ML19260A244

Forwards Revised Responses to NUREG-0578, Lessons Learned Requirements.

ML19260A244
Person / Time
Site:	Sequoyah
Issue date:	10/31/1979
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Vassallo D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7911080302
Download: ML19260A244 (105)

Text

. ... ' I -

TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 374o) 400 Chestnut Street Tower II A

v October 31, 1979 Mr. Domenic Vassallo, Acting Director Division of Proj ect Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 -

Dear Mr. Vassallo:

In the Matter of the Application of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 Enclosed are ten copies of TVA's revised responses to hTREG 0578, Lessons Learned Requirements, for the Sequoyah Nuclear Plant. Our responses fully address the NRC's letter (To All Operating Nuclear Power Plants) from Harold Denton. Mr. Denton's letter provided additional guidance on les-sons learned short term requirements.

For your convenience, the responses are presented in binders with dividers separating each item. Each item includcs the NRC position statement with the associated TVA response.

Very truly yours, TENNESSEE VALLEY AUTHORITY

,j Ab L. M. Iaills, Manage'r Nuclear Regulation and Safety Enclosure (10')

p 1296 196 7911 080 3,01 g

An Equal Opportunity Employer

EMERGENCY POWER SUPPLY (2.1.1)

Pressurizer Heaters POSITION Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17 and 20 of Appenaix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:

Pressurizer Heater Power Supply

1. The pressurizer heater power supply design shall provide the capaDility to supply, frca cither the offsite power source or the emergency power source (wher offsite power is not availaole),

a predetermined number of pressurizer heaters and associated controls necessary to establish ared maintain natural circula-tion at hct standoy conditions. The required heaters and their controls shalI be connected to the emergency buses in a manner that will provice redundant power supply capability. -

2. Procedures ar.d training shall be established to.make the operator aware of when and how the required pi s5surizer heaters shall be connected to the emergency buses. I? required,.the procedures shall identify under what conditions selectea emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressuri' zer heaters.

3. Tne time required to accomplish the connection of the preselectec pressurizer heater to the emergency ouses shall be consistent with the timely initiation and maintenance of natural circulation conditions.

4. Pressurizer heater motive and control powee interfacet with the emergency buses shall be accompliued through devices that have been qualified in accordance with safety-giade requirements.

CLARIFICATION

1. In orcer not to compromise independence between the sources of emergency ' power and still provide redundant capability to provide emergency power to the pressurizer heaters, eacn recundant heater or group of heaters should have access to only one Class lE division power supply.

1296 197

a

2. The number of heaters required to have access to each emergency '

power source is that number required to maintain natural circula-tion in the hot stancby condition.

3. The power sources need not necessarily have the capacity to provide power to the heaters concurrent with the loads required for LOCA.

4. Any cnange-over of the heaters from nomal offsite power to emergency onsite power is to be accomplished manually in the control room.

5. In establisning procedures to manually reload the pressurizer heaters onto the emergency power sources, careful consideration must be given to:

a. Which ISF loads may be appropriately shed for a given situtation.

b. Reset of the Safety Injection Actuation Signal to permit the operation of tne heaters.

c. Instrumentation and criteria for operator use to prevent overload-ing a diesel generator,

b. The Llass IE interfaces for main power ar.d control power are to be protected by safety-grade circuit breakers. (See also Reg. Guide 1.75)

7. Being non-Class IE loads, the pressurizer heaters must be automatically shed from the emergency power sources upon the occurrence of a safety inaection actuation signol . (See item 5.b. above) 1296 198

EMERGENCY POWER SUPPLY (2.1.1)

Pressurizer Heaters SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

TVA policy on pressurizer heaters is consistent with the NRC position. The ability of the Sequoyah Nuclear Plant design to meet each of the NRC recom-mendations is addressed in the following response.

RESPONSE

The SQN pressurizer heaters are powered and controlled from Class 1E 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 w hh safety-grade requirements. All four heater banks will trip on a Safety Injection signal when in the normal mode. After safety injection reset and 1cvel 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 signal, two backup 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 required operator actions are specified in the Sequoyah Emergency Operating Instructions (E01 5).

CLARIFICATION ITEMS

1. As specified in the above response, the Sequoyah design provides redundant capability for providing emergency power to each bank of heaters. The independence of the Class IE division power supply for each heater bank is shown by the following load group designation.

Power Train Heater Bank 1A-A 1A-A 1B-B Ic (automatic) 2A-A 2A-A 2B-B 2c

2. Emergency power is availatle to heaters required for riaintaining natural circulation in a hot standby condition.

1296 199

i ,

EMERGENCY POWER SUPPLY Pressurizer Heaters (cont)

CLARIFICATION ITES (continued)

3. The Sequoyah design provides for the required power for each bank of pressurizer heaters.

4. As specified in the above 64sponse, this cability is available at Sequoyah.

5. The existing Sequoyah Emergency Operating Instructio.is (E01) acount for the considerations specified in this clarification item.

i

6. The above response specifies that Class 1E interf aces for main power and control a power are protected by safety-grade circuit breakers.

7. The pressurizer heaters are automatically shed from the emergency power sources upon the occurrence of a safety injection actuation signal (SIS). SIS reset is covered in the Sequoyah E01.

1296 200

a Emergency Power Supply (2.1.1)

Pressurizer Level and Relief Block Valves POSITION .

Consistent with satisfying the requirements of General Design Criteria 10,14,1:s 17 and 20 of Aprendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:

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 capaDie 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 wii;h the PORV block valves shall be capable of being supplied from aither the offsite power source or the emergency power source when the offsite power is not available.

3. Motive and control power connections to the emergency ouses for the PORVs and their associated block valves shall be through devict.s that have Deen qualified in accordance with safety-grade requirements.

4. The pressurizer level indication instrument cnannels shall De powereo from the vital instrument buses. The buses shall nave tne capability of being suppied from either the offsite power source or the emergency power source when offsite power is not available.

CLARIFICATION

1. While the prevalent consideration from TMI Lessons Learned is being aDie to close the PORV/ block valves, the design shoula retain, to the extent practical, the capability to open these valves.

2. The motive and control power for the block valve shoulo De supplieo from an emergency power bus different from that which supplies the PCRV.

3. Any changover of the PORV and block valve motive and control power from the normal 7ffsite power to the emergency offsite power is to be accomplished manually in the control' room.

3_ 1296 201

4. For those designs where instrument air is needed for operation, appropriate electrical power must also be applied via the emergency power sources.

4 6

1296 202

Emergency Power Suoply (2.1.1)

Pressurizer Level anc Relief Block Valves SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

The design f or Sequoyah Nuclear Plant is consistent with the NRC positions concerning power supply for pressurizer relief and block valves and pressurizer level indication.

RESPONSE

The power-operated relief valves (PORV) and their associated block v4. ves and control components are classified as Class 1E and are supplied from the emergency onsite power supply if offsite pcwer is lost. The relief valves and their associated block valves are powered from 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.

CLARIFICATION ITEMS

1. Since the Sequoyah design meets NRC recommendations, no changes are anticipated and therefore, the capability to open PORV/ block valves will not be affected.

2. The redundancy built into the Sequoyah systems provides for independent energency power for PORV and block valves.

3. Any changeover of the PORV and bicek valve motive and control power from the normal offsite power to emergency power is accomplished manually in the control room as specified in the Sequoyah Emergency Operating Instructions.

4. Where instrument air is needed for operation, any associated electrical power required can be supplied by emergency power sources.

_,_ 1296 203

PERFORitAhCE RESTING FOR BWR AND PWR RELIEF AND SAFETY VALVES (2.1.2)

POSITION Pressurized Water Reactor anc Boiling Water Reactor licensees and applicants snall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and acC idents.

CLARIFICATION

1. Expected operating conditions can be determined through the use of analysis of accidents and anticipated operational cccurrences referenced in Regulatory Guide 1.70.

2. ibis testing is intended to demonstrate valve operability under various flow conditions, that is, the ability of the valve to open and shut under the various flow conditions should be demonstrated.

3. Not all valves on all plants are required to be tested. The valve testing may be conducted on a pr9totypical basis.

4. The effect of piping on valve operability should be included in the test conditions. Not every piping configuratlan is required to be ,ested, but the configurations that are tested should produce the appropriate feedback effects as seen by the relief or safety valve.

5. Test data should include data that would permit an evaluation of discharge piping and' supports if thnse components are not tested directly.

1296 204

6. A description of the test program and the scneaule for testing shoulo be submittea by January 1,1980.

7. Testing shall be complete by July 1,1981.

T296 'c05

PERFORMANCE TESTING FOR B'aR AND PkR RELIEF AND SAFETY VALVES (2.1.2)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

TVA,s commitment to a program for testing of relief and safety valves is presented below. This program is consistent with the NRC position.

RESPONSE

TVA is actively pursuing a joint ef fort with other members of the utility industry which will develop requirements for a generic test facility and program f or reactor coolant system relief and saf ety valve prototypical testing. This joint effort will identify expected valve operating conditions through analytical studies 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 available.

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.

CLARIFICATION ITEMS

1. See paragraph 1 cf the abo'a response.

2. Testing will demonstrate volve operability under various flow conditions.

3. See paragraph 1 of the above response.

4. Test conditions will include the effect of piping on valve operability.

5. The test results will provide data that would permit an evaluation of discharge piping and supports for those components not tested directly.

6. See paragraph 2 of the above response.

7. The testing is expected to be complete by July 1,1981.

l-

} 2h b

E DIRECT INDICATION OF POWER-OPERATED RELIEF a

VALVE AND SAFETY VALVE P6;' TION FOR PWRs AND BWRs (2.1.3a)_

POSITION Reactor System relief and safety valves shall be provided with a positive indication in the control room derived from a relidale valve position detection device or a reliable indication of flow in the discharge pipe.

CLAR'FICATION

1. The basic requirement is to provide the operator with unambiguous indication of valve position (open or closed) so that appropriate operator actions can be taken.

2. The valve position should be indicated in the control rcom. An alarm should be provided in conjunction with this indication.

3. The valve position indication may be safety grade. If the position indication is not safety grade, a reliable single channel direct indica-tion powered frum a vital instrument bus may be provided if backup methods of determining valve position are available and are discussed in the emergency procedures as an aid to operator diagnosis and action.

4. The valve position indication should be seismically qualified consistent with the component or system to which it is attached.

If the seismic qualification requirements cannot be met feasibly by January 1,1980, a justification should be provided for less than seismic qualification and a schedule snould be suomitted for upgrade to the required seismic qualificiation.

1296 207

5. The position inoication should be qualified, for its appropriate envir , , 'n t (ac-y transient or accident which would cause the relief or safety valve to lif t). If the environmental qualification program for this position indication will not be completed by January 1, IWO, a proposed schedule for completion of the environmental qualification program should be provided.

1 1296 208

DIREC',(NDICA7IONOFPOWER-OPERATEDRELIEF VnLVE.AND SAFETY VALVC POSITION FOR PLRS AI.'J BWRs (2.1.3a)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Position indication in the main centrol rova ror power operateu relief valves is currently available at Sequoyah. TVA,will provide main control room indicati<

of valve position of the pressurizer safety valves as specified in the followin' response.

RESPONSE

The power operated relief valves have a reliable direct, stem-mounted position indication in the main control room. Valve position of the pressurizer safety valves is ,,crently 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 has temperature, pressure, and fluid level indication rnd alarms in the main control room.

3. The pressurizer has high pres.ure alarms in the main control roca.

An environmentally qualified acoustic monitoring system for tne three safety relief valves on each unit will be provided. An accelerometer will be mounted on the valve discharge line just downstream of each valve. The accelerometer signals will go to a charge converter incide containment which will be mounted in a NEXA-4 enclosure. A valve flow indicator module will be located in the main contiel room. The flow indicator module will give positive indication of the fully open and .Q closed position of each valve. An alarm in the main control room will indicate when any valve is not in the fully closed position.

1296 209 DIRECT INDICATION OF POWER-OPERATEE VALVE AND SAFETY VALVE POSITION FOR PWRs AND BWRs (2.1.3a)

CLARIFICATION ITEMS

1. This design provides the operator with unambiguouc indication of valve position as specified in the above response.

2. Valve position is indicated in the main control room and alarmed as discussed in the above response.

3,4,5. Valve position indication for Sequoyah Nuclear Plant will meet seismic and environmental qualification requirements as specified for Sequoyah.

i10

INSTRUMENTATION FOR DETECTION OF INADE0VATE CORE C0 CLING (2.1.3.b)

SUBC00 LING METER POSITION Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation. The licensee shall provide a cescription of the existing instrumentation for the operators to use to recognize these conditions. A detailed description of the analyses 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 Circulation " (see Section 2.1.9 of NUREG-0578)

In aadition, eacn PWR shall install a primary coolant saturation meter to provide on-line indicatiun of coolant saturation condition. Operator instruction as to use of this meter shall incluce consideration that is not to be used exclusive of other relatec plant parameters.

CLARIFICATION

1. The analysis and procedures addressed in paragraph one abcve will revieweo ano should be submitted to the NRC " Bulletins and Orders Task Force" for review.

2. The purpose of the subcooling meter is to provide a continuous indication of margin to saturated conditions. This is on important diagnostic tool for the reactor operators.

3. Redunaant safety grade temperature input from each hot leg (or use of multiple core exit in T/C.'s) are required.

4. Redundant-safety grace system pressure measures should be provided.

5. Continuous display of the primary coolant saturation conditions should be provided.

1296 211

6. Each PWR should have: (A.) Safety grade calculational devices and a

display (minimum of two meters) or (B.) a highly reliable single channel environmentally qualified, ano testable system plus a oackup procedure for use of steam tables. If the plant computer is to oe used, its availability must be documented.

7. In the long term, the instrumentation qualifications must be required to be upgraded to meet the requirements of Regulatory Guide 1.97 (Instrumentation for Light Water Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident) which is under development.

8. In all cases appropriate steps (electrical, isolation, etc.) must De taken to assure that the addition of the subcooling meter ~ does not aaversely impact the reactor protection or engineered safety features systems.

e

)29b

INFORMATION REQUIRED ON THE SUBC00ING FETER Display Infomation Displayed (T-Tsat, Tsat, Press, etc.)

