ML19250A835
| ML19250A835 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/18/1979 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7910240674 | |
| Download: ML19250A835 (27) | |
Text
)SMUD 4
=# SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s street, Box 15830, sacramento, California 95813; (916) 452-3211 October 18, 1979 Mr. D. G. Eisenhut, Acting Director Division of Operating Reactor U.S. Nuclear Regulatory Cornission Washington, D.C.
20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station, Unit No.1
Dear Mr. Eisenhut:
Your letter of September 13, 1979 requires the Sacramento Municipal Utility District to commit to the action listed in the report titled TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendation (NUREG-0578).
The letter also details additional requirements related to emergency planning, to containment status monitoring and to reactor coolant system venting. At the September 26, 1979 regional meeting in Las Vegas, Nevada you distributed more implementation information in the handout titled " Regional Meetings TMI Short-Term Implementation".
The Sacramento Municipal Utility District intends to implement the applicable requirements of NUREG-0578 as modified by the documents listed in the above paragraph.
In general, items listed for completion by January 1,1980 will be done during the refueling outage currently scheduled to start in January, 1980. Similarly, the District intends to complete the items requirinc imple-mentation by January 1,1981 during the refueling outage scheduled for 1981.
The attachment to this letter gives currently available detail for each imple-mentation item. Anticipated technical, timing and procedural problems are noted.
Many of the implementation items require " safety grade" installation. A defendable definition of " safety grade" will be available for each installa-tion.
If you require additional infonnation concerning implementation of NUREG-0578 and related item;., please contact me.
Sincerely,
(
Y Yh $
v John J. Mattimoe Assistant General Manager D
- \\
'6 and Chief Engineer 7910240b s
1199 246-
Attachment NUREG - 0578 Implementation Status 2.1.1 Emergency Power Supply Requirements for Pressurizer Heater, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRS.
NRC Position on Pressurizer Heater Power Supply 1.
The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions.
The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.
2.
Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses.
If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connec-tion of the pressurizer heaters.
3.
The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
4.
Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.
Sacramento Municipal Utility District Commitment Power for at least 126 KW of pressurizer heaters will be available from a diesel generator supplied emergency bus when off-site power is unavailable. Transition to the emergency power supply will require less than two hours.
The District intends to complete system modifications and associated pro-cedures and training during the next refueling outage.
1199 247 ggyA
NRC Position on Power Supplies for Power-0perated Relief Valves and Block Valves and for Pressurizer Level 1.
Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
2.
fiotive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
3.
Motive and control power connections to the emergency buses for the PORVs and their associated block va'ves shall be through devices that have been qualifieu in accordance with safety-grade requirements.
4.
The pressurizer level indication instrument channels shall be powered from the vital instrument buses.
These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.
Sacramento Municipal Utility District Commitment The PORV is currently powered from a battery which may be charged by a diesel generator. Pressurizer level indication power comes from an inverter which receives its power input from a battery which may be charged by a diesel ger erator. As such, the existing PORV and pressurizer level power supplies satisfy the requirements of NUREG 0578.
The PORV Block valve power supply will be shifted to a bus which can be supplied by the diesel generator. The shift will be completed during the next refueling outage.
2.1.2.
Performance Testing for BWR and PWR Relief and Safety Valves NRC Position Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valvec under expected operating conditions for design basis transients and accidents.
The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
The single failures applied to these analyses shall be chosen so that the dynamic forces on the
} }99 h kJ safety and relief valves are maximized. Test procedures shall be the highest predicted by conventional safety analysis pro-cedures.
Reactor coolant system relief and safety valve quali-fication shall include qualification of associated control circuitry piping and supports as well as the valves themselves.
Sacramento Municipal Utility District Commitment The Sacramento Municipal Utility District will participate in the EPRI/NSAC program to conduct performance testing of PWR relief and safety valves. We will verify that the program is applicable to our plant.
It is understood that this program will be reviewed vith the NRC prior to testing to ensure that the intent of NURUi 0578 is satisfied.
The District believes that substantive test data can be obtained by July, 1981. However, scheduling of the test facility, acquisi-tion of the valves to be tested, and the possibility of extensive retesting could result in delays.
2.1.3.a Direct Indication of Power-0perated Relief Valve and Safety Valve Position Indication for PWRs and BWRs.
