ML19337B653
| ML19337B653 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/29/1980 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| FRN-45FR54708, RTR-NUREG-0696, RTR-NUREG-696, RULE-PR-50 45FR54708-7, 653, NUDOCS 8010080427 | |
| Download: ML19337B653 (39) | |
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TOLEDO
%s EDISON RCHARO P. Caoust
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"' S S-588' Serial No. 653 September 29, 1980 s
Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Chilk:
Toledo Edison appreciates the solicitation of comments on. UREG-0696, u
Functional Criteria for Emergency Response Facilities as announced in the Federal Register of August 15, 1980 (45 F.R. 54708). We feel, by sharing the comments attached hereto, a valuable insight can be gained into developing and implementing functional criteria that will truly improve the capability of utilities and regulators in upgrading their emergency preparedness posture.
Toledo Edison has dedicated all levels of management attention to emergency preparedness, an area that we feel has been one of our most productive efforts in light of the Three Mile Island Unit 2 accident in March of 1979. As a result, a functional criteria has been developed and is currently in the process of implementation. The resolution of pitfalls and the problems of integrating such a comprehensive response capability into an organizational structure has led to the development of a truly effective emergency response capability in our company.
Due to our commitment and an accelerated schedule, our facility is to be in full operation by late Summer 1981. In light of this, we would be happy to make the facilities availabla to the Commissioners and their staff to observe first hand its role in our overall emergency response program.
In retrospect. the development of our functional criteria has been both difficult and exhilerating. Our conceptual development was =ided greatly by the insights of a Toledo Edison Review Team that went to Threa Mile Island in July of 1979. They gathered first hand emergency response experiences of the Metropolitan Edison / General Public Utilities organiza-tion.
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/ l THE TOLECO EDISON COMPANY EOtSON PLAZA 300 MAOISON AVENUE TOLEDO. OHIO 43652 8010080 4 3,7
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4 The emergency capabilities of the utility, industry, local, state and federal organizations have to function to direct all the resources at their command to insure the public health and safety. The purposes of emergency response facilities, soc 6 af which are described in NUREG-0696, are to support centralized management of technical assessment, radiological assessment, governmental / industry interface, public infor-mation, and recovery activity. By proper management, the available resources can be properly directed under any conditions, foreseen or not.
The centralized management philosophy of overall emergency response was formally adopted by Toledo Edison in December of 1979. The functional requirements of our ciganization have been defined and t anslated into support facility criteria. These are being implemented n an optimized fashion to insure the overall centralized management concept is intact.
One of the main elements in our response facility upgrade is a combined function facility currently in construction at the Davis-Besse site boundary. This multi-million dollar facility fulfills our functional criteria. I consider this basic approach of centralized accident manage-ment as part of my defined responsibility in insuring the health and safety of the public. Degradation of this concept is not considered jrstifiable.
Comments attached reflect areas that need be seriously addressed prior to the final issuance of NUREG 0696. To provide details on the logic behind our comments, Toledo Edison's approach to these support facilities is used as examples where appropriate.
Very truly yours, 11 "
R. P. Crouse Vice President, Nuclear RPC:TJM db b/1-2 cc:
Commissioner John F. Ahearne Commissioner Victor Gilinsky Commissioner Peter A. Bradford Commissioner Joseph M. Hendrie Dr. Milton S. Plesset, Chairman, ACRS (16)
Harold R. Denton, Director, NRR Carl Walske, President, AIF
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i COMMENTS ON NUREG 0696 BY THE TOLEDO EDISON COMPANY FUNCTIONAL CRITERIA FOR EMERGENCY RESPONSE FACILITIES 1
i SEPTEMBER 1980 j
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4 Contents i
I.
Introduction II.
Comments on Major Items III. Other Conuments i
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I.
Introduction Toledo Edison has participated jointly through representation on the Atomic Industrial Forum's Safety Parameter Integration Group Subcom-mittee and the utility sponsored KMC, Inc. Coordinating Group on Emergency Preparedness Implementation. Additionally direct interfacing with the NRC staff on acceptability of Toledo Edison's centralized accident management support facilities has provided dialogue with the drafters of NUREG 0696 throughout its development.
Comsants have been provided before and are being provided again jointly. Toledo Edison endorses comments from both the AIF and KMC groups. However, the importance of developf.ng appropriate functional guidance for facilities to support emergency response dictates that our Company provide comments independently on several areas of major concern as well as minor comments.
In general, these comments arise from what we believe is a basic lack of appreciation that the equipment and facilities only support the response organizations.
It is the organization that responds to the emergency not the facilities.
II.
Comments on Major Items A.
Data Information Systems 1.
Basis of Concern An obvious shortcoming of the onsite and offsite response functions in the past has been the lack of accurate information.
