ML20136B768
| ML20136B768 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/07/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7909110621 | |
| Download: ML20136B768 (40) | |
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O,~b TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 400 Chestnut Street Tower 11 September 7, 1979 Mr. Dominic B. Vassallo, Acting Director Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatocy Commission Washington, DC 20555
Dear Mr. Vassallo:
In the Matter of the Application of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 TVA has reviewed NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July 1979 and the memorandum from H. R. Denton to the NRC commissioners, " Resumption of Licensing Reviews for Nuclear Power Plants," August 20, 1979. We have evaluated each of the recommendations with respect to Sequoyah Nuclear Plant.
The enclosed copy of our evaluation is submitted for your review.
Except as noted in the enclosure, we believe that the schedule given in Enclo-sure 6 of H. R. Denton's memorandum is reasonable and hereby commit to implementation of the recommendations accordingly.
We believe it would be appropriate to meet with the NRC staff to discuss our submittal at your earliest convenience. We will coordinate arrange-ments for such a meeting with Carl Stable of your staff.
Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Ib ager Nuclear Regulation and Safety Enclosure s1 9909 f(a & Al Dy-An Equal Opportunity Employer 7 7 / 4 h >-
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ENCLOSURE 2.1.1 - Emergency Power Supplies NRC Position I
Consistent with satisfying the requirements of General Design Criteria 10,14,15,17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be bnplemented:
Pressurizer Heater Power Supply 1.
The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions.
The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.
2.
Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses.
If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
3.
The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
4.
Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators 1.
Motive and control components of the power-operated relief valves (PORVs) shall be capabic of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
2.
Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
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3.
Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety grade requirements.
4.
The pressurizer level indication instrument channels shall be powered f rom the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.
Response
TVA policy has been to provide for all PWR's qualified emergency power to PORV's, block valves, pressurizer level indicators, and some pressurizer heaters because we considered that these items were important to plant safety.
The SQN pressurizer heaters are powered and controlled from Class lE sources (see FSAR figures 8.3-10, 8.3-11, 8.3-12, and 8.3-13).
The motive and control power interfaces with the emergency buses are qualified in accordance with safety grade requirements.
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 (lc) would operate automatically.
The other two backup heater banks and the control bank would not come on automatically but are manually activated.
In the event of a loss of offsite power and safety injection, two back up heater banks rated at 485 KW each can be manually activated by hand switches in the main control room, 90 seconds after emergency power becomes available.
The power-operated relief valves (PORV) and their associated block valves and control components are classified as Class 1E and are supplied from the emergency onsite power supply if offsite power is lost. The relief valves and their associated block valves are powered f rom opposite power trains. All connections to the emergency power supply are through devices that are qualified in accordance with safety grade requirements. For a description of the PORV and block valves, see FSAR Sections 5.1 and 5.2.2 and figure 5.1-6.
The pressurizer level indication instrumentation power is taken from the vital power bus (see FSAR Section 7.5).
These buses are supplied from the emergency power source when offsite power is unavailable.
The level indication is safety grade and classified as Class lE.
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2.1.2 - Performance Testing for Relief and Safety Valves NRC Position Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents. The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
E c singic failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized.
Test pressures shall be the highest predicted by conventional safety analysis procedures.
Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping, and support as well as the valves themselves.
Response' TVA is actively pursuing a joint effort with other members of the utility industry which will develop requirements for a generic test facility and program for reactor coolant system relief and safety valve prototypical testing. his joint effort will identify expected valve operating conditions through analytical studios and through these bounding analyses develop performance specifications for the test facility.
TVA will submit to NRC a description of and schedule for the generic performance testing of these valves as soon as this is availabic.
Upon completion of sufficient analysis to identify the environmental conditions which may exist, TVA will provide associated control circuits, piping, and supports which are qualified for such an environment.
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2.1.3.a - Direct Indication of Valve Position NRC Position Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliabic valvo position detection device or a reliable indication of flow in the discharge pipe.
Response
The power operated relief valves have a reliable direct, stem-mounted position indication in the main control room.
TVA will provide main control room indication of.valvo position of the pressuriecr safety valves by use of downstream acoustic monitoring.
Differential pressure measurements, direct stem-mounted switchen or multiple thurmocouple monitoring in the downstream piping are also being considered possibic alternate means to provido positive safety valve position indication.
The monitoring system will not be availabic for SQN unit 1 by the requested implementation date since delivery cannot be made by that date. llowever, for this short period, safety valve position indication is provided in the following manner:
1.
Temperature is sensed downstreatm of the valves and displayed in the main control room including high temperature alarms.
2.
The pressurizer relief tank han temperature, pressure, and fluid level indication and alarms in the main control room.
3.
The pressurizer has high pressure alarms in the main control room.
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b 2.1.3.b - Instrumentation for Inadequate Core Cooling NRC ponition 1.
Licenseca shall develop procedures to be used by the operator to recognize inadequate core cooling with currently availabic instru-mentation.
The licensec shall provide a description of the existing instrumentation for the operators to use to recognize these conditions.
A detailed description of the analyscs needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement, " Analysis of Off-Normal Conditions.
Including Natural Circulations" In addition, each PWR shall install a primary coolant saturation motor to provide on-line indication of coolant saturation condition.
Operator instruction as to use of this meter shall include consid-eration that is not to be used exclusive of other related plant parameters.
2.
Licenneen shall provide a description of any additional instrumenta-tion or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguoua, cany-to-interpret indication of inadequate core cooling.
f A description of the functional design requirements for the system shall also be included. A description of the procedures to be used l
with the proposed equipment, the analysis used in developing these proceduren, and a schedulo for installing the equipment shall bc l
provided.
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Itenponno TVA will provide an indication in the main control room of the deviation from saturation conditions.
This saturation readout will utilize output l
f rom the plant cosputer to perform this function.
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The plant computers presently monitor reactor system hot leg temperatures and pressurizer pressure.