Display Type (Analog, Digital, CRT)

Continuous or on Demand Single or Redundant uisplay Location of Display Alarms (include setpoints)

Overall uncertainty (*F, PSI)

Range of Display Qualifications (seismic, environmental, IEEE279) ,

Calculator Type (process computer, deaicated digital or analog calc.)

If process computer is used,specify availability. (f. of time)

Single or redundant calculators Selection Logic (highest T., lowest press)

Qualifications (seismic, environmental, IEEE279)

Calculational Technique (Steam Tables, Functional Fit, ranges)

Input Temperature (RTD's or T/C's)

Temperature (number of sensors and locations) dange of temperature sensors j]g6 })3

Uncertainty

• of temperature sensors (*F at le-)

Qualifications (seismic, es.vironnental, IEEE279)

Pressure (specify instrument used)

Pressure ~(number of senssrs and locations) ,

Ronye of Pressure sensors uncertainty

• of pressure sensors (PSI at Ir-)

Qualifications (seismic, environmental, IEEE279)

Backup Capaoility Availability of Temp & Press Availability of Steam Tables etc.

Training of operators P rocecures

• Uncertainties must address conditions of forced flow and natural circulation I296 214 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

SUBC00 LING METER SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

TVA will provide continuous monitoring of the deviation from saturation conditions. The plant computer will be used to perform this function.

Procedures are being developed which will be used by the operator to recognize inadequate ccre cooling with currently available instrumentation.

Operator instruction for primary coolant saturation indication will emphasize the need to use related plant parameters.

RESPONSE

TVA will provide continuous monitoring of the deviation from saturation conditions. This saturation readout will utilize output. The plant computer will be used to perform this function.

The plant computers presently monitor reactor system hot leg temperatures and pressuri-er pressure. In addition, steam table conversion routines are a part of the computer software. Programs will be added to calculate saturation temperature corresponding to the measured pressurizer pressure.

In the event any hot leg temperature measurement approaches the saturation temperature by a predetermined amount, an alarm will occur in the control room. The operator will be able to observe the saturation pressure and system pressure and compare the two by trending them on computer output recorder in the main control room.

TVA is developing procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation.

CLARIFICATION ITEMS

1. The guidelines for procedures specified in the above response are being developed by the Westinghouse Owners' Group in response to the Bulletins and Orders task force. TVA will provide plant procedures based on these guidelines.

2. TVA recognizes that a continuous monitoring of margin to saturation conditions will be provided.

3. Redundant safety grade temperature input from each hot leg and/or multiple core exit thermocouples are provided for measurement of saturation conditions.

4. Redundant safety grade system measurement is provided at Sequoyah.

) 9b

INSTRITMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

SUBC06 LING METER (continued)

CLARIFICATION ITEMS (cont. )

5. Continuous monitoring of the primary coolant saturation conditions will oe provided as specified in the above response.

6. The margin to saturation can be continuously displaced on a trend recorder in the main control room. The backup trend recorder is available. Saturation curves are provided in the main control room and procedures will require the use of these curves on the loss of indication of margin to saturation. We expect the computer availability to exceed 99 percent.

7. Required instrumentation will be upgraded to the applicable requirements of Regulatory Guide 1.97 when they are fully developed. We consider the present design we are pursuing to be adequate.

8. Changes to the Sequoyah design will not effect the reactor protection or engineered safety features systems.

}2hh .

+'

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.0)

ADDITIONAL INSTRUMENTATION POSITION Licensees shall provide a decription of any additional instrumentation or controls (primary or backup) proposed fr.r the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

CLARIQCATION

1. Design of new instrumentation should provide an unamoiguous indication of inadequate core cooling. This may require new measurements to or a syntnesis of existing measurements ehich meet safety-grade criteria.

2. The evaluation is to include reactor water level indication.

3. A commitment to provide the necessary analysis and to study aavantages of various instruments to monitor water level and core cooling is required in the response to the September 13, 1979 letter.

4. The indication of inadequate core cooling must be unambiguous, in that, it :Pould have the following properties:

a) it must indicate the existence of inadequate core cooling caused by various phenomena (i.e., high void fraction pumped flow as well as stagnant boiioff).

b) it rust not erroneously indicate inadequate core cooling because of the presence of an unrelated phenomenon.

1296 217

5 :he indication must give advanced warning of the approach of inadequate core cooling.

6. The indication must cover the full range from normal operation to complete core uncavering. For example, if water level is chosen as the unambiguous indication, then the range of the instrument (or instruments) must cover the full range from normal water level to the bottom of the core .

m 22 1296 218

~

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

ADDITIONAL INSTRUMENTATION SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Analysis and procedures for the datection of inadequate core cooling using existing instrumentation are currently being developed in conjunction with the Westinghouse Owners' Group. This will be the primary method for detecting inadequate core cooling. In addition, TVA will provide instrumentation to measure water level in the reactor vessel down to the bottom of the hot leg piping.

RESPONSE

Analysis and procedures for the detection of inadequate core cooling using existing instrumentation are currently being developed in conjunction with the Westinghouse Owners' Group In addition to the above primary method for detecting inadequate core cooling described above, TVA will provide instrumentation to measure water level in the reactor vessel down to the bottom of the hotleg piping. This instrumentation will be designed and qualified in accordance with safety grade, Class IE, requirements including redundancy and emergency power.

The Reactor Vessel Level Instrumentation System was designed to provide direct readings of vessel level which can be used by the operator. This Reactor Vessei Level Instrumentation System does not replace existing systems and is not coupled to safety systems, but acts only to provide additional information to the operator.

23 \

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

ADDITIONAL INSTRUMENTATION RESPONSE (cont. )

The Reactor Vessel Level Instrumentation System consists of differential pressure measurement across the upper region of the reactor vessel. The system utilizes two differential pressure cells measuring the pressure drop from the bottom of the reactor coolant hot leg piping to the top of the reactor vessel head. The system provides an indication of reactor vessel water level above the bottom of the hot leg pipe when the pump in the loop with the hot leg connection is not operating. The number of pumps operating in the other loops has an effect of less than 10 percent of this indication.

When the pump is operating in the loop with the hot leg connection, the instrument reading will be off scale.

To provide the required accuracy for water level measurement, temperature measurements of the reference legs are provided. These measurements together with the reactor coolant temperature measurements are used to compensate the differential pressure transducer outputs for differences in reference leg temperature, particularly during the environment inside the containment structure following an accident.

The Reactor Vessel Level Instrumentation System utilizes differential pressure cell instrumentation in two of the hot leg pipes. The instru-mented hot leg piping will not be adjacent, but with respect to the plant layout, will be on opposite sides of the reactor vessel. The differential pressure cells for either of these options are to be located outside of containment such that calibration cell replacement, reference leg checks and filling, and operation are made more easily and the overall system accuracy is improved.

23A L

} 29 ()

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

ADDITIONAL INSTRUMENTATION RESPONSE (cont.)

Instrumentation for the operator for the Reactor Vessel Level Instrumentation System is intended to be unambiguous and reliable so that operator error or misinterpretation is avoided. The system would includ the following control board indicators:

An indication of upper region water level on each instrumented loop displaying water level in feet from 0 to -16 feet after compensation for any reactor coolant temperature and density effects. Indicator lights are included to indicate whether or not the pump in the loop is operating.

The Reactor Vessel Level Instrumentation is to be used in conjunction with a coolant subcooling readout to determine the state and transient behavior of the reactor coolant system. During normal operation, the reactor vessel level indicators would read off scale since the dynamic pressure drop due to coolant flow would be greater than the meter range. With all pumps shut down, the indicators will provide a direct indication of water level in the reactor vessel.

TVA will extend the range of incore thermocouples to give readout of fuel temperatures that could be expected if the core was partially uncovered.

1296 221

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)

ADDITIONAL INSTRUMENTATION CLARIFICATION ITEM 3

1. See the above response.

2. See the above response.

3. TVA's commitment is as submitted in response to the September 13, 1979, letter.

4. See the above respor.se.

5. Existing instrumentation and subcooling monitor is used to give advance warning of the approach of inadequate core cooling.

6. See the above response.

25 9

' 27-

\ iL9 6

CONTAINMENT ISOLATION (2.1.4)

POSITION,

1. All ;ontainment isolation system designs shall ccaply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.

2. All plants shall 91ve careful reconsideration to the definition of essential and non-essential systems, shall icentify each system ceterminea to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs dCCordingly, and shall report the results of the re-evaluation to NRC.

3. All non-essential systems shall be automatically isolated by the containment isolation si3nal.

4. 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 containment isolation valves shall require deliberate operator action.

CLARIFICATION

1. Provide diverse containment isolation signals that satisfy safety-grade requirements.

2. Identify essential and non-essential systems and provide results to NRC.

3. Non-essential systems should be automatically isolated by containment isolation signal s.

4. Resetting of containment isolation signals shall not result in the automatic loss of containment isolation 0

26 9

'1 12 %

s CONTAINMENT ISOLATION (2.1.4)

SEQUOYAH NUCLEAR PLANT RESPONSE SU WARY The Sequoyah Nuclear Plant Meets all of the NRC positions concerning containment isolation. Specific information pertaining to each of the psoitions is given below.

RESPONSE.

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, essential raw cooling water, and control air.

Phase B5 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 additica, isolation 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 each system which penetrates the containment, (b) determine whether or not it is essential, (c) describe basis for this determination, (d) modif y design if required, and (e) report results to NRC.

Every system that penetrates containment has been reevaluated to determine if it should be claccified as eccentid or noneccentia.l. S e errrent classifications have been fo"nd to be acceptable a .i no cL.:.get in classification are planned.

3. The Sequoyah Nuclear Plant design complies with NRC requirements on the automatic isolation of nonessential systems.

4. The Sequoyah Nuclear Plant design complies with the NRC's requirements by requiring nanual actions on the controls of individual components should it be necessary to change their status after the containment isolation signal has been cleared.

-2 7-

\ b

CONTAIMMENT ISOLATION (2.1.4)

CLARIFICATION ITEMS

1. Qualified diverse containment isolation signals are provided at Sequoyah.

2. As specified in the above response, ait evaluation of essential and non-essential systems has be performed and Sequoyah complies with NRC requirements. This information will be made available for NRC review.

3. See section *,of the above response.

4. See section 4 of the above response.

b

DEDICATED H 2

CONTROL PENETRATIONS (2.1.5.a)

POSITION Plants using external recomoiners or purge systems for post-accioent comoustiole 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 the redundancy and single failure requirements of General Design Criterion 54 ana 56 of Appencix A to 10 CFR 50, and that are sizec to satisfy the flow requirements of the recombiner or purge system.

CLARIFICATION

l. This requirement is only applicable to those plants whose licensing basis includes requirements for external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere.

2. An acceptable alternative to the dedicated ?enetration is a combinea cesign that is single-failure proof for containment isolation purposes and single-failure proof for operation of the recombiner or purge system.

3. The dedicated penetration or the comoined single-failure proof alternative should be sized such that the flow requirements for the use of the recombiner or purge system are satisfied.

4. Components necessitated by this requirement shoula be safety grace.

5. A description of required design changes and a schedule for accomplishing these changes should be provided by January 1,1980. Design cnanges should be completed by January 1,1961.

29 1296 226

OEDICATED H " " ^ " ***

2 SEQUOYAH NUCLEAR PTANT RESPONSE

SUMMARY

Sequoyah does not use external recombiners or purge systems for post-accident combustible gas control.

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).

CLARIFICATION ITEMS

1. Not applicable to Sequoyah Nuclear Plant.

2. Not applicable to Sequoyah Nuclear Plant.

3. Not applicable to Sequoyah Nuclear Plant.

4. See the above response.

5. Not applicable to Sequoyah Nuclear Plant.

30 1296 227 ,

~*

CAPABILITY TO INSTALL HYDROGEN RECOMBINER AT EACH LIGHT WATER NUCLEAR POWER PLANT (2.1.5.c)

POSITION The procedures and bases upon which the recombiners would be usea on all plants should be the subject of a review by the licensees in consicering sheilding requirements and personnel exposure limitations as demonstratea to be necessary in the case of TMI-2.

CLARIFICATION

1. This requirement applies only to those plants that included Hycrogen Recombiners as a design basis for licensing.

2. The shielding and associated personnel exposure 1. imitations associatea with recombiner use should be evaluated as part of licensee response to requirement 2.1.6.B, " Design review for Plant Shielding."

3. Each licensee should review ano upgrace, as necessary, those criteria and procedures dealing with recombiner use. Action taken on this requirement should be submitted by January 1,1960.

e 31

)2hb 2.

CAPABILITY TO INSTALL HYDROGEN RECOMBINER AT EACH LIGHT WATER NUCLEAR POWER PLAliT (2.1.5.c)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

This requirement is not applicable to Sequoyah

RESPONSE

The Sequoyah design has an ESF recombiner system inside containment which is redundant and fully qualified (see FSAR Section 6.2.5) and is manually actuated from the main control room.

CLARIFICATION ITEMS

1. Combustible gas control has been accounted for in the Sequoyah design as stated in the above response.

2. Their is no personnel exposure associated with recombiner use at Sequoyah.

3. Procedures for recombiner use are adequate for Sequoyah Nuclear Plant and no upgrading is required.

1296 229

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 iluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:

1. Immediate Leak Reduction

a. Implement 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-practicel levels. This program shall include periodic integrated leak tests at a frequency not to exceed rerueling cycle intervals.

Response

, TVA will investigate practical leakage reduction measures on systems which may contain radioactive fluids post-LDCA 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 the above systems that may be leak checked and will implement a periodic leak check program on these systems. System leakages will be reported to the NRC.

33 1296 230

DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICAT!aN OF EQUIPMENT FOR SPACES /SYSEMS WHICH MAY a: USED IN POST ACCIDENT OPERATIONS (2.1.o.8)

POSITION With the assumptior, of a post-accicent release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radiciodine,100% of the core noble gas inventory,and 1% of the core solids, are contained in the primary coolant), each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a r.esult of an accident, contain highly radioactive materials. The design review should inentify the location of vital areas and equipment, such as the control room, raawaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limitea or safety equipment may be unduly degraded by the radiation fielos during post-accicent operations of these systems.

Each licensee shall provice for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielaing, or post-accident procedural controls. The design review shall determine which types of corrective actions are neeced for vital areas throughout the f acility.