It is recognized that each activity center needs information. However, the determination of what information, its timeliness and its display format needs to be developed based on the individuals to whom the data is directed and their function in the response organization.
The only information display systems dicussed in NUREG 0696 include an SPDS in the control room, a TSC informa-tion system, an Emergency Operation Facility information system and an information system _to the NRC Operations Center, the Nuclear Data Link. These systems are discussed as if there is only one type of individual in each activity center and that the need for information is instantaneous and digital.
Further both the text of the document and meetings with its authors treat the SPDS as the sole information source illustrative of the plant conditions during all modes of plant operations.
If one starts from this assumption, the design criteria evolved would logically be as pre-scriptive and harsh as currently identified in NUREG 0696.
However, it is our major contention that not only is the basic assumption in error, but the operational philosophy that spawned this approach goes directly counter to one of the root problems during the accident at TMI-2 -- the reliance on one device to interpret plant status.
l No ma, e how available, reliable or omniscient one infor tion source is designed to be, we have learned the i
absolute ecessity to verify, diversely, plant data and j
em status. This is one lesson we do not intend to rescind.
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.o focus on the true need of strict requirements of the SPDS, let's reflect on several activities related to the functional response located in the control room. These activities include:
Expanded Controi Room Organization - This includes a.
an operations team of reactor operators and senior reactor operators that have undergone intensive retraining. A Shift Techcical Advisor is now assigned to provide a broader technical expertise to advise the shift operations organization. This staff is augmented during emergencies with addi-tional senior operations staff and high level sta-tion management and communications persons.
The important relationship here. is that different information is important to different functions in tha control room organization chain. An SPDS dis-play of interest to a Shift Technical Advisor would not necessarily be the same display provided for a reactor operator. An extremely flexible computer based data acquisition and display system is impor-tant to be able to address the functional differ-ences of the individual's needs in the control room.
b.
Control Room Information Systems - The SPDS is only a small part of the control room. A control room evaluation is to be done at every nuclear power plant to ensure the man-machine interface is good enough to overcome confusion due to expected events.
New instrumentation has been added with more being upstaded. The SPDS is one more operator aid that, like any other individual device, must be able to be done without. This is regardless of any high relia-bility and availability design goals.
c.
Long Term Procedure Upgrades - A complete change in procedural philosophy is in development.
It includes the conversion from event-oriented plant procedures to symptom-oriented procedures. The goal of this effort is to be able to initially respond to a plant upset in a manner that protects the core during any type event without requiring the control room organi-zation to know the initiating cause of the condition.
In our perspective this effort is the most safety effective item in the post-TMI-2 activities. These symptoms are to be recognizable with or without an SPDS operable or any other individual device.
d.
Short Term Procedure Upgrades - All plant emergency procedures for upset conditions have already been modified to require verification of vital information regardless of the reliability or design pedigree of the source instrument.
2.
Toledo Edison Approach In the Area of~0ata Information Systems it was recognized that the functional needs of the Technical Support Center, Emergency Operations Facility, Control Room and Nuclear Data Link could be aided by a flexible system that could be used in different formats for appropriate evaluation.
The development of the human factors relationship between all the different functions will evolve over a considerable time. Recognizing this, the system approach selected by Toledo Edison is computer based, flexible, powerful, reliable and proven. Figure 1 depicts the basic system.
This will allow the effective development of different display formats to be optimized through an evaluation process by the functional users.
Initially, Toledo Edison is placing a high priority on the capability to assess the immediate post-reactor trip time period. Because this is essentially the first 10-15 minutes of an event this function is required to be done by the control room shift organization without the support of a TSC or EOF. The manipulation and display of key parameters has been shown to provide a significant operator aid in the initial assessment period. Appendix A attached.
describes one such approach. Entitled A Real Time Method For Analyzing Nuclear Power Plant Transients, Messrs. Broughton and Walsh of the GPU Service Corporation described one method that 'could fulfill an operator aid requirement. A computer-based display system is ideal for this approach in the man-iachine interface. However, the technique is equally functional being hand plotted from information available in thE control room without an SPDS.
This tackup method to support such a function allows the approach of maintaining flexibility with one system to provide a wide variety of information and displays to all emergency response activity centers with off-the-shelf, highly dependable, maintainable and available equipment.
Figure 1 illustrates the syster selected by Toledo Edison. The basics are modeled after the data acquisition and display system at the Loss of Fluid Test Facility (LOFT).
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Additionally, a special loop communications system has
'been devised to allow a dedicated voice system for verbal verification of information at each of the activity centers. Closed circuit television viewing of the control room will be in the Technical Support Center.
3.
Proposed Changes to NUREG 0696 Page 8 - Revise Section II.F. to read:
F.
Safety Parameter Display Design Criteria The total SPDS need not be Class IE or meet the single failure criterion.