NRC Position Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indica-tion of flow in the discharge pipe.
Sacramento Municipal Utility District Commitment The Sacramento Municipal Utility District intends to install positive status indication for the PORV and safety valves.
Status indication will alarm and indicate in the control room.
The valve status indication system will be installed during the next refueling outage; however, full compliance with safety-grade criteria may not be incorporated until the 1981 refueling outage.
2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs.
NRC Position 1.
Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with99 currently available instrumentation. The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions. A detailed description of the analyses needed to fom the basis for operator training and procedure develop-ment shall be provided pursuint to another short-term requirement, " Analysis of C/f-Normal Conditions Including Natural Circulation" (see Section 2.1.9 of this appendix). In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that is net to be used exclusive of other related plant parameters. 2. Licensees shall provide a description of any additional instrumentation or controls (primary or backup) pro-posed for 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 equip-ment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided. Sacramento ftunicipal Utility District Commitment The District is currently participating in a study to develop operational guidelines for inadequate core cooling. The B&W Owners Group described the effort to the NRC Staff at a meeting on September 13, 1979. Guidelines for inadequate core cooling without Reactor Coolant Pumps will be supplied to the District by October 31, 1979. Guidelines for additional postulations of inadequate core cooling will be supplied ':0 the District by December 21, 1979. These guidelines will be incorporated into Rancho Seco Operating Procedures by the end of the next refueling outage. Presently installed instrumentation at Rancho Seco is satis-factory. Should operating guideline developments show that complimentary instrumentation would be useful, the District will endeavor to install the instrumentation and complete procedures for its use during the refueling outage scheduled for 1981. The District will procure reactor coolant system saturation meters and install them during the next refueling outage. Each loop will have a meter with redundant temperature measure-ment inputs. 1199 250 2.1.4 Containment Isolation Provisions for PWR and BWRs NRC Position 1. All contairment isolation system designs shall comply 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 give careful reconsideration to the definition of essential and non-essential systems, shall identify each system deteimined 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 accordingly, and shall report the results of the re-eval-uation to the NRC. 3. All non-essential systems shall be automatically isolated by the containment isolation signal. 4. The design of control systems for automatic containment isolation valves shall be such that resetting the isola-tion signal will not result in the automatic reopening of containment isolation valves. Reopening of contain-ment isolation valves shall require deliberate operator action. Sacramento Municipal Utility District Commitment Rancho Seco containment isolation systems currently isolate on diverse parameters -- low reactor coolant system pressure (1600 psig) and high containment pressure (4 psig). In addition, reset of either of the containment isolation signals does not result in automatic reopening of containment isolation valves. As such, modification of the system with respect to isolation initiation or reset action is not necessary. The status of the current essential /non-essential system review indicates no system changes are necessary. The final review will be done by January 1,1980. 2.1.5.a Dedicated Penetration for External Recombiner on Post-Accident Purge Systems NRC Position Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atr:osphere 1199 251 should provide containment isolation systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 Cr~R Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system. Sacramento Municipal Utility District Commitment The District will review the Rancho Seco post-accident purge system and submit intentions to the NRC staff by January 1,1980. Should system modifications be required they will be completed during the refueling outage scheduled for 1981. 2.1.5.b Inerting BWR Containments NOT APPLICABLE TO RANCHO SEC0. 2.1.5.c Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant NOT APPLICABLE TO RANCHO SEC0 per September 26, 1979 Regio.Tl Meeting in Las Vegas. 2.1.6.a Integrity of Systems Outside Containment Like?y to Contain Radioactive Materials (Engineered Safety Systnis and Auxiliary Systems) for PWRs and BWRs NRC Position Applicants and licensees shall immediately implement a program to reduce leakage from systems outside cr'tainment that would or could contain Sighly radioactive flu durino a serious transient or accident to as-low-as pratical s. 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 out-side of containment. b. Measure actual leakage rates with system in operation and report them to the NRC. 2. Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as practical levels. This 1199 252 program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals. Sacramento Municipal Utility District Commitment The District will implement a continuing leak reduction program at Rancho Seco. As some system tests can be done only during plant shutdown, implementation will be completed during the refueling shutdown currently scheduled for January of 1980. 2.1.6.b Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May be Used in Post-Accident Operations NRC Position With the assumption of a post-acci4nt 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 and 100% of the core noble gas inventory 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 result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety eouip-ment may be unduly degraded by the radiation fields during post-accident operations of these systems. Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility. Sa cramer, t Municioal Utility District Commitment The District intends to perform a shielding and radiation design review of systems outside of the containment which might contain highly radioactive fluids in accident situations and of spaces about those systems. Portions of the following systems will be reviewed: Letdown System Makeup System Decay Heat System Reactor Building Spray System 1199 253 _7