CLARIFICATION Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recuvery from an accident is designatea as a vital area. In order to assure that personnel can perform necessary post-accident operations in the vital arear we are providing the followi"-

guidance to be used by licensees to evaluate the acequacy of radiation protection to the operators:

1. Source Term The minimum r*mioactive source term should be equivalent to the source terms recommended, in Regulatory Guides 1.3,1.4, 1.7 and Stancaro Review Plant 15.6.5. witn appropriate cecay times based on plant cesign.

34 1296 231

a. Liquid Containing Systems: 100% of the core equilibrium noble gas inventory, 50% of the core equilibrium halogen inventory and 1% of all others are assumed to be mixed in the reactor coolant and liquids injected by HPCI and LPCI.

b. Gas Containing Systems: 100% of the core equilibrium noole gas inventory and 25% of.the core equilibrium halogen activity are assumed to De mixed in the containment atmospnere.

For gas containing lines connected to the primary system (e.g.,

BWR steam lines) the concentration of radioactivity shall De determined assuming the activity is contained in the gas space in the primary coolant system.

2. Oose Rate Criteria The dose rate for personnel in a vital area should be such that the guidelines of GDC 19 should not be exceeded during the course of the accident. GDC 19 limits the dose to an operator to 5 Rem whole boef or its equivalent to any part of the body. When determining the dose to an operator, care must be taken to determine the necessary occupancy time in a specific area. For example, areas requiring continuous occupancy will require much lower dose rates than areas

~

where minimal occupancy is required. Tnerefore, allowable dose rates will be based upon expected occupancy, as well as tne radioactive source terms and shielding. However, in order to provice a general design objective, we are providing tne following cose rate criteria 33 1296 232

with alternatives to be accumented on a case-Dy-case basis.

The recommended dose rates are vic.oye rates in the area. Local hot spots may exceeit the dose rate guidelines proviaed occupancy is not required at the location of the hot spot. These cases are design oojectives and are not to be used to limit access in the event of an accicent.

a. Areas Requiring Continuous Occupancy: <16mr/hr. These areas will require full time occupancy curing the course of the accident. The Control Room and onsite technical support

. center are areas where continuous occupancy will De required.

The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4.

b. Areas Requiring Infrequent Access: GDC 19. These areas may require access on a regular basis, but not continuous occupancy. Shielding should be proviced to allow access at a frequency and duration estimated oy the licensee. The plant Radiochemical / Chemical Analysis Laboratory, racwaste panel, motor control center, instrumentation loc .tions, and reactor coolant and containment gas sample station, are examples

, where occupancy may be needed of ten but not continuously.

1296 233

DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES /SYSEMS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS (2.1.o.6)

SE(KIOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Sequoyah Nuclear Plant is designed to mitigate major design basis events with no access outside the main control room required. Although the plant was not designed for access outside the control room, the current design may allow considerable capability for access for short times. TVA is evaluating the shielding requirements and will make design changes in shielding if the evaluation identifies feasible modifications which would significantly enhance desirable access.

RESPONSE

The Sequoyah design bases include the assumption of TID 14844 sources.

TVA plants ai- specifically designed to mitigate major design basis events with ro access outside the MCR being required. With this goal in mind, the plants wer, not specifically designed.for any access outside the main control room. fo specifically design for guaranteed access at anytime in most parts of the auxiliary building is not fassible. 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 instantes, 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.

37 1?06

'/ 234

DESIGN REVIEW OF PLANT SHIELDING AND ENVIR0hWENTAL QUALIFICATION OF EQUIPMENT FOR SPACES / SYSTEMS WHICH MAY.BE USED N POST ACCIDENT OPERATIONS (2.1.6.b)

RESPONSE (cont.)

TVA will 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 shielding in these vital areas will be reevaluated to assess its effectiveness in such a circumstance. The occupancy time limits, taking into consideration transit time and gamma shine intensities will then be calculated for the vital auxiliary building areas.

CLARIFICATION ITEMS

1. As specified in the above response, the Sequoyah design bases include the assumption of TID 14844 sources.

Sequoyah Nuclear Plant will meet the requirements of GDC 19.

J 38 t-

AUTO INITIATION OF THE AUXILIARY FEEDWATER SYSTEM (AFWS1 (2.1.7.A) a POSITION Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 witn respect to the timely initiation of tne auxiliary feecwater system, the following requirements shall be implementec in the short term:

1. The design shall provice for the automatic initiation of the auxiliary feedwater system.

2. The automatic initiation signals anc circuits shall be cesigncd so that a single failure will not result in tne loss of auxilir.ry feecwater system function.

3. Testability of the initiating signals and circuits shall De a feature of the design.

4. The initiating signals and circuits shall be powereo from the emergency buses.

5. Manual capaoil'ty to initiate the auxiliary feecwater system from the control room shall be retained and shall be implementec so that a single failure in tne manual circuits will not result in the loss of system function.

6. The a-c motor-criven pumps and valves in the auxiliary feecwater system shall be included in the automatic actuation (simultaneous anc/or sequer.tial) of the loacs onto the emergency buses.

7. The automatic initiating signals anc circuits shall De cesignec so that their failure will not result in the loss of manual capability to initiate the AFWS frcm the control room.

In the long tern, the automatic initiation sigaals anc circuits snall be upgracea in accordance with safety-grace requirements.

39 1296 236

CLARIFICATION Control Grade (Short-Term)

1. Provide automatic / manual initiation of AFWS.

2. Automatic initiation signals and circuits shall satisfy the single failure criterion.

3. Testability of the initiating signals ano circuits is requied.

4. Initiating signals and circuits shall De powered from the emergency buses.

5. Necessary pumps anc valves shall be '.nclucea in the automatic sequence of the loads to the emergency Duses. Verify that the adi'ition of these loacs does not comprimise the emergency diesel generating capacity.

6. Failure in the automatic circuits shall not result in the loss of manual capability to initiate the AFWS from the control. room.

7. Other Consicerations

a. For those designs wnere instrument air is neecea for operation, appropriate electric power must also De suppliea via the emergency power sources.

40 296 237

AUTO INITIATION OF THE AUXILIARY FEEDWATER SYSTEM (xFWS) (2.1.7.A)

POSITION Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 witn respect to the timely initiation of tne auxiliary feecwater system, the following requ'- ments shall De implementec in the short term:

1. The design shall provide for the automatic initiation of the auxiliary feedwater system.

2. The automatic initiation signals and ci cuits shall De cesigned so that a single failure wil' not result in the loss of auxiliary feecwater system function.

3. Testability of the initiating signals ard circuits shall De a feature of the design.

4. The initiating signals and circuits shal . be powereo from the emergency Duses.

5. Manual capability to initiate the aux'liary feecwater system f rom the control -com shall be retai'.ed and shall De implementec so that a single failure in the manual circuits will not result in the loss of system function.

6. The a-c motor-criven pumps and valves in the auxiliary feecwater system shall be incluced in the automatic actuation (simultaneous and/or sequential) of the loacs onto tne emergency ouses.

7. The automatic initiating signals ano circuits shall De designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.

In the long term, the automatic initiation signals and circuits shall be upgraced in accordance with safety grade requirements.

u 1296 238

CLARIFICATION Control Grace (Short-Term)

1. Provide automatic / manual initiation of ArNS.

2. Automatic initiaticn signals and circuit, shall satisfy the single failure criterion.

3. Testability of the initiating signals ano circuits is requied.

4. Initiating signals and circuits shall be powered from the emergency buses.

5. Necessary pumps anc valves shall be incluceo in the automatic sequence of the loads to ti.9 emergency buses. Verify that the addition of these loacs does not comprimise the emergency diesel generating capacity.

6. Failure in the automatic circuits shall not result in the loss of manual capability to initiate the AFWS f rom the control. rcom.

7. Other Considerations

a. For those designs where instrument air is neecea for operation, appropriate electric power musc also be suppliec via the emergency power sources.

1296 239 42

AUTO INITIATION OF AUXILIARY FEEDWATER (AFW) 2.1.7.A SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Sequoyah complies with all of the requirements of 2.1.7.A.

Response.

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 power system. Each auxiliary feedwater pump has ..anual initiation capa-The ac motor-driven pumps bility independent of the automatic initiation.

and valves are included in the automatic alignment of the loads to the emergency power system.

CLARIFICATION ITEMS

1. Automatic and manual initiation of AFW are provided at Sequoyah.

. . The AFW system is automatically initiated by redundant, coincident logic to preclude loss of function due to a single failure.

3. On line testability is provided.

4. Initiating signals are powered from the emergency power system.

5. The ac motor driven pumps and valves are included in the automatic alignment of loads to the emergency power system.

6. Manual initiation capability is provided independent of the automatic initiation.

7. Appropr electric power is supplied via the emergency power system for all valves wi.'re control air is needed for operation.

43 1.m? ?- 6 ?

m 4 0

tuXILIARY FEEDWATER FLOW INDICATION

. TO STEAM GENERATORS (2.1.7.b)

POSITION Consistent with satisfying the requirements set forth in 60C 13 to provice the capability in the control room to ascertain tne actual performance of the AFWS when it is callec to perform its intended function, the following requirements shall be implementec:

1. Safety-grade indication of auxiliary feecwater flow to each steam generator shall be proviced in the control room.

2. The auxiliary feedwater flow instrument channels shall oe powerec from the emergency buses consistent with satisfying the emergency ocwer diversity requirements of the auxiliary feecwater system set

~

forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

CLARIFICATION A. Control Grace (Short-Term)

1. Auxiliary feec4ater flow indication to each steam generator shall satisfy tne single failure criterion.

2. Testability of the auxiliary feedwater flow incication channels shall be a feature of the cesign.

3. Auxiliary feecsater flow instrument channeln shall be powerec from the vital instrument uuses.

B. Safety-Grade (Long-Tern)

1. Auxiliary feecwater flow indication to eacn steam generator shall satisfy safety-grace requirements.

C. Other

1. For the Short-Term the flow indication channels should oy themselves satisfy the single failure criterion for eacn steam generator. As 44 1296 c4i

a fall-back position, one auxiliary feed water flow channel may De oacked up by a steara generator level channel.

2. Each auxiliary feed water channel should provide an inoication of feed flow with an accuracy on the orcer cf + 10%.

1296 242

- AUXILIARY FEEDWATER FLOW INDICATION 2.1.7.B SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

AFW flow indication at Sequoyah is not calssified as safety grade; however, the components and design are similar to those used for safety grade systems.

The AFW flow instrumentation channels are powered f rom the emergency buses.

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 transmitters 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 inser ment channels are powered from the emergency buses consistent with the d .ersity requirements of the auxiliary feedwater system.

CLARIFICATION ITEMS A. Control Grade (Short Ters)

1. AFW flow indication to each steam generator does not satisfy the single failure criterion. (See item C.1)

2. Testability of the AFW flow indication channels is provided.

(See item C.1)

3. The AFW flow instrument channels are powered from the emergency buses.

B. Safety Grade (Long Term)

1. AFW flow indication at Sequoyah is not classified as safety grade; however, the components and design are similar to those used for safety grade systems. The AFW flow instrumentation channels are powered from the emergency buses.

C. Other

1. The AFW flow indication channels do not by themselves satisfy the safety-grade requirements; however, the components and design are similar so those used for safety grade systems. In addition, the steam generator water level indication is safety-grade and satis-fies the single failure and testability requirements. The steam generator water level provides backup indication of feedwater fle"

2. Each AFW flow instrument channel provides an irdicaiton of feed flow with an accuracy on the order of *10 percent.

1296 243

IMPROVED POST-ACCIDENT SAMPLING CAPABILITY (2.1.c.a)

POSITION A design and operational review of the react. coolant and containment atmosphere sampling systems shall be performea to cetermine tne capacility of personnel to promptly obtain (less than I hour) a sample under accicent conditions without ircurring a radiation exposure to any indivicual in excess of 3 anc 18 3/4 Rems to the whole Dody or extremities, 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 sampias, additional cesign features or shielaing should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capaoility 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 inoicators of tne degree of core camage. Such racionuclides are noole gases (which indicate cladding failure), iodines and cesiums (which indicate nigh fuel temperatures), and non-volatile isotopes (whicn indicate fuel melting).

The initial reactor coolant spectrum should corresponc to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct raciation from piping and components in the auxilie Duilding and possible contamination and direct raciation from airDorne :uents. If the review incicates that the analyses required cannot be crformed in a prompt manner with existing equipment, then design mooifications or equipment procurement shall De undertakan to meet the criteria.

In addition to the radiological analyses, certain cnemical analyses are necessary for monitaring reactor concitions. Procedures shall be provicec to perform boron anc chloride chemical analyses assuming a highly racio-active initial sample (Regulatory Guide 1.3 or 1.4 source term). Botn analyses shall be capaole of oeing completea promptly; i.a., the coron sample analysis within an hour and the chloride sample analysis within a shif t.

DISCUSSION The primary purpose of implementing Improvec Post-Accicent Sampling Capacility is to improve efforts to assess and control the course of an accicent cy:

1. Providing information related to tne extent of core damage that has occurrec or may be occurring curing ar. accicent;

2. Determining the types and quantities of tission products releasec to the containment in the liquid and gas pnase and wnich may be releasea to the environment; 1296 244

3. Providing information on coolant chemistry (e.g., dissolved gas,

- boron and pH) and containment hydrogen.

The above information requires a capability to perform the following analyses:

1. Radiological and chemical analyses of pressurized and unpressurized reactor coolant liquid samples;

2. Radiological and hydrogen analyses of containment atmosphere (air) samples.

CLARIFICATION A. The licensee shall have the capability to promptly obtain (in less than I hour) pressurized and unpressurized reactor coolant samples and a containment atmosphere (air) sample.

The licensee shall establish a plan for an onsite radiological and chemical

~

B.

analysis facility with the capability to provide, within i hour of obtaining the sample, quantification of the following:

1. certain isotopes that are indicators of the degree of core damage (i.e., noble gases, iodines and cesiums and non-volatile isotopes),

2. hydrogen levels in the containment atmosphere in the range 0 to 10 volume percent,

3. dissolved gases (i.e., H 2 , 0 2) and boron concentration of liquids.

4. pH of liquids, or have in-line monitoring capabilities to perform the above analysis. Plant procedures for the handling and ar_alysis of samples, minor plant modifications for taking samples and a design review and procedural modifications (if necessary) shall be completed by January 1,1981.'- Major plant modifications shall be completed by January 1, 1981.