The data acquisition system for the SPDS, consisting of sensors and signal conditioners, shall be designed and qualified to Class IE standards.
The processing and display devices of the SPDS shall be of proven high quality and reliability.
2 A Limiting Condition for Operation in the Technical Specifications shall be established that is consistent with the unavailability goal of the SPDS and with the compensatory measures defined during periods when the SPDS is inoperable.
Since the function of the SPDS is to aid in the detection and monitoring of transients and accidents, the SPDS shall be capable of functioning during and following most events expected to occur during the life of the plant.
Emergency operating procedures shall specify the limitations I
of the SPDS.
B.
Facility Locations 1.
Basis of Concern Emergency planning is just that, planning organizations and support facilities established to be able to effectively reJpond to events.that are expected.
Whether it is in
. the commercial nuclear power industry or any other field, the best planning does not account for every possible contingency In fact, the broader the spectrum of emer-gencies the 6:.rder it is to define effective organizations and managenett support facilities. One basic reason is the nature of the required response to one particular emergency may.ury to a point that a preconceived approach j
may need to be altered because it is not appropriate for the particular condition.
The correct approach then is to optimize support for a management organization in the i
expected modes of response while providing for flexibility to cope with the unexpected.
-S-Effective, flexible management depends on information availability, internal organization communication and external communicati.on. Effective communication includes not just technical but personal interchanges.
In the commercial nuclear power production arena, these inter-changes are vital cue to the many organizations and responsibilities involved. Each utility has to determine the optimum interrelationship it needs to support expected emergencies. Areas of potential weakness are then compen-sated for by strengthening other facility support capabil-ities.
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It is Toledo Edison's contention that this evaluation could be unique to different organizations and therefore support facility types and locations can be quite varied f
but still be adequate in performance.
As an example, the Technical Support Center in NUREG 0696 has been prescribed to be "within approximately two minutes comfortable walking time of the control room."
The funct!onal items this is to support is face-to-face communicatiocs "between coutrol room personnel and senior i
plant management working in the TSC" and information availability of ites: not transmitted to the Technical Support Center. By evaluating the expected conditions requiring face-to-face communication two time periods arise. The first is some unplanned event of concern.
Under this condition however, such removal of activity center managers for any time period could critically impair response capability.
In addition congregation of personnel in the control room could interfere with or distract the operations staff. Therefore, during unplanned events such face-to-face exchanges between control room personnel and senior plant management should be minimized The other events of concern would be events that had been pre-planned and may require briefings or observation. By their nature of being pre planned _a two minute walking distance can certainly be expandable.
The other functional aspect of the location is to acquire information not available in the TSC. Certainly, in today's era of information transmittal, addtional viable options can be provided for information gathering.
2.
Toledo Edison's Approach Figure 2 identifies the basic functional criteria utilized to determine locations of support facilities.
In developing our facilities, several options were evaluated. All had shortcomings to varying degrees. The final corporate commitment was made on the ability of facilities to support a controlled, centralized accident management organization. As a result, e
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6-Toledo Edison's emergency response organization is set up in three locations:
the Control Room, an Emergency Plan Facility at the site boundary (containing the TSC and EOF functions) and the Toledo Edison Plaza offices twenty-seven miles away.
Figure 3 identifies the internal arrangement of the Emergency Plan Facility while Figures 6 and 7, locate the facility with respect to the site.
There are key areas of the response organization, equipment and facilities that were upgraded to address initial shortcomings.
I They include:
Need for control room access for information - an elaborate a.
data acquisition and display system was selected to make available not only a minimum number of plant parameters to the TSC and EOF but to essentially access the total library of the plant computer. A video link from the Control Roca to.the TSC is being provided as well as a dedicated loop communications system that will place the Control Room on line with the technical assessment area and the radiological assessment area.
I b.
Need to have face-to-face interaction between senior station management personnel in the TSC and Control Room personnel - the response organization at Toledo Edison supplements the Control Room staff with senior station operations personnel and the Assistant Station Superin-l tendent. The Technical Support Center response organization includes the Station Superintendent and the Manager of Nuclear Engineering. The detailed day-to-day interactions of these persons provide valuable experience that will aid in the capability of handling verbal discussions and directions.
Additionally, to allow rapid transportation between the TSC and Control Room, a dedicated road is being constructed totally within Toledo Edison's control. Transport time utilizing a dedicated vehicle from the TSC to the Control Room is easily under five minutes.
Emergency Operation Facility Survivability - Toledo c.
Edison located the EOF functions at the Emergency Plan Facility. The entire first floor is designed to be habitable for the same radiological event as the Control Room (Ceneral De<Jgn Criteria 19).