Hydrogen Purge and tionitoring System Waste Gas System Reactor Coolant Sampling System The District expects to be able to complete a thorough review of the systems and surrounding spaces no sooner than February 15, 1980. The high source terms may require considerable plant modifica-tion. Until the nature of these modifications is defined by the study described above, and until the initial design for the modifications is known, a construction completion date can not be realistically supplied. The District will supply real-istic completion information when the initial design is complete. 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs NRC Position Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implemented in the short term: 1. The design shall provide for the automatic initiation of the auxiliary feedwater system. 2. The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function. 3. Testability of the initiating signals and circuits shall be a feature of the design. 4. The initiating signals and circuits shall be powered from the emergency buses. 5. Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the less of system function. 6. The A ^ motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses. 7. The automatic initiating signals and circuits sha designed so that their failure will not result in.ne loss of manual capability to initiate the AFWS from the control room. 1199 254 In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements. Sacramento Municipal Utility District Commitment The District believes that the Rancho Seco Auxiliary Feedwater satisfies the intent of the short term NRC position on this system. The point-by-point comparison below shows that it comes close to satisfying each detail of the NRC position. In those instances where it does not satisfy a detail, there are one or more backup capabilities to insure that water reaches the once through steam generation (OTSG). By January 1, 1980 the District feels that it could not design and install a system which would satisfy each detail of the NRC position, and yet operate as reliably as the existing system. In the long term, the District intends to install a sa'ety grade system independ-ent of the ICS. The system will be installed during the refueling outage in 1981. A comparison of the existing system to the NRC position is made below: 1. "The design shall provide for automatic initiation of the auxiliary feedwater system." In the existing system, one of the valves in the table below must open and one of the pumps in the table below must operate in order to provide water to an 0TSG. AUTOMATIC INITIATION SIGNAL Loss of Main Auxiliary Feed Feed Pump loss of SFAS Actuation System Component Discharge Pressure Four RCP's Channel Channel A B AFW Pump P-318 Starts X X X X AFW Pump P-319 Starts X X AFW Flow Control Valve FV-20527 Opens X X AFW Flow Control Valve FV-20528 Opens X X AFW Bypass Valve SFV-20577 Opens X AFW lypass Valve SFV-20578 Opens X 2. "The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function." i199 255 -9
i The function of the auxiliary feedwater system is to provide water to the OTSG's. Failure of any active ccmponent including initiating circuits and signals will not result in a loss of auxiliary feedwater system function. In addition, only two single failure cases would require operator action to provide water to the OTSG's. One case is when the turbine driven and pump P-318 fails af ter loss of offsite power. In that case, the motor drives on pumps P-318 and P-319 can be manually loaded onto the diesel generator supplied emergency bus. The other case is failure of the ICS. Failure of the ICS could prevent automatic opening of AFW flow control valves, FV-20527 and FV-20528 to the OTSG.(AFW pump starts are not controlled by the ICS). However, AFW bypass valves SFV-20577 and SFV-20578 are independent of the ICS and they can be opened from the control room. In addition, from the control room the air may be dumped from the pneumatic operators of AFW flow control valve FV-20527 and FV-20528 using circuitry independent of the ICS. Relieving air pressure fails the valves full open. 3. " Testability of the initiating signals shall be a feature of the design". Each of the auxiliary feed system components discussed for NRC position 1 above can be tested for initiation by simula-tion of appropriate signals. Prudence dictates that some of the tests be done only while shut down. 4. "The initiating signals and circuits shall be powered from the emergency buses." All AFW motive and initiation power comes directly from diesel backed emergency buses or from battery backed power supplies. Batteries are charged from diesel backed emergency buses. 5. "Itanual caoability to initiate the auxiliary feedwater system from the control ros... : hall be retained and shall be imple-mented so that a single failure in the manual circuits will not result in loss of system function." Manual initiation capability is not prevented by any single system failure. 6. "The A-C motor-driven pumps and valves in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses." Refer to the table in the discussion of NRC position I when following this discussion. 