In addition, steam tabic conversion routines are a I
part of the computer noftware. Programs will be added to calculate saturation temperature correnponding to the measured pressurizer prensure.
l In the event any hot leg temperature measurement approaches the saturation l
temperaturu by a prodotermined amount, an alarm will occur in the control room. The operator will be abic to observe the saturation temperaturc l
and hot Icg temperature and compare the two by trending them on computer l
input / output (I/0) devices.
I TVA will provido inntrumentation to measure water icvel in the reactor vessel down to the bottom of the hotica piping.
This instrumentation will be designed and qualificd in accordance with safoty grado, Clann 1E, require-(
ments including redundancy and emergency power, i
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TVA will extend the range of incore thermocouples to give readout of fuel temperatures that could be expected if the core was partially uncovered.
TVA is developing procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation.
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2.1.4 Containment Isolation Provisions for PWR's and BWR's NRC Position All containment isolation system designs shall comply with the recommenda-tions o f SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.
All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be nonessential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the reevaluation to the NRC.
All non-essential systems shall be automatically isolated by the containment isolation signal.
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.
Reopening of contain-ment isolation valves shall require deliberate operator action.
Response
Sequoyah either complies or will comply with the NRC positions on the four portions of this item in the following manner:
1.
The Sequoyah containment isolation system is designed to operate in two stages:
Phase A and Phase B.
Phase A isolates all process lines except safety injection, containment spray, portions of component cooling water, and essential raw cooling water.
Phase B isolates all remaining process lines except safety injection, containment spray, and auxiliary feedwater.
The Sequoyah containment isolation design utilizes the concept of diversity of initiating signals.
Phase A isolation can be initiated manually and is initiated by auto-matic or manual safety injection (SI) actuation.
The SI signal is derived from (1) high steam line flow concident with low steam line pressure or low-low average reactor coolant average temperature, (2) high steam line differential pressure between loops, (3) low pressurizer pressure, or (4) high containment pressure.
Phase'B isolation can be initiated manually or by high high containment pressure.
In addition, isointion valves in the primary containment ventilation system actuate on manual initiation of Phase A, Phase B, or SI and automatically on SI or high radiation signals.
2.
TVA has undertaken a study to (a) examine cach system which penetrates the containment, (b) determino whether or not it is essential, (c) describe basis for this determination, (d) modify design if required, and (e) report results to NRC.
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3.
The Sequoyah Nuclear Plant design complies with NRC requirements on the automatic isolation of nonessential systems. Any changes necessary as a result of study in item 2 will be made.
4.
The Sequoyah Nuclear Plant design complies with the NRC's requirements by requiring manual actions on the controls of individual components should it be necessary to change their status af ter the containment isolation signal has been cleared.
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i-u 2.1.5.a - Dedicated Hydrogen Control Penetrations t-NRC Position i
Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment. isolation systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of
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Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow t,.
requirements of the recombiner or purge system.
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Response
This requirement is not applicable to Sequoyah.
i The Sequoyah design has a manually actuated ESF recombiner system l
inside containment which is redundant and fully qualified (see FSAR Section 6.2.5).
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C 2.1.5.b - Inerting BWR Containments ilRC Position It shall be required that the Vermont Yankee and Hatch 2 Mark I BWR containments be inerted in a manner similar to other operating BWR plants.
Inerting shall also be required for near tem OL licensing of Mark I and Mark II BWR's.
Response
Noe applicable to PWR's.
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2.1.5.c - Installing Hydrogen Recombiners in LWR'e NRC Position All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-tenn post-accident combustible gas control.
The procedures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering shielding requirements and personnel exposure limitations as demonstrated to be necessary in.the case of TMI-2.
Response
This requirement is not applicable to Sequoyah.
The Sequoyah design has a manually actuated ESF recombiner system inside containment which is redundant and fully qualified (see FSAR Section 6.2.5).
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i 2.1.6.a - Systems Integrity for High Radioactivity NRC Position Applicants and licensees shall immediately implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. -This program shall include the following:
1.
Immediate Leak Reduction a.
Unplement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
b.
Measure actual leakage rates with system in operation and report them to the NRC.
2.
Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels.
This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.
Response
TVA will investigate practical leakage reduction measures on systems which may contain radioactive fluids post-LOCA and will examine such systems as the residual heat removal (normal letdown path), containment spray and safety injection (recirculation mode), chemical volume and control, sampling, and waste disposal systems.
This examination will include a study of valve stem packing leakoffs, rotating seals on equipment, gasketed connections or joints, drain pipes to open connections, and building drainage systems.
TVA will identify all systems from which leakage may quantitatively be measured. Measured system leakages will be reported to the NRC.
TVA will identify systems that may be leak tested and will implement a periodic leak test program on these systems at a frequency consistent with ASME Section XI requirements.
Existing plant surveillance instructions will be modified to include the periodic inspection of pump seals when the ESF systems are tested.
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Item 2.1.6.b - Plant Shielding Review NRC Position
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With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.
Each licensee shall provide for adequate access to vital areas and protec-tion of safety equipment by design changes, increased permanent or tempo-rary shielding, post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.
Response
The Sequoyah design bases include the assumption of TID 14844 sources.
TVA plants are specifically designed to mitigate major design basis events with no access outside the MCR being required.
With this goal in mind, the plants were not specifically designed for any access outsi'e the main control room.
To specifically design for guaranteed access at anytime in most parts of the auxiliary building is not feasible.
However, the current designs may allow considerable capability for access for short times if the entry time into the area can be selectively chosen.
The current arrangements and shielding for normal operation will help minimize the impact from post-accident contained sources even though the shielding was not intended for that purpose.
In certain instances, TVA has provided some shielding for post-accident access. TVA will make design changes in shielding if evaluations identify feasible modifications which should significantly enhance desirable access. The guidelines for the evaluations are given below.
TVA will assume a TID 14844 radioactivity release in the reactor containment.