Location of this facility here is considered to aid greatly the more critical face-to-face communications between different governmental / utility / industry / media response organizations and allow timely briefings within or among any of these crganizations.
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1 To arrange for flexibility to cope with unexpected events the Emergency Plan Facility provides for:
a.
An entire second floa-of offices that normally will i
house selected elements of the station staff.
I b.
An extensive communication system to ensure indepen-I dence from local phone overload conditions.
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An internal security and badging system to aid in the recognition of organizational authorities.
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d.
A power supply system that provides diesel generator backup as well as a battery backed uninterruptable power supply (UPS) system for critical data and communications functions.
Beyond the site response organization is a support engi-
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neering function that reports to the Toledo Edison corporate offices at the Edison Plaza. A data link provides all information available to the SPDS, TSC and EOF such that j
this organization can support, monitor and assume a i
direct response function if required due to extreme unforeseec site conditions.
j The accident management organization is well supported by
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the locations of its facilities. Additionally, it has the flexibility important when trying to design facilities for all possible events, foreseen or not.
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3.
Proposed Changes to NUREG 0696.
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Page 10 - Revise Section III.B. to read:
I B.
Technical Support Center Location I
The requirement for an inplant TSC was established to provide facilities for derailed analysis of plant 4
conditions and to alleviate the problem of control room overcrowding during an accident. The TSC shall be the emergency operations work area for designated senior plar?. management personnel, designated licensee engineering and technical personnel, a small staff of NRC personnel, and any other licensee designated personnel needed to provide the required technical support.
The TSC shall be located to readily allow face-to-face interaction between control room personnel and a
the senior plant management working in the TSC.
The TSC shall normally be in a location that is within approximately five minutes of the control room.
4 Provisions shall be made for the safe and timely movement of personnel between the TSC and the control room under emergency conditions.
III. OTHER C0KKENTS A.
Nuclear Data Link (NDL)
The Nuclear Data Link is still in its formative stages, but is being based on a consideration that has not yet matured.
Fundamental to the establishment of the NDL is a clear deter-mination by the Comm!ssion and an understanding by the staff of the function and role of the NRC in an emergency.
T5is is identified as Action Plan Task III.A.3.1, and is still ongoing.
Resolution of this concern is an importaat prerequisite to development of the NDL (which is Action Plan Task III.A.3.4).
As such, that part of NUREG-0696 that relates to the NDL, if retained in the final version, should be considered as informa-tion only, with no implementation inference at this time.
1 B.
NRC Regulatory Guide 1.97 All references to Regulatory Guide 1.97 (Instrumentation to assess and follow the control of an accident, etc.) should be 1
qualified. Reg Guide 1.97 is not final yet.
The parameter sets for the emergency facilities should be based on the function of each facility. The Atomic Industrial Forum's Safety Parameter has recommended to the Advisory Committee on Reactor Safeguards (ACRS) that a systematic. approach be used to establish the data requirements for emergency facilities'.
This approach, is contrast to Reg Guide 1.97, integrated the consideration of human factor engineering, the need for and importance of the information, and the-function for which the information is going to be used. As a result, the ACRS did not endorse Reg Guide 1.97 in its present form and recommended that additional effort be made to resolve some of the rather major differences in the approach between NRC staff and industry.
C.
Availability Availabilty of information and power supply requirements need to be defined with respect to function (purpose) of the SPDS, TSC, etc.
Unavailability.1hould not mean loss of a single input parameter but loss of the function of each Emergency Response Facility.
Design availability (or unavailability) should be defined using standard manufacturers data such as Mean Time Between Failures and Meau Time To Repair and should be based-upon actual historical or generic data.