1199 256 AFW valves SFV-20577 and SFV-20578 are automatically supplied by diesel backed emergency buses with or without offsite power. AFW flow control valves FV-20527 and FV-20528 are pneumatically operated and have battery backed control power supplies. When called upon by an automatic initiation signal, Pump P-318 starts on its own turbine whether or not offsite power is available. If the turbine fails to start, a backup notor may be started by merely depressing a push button in the control room if offsite power is available. If offsite power is not available, a key-lock bypass switch in the control room must be turned before start using the pushbutton. P-319 starts automatically on its A-C motor when called upon by an automatic initiation signal if offsite power is avail-able. If offsite power is not available, a key-lock bypass switch must be turned prior to loading the pump motor onto the emeMency bus. 7. "The automatic initiating signals and circuits shall be designed so that their failure will not result in loss of manual capability to initiate the AFWS from the control room." No automatic initiation signal will prevent manual control of system components. 2.1.7.b Auxiliary Feedwater Flow Inidcation to Steam Generator for PWRs NRC Position Consistent with satisfying the requirements set forth in Gb.13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented: 1. Safety grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room. 2. The auxiliary feedwater flow instrument channels shall be powued from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9. Sacramento Municipal Utility District Commitment The auxiliary feedwater (AFW) system at Rancho Seco currently has control grade flow indication for flow to each 0TSG. OTSG level instruments act as backup instruments. Power to the AFW flow and OTSG level instruments is from a battery backed inverter. 1199 257 The District intends to install' safety grade AFW flow instrumentation during the 1981 refueling outage. 2.1.8.a Improved Post-Accident Sampling Capability NRC Position ~ A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess 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 of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design teatures or shielding should be provided to meet the criteria. A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the cap-ability to promptly quantify (less than 2 hours) certcin radio-isotopes that are indicators of the degree of core damage. Such radionuclidas are noble gases (which indicate clading failure), iodines and cesiums (whic'1 indicate high fuel tempera-tures), and non-volatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and com-ponents in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria. In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample, (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly, i.e., the boron sample analysis within an hour and the chloriae sample analysis with'n a shift. Sacramento Municipal Utility District Position By January 1,1980 the District will develop procedures to minimize exposure during high activity sampling and to extend the activity levels at which chemical and spectrum analysis can be done. In addition, the District will complete a design review of high activity level sampling and analysis capabilities by January 1,1980. Design improvements suggested by the review will be installed during the refueling outage in 1981. 2.1.8.b Increased Range of Radiation Monitor NRC Position The requirements associated with this recomendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instru-mentation to Follow the Course of an Accidnet," which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term. 1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident con-ditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest. a. Noble gas effluent monitors with an upper range capacity of 105 pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants. b. Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (ALARA) concentrations to a maximum of 105 uCi/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors shall overlap by a f&ctor of ten. 2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for tht accident condition shall be provided with sampling concu.ted by adsorption on charcoal or other media, followed by un-site laboratory analysis. 3. In-containment radiation level monitors with a maximum range of 10s rad /hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment. 1199 259 Sacramento f!unicipal Utility District, Commitment Noble Gas Effluent The District intends to develop procrdures for estimating high level noble gas releases by January 1,1980. The District will install high range noble gas monitoring equip-ment during the refueling outage scheduled for 1981. Iodine Gaseous Effluent The District intends to develop procedures for estimating high level ga5eous radiciodine releases by January 1,1980. Containment Radiatior. Mcnitors The District will install high range containment radiation level monitors during the refueling outage scheduled for 1981. 2.1.8.c Improved In-Plant Iodine Instrumentation MC Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions. Sacramento Municipal Utility District Position Equipment selection, procurement and calibration, procedure writing and personnel training is expected to take five months. Therefore, implementation of new radiciodine measurement equip-ment will be complete in five months. 2.1.9 Analysis of Design and Off-Normal Transients and Accidents NRC Position Analyses, procedures, and training addressing the following are required: 1. Small break loss-of-coolant accidents; 2. Inadequate core cooling; and 3. Transients and accidents. Some analysie requirements for small breaks have already been specified by the Bulletins and Orcers Task Force. These should be completed. In addition, pretes'. calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start \\\\99 260