The leakage from the containment into the auxiliary building will then be calculated using the Technical Specification containment leakage rate value and assuming the DBA LOCA pressure surge described partially in FSAR Figure 6.2-29 as the driving force for the leakage. The leakage paths from the containment into the auxiliary building will be located and each leakage path will be assigned a source term.
The airborn activity levels in the various parts of the auxiliary building will then be calculated using these source terms and the airflow patterns within the building.
Credit for holdup and dilution will be taken.
It will then be shown qualitatively that once the containment driving force for the leakage drops to a negligible value, no accident resulting in a RG 1.4 radioactivity release will continue to produce airborne radioactivity in the auxiliary building.
n TVA will also calculate the containment sump water leakage and the resulting airborne radioactivity introduced into the auxiliary building during the recirculating mode of the accident recovery period.
This analysis will assume the leak reduction measures provided to satisfy the. requirements of section 2.1.6.a of NUREG-0578 are implemented.
Credit will be taken for holdup and dilution.
A summation of the airborne radioactivity levels from containment leakage and sump water leakage from process systems in the auxiliary building will be made.
The next step will be to calculate the source terms for the sump water recirculating piping, pumps, and valves installed in the auxiliary building. TVA will then identify the vital areas in the auxiliary building which may need to be entered for servicing during an accident recovery period. The shic1 ding in these vital areas will be reevaluated to assess its effectiveness in such a circumstance.
The occupancy time limits, taking into consideration transit time, airborne radioactivity levels and gamma shine intensities, will then be calculated for the vital auxiliary building areas.
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s 2.1.7.a - Auto Initiation of Auxiliary Feedwater NRC Position a.
l Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be imple-mented in the short term:
1.
The design shall provide for the automatic initiation of the auxiliary feedwater system.
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The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.
3.
Testability of the initiating signals and circuits shall be a feature of the design.
4.
The initiating signals and circuits shall be powered from the emergency buses.
I 5.
Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
4 6.
The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.
7.
The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of mauual capability to initiate the AFWS from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety grade requirements.
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Response
i The auxiliary feedwater system is automatically initiated by redundant, coincident logic to preclude loss of function due to a single failure and to provide on line testability. The auxiliary feedwater system and initiating logic are described in TVA's response to NRC-0IE Bulletin 74-06A and in Sequoyah FSAR Section 10.4.7.2.
The auxiliary feedwater control circuitry _ including the automatic initiating circuitry is safety grade, Class 1E, and is powered from a power source connected to the emergency l
power system.
Each auxiliary feedwater pump has manual initiation capa-
- bility independent of the automatic initiation. The ac motor-driven pumps and valves are included in the automatic alignment of the loads to the i
emergency power system.
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2.1.7.b - Auxiliary Feedwater Flow Indication NRC Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented.
1.
Safety grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
2.
The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
Response
Auxiliary feedwater flow is indicated in the main control room for each of the four steam generators. The flow indication has not been classed as safety grade; however, it utilizes the same type of tran:a. utters which are used in other safety grade circuits. The transmitters are mounted on two separate seismically qualified panels and powered from power sources connected to the emergency power system. The cables are in low level signal trays and are kept separate from all power cables.
In addition, the total flow from the turbine driven auxiliary feedwater pump is indicated in the main control room.
The auxiliary feedwater flow instrument channels are powered from the emergency buses consistent with the diversity requirements of the auxiliary feedwater system.
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2.1.8.a - Sampling System NRC Position A' design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of per,onnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 10 3/4 Rems to the whole body or extremeties, respectively.
Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.
If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A design and operational review of the radiological spectrum analysis facili-ties shall be performed to determine the capability to promptly quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radioisotopes that are indicators of the degree of core damsge. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and non-volatile isoto,es (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.
If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Procedures shall be pro-vided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term).
Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift.
Response
A design and operational review of the reactor coolant and containment atmosphere sampling systems and analysis facilities has been performed.
To enhance the capability at Sequoyah for post-LOCA sampling TVA will:
1.
Make provisions for sampling water from the reactor coolant system and the residual heat removal system for the degraded accident condition.
2.
Install new lines with connections to the existing gaseous radiation sampling system for use in sampling the containment atmosphere for the degraded accident conditions.
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3.
Route sample lines to a shielded sampling station in an access-ible area and provide for taking samples which could be removed offsite for analysis.
TVA will also identify the type and nature of onsite analysis required by NRC.
If practical, TVA will procure the required analysis equipment, locate, design, and build an onsite analysis facility.
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2.1.~8.b Increased Range of Radiation Monitors NRC Position The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a l
revision to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident," which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term.
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Noble gas effluent monitors shall be installed with an extended range designed La function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.
Noble gas effluent monitors with an upper range capacity a.
of 105 pCi/cc (Xe-133) are considered to be practical and 1
should be. installed in all operating plants.
i b.
Noble gas effluent monitoring shall be provided for the total range of concentration extending from a minimum of 10~7 pCi/cc (Xe-133) to a maximum of 105 pCi/cc (Xe-133).
Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors shall overlap by a factor of ten.
2.
Since iodine gaseous effluent monitors for the accident condi-tion are not considered to be practical at this time, capability I
for effluent monitoring of radioiodin'es for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
t 3.
In-containment radiation level monitors with a maximum range l
of 10s rad /hr shall be installed. A minimum o'f two such monitors that are physically separated shall be provided.
Monitors shall be designed and qualified to function in an accident environment.
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Response
2 Redundant safety grade high range noble gas effluent monitors will be i
provided at Sequoyah.
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A method or methods of sampling effluent particulates and iodine will be l
chosen and redundant particulate and iodine effluent monitoring systems qualified to the present state-of-the-art will be installed.
l The present SQN design has one high range radiation monitor outside the i
containment in the auxiliary building, opposite the personnel hatch to detect high levels of radioactivity inside the containment. However, its range is not as high as required by the NRC.
