ML19254C721
| ML19254C721 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/12/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910170263 | |
| Download: ML19254C721 (29) | |
Text
TENNESSEE VALLEY AUTHORITY CH ATTANOOG A. TEN NESS E E 374n t 400 Chestnut Street Tower 11 October 12, 1979 Mr. Domenic B. Vassallo, Acting Director Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Ccmmission Washington, DC 20555
Dear Mr. Vassallo:
In the Matter of the Application of ) Docket No. 50-327 Tennessee Valley Authority )
Enclosed are TVA's responses to the Advisory Committee on Reactor Safeguards (ACRS) recommendations listed in their interim reports, Nos. 2 and 3, dated May 16, 1979. We were asked to respond to all ACRS concerns on the Three Mile Island 11 incident in S. A. Varga's letter to' H. G. Parris dated June 1, 1979. TVA previously responded to this request in my letter to you dated July 12, 1979.
Very truly yours,
\. Dvbd L. M. Mills, Manager Nuclear Regulation and Safety Enclosures (40) i 1
- N t 1157 001 7910170 el
~~~~~en A
RESPONSES TO ACRS CONCERNS IN INTERIM REPORT NO. 2 ON T!!I-2 1157 002
ACRS Statement--Natural Circulation--Procedures It is evident from the experience at T.'ll-2 that there was failure to establish natural circulation of water in the primary system and failure to recognize in a t.imely manner that natural circulation had not been achieved. The need for natural circulation under certain circumstances is common to all PWR's.
The Committee reconunends that procedures be developed .by all operators of PWR's for initiating natural circulation in a sa fe manner and for providing the operator with assurance that circulation has in fact been established. These procedures should take into account the behavior of the systems under a variety of abnormal conditions.
As a first step, the NRC Staff should initiate immediately a survey of operating procedures for achieving natural circt.lation, including the case when offsite power is lost. At the same time, the operators of all PWR plants should be requested to develop detailed analyses of the behavior of their plants following anticipated transients and small breaks in the primary system, with appropriate consideration of poten-tial abnormal conditions, operator errors and failures of equipment, power sources, or instrumentation. These analyses are necessary for the development of suitable cperating procedures. The re"lew and evaluation of these analyses by the NRC Staff should receive a priority consistent with the priority being given to changes in operating procedures.
l) f I Response P00 g ] g j-(
In order to prevent confusion, the definition of natural circulation must be established prior to discussing this phenomena. Natural circu-lation is a condition in the reactor coolant system (RCS) wherein the RCS fluid is predominantly single-phase water, no forced circulation of the water exists, but water density differences between the water in the reactor pressute vessel and the steam generators exist such that a driving head acrcss the core results. This definition is apparently consistent with the ACRS dafinition of natural circulation, but may no.
be consistent with the NRC definition.
The implications of Three Mile Island (TMI) plus the traditional single-failure licensing philosophy have been considered in our evaluation of natural circulation at Sequoyah.
Natural circulation is one of the important modes of decay heat removal during the course of an entire family of loss-of-coolant accidents (LOCA's) characterized as small break loss-of-coolant transients and other SAR Chapter 15.0 cvents. The other modes are heat removal through the break and steam condensation in the reflux boiling mode. Any btcak in the reactor coolant pressure boundary larger than 0.375 inch ID (0.008 sq ft) and smaller than 9.57 inches ID (0.50 sq ft) is catego-rized as a small break on a Westinghouse pressurized water reactor, The following discussion on natural circulation following a small break loss-of-coolant transient is based on the latest analyses parformed by Westinghouse in light of TMI. The base plant considered in these analyses is a four-loop RESSAR-3 plant.
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- - - . - - .. w
3-The break size in a small break loss-of-coolant transient is the deter-mining factor as to whether or not the stean generators are relied upon as a heat sink during the initial portion (approximate 1;. the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of the event. Westinghouse has shown that for breaks two incl.es ID (0.022 sq ft) and smaller, the steam generators are relied upon as a heat sink during the initial portion of the transient until the break flow is capable of removing decay heat. Gypically for a 1-inch ID line break, it would take approxitaately 2'. hours before the break flow can .cmove decay heat.) For breaks larger than 2 inches ID, the steam generators are not relied upon as a heat sink daring the initial portion of the transient because the breaF flow is large enough to remove decay heat very early in the event.
Westinghouse has concluded that natural circulation as cefined above will not be interrupted during small breaks larger than 3/8 :nch ID and smaller than 1 inch ID. Their bases for this conclusion are as follows:
- 1. The system will reach an equilibrium pressure whict corresponds to the pressure at which the liquid phase break flo'., equals the high head centrifugal pump injection rate.
- 2. This equilibrium pressure will be established below the steam generator safety valve setpoints for these break sizes.
