NRC Generic Letter 1979-56

NRC Generic Letter 1979-056: Discussion of Lessons Learned Short Term Requirements

ML031320403
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Indian Point, Kewaunee, Saint Lucie, Point Beach, Oyster Creek, Cooper, Pilgrim, Arkansas Nuclear, Three Mile Island, Prairie Island, Brunswick, Surry, North Anna, Turkey Point, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Duane Arnold, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Fort Calhoun, FitzPatrick, Fort Saint Vrain, Trojan
Issue date:	10/30/1979
From:	Denton H R Office of Nuclear Reactor Regulation
To:
References
NUREG-0578 GL-79-056, NUDOCS 7911190148
Download: ML031320403 (75)

UNITED STATES0 NUCLEAR REGULATORY COMMISSIONaWASHINGTON. D. C. 20555 (1 79f 4October 30, 1979(TO ALL OPERATING NUCLEAR POWER PLANTS)Gentlemen:SUBJECT: DISCUSSION OF LESSONS LEARNED SHORT TERM REQUIREMENTSOn September 13, 1979, a letter was issued to each power reactor licenseewhich defined a set of "short term" requirements resulting from the NRCstaff investigations of the TMI accident. Since the letter was issued, thestaff has attempted to further define these requirements. During theweek of September 24, 1979, seminars were held in four regions of thecountry to encourage industry feedback and dialogue on each short termrequirement. As a result of these discussions, four topical meetings wereheld in Bethesda to discuss certain issues in further detail.Enclosure 1 provides additional clarification of the NRC staff requirements.It should be noted that the intent of these requirements have not changedthroughout this process and are restated in Enclosure 1.Enclosure 2 is a chart of the NUREG-0578 items and their correspondingimplementation schedules. The chart indicates which of the items requireprior NRC review and approval and those for which post implementation NRCreview is acceptable.For those items requiring prior NRC approval, your design details should besubmitted in a timely manner so that this approval and your implementation ofthe item can be completed by the required date. For those items which do notrequire prior NRC approval, you must document your method of implementation bythe required completion date. These schedules assume that your methods are incomplete agreement with the staff's requirements as previously documented inNUREG-0578, our September 13, 1979 letter, and clarified herein. Where yourmethods are not in complete agreement with the staff's requirements, a detaileddescription of your proposed methods along with justification for the differences,is required. Please provide this description and justification as soon aspossible but no later than 15 days following receipt of this letter.mPg7?///f/ffeo Is-2 -October 30, 1979The schedule for completing each of the short term TMI followup requirementsis firm. Some licensees, in responding to our September 13, 1979 letter,have indicated an inability to meet the established implementation schedule.If your response was in this category you are requested to reconsider yourimplementation schedule with the purpose of improving your implementationdates to meet those required by the staff. Within fifteen days from receiptof this letter, you are requested to submit your revised schedule forimplementation. If you are unable to commit to meeting any of theJanuary 1, 1980 requirements, you must provide, for each item, a report onthe degree of compliance expected on January 1, 1980, and a detailedjustification for the delay.Sincerely,Harold R. Denton, DirectorOffice of Nuclear Reactor RegulationEnclosures:1. Discussion of TMI Lessons LearnedShort Term Requirements2. Implementation Schedule A-.Enclosure 1DISCUSSION OFTMI LESSIONS LEARNEDSHORT TERM REQUIREMENTS

IITABLE OF CONTENTSSection2.1.12.1.22.1.3.a2.1.3.b2.1.42.1.5.a2.1.5.c2.1.6.a2.1.6.b2.1.7.a2.1.7.b2.1.8.a2.1.8.bTables2.1.8.b.12.1.8.b.22.1.8.b.3Title PageEmergency Power Supply-Pressurizer Heaters ..... 1-Pressurizer Level and Relief B lock Valvess. ...3...............3Performance Testing for BWR and PWR Relief and ..... ... 5Safety ValvesDirect Indication of Power-Operated Relief Valve .. .7Valve and Safety Valve Position for PWRs and BWRsInstrumentation for Detection of Inadequate Core Cooling-Subcooling Meter., .................................. .9-Additional Instrumentation ............. .... .... 13Diverse Containment Isolation ....... ....... .15Dedicated H2 Control Penetrations ....16Capability to Install'Hydrogen Recombiner at each ...... 17Light Water Nuclear Power PlantIntegrity of Systems Outside Contaimment Likely ........ 18to Contain Radioactive Materials for PWRs andBWRsDesign Review of Plant Shielding and Environmental ......19Qualification of Equipment for Spaces/SystemsWhich May Be Used in Post Accident OperationsAuto Initiation of the Auxiliary Feedwater Systems.... 22(AFSW)Auxiliary Feedwater Flow Indication to Steam .......... 24GeneratorsPost-Accident Sampling Capability ................. ... 26Increased Range of Radiation Monitors .... .. 31Interim Procedures for Quantifying High Level ......... 37Accidental Radioactivity ReleasesHigh Range Effluent Monitor ........................... 38High Range Containment Radiation Monitor ... 39 TABLE OF CONTENTS(continued)Section Title Page2.1.8.c Improved In-Plant Iodine Instrumentation Under ......... 40Accident Conditions2.1.9 Transient and Accident Analysis ...42Containment Pressure Indication (ACRS) ...43Containment Water Level Indication (ACRS) .............. 44Containment Hydrogen Indication (ACRS) ...45Reactor Coolant System Venting (NRR) ...462.2.1.a Shift Supervisor Responsibilities ...................... 512.2.1.b Shift Technical Advisor ...... 532.2.1.c Shift and Relief Turnover Procedures ...562.2.2.a Control Room Access ................. ........................ 572.2.2.b Onsite Technical Support Center ........................ 582.2.2.c Onsite Operational Support Center ...................... 64Enclosure 2Implementation Schedule ................................ 65Analyses and Training Schedule ........................ 70

EMERGENCY POWER SUPPLY (2.1.1)Pressurizer HeatersPOSITIONConsistent with satisfying the requirements of General Design Criteria 10, 14, 15,17 and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power,the following positions shall be implemented:Pressurizer Heater Power Supply1. The pressurizer heater power supply design shall provide thecapability to supply, from either the offsite power source orthe emergency power source (when offsite power is not available),a predetermined number of pressurizer heaters and associatedcontrols necessary to establish and maintain natural circula-tion at hot standby conditions. The required heaters and theircontrols shall be connected to the emergency buses in a mannerthat will provide redundant power supply capability.2. Procedures and training shall be established to make the operatoraware of when and how the required pressurizer heaters shall beconnected to the emergency buses. If required, the proceduresshall identify under what conditions selected emergency loadscan be shed from the emergency power source to provide sufficientcapacity for the connection of the pressurizer heaters.3. The time required to accomplish the connection of the preselectedpressurizer heater to the emergency buses shall be consistentwith the timely initiation and maintenance of natural circulationconditions.4. Pressurizer heater motive and control power interfaces with theemergency buses shall be accomplished through devices that havebeen qualified in accordance with safety-grade requirements.CLARIFICATION1. In order not to compromise independence between the sources ofemergency power and still provide redundant capability to provideemergency power to the pressurizer heaters, each redundant heateror group of heaters should have access to only one Class 1E divisionpower supply.-1 -

2. The number of heaters required to have access to each emergencypower source is that number required to maintain natural circula-tion in the hot standby condition.3. The power sources need not necessarily have the capacity to providepower to the heaters concurrent with the loads required for LOCA.4. Any change-over of the heaters from normal offsite power to emergencyonsite power is to be accomplished manually in the control room.5. In establishing procedures to manually reload the pressurizer heatersonto the emergency power sources, careful consideration must be given to:a. Which ESF loads may be appropriately shed for a given situtation.b. Reset of the Safety Injection Actuation Signal to permit theoperation of the heaters.c. Instrumentation and criteria for operator use to prevent overload-ing a diesel generator.6. The Class IE interfaces for main power and control power are to beprotected by safety-grade circuit breakers. (See also Reg. Guide 1.75)7. Being non-Class IE loads, the pressurizer heaters must be automaticallyshed from the emergency power sources upon the occurrence of a safetyinjection actuation signal. (See item 5.b. above)-2 -

Emergency Power Supply (2.1.1)Pressurizer Level and Relief Block ValvesPOSITIONConsistent with satisfying the requirements of General Design Criteria 1U, 14, 1l17 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 PressurizerLevel Indicators1. Motive and control components of the power-operated relief valves(PORYs) shall be capable of being supplied from either the offsitepower source or the emergency power source when the offsite power isnot available.2. Motive and control components associated with the PORV block valvesshall be capable of being supplied from either the offsite powersource or the emergency power source when the offsite power is notavailable.3. Motive and control power connections to the emergency buses forthe PORVs and their associated block valves shall be through devicesthat have been qualified in accordance with safety-grade requirements.4. The pressurizer level indication instrument channels shall be poweredfrom the vital instrument buses. The buses shall have the capabilityof being suppied from either the offsite power source or theemergency power source when offsite power is not available.CLARIFICATION1. While the prevalent consideration from TMI Lessons Learned is being ableto close the PORY/block valves, the design should retain, to the extentpractical, the capability to open these valves.2. The motive and control power for the block valve should be supplied froman emergency power bus different from that which supplies the PORY.3. Any changover of the PORV and block valve motive and control power fromthe normal offsite power to the emergency onsite power is to beaccomplished manually in the control room.-3 -