C. During the review of the post accident sampling capability consideration should be given to the following items:

1. Provisions shall be made to permit containment atmosphere sampling under both positive and negative containment pressure.

2. The licensee shall consider provisions for purging samples lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose meterial in the RCS or containment, for appropriate disposal of the samples, and for passive flow restrictions to J'mit reactor coolant loss or containment air leak from a rupture of the sample line.

48 1296 245

3. If changes or modifications to the existing sampling system are required, the seismic design and quality group classification or sampling lines and esponents shall conform to the classification of the system to which each sampling line is connected. Components and piping downstream of the second isolation valve can be designed to quality Group D and nonseismic Category I requirements.

D. The licensee"s radiological sample analysis capability should include provisions to:

a. Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms given in Lessons Learned Item 2.1.6.b. Where necessary, ability to dilute samples to provide capability for measurement and reduction of personnel exposure, should be provided. Sensitivity of onsite aaalysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 uli/gm to the upper levels indicated here.

b. Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with an acceptably small crror (approximately a factor of 2). This can be accomplished through the use of suf ficient shielding around samples and outside sources, and by the use of ventilation systen design which will control the presence of airborne radioactivity.

c. Maintain plant procedures which identify the analysis required, measurement techniques and provisions for reducing background levels.

E. The licensees chemical analysis capability shall consider the presence of the radiological source term indicated for the radiological analysis.

F. In performing the review of sampling and analysis capability, consideration shall be given to personnel occupational exposure. Procedural changes end/or plant modit'ications must assure that it shall be possible to obtain a sd analyze a sample while incurring a radiation dose to any individual tha*. is as low as reasonable achievable and not in excess of DGC 19.

In assuring that these limits are met, the following criteria will be used by the staff.

1. For shielding calculations, source terms shall be as given in Lessons Learned Item 2.1.6.b.

2. Access to the sample station and the radiological and chaaical analysis facilities shall be through areas which are accessible ir. post accident situations and which are provided with sufficient shielding to assure that the radiation dose criteria are met.

3. Operations in the sample station, handling of highly radioactive samples from the sample station to the analysis facilities, and handling while working with the samples in the analysis _acilities shall be such that the radiation dose criteria are met. This may involve sufficient shielding of personnel from the samples and/or the dilution of samples for analysis. If the existing facilities do not satisfy these criteria, then additional design features, e.g., additional shielding, remote handling etc. shall be provided. The radioactive sample lines in the sample station, the samples themselves in the analysis facilities, and other radioactive lines of the vicinity of the sampling station and analysis facilities shall be included in the evaluation.

49

4. High range portable survey instruments and personnel dosimeters should be provided to permit rapid assessment of high exposure rates and accumulated personnel exposure.

G. The licensee shall demonstrate their capability to obtain and analyze a sample containing the isotopes discussed above according to the criteria given in this section.

1296 247 50

IMPROVED POST-ACCIDENT SAMPLING CATABILITY 2.1.8.A SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

A design and operational review of the reactor coolant and containment atmosphere sampling systems and analysis facilities is being performed.

RESPONSE

A design and operational review of the reactor coolant sampling systems and analysis facilities is being performed and will be complete by January 1, 1980. TVA expects to complete required modifications by January 1, 1981, provided that equipment procurement / installation con-flicts are not encountered. These modifications will make provisions for sampling water from the reactor coolant system for the degraded accident condition. TVA will also identify the type and nature of onsite analysis required. If practical, TVA will procure the required analysis equipment and locate, design, and build an onsite analysis facility.

Until the design modifications are complete, procedures will be devised to evaluate the primary coolant system activity depending on the accessi-bility of the sampling stations for particular degraded conditions.

To enhance the capability at Sequoyah for pos:-LOCA s.apling 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 connecitons to the existing gaseous radiation sampling system for use in sampling the containment atmosphere for the degraded accident conditions.

3. Route sample lines to a shielded sampling station in an accessible area and provide for taking samples which could b6 removed offsite for analysis.

CLARIFICATION ITEMS A. TVA will provide the capability to obtain (within one hour) Reactor Coolant samples and containment air samples under accident conditions.

This capability w ill be provided by January 1,1981.

51 1296 248

B. Plant procedures for the handling and analysis of samples, minor plant modifications for taking samples, and a design review and procedural modificaticns (if necessary) will be completed by January 1, 1980. TVA will provide, as practical, onsite radiological and chemical analysis capacities in order to quantify the following:

1. core damage (RES)

2. hydrogen level in containment

3. dissolved gases and boron content (RCS)

4. pH (RCS)

C. Provisions will be made:

1. to permit sampling under both positi'e and negative pressure.

2. ior purging sample lines, for reducing plateout in sample lines, fcr minimizing sample loss or distortion, for preventing blockabe of sample lines, for appropriate disposal of samples, and for passive flow restrictions.

3. to qualify the sampling system to appropriate seismic and environmental requirements.

D. The radiological sample analysis capability will include provisions to:

a. Identify and quantify isotopes to levels corresponding tc the source terms given in item 2.1.6.B. The ability to dilute samples and to measure nuclide concentrations as low as 1 pCi/gm will be provided.

b. Restrict background levels in the health physics laboratory.

c. Maintain plant procedures to identify the analysis required, measurement techniques and provisions for reducing background.

E. The chemical analysis capability will consider the presence of the radiological source term indicated by the radiological analysis.

F. Procedural changes and plant modifications will be made to assure that radiation exposures are as low as reasonably achievable. TVA will ensure that these criteria are met using tha criteria identified.

G. TVA will demonstrate the capability to obtain and analyze a sample containing radio isotopes according to the criteria discussed above.

32 1296 249

INCREASED RANGE OF RAu1ATION MONITORS (2.1.8.b)

POSITION The requirements associatec with this recorrmendation shoulo De consicerec as advanced implementation of certain requirements to be incluceo in a revision to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accicent",

which has alreacy been initiated, and in other Regulatory Guices, wnich will De promulgated in the near-term.

1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as curing normal operating conoitions; multiple monitors are consicerec to be necessary to cover the ranges of interest.

a. NoDie gas effluent monitors with an upper range capacity of 10" uti/cc (Xe-133) are considered to be practical ano shoulo ce installed in all operating plants.

b. Noble gas effluent monitoring shall be proviced for the total range of concentrativ extencing from normal condition ( ALARA) concentrations to a maximum of 10 uCi/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges or interest. The range capacity of indivicual monitors should overlap by a factor of ten.

2. Since iodine gaseous effluent monitors for tne accicent conoition are not considered to be practical at this time, capability for effluent monitoring of radiciocines for the accident condition snail De provided with sampling conducted by adsorption on charcoal or other meaia, followed by onsite laboratory analysis.

3. In-containment radiation level monitors with a maximum range of lub rad /hr shall be installed. A minimum of two such monitors that are physically separatec shall De proviced. Monitors snall be cesigned and qualified to function in an accicent environment.

DISCUSSION The January 1,1960 requirement, were specifically acced Dv the Commission and were not incluqed in NUREG-0578. The purpose of the interim January 1, 1980 requirement is to assure that licensees have methocs of quantifing racioactivity releases shoula the existing effluent instrumentation go offscale.

CLARIFICATION

1. Raaiological Noble Gas Effluent Monitors A. January 1,1980 Requirements until final implementation in January 1,1961, all operating reactors must provice, by January 1,1960, an interim metnod for 53 1296 250

quantifying high level reinses which meets the requirements of Table 2.1.8.D.l. This method i, to serve only as a provisional fix with the more detailec, exaco methocs to follow. Methcos are to De cevelopeo to quantify release rates of up to 10,000 Ci/sec for noble gases from all potential release points, (e.g. , auxiliary ouiloing, rauwaste building, fuel handling builcing, reactor ouilding, waste gas cecay tank releases, main condenser air ejector, BWR main concenser vacuum pump exhaust, PWR steam safety valves and atmosphere steam cump valves ana BWR turoine buildings) and any other areas that communicate cirectly witn systems which may contain primary coolant or containment gases, (e.g., letcown and emergency core cooling systems ano external recombiners).

Measurements / analysis capabilities of the effluents at the final release point (e.g., stack) should De such that measurements of indivicual sources which contribute to a common release point may not ce necessary.

For assessing radioicaine and particulate releases, special procedures must be developed for the rem..al and analysis of the racioiocine/

particulate sampling mea'a (i.e., charcoal canister / filter paper).

Existing sampling locations are expectea to De acequate; however, special procecures for retrieval ano analysis of the sampling media under accident conditions (e.g., hign air and surface contamination and direct radiation levels) are needec.

It is intenced that the monitoring capabilities calleo for in the interim can be accomplished with existing instrumentation or reactly available instrumentation. For noole gases, modifications to exist-ing monitoring systems, such as the use of portacle nigh range survey 1296 2SI

instruments, set in shielded collimators so tha,t they "see" small sections of sampling lines is an acceptable methoc for mceting the intent of this requirement. Conversion of the measureo cose rate (mR/hr) into concentration (uCi/cc) can be performed using stancara volume source calculations. A method must ce oeveloped with sufficient accuracy to quantify the iocine releases in the presence of nign background raciation from noble gases collectea on charcoal filters.

Seismically qualifiea equipment and equipment meeting IEEE-279 is not required.

The licensee shall provice the following information on his methocs to quantify gaseous releases of racioactivity from the plant curing an accicent.

1. Noble Gas Effluents

a. System / Method description including:

i) Instrumentation to be used incluaing range o' sen-sitivity, energy depencence, anc calibration frequency and technique, ii) Monitoring / sampling locations, incluaing methocs to assure representative measurements and backgrouna radiation correction, iii) A cescription of method to De employed to facilitate access to radidtion readings. For January 1,1960, Control room read-out is perferrea: however, if impractical, in-situ readings by an incivicual witn veroal communication witn tne Control Room is acceptacle basec on (iv)

Delow.

1296 252

iv) CapaDility to oDtain radiation readings at least every 15 minutes curing an accicent.

v) Source of power to be usea. If normal AC power is useo, an alternate Dack-up power supply shoula be provided. If DC power is used, the source shoula De capable of proviaing continuous readout for 7 consecutive days.-

D. Procedures for concucting all aspects of the measurement /

analysis incluaing:

1) Procecures for minimizing occupational exposures

11) Calculational methods for converting instrument reacings to release rates cased on exnaust air ficw and taking into consideration radionuclide spectrum distrioution as function of time af ter shutcown, iii) Procedures for dissemination of information.

iv) Procecures for calibration.

B. January 1,1961 Requirements By January 1,1961, the licensee shall provice high range noDie gas effluent monitors for each release path. The noble gas effluent monitor should meet the requirements of Table 2.1.6.D.2.

The licensee shall also provice the information given in Sections 1. A.I.a.i, l.A.l.a.ii, l.A.l.a.11, l . A.I .b. iii, and 1. A.I .D. iv above f or tne noole gas effluent monitors.

56 1296 253

B. For January 1,1981, the licensee should have the capability to continuously sample and provide onsite analysis of the sampling media. The licensee should also provide the information required in 2. A above.

3. Containment Radiation Monitors Provide by January 1,1981, two radiation monitor systems in containment which are documented to meet the requirements of Table 2.1.u.b.2.

It is possible that future regulatory requirements for emergency planning interfaces may necessitate identification of different types of radionuclices in the containment air, e.g., noble gases (indication of core camage) and non-volatiles (indication of core melt). Consequently, consideration should be given to the possible installation or future conversion of these monitors to perform this function.

se 1296 254

2. Radiciodine anc Particulate Ef fluents .

A. For January 1,1960 tne licensee shoulo provide the following:

1. System /Methoc cescription incluaing:

a) Instrumentation to be used for analysis of the sampling meaia with discussion on methods usec to correct for potentially interfering backgrcuno levels of racioactivity, b) Monitoring / sampling location.

c) Methoc to be useo for retrieval and nandling of sampling media to minimize occupational exposure.

c) Metnod to De usea for cata analysis of incivicual radionuclides in the presence of high levels of radioactive noole gases.

e) If r.ormli AC power is used for sample collection ano analysis equipment, an alternate bacK-up power supply shoula De proviced. If DC power is used, the source should be capable of providing continuous read-out for 7 consecutive cays.

2. Procedures for conducting all aspects of the measurement analysis incluaing:

a) Minimizing occupational exposure b) Calculational methods for cetermining release rates c) Procedures for cissemination of information d) Calibration frequency and tecnnique 57 1296 255

e TABLE 2.1.8.b.1 INTERIM PROCEDURES FOR quANTIFYIf4G HIGH LEVEL ACCIDENTAL RADICACTIVITY RELEASES

. Licensees are to implement procecures for estimating noole gas and racioiodine release rates if the existing effluent instrumentation goes off scale.

. Examples of major elements of a highly radioactive effluent release special procedures (noble gas).

- Preselectea location to measure radiation from the exhaust air, e.g. , exhaust duct or sample line.

- Provide shielding to minimize bactarouno interference.

- Use of an installed monitor (preferable) or dedicatec portable monitor (accepta*.le) to measure the radiation.

Predetermined calculational method to convert the radiation level to radioactive effluent release rate.

\7.96 59

TABLE 2.1.8.b.2 HIGH RANGE EFFLUENT MONITOR

. NOBLE GASES ONLY

. RANGE: L0verlap with tiormal Effluent Instrument Range)

UNDILUTED CUNTAINMENT EXHAUST lu+S uCi/CC DILUTED (>10: 1) CONTAlhMENT EXHAUST 10*4 uCi/CC MARK I BWR REACTOR BUILDING EXHAUST 10 " uCi/CC PWR SECONDARY CONTAINMENT EXHAUST 10*4 uCi/CC BUILDINGS WITH SYSTEMS CONTAINING PRIMARY COOLANT OR LASES 10,3 pCi/CC OTHER BUILDINGS (E.G., RADWASTE) 10 uCi/CC

. NOT REDUNDANT - 1 PER NORMAL RELEASE POINT

. SEISMIC - NO

. POWER - VITAL INSTRUMENT BUS

. SPECIFICATI0t45 - PER R.G.1.97 AND ANSI N320-1979

. DISPLAY *: CONTINUQUS AND RECORDING WITH READ 0VTS IN THE TECHNICAL SUPPORT CENTER (TSC) AND EMERGENCY OPERATIONS CENTER (EOC)

. QUALIFICATIONS - NO

• Althougn not a present requirement, it is likely that tnis information may have to be transmitted to *.he NRC. Consequently, consiceration snould be giv?n to this possiDie future requirement wnen cesigning the cisplay interfaces.

