ML20038A877
| ML20038A877 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 11/16/1981 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 14852, NUDOCS 8111240376 | |
| Download: ML20038A877 (2) | |
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C011SumBIS Power Jernes W Cook Vice President - Projects, Enginer,ing and Constructwn General offices: 1945 West Parnell Road, Jackson, MI 49201 * (517) 788-0453 1 ' 1 O)',
November 16, 1981
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US Nuclear Regulatory Commission Washington, DC 20555 l
MIDLAND PROJECT M:...:,'D DOCKET NOS 50-329, 50-330 RESPONSE TO OCTOBER 23, 1981 MEETING ON VERIFICATION TESTING FOR SB LOCA METHODS FILE:
0926.2 SERIAL:
14852 i
This letter expresses our understanding of the agreements reached in the B&W Owners Group Meeting of October 23, 1981 with you and members of your Staff.
i In summary, the Staff and Owners concluded that an in depth review of the current B&W SB LOCA Methods Program including the verification testing base l
was required before it could be determined if further verification testing of our SB LOCA methods is required. The Staff agreed to participate in this
-review which will begin as soon as practical and include the analysis results achieved with the improved models and methods now being developed. This 4
review is to be scheduled for completion concurrent with the completion of the SB LOCA Methods Program, June 1, 1982.
In addition, the Staff stated that a commitment to RCS vents and water level measurement would enhance their confidence in the Babcock & Wilcox system.
l Based on the above understanding, the following is agreed to:
l 1.
We will participate in a joint effort with B&W Owners and the Staff to
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ensure that current SB LOCA methods and ATOG efforts are fully understood
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and considered in an integrated manner. This program will include the i
following:
i Code parameters, models, assumptions, etc, which are important in controlling dynamics of interest will be identified and available experimental data substantiating their validity will be reviewed. This will be done using results of the improved evaluation model in order that l
the most accurate dynamic response characteristics are reviewed.
i If as a result of the joint effort to review the above, specific technical I
gaps, significant to the understanding and confidence of the B&W system, $cO are determined to exist, experimental tests will be defined to develop S
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2 additional data bases. This process should assure that questions which must be answered are answered in the most appropriate technical and cost-effective manner.
This program is scheduled for completion by June 1, 1982 concurrent with the SB LOCA program schedule and is based on adequate availability of appropriate NRC Staff.
Consistent with this effort, the Staff is expected to continue to review and approve the SB LOCA code revisions by April 1982 as previously scheduled.
2.
As requested in the meeting, we are providing a summary of the philosophy underlying AToG. The attached paper, " Abnormal Transient Operating Procedures for Nuclear Power Plants" by J J Kelly and D H Williams, includes the total scope of the ATOG approach as well as the philosophy presented by Dr Roy in this presentation on October 23.
As noted in Item 1 above, an understanding of ATOG is an essential element in the SB LOCA program analysis.
3.
In order to provide further support to the Staff's confidence in the B&W designed systems, we will install RCS vents in the pressurizer and the hot leg as described in FSAR Section 5.4.15.
A water level measurement system for the RCS will be provided as committed to in FSAR Volume " Response to NRC Questions, Vol 3",
Post TMI-2 Issues and Events, Q221.16 (Rev 38 dated 9/81).
4.
In response tc Dr Mattson's question, information on the formation of steam during normal operating transients, including rapid cooldown, is included as part of the ATOG program development.
We believe this letter reflects the major conclusions and agreements reached in the October 23 meeting.
The B&W Owners Group is prepared to begin the review of our SB LOCA Methods Program with the Staff and request that the assigned Staff reviewer contact Mr Lou Lanese of GPUNC to establish the details for carrying out this review.
It is our intent to implement this review without impacting the current schedule for the completion of the SB LOCA Methods Program.