4. For those designs where instrument air is needed for operation,the electrical power supply requirement should be capable of beingmanually connected to the emergency power sources.-4 -

PERFORMANCE TESTING FOR BWR AND PWR RELIEF AND SAFETY VALVES (2.1.2)POSITIONPressurized Water Reactor and Boiling Water Reactor licensees and applicantsshall conduct testing to qualify the reactor coolant system relief and safetyvalves under expected operating conditions for design basis transients andaccidents.CLARIFICATION1. Expected operating conditions can be determined through the use of analysisof accidents and anticipated operational occurrences referenced inRegulatory Guide 1.70.2. This testing is intended to demonstrate valve operability under variousflow conditions, that is, the ability of the valve to open and shutunder the various flow conditions should be demonstrated.3. Not all valves on all plants are required to be tested. The valve testingmay be conducted on a prototypical basis.4. The effect of piping on valve operability should be included in the testconditions. Not every piping configuration is required to be tested, butthe configurations that are tested should produce the appropriate feedbackeffects as seen by the relief or safety valve.5. Test data should include data that would permit an evaluation of dischargepiping and supports if those components are not tested directly.-5 -

6. A description of the test program and the schedule for testing should besubmitted by January 1, 1980.7. Testing shall be complete by July 1, 1981.-6 -

\-..-DIRECT INDICATION OF POWER-OPERATED RELIEFVALVE AND SAFETY VALVE POSITION FOR PWRs AND BWRs (2.1.3.a)POSITIONReactor System relief and safety valves shall be provided with a positiveindication in the control room derived from a reliable valve positiondetection device or a reliable indication of flow in the discharge pipe.CLARIFICATION1. The basic requirement is to provide the operator with unambiguousindication of valve position (open or closed) so that appropriateoperator actions can be taken.2. The valve position should be indicated in the control room. An alarm shouldbe provided in conjunction with this indication.3. The valve position indication may be safety grade. If the positionindication is not safety grade, a reliable single channel direct indica-tion powered from a vital instrument bus may be provided if backupmethods of determining valve position are available and are discussed inthe emergency procedures as an aid to operator diagnosis and action.4. The valve position indication should be seismically qualified consistentwith the component or system to which it is attached. If the seismicqualification requirements cannot be met feasibly by January 1, 1980,a justification should be provided for less than seismic qualificationand a schedule should be submitted for upgrade to the required seismicqualificiation.-7 -

k-5. The position indication should be qualified for its appropriate environment(any transient or accident which would cause the relief or safetyvalve to lift). If the environmental qualification program for thisposition indication will not be completed by January 1, 1980, aproposed schedule for completion of the environmental qualificationprogram should be provided.-8 -

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)SUBCOOLING METERPOSITIONLicensees shall develop procedures to be used by the operator to recognizeinadequate core cooling with currently available instrumentation. The licenseeshall provide a description of the existing instrumentation for the operatorsto use to recognize these conditions. A detailed description of the analysesneeded to form the basis for operator training and procedure development shall beprovided pursuant to another short-term requirement, "Analysis of Off-NormalConditions, Including Natural Circulation " (see Section 2.1.9 of NUREG-0578)In addition, each PWR shall install a primary coolanton-line indication of coolant saturation condition.use of this meter shall include consideration that isother related plant parameters.saturation meter to provideOperator instruction as tonot to be used exclusive ofCLARIFICATION1. The analysis and procedures addressed in paragraph one above willreviewed and should be submitted to the NRC "Bulletins and OrdersTask Eorce" for review.2. The purpose of the subcooling meter is to provide a continuous indicationof margin to saturated conditions. This is an important diagnostic toolfor the reactor operators.3. Redundant safety grade temperature input from each hot leg (or use of multiplecore exit in T/CVs) are required.4. Redundant safety grade system pressure measures should be provided.5. Continuous display of the primary coolant saturation conditions should beprovided.-9 -

6. Each PWR should have: (A.) Safety grade calculational devices anddisplay (minimum of two meters) or (B.) a highly reliable singlechannel environmentally qualified, and testable system plus a backupprocedure for use of steam tables. If the plant computer is to beused, its availability must be documented.7. In the long term, the instrumentation qualifications must be requiredto be upgraded to meet the requirements of Regulatory Guide 1.97(Instrumentation for Light Water Cooled Nuclear Plants to Assess PlantConditions During and Following an Accident) which is under development.8. In all cases appropriate steps (electrical, isolation, etc.) must be takento assure that the addition of the subcooling meter does not adverselyimpact the reactor protection or engineered safety features systems.9. The attachment provides a definition of information required on thesubcooling meter.-10 -

\- /INFORMATION REQUIRED ON THE SUBCOOLING METERDisplayInformation Displayed (T-Tsat, Tsat, Press, etc.)Display Type (Analog, Digital, CRT)Continuous or on DemandSingle or Redundant DisplayLocation of DisplayAlarms (include setpoints)Overall uncertainty ('F, PSI)Range of DisplayQualifications (seismic, environmental, IEEE323)CalculatorType (process computer, dedicated digital or analog calc.)If process computer is used specify availability. (% of time)Single or redundant calculatorsSelection Logic (highest T., lowest press)Qualifications (seismic, environmental, IEEE3Z3)Calculational Technique (Steam Tables, Functional Fit, ranges)InputTemperature (RTD's or T/C's)Temperature (number of sensors and locations)Range of temperature sensors-.--11 -

IUncertainty* of temperature sensors (°F at 1 )Qualifications (seismic, environmental, IEEE323)Pressure (specify instrument used)Pressure (number of sensors and locations)Range of Pressure sensorsUncertainty* of pressure sensors (PSI at 1 )Qualifications (seismic, environmental, IEEE323)Backup CapabilityAvailability of Temp & PressAvailability of Steam Tables etc.Training of operatorsProcedures*Uncertainties must address conditions of forced flow and natural circulation-12 -

INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (2.1.3.b)ADDITIONAL INSTRUMENTATIONPOSITIONLicensees shall provide a decription of any additional instrumentation or controls(primary or backup) proposed for the plant to supplement those devices cited in thepreceding section giving an unambiguous, easy-to-interpret indication of inadequatecore cooling. A description of the functional design requirements for the systemshall also be included. A description of the procedures to be used with theproposed equipment, the analysis used in developing these procedures, and a schedulefor installing the equipment shall be provided.CLARIFICATION1. Design of new instrumentation should provide an unambiguous indication ofinadequate core cooling. This may require new measurements to or a synthesisof existing measurements which 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 advantages ofvarious instruments to monitor water level and core cooling is required inthe response to the September 13, 1979 letter.4. The indication of inadequate core cooling must be unambiguous, in that, it shouldhave the following properties:a) it must indicate the existence of inadequate core cooling caused by variousphenomena (i.e., high void fraction pumped flow as well as stagnantboil off).b) it must not erroneously indicate inadequate core cooling because of thepresence of an unrelated phenomenon.-13 -

5. The indication must give advanced warning of the approach of inadequatecore cooling.6. The indication must cover the full range from normal operation tocomplete core uncovering. For example, if water level is chosenas the unambiguous indication, then the range of the instrument (orinstruments) must cover the full range from normal water level tothe bottom of the core .-14 -

CONTAINMENT ISOLATION (2.1.4)POSITION1. All containment isolation system designs shall comply with therecommendations of SRP 6.2.4; i.e., that there be diversity in theparameters sensed for the initiation of containment isolation.2. All plants shall give careful reconsideration to the definition ofessential and non-essential systems, shall identify each systemdetermined to be essential, shall identify each system determined tobe non-essential, shall describe the basis for selection of eachessential system, shall modify their containment isolation designsaccordingly, and shall report the results of the re-evaluation to NRC.3. All non-essential systems shall be automatically isolated by thecontainment isolation signal.4. The design of control systems for automatic containment isolation valvesshall be such that resetting the isolation signal will not result in theautomatic reopening of containment isolation valves. Reopening ofcontainment isolation valves shall require deliberate operator action.CLARIFICATION1. Provide diverse containment isolation signals that satisfy safety-graderequirements.2. Identify essential and non-essential systems and provide results to NRC.3. Non-essential systems should be automatically isolated by containmentisolation signals.4. Resetting of containment isolation signals shall not result in theautomatic loss of containment isolation-15- DEDICATED H2 CONTROL PENETRATIONS (2.1.5.a)POSITIONPlants using external recombiners or purge systems for post-accident combustiblegas control of the containment atmosphere should provide containment isolationsystems for external recombiner or purge systems that are dedicated to that serviceonly, that the redundancy and single failure requirements of General DesignCriterion 54 and 56 of Appendix A to 10 CFR 50, and that are sized to satisfy theflow requirements of the recombiner or purge system.CLARIFICATION1. This requirement is only applicable to those plants whose licensing basisincludes requirements for external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere.2. An acceptable alternative to the dedicated penetration is a combined designthat 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 combined single-failure proof alternativeshould be sized such that the flow requirements for the use of therecombiner or purge system are satisfied.4. Components necessitated by this requirement should be safety grade.5. A description of required design changes and a schedule for accomplishingthese changes should be provided by January 1, 1980. Design changesshould be completed by January 1, 1981.-16 -

