ML20052F008
| ML20052F008 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/05/1982 |
| From: | Mardis D FLORIDA POWER CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM NUDOCS 8205110624 | |
| Download: ML20052F008 (35) | |
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Power C O A P O R A 7 I O at May 5,1982 c1 f
- 3F-0582-06 ff File: 3-0-26 3
RECEtygg Mr. John F. Stolz, Chief 2
M47J0 Operating Reactors Branch #4
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-l Division of Licensing garYs U.S. Nuclear Regulatory Commission 9
8 Washington, D.C. 20555 N
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Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 NUREG-0737; Items I.A.2.1 and II.B.4 Upgrading SRO and RO Training and Training for Mitigating Core Damage
Dear Mr. Stolz:
By letter dated April 1,1982, the Nuclear Regulatory Commission (NRC) requested additional information on NUREG-0737, item I.A.2.1, " Upgrade of Senior Reactor Operator and Reactor Operator Training," and Item II.B.4, " Training for Mitigating Core Damage."
Florida Power Corporation (FPC) hereby provides responses to your eight questions as follows:
Item I. A.2.1 Upgrade Reactor Operator and Senior Reactor Operator Training Question 1.
Is the material on heat transfer, fluid flow, and thermodynamics which is taught in the training and requalification programs at a level of detail comparable to that identified in Enclosure 2 of Denton's March 28 letter?
Response 1.
The material taught on heat transfer, fluid flow, and thermodynamics is of a comparable level of detail as that identified in Denton's March 28, 1980 letter.
Question 2.
Are the training and requalification courses on mitigating the conse-quences of an accident involving a severely damaged core complete, and in effect at the present time? Do they follow the guidelines in Denton's March 28,1980 letter?
Response 2.
The training and requalification courses on mitigating core damage are complete and in effect at the present time. They follow the guidelines in Denton's March 28,1980 letter.
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u Mr. John F. Stolz May 5,1982 Page 2 Question 3.
Are the lectures and quizzes on the subject of accident mitigation given to shift technical advisors and operating personnel from the plant managers through the operations chain to the licensed operators? If they are, provide the titles of the people who are trained and an organization chart which illustrates their position in the operations chain.
Response 3.
The lectures and quizzes on accident mitigation are given to the shif t technical advisors and operating personnel from the plant managers down through the operations chain to the licensed operators. For the latest organizational chart, reference Techr.ical Specification Change Request Number 67, Rev. I dated January 19, 1982. The titles of personnel who attend the training are as' follows:
(1) Nuclear Plant Manager; (2)
ChemRad Manager; (3) Technical Services Superintendent; (4) Operations Superintendent; (5) Health Physics Supervisor; (6) Health Physics Technician; (7) Instrument and Controls Supervisor; (8) Instrument and Controls Technician; (9) All Licensed Operators; (10) Nuclear Operations Technical Advisor; and (11) Licensed Operator Candidates.
Question 4.
Do the portions of the training program and requalification program which address heat transfer, fluid flow, thermodynamics, and accident mitiga-tion involve 80 contact hours? (A contact hour is a one-hour period in which the course instructor is present or available for instructing or assisting students; lectures, seminars, discussions, problem-solving sessions, and examinations are considered contact periods under this definition.)
Response 4.
The training and requalification program does not involve 80 contact hours of training in heat transfer, fluid flow, thermodynamics, and accident mitigation. Each program does include sixteen (16) hours of accident mitigation and forty (40) hours of fluid flow, thermodynamics, and heat transfer training.
Question 5.
Does the requalification program which the instructors take address current operating history, problems, and changes in procedures and administrative limits?
l Response 5.
The instructor requalification program does address current operating l
history, problems, and changes in procedures and administrative limits.
t Question 6.
As specified in Enclosure 1 of Denton's March 28,1980 letter, does the operator requalification program call for accelerated requalification if the overall score is less than 80%, and the score in any category is less than 70%?
Response 6.
The operator requalification program does call for accelerated requali-fication if the overall score is less than 80%, and if the score in any category is less than 70%
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Mr. John F. Stolz May 5,1982 Page 3 Question 7.
Does the operator requalification program call for control manipulations as specified in Enclosure 4 of Denton's letter of March 28,1980, to all power reactor applicants and licensees?
Response 7.
The operator requalification program does call for control manipulations as specified in Denton's March 28,1980 letter.
Item II.B.2 Training for Mitigating Core Damage Question 8.
Provide an outline of the training program for mitigating core damage, including the number of training hours involved. Your outline can include any training program which relates to the training for mitigating core damage. Follow the guidelines given in the Enclosure 3 of Denton's letter dated March 28,1980, and INPO Guidelines for Training to Recognize and Mitigate the Consequences of Core Damage (Document Number STG-01, Rev.1, January 15, 1981). NRC requires minimum of 80 contact hours of training for mitigating core damage.