JWC/LSG/fms CC RJCook, Midland Resident Inspector DSHood, US NRC TPSpeis, US NRC DBMiller, Midland Construction (3) (w/o enc)
RWHuston, Washington (w/o enc) oc1181-0432a131 J
Abnormal transient operating procedures for nuclear power plants J. J. Kelly, Jr., Supervisory Engineer Engineering Department Nuclear Power Generation Division Babcock & Wilcox Company Lynchburg, Virginia D. H. Williams Special Products Coordinator Arkansas Power & Light Little Rock, Arkansas PGTP 81-31 Presented to American Power Conference Chicago, Illinois April 27-29,1981 I
The aims, objectives and methodology involved in symptoms and immediate actions. If a loss of producing abnormal transient operating procedures feedwater occurs, he is expected to recognize it, for nuclear power plants were discussed in detail at perform the appropriate immediate actions, and the American Power Conference in April,1980.*
then use the event-oriented loss-of feedwater Abnormal Transient Operating Guidelines now procedure for determining follow-up actions. This exist in draft form for Arkansas Power and Light, approach has several inherent drawbacks:
and it is possible to detail a truly symptom-oriented
- 1. At time zero, the operator must correctly approach to transient management. Using this diagnose the initiating event. Ile does this approach, the operator does not have t mentally, based on training or prior experience.
immediately diagnose the initiating casualty and After taking several actions, depending on this locate the event-specific procedure for that casualty.
instant evaluation, he then refers to the event-Instead, the operator can pick up and use one.
oriented procedure that fits his diagnosis. If he simple procedure for all transients startmg with a were to treat a small steam line break inside the reactor trip. This paper describes the approach and reactor building, but actually had a smal' loss of provides several examples of the simplified decision coolant accident (LOCA) inside the building, he making process now available to the operator. Als would be tracking through the wrong procedure.
discussed is one possible anrroach to implementmg He would eventually recognize this these new guidehnes m', ue existing procedure misinterpretation: however, by then he would be structure.
well into the transient and possibly confused.
- 2. Procedures must be written to cover every
Background
conceivable initiating event. If the operator The traditional approach to transient and accident correctly diagnoses a loss of nonnuclear control has been to deselop many " emergency" instrumentation power and no procedure covers procedures, each based on a postulated event such that event, his actions will be based only on as loss of main feedwater. The operator is then experience.
required to study this event and memorize its
- 3. If more than one event contributes to the
- " Engineering basis for operator control of nuclear power transient, the operator will find himself working stations in abnormal operations - closing the loopf two or more procedures at the same time. For E. A. Womack. J. J. Kelly. and N. S. Elliott American Power instance, if a main steam safety valve failed to Conference. Chicago Illinois. April 21-23. 1950 IBabcock reseat following the loss of main feedwater, the
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operator would have to use the loss of feedwater 1
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procedure and small-steam-line-break procedure identify a transient. Similarly, some parameters (if available). These procedures may conflict and were common to all transients. One event found he would have to decide a priority between them throughout the study was a reactor trip.
- with no convenient method of shifting Consequently, it was used as the key for f
between the two procedures. Writing e procedure entering the abnormal guidelines procedure.
to combine these two events is ponible;
- 2. Event trees for various initiating events
- and however, if just a few more failures are consequential failures were developed. These considered (e.g., the power-operated relief valve included various multiple failures (including or spray valve remains open), the number of operator error), and therefore covered a large combinations of failures, along with possible number of possible scenarios. Event trees were initiating events, quickly increases. Even if studied to find repetitive patterns and common writing the approp-iate procedures was end points. The study showed that although atterapted, the operator's ability to pick the many failures can occur, the symptoms of correct procedure would certainly diminish.
unbalanced heat transfer that result from these
- 4. Because of these limitations, most operators are failures followed a few common patterns or likely to use no specific procedure. They will use trends.
training, experience, intuition, etc., to bring the
- 3. Actual operating transients were investigated, plant under coritrol. They will then choose what again looking for patterns. This time the they think is the closest procedure to what is emphasis was on parameter trends and the time happening and confirm their actions or see if avai:able for operator action.
they overlooked anything.