1296 257 60

TABLE 2.1.8.o.3 HIGH RANGE CONTAIhMENT RADIATION MONITCR

. RADIATION: TOTAL RADIATION (ALTERNATE: PHOTON ONLY)

. RANGE:

O UP T010 RAD /HR (TOTAL RADIATION) 7 ALTERNATE: 10 R/HR (PHOTON RADIATION ONLY)

- SEdSITIVE DOWN TO 60 KEY PHOTONS *

. REDUNDANT: TWO PHYSICALLY SEPARATED UNITS

. SEISMIC: PER R. G.1.97

. POWER: VITAL INSTRUMENT BUS

. SPECIFICATIONS: PER R.G.1.97 REV. 2 AND ANSI N320-197d

. DISPLAY: CONTINU0US AND RECORDING

. CALIBRATION: LABORATORY CALIBRATION ACCEPTABLE

• Monitors must not provice misleading information to the operators assuming celayed core damage when the 60 KEV photon Xe-133 is the major noble gas present.

1296 258 61

• INCREASED RANGE OF RADIATION MONITORS 2.1.8.B SEQUOYAH NUCLEAR PLANT RESPONSE _

SUMMARY

Sequoyah will comply with the requirements of 2.1.8.B by January 1, 1981, except that high level radiation monitors will bemeasures Interim located outside the will be provided annulus instead of inside containment.

before fuel loading in the respeutive units for quantifying high level releases.

Response

Redundant safety grade high range noble gas ef fluent monitors will be provided at Sequoyah on the shield building vents.

A method or methods of sampling effluent particulates and iodine will be chosen and redundant particulate and iodine effluent sampling systems to the present state-of-the-art will be provided.

The present bQN design has one high range radiation monitor outside the containment in the auxiliary building, opposite the personnelHowever, hatch toits detect high levels of radioactivity inside the containment.

range is not as high as required by the NTC. Redundant radiation monitors will be provided outside the annulus to meet the NRC's high-range requirement.

These monitors will be safety grade and will be designed and qualified to function in an accident environment.

Interim Procedures for Quantifying Hieh Level Accidental Radioactivity Releases To provide interim measures to estimate high level releases, TVA now plans to install a temporary high-range detector external to the sampling tubing of the shield building vent monitor. The detector will monitor only gcc;s radioactivity releases and will not be able to distinguish the radioiodine contribution of the total release. TVA will provide a method for easily converting the detector readings and vent flow rate to activity release rates.

CLARIFICATION ITEMS

1. Noble Gas Effluent Monitors A. Requirements for January 1,1980 - TVA will provide an instrument to monitor gross releases of radioactivity from the shield building vent.

Our present shield building vent monitor provides a gaseous sample for laboratory analysis. Special procedures will be developed for estimating noble gas effluent in the event present instrumentation saturates. A description of these systems and methods will be made available to the NRC.

62 1296 259 . _ _ _ _

e

B. By January 1, 1981, TVA will provide high range noble gas effluent monitors for all identified release paths. This monitor will meet the requirements of Table 2.1.8.B.2. Information requested on these monitors will be made available to the NRC.

2. Radioiodine and Particulate Effluents A. Requirements for January 1, 1980 - Methods used to sample radio-iodine and particulates will be provided.

B. By January 1,1981, TVA will provide the capability to continuously sample effluents and onsite analysis for radioiodine and particulates with state-of-the-art equipment. The requested information will be made available to the NRC.

3. Containment Radiation Monitors By January 1, 1981, TVA will provide two radiation monitors outside the annulus which meet the intent of thc requirements of Table 2.1.8.B.3.

1296 260 63

IMPROVED IN-PLA!1T IODINE INSTRUNE!4TATION UNDER ACCIDENT CONDITIO!45 (2.1.o.c)

POSITION Each licensee shall provide equipment and associatec training and procedures for accurately determining the airborne iocine concentration in areas within the facility where plant personnel may be present curing an acci, cent.

CLARIFICATI0fi Use of Portable versus Stationary Monitorina Eauipment Effective monitcring of increasing iodine levels in the builuings uncer accident conditions must include the use of portaole instruments for the following reasons:

a. The physical size of the auxiliary / fuel handling builcing precludes locating stationary monitoring instrumentation at all creas where aircorne iocine concentration data raight be requirea.

b. Unanticipated isolated " hot spots" may occur in locations where no stationary monitoring instrumentation is located.

c. Unexpectedly high background radiation levels near stationary monitoring instrumente. .on af ter an accicent may interfere with filter radiation reacings.

64 1296 261

d. The time requirec to retrieve samples af ter an accicent may result in high personnel exposures if these filters are located in high dose rate areas.

locine Filters and iteasurement Technioues A. The following are snort-term recommencations anc sna11 De implemented by the licensee oy January 1,1960. The licensee shall have the capability to accurately cetect the presence of iodine in the region of interest following an accident. This can De accomplished oy using a portable or cart-mounten incine sampler with attached single channel analyzer (SCA). The SCA wincew shoula ce caliorated to the 365 key of I.

A representative air sample shall be taken and then countea for I using the SCA. This will give an initial cMservative estimate of presence of iodine and can be used to deterrline if respiratory protection is requirea. Care must be taken to assure that the counting system is not saturateo as a result of too must activity collectea on the sampling cartricge.

B. By January 1,1981:

The licensee shall have the capability to remove the sampling cartridge to a low backgrounc, low contamination area for further analysis. This area shoula De ventilated with clean air containing no airoorne racionuclides which may contribute to inaccuracies in analyzing the sample. Here, the sample should first be purgec of any entrappec noole gases using nitrogen gas or clean air free of noole cases. The licensee shall have the capability to measure accurately the iocine concentrations present on these samples ano effluent charcoal samples under accident conoitions.

1296 262

IMPROVED IN-PLANT IODINE INSTRUME7TATION UNDER ACCIDENT CONDITIONS 2.1 8.C SEQUOYAH NUCI. EAR PLANT RESPONSE

SUMMARY

Sequoyah ahs low-volume portable air monitore equipped with a charcoal filter to absorb iodine isotopes. These filters will be analyzed in the health physics laboratory. This capability meets the requirements of 2.1.8.C.

Response

Sequoyah has portable lew-volume air samplers, each equipped with a particulate filter followed by a charcoal absorber to collect iodine isotopes. The particulate filter will be counted in the health physics laboratory for gross activity and the charcoal absorber sent to the radiochemical laboratory for a gamma isotopic analysis for radioactive iodines. If necessary, as necessitated by a hig' -gross activity, the particulate filter will also be sent to the radiochemical laboratory for an isotopic analysis. The primary dif f erence in obtaining in-plant airborne isotop;c concentrations for accident and routine operating condi-tions is the time required for sampling. A shorter sample time co"1d be necessary for accident conditions because of the presence of high _sotopic concentrations.

The plar.: has procedures for sampling and analysis of in-plant air spaces incorporated in the Health Ohysics Laboratory Instruction Manual and the Radiation Control Instruction Manual.

Pla L bealth physics technicians are required to complete a formal training program plus receive in-plant training which includes the use of health physics procedures and instrumentation.

CLARIFICATION ITEMS A. Requirements before fuel loading.

Sequoyah has portable low-volume air samplers equipped with a charcoal absorber to collect iodine. The filters are removed and taken to the health physics laboratory for analysis. Single channel analyzers are not provided for detection of iodine.

B. January 1, 1981, requirements.

Sequoyah's low-volume air samplers and health physics lab meet these requirements.

1296 263

TRANSIENT AND ACCIDENT ANALYSIS (2.1.9)

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.

Soma analysis requiremente 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 natura 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 .epeated taking credit for correct operator action. TF. additicaal cases will provide the basis for developing appropriate er ergenc' pr >cedures. These calcula-tions should also provide the analyt'.a] basis ror ' ' design of any additional instrumentation needed to provide ope . ors with an unambiguous indication of vessel water level ar.d core cooling adequacy (see Section 2.1.3.b in this appendix).

The analyses of transients and ac:idents shall include the design basis events specified in Section 15 c. each FSAR. The analyses shall include a single active failure for ea< a 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. Operator actions that could cause the complete loss of function ~ 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 67 b 2h4

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 available 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 defiaing the event trees and would be useful in studying the informa-tion available to the operators. The transient and accident analyses are to bc ,eriormed 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 oe performed by the NRC Office of Research, using the best available compt.ter 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, includi,g 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.

DISCUSSICj The scope of OSe required transient and eccident analysis is discussed in NUREG-0578. The schedule for these analyses is incluited in NUREG-0578 and is reproduced in the Implementation Schedule atta:hment to this letter.

The Bulletins and Orlers Task Force has been implementing these required analyses on that schedule. The analysis of the small treak loss of coolant accident has been submitted by each of the owners groups. These analyses are presently under review by the 560 Task Force. The scope and schedule for the analyeis of inadequate core cooling have been discussed and agreed upon in meetings between the owners groups and the B&O Task Force, and are documented in the minutes to those meeting-The analysis of transients and accidents for the purpose of upgrading emergency procedures is due in early 1980 and the detailed scope and schedule of this analysis is the subject of continuing discussions between tF; owners groups and the B&O Task For' 68 1296 265

TRANSIENT AND ACCIDENT ANALYSIS 2.1.9 SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

TVA is pursuing the required analyses and the development of new procedures and training guidelines with other utilities through the Westinghouse owners group. We doubt that the extremely ambitious implementation schedule of NUREG-0578 can be met without extraordinary effort on all parts.

Response

TVA is pursuing the required analyses and the development of new pro-cedures and training guidelines with other utilities through the Westinghouse TML 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.

Inadequate core cooling is an item where further definition of the scope, such as system fai. lure and operato*r error assumptions, is needed from the NRC. At present model preparation is in progress to permit response to identified actica. 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 conditians. 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 consequences of accidents using realistic sssumptions incorporating 'he 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.)

1?96 266

Appropriate changes can then be incorporated into the existing procedures, designs, and training programs.

Development of the models to incorporate such effects is in itself a 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 anhitious. 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.

70 1296 267

C0t4TAINNENT PRESSURE INDICATION P051TI014 A continuous indication of containment pressure shoula be provicea in tne control room. Measurement and indication capaoility shall incluce three times the oesign pressure of the containment for con: rete, four times the design pressure for steel, anc minus five psig for all containments.

CLARIFICATION .

1. The containment pressure indication shall meet the cesign provisions of Regulatory Guide 1.97 including qualification, recuricancy, and testabili ty.

2. The containment pressure monitor shall be installeo oy January 1,19ol.

71 1296 268

CONIAINMENT PRESSULF "")ICATION 2.1.9(a)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Sequoyah will compty with all of the requirements of this position before January 1, 1981.

Response

Four qualified, continuous indications of the coi.tainment pressure are provided in the main control room. The 5 ps g; negative pressure requirement is not applicable to Sequoyah since gualified vacuum relief of the containment maintains the pressurr at greater than negat*c; 0.5 psig. The negative range of the r41st17g pressure indf;ators envilopes this negative 0.5 psig liuf.t; Redundant, continuous con:ainment pressure indication with a range up to four times the des! 3n pressure of the steel containment will be provided.

CLARIFICATION ITEMS

1. The monitors will meet the applicable design requirements for qualification, redundancy and testability.

2. The monitors will be installed and operational by Jan.iary 1, 1981.

1296 269 72

CONTAINMENT HYDROGEf4 INDICAT10ii POSITION A continuous ir.dicaton of hyorogen concentration in the containment atmospnere shall be provided in the conte ' room. Measureuent capacility shall be provided over the range of 0 to 10% hydrogen concentration uncer Doth positive anc negative ambient pressure.

CLARIFICATION

1. The containment nydrogen inoication shall meet the cesign provisions of Regulatory Guice 1.97 incluaing qualification, recunaancy. and testabili ty.

2. The containment hydrogen indication shall De ir.stallea Dy January 1,1961.

1296 270

CONTAINMENT HYDROGEN INDICATION 2.1.9(b)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

Sequoyah has redundant safety-grade hydrogen analyzers located in the annulus. These monitors have a range of 0 to 10 percent hydrogen concentration. Sequoyah complies with all of the requirements of this NRC position.

Response

Redundant, safety-grade 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 from negative 2 psig to positive 50 psig pressure.

CLARIFICATION ITEMS

1. The hydrogen analyzers of Sequoyah meet the applicable requirements for qualification, redundancy and testability.

2. These analyzers are installed and operational.

74 1'?96 771

CONTAINMENT WATER LEVEL INDICATION POSITION A continuous indication of contair. ment water level shall be providea in the control room for all plants. A narrow range instrument shall De proviced for PWRs and cover the range from the Dottom to the top of the containment sump. A wice range instrument shall also be provicea for PWRs anc shall cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity. For BWRs, a wice range instrument snali ce provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool.

CLARIFICATION

1. The narrow range sump level instrument shall monitor the normal contain-ment sump level vice the containment emergency sump level.

2. The wice range containment water level .nstruments shall meet the require-ments of Regulatory Guice 1.97 (Instrumentation for Light-Water Cooled Nuclear Power Plant to Assess Plant Concitions During ana Following a Accident).

3. The narrow range containment water level instruments shall meet tne requirements of Regulatory Guide 1.89 (Qualification of Class IE Equipment of Nuclear Power Plants).

4. The equivalent capacity of the wide range PWR level instrument has been cnangeo from 500,000 gallons to 600,000 sallons to ensure consistency with the proposed revision to Regulatory Guide 1.97.

S. The containment water level indication shall ce installea oy January 1,1961.

75 1296 272

CONTAINMENT WATER LEVEL INDICATION 2.1.9(c)

SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

The sump water level is indicated by four separate qualified, and continuous level instruments with readout in the main control room.

These instruments provide adequate indication of the water level in the sump. Sequoyah complies with all of the requirements of this NRC position.