Response 8.
FPC's training program includes sixteen (16) hours of mitigating core damage training and forty (40) hours of training in fluid flow, thermodynamics, and heat transfer. An outline of this training program is attached. From conversations with the NRC staff, FPC understands that the minimum of 80 contact hours of training for mitigating core damage (and related subjects)is not required by the NRC. FPC's training program does address and satisfy the concerns and intent of the INPO recommendations.
If you have any further questions, please contact this office.
Very truly yours, Dsd) 0 MO David G. Mardis Acting Manager Nuclear Licensing Attachment RAW:mm cc:
Mr. 3. P. O'Reilly, Regional Administrator Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 3100 Atlanta, GA 30303
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COURSE OUTLINE E^
f DEGRADED CORE TRAINING Part I: Preventing Degraded Core Conditions Lesson 1 - Core Cooling Mechanics Lesson 2 - Gas / Steam Binding Affects on Core Cooling Lesson 3 - Boron Precipitation Concerns Following a LOCA Lesson 4 - Equipment Failure Sequences that Could Lead to a Degraded Core Lesson 5 - Avoiding Degraded Core Conditions Part II: Recognizing a Degraded Core Lesson 6 - Consequences of Inadequate Core CoolirIg and Likely Core Da= age Effects Lesson 7 - Use of SPND's in the Recognition of Degraded Core Conditions n a:;.
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Lesson 8 - Detection and Treatment of Inadequate Core Cooling Using l
Lesson 9 - Relationship of Out-of-Cor'e (OCD) Source Range Detectors to Degraded Core Conditions Lesson 10- Incore Ther:occuples and Core Flow Blockage Lesson 11-Release of Fission Products from Damaged Fuel Lessen 12-Fission Product Transpo'rt Characteristics and Release Psthways l:
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Lesson 13-Response of Ga==a Radiation Monitors Lessen 14-Chemical and Radioche=ical Sa=pling Problems I
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s CONTENTS
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Page Lesson 1 - CORE COOLING lECHANICS.
1-1 Lesson 2 - GAS / STEAM BINDING 2-1 Lesson 3 - BORON PRECIPITATION CONCERNS POLLOWING A LOCA 3-1 Lesson 4 - EQUIPMENT FAILURE SEQUENCES THAT COULD LEAD TO A l
DEGRADED CORE 4-1 l
Lesson 5 - AVOIDING DEGRADED CORE CONDITIONS 5-1 II e
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Lesson 7 - USE OF SPNDs IN RECOGNITION OF DEGRADED CORE 7-1 CORDITIONS Lesson 8 - DETECTION AND TREATMENT OF INADZQUATE CORE COOLING 8-1 USING CORE EXIT THERMOCOUPLES.
Lesson 9 - RELATICNSHIP OF OCD SOURCE RANGE DETECTORS TO -
9-1 DEGRADED CORE CONDITIONS' W4 Lesson 10 - INCORE THERMOCOUPLES AND CORE FLOW BICCFAGE.
10-1 Lesson RELEASE OF FISSION PRODUCTS FROM DAMAGED FUEL..
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12-1 RELEASE PATHWAYS 0
13-1 Lesson 13 - RESPONSE OF GAMMA RADIATION MONIT,RS 14-1 Lesson 14 - CHEMICAL AND RADIOCHEMICKL PROBLEMS..
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f Lesson 1 - CORE COOLING MECHANICS 1
Introduction 1.
Lecturer -
2.
Purpose - To describe the basic principles of heat transfer from the core, the effects of Icss of this heat transfer, and the reasons for disruption of nomal circulation patterns.
Objectives The following subjects will be covered during this lesson:
1.
Basic heat transfer from the core to the ultimate heat sink.
2.
The factors affecting this heat transfer and the operator's con-trol over them.
3.
Heat transfer characteristics of the vario'us states of water.
4.
Natural circulation theory.
5.
High-pressure injection (once-through) method of core cooling.
11 Key points to be retained are as follows:
1.
The various methods available for cooling the ' core.
l 2.
How to interpret primary system status from saturation curves.
I 3.
Factiors that affect natural circulation.
4.
How the operator can effect heat transfer from the core to the h
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Lesson 2 - GAS / STEAM BINDING I
i Introduction 1.
Lecturer -
Purpose - To' describe situations that could lead to introduction of large 2.
To list potential gas sources - conden-quantities of gases into the RCS. To discuss potential gas effects on nat-sible and non-condensible gases.