- 4. Where necessary, computer simulations were run to complete the baseline and fill in gaps in To correct these deficiencies, it is necessary to understandm, g plant response. Because the step back from the traditional approach and utput was mtended for use m developmg g
examine what the operator is attempting to do peratm, g guidelines, realistic mput was used (as during reactor posttrip abnormal transient control.
pposed to boundmg safety analysis He can best protect the health and safety of the assumptions).
public by guarding the integrity of the core. To do this he must ensure the continuous removal of This investigation's conclusion was that the decay heat from the fission products to the ultimate operator can track the removal of decay heat from heat sink. By adjusting the priorities and the core to the ultimate heat sink by monitoring concentrating efforts on maintaining proper heat just a few symptoms which reflect the " health" of transfer along this path, he can protect the core and the thermodynamic process around the reactor minimize radioactive release. To give the operator coolant system and its coupling to the secondary this capability, a concept of symptom-oriented (as side.
i opposed to event-oriented) procedures was investigated. The symptoms are based on upsets in Symptoms identified heat transfer from the core to the coolant and from the coolant to the steam generators. The symptom.
The three symptoms of primary interest to the oriented procedures thus focus on core cooling first pressurized water reactor (PWR) operator are and on event identification second. The result of adequate subcooling of the primary system this investigation is the Abnormal Transient inventory, inadequate primary-to-secondary heat Operating Guidelines (ATOG).
transfer, and excessive primary-to secondary heat transfer. These symptoms are important for the Expected plant response following reasons:
To produce a symptom-oriented procedure, B&W
- 1. Adequate primary inventory subcooling. If the developed a thorough understanding oi' expected operator knows the primary fluid is in a liquid plant responses during many varied abnormal state, he is assured that it is available and transients. These transients included classic capable of removing heat from the core and I
singular initiating events as well as additional transferring it to the steam generators. If single and multiple component failures. The subcooling is lost, these issues are in doubt, and procedure was developed through the following he is therefore directed to make every effort to
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regain subcooling.
- 1. Existing plant casualty procedures were
- Included as initiating events were loss of main feedwater. loss investigated for common symptoms. Fev single of offsite power. excessive main feedwater small steam tine alarms or parameters were found to uniquely break. and steam generator tube rupture.
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- 2. Inadequate primary-to-secondary aeat transfer.
This symptom addresses the heat transfer
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- 3. Excessive primary-to-secondary heat transfer. In 400 450 500 550 600 650 this case, the symptom is mdicative of a Reactor coolant and steam outlet ten.oerature, F secondary side malfunction (e.g., loss of steam pressure control or steam generator overfill). The
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L and for Natural Circulation (Twa) heat transfer is again unbalanced and the operator's attention is directed toward generic s Normal Operating Point - Power Operation (Tng) actions to restore this balance.
By tracking these basic symptoms the operator
] End Point - Post Trip with Natural Circulation (Tng) can quickly focus on problems without checking a Figure 1 Basic pressure-temperature (P-T) display large number af parameters. At the same time, by their nature the symptoms allow rapid elimination is also input. The saturation temperature for this of problem sources and continue to emphasize core input pressure is displayed as a vertical line. The protection. Additionally, the symptoms are so basic subcooled margin line accounts for potential instrumentation inaccuracies with the objective of that the procedure inherently covers many more assuring subcooling above that line.
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initiating events than those first ctudied. This happens because initiating events cause equipment A typical plant response to a reactor trip is shown in Figure 2. For simplicity, only reactor to fail, and equipment failures affect these symptoms. As the operator follows the procedure to y
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the symptoms of interest. The solution developed in
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400 450 500 550 600 650 a pressure-temperature (P-T) display with a Reactor coolant and steam outlet temperature. F saturation curve included. The area above and to the left of this curve is the subcooled region. The Figure 2 Typical posttrip response area below and to the right is the superheated coolant hot leg temperature is plotted. With the region. Reactor coolant system hot leg temperature reactor coolant pumps running (forced circulation)
(T ot) and cold leg temperature (Teoid) are input to and the comparatively small amount of energy h
this display and plotted as functions of reactor being added to the coolant by decay heat, the cold coolant system pressure. Steam generator pressure leg temperature is also expected to settle out close 3
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I the a T across the steam generator tubes is small, remain within a "posttrip window," the plant is both of these temperatures should approach the responding normally.