CAPABILITY TO INSTALL HYDROGEN RECOMBINERAT EACH LIGHT WATER NUCLEAR POWER PLANT (2.1.5.c)POSITIONThe procedures and bases upon which the recombiners would be used on allplants should be the subject of a review by the licensees in consideringsheilding requirements and personnel exposure limitations as demonstratedto be necessary in the case of TMI-2.CLARIFICATION1. This requirement applies only to those plants that included HydrogenRecombiners as a design basis for licensing.2. The shielding and associated personnel exposure limitations associatedwith recombiner use should be evaluated as part of licensee responseto requirement 2.1.6.B, uDesign review for Plant Shielding."3. Each licensee should review and upgrade, as necessary, those criteriaand procedures dealing with recombiner use. Action taken on thisrequirement should be submitted by January 1, 1980.-17 -

INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELYTO CONTAIN RADIOACTIVE MATERIALS FOR PWRs AND BWRs (2.1.6.a)POSITIONApplicants and licensees shall immediately implement a program to reduceleakage from systems outside containment that would or could contain highlyradioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:1. Immediate Leak Reductiona. Implement all practical leak reduction measures for all systemsthat could carry radioactive fluid outside of containment.b. Measure actual leakage rates with system in operation and reportthem to the NRC.2. Continuing Leak ReductionEstablish and implement a program of preventive maintenance to reduceleakage to as-low-as-practical levels. This program shall includeperiodic integrated leak tests at intervals not to exceed each refuelingcycle.CLARIFICATIONLicensees shall, by January 1, 1980, provide a summary description of theirprogram to reduce leakage from systems outside containment that would orcould contain highly radioactive fluids during a serious transient oraccident. Examplesof such systems are given on page A-26 of NUREG-0578.Other examples include the Reactor Core Isolation Cooling and Reactor Water CleaCleanup (Letdown function) Systems for BWRs. Include a list of systems whichare excluded from this program. Testing of gaseous systems should includehelium leak detection or equivalent testing methods. Consider in your programto reduce leakage potential release paths due to design and operator deficienciesas discussed in our letter to you regarding North Anna and Related Incidents datedOctober 17, 1979.-18 -

N-;DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTALQUALIFICATION OF EQUIPMENT FOR SPACES/SYSEMS WHICHMAY BE USED IN POST ACCIDENT OPERATIONS (2.1.6.b)POSITIONWith the assumption of a post-accident release of radioactivity equivalentto that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalentof 50% of the core radioiodine, 100% of the core noble gas inventory,and1% of the core solids, are contained in the primary coolant), eachlicensee shall perform a radiation and shielding design review of thespaces around systems that may, as a result of an accident, contain highlyradioactive materials. The design review should identify the locationof vital areas and equipment, such as the control room, radwaste controlstations, emergency power supplies, motor control centers, and instrumentareas, in which personnel occupancy may be unduly limited or safety equipmentmay be unduly degraded by the radiation fields during post-accident operationsof these systems.Each licensee shall provide for adequate access to vital areas and protectionof safety equipment by design changes, increased permanent or temporaryshielding, or post-accident procedural controls. The design review shalldetermine which types of corrective actions are needed for vital areasthroughout the facility.CLARIFICATIONAny area which will or may require occupancy to permit an operator toaid in the mitigation of or recovery from an accident is designated asa vital area. In order to assure that personnel can perform necessarypost-accident operations in the vital areas, we are providing the followingguidance to be used by licensees to evaluate the adequacy of radiationprotection to the operators:1. Source TermThe minimum radioactive source term should be equivalent to thesource terms recommended, in Regulatory Guides 1.3, 1.4, 1.7and Standard Review Plant 15.6.5. with appropriate decay timesbased on plant design.-19 -

a. Liquid Containing Systems: lOOt of the core equilibriumnoble gas inventory, 50% of the core equilibrium halogeninventory and 1X of all others are assumed to be mixedin the reactor coolant and liquids injected by HPCI andLPCI.b. Gas Containing Systems: l00 of the core equilibrium noblegas inventory and 25X of the core equilibrium halogenactivity are assumed to be mixed in the containment atmosphere.For gas containing lines connected to the primary system (e.g.,BWR steam lines) the concentration of radioactivity shall bedetermined assuming the activity is contained in the gas spacein the primary coolant system.2. Dose Rate CriteriaThe dose rate for personnel in a vital area should be such thatthe guidelines of GDC 19 should not be exceeded during the courseof the accident. GDC 19 limits the dose to an operator to 5 Rem wholebody or its equivalent to any part of the body. When determining thedose to an operator, care must be taken to determine the necessaryoccupancy time in a specific area. For example, areas requiringcontinuous occupancy will require much lower dose rates than areaswhere minimal occupancy is required. Therefore, allowable dose rateswill be based upon expected occupancy, as well as the radioactive sourceterms and shielding. However, in order to provide a general designobjective, we are providing the following dose rate criteria-20 -

with alternatives to be documented on a case-by-case basis.The recommended dose rates are average rates in the area. Local hotspots may exceed the dose rate guidelines provided occupancy is notrequired at the location of the hot spot. These doses are designobjectives and are not to be used to limit access in the event ofan accident.a. Areas Requiring Continuous Occupancy: <15mr/hr. These areaswill require full time occupancy during the course of theaccident. The Control Room and onsite technical supportcenter are areas where continuous occupancy will be required.The dose rate for these areas is based on the control roomoccupancy factors contained in SRP 6.4.b. Areas Requiring Infrequent Access: GDC 19. These areasmay require access on a regular basis, but not continuousoccupancy. Shielding should be provided to allow access ata frequency and duration estimated by the licensee. The plantRadiochemical/Chemical Analysis Laboratory, radwaste panel,motor control center, instrumentation locations, and reactorcoolant and containment gas sample stations are exampleswhere occupancy may be needed often but not continuously._ 21 -t AUTO INITIATION OF THE AUXILIARYFEEDWATER SYSTEM (AFWS) (2.1.7.a)POSITIONConsistent with satisfying the requirements of General Design Criterion 20 ofAppendix A to 10 CFR Part 50 with respect to the timely initiation of theauxiliary feedwater system, the following requirements shall be implemented inthe short term:1. The design shall provide for the automatic initiation of theauxiliary feedwater system.2. The automatic initiation signals and circuits shall be designed sothat a single failure will not result in the loss of auxiliaryfeedwater system function.3. Testability of the initiating signals and circuits shall be afeature of the design.4. The initiating signals and circuits shall be powered from theemergency buses.5. Manual capability to initiate the auxiliary feedwater systemfrom the control room shall be retained and shall be implementedso that a single failure in the manual circuits will not result inthe loss of system function.6. The a-c motor-driven pumps and valves in the auxiliary feedwatersystem shall be included in the automatic actuation (simultaneousand/or sequential) of the loads onto the emergency buses.7. The automatic initiating signals and circuits shall be designed sothat their failure will not result in the loss of manual capabilityto initiate the AFWS from the control room.In the long term, the automatic initiation signals and circuits shall be upgradedin accordance with safety-grade requirements.-22 -;

CLARIFICATIONControl Grade (Short-Term)1. Provide automatic/manual initiation of AFWS.2. Testability of the initiating signals and circuits is requied.3. Initiating signals and circuits shall be powered from the emergency buses.4. Necessary pumps and valves shall be included in the automatic sequence ofthe loads to the emergency buses. Verify that the addition of these loadsdoes not comprimise the emergency diesel generating capacity.5. Failure in the automatic circuits shall not result in the loss of manualcapability to initiate the AFWS from the control room.6. Other Considerationsa. For those designs where instrument air is needed for operation,the electric power supply requirement should be capable of beingmanually connected to emergency power sources.-23 -

I7AUXILIARY FEEDWATER FLOW INDICATIONTO STEAM GENERATORS (2.l.7.b)POSITIONConsistent with satisfying the requirements set forth in GDC 13 to provide thecapability in the control room to ascertain the actual performance of the AFWSwhen it is called to perform its intended function, the following requirementsshall be implemented:1. Safety-grade indication of auxiliary feedwater flow to each steamgenerator shall be provided in the control room.2. The auxiliary feedwater flow instrument channels shall be poweredfrom the emergency buses consistent with satisfying the emergencypower diversity requirements of the auxiliary feedwater system setforth in Auxiliary Systems Branch Technical Position 10-1 of theStandard Review Plan, Section 10.4.9.CLARIFICATIONA. Control Grade (Short-Term)1. Auxiliary feedwater flow indication to each steam generator shallsatisfy the single failure criterion.2. Testability of the auxiliary feedwater flow indication channels shallbe a feature of the design.3. Auxiliary-feedwater flow instrument channels shall be powered fromthe vital instrument buses.B. Safety-Grade (Long-Term)1. Auxiliary feedwater flow indication to each steam generator shallsatisfy safety-grade requirements.C. Other1. For the Short-Term the flow indication channels should by themselvessatisfy the single failure criterion for each steam generator. As-24 N-a fall-back position, one auxiliary feed water flow channel may bebacked up by a steam generator level channel.2. Each auxiliary feed water channel should provide an indication offeed flow with an accuracy on the order of + 10X.-25 -