To give guidance for recog-ural circulation and on OTSG heat removal.
tiizing and dealing with those gases.
Objectives The following subjects will be covered during this lesson:
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1.
Why gases might enter the RCS.
gases might do once inside the RCS.
2.
What How an operator might recognize and deal with gas presence and 3.
effects.
Key points to be retained are as follows:
There are situations which could lead to gas volumes in the RCS.
1.
Once there, gases can affect natural circulation and heat removal 2.
through the OTSGs.
These effects can be recognized and dealt with by the operator.
3.
(e.g.,
Actions are dependent on availability of certain equipment 4.
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Lesson 3 - BORON PRECIPITATION CONCERNS FOLLOWING A LOCA Introduction 1.
Lecturer -
2.
Purposes -
To explain mechanisms for obtaining high boron concentrations in the a.
core following,a LOCA.
b.
To discuss the dangers of high boron concentrations relative to long-term, post-LOCA core cooling.
c.
To discuss the interface between natural circulation and operator-induced circulation in preventing boron precipitation.
d.
To outline the long-term flow requirements to prevent boron precipi-tation.
Objectives i
The following material will be presented during this lesson:
1, 1.
How the boron precipitation problem can arise.
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2.
The dangers of boron precipitation and how these could lead to a I
degraded core situation.
3.
Natural circulation and subsequent operator-initiated circulation to l
prevent boron precipitation.
I 4.
Specific results of operator actions.
l The key points to be retained are as follows:
1.
How and why boron precipitation can become a problem following a TOC.
l 2.
How long the operator can wait to take action before the danger l
point is reached, l
3.
In general, what actions (including NRC requirements) must be taken I
to manage the boron precipitation problem on a long-term basis.
4.
There is nothing to indicate to the operator that boron precipitation is causing a problem. The best approach is to prevent rather than correct.
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GF Lesson 4 - EQUIPMENT FAILURE SEQUENCES THAT COULD LEAD TO A DEGRADED CORE 1.
Functional Failures A degraded core could evolve from several vulnerable plant conditions.
Such conditions include small LOCAs (including a stuck-open pressurizer safety valve and RC pump seal rupture), loss of normal heat sink, and loss of offsite power.
In all cases, additional failures are required to cause inadequate core cooling. Any event rath leading to a degraded core has a low probability of occurrence. Before addiessing combinations of equipment f ailure and time frames, it is important to note failures in system functions - for example, a small LOCA with a loss of the HPI function could lead to a degradfd core con-dition.
Similarly, loss of the secondary side cooling function with loss of I
the HPI function or tripping of RC pumps during a small break while the RC void
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fraction is 70% or greater could also lead to a degraded core condition.
l Loss of any system function can result from combinations of independent equip-t j
' ment failures, operator errors, or common-cause failures. Combinations of in-l dependent equipment failure in different systems, e.g., MFW pumps, auxiliary I
feed pumps, and HPI pumps, are a remote possibility.
It is the intent of these lesson plans and ATOG to lessen the probability of operator. error that would propogate a degraded core condition (e.g., turning off EPI and auxiliary FW),
leaving common-cause failures as the most probable (but still very small) path to inadequate core cooling conditions. These include hardware common causes and the cperator common cause (which is the result of conflicting information 1
being displayed to the operator). Lesson 5 describes recognition and proce-dures to employ for these conditions, i
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Scenarios 2.1.
Natural Catastrophic, Sabotage, and ATWS Scenarios These sequences are beyond the immediate scope of these lesson plans although f
The ini-operator corrective actions can be taken to mitigate their effects.
tiating event itself is outside the control of the operator, but depending on the resulting effects of the initiating event, some portions of the ATOG guide-For exagle, if the initiating event caused a loss of lines are applicable.
secondary heat removal, operator actions would coincide with guidelines for It should b'e noted that ATWS sequences are manageable loss of all feedwater.
if sec.ondary heat removal is avcilable and if peak ATWS pressure does not re-B&W in primary system rupture (if it ruptures, it is a LOCA condition).
sult plants also have a manual trip and runback capability in the event of failure HPI boration should also be attempted as soon as possi-of the RPS breakers.
ble (the initial primary system pressure spike may be'too large for the HPI pumps to deliver water, but they should be flowing as pressure drops).
HPI To summarize, if the RPS fails initiation will be by manual operator action.
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to drop the rods, try to run the rods in manually while initiating HPI flow.
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It should be noted that this course of action is consistent with ATOG proce-dures in that HPI flow should be established on loss of subcooling margin.
2.2.
Scenarios Terminated by Procedural Action Most of the corrective actions for the INPO vulnerable plant conditions have 1
This section addresses only those been accounted for in the first lesson plan.
items not discussed in Lesson 1.