saturation temperature of the secondary side of the With this type of display, the symptoms of steam generator (SG Tsat). The Figure also shows interest are highlighted and brought into focus for steam pressure moving from its pretrip value up to the operator. Consider the example in Figure 3.
the steam safety valve setpoint and back to its Combinations of these symptoms are also easily P
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400 450 500 550 600 650 400 450 500 550 600 650 Reactor coolant and steam outlet temperature. F Reactor coolant and steam outlet temperature. F Figure 3A Inadequate subcochng margin:
Figure 3C Excessive primary-to-secondary heat transfer:
Tnot is not progressing toward its target value; in SG Tsat has decreased below its estabhshed limit.
fact, it has rapidly dropped through the subcooled Tnot and Tcoio have reached equal values but both margin kne. This condition is diagnosed as loss of have gone out of the posttrip window following SG adequate primary inventory subcoohng, or simply Tsat. This condition is diagnosed and treated as
" inadequate subcoohng margin." and the excessive primary-to-secondary heat transfer.
procedure is written with directions to take care of inadequate subcochng margin, recognized. Consider the example m, Figure 4, taken from the first twenty minutes of the Tall-2 9 2400 l
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Other display arrangements of basically the same fundamental parameters have been developed with similar effectiveness, still relying on basic patterns 400 450 500 550 600 650 Reactor coolant and steam outlet temperature. F of parametrlC change to mdicate an overall plant ad Figure 3B Loss of primary to-secondary heat transfer:
Tnot is increasing as SG Tsat is decreasing. A.1T between the two is growing larger. The secondary is ATOG organizat. ion no longer removing heat and has lost couphng with Once the symptoms are identified and a method of the primary. This condition is diagnosed and treated as loss of (inadequate) primary to-secondary heat monitoring those symptoms developed, the ne3+
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Figure 4C 8 to 15 minutes:
At time zero the reactor has tripped on loss of Steam pressure and temperature have recovered to feedwater. At % minute Tnot and Tcoioare essentially their normal posttrip values. A substantial coding the same temperature. At 2% minutes the ESFAS of the primary is also in progress.
pressure setpoint is reached and high pressure injection (HPI) is automatically started. At 3%
minutes subcooling margin is lost, and at 4%
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Primary-to-secondary heat transfer (coupling) is y 400 a,aproaching their normal posttrip values. However, now almost completely restored. Tnot and Tcoio are the inadequate subcooling margin is evident.
400 450 500 550 600 650 Reactor coolant and steam outlet temperature. F design bases for, and the use of, the procedures.
Figure 5 outlines the organization of Part I. The Figure 4B 5 to 8 minutes:
The primary continues to heat up along the immediate actions are common to every reactor trip saturation line while secondary temperature and and must be performed regardless of the cause. The pressure drop. At 8 minutes the primary-to-Vital system status verification is a short checklist feedwater is frst di used to determine a baseline for possible operator ted o he st m gener tors actions. This checklist considers mstrumentation useful to the operator. The Abnormal Transient power supplies, engineered safety features Operating Guidelines consist of two parts. The first activation system (ESFAS) status, steam line break part is procedural guidance to be used in the protection system status, etc. Included in this control room during transients. The second part, a checklist is a requirement to monitor the ATOG much larger volume, is a training aid explaining the display. If everything is normal, the plant has 5
i intent as to why various steps are taken in Part 1.