IMPROVED POST-ACCIDENT SAMPLING CAPABILITY (2.1.8.a)POSITIONA design and operational review of the reactor coolant and containmentatmosphere sampling systems shall be performed to determine the capabilityof personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accidentconditions without incurring a radiation exposure to any individual inexcess of 3 and 18 3/4 Rems to the whole body or extremities, respectively.Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release offission products. If the review indicates that personnel could not promptlyand safely obtain the samples, additional design features or shieldingshould be provided to meet the criteria.A design and operational review of the radiological spectrum analysisfacilities shall be performed to determine the capability to promptlyquantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radioisotopes that are indicators ofthe degree of core damage. Such radionuclides are noble gases (whichindicate cladding failure), iodines and cesiums (which indicate highfuel temperatures), and non-volatile isotopes (which indicate fuel melting).The initial reactor coolant spectrum should correspond to a RegulatoryGuide 1.3 or 1.4 release. The review should also consider the effects ofdirect radiation from piping and components in the auxiliary building andpossible contamination and direct radiation from airborne effluents. Ifthe review indicates that the analyses required cannot be performed in aprompt manner with existing equipment, then design modifications or equipmentprocurement shall be undertaken to meet the criteria.In addition to the radiological analyses, certain chemical analyses arenecessary for monitoring reactor conditions. Procedures shall be providedto perform boron and chloride chemical analyses assuming a highly radio-active initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analysesshall be capable of being completed promptly; i.e., the boron sample analysiswithin an hour and the chloride sample analysis within a shift.DISCUSSIONThe primary purpose of implementing Improved Post-Accident Sampling Capabilityis to improve efforts to assess and control the course of an accident by:1. Providing information related to the extent of core damage thathas occurred or may be occurring during an accident;2. Determining the types and quantities of fission products released tothe containment in the liquid and gas phase and which may be releasedto the environment;-26 -

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 reactorcoolant liquid samples;2. Radiological and hydrogen analyses of containment atmosphere (air) samples.CLARIFICATIONThe licensee shall have the capability to promptly obtain (in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)pressurized and unpressurized reactor coolant samples and a containmentatmosphere (air) sample.The licensee shall establish a plan for an onsite radiological and chemicalanalysis facility with the capability to provide, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of obtainingthe 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 U to10 volume percent,3. dissolved gases (i.e., H2 ' 02) and boron concentration ofliquids.or have in-line monitoring capabilities to perform the above analysis.Plant procedures for the handling and analysis of samples, minor plantmodifications for taking samples and a design review and procedural modifi-cations (if necessary) shall be completed by January 1, 1980. Major plantmodifications shall be completed by January 1, 1981.During the review of the post accident sampling capability consideration shouldbe given to the following items:-27 -

01. Provisions shall be made to permit containment atmosphere samplingunder 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 ordistortion, for preventing blockage of sample lines by loosematerial in the RCS or containment, for appropriate disposal ofthe samples, and for passive flow restrictions to limit reactorcoolant loss or containment air leak from a rupture of the sampleline.3. If changes or modifications to the existing sampling system arerequired, the seismic design and quality group classification orsampling lines and components shall conform to the classification ofthe system to which each sampling line is connected. Components andand piping downstream of the second isolation valve can bedesigned to quality Group D and nonseismic Category I require-ments.The licensee's radiological sample analysis capability should includeprovisions to:a. Identify and quantify the isotopes of the nuclide categories discussedabove to levels corresponding to the source terms given in LessonsLearned Item 2.1.6.b. Where necessary, ability to dilute samplesto provide capability for measurement and reduction of personnel_ 28 -

exposure, should be provided. Sensitivity of onsite analysis capabilityshould be such as to permit measurement of nuclide concentrationin the range from approximately I uCi/gm to the upper levels indicatedhere.b. Restrict background levels of radiation in the radiological andchemical analysis facility from sources such that the sample analysiswill provide results with an acceptably small error (approximatelya factor of 2). This can be accomplished through the use of sufficientshielding around samples and outside sources, and by the use ofventilation system design which will control the presence of airborneradioactivity.c. Maintain plant procedures which identify the analysis required,measurement techniques and provisions for reducing background levels.The licensees chemical analysis capability shall consider the presence of theradiological source term indicated for the radiological analysis.In performing the review of sampling and analysis capability, consideration shallbe given to personnel occupational exposure. Procedural changes and/or plantmodifications must assure that it shall be possible to obtain and analyze asample while incurring a radiation dose to any individual that is as low asreasonably achievable and not in excess of GDC 19. In assuring that theselimits are met, the following criteria will be used by the staff.1. For shielding calculations, source terms shall be as given in LessonsLearned Item 2.1.6.b.-29 -

-2. Access to the sample station and the radilogical and chemical analysisfacilities shall be through areas which are accessible in post accidentsituations and which are provided with sufficient shielding to assure thatthe radiation dose criteria are met.3. Operations in the sample station, handling of highly radioactive samplesfrom the sample station to the analysis facilities, and handling whileworking with the samples in the analysis facilities shall be such thatthe radiation dose criteria are met. This may involve sufficientshielding of personnel from the samples and/or the dilution of samplesfor analysis. If the existing facilities do not satisfy these criteria,then additional design features, e.g., additional shielding, remotehandling etc. shall be provided. The radioactive sample lines in thesample station, the samples themselves in the analysis facilities, andother radioactive lines of the vicinity of the sampling station andanalysis facilities shall be included in the evaluation.4. High range portable survey instruments and personnel dosimeters shouldbe provided to permit rapid assessment of high exposure rates andaccumulated personnel exposure.The licensee shall demonstrate their capability to obtain and analyze a samplecontaining the isotopes discussed above according to the criteria given inthis section.-30 -

','-INCREASED RANGE OF RADIATION MONITORS (2.1.8.b)POSITIONThe requirements associated with this recommendation should be considered asadvanced implementation of certain requirements to be included in a revisionto Regulatory Guide 1.97, Instrumentation to Follow the Course of an Accident",which has already been initiated, and in other Regulatory Guides, which willbe promulgated in the near-term.1. Noble gas effluent monitors shall be installed with an extended rangedesigned to function during accident conditions as well as duringnormal operating conditions; multiple monitors are considered to benecessary to cover the ranges of interest.a. Noble gas effluent monitors with an upper range capacity of10 pCi/cc (Xe-133) are considered to be practical and shouldbe installed in all operating plants.b. Noble gas effluent monitoring shall be provided for the totalrange of concentration extending grom normal condition (ALARA)concentrations to a maximum of 10 vCi/cc (Xe-133). Multiplemonitors are considered to be necessary to cover the ranges ofinterest. The range capacity of individual monitors shouldoverlap by a factor of ten.2. Since iodine gaseous effluent monitors for the accident condition arenot considered to be practical at this time, capability for effluentmonitoring of radioiodines for the accident condition shall beprovided with sampling conducted by adsorption on charcoal or othermedia, followed by onsite laboratory analysis.3. In-containment radiation level monitors with a maximum range of 10rad/hr shall be installed. A minimum of two such monitors that arephysically separated shall be provided. Monitors shall be designedand qualified to function in an accident environment.DISCUSSIONThe January 1, 1980 requirement, were specifically added by the Commissionand were not included in NUREG-0578. The purpose of the interim January 1,1980 requirement is to assure that licensees have methods of quantifingradioactivity releases should the existing effluent instrumentation gooffscale.CLARIFICATION1. Radiological Noble Gas Effluent MonitorsA. January 1, 1980 RequirementsUntil final implementation in January 1, 1981, all operating reactorsmust provide, by January 1, 1980, an interim method for-31 -

quantifying high level releases which meets the requirements of Table2.1.8.b.l. This method is to serve only as a provisional fix withthe more detailed, exact methods to follow. Methods are to be developedto quantify release rates of up to 10,000 Ci/sec for noble gasesfrom all potential release points, (e.g., auxiliary building, radwastebuilding, fuel handling building, reactor building, waste gas decaytank releases, main condenser air ejector, BWR main condenser vacuumpump exhaust, PWR steam safety valves and atmosphere steam dump valvesand BWR turbine buildings) and any other areas that communicate directlywith systems which may contain primary coolant or containment gases,(e.g., letdown and emergency core cooling systems and external recombiners).Measurements/analysis capabilities of the effluents at the finalrelease point (e.g., stack) should be such that measurements of individualsources which contribute to a common release point may not be necessary.For assessing radioiodine and particulate releases, special proceduresmust be developed for the removal and analysis of the radioiodine/particulate sampling media (i.e., charcoal canister/filter paper).Existing sampling locations are expected to be adequate; however,special procedures for retrieval and analysis of the sampling mediaunder accident conditions (e.g., high air and surface contaminationand direct radiation levels) are needed.It is intended that the monitoring capabilities called for in theinterim can be accomplished with existing instrumentation or readilyavailable instrumentation. For noble gases, modifications to exist-ing monitoring systems, such as the use of portable high range survey32 -

instruments, set in shielded collimators so that they "see" smallsections of sampling lines is an acceptable method for meeting theintent of this requirement. Conversion of the measured dose rate(mR/hr) into concentration (vCi/cc) can be performed using standardvolume source calculations. A method must be developed with sufficientaccuracy to quantify the iodine releases in the presence of highbackground radiation from noble gases collected on charcoal filters.Seismically qualified equipment and equipment meeting IEEE-279 isnot required.The licensee shall provide the following information on his methodsto quantify gaseous releases of radioactivity from the plant duringan accident.1. Noble Gas Effluentsa. System/Method description including:i) Instrumentation to be used including range or sen-sitivity, energy dependence, and calibrationfrequency and technique,ii) Monitoring/sampling locations, including methods toassure representative measurements and backgroundradiation correction,iii) A description of method to be employed to facilitateaccess to radiation readings. For January 1, 1980,Control room read-out is perferred: however, if impractical,in-situ readings by an individual with verbal communicationwith the Control Room is acceptable based on (iv)below._33 -

iv) Capability to obtain radiation readings at leastevery 15 minutes during an accident.v) Source of power to be used. If normal AC poweris used, an alternate back-up power supply shouldbe provided. If DC power is used, the source shouldbe capable of providing continuous readout for 7consecutive days.b. Procedures for conducting all aspects of the measurement/analysis including:i) Procedures for minimizing occupational exposuresii) Calculational methods for converting instrumentreadings to release rates based on exhaust air flowand taking into consideration radionuclide spectrumdistribution as function of time after shutdown.Procedures for dissemination of information.Procedures for calibration.iii)iv)B. January 1, 1981 RequirementsBy January 1, 1981, the licensee shall provide high range noble gaseffluent monitors for each release path. The noblegas effluent monitor should meet the requirements of Table 2.1.8.b.2.The licensee shall also provide the information given in Sections l.A.l.a.i,l.A.l.a.ii, l.A.l.b.ii, l.A.l.b.iii, and l.A.l.b.iv above for the noblegas effluent monitors.-34 -