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 verifica-tion of compliance with Appendix K of 10 CFR Part 50. In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods: 1. Low reactor coolant system inventory (two examples will be required - LOCA with forced flow, LOCA without forced flow). 2. Loss of natural circulation (due to loss of heat sink). These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists. The calcula-tions shall be carried out in real time far enough that all important phenomena and instrument indications are included. Each case should then be repeated taking credit for correct operator action. These additional cases will provide the basis for developing appropriate emergency procedures. These calculations should also provide the analytical t, asis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy. The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR. The analyses shall include a single active failure for each system called upon to function for a particular event. Consequential failures shall also be considered. Failures of the operators to perfom required control manipulations shall be given consideration for permutations of the analyses. Operator actions that could cause the complete loss of function of a safety system shall also be considered. At present, these anlyses 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 ansidered. The complete loss of auxiliary feed-water may be added to the fc W res being considered in the analysis of transients and accidents if it is concluded that more is needed in operator training 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 I could lead to core uncovery for some transients, and a computer calculation could provice information on the amount of time available for corrective action. Reactor simulators may pro-vide some information in defining the event trees and would be useful in studying the information available to the operators. The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate oper-ator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and pre-vention 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 perfomed by the NSSS vendors will be put in the form of emergency procedure guidelines and tnat the changes in the procedures will be implemented by eac5 licensee or applicant. In addition to the analyses performed by the reactor vendors, analyses of selected transients should be periormed by the NRC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor vendors. These comparisons together with comparisons to data, including LOF1~ small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emer-gency procedures. Sacramento Municipal Utility District Position The District is participating in a small break analysis effort which has been in progress for several months. Much of the effort is related to Inspection and Enforcement Bulletins 79-05A, 79-05B and 79-05C. Status of outstanding small break items is contained in the District's letter dated October 8,1979 to D. F. Ross, Jr. of the NRC staff. The District is also participating in development of guidelines for inadequate core cooling and for abnormal transients. Details of the effort were presented to the NRC staff at a meeting on September 13, 1979. 2.2.1.a Shift Suoervision Responsibilities NRC Position 1. The highest tevel of corporate maw qement of each licensee shall issue and periodically reissue a managerent directive that emphasizes the primary management responsibility of 1199 262 the shift super"isor for safe operation of the plant under all conditions ;n his shif t and that clearly establishes his command (.ies. 2. Plant procr.dures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of comand and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis snall be placed on the following: a. The responsibility and authority of the shift supervisor shall be to maintaii. the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when on duty in the control room. The idea shall be rein-forced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room, b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the shift supervisor shall be specified. c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified. 3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the manage-ment function the shift supervisor is to provide for assuring safety. 4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations. Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room. Sacramento Municipal Utility District Commitment The District intends to review the management and command aspects of shift supervision responsibilities and to change procedures accordingly. Procedural changes will be implemented by January 1, 1980. 1199 263
2.2.1.b Shift Technical Advisor NRC Position Each 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 site if qualified to per-form the advisor function for the various units. The shift technical advisor shall have a bachelor's degree or equivalent in r scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and lay-out, including the capabilities of instrumentation and 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 operations of the plant, including the review and evaluation of operating experience. S cramento Municipal Utility District Commitment 1 By January 1,1980 the District is considering assigning a sh:ft technical ad, or to the Rancho Seco site. The individual would be assigned to the position for 24 hour periods during which he would work and sleep on site such as discussed at the Shift Technical Advisor Topical Meeting in Bethesda, Maryland on October 12, 1979. The position would be covered during periods in which the reactor is not in cold shutdown or hot shutdown. Methods of upgrade of the position by January 1,1981 are also being studied. In conjunction with these considerations, the District is analyzing the alternatives presented in enclosure 2 to the September 13, 1979 NRC staff letter which implemented NUREG-0578. 2.2.1.c Shift and Relief Turnover Procedures NRC Position The licensees shall review and' revise as necessary the plant procedure for shift and relief turnover to e7re the fcP. awing: 1. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a ninimum, shall be included in the checklist: }}99 2b a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist). b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist); c. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of time in the degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a separate entry on the checklist). 2. Checklists or logs shall be provided for completion by the offgoing and oncoming auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance of test that by themselves could degrade a system critical to the prevention and mitigation of operational i aansients and accidents or initiace an operational transients (what to check and criteria for acceptable status shall be included on the checklist); and 3. A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments.) Sacramento flunicipal Utility District Commitment The District will review and implement shift turnover procedures by January 1, 1980. 2.2.2.a Control Room Access NRC Position The licensee shall make provisions for limiting access M the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations super-visor, shif t s-visor, and control room operators), to technical advisvrs who may be requested or required to support the operation, and to predesignated NRC personnel. Provisions shall include the following: 1199 265 1. Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access. 2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of comunication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room. Sacramento Municipal Utility District Position The District will review control room access policy and add provisions for control by January 1,1983. 2.2.2.b Onsite Technical Support Center NRC policy Each operating nuclear power plant shall maintain an ensite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledge-able of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and loca-tion of the technical support center. Records that pertain to the as-built conditions and layout of structures, systems and compor.e.its shall be stored and filed at the site and accessible to the technical support center under emergency conditions. Examples of such records include system descriptions, general arrangement drawings, piping and instru-ment diagrams, piping system isometrics, electrical schematics, wire and cable lists, and single line electrical diagrams. It is not the intent that all records described in ANSI N45.2.9-1974 be stored and filed at the site and accessible to the technical support center under emergency conditions; however, as stated in that standard, storage systems shall provide for accurate retrieval of all pertinent inforr.:ation without undue delay. \\\\99 266 Sacramento Municipal Utility District Commitment The District will have an interin onsite technical supp:vt center by January 1,1980.. Preliminary plans and schedules for upgrade of the center will be available by March 1,1980. 2.2.2.c Onsite Operational Support Center NRC Position An area to be designated as the onsite operational support center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel will report in an emergency situation. Communications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of communica-tion and management. Sacramento Municipal Utility District Commitment The District currently has onsite operation.;cport centers. NUREG-0578 Implementing Letter CONTAINMENT PRESSURE MOMITOR of September 13, 1979 Encl. 3, Item 3(1) NRC Position A continuous indication of containment pressure shall be pro-vided in the control room, tieasurement and indication capa-bility shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus five psig for all containments. Sacramento Municipal Utility District Commitment The listrict intends to install a safety grade containment pressure monitor. Installation will be completed during the refuel 4.ig outage in 1981. 1199 267 NUREG-0578 Implementing Letter CONTAINMENT HYDROGEN CONCENTRATION MONITOR of September 13, 1979 Encl. 3, Item 3(2) NRC Position A continuous indication of hydrogen concentration in the con-tainment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 107. hydrogen concentration under both position and negative ambient pressure. Sacramento Municipal Utility District Commitment The District now has.a containment hydrogen monitor at Rancho Seco. We currently do not know of a safety grade hydrogen monitor which covers the requirements listed above. The District will attempt to obtain a qualified monitor prior to January 1,1981. If the device is obtained, it will be installed during the refueling outage scheduled for 1981. NUREG-0578 Implementing Letter CONTAINMENT WATER LEVEL of Septembec 13, 1979, Encl. 3, Item 3(3) NRC Position A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for PWR's and cover the range from the bottom to the top of the containment sump. Also for iWR's a wide range instrument shall be provided and cover the range from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity. Sacramento Municipal Utility District Position The District will install reactor building sump and wide level indication during the refueling outage scheduled for 1981. NUREG-0578 Implementing Letter UPGRADE EMERGENCY PLANS of September 13, 1979 Encl. 7, Item (1) NRC Position Upgrade licensee emergency plans to satisfy Regulatory Guide 1.101, with special attention to the development of uniform action level criteria based on plant parameters. l99 2bb Sacramento Municipal Utility District Cocmitment Reformatting, expansion and assessment of plant parameters will take five months. Therefore, implementation is scheduled for April 1,1980. NUREG-0578 Implementing Letter INSTRUMENTATION IMPROVEMENTS of September 13, 1979 Encl. 7, Item (2) NRC Position Assure the implementation of the related recommendations of the Lessons Learned Task Force involving instrumentation to follow the course of an accident and relate the infomation t rovided by this instrumentation to the emergency plan action 11vels. This will include instrumentation for post-accident
- ampling, high range radioactivity monitors, and improved in-plant radiciodine instrumentation. The implementation of the Lessons Learned Task Force's recommendations on instru-mentation for detection of inadequate core cooling will also be factored into the emergency plan action level criteria.