Vulnerable plant conditions identified by INPO include the following:
Loss of offsite power while one onsite power source is out of ser-1.
If Loss of offsite power results in loss of main feedwater.
vice.
the diesel supplying the motor-driven emergency feed pump is the one that is out of service, secondary side heat removal is dependent on the turbine-driven pump (unless condensate / fire pumps, etc. can If the HPI that be aligned and'are supplied by the good diesel).
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is aligned (or can be through a swing arrangement *) to the good Unless these multiple equipment failures (feedwater and diesel.
HPI pumps) occur, normal procedural action terminates this event If the multiple failures do occur, this event would be sequence.
classified under section 3 (common cause and causal f ailures).
2.
Extended station blackout. If the emergency turbine feed pump is available, secondary side cooling is initiated and the. event is within the scope of the guidelines. However, see section 3 for a discussion of vulnerability while operating in this mode.
3.
Stuck-open pressurizer safety valve. This is a small LOCA, and guidelines are in place; however, depressurization to decay heat removal pressure is encouraged before the BWST is depleted.
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4.
BWR event.
valve or steam line break. The affected steam generator should be isolated by stopping feed flow, and cooldown should be initiated under the "one good OTSG" guideline.
For loss Loss of normal heat sink following reactor / turbine trip.
5.
O of the secondarv heat remevat function. emeter HP1 coe11ng under Lesson 1 (part 5).
If Loss of d-c control power to a 4160 V emergency safeguard bus.
6.
this should occur, the diesel supplying the good 4160 V bus should If be started and left running until the situation is rectified.
the diesel fails to start or fails to continue to run, a shutdown A loss of offsite power during this time is unlikely, should begin.
but in case it should occur, a station blackout results as discussed in item 2 above.
If the motor-Loss of automatic emergency turbine feedwater control.
7.
driven emergency feedwater pump (s) are not available, then manual Two methods of control control of the turbine pump is necessary.
if the third HPI pump is a swing pump, all supporting
- However, remember that subsystems, such as cooling water, should also be aligned to some power ser-vice.
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(1) manually start /stop pump based on steam genera-tor level, or (2) continue to run the pump but regulate. flow to the steam generator by varying the amount of recirculation to the con-denser. The second method is preferable (as long as the bearing temperature on the turbine and pump are maintained in the safe re-gion).due to the relatively large probability of failure of the tur-bine pump to start on demand (e.g., overspeed trips while coming up to j
load) and the fact that even small flow to the generators has a quenching (cooling) effect.
2.3.
Scenarios Unaffected by Procedural Action There are scenarios for which no guidelines are written.
In short, these are
- the ones in which nothing operates properly, so the operator has little if any-thing to do concerning the primary system condition. No guidelines address the actions to be taken in the event of losses in th& secondary heat transfer and high-pressure injection functions, or in fact for any combination of losses I
of functions that would prohibit eventual depressurization of the primary sys-()
tem.
Under these conditions, the operator could adjust the containacnt func-The reason guide' lines 're not written for these tion if it were operable.
a "everything f ails" scenarios is that their probability is remote. The most likely of these improbable scenarios are the common-cause sequences, such as loss of all a-c power, and those in which the operator fails to act (analogous to "everything fails") or acts incorrectly due to inadequate feedback of plant conditions.
The next section discusses the loss of all a-c power event se-quence viewed from a common-cause/ causal viewpoint. Some simplified event tree diagrams follow in section.4.
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Common-Cause and Causal Failures g.
A loss of power is a common cause event that affects equipment in more than one system. There are three sources of high-voltage a-c power:
(1).in-house, s
(2) of f site, and (3) emergency ot site. The pre-TMI setpoint (on PORV and high presisure trip) for operation of B&W plants allowed for a possible runback of station power to house locd in the event of load rejection. With the post-TMI setpoints, this capability is no longer available. Therefore, if offsite power is lost, only emergency onsite power is available. Most plants have two diesel generators for emergency onsite power. Safety systems, such as high-pressure injection, are loaded on their respective diesels (train A on diesel A, train B on diesel B), minimizing the consequences of, failure of one diesel If both diesels fail to start, only t.1e steam-driven auxiliary feed to start.
pump is available as an active source for heat removal.
In the event that flow blockages (e.g., closed valves, pump failure, or inadequate steam supply to drive the turbine pump) exist, then this source of heat removal is also lost.