Redundant radiation monitors
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will be provided outside the annulus to meet the NRC's high-range require-ment.
These monitors will be safety grade and will be designed and' qualified to function in an accident environment.
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- 2.1.8.c - Improved In-Plant Iodine Instrumentation NRC' Position 4
Eadh licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions.'
Response
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Sequoyah has portable low-volume air samplers, each equipped with a
-particulate filter followed by a charcoal adsorber to collect iodine isotopes. The particulate. filter will be counted in the health physics laboratory for gross activity and the charcoal adsorber sent to the radiochemical laboratory for a gamma isotopic analysis for radioactive iodines.
If necessary, as necessitated by a high-gross activity, the particulate filter will also be sent to the radiochemical laboratory for an isotopic analysis.
The primary difference in obtaining in-plant l
airborne isotopic concentrations for accident and routine operating conditions is the time' required for sampling. A shorter sample time could be necessary for accident conditions because of the presence of high isotopic J,
concentrations.
The plant has proce'dures for sampling and analysis of inplant air spaces incorporated in the Health Physics Laboratory Instruction Manual and the Radiation Control Instruction Manual.
Plant health physics technicians.are required to complete a formal training program plus receive inplant training which includes the use of health physics procedures and instrumentation.
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2.1.9 Transient and Accident Analyses NRC Position Analyses, procedures, and training addressing the following are required:
1.
Small break loss-of-coolant accidents; 2.
Inadequate core cooling; and 3.
Transients and accidents.
Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force.
These should be completed.
In addition, pretest calculations of fome of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long-term verification of compliance with Appendix K of 10 CFR Part 50.
In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:
1.
Low reactor coolant system inventory (two examples will be required--LOCA with forced flow, LOCA without forced flow).
2.
Loss of natural circulation (due to loss of heat sink).
These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included.
Each case should then be repeated taking credit for correct operator action. These additional cases will provide the basis for developing appropriate emergency procedures.
These calcula-tions should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b in this appendix).
The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR.
The analyses shall include a single active failure for each system called upon to function for a particular event.
Consequential failures shall also be considered.
Failures of the operators to perform required control manipulations shall be given consideration for permutations of the analyses. Gperator actions that could cause the complete loss of function of a safety system shall also be considered. At present, these analyses need not address passive failures or multiple system failures in the short term.
In the recent analysis of small break LOCAs, complete loss of auxiliary feedwater was considered. The complete loss of auxiliary feedwater may be added to the failures being considered in the analysis of transients and accidents if it is concluded that more is needed in operator training
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beyond the short-term actions to upgrade auxiliary feedwater system reliability. Similarly, in the long term, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-term analyses.
The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations could be used to provide important quantitative information not availble from an event tree.
For example, failure to initiate high pressure injection could lead to core uncovery for some transients, and a computer calculation could provide information on the amount of time available for corrective action.
Reactor simulators may provide some information in defining the event trees and would be useful in studying the informa-tion available to the operators. The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considera-tions such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.
The information derived from the preceding analyses shall be included in the plant emergency procedures and operator training.
It is expected that analyses performed by the MASS vendors be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant.
In addition to the analyses performed by the reactor vendors, analyses of selected transients should be performed bJ the NRC Office of Research, using the best available computer codes, to provide the basis for compari-sons with the analytical methods being used by the reactor vendors.
These comparison, together with comparisons to data, including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures.
Response
TVA is pursuing the required analyses and the development of new proced-ures and training guidelines with other utilities through the Westinghouse TMI Owners Group.
The transient and accident analyses should use realistic codes and include event tree analyses. The analyses should consider permutations and combinations of operator errors and equipment failures, including single failures in multiple systems and multiple operator errors. The operating procedures and operator training that will evolve from these analyses are essential to enhancing safety by improving reactor operator performance during transient and accident conditions.
Small break loss-of-coolant accident analyses have been performed and submitted to NRC in WCAP 9600.
The report presents a comprehensive study of Westinghouse system response to small breaks. Westinghouse has already discussed continuing efforts aimed at improving emergency operating procedure guidelines with the NRC.
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Inadequate core cooling is an item where further definition of the scope, such as system failure and operator error assumptions,. is needed from the NRC. At present model preparation is in progress to permit response to identified action. Westinghouse does plan to perform pre-test calculations of the LOFT tests when we are provided with the necessary input information.
The purpose of this action is to improve the performance of reactor operators during transient and accident conditions.
The primary concern is that the operator training and emergency operating procedures are based on the conservative plant FSAR Chapter 15 analyses.
Chapter 15 should continue to be used for design basis analyses since these show the most limiting initial approach to safety limits. What is needed is to evaluate the longterm consequehces of accidents using realistic assumptions incorporating the effects of the following:
1.
Operator's failure to act when required.
2.
Operator's inappropriate actions during an accident.
3.
Additional failures.
4.
Selected system operations (e.g., restarting of RCP's etc.)
Appropriate changes can then be incorporated into the existing procedures, j
designs, and training programs.
Development of the models to incorporate such effects is in itself a i
longterm effort before detailed analyses can be run.
Significant inter-action between industry and the NRC is required to agree on the assump-tions, bases, appropriate actions or misactions to be modeled, and best estimate boundary conditions.
Based on TVA's perception of NRC intent, the proposed implementation schedule in NUREG 0578 is extremely ambitious. We believe that it cannot be met without an extraordinary effort on the part of NSSS vendors, utilities, and the NRC staff.
While we agree with the urgency attached to this effort, we caution that undue haste, just to meet the implemen-tation schedule, is unwarranted, d
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2.2.1.a Sh i f t. %..>nerv i no c ' n Rt.;porm '.i 1 i t ic-MP.C Position 1.
The highest level at cor porate manages;.cnt o f ei.cb lam.isee ahall issue and periodically reiraue a management directive that etcphanizes the primary management reslonsibility of the shif t sancrviscr for safe operation of the plant under all conditions on his chift and that clearly establishes his command dut.ies.