- 3. The fluid in the reactor coolant system is saturated or subcooled liquid ' except in the core and hot legs, where small values of vc id fraction exist.
- 4. The steam generator tubes do not drain and the natt.ral circulation mode of decay heat removal will continue to function until the time that the break can remove all the decay heat.
- 5. No core uncovery is predicted to occur for breaks vf tnis size.
For breaks greater than 1 inch ID, Westinghouse has concluded that natural circulation, as defined above, will be interrupted. dowever, suf ficient heat removal capability by way of either a combination of break flow and steam condensation in the steam generators or break flow alone exists such that peak clad temperatures are limited to below 1800 F and the amount and duration of core uncovery are of little consequence to the outcome of the transient.
Westinghouse has concluded that decay heat removal via the steam geneca-tors during small break loss-of-coolant accidents with the P.CS in the natural circulation or steam condensation (reflux boiling) raodes of operation is not susceptible to interruption due .o ti.e int roductiori and/or presence of noncondensibles. Their bases for this conclusion are as follows:
- 1. There is not a large enough source of noncondensibles during any of the small breaks analyzed which has the potential tu bind up the U-tubes in the steam generators.
g ORGR 1157 004
2 .< The physical characteristics of the U-tube steam generators used in Westinghouse plants prevent them from being susceptible to noncon-densible binding; any steam and noncordensibles that enter the steam generator will pass through an area of the steaa generator that is surrounded by a substantial amount of water or. the secondary side, causing the steam to condense, and reducing the steam and noncondensible bubble size to the point that it cannot cause binding of the U-tubes in the steam generators.
- 3. Even if large amounts of noncondensibles were present in the reactor coolant system, Westinghouse had modeled, calculated, and concluded that any noncondensibles that enter the ste. .n genera tor U-tubes will be swept out due to the inherent differences between the water and noncondensible velocities Subsequentl;., buildup cf noncon-densibles in the high points of the reactor coolant system will be prevented.
However, TVA will continut to work with Westinghouse ta ensure that.
both organizations' understanding of natural circulation during small break loss-of-coolant accidents and other Chapter 15.0 cvents remains valid as the understanding and implications of IM1 evolve.
At the present time, TVA supports t.he Westinghouse conclusions based on TVA's current understanding of a Westinghouse PWR's response during small break loss-of-coolant accidents.
Instructions for initiation of natural circulation, including the case when offsite power is lost, are contained in a- existing eraergency operating procedure. These instructions dest he expected response of existing controi room instrur..entation used to verify that natural circulation has been established. In addition, these instructions describe the actions required to enhance natural circulation including those conditions during which saturation temperature and pressure conditions are reached in the primary system.
1157 005
ACRS Statement--Natural Circulation--Pressurizer lleaters The use of natural circulation for decay heat removal following an accident in a PU" normally requires the maintenance of a suitable over-pressure on t ictor coolant system in order to prevent the genera-tion of ster.. . ch can impede circulaticn. For many transients, maintenance of this overpressure is best accomplished by use of the pressurizer heaters.
Although the pressurizer ht aters at T2!I-2 continued to recei.ve power from offsite sources during the entire accident, the availability of of fsite power cannot be assured for all transients or accidents during which, or following which, natural circclat. ion must be established. The Conunittee recommends that the NRC Sta f f initiate immedir t ely a survey of all PWRs licensed for operation to determine whether the pressurizer heaters are now or can be supplied with power frca qualified onsite sources with suitable redundancy.
Response
Ther are fouc banks of pressurizer heaters.
1 automatic control group at 415 kW 3 backup groups at 485, 485, and 415 kh Pres urizer low level will trip all four banks of the heaters and prevent them from coming back on until level is recovered in the pressurizer.
All four heater banks will trip on a safety injection signal when in the normal mode. After safety injection reset and level recovery in the pressurizer, one backup heater bank would cperate automatically. The other two backup heater banks and the control bank would not come on automatically, but could be manually acticated. In the event of a loss of offsite power and safety injectior, two backup heater banks rated at 485 kW cach and powered from different trains of emergency power can be manually activated from the main control room 90 seconds af ter en.ergency power becomes available.
I157 006
ACRS Statement--Natural Circulation--Saturation Conditions The plant operators should be informed adequately at all times of those conditions in the reactor coolant system that might affect their capa-bility to place the system in the natural circulation mode or to sustain it in such a mode. Information indicating that coolant pressure is approaching the saturation pressure corresponding to the core exit temperature would be especially useful, since an impending loss of overecessure would signal to the operator a potential loss of natural circulation. This information can be derived from available prescurizer pressure and hot leg temperature measurements, in conjunction with conventional steam tables.