Response

The floor of the reactor building serves as the sump for the contain-ment. It is instrumented with four reparate, qualified, and continuous level instruments which indicate 1. tae main control room. The range of the instruments is from less then six inches above the floor up to 20 feet above the floor. If 600,000 gallons of water were introduced into containment in addition to the fluid volume of

' the reactor coolant system, safety injection accumulators, and a tota? ice melt, the containment water level would not exceed the 20 ft. range of the level instruments. A small sump suction pocket (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.

CLARIFICATION ITEMS

1. The narrow range sump level instrument monitors the normal contain-ment sump level and the wide range sump level instrument monitors the emergency sump level.

2. The wide range sump level instrument meets the appropriate require-ments of Regulatory Guide 1.45.

3. The narrow range sump level instrument meet the applicable require-ments for qualification, redundancy, and testability.

T. If 600,000 gallons of water were introduced into containment, in addition to the entire 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 for the wide range water level monitor.

5. The sump water level monitors are installed and operational.

76 1296 273

. REACTOR COOLANT SYSTEM VENTING POSITION Each applicant ana licensee shall install reactor coolant system ano reactor vessel head high point vents remotely operated from tne control rcom. Since these vents form a part of the reactor coolant pressure bouncary, tne cesign of the vents shall confera to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria. In particular, these vents snall De safety grade, ano shall satisfy the single failure criterion anc the raquirements of IEEE-279 in orcer to ensure a low probability of inadvertent actuation.

Each applicant and licensee shall provide the following information concerning the design and operation of these high point vents:

1. A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accicents initiatea by a break in the vent pipe. The results of the analyses shoulo De demonstratea to be acceptable in accordance with the acceptance criteria of 10 CFR bO.46.

2. Analyses deomcnstrating that the cirect venting of nonconcensaole gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentration limits in containment as describec in 10 CFR Part 50.44, Regulatory Guice 1.7 (Rev. 1), anc Stancard Review Plan Section 6.2.5.

3. Procecural guidelines for the operators' use of tne vents. The information available to the operator for initiating or terminating vent usage shall De discusseo.

CLARIFICATION A. ueneral

1. The two important safety functions ennanceo by tnis venting capability are core cooling and containment integrity. For events witnin the present design basis for nuclear power plants, the capability to vent non-condensiole gases will provice additional assurance of meeting tne requirements of IUCFR50.46 (LOCA criteria) ano 10CFRou.44 (containment criteria for nyarogen generation). For events oeyond the present cesign basis, this venting capability wi'l substantially increase tne plant's ability to ceal with large quantities of non-concensible gas without the loss of core cooling or containment integrity, 77 )2hb 274

2. Procedures addressing the use of the RCS ver.ts are requirec Dy January 1,1961. The procedures shoula cefine the conditions under which the vents should De used as well as the conoitions uncer which the vents should not be used. The procecures shoulc be cased nn the followi69 critcrN (1) assurance that tne plant can meet the requirements of 10CFR50.46 and 10CFR50.44 for Design Basis Accicents; and (2) a substantial increase in the plants acility to maintain core cooling and containment integrity for events beyond the Design Basis.

B. BWR Design Considerations -

1. Since the BWR owners group has suggested that the present BWR designs inherent capability of venting, this question relates to tne capability of existing systems. The ability of these systems to vent the RCS of non-condensible gas must be demonstrateo. In adaition the, ability of these systems to meet tne same requirements as the PWR vent systems must be cocumented. Since there are important differences among BWR's, each licensee should accress the specific design features of his plant.

2. In addition to reactor coolant system venting, each BWR licensee shoula address the ability to vent other systems sucn as the isolation condenser, which may be required to maintain acequate core ccoling.

If the production of a large amount of non-condensible gas woulo cause the loss of function of such a system, remote venting of tnat system is required. The qualifications of sucn a venting system should be the same as that required for PWR venting systems.

78 1296 275

C. PWR Vent Design Considerations

1. The locations for PWR Vents are as follows:

a. Each PWR licensee should provice the capaDility to vent the reactor vessel head.

b. The reactor vessel head vent shoula be capaole of venting non-condensible gas from the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and cold legs (through head jets and other leakage paths). Accitional venting cap 3Dility is required for those portions of eacn hot leg which can not be vented through the the reactor vessel head vent. The NRC recognizes that it is impractical to vent each of tne many thousanos of tubes in a U-tube steam generator. however, we Delieve that a procecure can be ceveloped which assures that sufficient liquia or steam can enter tne U-tuce region so tnat cecay heat can De effectively removea from the reactor coolant system. Such a procecure is required by January 1961.

c. Venting of the pressurizer is required to assure its availaoility for system pressure and volume control. Tnese are important considerations especially curing natural circulation.

2. The size of the reactor coolant vents is not a critical issue.

The desired venting capability can De achievec witn vents in a fairly large range of sizes. The criteria for sizing a vent can De developeo in several ways. One approacn, which we consider reasonaole, is to specify a volume of non-concensible gas to De venteo and a venting time i.e., a vent capaole of venting a gas volume of 1/2 the RCS in one hour. Other criteria and engineering approaches shocid De consicered if cesirea.

1296 276 79

9. Since the RCS vent system will be part of the reactor coolant systems bouncary, efforts should be mace to minimize the prooability of an inacvertent actuation of the system. Removing power from the vents is one step in the direction. Other steps are also encouragec.

10. Since tne generation of large quantities of non-concensible gas could De associated with substantial core camage, venting to atmosphere is unacceptable because or the associated released racioactivity. Venting into containment is the only presently available alternative. Within containment those areas which provice gooa mixing with containment air are preferrea. In adoition, areas wnich provide for maximum cooling of the ventec gas are preferred. Therefore the selection of a location for venting should take advantage of existing ventilation anc heat removal systems.

11. The inadvertent opening of an RCS vent must be accressed. For vents smaller than th .vCA definition, leakage detection must be sufficient to identify the leakage. For vents larger than the LOCA definition, an analysis is required to cemonstrate compliance with IUCFR50.4b.

So 1296 277

REACTOR COOLANT SYSTEM VENTING 2.1.9(d) l SEQUOYAH NUCLEAR PLANT RESPONSE

SUMMARY

TVA will provide the capability to vent the reactor vessel head by January 1, 1981. The design for this vent will be made available for NRC review by January 1, 1980.

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.

CLARIFICATION ITEMS A. Procedures for use of the reactor vessel head vent at Sequoyah will be made available to the NRC before January 1,1981.

B. (Not applicable to Sequoyah)

C. PWR Vent Design Consideration

1. a) A reactor vessel head vent will be installed by January 1, 1981, to provide the capability to vent noncondensible gas from the reactor coolant system.

b) No additional vents are required. Natural circulation in the primary system will ensure that sufficient liquid or steam can enter the U-tube region so that decay heat can be effectively removed, c) Venting of the pressurizer is provided as part of the Sequoyah design.

2. Appropriate design considerations will be implemented in design of the reactor vessel head vent.

I 81 1296 278

SHIFT SUPERVISOR RESPONSIBILITIES (2.2.1.a)

POSITION I

1. The highest level of corporate management of each licensee shall issue and pericaically reissue a management airective that empnasizes the primary management responsibility of the shif t supervisor for safe operation of the plant under all conditions on his shif t and that clearly establishes his corrmana cuties.

2. Plant procedures shall be reviewed to assure that the duties, responsi-bilities, and authority - the shif t supervisor and control roca operators are properly defined to effect the establishment of a cefinite line of command and clear delineation of the command cecision authority of the shif t supervisor in the control room relative to other plant management personnel.

Particular emphasis shall be placed on the fol'owing:

a. The responsit flity and authority of the snif t supervisor shall be to maintain the broacest perspective of operational conaitions affecting the 3afety of the plant as a matter of highest priority at all times wnen on duty in t?.e control room. Tne idea snali ce reinforcea that the shif t supervisor snould not occcme totally involved in any single operation in times of emergency wnen multiple operations are required in the control room.

b. The shif t supurvisor, until properly relieved, shall remain in the control room at all times curing accident situaticns to direct the activities of control room operators. Persons authorized to relieve the shif t supervisor shall be specified.

c. If the shif t supervisor is temporarily absent from the control room curing routine operations, a lead control room operator shall oe de:ignated to assume the control room commano function.

Tnese temporary cuties, responsibilities, and authority shall De clearly secifiec.

3. Training programs for shif t supervisors snali emphasize and reinforce the responsibility for safe operation anc the management function tne shif t supervisor is to provice for assuring safety.

4. The administrative cuties of the shif t supervisor shall De reviewed oy the senior officer of each utility responsible for plant operations. Aaministra-tive functions that cetract from or are subordinate to the management responsibility for assuring the safe operation of tne plant shall be celegated to other o::erations perscnnel not on cuty in the control room.

CLARIFICATION The attachment provices clarification to the above position.

1296 279 82

- Attachment i

SHIFT SUPERVISOR RESPONSIBILITY (2.2.1.A)

NUREG-0576 POSITION (POSITION NO.) CLARIFICATION Highest Level of Corporate Management (1.) V. P. For Operations Periodically Reissue (1.) Annual Reinforcement of Company Policy Management Direction (1.) Formal Documentation of Shift Personnel, All Plant Management, Copy to IE Region Properly Defined (2.0) Defined in Writing in a Plant Procecure until Properly Relievec (2.B) Formal Transfer of Autnority, Valic SR9 License, Recorced in Plant Log Temporarily Absent (2.C) Any Aasence Control Room Defined (2.C) Incluces Snif t Supervisor Office Acjacent to the Con' i Room Designatea (2.C) In Acministrative Procecures Clearly Specifiec Defined i t.cministrative Procecures SR0 Training Specifisc in ANS 3.1 (Draft)

Section 5.2.1.o Administrative Duties (4. ) Not Affecting Plant Safety Administrative Duties Reviewea (4.) On Sari.e Interval as deinforcement:

i .e. , Annual by V. P. for Operations.

1296 280 83

SHIFT SUPERVISOR RESPONSIBILITIES (2.2.1.a)

SEQUOYAH NUCIEAR PLANT UNIT 1 RESPONSE

SUMMARY

The requirements are to be implemented by OL for SNP unit 1. The duties of the shif t supervisor, as discussed in NUREG-0578, are performed by the assistant shift engineer on each unit. The V.P. for Operations is the Manager of Power Operetions. SR0 training is specified in the SNP FSAR 13.2 which meets the intent of section 5.2.1.8.

RESPONSE

1. TVA's administrative procedures, shift supervisor job descriptions, and training programs emphasize the primary management responsibility of the shif t engineer. In addition, periodic retraining acts to reinforce his command responsibilities. While these existing measures provide a high level of confidence that the shift supervisor has primary management responsibility for safe operation of the plant, TVA will periodically issue a management directive which emphasizes this assignment of responsibility.

2a. Plant administrative procedures have been reviewed to ensure that they clearly define the authority and responsibilities of each position on shift. The duties and responsibilities of the shift supervisor, as specified in the job description, are consistent with position statement 2a.

2b. The shif t crew in TVA plants consiets of the following: (1) a shift enginecr who has an SRO license and who has overall respon-sibility for the plant when higher level "in-line" management personnel are not present, (2) an assistaat shift engineer (also has an SRO license) for each unit who has supervisory responsibility for all normal, abnormal, and emergency activities on his assigned unit, (3) a unit operator (with an RO licer.se) for each unit, and (4) other personnti as appropriate. The duties of the shift supervisor as discussed in NUREG-0578 are performed by the assistant shift engineer on each unit. For purposes of our responses, we will use the tern assistant shift engineer for shift supervisor.

The assistant shift engineer's normal work station is in the control room, but he periodically makes inspections of plant equipment. He will immediately go to the control room during emergency situations.

He remains in the control room at all tLmes 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) .

1296 281 84

2c. In the event that the assistant shif t engineer (shif t supervisor) is absent, the unit operator will be the lead operator on the unit to which he is assigned. For multiple unit plants, an additional licensed operator will be available in the control complex to act as as assistant to the unit operator in abnormal or emergency situations. The line of command is clearly specified in administra-tive procedures.

3. The shif t engineer and assistant shif t engineers 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 shif t engineer's office on each shift to perform administrative details formerly done by the shif t engineer.

2. Part of the routine "non-management" duties of the assistant shift engineer have been assigned to other employees.

CLARIFICATION Not required.

1296 282

SHIFT TECHNICAL ADVIE ,A (Section 2.2.i.b)

POSITION Each licensee shall provide an on-shif t technical aavisor to tne snif t supervisor.

The shift techrical advisor may serve more than one unit at a multi-unit site if qualified to perform the aavisor function for the virious ur its.

The Shift Technical Aavisor shall have a bachelor's .egree or equivalent in a scientific or engineering discipline and have receit ed specific training in the response and ana',ysis of the plaitt for transients i as accidents. The Snift Technical Advisor shall also receive training in plant design anc layout, incluaing the capabilities of instrumentation anc controls in the control room. The licensee shall assign normal duties to the Shift Technical Advisors that pertain to the engineering aspects of assuring safe operatim of the plant, including the review and evaluation of operating experience.

DISCUSSION The hRC Lessons Learnea Task Force has recommenced the use Of Shif t Technical Adviors (STA) as a methoa of immeciately improving the plant operating staff's capabilities for response to off-normal conditions anc for evaluating operating experience.

In oefining the characteristics of the STA, we have useo the two essential functiuns to be proviced by the STA. These are accicent assessment anc operating experience assessment.

1. Accident Assessment The STA serving the accident assessment function must be decicatea to concern for the safety of tne plant. The STA's duties will be to diagnose off-normal events ano advise the shift supervisor. The duties of the STA should not incluce tne manipulatin of controls or supervision of operators. The STA must be available, in the control room, within 10 minutes of Deing summoned.

The qualifications of the STA shoulo include college level ecucation in engineering and science subjects as well as training in reactor operations Doth norpal and off-normal. Details regarding these qualifications are provided in paragraphs A.1, 2 and 3 of Enclosure 2 to our Septemoer 13, 1979 letter. In addition, the STA serving the accident assessment function must be cognizant of the evaluations performed as part of the operating experience assessment function.

86 1296 283

2. Operatina Experience Assessment The persons serving the opeating experience assessment function must be dedicatec to concei n for the safety of the plant. Their function will be to evaluate plant operations from a safety point of view and snould incluce such assignments as listed on pages A-50 and A-Sl of NUREG-0578. Tneir qualifica-tions are identical to those described previously uncer accicent assessment and collectively this group should provice competence in all technical areas important to safety. It is desirable that this function be performed by onsite personnel.