Commercially available computers typically have an advertised availability of 99.5% and when used in conjunction with available input / output devices and power supplies overall availability of 99.0% in achievable. To meet the 99.9% availability re-quirement would require redundant computer systems, input / output devices and power sources. To statistically demonstrate an availability of 99.9% with a confidence level of 95% would require a test period of approximately 400,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
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- Minimize Functional Impacts During Protective Action
- Address Known NRC Guidance and investigative Group Recommendations 1
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APPENDIX A A REAL-TIME ME" HOD FOR ANALY ING NUCLEAR POWER PLANT TRANSIENTS T. G. Broughton P. S. Walsh GPU Service Corporation l 100 Interpace Parkway { Parsippany, N.J. 07054 i )
- Discussed in ANS Transactions, Voltme 34, TANSAD 34 1-899 (1980) pages 723-724 l
A REAL-TIME METH0D TOR ANALYZING NUCLE AR POJER PLANT TRANSIENTS T. G. Brougacon P. S. Walsh GpU $ervice Corporation 100 Interpace Parkway Parsippany, N.J. 07054 Introduction Monitoring power planc performance during transients and detecting abnormal performance requires comparing many parameters to limi:ing values, deter-mining correlations among parameters, evaluating trends and verifying ene status of key systems and components. This task may be difficult to perform in real time during unexpected cransients. After-the-face analysis of cae Int-2 accident, for example, revealed enac the multiple malfunc: ions wnica occurred could have been diagnosed from the values of and relacionships between a few key parameters. The presence of additional data of lesser importance, however, resul:ed in confusion and innibited mitigating accions. This accidenc and other plant transients demonstrate a clear need for a real-cime analysis method useable by operacoes. Effoets to develop a diagnostic method concentraced on heat transfer and pressure control following a' reactor trip. A method of determining the ef fectiveness of steam generator heat removal was particularly desiracle. Many alternece approaches were evaluated and pocential machods were refined j or eliminated in an attempt to develop a cool which was based on sound j principles, could be consiscencly applied over a wide range of expected events and could be easily learned and used by operacors. An analysis method using primary and secondary pressures and primary temperatures proved to provide operations personnel wi:a the desired cool to diagnose power planc status in real cime following a reactor trip. This machod has been specifically applied to pressurized water reactors vita onca carough sceam generators. However, the principles of tais mecnod are applicable :o other light water reactor :ypes. As a test of tnis method, accaal plant data froi~more chan twenty transients as well as data from compucer simulacions of ::ansients have seen ploccad to evaluace the ef fectiveness of cne cachnique. All cne examples used in cnis paper are based on daca from actual planc cransients. Mecnod During the cransition from power operacion to post trip decay heat removal four key parameters (primary and secondary pressure and hoc and cold leg
- emperatures) are monitored using pressure /:emperature plocs (Figure 1)..If normal pos c : rip decay heac removal condi:iore. are accained, camperacare and pressure will de wi nin eneir expec:ed range, as indicated by :he box on :ne 9
O I ploc. Normal operacion at power is outside the expected pose crip range. The normal cransition to stable pouc crip condicions requires 5 to 10 minutes. Ac any point during the transition, che ploc indicates the proximity to che expected value and any limiting values. In addition, the crend allows the "analysc" to anticipace the course of ene transienc. A single plot which inclcdes primary and secondary condicions has proven most useful (Figure 2). The expected post crip ranges are indicated for cae primary and secondary systems. The saturacion curve is also included. The i primary plot of primary pressure versus hoc leg temperature directly indicaces primary system subcooling (saturacion margin). The secondary plot of steam generator pressure versus cold leg temperature indicates the effects of the steam generator on the primary system. Effective steam I generacor heat transfer is reflected by cold leg temperature (steam generacor'ouclec camperature) nearly equal to che sar1 ration camperature for steam generacor pressure. This relacionship should. isc following trip i since at decay heat levels the relatively low heat transfer in ene steam generator results in a small temperature difference across cne sceaa j generacor cubes. Assumocions and Limitations Since this monitoring agchod only considers four key paramecers, it cannot diagnose all abnormal power plant condicions nor can it determine che specific cause of cae abnormal condicions wnich exisc. For example, a small steam generator tube leak (leak rate less enan makeup capacity) would not affect these enormal hydraulic paramacers but would be indicated by secondary system radioactivity. The cause of overcooling events cannot be I determined without additional steam and feedwater system data. Therefore, this mached should be considered a supplement to che normal monitoring of systeh and component scacus and other parameters import ne to safecy. Other assumptions are implicit in the use of this machod. The most important is taac the reactor is shutdown since cne expected values and ) crends whien are che casis for diagnosis are valid only ac decay heat levels. Of course, instrumencacion accuracely displaying the monitored paramecers mu,c be available to pernic analysis. Also, it is necessary to i know wnecher t'orced primary flow or natural circulacion exiscs since ene expected values and aonormal condicions limits are dependent on cne flow race. Normal pose Trio Perfcesance Figure 3 is an example of normal pose crip performance following a loss of feedwater. The condicions juse prior to che crip (cime = 0 minutes) are normal for full power. The loss of feedwater causes tne reactor and curoine to crip. During the first minute steam pressure rises to che safety valve secpoint due to reduced steam generator heat removal following turbine crip, then decreases to the post crip curoine bypass valve control secpoinc. Hoc les camperature decreases and cold leg camperature incree -es as a result of reduced reactor power generacion. From t! tse condicions the transition to. e r ~