2. Radioiodine and Particulate EffluentsA. For January 1, 1980 the licensee should provide the following:1. System/Method description including:a) Instrumentation to be used for analysis of thesampling media with discussion on methods used tocorrect for potentially interfering background levelsof radioactivity.b) Monitoring/sampling location.c) Method to be used for retrieval and handling ofsampling media to minimize occupational exposure.d) Method to be used for data analysis of individualradionuclides in the presence of high levels ofradioactive noble gases.e) If normal AC power is used for sample collection andanalysis equipment, an alternate back-up power supplyshould be provided. If DC power is used, the sourceshould be capable of providing continuous read-outfor 7 consecutive days.2. Procedures for conducting all aspects of the measurementanalysis Including:a) Minimizing occupational exposureb) Calculational methods for determining release ratesc) Procedures for dissemination of informationd) Calibration frequency and technique-35 -

IB. For January 1, 1981, the licensee should have the capabilityto continuously sample and provide onsite analysis of thesampling media. The licensee should also provide theinformation required in 2.A above.3. Containment Radiation MonitorsProvide by January 1, 1981, two radiation monitor systems in containmentwhich are documented to meet the requirements of Table 2.1.b.b.2.It is possible that future regulatory requirements for emergency planninginterfaces may necessitate identification of different types of radionuclidesin the containment air, e.g., noble gases (indication of core damage) andnon-volatiles (indication of core melt). Consequently, considerationshould be given to the possible installation or future conversion ofthese monitors to perform this function.-36 -

TABLE 2.1.8.b.1INTERIM PROCEDURES FOR QUANTIFYING HIGH LEVELACCIDENTAL RADIOACTIVITY RELEASESLicensees are to implement procedures for estimating noble gas andradioiodine release rates if the existing effluent instrumentationgoes off scale.Examples of major elements of a highly radioactive effluent releasespecial procedures (noble gas).-Preselected location to measure radiation from the exhaustair, e.g., exhaust duct or sample line.-Provide shielding to minimize background interference.-Use of an installed monitor (preferable) or dedicated portablemonitor (acceptable) to measure the radiation.-Predetermined calculational method to convert the radiationlevel to radioactive effluent release rate.-37 -

.0TABLE 2.1.8.b.2HIGH RANGE EFFLUENT MONITORNOBLE GASES ONLYRANGE: (Overlap with Normal Effluent Instrument Range)-UNDILUTED CONTAINMENT EXHAUST pc 5 PCi/CC-DILUTED (>10: 1) CONTAINMENT EXHAUST 10 4 VCi/CC-MARK I BWR REACTOR BUILDING EXHAUST 10 4 VCi/CC-PWR SECONDARY CONTAINMENT EXHAUST 10 4 VCi/CC-BUILDINGS WITH SYSTEMS CONTAINING +3PRIMARY COOLANT OR GASES 10 VCi/CC-OTHER BUILDINGS (E.G., RADWASTE) 10 2 PCi/CCNOT REDUNDANT -1 PER NORMAL RELEASE POINTSEISMIC -NOPOWER -VITAL INSTRUMENT BUSSPECIFICATIONS -PER R.G. 1.97 AND ANSI N320-1979DISPLAY*: CONTINUOUS AND RECORDING WITH READOUTS IN THE TECHNICALSUPPORT CENTER (TSC) AND EMERGENCY OPERATIONS CENTER (EOC)* QUALIFICATIONS -NO*Although not a present requirement, it is likely that this information mayhave to be transmitted to the NRC. Consequently, consideration should begiven to this possible future requirement when designing the displayinterfaces.-38 -

TABLE 2.1.8.b.3HIGH RANGE CONTAINMENT RADIATION MONITORRADIATION: TOTAL RADIATION (ALTERNATE: PHOTON ONLY)RANGE:-UP TO 10 8 RAD/HR (TOTAL RADIATION)-ALTERNATE: 10 7 R/HR (PHOTON RADIATION ONLY)-SENSITIVE DOWN TO 60 KEY PHOTONS*REDUNDANT: TWO PHYSICALLY SEPARATED UNITSSEISMIC: PEROR. G. 1.97POWER: VITAL INSTRUMENT BUSSPECIFICATIONS: PER R.G. 1.97 REV. 2 AND ANSI N320-1978DISPLAY: CONTINUOUS AND RECORDING* CALIBRATION: LABORATORY CALIBRATION ACCEPTABLE*Monitors must not provide misleading information to the operators assumingdelayed core damage when the 80 KEV photon Xe-133 is the major noble gaspresent.-39 -

/IMPROVED IN-PLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS (2.1a.c)POSITIONEach licensee shall provide equipment and associated training and proceduresfor accurately determining the airborne iodine concentration in areas withinthe facility where plant personnel may be present during an accident.CLARIFICATIONUse of Portable versus Stationary Monitoring EquipmentEffective monitoring of increasing iodine levels in the buildings underaccident conditions must include the use of portable instruments for thefollowing reasons:a. The physical size of the auxiliary/fuel handling buildingprecludes locating stationary monitoring instrumentation atall areas where airborne iodine concentration data might berequired.b. Unanticipated isolated "hot spots" may occur in locations whereno stationary monitoring instrumentation is located.c. Unexpectedly high background radiation levels near stationarymonitoring instrumentation after an accident may interfere withfilter radiation readings.-40 -

d. The time required to retrieve samples after an accidentmay result in high personnel exposures if these filters arelocated in high dose rate areas.Iodine Filters and Measurement TechniquesA. The following are short-term recommendations and shall be implementedby the licensee by January 1, 1980. The licensee shall have the capabilityto accurately detect the presence of iodine in the region of interestfollowing an accident. This can be accomplished by using a portableor cart-mounted iodine sampler with attached single channel analyzer(SCA). The SCA window should be calibrated to the 365 keV of 13I.A representative air sample shall be taken and then counted for 1 1Iusing the SCA. This will give an initial conservative estimate of presenceof iodine and can be used to determine if respiratory protection isrequired. Care must be taken to assure that the counting system isnot saturated as a result of too must activity collected on the samplingcartridge.B. By January 1, 1981:The licensee shall have the capability to remove the sampling cartridgeto a low background, low contamination area for further analysis. Thisarea should be ventilated with clean air containing no airborneradionuclides which may contribute to inaccuracies in analyzing thesample. Here, the sample should first be purged of any entrapped noblegases using nitrogen gas or clean air free of noble bases. The licenseeshall have the capability to measure accurately the iodine concentrationspresent on these samples and effluent charcoal samples under accidentconditions.-41 -

TRANSIENT AND ACCIDENT ANALYSIS (2.1.9)POSITIONSee NUREG-0578, page A-44.DISCUSSIONThe scope of the requied transient and accident analysis is discussed inNUREG-0578. The schedule for these analyses is included in NUREG-0578and is reproduced in the Implementation Schedule attachment to this letter.The Bulletins and Orders Task Force has been implementing these requiredanalyses on that schedule. The analysis of the small break loss of coolantaccident has been submitted by each of the owners groups. These analysesare presently under review by the B&O Task Force. The scope and schedulefor the analysis of inadequate core cooling have been discussed and agreedupon in meetings between the owners groups and the B&O Task Force, andare documented in the minutes to those meetings.The analysis of transients and accidents for the purpose of upgrading emergencyprocedures is due in early 1980 and the detailed scope and schedule of thisanalysis is the subject of continuing discussions between the owners groupsand the B&O Task Force.-42 -

CONTAINMENT PRESSURE INDICATIONPOSITIONA continuous indication of containment pressure should be provided in thecontrol room. Measurement and indication capability shall include threetimes the design pressure of the containment for concrete, four times thedesign pressure for steel, and minus five psig for all containments.CLARIFICATION1. The containment pressure indication shall meet the design provisionsof Regulatory Guide 1.97 including qualification, redundancy, andtestability.2. The containment pressure monitor shall be installed by January 1, 1981.-43 -