The Committee recommends that information for detecting an approach to saturation pressure be displayed to the operator in a suitable form at all times. Since there may be severa1 equally acceptable means of providing this information, C. _ _ is no need for the NRC Staff to assign a iagh priority to the development of prescriptive requirements for such displays. However, a reasonably early request that licensees and vendors consider and comment on the need for such a display wonid be appropriate.
Response
- 1. Presently, the Sequoyah process computer monitors four hot Icg temperatures (HLT's) and four pressurizer pressures (PP's) and obtains an average of each. The computer programs include steam table conversions. Also, the computer has trend recorders with dual pens.
- 2. TVA will add program (s) to calculate the saturation temperature corresponding to the measured pressurizer pressure { avg). We have the capability to trend the HLT (avg or any leg) on one pen and the calculated saturation temperature on the other pen. The degrees of subceoling can be observed as the difference between the two pens.
An alarm function would be added to indicate when the subcooling AT is abnormal. The operator could select the points for trend at that time. (The calculation would be performed evecy 64 seconds.)
- 3. TVA will also have steam tables and/or saturation curves available to the control room operator at all times.
.. I157 007
ACRS Statement--Core Exit Thermocouples The NRC Staff should request licensees and vendors to consider whether the core exit temperature measurements might he u*ilized, where avail-able, to provide additional indication regardint notural circulation or the status of the core. For the latter purpose . is recoimnended that the full temperature range of the core exit thermoccuples be utilized.
At TMI-2, the temperatures displayed and recorded did not include the full range of the thermocouples.
The Committee believes it would be appropriate for the NRC Staf f to request licensees and vendors to consider and comment on this recommen-dation. This request should be made as soon as convenient and the time allowed for responses should be such as not to degrade responses on higher priority matter > Plant changes that might result eventually from consideration of this recommendation would not, at this time, r em to require a high priority.
Response
- 1. Presently, the Sequoyah process computer monitors 65 incore CA (type K) thermocouples. They are now ranged from 0-700*F and calibrated for highest degree of accuracy between 400-700 F (13/8 percent). They should be within 12 F below 400 F.
- 2. TVA is in the process of changing the sof tware out-of-range index to 1800 F. Accuracy in the upper range will be considerably less than the 0-700 F range (120 F). The software change will be complete before Sequoyah unit I fuel loading.
I157 008L
_g.
ACRS Statement--Instrumentation to Follow the Course of an Accident The ability to follow and predict the course of an accident is essential for its mitigation and for the provision of credible and reliable predic-tions of potential offsite consequences. Instrumentation to follow the course of an accident in power reactors of all types has long been a concern of the ACRS, is the subject of Regulatory Guide 1.97 (which has not yet been implemented on an operating plant), and is the subject of an NRC Staif Task Action Plar. for the resolution of generic issues.
The Committee believes that the positions of Regulatory Guide 1.97 should be reviewed, and redefined as necessary, and that the Task Action Plan should be reexamined, as soon as manpower is available. The lessons learned from TMI-2 should be the bases for these reviews. For example, improved sampling procedures under accident conditions should be considered.
Although revie ad reexamination of existing criteria may take some time, the studies completed to date, together with the understanding gained fcom the accident at TM1-2, should provide sufficient basis for planned and appropriately phased actions. The Committee believes that the installation of improved instrumentation on operating reactors of all types should be underway within one year.
Response
- 1. The following post accident instrumentation is supplied to enable the operator to follow transients.
A. T hot or T cold (measured wide range)
B. Pressurizer water level C. RCS pressure (wide range)
D. Containment pressure E. Steam line pressure F. Steam generator water level (wide range)
G. Steam generator water level (narrow reage)
H. RWST water level I. Containment water level J. Pressurizer pressure K. Containment H2 monitors Each of the above channels is either recorded or logged.
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- 2. Containment Radioactivity Levels A. Airborne radioactivity levels in the primary containment during accident conditions can be indirectly obtained with the high range area monitor that is located outside the upper compartment personnel hatch. This monitor will remain on scale for contain-ment airborne radioactivity concentrations up to about 20 perct.t of those that could be experienced in a RG 1.4 Joss-of-coolant acci. 'nt.
There is no provision for direct measurement during accident conditions of exposure rates or nuclide radioactivity concen-trations in the primary containment. There are no radiation monitors inside the containment that have sufficient range and atmospheric qualification for the measurement of radia-tion levels in the containment during accident conditions corresponding to RG 1.4 assumptions.
Under normal conditions, real time detection of airborne parti-culate, iodine, and gross radioactivity concentrations is pro-vided by two 3-channel monitors per reactor unit. For these monitors, samples of containment air are pumped to the det ec-tion assemblies which are located in the auxiliary building.