Sacramento Municipal Utility District Commf tyent The District's commitment in this area is contained within the response to items 2.1.8.a, b and c. It is located in previous pages of this attachment. NUREG-0578 Implementing Letter EMERGENCY 0 % ATIONS CENTER FOR of September 13, 1979 FEDERAL, STATE AND LCCAL PERSONNEL Encl. 7, Item (3) NRC Position Determine that an emergency operations center for Federal, State and local personnel has been established with suitable communications to the plant, and the upgrading of the facility in accordance with the Lessons Learned Task Force's recommenda-tion for an in-plant technical support center is underway. Sacramento Municipal Utility District Commitment The District expects that an emergency operations control center site will be selected by July 1, 1980. The center will be up-graded in conjunction with the onsite technical support center as described in the response to NUREG-0578 item 2.2.2.b which was described in a previous page of this attachment. 1199 269
NUREG-0578 Implementing Letter 0FFSITE MONITORING CAPABILITIES of September 13, 1979 Encl. 7, Item (4) NRC Position Assure that improved licensee offsite monitoring capabilities (including additional thermoluminescent dosimeters or the equivalent) have been provided for all sites. Sacramento Municipal Utility District Commitment The District expects to complete the upgrrde of offsite monitoring capabilities by July 1,1980. NUREG-0578 Implementing Letter EMERGENCY PIAN COMPATABILITY of September 13, 1979 Encl. 7, Item (5) NRC Position Assess the relationship of State / local plans to the H~ sees' and Federal plan; so as to assure the capability to take appropriate emergency actions. Assure that this capability will be extended to a distance of ten miles. This item will be performed in conjunction with the Office of State Programs and the Office of Inspection and Enforcement. Sacramento Municipal Utility District Commitment The District will assess compatability of Rancho Seco emer-gency plans with state and local emergency plans by July 1, 1980. The assessment will be made using current cri' aria. The District will make the same type of assessment for upgraded criteria by January 1,1981. NUREG-0578 Implementing Letter EMERGENCY PLAN EXERCISES of September 13, 1979 Encl. 7, Item (6) NRC Pos; tion Require test exercises of approved emergency plans (Federal, State, local and licensees), review plans for such exercises, and participate in a limited number of joint exercises. Tests of licensee plans will be required to be conducted as soon as practical for =11 facilities and before reactor startup for new licosees. Exercises of State plans will be perfonned in con-junction with the concurrence reviews of the Office of State }l99 Programs. As a preliminary planning bases, assume that joint test exercises involving Federal, State, local and licensees will be conducted at the rate of about ten per year, which would result in all sites being exercised once each five years. Revised planning guidance may result from the ongoing rulemaking. Sacramento Municipal Utility District Commitment The District intends to conduct a test of its emergency plan by July 1,1980 and by the same date it will cooperate with the state in test of their emergency plan. The District will participate in a joint (Federal, State and local) emergency plan exercise within five years. NUREG-0578 Implementing Letter REACTOR C0OLANT SYSTEM VENTS of September 13, 1979 incl. 4 NRC Position Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vents remotely operated from the control room. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appenoix A to 10 CFR Part 50 General Design Criteria. In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of inadvertent actuation. Each applicant and licensee shall,,rovide 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 accidents initiated by a break in the vent pipe. The results of the analyses should be demon-strated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46. 2. Analyses demonstrating that the direct venting of non-condensable gases with perhaps high hydrogen concentra-tions does not result in violation of combustible gas concentration limits in containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev.1), and Standard Review Plan Section 6.2.5. .+ 3. Procedural guidelines for the operators' use of the vents. The information available to the operator for initiating or terminating vent usage shall be discussed. Sacramento Municipal Utility District Commitment The District will install reactor coolant system vents during the 1981 refueling outage. Procedures for venting will be completed prior to startup after the 1981 outage. The vent system design and analysis justifying the design will be complete by May 1,1980. 1199 272 8 }}