1 stabl. pose trip decay heat removal begins. The cooldown causes a primary system contraccion which reduces primary pressure. As cne cooldown race decreases, primary pressure recovers to che expected range. As the j secondary system pressure is controlled at a constant value by curoine bypass valves, tne cold leg cemperature decreases to cae corresponding sacuration camperature. These pressure /camperature traces are cypical of normal transitions. However, che condicions at the cime of,reaccor crip, che exacc transicion path and the cima to reach tne expected values vary wica che specific cransient. Detecting Abnormal Performance Several cacagories of aonormal performance can be diagnosed using this ploc. Certain regions on ene plot which indicace sonormal performance can de defined using limic lines. However, che crend of cae data provides the earliest indicacien of abnormal performance. 1. LOCA (Figure 4) - Loss of coolanc accidents and loss of pressure events are reflected by a continuous decrease of primary pressure to che saturation curve or below (superheac region) witnouc significant camperature decrease and no ef fect on sce.sm generator condicions. Pressures below the hign pressure injection secpoint or below a predefined subcooling margin may be considered indication of a LOCA. 2. Loss of Heat Sink (Figure 5) - Loss of sceam generacor heac sink events are re flected by the secondary ploc crending away from the saturation curve. This indicaces an increasing camperature difference between cold leg camperature and steam generator saturacion camperature indicacive of j d'egraded heac transfer condicions between the primary and secondary. This ploc defines a loss of steam generacor heat sink region as cold les camperatures greater than 35*F from cne sacu:acion curve or cola leg camperature more enan 25'T above ene expected value. The cocal loss of primary hsac sink is reflected by primary camperature increasing above che expected range. The limic for this zone has been sec ac 25'F above che expected value. 3. Overcooling (Figure 5) - Overcooling events are re.flected by camperature l decreases below cne expected range. On chir plot cne overcooling region is considered to de camperatures more enan 10*F below the expected value. The camperature decreases can cause or be caused by a steam generator pressure decrease. A rapid cooldown will produce a primary pressure decrease wnile a slower cooldown may not affect primary pressure. other limics can be indicated on the ploc, for example, primary system relief valve settings and steam line rupture isolacion system secpoint. One of the key advancages of this cachnique is the ability to distinguisn between a LOCA depressurizacion and a rapid overcooling which results in low prisary pressure. The LOCA is reflected only in ene primary crend but cne sajor indicator of overcooling is the secondary crend..
l 3 - ~ 1 I: is also possible to determine if a low steam generator pressure is due to j loss of heat transfer or overcooling. This distine: ion is importan: since ene correct action for loss of heat transfer (addi: ion of feedwater) is i'acorrecc for the overcooling case. j i 1 Asymmetric and Mul:iple Casualties Ploccing each loop individually on the same grapn ennances early recogni: ion of asymmetric transients. Figure 7 illustraces a stuck open turbine oypass valve in the A loop steam sys:em. At 2 minutes the overcooling trend on A is becoming visible on che secondary plot. Af ter 3 minutes the A loop continues to overcool, but the 3 Loop steam generator acts as a heat source to the primary (steam generacor saturation temperature is greater than cold ) leg temperature indicating heat transfer from secondary to primary). i Multiple casual:ies can also be diagnosed by this method. Many combinacions are possible including loss of heat sink in one steam generator with overcooling in another, LOCA combined with loss of heat sink or overcooling, or sequencially occurring faults. Figure S illustrates the multiple if fects of the TMI-2 accident. LOCA indications develop ac 2 minutes. High pressure injection (EPI) has no effect on the primary pressure decrease but does remove decay heac becween 2 and 4 L/2 minutes during the loss of steam generator heac sink. When EPI flow is reduced ac 41/2 minutes a loss of primary heat sink results. The primary reaches sa:uration condicions at 6 minutes and pressure increases as the system heats up. The steam generators are restored as a heat sink when emergency feedwater is iniciated ac S minutes. By 20 minutes the secondary system has oeen restored to the expected range but ene primary system still re fle,ces the LOCA. Na tural Circulacion Monitoring and diagnostic methods used during natural circulacion are idencical to forced circulation except tha t the expected value of ho: leg temperature is signer, che loss of primary heat sink limit is higner, and
- he increased loop transpor: cine presents additional caallenges in diagnosis.