It should be noted that after a loss-of-feedwater event there is a lim-ited amount of steam inventory (even if there are no stuck-open steam or at-mospheric dump valves, which could be caused by loss of control power), so it
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is imperative that the auxiliary feed pump be started (if it has failed, for example, on overspeed trip) as soon as possible,to supply feedwater not only for cooling purposes but also for steam generation to drive the pump.
In the loss of all a-c power scenario, there is a limited amount of time before the d-c power source gives out.
In addition, the loss of ventilation / air condi-tioning will eventually result in erroneous signals being displayed for opera-tor information or used in automatic steam generator level control, even if d-c power is still available. This introduces additional event sequence paths that can lead to a degraded core condition.
j Other common-cause initiating events include seismic events *, the effects of which are discussed in more detail in Lesson 5.
Before the plant modifications on the separation of NNI-X and -Y, loss of power to either NNI would have been a
- Note hat seismic is not classified as natural catastrophic (section 2.1) because the more probable seismic events are the less severe or mild-amplitude occurrences which will not cause massive equipment damage but would generate conflicting information to the operator.
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Now, however, with this separation - coupled with knowledge of where the instrumentation power is supplied from - the probability of this event sequence is minimized.
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Introduction 1.
Lecturer -
2.
P'urpose - To give the operator necessary background information on the likely consequences of sustained inadequate core cooling and the resulting m;
progression of core damage.
iL Objectives
' The following material will be presented during this lesson:
n 1.
General physical and thermochemical description of the fuel and cladding.
O 2.
Normal operating parameters, = then accident (LOCA) conditions that effect the integrity of this system.
3.
Direct consequences to the cladding and fuel of major decrease in cooling capacity at cladding surface and/or partial core uncovering.
4.
Primary consequences of loss of fuel integrity.
Key points to be retained are as.follows:
I k1 1.
The general configuration of the fuel / fuel assembly.
2.
Operational (i.e'., pressure / temperature) limits of fuel assembly materials.
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3.
Mechanics of fuel failure under accident conditions.
i 4.
% e consequences of fuel failure to the primary system.
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Lesson 6 Outline 1.
Introduction 2.
Fuel and Cladding System Description 2.1.
Fuel 4
2.2.
Cladding and Assembly Materials v
3.
Loss-of-Coolant Conditions 3.1.
Manual Operation 3.2.
LOCA 4.
Direct Lonsequences of Core Uncovering to Cladding / Fuel 4.1.
Cladding Rupture 4.2.
Cladding Oxidation / Hydrogen Generation 4.2.1.
Zircaloy Oxidation 4.2.2.
Stainless Steel Oxidation 4.2.3.
Radiolytic Decomposition of Water 4.3.
Fuel / Cladding Eutectic Formation 5.
Core LOCA Consequences and TMI-2 Estimates 5.1.
General Consequences 5.2.
TMI-2 Estimates 6.
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Introduction 1.
Lecturer --
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Purpose -- To investigate the use of self-powered neutron detectors (SPNDs) b)
in the analysis of degrading or degraded core concitions.
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Objectives I,
I The following material will be presented during this lesson:
1.
Description of the incore monitoring system (IMS).
2.
Operation of the IMS under normal conditions: at power and during f~
shutdown.
.On 3.
Response of SPNDs to high temperatures: thermionic current and re-lease of space charge.
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Interpretation of SPND alarm data following reactor trips.
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5.
Limitations of SPND post-trip alarms.
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Key points to be retained are as follows:
1.
Nuclear and high-temperature responses of the IMS.
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Use of alarm printer in interpreting SPND alarms.
3.
Limitations on SPND post-trip alarms.
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1.
Introduction 2.
Description of Self-Powered Neutron Detectors 3.
Operation of SPNDs Under Normal Conditions 4.
High-Temperature SPND Response 4.1.
Thermionic Emissions - 1979 LRC Experiment
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4.2.
Space Charge Release Mechanism 4.3.
Other Phenomena 4 3.1.
Interaction of Thermionic Emissions and Space Charge Release 4.3.2.
Presence of Gamma Radiation and Electronic Circuitry 91 5.
Interpretation of SPND Alarms After Reactor Trips 5.1.
Axial and Radial Distribution J
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High Temperature Vs High Flux
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Limitations of SPND Post-Trip Alarm Analysis q
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Conclusions 8.
Summary References J
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Lesson 8 - DETECTION AND TREATMENT OF INADEQUATE CORE COOLING USING CORE EXIT THERMOCOUPLES Introduction r.
1(,j 1.
Lecturer'-
2.
Purposes -
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Id a.
To discuss the changes in core cooling efficiency as the reactor coolant changes state and flow conditions.
F i-b.
To discuss core exit thermocouple (T ) use as a means of detecting L
dangerous conditions in the core.