After containment isolation, the isolation valves on the sample lines may be manually reopened from the main control room; however, this action cannot be taken until the contain-ment 'tmospheric conditions permit it since the monitors are not designed to operate with sample pressure, temperature, and humidity conditions that would exist during some accidents.
Even after sample pressure, temperature, and humidity condi-tions return to acceptable values, the monitor channels would be offscale for containment acitivity levels corresponding to RG 1.4 assumptions.
TVA will im ' all redundant radiation monitors outside of containment capable of monitoring airborne radiation inside containment corresponding to RG 1.4 assumptions. As soon as design details become available they will be submitted for NRC review.
B. Containment Air Sample Currently, there is no provision to take containment atmos-pheric samples for laboratory analysis during harsh contain-ment atmospheric pressure, tempe ra t ure , and humidity conditions.
During normal conditions, the monitors referenced in part (A) provide the following samples that can be analyzed in the laboratory: (1) particulate filter, (2) charcoal absorption cartridge, and (3) a gaseous sample. llowever, the sampling system for these nionitors is not qualified for operation whea containr. cia atoosplaric conditions correspond to RG 1.4 assump-tions. Furthermore, are such samples collectable with these monitor assemblies du. iag accident conditions, there is not sufficient radiation protection for personnel to remove the samples and analyze them in the laboratory.
1157 010
TVA will modify portions of the existing gaseous sampling system so that shielded samples of RG 1.4 containment atmos-phere can be taken in an accessible area. As soon as design details are available, they will be submitted ft; NRC review.
Under accident conditions, the hydrogen content of the contain-ment atmosphere is monitored with two analyzers located in the annulus between the containment and the shield building.
Remote indication is provided in the main control room. These analyzers are redundant safety grade and are on trained power.
C. Water Samples During normal operation,' reactor coolant samples, cooled with component cooling water, are available in the act sample room.
During accident conditions, the containment isolation valves on the sample lines can be opened and reactor coolant samples will again be available in the hot sample room. During normal reactor shutdown optrations, samples of the reactor coolant water being cooled ly the residual heat removal system (RHR) are taken from RHR pipes and routed to the-hot sample room.
During accident conditions, these samples, which are available in the hot sample room, would be samples of the sump water under the reactor vessel that is being recirculated. The radiation protection design ;or taking these samples ar analyzing them in the laboratory is based on operation with up to 1.0 percent failed fuel. The samples could not be taken and analyzed when sample specific activities are even a small fraction of those corresponding to RG 1.4 assumptions.
TVA will make provisions for campling water from the reactor coolant system (RCS) and the 'esidual heat removal system (RHR) for activities corresponding to RG 1.4 assumptions. The radiation monitor (s) will be placed on the RHR piping to monitor containment sump water activities correspond 1.g to RG 1.4 assumptions. As soon as design details are ava!)able, they will be submitted for NRC review.
r i157 011
ACRS Statement--Reactor Safety Research The ACRS recommends that safety research on the behavior of light-water reactors during anomalous transients be initiated as soon as possible and be assigned a high priority. The ACRS would expect to see plans and proposals within about three months, preliminary rest.lts within an additional six months, and more comprehensive results within a year.
Of particular interest would be the development of the capability to simulate a wide range of postulated transient or accident conditions, i luding various abnormal or low p obability mechanical failures,
- t. .rical failures, or human errors, ia order to gain increased insight into measures that can be taken to improve safety.
The new program of research to improve reactor safety has been initiated only recently, and then c . ty on a relatively small scale. The Committee reiterates its previous recommendations that this program be pursued and its expansion sought by the Commission with a greater sence of urgency.
Response
Not applicable.
k\ )
ACRS Statement--Status Monitoring Although the closed auxiliary feedwater system valves may not have contributed directly or significantly to the core damage or environmental releases at TMI-2, the potentially much more severe consequences of unavailability of engineered safety features in plants of any type is of concern and deserving of attention. Status monitoring not dependent chiefly on administrative control, and thus possibly less subject to human error, might help assure the availability of essential featur s.
A request should be made within the next few months that licensees consider additional status monitoring of various engineered safety features and their supporting services. The NRC Staff should begin studies on the advantages and disadvantages of such monitoring on about the same time scale. Responses from licensees should be expected in about one year, at which time the NRC Staff should be in a position to review and evalmate them.
Response
The status monitoring system automatically presents the operator in the main control room with a visual display and alc. indicating the status of any ECCS system which has been deliberately bypassed or deliberately made inoperable. This system meets the condition _ described in Section C of Regulatory Guide 1.47.
The visual display consists of a schematic flow diagram of the bypassed or inoperable system (s), the status of each component to which Section C of RG 1.47 is applicable is indicated on the face of a cathode ray tube.