CONTAINMENT WATER LEVEL INDICATIONPOSITIONA continuous indication of containment water level shall be provided in thecontrol room for all plants. A narrow range instrument shall be providedfor PWRs and cover the range from the bottom to the top of the containmentsump. A wide range instrument shall also be provided for PWRs and shallcover the range from the bottom of the containment to the elevation equivalentto a 600,000 gallon capacity. For BWRs, a wide range instrument shall beprovided and cover the range from the bottom to 5 feet above the normal waterlevel of the suppression pool.CLARIFICATION1. The narrow range sump level instrument shall monitor the normal contain-ment sump level vice the containment emergency sump level.2. The wide range containment water level instruments shall meet the require-ments of the proposed revision to Regulatory Guide 1.97 (Instrumentationfor Light-Water Cooled Nuclear Power Plant to Assess Plant ConditionsDuring and Following a Accident).3. The narrow range containment water level instruments shall meet therequirements of Regulatory Guide 1.89 (Qualification of Class IE Equipmentof Nuclear Power Plants).4. The equivalent capacity of the wide range PWR level instrument has beenchanged from 500,000 gallons to 600,000 gallons to ensure consistency withthe proposed revision to Regulatory Guide 1.97. It should be noted thatthis measurement capability is based on recent plant designs. For olderplants with smaller water capacities, licensees may propose deviationsfrom this requirement based on the available water supply capability attheir plant.5. The containment water level indication shall be installed by January 1, 19i1.-44 -

CONTAINMENT HYDROGEN INDICATIONPOSITIONA continuous indicaton of hydrogen concentration in the containment atmosphereshall be provided in the control room. Measurement capability shall be providedover the range of 0 to lO hydrogen concentration under both positive and negativeambient pressure.CLARIFICATION1. The containment hydrogen indication shall meet the design provisionsof Regulatory Guide 1.97 including qualification, redundancy, andtestability.2. The containment hydrogen indication shall be installed by January 1, 1981.-45 -

REACTOR COOLANT SYSTEM VENTINGPOSITIONEach applicant and licensee shall install reactor Coolant system and reactorvessel head high point vents remotely operated from the control room. Sincethese vents form a part of the reactor coolant pressure boundary, the designof the vents shall conform to the requirements of Appendix A to 10 CFRPart 50 General Design Criteria. In particular, these vents shall be safetygrade, and shall satisfy the single failure criterion and the requirementsof IEEE-279 in order to ensure a low probability of inadvertent actuation.Each applicant and licensee shall provide the following information concerningthe design and operation of these high point vents:1. A description of the construction, location, size, and power supply forthe vents along with results of analyses of loss-of-coolant accidentsinitiated by a break in the vent pipe. The results of the analyses shouldbe demonstrated to be acceptable in accordance with the acceptancecriteria of 10 CFR 50.46.2. Analyses deomonstrating that the direct venting of noncondensable gaseswith perhaps high hydrogen concentrations does not result in violationof combustible gas concentration limits in containment as described in10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard ReviewPlan Section 6.2.5.3. Procedural guidelines for the operators' use of the vents. The informationavailable to the operator for initiating or terminating vent usage shallbe discussed.CLARIFICATIONA. General1. The two important safety functions enhanced by this venting capabilityare core cooling and containment integrity. For events within thepresent design basis for nuclear power plants, the capability to ventnon-condensible gases will provide additional assurance of meeting therequirements of lOCFR50.46 (LOCA criteria) and lOCFR50.44 (containmentcriteria for hydrogen generation). For events beyond the present designbasis, this venting capability will substantially increase the plant'sability to deal with large quantities of non-condensible gas without theloss of core cooling or containment integrity._ 46 -

2. Procedures addressing the use of the RCS vents are required byJanuary 1, 1981. The procedures should define the conditions underwhich the vents should be used as well as the conditions underwhich the vents should not be used. The procedures should be basedon the following criteria: (1) assurance that the plant can meetthe requirements of lOCFR5U.46 and lOCFR5U.44 for Design BasisAccidents; and (2) a substantial increase in the plants abilityto maintain core cooling and containment integrity for events beyondthe Design Basis.B. BWR Design Considerations1. Since the BWR owners group has suggested that the present BWR designsinherent capability of venting, this question relates to the capabilityof existing systems. The ability of these systems to vent theRCS of non-condensible gas must be demonstrated. In addition theability of these systems to meet the same requirements as the PWRvent systems must be documented. Since there are important differencesamong BWR's, each licensee should address the specific design featuresof his plant.2. In addition to reactor coolant system venting, each BWR licenseeshould address the ability to vent other systems such as the isolationcondenser, which may be required to maintain adequate core cooling.If the production of a large amount of non-condensible gas wouldcause the loss of function of such a system, remote venting of thatsystem is required. The qualifications of such a venting system should bethe same as that required for PWR venting systems.-47 -

C. PWR Vent Design Considerations1. The locations for PWR Vents are as follows:a. Each PWR licensee should provide the capability to vent thereactor vessel head.b. The reactor vessel head vent should be capable of venting non-condensible gas from the reactor vessel hot legs (to the elevationof the top of the outlet nozzle) and cold legs (through headjets and other leakage paths). Additional venting capabilityis required for those portions of each hot leg which can notbe vented through the the reactor vessel head vent. The NRCrecognizes that it is impractical to vent each of the manythousands of tubes in a U-tube steam generator. However, webelieve that a procedure can be developed which assures thatsufficient liquid or steam can enter the U-tube region so thatdecay heat can be effectively removed from the reactor coolantsystem. Such a procedure is required by January 1981.c. Venting of the pressurizer is required to assure its availabilityfor system pressure and volume control. These are importantconsiderations especially during natural circulation.2. The size of the reactor coolant vents is not a critical issue.The desired venting capability can be achieved with vents in afairly large range of sizes. The criteria for sizing a vent canbe developed in several ways. One approach, which we consider reasonable,is to specify a volume of non-condensible gas to be vented anda venting time i.e., a vent capable of venting a gas volume of1/2 the RCS in one hour. Other criteria and engineering approachesshould be considered if desired.-48 -

3. Where practical the RCS vents should be kept smaller than the sizecorresponding to the definition of a LOCA (lOCFR5O Appendix A).This will minimize the challenges to the ECCS since the inadvertentopening of a vent smaller than the LOCA definition would not requireECCS actuation although it may result in leakage beyond TechnicalSpecification Limits. On PWRs the use of new or existing valveswhich are larger than the LOCA definition will require the additionof a block valve which can be closed remotely to terminate theLOCA resulting from the inadvertent opening of the vent.4. An indication of valve position should be provided in the controlroom.5. Each vent should be remotely operable from the control room.6. Each vent should be seismically qualified.7. The requirements for a safety grade system is the same as the safetygrade requirement on other Short Term Lessons Learned items, thatis, it should have the same qualifications as were accepted for thereactor protection system when the plant was licensed. The exceptionto this requirement is that we do not require redundant valves at eachventing location. Each vent must have its power supplied from anemergency bus. A degree of redundancy should be provided by poweringdifferent vents from different emergency buses.8. For systems where a block valve is required, the block valve shouldhave the same qualifications as the vent.-49 -

'I9. Since the RCS vent system will be part of the reactor coolant systemsboundary, efforts should be made to minimize the probability of aninadvertent actuation of the system. Removing power from the vents isone step in the direction. Other steps are also encouraged.10. Since the generation of large quantities of non-condensible gas could beassociated with substantial core damage, venting to atmosphere isunacceptable because of the associated released radioactivity. Ventinginto containment is the only presently available alternative. Withincontainment those areas which provide good mixing with containment airare preferred. In addition, areas which provide for maximum cooling ofthe vented gas are preferred. Therefore the selection of a location forventing should take advantage of existing ventilation and heat removalsystems.11. The inadvertent opening of an RCS vent must be addressed. For ventssmaller than the LOCA definition, leakage detection must be sufficientto identify the leakage. For vents larger than the LOCA definition,an analysis is required to demonstrate compliance with lOCFR50.46.-50 -

SHIFT SUPERVISOR RESPONSIBILITIES (2.2.1.a)POSITION1. The highest level of corporate management of each licensee shallissue and periodically reissue a management directive that emphasizesthe primary management responsibility of the shift supervisor for safeoperation of the plant under all conditions on his shift and thatclearly establishes his command duties.2. Plant procedures shall be reviewed to assure that the duties, responsi-bilities, and authority of the shift supervisor and control room operatorsare properly defined to effect the establishment of a definite line ofcommand and clear delineation of the command decision authority of the shiftsupervisor in the control room relative to other plant management personnel.Particular emphasis shall be placed on the following:a. The responsibility and authority of the shift supervisor shallbe to maintain the broadest perspective of operational conditionsaffecting the safety of the plant as a matter of highest priorityat all times when on duty in the control room. The idea shall bereinforced that the shift supervisor should not become totallyinvolved in any single operation in times of emergency whenmultiple operations are required in the control room.b. The shift supervisor, until properly relieved, shall remain inthe control room at all times during accident situations todirect the activities of control room operators. Personsauthorized to relieve the shift supervisor shall be specified.c. If the shift supervisor is temporarily absent from the controlroom during routine operations, a lead control room operatorshall be designated to assume the control room command function.These temporary duties, responsibilities, and authority shall beclearly secified.3. Training programs for shift supervisors shall emphasize and reinforce theresponsibility for safe operation and the management function the shiftsupervisor is to provide for assuring safety.4. The administrative duties-of the shift supervisor shall be reviewed by thesenior officer of each utility responsible for plant operations. Administra-tive functions that detract from or are subordinate to the managementresponsibility for assuring the safe operation of the plant shall bedelegated to other operations personnel not on duty in the control room.CLARIFICATIONThe attachment provides clarification to the above position.-51 -