Section i It also describes, using m my graphic examples, the immediate Actions expected plant response information gathered during the guideline development stage. Part II has been written to aid the operator's training and is tal s stem status venfication important to the guidehnes because an intelligent, Section ill capable operator is a basic part of the plant A. Treatment of lack of adequate subcooling margin operating structure in which the guidelines are built B Treatment of lack of primary-to-secondary heat transfer (i.e., the guidelines try to optimize the operator's C. Treatment of too much primary-to-secondary heat transfer gg Cooldown procedures
. Large LOCA Guideline validation Once written, the potential guidelines were tested ted RCS u
on a PWil simulator by imposing multiple e HPl cooling e Solid water cooldown casualties and using the guidelines to recover.
Guideline credibility was also established by back checking the guidelines against event tree paths Figure 5 ATOG - organization of Part I and benchmarking event tree paths and computer responded as designed and come to a steady post-simulations against actual plant transients. The trip condition. No further action is required.
event trees were also reviewed by the utility However, if the operator diagnoses an imbalance in operators to take advantage of their plant one of the basic symptoms, he is directal to the experience. The draft guidelines were sent to the appropriate section for follow up actions. These plant site for walk-through drills to test their sections treat the symptoms and do not require the applicability. Feedback from the operator to the operator to determine the cause. It is expected, plant designer served to greatly reduce however, that as he treats the symptoms he will communication errors and increase confidence in the find the original problem.
final guidelines.
Treating ute symptoms will allow returning the An important final step in validation involves plant to a stable condition. This stable condition implementing the guidelines into the plant could very well be abnormal compared to what the procedures system. This implementation tests the operator normally sees. Accordingly, varieus guidelines' scope and appropriateness since they cooldown procedures are provided to give him must be a workable part of the overall plant '
guidance on long-term recovery from these possible procedures system or their worth diminishes.
conditions.
Existing posttrip procedures must be checked Figure 6 outlines the organization of Part II.
against the guidelines to determine the following:
Intended to give the operator a thorough
- 1. Necessary actions outside the developrnent understanding of Part I, it conveys the writer's program scope but needed for a posttrip procedure in the same time frame. This assures f
volume 1 Fundamentals of reactor control for abnormal transients that, although everything may not be A. Heat transfer considered, the adoption of ATOG does not B. Use of P-T diagram decrease in any area the adequacy of procedures C. Abnormal transient diagnosis and mitigation vious level. Some actions in the D. Backup cooling methods E. Best methods of equipment operation previous procedures may be found good but not F. Stability determination necessary, and either be deleted or relegated to a lower level of instruction. The goal is to votume 2 maximize simplicity.
- 2. Actions that should be included in an instruction A Excessive fee w er for longer term action. Current posttrip B. Loss of feedwater C. Steam generator tube rupture procedures include many necessary follow-up D. Loss of off-site power actions that are not appropriate for ATOG, but E. Small steam line break must be included somewhere. Three actions, identification of these items, determination of the form in which they should be given, and Figure 6 ATOG - organization of Part 11 optimization of the interface between the form in 6
which they are given and the ATOG, are Summary necessary to make ATOG a workable part of the overall plant procedure system. Again, the goal By using the Abnormal Transient Operating is to maximize simplicity.
Guidelines, the operator can enhance plant safety
- 3. Any posttrip procedures not accommodated by by monitoring reactor posttrip parameters for only ATOG, but which must remain ir> tact. One goal a small number of symptoms and taking corrective of ATOG is to eliminate these procedures, but action as directed by the procedure. The guidelines that goal has not yet been proven consistently allow him to use one simple procedure for all attainable. Any such procedures identified must transients which start with a reactor trip. The be entered in a manner compatible with ATOG unique feature of this approach is that it provides a implementation.
common starting point, independent of initiating event, and leads the operator through a step-by-Although plant procedures vary from plant to step procedure to regain stable plant conditions plant, preliminary work m, dicates that portions of without having to identify either the cause of the all of the procedures, such as the followmg, may be transient or any additional posttrip malfunctions.
replaced by the ATOG:
- Reactor-turbine trip
- Degraded electrical power
- Loss of coolant!RC pressure
- Steam supply system rupture
- Loss of steam generator feedwater
- Steam generator tube rupture
- Loss of reactor cooling flow - RCP trip 9
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