted or partial
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during -fuel handling re;.acemer.1 cf the HEPA
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operations.
If the filter bank and after any
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system is not operable structural cair.ter.ance en fuel nove.ent shall not the HEFA filter syste:
6e started (any fuel housing which c uld affect
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asser.bly mosement in the HErA filter bank
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progress may be co..pleted).
bypass leakage.
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(2.) Halogenated hydrocarb:n testing shall be perferred ce the chartcal adsorber
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bank after each partial or cceplete re;1acement of the chart a1 adscrber
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bank or after any struc-
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tural eninter.ance on ths charcoal adscrter heusing
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A:end=ent No. 51
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Control Room High Efficiency Air 2.
Control R:c:- High Efficiency Air
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Filtration System Filtration Systen
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At least once' every 18
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Cxcept as specified in a.
Specification 3.7.B.2.c menths the pressure drop
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below, bath trains cf the *
acrcss each contined filter ControlkoonHighEfficiency.
train shall be de onstrated
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Air filtration System used to be less than 3 inches of i
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for the processing of inlet -
water at 1003 cfm.
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air to the cont.rol room under eccident tenditions and the b.
(1.) The tests and nr.alysis of.
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diesel generator (s) required 5;ecificatien 3.7.E.2.b for cperation of each shall be perferred once train of the system shall every 18 conths or
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be opecable whenever secondary folicwing painting, fire contair. cent integrity is
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or chemical release in re:;uired and during fuel any ventilaticn zone handling operations.
cc-unicating with the. *
sysiel while the system
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is o;erating.,,
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('1.) The results of the in-
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place cold DOP tests on
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FEFA filters sht11 show (2.) Inplace cold C P testing
199" DDP removal. The shall be pufemd afte results et the halogenat-each cc.plete or partial I ed hydrocarbon tests on replacemen't of the HEPA charcoal adsorber banks filter bank or after any shall show >99t halogenat-structural r.aintenance
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ed hydrocarten removal on the system housing whtn test results are which could affect the
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extrapolated to the HEPA filter bank bypass
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initiation of the test.
Jeakap.
(2.) The resu1*I of the (3.)Haiogenatedhydrocarbon laboratory carbon sample testing shall be performed
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analysis shall show >95.,
after each cc plete or
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rethyl iodide removaT at partial replacement of
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a velocity within 10% of
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the charccal adsorber
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syste design 0.05 to ba.k or after any
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0.15 rg/c3 inlet methyl structural mainte.ance
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icdide concentration, on the system housing 1 ". R.H., and >1250F.
which could affect the
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Tre: and after the date that
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lea ag *
one train of the Control Room High Efficiency Air Filtration (4.) Each ' train shall be System is made or found to be operated with the heaters
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incapable of supplying filtered in automatic for at least
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air to the control room for any 15 minutes every conth.
reason, reactor operation or refueling operations are (5.) The test and analysis of permissible only during the Smificaticn 3.7.E.2.b.(;
succeeding 7 days.
If the shall be perfereed after system is not made fully oper-every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system
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able within 7 days, reactor operation.
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- Conditional Reiter granted from
this LCO for the period February 5.1952 to startup for cycle 6.
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Amendment No, pf. Jf. 52 UNCONTROLLED
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s'hutdewn Yhall be the following shall be
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initiated and the reacter de-onstrated:
shall be in cold shutdown
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cithin the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and irradiated feel handling operations shall be terr.inated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. (Fuel hahdl-ing operations in progress r.ay be c:cpleted).
(1)
C;erability of heaters
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at rated power.
d.
Fans shall operate withiri 1,10 of 10C0 ci-.
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Perf:rm an instr; er.t fur.cticnal
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ATTACHMENT 2 Facility Comments on Written Examinations Made After Exam Review The following represents comments made by the facility as a result of the current exam review policy.
Only those ccmments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not listed.
i.e.: typo errors, relative acceptable terms, minor set point changes.
SRO EXAMINATION Question 6.01 e
(1) D/G lockout trip does not trip the D/G, it only opens the D/G output breaker (2) A local trip handle on the D/G will trip the D/G and is another method.
Question 6.05 b.
t 55 psig will trip the scram air dump valve, but is not an automatic scram signal. An auto scram would be due to closure of steam jet air ejector steam supply valve, causing a loss of condenser vacuum, which is an auto scram at 23" Hg.
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j Question 6.10 a.
The refueling refe ence text objective SLO 3 states:
" List the condi-tion... for all of the trips and interlocks associated with the Refueling System." This question asks for the condition and setpoints which ener-gize lights which are not trips and interlocks.
Specific setpoints for these lights were not taught or stressed in class.
R0 EXAMINATION
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Question 2.01 b.