2.
Plant procedures shall be reviewed to ent.ure that. the duties, responsibilitics, and authority of the shift supervisor and control room operators are properly defined to ef fect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel.
Particular emphasis shall be placed on the following:
a.
The responsibility and authorit y of the shif t supervisor shall be to maintain the broadest perspecLive of operational conditions a f f ecting t.he safety of the plant. as a matt er of highest priority at all times when on duty in the control room.
The idea shall be reinforced that the shif t sup.n vinor sheuld not become totally involved in any single operation in tinea of craergency. hen maltiplo veratiot.,.. re : e.,. ri o in the control room.
b.
The shift sulctvisor, unt :1 propt rly - 110., ;4 ail
.m:
in the control room at al} ti' tic.-
la r ing v.c ide;. s..tuat.cnn to direct the act.ivitim of control o,or. rp.oators.
h room authorized to reliove tbe r.hi f t c a;'et '. i sc s.a!i ;.
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If the shift sup..rvise r ir tenpot a 21; a r ' t ou. i h con'
..t room during routine oper.a t ions, lead cc.Fr.i room operator a
shall he desiqnated *o asnume the contro. ccon.. nam.au: f unc ti c...
These temporary duttes, responnihi 1 i t ter., und autherisy cha11 be cicarly specified.
3.
Tr aining programs for shift supervisor.. shall empt. asian and ceinfot.;c the respousibility for safe operation and the managtment tunct ion the shif t supervisor is t.o provide for ensaring saiety.
4 The administrativa duties of the shift supervisor shall be reviewcJ by t he s enior of ficer of each ati1ity resp.. ;ible for plata opera-tions. Adninistrative functions that detract from or are sut.ordinute t.o the management responsibility for ensuring the safe op.rration cf t.he plant shall be delegated to other operations personnel not on duty in the control room.
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s TVA Eer:ponse 1.
TVA's administ rative procedures, uhi! L supervi:;or jols de scr ipt.ioas,
and training progratas emphasize the primary management responsibility of the shift supervisor.
In ad:litior., pet: iodic retraining acts to reinforce his command responsibilities. While these existing neasures provide a high level of confidence that the shift supervisor has
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primary management responsibility for safe operation of the plant, TVA will periodically issue a management directive which emphasi::cs this assignment of responsibility.
2a.
Plant administrative procedures have been reviewed to casare that.
they cicarly define the authority and responsibilities of each position on shift. The dutice and responsibilities of the shift supervisor, as specified in the job description, are consistent with position statement 2a.
2b.
Th( shift crew in TVA plants consista of the following:
(1) 1 shif t engineer who has an 910 license aad who has overall r'escon-sibility for the plant vnen hijher level "in-l irw" inanagercent personnel are not present, (2) an..ns wt. ant shift engineer (als, has an SRO license) for each unit who ba; st'y'rvicory renponsibi1ts for all nonaal, abnor.na t,
..ud caerotr.cy ac t ivit. i...
o,.
hin w.:ign' d u n i t.,
(3) a unit operater (with an RO license) to. c..ch ;r.i t, ar. -
( i) other perronnel as appropriate. The dutics c: tha shi.ft supervisor as dincessed in EUPIG 057e c.re performe
- hy LV.w:a s at shif t.
engineer on.ach unit.
t'01 purro.ses of our :.'ercone3 n
will use the term aauistant 1;hift eng meer for shi.'t nu r:,.: rv i nor.
The assistant shift engineer's tormal work ntation is in the control room, but he perio.lically makes inspections of giant equipment. :le will be in the control roou during planned transient.u in addition, he will immediately go to the control room during emergency situations.
lie remains in the control room at all times during accident situations to direct the activities of the unit operator unless formally relieved of this function by the shift engineer.
The shift engineer may, in turn, be formally relieved by the assistant operations supervisor or the operations supervisor, (both also hold an SRO license).
bc.
In the event that the asuistaat shift engineer (chif r. supervisor) is absent, the unit operator will be the lead operator an the unir.
to which he is asnigned.
Por ;.:ultip;c unit plants, an additional licensed ogerator will be available in the control conple;- to act as an assistant to the unit oper4L.,r in abnormal or c.nergency s i t.ua ti on s.
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The shif t engineer and assistant shift eng'incers will receive such training.
4.
The administrative duties of the shift supervisor will be reviewed by the senior officer of TVA responsible for plant operations.
Administrative functions that detract from or are subordinate to ensuring safe operation of the plant will be assigned to other employees.
The following actions have already been taken:
1.
A clerk has been assigned to the shift engineer's office on each shift to perform administrative details formerly done by the shift engineer.
2.
Part of the routine "non-management" duties of the assistant shift engineer have been assigned to other employees.
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2.2.1.b Shi f t Technical Advinor flRC Position faci
- cnn."
- .h.e l l provide an on-shi f t technical advisor t o the shift sur svisor.
'The shi f t ti chnical advisor m,y serve more than one unit at a 'au l t i nnii L s a t e ii quaIified to perform the advisor function f or 1 hf! t.iri s'is nni t i I!e shiit lerhois.il aicisor sh111 llave a bachelor'.4 dey ee or equivalent b
in a stien: i f i-or enniocerine discipline and have rect ived f.pecific t ra i n i ng, se the r espoteie and arta ly-i t o f' the plant for transients an.1 are itent :
1he shift technical addisor shall also ren ive tre ining in p. ant.tm. i gn arni layout, inclinling 1.in: capabilities of instrumentat. inn and coet rol:. in t he cont rol roone.
The licensee shall assign narmal dutirs to the 8hifL technical advisors t. hat pertain to the engineering aspects of ensuring safe operations of the plant., including t he revies-
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and cualu.: lion of operat ing experience.
TVA Response TVA will provide an on-shiit technical advisor to the shift supervisor to support the diagnosis of of f-normal events anal to advise t.he shif t
- .itpervisor of act ions Lo terminate or mitigate the consequences of such events.