CLARIFICATION

1. Due to tne similarity in the requirements for dedication to safety, training and onsite location and the desire that the os.ident assessment function be performed by someone whose normal duties involve review of operating experiences, our preferred position is that the same people perform the accicent and operating experience assessment functions Tne performance of these two functions may be split if it can ce demonstratcc the persons assigned the accident assessment role are aware, on a current oasis, of the work Deing done by those reviewing operating experience.

2. To provice assurance that the STA will be dedicated to concern for the safety of the plant, our position has been that STA's must have a clear measure of independence from cuties associatec with the commercial operation of the plant. This would minimize possible cistractions from safety judgements by the demands of commercial operations. We have cetermined that, while desirable, independence from the operations staff of the plant is not necessary to provice this assurance. It is necessary, however, to clearly emphasize the decication to safety associateo witn the STA position both in the STA joo cescription and in the personnel filling this position. It is not acceptaole to assign a person, who is normally the innediate supervisor of the shift supervisor to STA cuties as defined herein.

87 1296 284

3. It is our position that the STA should be available within 10 minutes of Deing summoned anc therefore shoulc be onsite. The onsite STA may De in a cuty status for periods of time longer than one shift, and therefore asleep at some times, if the ten minute availability is assurea. It is preferable to locate those doing the operating experience assessment onsite. The cesirec exposure to the operating plant and contact with the STA (if these functions are to be split) may be aDle to be accomplished by a group, normally stationed offsite, with frequert onsite presence. We do not intend, at tnis time, to specify or advocate a minimum time onsite.

4. The implementation schecule for the STA requirements is to have the STA on duty by January 1,1980, anc to have STAS, who have all completea training require-ments, on duty by January 1,1981. While minimum training requirements have not been specifiec for January 1,19e0, the STAS on cuty by that time shoula enhance the acciaent and operating experience assessment function at the plant.

1296 285 88

SHIFT TECHNICAL ADVISOR (2.2.1.b)

SEQUOYAH NUCLEAR PLANT UNIT 1 RESPONSE

SUMMARY

The shift technical advisor requirements are to be implemented by January 1, 1980, or by initial criticality. The shift technical advisor training will not be complete until January 1981; however, minimum training requirements will be completed.

RESPONSE

TVA will provide an on-shift technical advisor to the shift supervisor to support the diagnosis of off-normal events and to advise the shift supervisor of actions to terminate or mitigate the consequences of such events.

The Shift Technical Advisor will .iave the following qualifications:

(1) additional training in basic engineering principles, (2) extensive training in plant transient and accident response, (3) technical specifi-cation trainir3 with emphasis on the basis for limiting conditions for operat ion, and (4) significant reactor training or systems and operating procet:res.

The duties of the Shif t Technical Advisor will include: (1) control room support in the diagnosis of off-normal events, (2) advice to the shift supervisor to terminate or mitigate the consequences of of f-normal events, (3) make engineering evaluations of plant conditions required for maintenance and testing, and (4) cognizant cf current information disseminated by TVA's operating experience review group.

On each shift, there will be one shif t technical advisor, however, this person will be assigned other duties when his duties as shift technical advisor are not required, provided that his availability is not compromised.

TVA is cptimistic that a substantial portion of the Shift Technical Advisor training may be completed by January 1, 1981.

As an interim policy by January 1, 1980, (1) an additional SRO will be pJaced on each shift to act as Shift Technical Advisor as circumstances require, and a duty engineer shall also be designated on call for advice in support of the shif t technical advisor, or (2) a plant experienced degreed engineer will be placed on shift to act as shift technical advisor.

TVA believes that a multi-disciplined review group is necessary to adequately investigate LER's. TVA's Nuclear Experience Review Panel presently reviews all licensee event reports. When applicable, results of the review will be incorporated in TVA's operator training and requalification programs. In addition, periodic training sessions are conducted for each shift crew. The material covered during these sessions include, but is not limited to, licens a event reports, operator errors, recttt equipment problems, changes to technical specifications, and general plant status. The Shif t Technical Advisors shall have additional responsibilities in being cognizant of the results of the LER review as applied to Browns Ferry.

89 1296 286

CLARIFICATION

1. In addition to the accident assessment function, the shift technical advisor will be cognizant of information determined by the TVA Operating Experience Review Group.

2. The shift technical advisor will be independent of duties that detract from his primary functions or dilute his dedication to these primary functions. Tha shift technical advisor will be an addition to the previously defined operating staff.

3. Although tne shift technical advisor will not be completely trained for his duties by January 1,1980, the STA will be c full-time shif t employee who will be available within 10 minutes of being summoned during any shift.

4. The shift technical advisor will be on duty by January 1, 1980, and trasning requirements will be met by January 1, 1981. The shift technical advisors on duty by January 1,1980, will provide additional accident and operating experience assessment.

90 1296 287

SHIFT AND RELIEF TURNOVER PROCEDURES (2.2.1.c)

, POSITION The licensees snall review anc revise as necessary the plant procecure for snif t anc relief turnover to assure the following:

1, A checklist shall be provided for the oncoming anc offgoing control root operators and the oncoming shift supervisor to complete anc sign. Th' following items, as a minimum, shall be included in the checklist.

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be lir.ted on the checkli st) .

b. Assurance of tne availability and proper alignment of all systems essential to the prevention anc mitigation of operational transients anc accicents by a eneck of the control console.

(what to check anc criteria for acceptacle status snall ce incluced on the checklist);

c. Identification of systems and components that are in a cegracea moce of operation permittea by the Tecnnical Specifications. For such systems anc components, the length of time in the cegracea moce snall be compared witn the Technical Specifications action statement (tnis shall be recorcec as a separate entry on the checklist).

2. Checklists or logs shall De provided for ccmpletion Dy the offgoing anc ongoing auxiliary operators anc technicians. Sucn cnecklists or logs shall include any equipment under maintenance or test that by themselves coula degrace a system critical to the prevention anc mitigation of operational transients and accicents or init. ate an operational transient (what to cneck and criteria for acceptable status shall ce incluced on the checklist);

and

3. A system shall De establishec to evaluate the ef fectiveness of the shift and relief turnover procecure (for example, periccic incepencent verification of system alignments).

CLARIFICATICN No clarification provided.

1296 288

~,

SHIFT AND RELIEF TURNOVER PROCEDURES (2.2.1.c)

SEQUOYAH NUCLEAR PLANT UNIT 1 RESPONSE

SUMMARY

The new shif t and relief turnover procedures have been developed and will be implemented for fuel load.

RESPONSE

TVA will develop and implement shift and relief turnover procedures that will provide assurance that the oncoming shif t possesses adequate knowledge of critical plant status information and system availability. A checklist or similar hard copy will be completed and signed by offgoing anu oncoming shifts at each shift turnover.

This checklist will include critical plant parameters and allowable limits, availability and proper alignment of safety systems, and a listing of safety system components in a de.raded mode along with the length of time in that mode. All shift personnel responsible for the status of critical equipment will have relief checklists for cncoming and offgoing shifts that will include any core cooling equipment under maintenance or test that could degrade a safety system. In addition, a system will be established to evaluate the effectiveness of the turnover procedures.

1296 289 92

CONTROL RCOM ACCESS (2.2.2.a)

POSITION The licensee shall make provisions for limiting access to the control room to those individuals responsible fcr the direct operation of the nuclear power plant (e.g., operations supervisor, snif t supervisor, and control room operators),

to technical advisors who may be requested or requireo to support the operation, and to predesignated NRC personnel. Provisions shall incluce +,he following:

1. Develop and implement an acministrative procedure that establishes the authority ano responsibility of the person in charge of the control room to limit access, and

2. Develop ano implement procecures that establish a clear line of authority ana responsibility in the control room in the event of an eraergency. Tne line of succession for the person in charge of the control room shall be es+.ablished and limited to persons possessing a current senior reactor ope: ator's license. The plan shall clearly define the lines of communication and authority for plant management personnel not in direct commana of operations, inclucing tnose who report to stations outside of the control room.

CLARIFICATION No clarification provided.

93 1296 290

ONSITE TECHNICAL SUPPORT CENTER (TSC) 2.2.2.b POSITION Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and tonsmit plant status to those individuals wno are knowledgeable of and responsible for engineering anc management s;'pport of reactor operations in the event of an accicent. The center shall ne haoitable to the same degree as the control room for pastulated accicent conaitions.

The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center. Records t%t pertain to the as-built conditions and layout of structures, systems anc components shall be readily available to personnel in the TSC.

CLARIFICATION

1. By January 1,1900, each licensee should meet items A-G that follow. Each licensee is encourageo to provice additional upgrading of the TSC (items 2-10) as soon as prc.ctical, but no later than January 1,1981.

A. Establish a TSC and provice a complete description, B. Provide plans and procecures for engineering / management support anc staffing of the TSC, C. Install cedicated communications between the TSC and the control room, near site emergency operations center, anc the NRC, D. Provide monitoring (either portable or pcrmanent' for both airect radiation and airborne radioactive contaminmants. The monitors should provide warning if the radiation levels in the support center are re.acning potentially cangerous levels. The licensee should cesignate action levels to cefine when protective measures should be taken (such as using breathir 3 apparatus anc potassium iocice taciets, or evacuation to the contrcl room),

E. Assimilate or ensure access to Technical Data, including the licensee's best effort to have direct display of plant parameters, necessary for assessment in the TSC, s'!7 0 d' 101 95

CONTROL ROOM ACCESS (2.2.2.a)

SEQUOYAH NUCLEAR PLANT UNIT 1 RESPONSE SUMMAPJ The new proceduras have been developed and will be implemented for fuel load.

RESPONSE

TVA will develop and implement plant specific administrative procedures that establish specific individual authority and responsibility as well as delineate various system or equiprent functions related to controlling A control room personnel access during normal and accident conditions.

access plan will be developed to provide direction to all members of the for safe operation plant staf f to ensure that those persons responsible of the plant are able to perfon effectitrely.

In addition, TVA will develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. These procedures will clearly define the lines of communication and authority for plant management personnel and will ensure that the shift supervisor, his assistant. or senior licensed management personnel are the only plant personnel who have the authority to direct licensed activities of licensed reactor operators.

94 1296 292

F. Develop procedures for performing this accident assessment function from the control room should the TSC become uninhabitable,.ana G. Submit to tne NRC a ..ger range plan for upgrading the TSC to meet all requirements.

2. Location It is recommenced that the TSC be located in close proximity to the control room to ease communications and access to technical information curing an emergency. The center should be located onsite, i.e., within the plant security bouncary. The greater the distance from the CR, the more sophisticated and complete should be the communications and availability of technical information. Consideration shc.ila De given to proviaing key TSC personnel with a means for gaining access to the control room.

3. Physical Size & Staffing The TSC shoulo be large enough to house 25 persons, necessary engineering cata and information displays (TV monitors, recorders, etc.). Eacn licensee should specify staffing levels and disciplines reporting to

.he TSC for emergencies of varying severity.

4 Activation The center shoulc be activated in accordance with the " Alert" level as Gef.ned in the hRC document " Draft Emergency Action Level Guicelines, NUREG-0610" dated SeptenDer,1979, aco currently cut for public comment. :nstrumentation in the TSC should be capable cf proviaing displays of vital plant paraineters from the time the accident began (t = 0 defined as either reactor or turbine trip).

The Shif t Technical Advisor should be consulte0 on the "Notificat; sn of Unusual Event" nc ever, the activation of the TSC is oiscretionary for tnat class of event.

96 k2

5. Instrumentaticn The instrumental. ion to be located in the TSC need im meet safety-grace requirements out sr.oula be qualitatively comparable (as regaras accuracy and reliability) to tnat in the control room. The TSC should have the capability to access and cisplay plant parameters independent from actions in the control room. Careful consiceration should De given to the design of the interface of the TSC instrumentat on to assure that addition of the TSC oill not result in any degrac i zian of the control rocm or other plant functions.

6. Instrumentatic ower Supply The power _epply to the TSC instrumentation need not meet safety-grace requirements, but should De reliable and of a quality compatible witn the TSC instrumentation requirements. To insure continuity of information at the TSC, the power supply provioed should ce continuous once the TSC is activatea. Consiceration shoulc be given to avoic loss of stored cata (e.g., plant computer) due to mcmentary loss of power or switcning transients. If the power supply is provided from a pir.nt safety-relatec power source, careful attention should De give to assure that the capability and reliacility of the safety-relatea power source is not degracea as a result of this modification.

7. Technical Data Each licensee shoula estaolish the technical cata requirements for the TSC, keeping in minc :ne accident assessment function that nas oeen estaclished for those person; reporting to tne TSC during an emerg ncy. As a minimum, 1796 294 97

~.

B drta (historical in acdition to current status) should be available to permit the assessment of:

Plant Safety Systems Parameters for:

. Reactor Coolant System

. Secondary System (PWRs)

. ECCS Systems

. Feedwater & Makeup Systems

. Containmen',

7-Plant Radiciogical Parameters for:

. Reactor Coolant System

. Containment

. Ef fluent Treatment

. Release Paths Offsite Radiological

. Meteorology

. Of fsite Raciation Levels

6. Data Transmissio3 In addition to providing a data transmission link between the TSC and the control room, each licensee s.toulu review current technology as regarcs transmission of tnose parameters identified for TSC cisplay.

Although there is not a recuirement at the present time, eacn licensee should investigate the capability to transmit plant data offsite to the Emergency Operations Center, the NRC, the reactor vencor, etc.

98 1296 295

9. Struct ical Integrity A. The TSC need not be cesignec to seismic Category I requirements.

The center snould be well built in accordance with sounc engineering practice with cue consideration to the effects of natural phenomena that may occur at the site.

B. Since the center need not os designec to tne same stringent requirements as the Control Room, each licensee si.oulc prepare a Dackup plan for responcing to an emergency f rom the control room.

10. Haoitability The licensee shoulc provice protection for the technical support center personnel from raciological nazarcs including cirect raciation anc airoorne contaminants as per General Design Criterion 19 and SRP 6.4.

A. Licensee s.kuld assure that personnel inside the technical support center (TSC) will not receive coses in excess of those specifiec in GDC 19 anc SRP 6.4 (i .e., S Rem wnole bocy ano 30 Rem to tne tnyroic for the curation of the accident). Major sources of radiation snoulo be consicerec.