AttachmentSHIFT SUPERVISOR RESPONSIBILITY (2.2.1.A)NUREG-0578 POSITION (POSITION NO.)Highest Level of Corporate Management (1.)Periodically Reissue (1.)Management Direction (1.)Properly Defined (2.0)Until Properly Relieved (2.B)Temporarily Absent (2.C)Control Room Defined (2.C)Designated (2.C)Clearly SpecifiedSRO TrainingAdministrative Duties (4.)Administrative Duties Reviewed (4.)CLARIFICATIONV. P. For OperationsAnnual Reinforcement ofCompany PolicyFormal Documentation of ShiftPersonnel, All PlantManagement, Copy to IE RegionDefined in Writing in aPlant ProcedureFormal Transfer of Authority,Valid SRO License, Recordedin Plant LogAny AbsenceIncludes Shift SupervisorOffice Adjacent to theControl RoomIn Administrative ProceduresDefined in AdministrativeProceduresSpecified in ANS 3.1 (Draft)Section 5.2.1.8Not Affecting Plant SafetyOn Same Interval as Reinforcement:i.e., Annual by V. P. forOperations.-52 -

SHIFT TECHNICAL ADVISOR (Section 2.2.1.b)POSITIONEach licensee shall provide an on-shift technical advisor to the shift supervisor.The shift technical advisor may serve more than one unit at a multi-unit siteif qualified to perform the advisor function for the various units.The Shift Technical Advisor shall have a bachelor's degree or equivalent in ascientific or engineering discipline and have received specific training in theresponse and analysis of the plant for transients and accidents. The Shift TechnicalAdvisor shall also receive training in plant design and layout, including thecapabilities of instrumentation and controls in the control room. The licenseeshall assign normal duties to the Shift Technical Advisors that pertain to theengineering aspects of assuring safe operations of the plant, including the reviewand evaluation of operating experience.DISCUSSIONThe NRC Lessons Learned Task Force has recommended the use of-Shift TechnicalAdviors (STA) as a method of immediately improving the plant operating staff'scapabilities for response to off-normal conditions and for evaluating operatingexperience.In defining the characteristics of the STA, we have used the two essentialfunctions to be provided by the STA. These are accident assessment and operatingexperience assessment.1. Accident AssessmentThe STA serving the accident assessment function must be dedicated to concernfor the safety of the plant. The STA's duties will be to diagnose off-normalevents and advise the shift supervisor. The duties of the STA should notinclude the manipulatin of controls or supervision of operators. The STA mustbe available, in the control room, within 10 minutes of being summoned.The qualifications of the STA should include college level educationin engineering and science subjects as well as training in reactor operationsboth normal and off-normal. Details regarding these qualifications areprovided in paragraphs A.1, 2 and 3 of Enclosure 2 to our September 13, 1979letter. In addition, the STA serving the accident assessment function must becognizant of the evaluations performed as part of the operating experienceassessment function.-53 -

92. Operating Experience AssessmentThe persons serving the opeating experience assessment function must bededicated to concern for the safety of the plant. Their function will be toevaluate plant operations from a safety point of view and should include suchassignments as listed on pages A-50 and A-51 of NUREG-0578. Their qualifica-tions are identical to those described previously under accident assessmentand collectively this group should provide competence in all technical areasimportant to safety. It is desirable that this function be performed byonsite personnel.CLARIFICATION1. Due to the similarity in the requirements for dedication to safety, trainingand onsite location and the desire that the accident assessment function beperformed by someone whose normal duties involve review of operatingexperiences, our preferred position is that the same people perform theaccident and operating experience assessment functions. The performance ofthese two functions may be split if it can be demonstrated the personsassigned the accident assessment role are aware, on a current basis, of thework being done by those reviewing operating experience.2. To provide assurance that the STA will be dedicated to concern for the safetyof the plant, our position has been that STA's must have a clear measure ofindependence from duties associated with the commercial operation ofthe plant. This would minimize possible distractions from safetyjudgements by the demands of commercial operations. We have determinedthat, while desirable, independence from the operations staff of theplant is not necessary to provide this assurance. It is necessary,however, to clearly emphasize the dedication to safety associatedwith the STA position both in the STA job description and in the personnelfilling this position. It is not acceptable to assign a person, who isnormally the immediate supervisor of the shift supervisor to STA dutiesas defined herein._54 _

3. It is our position that the STA should be available within 10 minutes of beingsummoned and therefore should be onsite. The onsite STA may be in a duty statusfor periods of time longer than one shift, and therefore asleep at some times,if the ten minute availability is assured. It is preferable to locatethose doing the operating experience assessment onsite. The desired exposureto the operating plant and contact with the STA (if these functions are tobe split) may be able to be accomplished by a group, normally stationed offsite,with frequent onsite presence. We do not intend, at this time, to specify oradvocate a minimum time onsite.4. The implementation schedule for the STA requirements is to have the STA on dutyby January 1, 1980, and to have STAs, who have all completed training require-ments, on duty by January 1, 1981. While minimum training requirementshave not been specified for January 1, 1980, the STAs on duty by that timeshould enhance the accident and operating experience assessment function atthe plant.-55 -

1 .SHIFT AND RELIEF TURNOVER PROCEDURES (2.2.1.c)POSITIONThe licensees shall review and revise as necessary the plant procedure for shiftand relief turnover to assure the following:1. A checklist shall be provided for the oncoming and offgoing control roomoperators and the oncoming shift supervisor to complete and sign. Thefollowing items, as a minimum, shall be included in the checklist.a. Assurance that critical plant parameters are within allowablelimits (parameters and allowable limits shall be listed on thechecklist).b. Assurance of the availability and proper alignment of all systemsessential to the prevention and mitigation of operational transientsand accidents by a check of the control console.(what to check and criteria for acceptable status shall be includedon the checklist);c. Identification of systems and components that are in a degraded modeof operation permitted by the Technical Specifications. For suchsystems and components, the length of time in the degraded mode shallbe compared with the Technical Specifications action statement (thisshall be recorded as a separate entry on the checklist).2. Checklists or logs shall be provided for completion by the offgoing andongoing auxiliary operators and technicians. Such checklists or logs shallinclude any equipment under maintenance or test that by themselves coulddegrade a system critical to the prevention and mitigation of operationaltransients and accidents or initiate an operational transient (what tocheck and criteria for acceptable status shall be included on the checklist);and3. A system shall be established to evaluate the effectiveness of the shiftand relief turnover procedure (for example, periodic independent verificationof system alignments).CLARIFICATIONNo clarification provided.-56 -

CONTROL ROOM ACCESS (2.2.2.a)POSITIONThe licensee shall make provisions for limiting access to the control room tothose individuals responsible for the direct operation of the nuclear power plant(e.g., operations supervisor, shift supervisor, and control room operators),to technical advisors who may be requested or required to support the operation,and to predesignated NRC personnel. Provisions shall include the following:1. Develop and implement an administrative procedure that establishesthe authority and responsibility of the person in charge of thecontrol room to limit access, and2. Develop and implement procedures that establish a clear line ofauthority and responsibility in the control room in the event of anemergency. The line of succession for the person in charge of thecontrol room shall be established and limited to persons possessinga current senior reactor operator's license. The plan shall clearlydefine the lines of communication and authority for plant managementpersonnel not in direct command of operations, including those whoreport to stations outside of the control room.CLARIFICATIONNo clarification provided.57 -

.IONSITE TECHNICAL SUPPORT CENTER (TSC) 2.2.2.bPOSITIONEach operating nuclear power plant shall maintain an onsite technical supportcenter separate from and in close proximity to the control room that has thecapability to display and transmit plant status to those individuals who areknowledgeable of and responsible for engineering and management support ofreactor operations in the event of an accident. The center shall be habitableto the same degree as the control room for postulated accident conditions.The licensee shall revise his emergency plans as necessary to incorporatethe role and location of the technical support center. Records that pertain tothe as-built conditions and layout of structures, systems and components shallbe readily available to personnel in the TSC.CLARIFICATION1. By January 1, 1980, each licensee should meet items A-G that follow. Eachlicensee is encouraged to provide additional upgrading of the TSC (items2-10) as soon as practical, but no later than January 1, 1981.A. Establish a TSC and provide a complete description,B. Provide plans and procedures for engineering/management support andstaffing of the TSC,C. Install dedicated communications between the TSC and the controlroom, near site emergency operations center, and the NRC,D. Provide monitoring (either portable or permanent) for both directradiation and airborne radioactive contaminmants. The monitorsshould provide warning if the radiation levels in the support centerare reaching potentially dangerous levels. The licensee shoulddesignate action levels to define when protective measures shouldbe taken (such as using breathing apparatus and potassium iodide tablets,or evacuation to the control room),E. Assimilate or ensure access to Technical Data, including the licensee'sbest effort to have direct display of plant parameters, necessary forassessment in the TSC,-58 -

F. Develop procedures for performing this accident assessment functionfrom the control room should the TSC become uninhabitable, andG. Submit to the NRC a longer range plan for upgrading the TSC tomeet all requirements.2. LocationIt is recommended that the TSC be located in close proximity to the controlroom to ease communications and access to technical information during anemergency. The center should be located onsite, i.e., within the plantsecurity boundary. The greater the distance from the CR, the moresophisticated and complete should be the communications and availabilityof technical information. Consideration should be given to providing keyTSC personnel with a means for gaining access to the control room.3. Physical Size & StaffingThe TSC should be large enough to house 25 persons, necessary engineeringdata and information displays (TV monitors, recorders, etc.). Eachlicensee should specify staffing levels and disciplines reporting tothe TSC for emergencies of varying severity.4 ActivationThe center should be activated in accordance with the "Alert" level as definedin the NRC document "Draft Emergency Action Level Guidelines, NUREG-0610"dated September, 1979, and currently out for public comment. Instrumentation inthe TSC should be capable of providing displays of vital plant parameters fromthe time the accident began (t = 0 defined as either reactor or turbine trip).The Shift Technical Advisor should be consulted on the "Notification of UnusualEvent" however, the activation of the TSC is discretionary for that class ofevent.-59-