The shift technical advisor will have the following spia t i f tcations:
(1) bachelor's degree or equivalent in a scienti fic or engineering disciiline, (2) cxtensive training in plant transient and accident i
response, (3) technical specification t raining with ea.phasis on the basis for l imi ting coneli t ions for operat ion, (4) si,cnificant reactor ope ra to r-rel a t ed t ra i ni ne. on sy:; ten.< and opera t i ng p mcedures, aad
(')) w i l l m." L TVA's plant-spec i fic. pia l i f i cation reqn i rement s foi sta. ion reactor,ny,in crs.
The dut.irr of the shi f t technical advisor will incin.lc (1) cont ro! ra ci'-
s a p po r i. in t he d i.q,tios i s o f o f f -na t a.a t events, i.2) adv i ce t o the shi:t suprrviser on a t t i o n:. to t emi na t<
or mitigate the sei.ncquenc.s of of;'-
no r saa ' < rent s,
() advi ce t o the uhiit supervisor on allowable conditions f r., i nope r.do t e egoipu nt anil c<pii pmcet out of nervice, (4) advice to the shift supervisor as requested on unscheduled maintenance of equipment and p9s t -m.ii n t en..nc <- i. e n t i n ;,, cinal (5) d i,nen.i nat ion of i n formation oa opera t i ng expe ience t i 18..
other plants incin lia:4 IJR*S-f
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TVA helieves it i :, not possible to meet the NRC requirement that the shift technical advisor be placed on sbi! L ley Jasesary 1980 ar.it trainc.1 by
. January 1981 because t ra ining and r.hi f t coverage cannut be accomplished at the same time.
Furtheimore, TVA doc.s not believe it is appropriate to place the shi f t Leciniital.advi::or on shi t t until all t r.i in t ne, r equi ren.ent;.
a re w> t. T51A there f oce propo:.e.s te select engineers.nal initiate training and to place them on shift by January 1981 after all training requirements have been satisfied.
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2. 2. l..c Shiit an1 Neliel Turnoves l' roced u re s
.NR.C Peset'on
'i h e l i cen:.ee. soa!! review anel res isc as necessar y the plant procedure for shilt and r ti"I tu rn<>ver t o cou.ce t he f ol lowing:
1.
A checklist sha!! he provided for the oncomir.g and offgoing control room <,p-ra t o r: and the oncoming t:hift supervisor to complete and sign.
The following items, ar. a ninimum, shill be included in the checklist:
a.
As s u ra nce that critical plant parameters are within allowable limits (parameters and allowable limi.ts shall be listed on the checklist).
b.
Assurance ot' the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist).
Identificat ion of systems and components that are in a degraded mode c.
of operation permitt ed by the Technical Specif icat ions.
For such systems and components, tiie length of tiir.' in tue degra, led mode shall be compared with the Technical Speci fications action st atement (this shall be recorded as a separate entry on trie check. list).
2.
Chorl< lists or logs shall be provided fr.r complet son by the offgoine and oncoming auxilia ry operat ors and t echnicians.
Such checklists or logs shalI inclueic any e<ps ipment under maintenance of test that by t hem *. elves coulet de +.; ca ele a system criticai to tha provant i..n and mi tipt ion c '
operati 2nal t ransient ope s a t. i on. ! ' rans i ent ; and arci.leius or initiate (t.ha t to che: k.:nd cr i t eria for.creptable status shall be included on ih.
rhestiit.t);
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estabIish"d to evaia.,te tha c;Iettivene.S of 1he shi f t and t"liet turnoves p ror r.ha r<-
( f o r e x a.rp i e, periodic i ndepe nden'.
veriiicalion of sy:.t om a 1 ignment:. i.
- 1VA Reguynse TVA will develop and impicment shift and relief turnover procedures that wil l provide a
- . ;nrance that t he oncoming shi f t possesses adequate I:nowled,te of crit.ical plant status inf.umation and system availability.
A checklist or similar ha r d copy will be completed by of fs;oing and encoming sh f t s at each shifL turnover.
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This checklist will include critical plant parameters am! allowable limits, availability anel proper.ilignment of sa fety systems, and a listing of sa fety syst em component s in a degraded ino<!c along with the lengt h of t inie in that mode. This checklist will be signed by the of t going unit ope ra to r and the oncoming assistant shi f t supervisor and unit ope ra t o r.
All shift personnel responsible for the status of critical equipsinnt wii!
l' ave reliel checklists for oncoming and offgoing shifts that will inclnde any core cooling equipmcHL under maintenance or test that "onld degra.ie a safety system.
In addiLion, system will be establisheti to evaluate the ef f ectiveness esf a
I he ti.: nover procedu res.
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2.2.2.a Cont rol Room Accc:.s NRC po_s i t_. i on_
The licensee shall make provisions f or limiting acce..s to the control room to those i nd i vichia l s respons ible for the direct operat ion of the nuel. ar power plint (e.g., operatinnn e npr rvi:.o r. shift supervisot, and control ro nn operat ors), La tech.. i ca l advisers who n-ay be rey:.eated os required to r.a ppo r t the oper at inn, and to predesignated NRC personnel.
Provisions r.ha l l tortude the following:
1.
Develop tind implement an.uhnin i s t ra ti ve procedure that establishes the authority anil responsibility of t he person in charge of the cont rol room to I in. i t a ccer.s.
2.
Develop arul implement procedures that establi:.h clear line ot'
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authori'.y and responsibility in t he cont.rol room in the event of an emergency. The line of succession for the person in charge of the control room shall he established and limited to persuas possessing a current senior reactor operator's license. The plan shall cleacly define the lines of communication and authority for plant management personnel not in di rect conunand of operat. ions, including those who report to stat ions out side of the control room.