In addition, a clock is provided indicating the time remaining before the system must be returned to normal or the unit shut down as required by technical specifications.
The SMS does nat currently .nonitar:
- 1. Solenoid valves for which the loss of power causes the valve to go to a safe position
- 2. Backpressure valves on the motor-driven pump discharges
- 3. Manual maintenance valves
- 4. Check valves
- 5. Auxiliary equipment and support systems TVA is proceeding to expand the Status Monitoring System capability for the Sequoyah Nuclear Plant. As soon as design details are available, they will be submitted for NBC review.
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RESPONSES TO ACRS CONCER"S IN INTERIM REPORT NO. 3 ON TMI-2 l'57 014
ACRS Statement--Reactor Pressure Vessel Level Indication The Committee believes that it would be prudent to consider expeditiously the provision of instrumentation that will provide an unambiguous indica-tion of the level of fluid in the reactor vessel. We suggest that licensees of all pressurized water reactors be requested to submit design proposals and schedules for accomplishing this action. This would assure the timely availability of reviewed designs if the Staff ongoing studies should indicate that early implementation is required.
The Committee believes that as a minimum, the level indication should range from the bottom of the hot leg piping to the reactor vessel flange area.
Response
To meet t!.e need for better information concerning the level of fluid in the reattor vessel, TVA will provide level measurement instrumentation for the Sequoyah Nuclear Plant. As soon as design details are available, they will be submitted for NRC review.
1157 015
ACRS Statement--Operator Training and Qualification The NRC Staff should examine operator qualifications, training, and licensing to determine what changes are needed. Consideration should be given to educational background, to training methods, and to content of the training program. Attention should also be given to testing methods, with specific concern for the ability of the testing methods to predict operator capability. Examination of licensing procedures should deter-mine whether they are responsive to new information that is developed about plcnt or operator performance. Effort should also be made to determine whether results of examinations can be correlated with operator ability. Requalification training and testing should be similarly examined to ensure that they take account of information that is developed by operation in the plant and to determine that relevant information about other plants is made available to operators and is made part of the training and requalification program. As part of thia and of other more extensive studies, continuing attention must be given to the amount of information which an operator can assimilate and use in normal and in emergency situations and to the best method of presenting the information to the operator. 'he use and limitations of simulators for operator training should receive careful consideration.
Response
TVA agrees that the Nuclear Regulatory Commission should examine operator qualification requirements, training, and licensing procedures. The NRC and the utilities should work together to determiae what changes are needed and to establish uniform performance and training requirement.
Educational background standards and the results of general aptitude tests are utilized by TVA in the selection of candidates for our student noerator training program. Although the admission standards increase ae probability of a student successfully completing the program, they do not guarantee the development of a better operator. This must be accomplished by establishing a rigorous and comprehensive program with many checks along the way. The TVA program meets these objectives. Our student operator training is 26 months in length and is divided into four sections or student levels. To progress from one level to the next, the student must pass comprehensive written and oral examinations.
The content of the student training program is outlined below:
Student I, Step 1
- 1. Mathematics
- 2. Chemistry
- 3. Physics
- 4. Thermal Hydraulics 1157 016
Student I, Step 2
- 1. Print Reading
- 2. Introduction to Secondary Plant Cycles and Components A. Valves B. Pumps C. Instrumentation
- 3. Secondary Plant Cycles Student II, Step 1
- 1. Physics (Electricity and Magnetia.a)
- 2. Plant Normal and Emergency Electrical Systems Student II, Step 2
- 1. Principle of Turbogenerator Operations
- 2. Turbine Construction
- 3. Turbine Control Systems
- 4. Generator Constructions
- 5. Generator Cooling and Excitations Systems Student III, Step 1
- 1. Nuclear Physics
- 2. Reactor Physics and Operations
- 3. Introduction to Primary Systems Student III, Step 2 (Conducted at Plant; Four Months)
- 1. In-depth study of all plant systems (Construction and Operation)
Student IV
- 1. On-the-job training. Assigned to shift for five months as student assistant operator.
Following completion of the 26-month training program, the stadent is reassigned as an assistant unit operator for a minimum of 14 months.
During this time the individual must complete a hot or cold licensing program and must pass an operator certification examination which is administered oq a simulator identical to the plant at which the assistant unit operator is assigned.
While written examinations are an acceptable method for determining the knowledge level of a stident, TVA believes that only through oral and performance examinations on a simulator can an individual's operating ability be properly evaluated.
The TVA requalification program consists of three weeks of retraining.