145. InstrumentationThe instrumentation to be located in the TSC need not meet safety-graderequirements but should be qualitatively comparable (as regards accuracyand reliability) to that in the control room. The TSC should have thecapability to access and display plant parameters independent fromactions in the control room. Careful consideration should be given tothe design of the interface of the TSC instrumentation to assure thataddition of the TSC will not result in any degradation of the controlroom or other plant functions.6. Instrumentation Power SupplyThe power supply to the TSC instrumentation need not meet safety-graderequirements, but should be reliable and of a quality compatible with theTSC instrumentation requirements. To insure continuity of informationat the TSC, the power supply provided should be continuous once the TSCis activated. Consideration should be given to avoid loss of storeddata (e.g., plant computer) due to momentary loss of power or switchingtransients. If the power supply is provided from a plant safety-relatedpower source, careful attention should be give to assure that thecapability and reliability of the safety-related power source is notdegraded as a result of this modification.7. Technical DataEach licensee should establish the technical data requirements for the TSC,keeping in mind the accident assessment function that has been establishedfor those persons reporting to the TSC during an emergency. As a minimum,_60 -

Wdata (historical in addition to current status) should be available topermit the assessment of:Plant Safety Systems Parameters for:* Reactor Coolant System* Secondary System (PWRs)* ECCS Systems* Feedwater & Makeup Systems* ContainmentIn-Plant Radiological Parameters for:* Reactor Coolant System* Containment* Effluent Treatment* Release PathsOffsite Radiological* Meteorology* Offsite Radiation Levels8. Data TransmissionIn addition to providing a data transmission link between the TSC and thecontrol room, each licensee should review current technology as regardstransmission of those parameters identified for TSC display.Although there is not a requirement at the present time, each licenseeshould investigate the capability to transmit plant data offsite to theEmergency Operations Center, the NRC, the reactor vendor, etc.-61 -

-19. Structural IntegrityA. The TSC need not be designed to seismic Category I requirements.The center should be well built in accordance with sound engineeringpractice with due consideration to the effects of natural phenomenathat may occur at the site.B. Since the center need not be designed to the same stringent requirementsas the Control Room, each licensee should prepare a backup plan forresponding to an emergency from the control room.10. HabitabilityThe licensee should provide protection for the technical support centerpersonnel from radiological hazards including direct radiation and airbornecontaminants as per General Design Criterion 19 and SRP 6.4.A. Licensee should assure that personnel inside the technical supportcenter (TSC) will not receive doses in excess of those specified inGDC 19 and SRP 6.4 (i.e., 5 Rem whole body and 30 Rem to the thyroidfor the duration of the accident). Major sources of radiation shouldbe considered.B. Permanent monitoring systems should be provided to continuouslyindicate radiation dose rates and airborne radioactivity concentrationsinside the TSC. The monitoring systems should include local alarmsto warn personnel of adverse conditions. Procedures must be providedwhich will specify appropriate protective actions to be taken in theevent that high dose rates or airborne radioactive concentrationsexist.-62 -

C. Permanent ventilation systems which include particulate and charcoal filtersshould be provided. The ventilation systems need not be qualified asESF systems. The design and testing guidance of Regulatory Guide 1.52should be followed except that the systems do not have to be redundant,seismic, instrumented in the control room or automatically activated. Inaddition, the HEPA filters need not be tested as specified in RegulatoryGuide 1.52 and the HEPA's do not have to meet the QA requirements of AppendixB to 10 CFR 50. However, spare parts should be readily available andprocedures in place for replacing failed components during an accident.The systems should be designed to operate from the emergency power supply.D. Dose reduction measures such as breathing apparatus and potassium iodidetablets can not be used as a design basis for the TSC in lieu of ventilationsystems with charcoal filters. However, potassium iodide and breathingapparatus should be available.-63 -

-/ONSITE OPERATIONAL SUPPORT CENTER (SECTION 2.2.2.c)POSITIONAn area to bedesignated as the onsite operational support center shall beestablished. It shall be separate from the control room and shall be theplace to which the operations support personnel will report in an emergencysituation. Communications with the control room shall be provided. Theemergency plan shall be revised to reflect the existence of the center andto establish the methods and lines of communication and management.CLARIFICATIONNo clarification provided.-64 -

IMPLEMENTATION SCHEDULESECTIONNUMBERIMPLEM.CAT. (1)PROPOSAL IMPLEMENTATIONREVIEW REVIEWTITLE2.1.1Emergency Power SupplyPressurizer HeatersPressurizer LevelPORV and Block ValveAAAxxx2.1.2Relief and Safety Valve TestProgram and ScheduleComplete TestA07/81xx2.1.3 aDirect Indication of ValvePositionAx(nl2.1.3.bInstrumentation forInadequate Core CoolingProceduresDesign of NewInstrumentationSubcooling MeterInstallation of New Instr.(E.G., Level Meter)AxAABxxxCm00U)IMPLEMENTATION COMPLETE BY JANUARY 1, 1980, rIMPLEMENTATION COMPLETE BY JANUARY 1, 1981.(1) CATEGORY A:CATEGORY B:

IMPLEMENTATION SCHEDULESECTIONNUMBERIMPLEM.CAT. (1)TITLEPROPOSAL IMPLEMENTATIONREVIEW REVIEWx2.1.42.1.5Containment IsolationDedicated H2 ControlPenetrationsADescription and ScheduleInstallationABAxxx2.1.5.c2.1.6.aRecombiner ProceduresSystems Integrity for HighRadioactivityLeak Reduction ProgramPreventative MaintenanceProgramPlant Shielding ReviewAAxxlIs2.1.6.bCDesign ReviewPlant ModificationsABxx(1) CATEGORY A:CATEGORY B:IMPLEMENTATION COMPLETE BY JANUARYIMPLEMENTATION COMPLETE BY JANUARY1, 1980,1, 1981.

a994 1IMPLEMENTATION SCHEDULESECTIONNUMBERIMPLEM.CAT. (1)PROPOSAL IMPLEMENTATIONREVIEW REVIEWTITLE2.1.7.aAuto Initiation of AFW2.1.7.bControl GradeSafety GradeAFW FlowControl GradeSafety GradeABABcxxxx2.1.8.aI8Post-Accident SamplingDesign ReviewProceduresDescription of PlantModificationsPlant ModificationsHigh Range Radiation MonitorsIn-ContainmentEffluents -ProceduresImplementAAxxABxx2.1.8.bBABxxx(1) CATEGORY A:CATEGORY B:IMPLEMENTATION COMPLETE BY JANUARY 1, 1980,IMPLEMENTATION COMPLETE BY JANUARY 1, 1981.

IMPLEMENTATION SCHEDULESECTIONNUMBER2.1.8.cIMPLEM.CAT. (1)TITLEImproved Iodine InstrumentationPROPOSAL IMPLEMENTATIONREVIEW REVIEWCAx2.1.9Transient and Accident AnalysisContainment Pressure MonitorContainment Water Level MointorContainment Hydrogen MonitorRCS VentingDesign CompleteInstallation CompleteShift Supervisor Responsibilities(2)xBBBABxxxxxx2.2.1.aAa}IoM((1) CATEGORY A:CATEGORY B:IMPLEMENTATION COMPLETE BY JANUARY 1, 1980,IMPLEMENTATION COMPLETE BY JANUARY 1, 1981.(2) SEE NUREG-0578-C

.IMPLEMENTATION SCHEDULESECTIONNUMBER2.1.2.B2.2.1.C2.2.2.A2.2.2.BIMPLEM.CAT. (1)TITLEShift Technical AdvisorAdvisor on DutyComplete TrainingShift Turnover ProcedureControl Room AccessOn Site Technical Support CenterEstablish CenterUpgrade to Meet All RequirementsOn Site Operational SupportCenterPROPOSAL IMPLEMENTATIONREVIEW REVIEWxxAB(AxAxABxxutl2.2.2.CAxc(1) CATEGORY A:CATEGORY B:IMPLEMENTATION COMPLETE BY JANUARY 1, 1980.IMPLEMENTATION COMPLETE BY JANUARY 1, 1981.

IANALYSIS AND TRAINING SCHEDULETask DescriptionCompletion Date1. Small Break LOCA analysis and preparationof emergency procedure guidelines2. Implementation of small break LOCAemergency procedures and retrainingof operators3. Analysis of inadequate core cooling andpreparation of emergency procedureguidelines4. Implementation of emergency proceduresand retraining related to inadequatecore cooling5. Analysis of accidents and transients andpreparation of emergency procedureguidelines6. Implementation of emergency proceduresand retraining related to accidentsand transients7. Analysis of LOFT small break testsJuly-September 1979*December 31, 1979October 1979January 1980Early 19803 months afterguidelines establishedPretest(Mid-September 1979)* Range covers completion dates for the four NSSS vendors-70 -