The expected camperature difference between hot and cold legs during natural circulacion is 20 to 40*F versus :he 2 :o 3*F expected during the forced flow. The box indicating :he expected range following reactor : rip during natural circulation reflects chase differences (Figure 9). The expanded range of hoc les camperature results in a higner limic for definicion of loss of heat sink, 10*F higher than the maxi.aum expected noc leg campera:ure (Figure 9). Ef fects of Low Flow Race The long loop transport :imes during natura[ circulacion al:er the : rends sligh ly from the forced flow case but, more importantly, introduce time diffarences setween primary and secondary plocs which give : hem the -a-
an appearance of independence. The loop ::ansport :ime increases from acouc 15 seconds during forced fice to about S minutes during natural circulation. The effects on plo:ced data are sien in Figure 10. Pressure changes are re'flected in the data immediacely. In :he secondary sys:em, wnere pressure controls cold les camperature, the cima delay associated with cold primary water flowing from the steam generator to the cold leg temperature detector is about 2 minutes. There fore, to evaluate steam generator performance, pressure at time O should be plocted with temperacare at time 2 minutes. Although this corree: ion improves the presentation of :ne data, cne direction of the trend remains unchanged, and therefore applying the correccions in real time is of little value and not recommended. Tne close proximity of hoc leg comperature to che pressure sensor renders a correccion to tne primary system p' oc unnecessary. This delayed response of indicacion during natural circula: ion makes i: impor tant to provide the operator with a method of decaccing asnormal condicions in time :o con:rol che planc. Detecting Abnormal Conditions Figure 11 illustraces an overcooling event during natural circulacion. The cause of overcoeLing is overfeeding whien would be indicated by increasing steam generator level. The secondary plot provides early indicacion of :nis abnormal condicion (4 minutes). The primary plot begins to reflec: over-cooling 2 or 3 minuces later. The existence of adequate natural circulation flow must os inferred from other parameters in cases where low range flow instrumentacion is not provided. Loss of natural circulacion flow prevents heat transfer from the primary to the secondary and hoc leg temperature can be expec:ed to rise indicating a loss of heat sink on the primary plot. If sceaming continues in the secondary, steam generacor pressure will oegin to drop without reducing cold leg :emperature, also indicating a loss of heat sink. This will also result in an increasing differencial between hoc and cold leg camperatures. However, other events say also produce large primary tempe.racure differencials even enough adequace flow exiscs. A severe overcooling evenc, such as a stuck open turbine bypass valve or s:aam system relief valve, may resul: in primary :emperature differencials in excess of 100*F. This diagnostic method allows :ne over:onling : rend :s be easily distinguished from the loss of nacural circula: ion flow (loss of neat sink) i event by the response of bots the primary and secondary plots. Isolamentation l Implementacion of this analysis method involves several areas. An automa:ed data display is desirable to enable operators :o davoce full :ime co' analy-sis. However, manual ploccing capability should be ?.onsidered as a backup and for i:s use as a training metnod. Procedures should be consistent wi:n the diagnos tic metnod. For example, il an overcooling event is diagnosed, there should be procedural guidance to mitigate overcooling. Introdue:ory and proficiency : raining is required in both ths analysis metnod and tne use of related procedures for operatnrs and appropria:e staff members. - - _ _
This analysis method has been applied to post trip data recorded during normal and abnormal plant transients and to transient data generated by computer simulation. Operator training using this data base has shown that the method can be quickly learned and ef fectively applied. In the classroom environment operators are able to manually plot and evaluate crer.ds in real l time. Operators who analyzed plant transients witnout this meth.'d seldom completed the analysis in real time, were less confident of eneir conclu-sions and made more errors in diagnosis. Additional uses for this method include basic training in plant dynamic re s ponse. The historical trend aids in communicating the actual or simulated dynamic plant response and may be used to supplement tse narrative reports of events of interest to the industry. This analysis method has been developed for use following reactor trips but the princip.les may be applied to other power plant operating conditions. l An example of the method applied to a PWR