Two weeks of the ret raining arc conducted on the plant simulator at the Power Production Training Center end one week at the plant. During the simulator training, the operations are continuously evaluated and graded) s
-y.
on their performance during normal and emergency operations. The pro-gram is tailored for each plant to ensure that all plant modifications, new plant information, and new operating procedures and techniques are considered as well as other specific piant or utility problems iacluding Licensee Event Reports. The results of the Training Center evaluation are documented, and a report is forwarded to the licensee's plant. All unsatisfactory evaluations . e reviewed by the plant Training Review Board which consists of the itant superintendent, assistant cuperin-tendent, operations supervisor, and assistant operations supervisor. Based on this review, the Training Review Board establishes additional retraining requirements that must be satisfactorily completed before an operator can resume licensed shift activities. Af ter completion of retraining, an' examination approximately 6-8 hours in length is administered, by the plant, to all licensed operators. If an operator receives a score of less than 80 percent in any category or receives an overall score of less than 70 percent, the Training Review Board reviews the individual's training record and assigns additional retraining in all categotics in which a deficiency has been demonstrated. Removal from all licensed activities and placement in an accelerated retraining program is mandatory when an overall score of less than 70 percent is obtained. Before an operator resumes licensed activities, he must demonstrate proficiency in the areas in which he was previously judged deficient and must be approved by the Training Review Board after the Board has conducted a review of the operator's performance in the accelerated retraining program. In addition to the above, each licensed operator periodically receives a shift performance evaluation. This evaluation is based on tne operator's performance on shift during transient, startup, shutdown, emergency, and abnormal conditions. i157 0i8
ACRS Statement--Evaluation of Licensee Event Reports Because of the pote:itially valuable information contained in Licensee , Event Reports (LER's), the Committee recommends that the NRC Staff afh7 establish formal procedures for the use of this information in the trair.ing of supervisory and maintenance staffs and in the licensing and requalification of operating personnel at commercial nuclear power < # plants. The information in LER's may also be useful in anticipating sa ety problems. At the present time, some utilities routinely request that they be provided copies of all LER's applicable to plants of the type they operate or to specific systems and components in a given class af plants similar to their plant. Certain reactor vendors have made similar requests and use the LER's.to review and evaluate the performance of their plants. In addition, the NRC operator licensing staff has indicated that they use LER's in reviewing operating experience at commercial facilities. The large number of LER's that attribute the cause to personnel error would tend to indicate that a formalized program of LER review would be useful in the training, licensing, and requalification of nuclear power plant personnel. The extent to which such a program could be used to anticipate safety problems should also be considered.
Response
TVA's Nuclear Experience Review Panel presently reviews all Licensee Event Reports. When applicable, results of the review will be incorpo-rated in TVA's operator training and requalification programs. In addition, monthly training sessions are conducted for each shift crew. The material covered during these sessions include, but is not limited to, Licensee Event Reports, over tor ercers, recent equipment problems, changes to technical specifications, and general plant status. 1157 019
ACRS Statement--Operating Procedures Se'.ty aspects of individual reactors during normal operation and under accident conditions are reviewed in detail by the NRC Staff and discussed with the ACRS. Acceptable limits for normal operations are formalized by technical specifications, submitted by the licensee, and approved by the NRC Staff. Operating procedures for severe transients have received less detailed review by the NRC Staff. It appears that such procedures would benefit from review by an interdisciplinary team which includes personnel expert both in operations and in system behavior. Also, for the longer tenn, there may be merit in considering the development of more standardized formats for such procedures. Respcase Sequoyah operating instructions undergo an independent review by the Plant Operations Review C imittee. This Committee consists of the plant superintendent, assistant superintendent, maintenance supervisor, opera-tions supervisor, results supervisor, health physics supervisor, and quality assurance supervisor. In addition, TVA will develop and imple-ment internal procedures to ensure that all abnormal and emergency operating procedures are reviewed by the Division of Engineering Design. This prceess will ensure that abnormal and emergency operating instruc-tions undergo an interdisciplinary review by personnel expert in both operations and system behavior.
} } h)7 02.0-
ACRS State _.mnt--Reliability of Electric Power Supplies (System Design and Testing) During the past several years, there have been several operating expe-riences involving a loss of ac power to important engineered safeguards. The ACRS believes it important that a comprehensive reexamination be made by the NRC and the reactor licensees of the adequacy of design, testing, and maintenance of offsite and onsite ac and dc power supplies. In particular, failure modes and effects analyses should be made, if not already performed, more systematic testing of power system reliability, including abnormal or anomalous system transients, should be considered, and improved quality assurance and status monitoring of power supply systems should be sought.