To describe analyses which resulted in the cor'e exit T readings q
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and cladding temperatures (T& d)*
d.
To provide instructions for converting T readings to T clad
- e.
To discuss required operator actions during high Tc d (inadequate O"""
core coo 11ns) cenditie s.
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f.
To briefly discuss T limitations.
C objectives The following material will be presented during this lesson:
1.
What might cause the reactor coolant to change. state and how will e ra that affect core temperatures?
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2.
How might the operator be able to use Ts to determine if dangerous i.
conditions exist, and the degree of dan ler?
3 How are T readings and cladding temperature related, and how was e
relationship established?
r 4.
What actions must operators take based on T readings?
5.
How might these actions affect the core and the rest of the plant?
t' 6.
What are the limitations of the T 's?
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What influence does the amount of available reactor coolant and the O
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What damage would be expected at high core temperatures?
3.
How T 's can measure existence and magnitude of danger to ' ore cooling.
c 4.
Some familiarity with anslyses used to develop T correlation.
c clad 5.
What the operator should be doing for given T readings.
c 6.
Jhat the'results of these actions might be for the core and the rest of the plant.
7.
How T readings might be misleading.
8.
What to expect for cladding temperatures above 1800F.
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Introduction I'
2.
Causes of. Inadequate Core Cooling (ICC) and High Core Temperatures b
3.
Normal and Off-Normal Modes of Heat Transfer n).'
4.
Why and How Cladding Temperature is Measured L./
5.
ICC Cuidelines Defining Operator Actions Based on T r,
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Possible Results and Consequences of Operator Actions 7.
Shortcomings of T 's a
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Lesson 9 -- RELATIONSHIP OF OCD SOURCE RANCE DETECTORS
',a TO DEGRADED CORE CONDITIONS
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'f~1 Introduction r.
1.
Lecturer --
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2.
Purpose - To provide interpretations of abnormal source range monitor read-(..
t; ings after a reactor trip.
L Although the primary function of out-of-core flux detectors is to monitor I.,
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reactor flux levels during approach to criticality and power operation, an analysis of post-accident TMI events has indicated that the source range monitors (SRMs) are sensitive to citanges in primary coolant " apparent den-sity" and changes in certain core conditions that affect neutron genera-O tion.
'&s Objectives The following material is covered in Lesson.9:
1.
General description of out-of-core detector (OCD) system.
I.,
E' 2.
Brief discussion of physical processes involved -- neutron generation in core, criticality, neutron transport.
G Lj 3.
Parametric effect of source generation and void fraction on SRM readings.
r)j' 4.
Discussion of normal and abnormal (TMI-2) SRM readings after a reactor trip.
The following key points are to be retained:
1.
Above-normal SRM readings after trip can be caused by any one or a combination of the following: recriticality, fuel failure with fission sj product release, core voiding, downcomer voiding, and/or coolant tem-perature increcse.
l r~
2.
" Normal" SRM readings after trip can vary due to startup source strength, i
I core multiplication, and/or coolant temperature.
3.
SRM readings should decrease continually after trip, and the operator L_
should be wary of any increase in count rate.
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9-1 Babcock & Wilcox l
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J LESSON OUTLINE 1.
Source Range Monitor Response After Reactor Trip 1.1.
Typical SRM Response (Normal Trip at Oconee 3) 1.2.
.iRM Response After.TMI-2 Accident 2.
Out-of-Core Detector System 3.
Neutron Transmission From Core to SRM 3.1.
Neutron Transport 3.2.
Homogeneously Distrib'uted Voids 3.3.
Coolant Temperature 3.4.
Coolant Level n
e So e
4.2.
Suberitical Multiplication 4.3.
Excore Sources 5.
Reactor Events 5.1.
Recriticality 5.2.
Loss of Coolant 5.3.
Coolant Boiling 5.4.
Core Damage i
5.5.
False SRM Signals Due to Gamma Radiation 6.
SRM Chart Interpretation 7.
Summary References O
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TO DEGRADED CORE CONDITIONS m
1J 1.
Source Range Monitor Response After Reactor Trip
(.
Source range monitors (SRMs) are part of the out-of-core detector (OCD) system in B&W reactors and, as such, are located in the reactor cavity (Figure 9-1).
y Their major function is to measure a reactor's approach to criticality prior to startup. Observations after the TMI-2 accident indicated that the SRM
(.
readings could be confusing, but if properly interpreted, they could be useful s
monitors of core behavior and system conditions during certain accidents.