Alternate answer:
Prevent drawing vacuum in exhaust line and drawing water into the turbine, causing water hammer on initiation.
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Attachment 2
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Question 2.02 a.
Also an alternate answer is drywell coolers, two of which are dedicated for the recirc pumps.
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Cooling units VAC-206A and B discharge air around the recirculating pumps and motors.
Question 2.04 b.
DC control power.(D-4(5) is supplied to the air start valve solenoids through auxiliary panels 103A (104A).
Students may have interpreted
question to mean "how can the diesel engine be started with a loss of control power?" This interpretation will be answered using Procedure 2.4.143, Attachment C.
This procedure states "using pencils or short rods, depress actuator buttons on top of air solenoids."
Question 2.08 a) 1.
Answer key from question "a" is different than the answers available in this question. Students may interpret Q #1 to be answered as "150 F"
since the real number should be 190 F as referenced in Reactor Water Cleanup reference text page 4-1, and 150 *F was closest to the correct answer.
Question 2.09 b.
Recirc pump motor B is tripped, but not to ensure that water is injected into loop B (the discharge valve is shut for that reason per Residual Heat Removal Ref. text page 13, paragraph 2). Students may not include recirculation pump as an answer. Request that recirculation pump trip not be required as an answer.
Question 3.03 a.
Dependirg on which stop valves close, you may get no action or a half-scram. Fequest this statement should be acceptable for full credit.
Question 3.06 b.
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Attachment 2 Question 3.09 Consider when grading:
Any hoist may be broken down into:
' fuel grapple
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frame mounted hoist
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trolley mounted hoist
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Also " power off to the refuel bridge" will prevent rod withdrawal.
Question 4.07 a.
Consider in grading that only one method to scram the reactor is deli-neated in Procedure 2.4.143.
Operators may list other methods to' scram for the second metho ATTACHMENT 3 NRC Resolution of Comments on Written Examinations The following represents the NRC resolution to those comments made by the facility as a result of the current exam review policy.
Only those comments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not listed.
i.e.: typo errors, relative acceptable terms, minor set point changes.
SR0 EXAMINATION 6.01 Comment accepted. The D/G lockout trip will be deleted and the local trip handle D/G trip will be added.
(Comment resolved at site)
6.05 b.
Comment accepted. The term automatic scram signal could be inter-preted as only those signals which are inputs to the RPS activation logic vice any mechanism which causes a scram.
(Comment resolved after leaving site)
6.10 a.
Comment not accepted. The student learning objective requires a knowledge of the setpoints of trips and interlocks associated with the refuel system. The question asks for setpoints that energize indication lights which are indications of interlocks on the refuel-ing system. The lesson plan represents these setpoints as conditions to energize indication lights and as conditions for interlocks.
Therefore, the question asks for information within the scope of facility required knowledge (Comment resolved after leaving site).
R0 EXAMINATION 2.01 b.
Resolved at Site. Accepted comment. This information was not in the training material sent to the examiner.
2.02 a.
Resolved after leaving Site. Comments accepted with the requirement that the answer include drywell coolers dedicated for the recirc pump.
This answer is acceptable since it is similar to the example given in the problem (MG set room area recirculating coolers)
although it was not included in the Recirculating System Reference Text.
2.04 b.
Resolved at Site. Accepted comment since it was legitimate interpretation of the questio Attachment 3 2.08 a) 1. Resolved at Site. Accepted comment. A typing error was the cause of this problem.
2.09 b.
Resolved at Site.
Comment accepted since it is the correct inter-pretation of material in the RHR System Reference Text.
3.03 a.
Resolved at Site. Comment accepted since either answer is correct.
i 3.06 b.
Resolved at Site.
Comment accepted. This information was not I
available in the training material sent to the examiner.
3.09 Resolved at Site. The latter part " Power off to the refuel bridge" was accepted. Although it was not in the training material sent to the examiner, it is correct.
Resolved after leaving Site. Breaking up "any hoist" into three parts was ac'cepted since they are indeed three separate conditions.
4.07 a.
Resolved at site.
It was agreed that consideration would be given to other methods.
It should be noted, however, that a second method is implied in Procedure 2.4.143. Again this is a matter of interpre-tation.
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ATTACHMENT 4 Details on Procedures Facility has committed to Review 1.
Procedure E0P-3 P.10.
This page contains a matrix of possible reactor conditions which direct the operator to certain actions.
Facility com-mitted to determine if one of the blocks in the matrix needed to specify any action since it was blank.
2.
Procedure E0P-4 P.24.
This page contains procedure text and a graph of containment parameters.
Facility committed to review question on maximum temperature in text on step III. G.5 and maximum temperature shown on graph since they are different and could mislead the operator as to what limit applied.