Response
The Sequoyah Nuclear Plant is supplied with electrical power from two major systems, an offsite power system and an onsite power system. The offsite electrical powe; is supplied by two physically and electrically independent connections from the Sequoyah 161-kV switchyard to the onsite electrical distribution system. The 161-kV switchyard is the terminus for the second nuclear unit, the 500-kV intertie bank and nine 161-kV transmission lines. The onsite power system consists of an ac power system cnd a de power system. The onsite ac power system is a Class IE system which consists of: (1) the standby ac power system and (2) the 120-V vital ac system. The standby power system is identified as the diesel generators, the 6.9-kV shutdown boards, the 480-V shutdown boards, and all motor control centers supplied by the 480-V shutd sn boards for both units. The 120-V vital system is powered by four inverters per unit each power-ing a separate channel through a separate instrument power board. The vital 120-V dc control power system is a Class IE system composed of four redundant channels fed from the fomr vital batgpries. Prior to operation, functional and preoperational tests are conducted on each system. During the preoperational test phase, system is demonstrated during both normal and abnormal conditi8ns. performance The results of these tests are then reviewed by TVA's Division of Ehgineering Design. During plant operation, system testing is controlled by plant technical specification surveillance requirements. If any major maintenance is performed, post-maintenance tests and/or surveillance tests are per-formed to ensure the operability of the system. TVA is aware of the importance of the reliability of electric power supplies and periodically reviews the adequacy of both the offsite and onsite ac and de powar supplies. Additional monitoring of the systems during system transients is being considered as well as including the power systems on t he emergency core cooling status monitoring system.
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9 ACHS Statement--Analysis of Transients The ACRS recommends that each licensee and holder of a construction permit be asked to make a detailed evaluation of his current capability to with-stand station blackout (loss of offsite and onsite ac power) including additional croplicating factors that might be reasonably considered. The evaluation should include examination of natural circulation capability. the continuing availability of components needed for long-term cooling, and the potential for improvement in capability to survive extended station blackout. The ACRS also recommends that each licensee and construction permit holder should examine a wile range of anomalous transients and degraded accident conditions which might lead to water hammer. Methods of controlling or preventing such conditions should be evaluated, as should research to pro-vide a better basis for such evaluations. The Committee expects it would be appropriate to have such studies done genet cally first, for classes of reactor designs and system types.
Response
The Sequoyah auxiliary feedwater system includes a separate turbine-driven pump and redundant electric-driven feed pumps to provide motive power diversity for the system. The valves and controls and necessary support systems of the turbine-driven pump are powered by a IE de power source. In addition to this major plant feature, Westinghouse has <Sne some preliminary work on this issue. The time that Sequoyah will be able to survive total blackout is largely dependent upon the loss rate of primary coolant through the reactor coolant pump seals. Further study will be necessary to develop a complete picture of plant response to this event. To be most efficient, this work should be done on a generic basis. TVA agrees that systematic studies of operating and accident situations that might lead to water hammer could be valuable in improving the reliability and safety of power plants. TVA is willing to participate in generic studies of this phenomenon. 1157 J22 W
-lC-ACRS Statement--Emergency Planning An effort should be undertaken to plan and define the role NRC will play in emergencies and what their contribution and interaction will be with the licensee and other emergency plan participants including other gov-ernment agencies, industry representatives, and national laboratories.
Such planning should consider:
- 1. Assurance that formal documentation of plans, procedures, and organization are in place for action in an emergency.
- 2. Designation of a technical advisory team with names and alternates for the anticipated needs of an emergency situation.
- 3. Compilation of an inventory ot equipment and materials which may be needed for unusual conditions including its description, location, availability, and the organization which controls its release.
The Committee recommends that each licensee be asked to review and revise within about three months:
- 1. IIis bases for obtaining offsite advice and assistance in emergencies, from within and outside the company.
- 2. Current bases for notifying and providing information to authorities offsite in case of emergency.
This review and evaluation should be in terms of accidents having a broad range of consequences. The results of this review should be reported to the NRC.
Response
TVA has reviewed the bases for obtaining of fsite advice and assistance in emergencies. TVA's Radiological Emergency Plan (REP) identifies the bases and mechanisms fo. requesting assistance for emergencies requiring of fsite response. TVA also has reviewed its current bases for notifying and providing infor-mation to.offsite authorities. The TVA REP provides criteria and a flow-chort for notifying and providing information to the lead State agency, NRC, and DOE for all potential or accidental radiation releases. l) )7 .
ACRS Statement--Decontamination and Recovery The Committee wishes to call attention to the importance of a program designed to learn directly about the behavior, failure modes, surviv-ability, and other aspects of component and system behavior at TMI-2 as part of the long-term recovery process. This program should also examine the lessons learned at TMI-2 to determine if design changes are necessary to facilitate the decontamination and recovery of major nuclear power plant systems. Respons_e The long-term recovery ef fort at TM' 2 has oeen and will continue to be a valuable classroom for the industry. This experience should con-tinue to yield insights and data that will be aseful in the evaluation and design of plant systems for accident recovery.