Be-cause of the SRM locations - away from the harsh environment in the core - SRM L;
signals should be a rapid, reliable measure of neutron flux in the reactor O"
J cavity which is normally. proportional to core flux. However, interpretation of SRM signals is not straightforward during accidents that could lead to a degraded core because the signal responds to changes in both neutron transmis-Le sion and neutron generation. Thus, supporting data from other installations must be considered along with operator action to' define,the cause of anomalous SRM readings following accidents.
g Neutrons that are born in the core region must pass through several steel and
,~
h water regions before reaching the detector locations. The shaded area in Fig-ure 9-1 represents the peripheral fuel assemblies in the core where neutrons that are born have a good chance of escaping the core. The unshaded area rep-L.
resents the inner fuel assemblies (later referred to as the inner core), where changes in neutron generation rate will not significantly affect the core es-cape flux or the SRM response.
The fraction of neutrons transmitted (or its reciprocal, the attenuation fac-tor) is affected by the mass of water between core and cavity. Water mass can vary due to density changes associated with temperature changes and/or voiding due to loss of coolant via some mechanical malfunction or human error. The neutron generation rate is affected by the mass of coolant in the core region u
9-1 Babcock s,Wilcox L
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and any other parameters that alter core reactivity, such as coolant void, coolant temperature, boron concentration, control rod location and integrity.
and fuel pin integrity. The complexity of the problem becomes apparent when one considers that these phenomena have competing dependent and/or multiple effects on SRM responses.
j In this ' lesson the SRM responses after a typical reactor trip and after the TMI-2 accident trip are discussed. We will consider some of the reactor con-ditions that perturb SRM responses and then provide suggestions for interpret-ing SRM charts in accident situations.
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1.1.
Typical SRM Response (Normal Trip at Oconee 3)
]!
The SRM response to a normal reactor trip is shown in Figure 9-2.
Initially,
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the power (and therefore the SRM signal) drops with an 80-second period, con-sistent with the 55-second half-life delayed neutron group. During the first 15 to 20 minutes following reactor trip, the delayed neutrons from 2nU fission
~
products are the dominant source of neutron flux.
J As the delayed neutrons die out (becoming relatively insignificant af ter about
]
20 minutes), the photoneutrons from y,n reactions with D20 (in the primary 5-l coolant) become the dominant source of neturons. This source decays with a variable half-life of 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (over the' time period of about 0.5 to 4 1
hours af ter trip) and becomes relatively small by about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> af ter trip.
- 1 After about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> the SRM signal tends to be essentially constant because it is then responding to neutrons from the startup source.
l The absolute value of the SRM signal following a normal reactor trip may vary
)
from one reactor to another and even between subsequent trips in the same re-3, actor. However, the general shape should be maintained. Some of the following j
items could affect the SRM reading:
1.
Power history several days prior to trip and, to some extent, burnup -
This determines fission product concentration, which is the source of gama flux for photoneutron production.
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2.
Coolant temperature variations - This changes the attenuation of neu-i trons from core to detector.
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3.
Startup source activity - Depending en the type of startup source, the p
neutron output may either decrease or increase with time after instal-V lation in the reactor.
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.i Lesson 10 - INCORE THERMOCOUPLES I
AND CORE FLOW BIDCKAGE i n
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Introduction r m 1,
1.
Lecturer -
Uniqueness - Incore thermocouples are unique in that they are the only 2.
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temperature instruments monitoring below a corevide scale.
3.
Purpose - To evaluate the capability of incore thermocouple to provide s
t adequate indication for reactor operators to recognize core flow i
j blockage.
'1
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Objectives The following material.will be presented during this lesson:
1.
Use of incore thermocouples to estimate ' core flow.
i 5'
2.
Comparing core flow to reactor vessel flow.
3.
Estimating core flow blockage.
4.
TMI-2 experience, The key points to be retained are as follows:
m 1.
How to calculate and compare core flow and reactor vessel flow, 2.
How to detect and estimate core flow blockage.
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1.
Introduction 2.
RC or HPI Pumps Running 2.1.
Calculation of Effective Core Flow 2.2.
Determination of Reactor Vessel Flow
~
2.3.
Comparisons and Conclusions 3.
Natural Circulation 3.1.
Incore Thermocouple Readings 3.2.
Core Flow Distribution 3.3.
Core Flow Blockage i
l 4.
TMI-2 Experience J
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(J Lesson 11 - RELEASE OF FISSION PRODUCTS
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FROM DAMAGED FUEL r
{.3 Introduction r
1.
Lecturer --
g 2.
Purpose -- To describe the buildup and distribution of fission products in
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the fuel and to quantify the amount of fission product activity expected to be released after various fuel damage scenarios.
r
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Objectives lesson 11 covers th'e following material:
1 L-1.
The basic characteristics that control the release of key fission products from the fuel (qualitative).