with a U-tube steam generator is shown in Figure 12. The transient was a reactor trip test in which a turbine bypass valve malfunctioned and stuck open. The pressure-temperature plot shows an abnormal cooldown below the expected range in 1-1/2 Jinutes. The primary pressure drop was rapid but the steam generator response and the large saturation margin clearly indicate an overcooling problem and not a LOCA. Al th ou gh an extensive amount of data has not been plotted for this type of plant, this example demonstraces that the meched ~ has potential applicability. Summary A method for analyzing power plant performance in real time following a reactor trip has been developed. Using trends of primary camperatures and primity and secondary pressures on a pressure / temperature plot, plant response can be evaluated as either normal or abnormal with respect to one or more of the following categories: Loss of Coolant Accident, Loss of Heat Sink or Overcooling. This method facilitates distinguising LOCA from overcooling, highlignts asymmetric ef fects and diagnoses multiple casualties. ine analysis and diagnostic capability is applicaole during forced flow and natural circulation. Implemen-cation in an operating environment involves training, procedures and methods to provide plots. Other uses include communicating dynamic plant response to operations personnel. The potential exists to extend thit_ method to ocner plant conditions and to pressuriced water reactors wita U-tube steam generators. -5
i e l i 1 Figure 1 Pressure Temperature Plot I I 1 I 1 4 l i I 1 I I Limit Expected I e m '3 1 m 1 m 2 1 Q-1 0, sb. }/ \\64 N 7 /1 / / I / I .--------_-----l--._1_- I Limit / / I I /6 i / 6o I s' s Temperature
Figure 2 Expected Post Trip Conditions Expected Ranges of 2400 Primary Pressure and Hot Leg Temperature r/ 2200 L..) 2000 1800 3 [1600 0 1400 n. 1200 1000 ~1
- y Expected Ranges of 800 Steam Generator Pressure and Cold Leg Temperature 600 I
t I f 520 540 560 580 600 620 Temperature ( F)
Figure 3 Normal Post Trip Response i i I I 2400 i w l I I I 1i-1 2200 Ii 9: I 00 l !...6.3i I I I 2000 I s <(4 - & Yl 1 I 3 l l 1 1800 l l l l 6 f1600 ~~~J ~ ~]~~~~~ 1 I e 5 I I m 1400 1 i E I i E I I i 1200 i i l I w I 1000 gyg I o I o,/ I 800 le,/ F I 600 ---r--------------' I i i i 520 540 560 580 600 620 Temperature ( F)
j Figure 4 i Loss o'f Coolant Accident I 2200 r-; l _e en E 1600 e ws m .f40wc. 'l 1000 I I 550 600 Temperature ( F)
s l Figure 5 Loss of Heat Sink i I I I 1 2200
- ---j i
I I I I I I. l C i G I l E 1600 2 1 o I m m g 2 1 c I 1 l I I 800 / ~ 600 I f i 550 580 600 Temperature ( F)
Figure 6 Overcooling i I 1 I I i 2200 i; !!...j I i g l I 55 E 1600 e3mm i 2 I 1 I 1 I I i I 1000 _, ]: 800 1 I i l 600 I I f i 520 540 550 600 Temperature ( F)
Figure 7 Stuck Open Turbine Bypass Valve "A" Loo I l 2400 1 I I i 1 I i j-j l 2200 I! I _,o I o i l 2000 _A B l 15 8 l 4 1800 I I I i 1 2 B l [ 1600 :- j r]4 IA ~ 2 I I 3 i l g 1400 i 2 i n. 1 i 1 I 1200 I I I ol 4! }] ~ 1000 B Af -1o 2 15 3 / I / 800 a 15 / \\ r - l - - - - - - - - -- - - - - - - 600 41 3 i i i 520 540 560 580 600 620 Temperature ( F) n e
1 Figure 8 TMI-2 Accident 1 I 1 I 2400 i 1 o i i i I 2200 i r- ; i i s... :! i 2000 l l j 1 I I w I ,,/ 1800 l c 1 i l I o \\ T-----2<h---- ~ 2 1600 i 1 3 3 01 1 j 8 i 4s(! 10 8 E.1400 I s i E l I e e F i 15 l I 1200 1 a% li_1_o j zo I 2 1000 1.2 9.".) 1 1 3 9 1 ,/ 4 1 / 800 I / 8 l'/ t 1 600 ----i--------------- I 520 540 560 580 600 620 Pressure (PSIG) ~+ -, - --
Figure 9 Natural Circulation Expected Post Trip Ranges i I I i Expected Ranges of I 2400 I Primary Pressure and i I Hot Leg Temperature ll i 2200 I 3..-...-............-.- I i u. I I 2000 i I I I 1 I i 1800 l l i .___L___________ E 1600 1 I @d' 2 I S 1400 l 1 cE I i 1200 1 I I I I i I b 1000 I ~1 I L; ,e g i ' Expected Ranges of 800 i ,/ Steam Generator Pressure f, and Cold Leg Temperature 600 ---4--------------- l 1 f f f 520 540 560 580 600 620 i Temperature ( F)
Figure 10 Effects of Steam Generator Pressure to Cold Leg Temperature Delay 2 o il Corrected m 1 0 i Q-2, 2 ..3 i 4 Uncorrected s ' 'emperature Data gathered at time 0 is pressure at 0, temperature at -2 Plotting pressure at 0 against temperature at 2 shows direct effect of pressure on temperature. i 9
Figure 11 Overcooling During Natural Circulation i i I I 2400 4 i i i ,I l 1 I I I 2200
------ ------ --- -- i i
i O [ t.............. I 1 1 1 4 2000 Iu 3 1 l10 ~g 3 5 i 1 I I 1800 I I i/ 1 I C [ 1600 i I e I ws I = 1400 m I l 4 i a I i i 1200 1 I i i i 1 I i i i i 1000 q,, r 2 1 3 i 5 I e' i 800 7 1 I / _10 p' I 600 ----!--------------- I i i i 520 540 560 580 600 620 Temperature ( F) 1
Figure 12 Reactor Trip Stuck Open TBV U-Tube Steam Generator 2400 !....I 2200 o 2000 20 1 1s 1% 1800 2 m S: 1600 e 4 10 8 y1400 3 7 8 1200 s 1000 w 2o ~~ o 800 1s lh 7 2 8 600 5 3 t t 460 480 500 520 540 560 580 600 Temperature ( F) .}}