TVA Response TVA will develop and implement plant speci fic administ r.stive procedures that er.tablish specific inniividual anthority and responsibilit.y as well as delineate varions system nr equipment innetions related to controlling personnel access during normal and accident conditiona. A control room access plan will he developed to provide direction to all members of the plant st af f to ensure t hat those personr. responsible for.;a f e o;3 erat ion of the lilant are able t o perf orm el feet ively.
l fo addition, TVA wili develop and i..iplement proi eden rs t hat establ tsh r ricar line or gulhorit y aw! responsibility in the cont rel roon, in the event of an emergency.
Th.se proce hares will clea rly d. fine the lines of comunnication arnt nit hu s t y f or plant..nanan.m at permanel and will cusure t hat t he shift :.upe s v i:.u, h i s a t. : i t a n t, o r :.cn i..r
- i ceau.ed...ana gement personnel aie tie-only piant pe rsor.ne l wh. ha v.-
the aothoritf to di.ect a
4 license I act ive t ice of l i rcoser' reactor operators.
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.)n. i t,o,Techniial %ippo t;t, C' nt e r
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l..it:h n;ici aliar, nisclea r power plant rh il l ma tntain an on.;ite tes nni ca l support center sep.irite t rom.ind in s lor.c proAlmily to the control room that has t he-ca}sabality to elisplay and transmit plant : stat.us to those individuals. s.ho are hnowledgeable of and responsible for engincering and managem<nt ;npport on reactm operat ions in the ovent of an accidont.
The center shall be hahitable ta the 8.nce degree as the i ont.rol reom for postulated accident conditions. The licensee shall revice.his emergency plans as necessary to incorporate the role and location of the technical support retiter.
A complete set. of as-bnilt drawings and other records, as described in ANSI ;J45.2.9-l')74, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions.
These documents shal1 include, but. not. be limited to, general arrangement drawings, P&lDs, piping system isometrics, electrical schemat.ies, and photographs of component.s installed without layout specifications (e.g.,
lield-rnn piping and instrument tubing).
TVA Response The onsite technical support center will be established on the same floor as t he m: in control room (flCR) hut outside of th.* HCR.
I t wil l be bahi t.ab ! c to the same extent as the MCR and will have ready accesn t.o a complet e set of as-btill drasings.
l'eliable con.monicat ion.s will be provioed to the ilCR.
The technical support center will he established before receipt of the operating license.
I'rion t.e lanua.y I, !'iS I, caps ipmen'. wi l I bc :esLalled in the saptwr* c.nten t., i aq i..va the i.lant me,nitoring o pability of terbnical support perra.uncl.
Ihr p l..n t *<ait i o l o; i ue l Eu.c rgency Pl a n wi ' l be a~ien. led t.. es t.iol ish t he t ecimis a s s u pp..r t
?nt er and cpec i t e the personnol who will sta f f it i;. the e.. st t 9t a:t emerge ry.
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2.2.2.c Gys i: e (Jpe r a t ional, Support, C. qt e r N!iC Position An area to be designateil as t he onsi te operationa l 3.uppot L cen'et snail be established.
It shall be separate from the contro! ronia and shall be the place to which the operations support personnel will report in an en.cegency s i t u a t.i un.
Conmninications with the control roora shalI be provided.
The ernergency plan shall be revised to reflect the e::ictence of the center and to estabt ish the methods and Iines of comruinication and manage:avat.
TVA Response An opes ational support center with co:imittnications to the main control room will be est.ablished.
The plant Itadiological Liacrgency Plan w ill be awmded to establir,h thir, cc. iter and to specif y the personnel wl'o vill report to Litis center ist the event of an emergeticy.
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2.2.3 1:evired himiting, Conditions for Operation of Nuclear pow"r plants 11ased t!pon Saiety System Availat'iiity NRC position All NRC nuclear jumer plant l icensee: sh.ill provide inform; tion to define limiting up cational rondition based on a threshold of comp:cte loss of a
s e fet y finn t ion.
f.fenii1iritinn el a hnman or operat innal error tha' prevents or ronid prevent t he.s n.oi.igil s sini.ent ot a safety function required by Nhc iegnlat iona an.1 an.siyxcel in 1De Iicense applieaLion she.!I require platement of Lin plant in a hot shutda n condition within S hours aint in a co l ti shut down condi tioa wi t hin 20 fu,urs.
The l o.-a.
o f ope. ability..t a
sa fety f *nict ic.n sir:11 include consideration o f the nocer.n ry i ns t rument<:t ion, controls, emergency electrirai power
- .ou r ces, cool i ny. o r sea I wa t er, I uh ri ca t ion, opera t i ng procedu ren,
maintenance procedures, test procedures anni opt.rator interface with the system, which rwst a i r.o lic rapable of performing theic auxiliary or cupport inn iun< t ions.
Toe Iimiting conditiens for operation chall define the minimum safety functions for modes I,
2, 3, 4, and 5'of operation.
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The limiting corolitions of operatien shall require the following:
1.
If the plant is critical, restore the safety innetion (if possible) and place the plant in a hot shntdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
2.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, bring the plant to cold shutdown.
3.
Determine the cause of the loss of operability of the safety Innction.
Organizational accountability f or the loss of operabilit.y of the safety system shall he established.
4.
Determine...rrective actions and measures to prevent recurrence of the specifir loss of operability for the part icnlar sa fety funct ion r.nd generally for any sa fet y. t unction.
's.
Report the event ei thin 24 honrs by telephone and confirr.: hy telegraph.
masigr.ua, or lacsituito t i ansm i:.s i on to t he Di rector of th.' Regiona l Office, or his designee.
6.
prepare an.i dri i v. r a ;p.u i a l th port to the NRC'- Director of Nec; ear 19 i. t o r k re l a t i nn an.t i. o the Oi.i< tor of the appropriate regional offire of *he Office.: I n.< p, e t i.,n and CuIorrement.
fhe repo rt shafI cc.u t a ' n t he rosn i t s r.
stepn n.1 4
. h.. u. elong m th s basis for s
ailooing f.he pl.sai t.,
return t.+ power operistion.