ACRS Statement--Safety Review Procedures The TMI-2 accident has imposed large new pressures on the availability of manpower resources within the hPC Staff. If progress is to be expedited on the new questions which have arisen and on existing unresolved saft y issues, the ACRS believes that new mechanisms should be sought and imple-mented. For those safety concerns where such a mechanism is appropriate the Committee recommends that the Commission should request licensees to perform suitable studies on a tin.ely basis, including an evaluation of the pros and cons, and prcposals for possible implementation of safety improvements. The NRC Staff should concurrently establish its own capa-bility to evatuate such studies by arranging for support by its consul-tants and cuatractors. In this fashion, the Committee anticipates that the information on which judgments will be based can be d < eloped much more expeditiously, and an earlier resolution of many safety concerns may be achieved.
Response
Not applicable. 1157 025
ACRS Statement--Capability of the NRC Staff The Committee recommends that the copability of the NFC Staff to deal with basic and engineering problems in what may be termed broadly as reactor and fuel cycle chemistry be augmented expeditiously. This should include establishment of expertise within the NRC, with assis-tance arranged from consultants and contractors, in such important technical areas as the behavior of PWR and BWR coolants and other materials under radiation conditions; generation, handling, and dis-po,al of radiolytic or other hydrogen at nuclear factlities; perfor-maace of various chemical additives in containment sprays; processi , and disposal techniques for low- and high-level radioactive wastes; chemical operations in other parts of the nuclear fuel cycle; and in the chemical treatment operations involved in recovery, decontamination, or decommissioning of nuclear facilities. The Committee wishes to emphasize the importance of providing this expertise in both the research and licensing management elements of the NRC.
Response
Not applicable. 1157 026
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ACHS Statement--Single Failure Criterion The NRC should begin a study to determine if use of the single failure criterien establishes an appropriate level of reliability for reactor safety systems. Operating experience suggests that multiple 'ailures and common mode frilures are encountered with sufficient frequency that they need more specific consideration. This study should be accompained by concurrent consideration of how the licensing process can be modified to take account of a new set of criteria as appropriate.
Response
Not applicable. 1157 327
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ACRS Statement--Safety Research The ACRS believes that, as a result of the TMI-2 accident erious safety research areas will warrant initiation or much gre-,<. emphasis, as appropriate. The Committee suggests that consideration de give.i to an augmentation of the NRC safety research budget for FY 80. Also, the Committee believes that a larger part of the safety research p rog; a should be oriented toward exploratory research as contrasted to confirmatory research, with some degree of freedom from immediate licens-ing requirements. The ACRS plans to have a Subcommittee meeting on this subject with representatives of the N;u Office of Nuclear Regulatory Research in the near future. The Committee is continuing to review these matters and will report further as additional recommendations are developed. Rgsgense Not applicable.
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Additional Comments by Messrs. H. Lewis, D. Moeller, D. Okrent, and J. Hay The potential for a reduction in risk to the public in the case of a serious reactor accident Sy the implementation of a means for controlled, filtered venting of a containment which could retain particulates and the bulk of the iodine has been recognized for more than a decade. The concept was recommended for study more recently in the American Physical Society Report or, light-water reactor safety and in the Ford Foundation-Mitre Report, " Nuclear P.,wer - Issues and Choices." It is a high pri-ority item in the NRC plan submitted to Congress for Research to Improve the Safety of Light-Water Nuclear Power Plants (N UREG-0438 ) . The study performed for the State of California on underground siting concluded that filtered, vented containment was a favored option to explore in connection with possibic means to mitigate the consequences of serious reactor accidents. However, little progress has been made on the devel-opment of sufficiently detailed design information on which to evaluate the efficacy and other factors relevant to a decision on possible implementation of such consequence ameliorating systems. The TMI-2 accident suggests that the probability of a serious accident in which a filtered, vented containment could be useful is larger than many had anticipated. We recommend that the Commission request each power reactor licensee and construction permit holder to perform design studies of a system which adds the option of filtered venting or purging of containment in the event of a serious accident. The system should be capable of withstand-ing a steam and hydrogen environment and of removing and retaining for as long a time as necessary radioactive particulates and the great bulk of the iodine for accidents involving degraded situations up to and including core melt. Such studies could be done generically for several reactor-containment types, and should evaluate the practicality, pros and cons, the costs, and the potential for risk reduction. A period of about 12 months for a report to the NRC by licensees and construction permit holders appears to represent a possible schedule.
Response
TVA has been involved in a number of studies of advanced contaimnent concepts over the years. Although the vcnted and filtered concept has merit, it is not clear that this particular approach will always be the most efficient if additional containment capability is deter-mined to be necessary. TVA is willing to participate in industry studies to evaluate the need for enhanced containment capability and to consider design proposals.
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