B. Permane:2 monitoring systems snou;d De provicea to continuously in,cicate radiation dose rates anc airoorne radioactivity concentrations inside the TSC. The monitoring systems shoulu incluce local alarms to warn personnel of adverse concitions. Procecures must ce provicec which will specify appropriate protective actions to De taken in the event that high dose rates or airoorne racioactive concentrations exist.

99 1?96 296

o C. Permanent ventilation systems which include particulate and charcoal filters should be provided. The ventilation systems neea not be qualifiec as ESF systems. The cesign anc testing guicance of Regulatory Guide 1.52 shoula be follcwed except ti.at the systems co not have to be recundant, seismic, instrumented in the control room or automatically activatea. In addition, the HEPA filters need not De tested as specified in Regulatory Guice 1.52 anc the HEPA's do not-have to meet the (A requirements of Appendix B to 10 CFR 50. nowever, spare parts should te reacily availaole anc procecures 1: place for replacing failed components during an accident.

The systems should be designed to operate from the emergency power supply.

O. Dose recuction measures sucn as creatning apparatus and potassium iodice tablets can not be used as a cesign basis for tne TSC in lieu of ventilation cystems with charcoal filters. However, potassium iodice anc breathing apparatus should De availaDie.

1296 297 100

.s.

j..**

.W

\ \ k l g s ', *I' J

<q l t N

< s'a 4 f

f

'. j; .i ) a 4 M.s ,N ',;; '--

x .>.

,.s i

i

/ . l- . . ! ' d (, h '. / ,

! N, \ ,s-, '7 J. '. ! g , j '4 a

,4 6[ 6i '.-l!

6

i. s. M isd .a i t i. .

i

[  ! \, Y . ' ,is s ,I 1, -]g } j *; i h,' e g j '/

6.C gi f

/ '

-( ~ L . a f [.; 'J !.(-

? . 2,<,'.'. (

l 4 s\\ # '

s,

' - -4

} . 1,I.. '. ' J r ' p4 f - 4t r

4- 6-si

• J' ' '

, ( .- -; .*

,, i g - - - ~ - W l' -

i

, , ,g ,,

, ) ,

'- ,..i { ,.,"w - 4 . ,

I f. . . j -

. ,e d' I e. 3 ., p,, ,.*#  ;

. ;, 4

.(.>=~*.1*. #

._y.,___._ .. _

i, , . . ..

.- ' , ,i,, , , , .

. i ., 3  %. .

y

,1j "  : ; L

} t ,; y f'.a{-0 ..

i ;-

'l r '

, I

) .4 ,

S .

,j $,i,) s[,) .j '

!l

, e

( .E . : '

( n, ,

. 7. ., l1.l. .

~ f[ ,*

\ i t. 4j., .  ! l l

.. 1 g

• 1 O".,

, 8 ,-

-(' . F t g l t. 6l .

1

,i / /

>  ; l ' , !' t ' ,? ' , *'/g , , 1 i

' ,u h

~

j l/,P..  ![ '. ., .' \ n'.,

y;y ,.s.

'[t N ' ]'~ '

p . .Ef.y,

,' k  ;

..,'l- s. , -

M t.,

r'4:4 }h ?i .;. { .3 ,

,f

'D ,..s'* 3 (r f t ', fj q ,1 '

.- - R ,,-- --l s. "p/

i

's., ." ,,. { i ,- - *' ;},

? . . , s. .'. ',, ., 't,,3+ } -.. s . .

,e g ~.

i.

• ?,

., ,. ;a ', '.fp

s . {< ? ,

a,  ; ; i-) y I . 3 . J' c' N ,:.

'- -+-.

' s 'f1- -

.'c.--.-__.-',A',, if Q' Te'

' ' . . Y l i,'-

.I,' .$ 't ,, A*}- i*yy. g' .L

'- - - - . M. g,~ ,,, s.

i.

trI

..f 1,

+

j, >I, { d , ,, . t d - , , ., t, .',0>.

1' . 1_4. 4. a. ,.e, ,

M. , .q '. 1.. . . ... . .4, f, 4, ,,

fd"..,Mj'

, s 4,

e. t, :  ?

i C ..'..OJll;Q'*)l } i p f [(M it L f

' .1 : fp r.;6 n'$-Q # ' g I'$ l s

fl L [* N M .I',. c'-,4

'A e 'i t !'I 7 *

~ 'vI b7 '

y s

..:- "IM.- _

f h,3 c m. . .,

y;Q

'4 T' '

r bk_ . . h . 'a,}

s

.., M Mc.. - il , 9 - = ,' (r='d

~i a. .,

a,'. . _'ir M ,s i [t. rd. /. .<,:,0.')/ ' .' m .

MMF' p ",/- p'j ? N- -

e' .lh ._lll '%! k'w..l7.<, ,t[),;m; . fi,p I 'I( y: % d ., ,*

.. .- s . . .

-\ . %j P )Atg4IN U[

q Q].D " , s T NS t_ "r t I t n y,a-- i.$.. -

t I ' h ;,9 f  ? e

' M.g. 4 4s-,k - p i I4 o A- 3 , ,4 ti

a. s- t. 9. I c ?. /j y s a v -

I

-.}e. .- -

. 3 ..- Is84g*L, I g t'

_'-..:u;. ;g ,,,.

w-,--.e.W - W .'.*+ i -4 l e

V. '

t i s,-n @i i. V c g<4 %

.n ,-a.p g M%.es w

j ! ~*  ;";. -'

t-t . j'-

, '* ' C. 06 ;jl dll :s ; k i ;,', , P' P l+

• mv !!

i

. .Is . . -

l l 6' t $ 'd, [E> q d "' '

T-~lt '. g . l ,,, ,

'{  !{-

f 4 = . - ,

< rl- ' I?;\l p. , '.'- t , t h '. 4y.. L ~l. ic , . i-i i -

- )- tai ---.

i .4 ,4 i s .1. v mj  % 4 p--s- ,. gi,. ,3,

,.(

i

a- 4' I I h r:'1 1

le)

'; r, h, -

+' h ---

1' M Q;g ,

IF C i w 3 ' ;'

g 4j ',*,j I 2

A; ' l'L )lp j 7',a-

- g ,I(' l.

* .. i

,;t

.p'..x,. .,

. f- . -6 jd* .

, / - . */ - ' i

'- - f2s l-- - t.

t Y ' g'" ' A' ' ' f 'a 2 3 I q .

1

• h~fl' *'

t , .

' ; ' >..i ?,l1' t s s' 9

, p

'~f}'N.

' l.' l ll -}y

,'l_

l 9 , - , .__ ..

9I s.

k1 - ;, ; ,',s ' s *e. l i

' ,l r t, 4 ; .- I P- i . {. ,

. .s },. .'e e' & u4 e i -

s j 3, p ,., .g , @,

e

--.7,- .-4 . - .

, ,s.; ,,

~

h s' N ~ T,'[ dv ] $ d s'- ~

i y' ^h# [ H, [" ' i w

i-

• )t[ m W - <

t

% w; , u.a ,u,o :. 6 p

,__m e

ti .. o m'p . .m ,0.Cj --*i ij---

• -p-* m g".3

. ' . 7_ yg,. . w.,p '. ' . _

..g __ '. A,- .lv- . . ," -_  !- <

_2.-. '.c__l i'-

a- .m a m , --._2 _., . .

j .

'f,M  % ,! N[. ...EJh.'-s',V @

. p.. _gpi[kg_ ,.: r, .,' .S[.  %'A,$' L.IIIL

.'b M t, '$> e

' t

%g.,t% !

l ' E' n w h .c 1qm=,ubg g W p

% 0[ g:;,

t ..

st

-- %g = => d ,

. kVj j gi

+ '

W,bHW .--- 4m. ;

, %@ g;.I **

. ... ; i - h....%

p

.u 1,,. . . :M.l1 y5 m.

.L. L:, b,'._ .

. _ .,. z ,a t . .

Wp.*

N-

,V i:

$,1

& . . r. -. , , g .,

J ; 9_. ._i;'t .W'..*.**'".7,, . .J b; ,. ," . . g ..

. r_ . , _ - , _4,.',,, 4...

.- _. t_-

g; ., J._ l....- . _ . _ . . .

..'.s,.. . .

, .. p, ,:. ( ,n'd s ;

..( ' ;. \b

,m[_7 . .,jjl s

' . . . y{ [. .Y.I v

., ..~

' . . t[ U {h b..'

6 * [i

, p

'  ; i

. e' l { ;f ,* I }

9

-- .[g \

i 4 i- y

' l 4 ,

.,I

( ..; i

.,t .. .

' i L-4,4,. c.d ' ,. . 3 :h. M ttt t q;: ,,

. t - < I! c: p i r 1 s.

- ~

, 'g 3 b g 'l. .

?! ft +,

N ii! I' 't L'l' . SN I' -f'

't

.@fd> q.f, .; ._ . ..

, 4 3 ,

4

-m .c ._.- . ,,,

a

> . , H, _., .. l. . p .  ;-'

.w , ,

g l

>to '

. 2.1 ; ' ;i 8 l' l 3- .M ' A 2  : .- . , '

l

,c 'l , ' 4 i  ;

he i t' . 6 s d  ;, O ,

[\ y; N ",44M , b qQ } ]' , }

a, . . .. m s _. a vi .

y ..; .

.* ,,.-' l .,'

i',

> k. p g

' i i ,. p t s h.3

)

, 4 I , [.p,ylP t

.- .y  ! ,

., e 'Y s>

, D ', q - " i --- m-l lhl i

t j,I, jy.

4.p i .,

,L..->,

• i, t_

s- f

],

c i

c r.

p ' .}_.,l_,

.ji, r gl.,j_-__Q{. J-j

,i t

y~!- ---. g- - 3 _ ,;".. ,

3 -

c nhI4.w pi-

) _ _ , _ _ ,=x

' iHs 3 . . n, ,g it 1

. . i i ,

m. .I.,,. > 6, k i l. .+l , ea s , ,,,3 .__. ; .- $ 4 .a H .JL

,a - . gg p!

-3q.

ii , l l ,j i.

' 3

[f t .. } }- frg_ ,e.,c#9 ,  ! l' h.

la f:ll l a l I i ', j o "

i L , ! Q{ l ; [f9 i

i ,

' ti .

.. t

.1 i , d: } d g!,  !

j i C ; > i , l, ' {

,; ': 1 j d,

' t "Li li7 ' 8 101 ,! . ,? J.. [ ~ 1 W, {TI ! ,;. g

~.

)

ONSITE TECHNICAL SUPPORT CENTER (TSC) (2.2.2.b)

SEQUOYAH NUCLEAR PLANT UNIT 1 RESPONSE

SUMMARY

The onstie technical support center (TSC) has been ettablished and meets the criteria for the interim TSC, as well as many of the criteria for the perma-nent TSC.

RESPONSE

The onsite technical support center will be established on the same floor as the rain control room (MCR) but outside of the MCR. It will be habitable to the same extent as the MCR and will have ready access to a complete set of as-built drawings. Reliable communications will be provided to the MCR The technical support center will be established before receipt of the operating license.

Prior to January 1, 1981, equipment will be installed in the support center to improve that plant monitoring capability of technical support personnel.

The plant Radiological Emergency Plan will be amended to establish the technical support center and specify the personnel who will staff it in the event of an emergency.

CLARIFICATION 1.A. The TSC description will be provided by January 1, 1980.

1.B. The plans and procedures for engineering / management support and staf fing of the TSC will be provided by December 1, 1979.

1.C. The requested coatmunications betweca the TSC and the control room, the emergency operations center in Chattanooga, and the NRC have been installed, l.D. Portable radiation monitors will be provided for the TSC until permanent monitors are available.

1.E. This iten will be addressed in our submittal of the TSC description.

1.F. Procedures will be provided for performing the accident assessment function from the control room should the TSC become uninhabitable and will be revised as the TSC is upgraded.

1.C. The long-range plan for upgrnding the TSC will be submitted before January 1, 1981.

2. The TSC is located next to the control room (see attached sketch).

1296 299 102

3. The TSC can accommodata 25 persons. Specifics as to physical size and staffing will be provided (see item 1.A,).

4. The TSC activation is defined in the SNP Emergency Plan. The clz.esifi-cation nomenclature of NUREG-0610 will implemented by July 1, 1980. The intent of the NUREG-0610 alert levels are implemented. Instrumentation in the TSC will provided in the submittals on TSC design (see Item 1.A.).

5. Design and capability of instrumentation for the permanent TSC will be provided before January 1, 1981.

6. Design criteria for TSC instrumentation will be provided before January 1, 1981

7. The TSC data requirements will be addressed in a description of the upgraded TSC before January 1, 1981.

8. The TSC data transmission links to offsite centers is being discussed with the Westinghouse Owner's Group for TMI-II.

9.A. The TSC is located in a seismic Category I structure.

9.B. The backup plan for responding to an emergency from the control room will be defined (see item 1.F.).

10.A. The TSC is located in the control room complex. The permanent design meets GDC 19 an-d SRP 6.4 specifications, as well as TVA's radiological health and s:fety requirements.

10.B. Permanent monitoring systems will be provided. Procedures for response to high dose rates will be established by January 1, 1980.

10.C. Ventilation systems will be addressed in the TSA description submittals.

10.D. Dose reduction measures and breathing apparetus will be addressed in the TSA description submittals.

1296 300

~.

'. ~

ONSITE OPERATIONAL SUPPORT CENTER (SECTION 2.2.2.c)

POSITION An area to bedesignated as the onsite operational support center shall De established. It shall be separate from the control room and shall be the place to which the operations support pe'rsonnel will report in an emergency s i tuati on. Communications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center anc to establish the methoos and lines of communication ano management.

CLARIFICATION No clarification provided.

4 e

104

9 9

ONSITE OPERATIOi'AL SUPPORT CENTER SEQUOYAH NUCLEAR PLANT UNIT 1 RESPONSE

SUMMARY

The operational support center has been established and meets the criteria.

RESPONSE

The operational support center with communications to the main control room has been established. This center is located on the control room floor next to the shift engineers office.(see figure in response to item 2.2.2.b.). The plant Radiological Emergency Plan will be revised by January 1, 1980, to reflect this center and to specify the personnel who will report in the event of an emergency.

I 9b 105