The senior corporate e.verntive of the l i c em ee responsible and acconnL.bl a for sa fe plant ni v ra tina sha l l del i ver.nul dit.russ the roi. tent s of the report in a public meation with the ottice ut Nuclea r l'eactor Kcgulation and the Oflice of inspect ion and Enforcement at a location t.o be c!.osen by the 1)i rector of Nuclea r Reactor Regulation.
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7.
A I i nd in,t o f adequacy of the licensee's Special Report 1y the Director et Nuclear Reactor Kenulation will be reipiired be fore the Iicensee returns the pl. int to power.
IVA Hesponse in t he event a i,i m i t i n y, Condition f or Operal ion and associated ACTION s t ;s t eno n t. c.tu.ol be s21istied, technical specifical. ions require
- f. ha t the pl snt tio in at ! cant hot.st acolby wi th i n I honr and in at least cc,ld rhntdo.-n within.ta bours unics. co t rec t i ve me.uin re-are completed ihat pe rm i t ope ra t ion un*'e r t he per mi s :ibl e A.: TION st.at ement.
In a.bli t ion, technical :.;v c i i i ca l t dn s require that the event:, 1isted belw. cvents wh:ch result in LCO's not being met, he rt. ported within 24 hour:. by telephoin and c on firmed by t elegraph, mailgram, or f a c s i tai l e t ranmni.ss i on to Llie Direttoi of t.he Regior.a i Of fice no lat er than the first working day following the event, with a writt.en followup report.
wi t hin 14 days.
Technical spec i fica t. ions further require that correct.ive act ions and mea:.ures to prevent. recurrence of the event be reported to the NRC.
Failure or mal function of one or more component.s which prevents os a.
could prevent, by itself, the fulfillment of the functional requirements of systevi(s) used to cope wit.h accidents analyzed in the SAR.
h.
Personnel error or procedural inadequacy which prevents or could prevent, by itself, the IutiiiIment of the functiona! requi remen t s of systems required to cope with accidents analyze <1 in the SAR.
In the event a Limiting Condition for Operation and associated ACTION statement cannot be satisfied, TVA will notify the NRC within the required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
At that time, the NRC may, if deemed necessary, preclude further operation of the unit until TVA has completed any action considered neces-sary by the NRC.
TVA believes this current process fulfills the intention of this NRC position.
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. Position (1) - Containment Pressure (2) - Hydrogen Monitoring (3) Water Level Monitoring
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NRC Position Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain containment conditions during the course of an accident, the following requirements shall be implemented.
(1) A continuous indication.of containment pressu e shall be provided in the control room.
Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus five psig for all containments.
(2) A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room.
Measurement capability shall be provided over the range of 0 to 10 percent hydrogen concentration under both positive and negative ambient pressure.
(3) A continuous indication of containment water level shall be provided in the control room for all plants.
A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump.
Also for PWRs, a wide range instrument shall be provided and cover the range from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity.
For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool.
The containment pressure, hydrogen concentration and wide range contain-ment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy,
'and testibility. The narrow range containment water level measurement instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.
Response
1.
Four qualified, continuous indications of the containment pressure are provided in the main control room.
The 5 psig negative pressure requirement is not applicable to Sequoyah since qualified vacuum telief of the containment maintains the pressure at greater than negative 0.5 psig. The negative range of the existing pressure indicators envelopes this negative 0.5 psig limit.
Redundant, continuous containment pressure indication with a range up to four times the design pressure of the steel containment will be provided.
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2.
Redundant, sa fetygrade hydrogen analyzers are located in the annulus between the containment and shield building. These monitors provide continuous indication in the main control room within a few minutes of being remotemanually actuated in the main control room.
The range of these monitors is from 0 to 10 percent hydrogen concentration f om negative 2 psig to positive 50 psig pressure.
3.
The floor of the reactor building serves as the sump for the contain-ment. It is instrumented with four separate, qualified, and continuous level instruments which indicate in the main control room.
The range of the instruments is from less than six inches above the floor up to 20 feet above the floor.
If 500,000 gallons of water were introduced into containment in addition to the fluid volume of the reactor coolant system, safety injection accumulators, and a total ice melt, the containment water level would not exceed the design basis level.
This level is well within the 20 ft. range of the level instruments. A small sump suction procket (about 120 cubic feet) in the reactor building floor serves as a collector for the recirculation piping exiting the containment and does not require qualified level instrumentation.
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. Reactor Coolant System Venting NRC Position 10 CFR Part 50.46 requires that af ter any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the Additionally, Criterion 35 of 10 CFR Part 50 Appendix A requires core.
that a system to provide abundant emergency core cooling shall be provided.
The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and, (2) metal-water reaction is limited to negligible amounts.
During the TMI-2 accident, a condition of low water level in the reactor vessel and inadequate core cooling existed and was not rectified for a long period of time.
The resultant high core temperature produced a metal-water reaction with the subsequent production of significant amounts of hydrogen.
The collection of noncondensable gases Lmpaired natural circulation cooling capacility. Additionally, the collection of noncondensables gases impaired natural circulation cooling capability.
Additionally, the collection of noncondensables gases limited reactor coolant pump operational capability because of coolant voids in the system occupied by the gases.
Even when reactor coolant pump operation was possible, the installed plant venting system was capable of removing the noncondensable gases only through an extremely slow process.
The purpose of this recommendation is to provide reactor coolant system and reactor vessel head high point vents remotely operated from the control room for the purpose of removing noncondensable gases collected in the system in order to allow satisfactory long-term core cooling.
Response
TVA will provide the capability to vent the reactor vessel head in addition to the existing venting capability from the pressurizer.
The new reactor vessel head vent system will meet all of the NRC requirements.
It is, of course, not feasible to directly vent the reactor coolant system high points in the U-tubes of the steam generators.
This venting capability is not required.