D 2.
The buildup of key fission products as a function of irradiation L
time and the distribution of fission product activity (quantita-r1 tive) between the fuel pellets, th,e fuel rod gaps, and the reac-tor cociant.
s 3.
The progressive increase in fission product release associated with the various stages of core damage.
4.
Rapid assessment of fuel damage based on the amount of key nuclides in the reactor coolant and the ratio of long-to short-half-life nuclides.
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Behavior of Fission Products
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2.
Fission Product Buildup and Distribution
,e 3.
Assessing Degree of Core Damage
- 4.
Summary lD j
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Lesson 12 - FISSION PRODUCT TRANSPORT CHARACTERISTICS AND RELEASE PATHWAYS s
~
Introduction j
1.
Lecturer --
2.
Purpose - To describe the basic chemical characteristics that affect the
[]
transport of several groups of key fission product nuclides and to iden-tify potential pathways for the release of these fission products to the m'
containment or to the environment.
L Objectives The following material is covered in Lesson 12:
... r 1.
The basic chemical characteristics (qualitative) that affect the transport of key fission product nuclides.
2.
Basic facts on how fission products will behave following a reac-tor accident.
_a 3.
Identification of potential pathways for the release of fission ri products into the containment or into the environment.
4.
Rapid identification of fission product release pathways so that they can be expeditiously terminated.
l The following key points are to be retained:
1.
The noble gases will follow.the steam and will accumulate wherever there is a gas phase, i.e., the pressurizer steam space, the makeup tank gas space, or the containment atmosphere.
2.
The iodine activity can exist in many forms, but in general if
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water is readily available, 99.9%+ will tend to be in the liquid r-phase.
i
'J 3.
If for some reason the accident is a dry one, i.e., no water in l
the reactor and no steam in the ' containment atmosphere, there is likely to be a very large source of iodine and cesium aerosols l
airborne in the containment, which shoald be removed as quickly as possible using the reactor building sprays.
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The activity release pathways that develop during-a reactor
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accident are not necessarily the intuitively obvious pathways v
but are more likely to be obscure, forgotten pathways.
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LESSON OUTLINE 1.
Chemical and Transport Characteristics of Fission Products 1.1.
Noble Gases 1.2.
Iodine 1.2.1.
Elemental Iodine 1.2.2.
Hypoiodous Acid 1.2.3.
Organic Iodide 1.2.4.
Cesium Iodide 1.2.5.
Particulates 1.3.
- 3..:,...
2.
Fission Product Release Pathways 2.1.
Pathvays From RCS to Containment 2.2.
Pathways Into Auxiliary Building 2.3.
Pathways to the Environment 2.4.
Release Pathways Identified at TMI-2 3.
Summary i
O 12-11 Babcock a.Wilcox 4
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Introduction 1.
Lecturer -
u 2.
Purpose - To describe the responses that should be expected from various gamma radiation monitors following a reactor accident and to alert the b
operators to anomalous readings that may occur due to high background radiation, instrument failure, improper calibration, and source concen-
{
tration gradients.
Objectives l
The following material is covered in Lesson 13:
u 1.
The magnitude of the letdown line monitor response as an indication C',S d
of the degree of core damage.
2.
Anomalous readings that have occurred in process radiation monitors
~
1 and area radiation monitors following the TMI-2 accident and the CR-3 loss of non-nuclear instrumentation event.
[
The following key points are to be retained:
1.
1.
Expected letdown monitor response.
2.
Effects that can cause anomalous readings:
y a.
High background radiation.
('
a b.
Scale and calibration errors.
ta c.
Shielding of detectors.
d.
Mixture of fission product nuclides.
e.
Source concentration gradients.
f.
Electrical power supply failures.
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Electronic component failures.
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1.
Letdown Line Monitor W
A 2.
Pro. cess Monitors 3.
Area Monitors x
4 Reactor Building Dome Monitor 5.
Failure Modes 6.
Summary 9
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e-Lesson 14 - CHEMICAL AND RADI0 CHEMICAL PROBLDiS J
i Introduction r-1.
Lecturer -
}'
2.
Purpose - To describe some of the chemical and radiochemical problems t'
that can occur following a reactor accident that results in degraded core L.
conditions.
Objectives The following material is presented in Lesson 14:
"~
1.
Chemical and radiochemical problems to be expected following reactor accidents that result in degraded core conditions.
i-2.
Interpreting the analytical results, which will appear to be 0
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3.
How to avoid making operational or procedural errors that might lead
['.
to plant and equipment contaminat; ion as excessive radiation exposure to operating personnel.
4.
Avoiding both sampling problems and the use of incorrect analytical methods.
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