ML18081A359
| ML18081A359 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/12/1979 |
| From: | Librizzi F Public Service Enterprise Group |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910160498 | |
| Download: ML18081A359 (52) | |
Text
_,
t PS~G Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /430-7000 October l2, l979 Director of Nuclear Reactor Regulation
- u. s. Nuclear Regulatory Conunission Washington, D. C.
20555 Attention:
Mr. D. G. Eisenhut, Acting Director Division of Operating Reactors Gentlemen:
REQUEST FOR ADDITIONAL INFORMATION RESULTING FROM THE THREE MILE ISLAND 2 ACCIDENT NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 Public Service Electric and Gas Company hereby submits its response to your letter of September 13, 1979, requesting additional information in the areas of Lessons Learned and Emergency Preparedness as a result of the Three Mile Island 2 accident.
This information is contained in Attachments 1 and 2 to this letter, respectively.
The attachments to this letter are applicable to both No. 1 and Nb. 2 Units of the Salem Nuclear Generating Station and are responsive to your concerns.
This information has been transmitted under a separate cover to the Acting Director, Division of Project Management for the Salem No. 2 Unit (Docket No. 50-3ll).
Attachments The Energy People Very truly yours, JI F. P. Li~
General Manager -
Electric Production 95-2001 (400M) 9-77
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ATTACHMENT 1 LESSONS LEARNED Salem 1 & 2
l l
Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators 1n PWRs (Section 2.1.l)
NRC POSITION Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:
Pressurizer Heater Power Supply
- 1.
The pressurizer heater power supply design shall pro-vide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predeter-mined 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 pres-surizer 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 suf-ficient capacity for the connection of the pressuriz-er 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 inter-faces with the emergency buses shall be accomplished through devices that have been qualified in accord-ance with safety-grade requirements.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators
- 1.
Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
M P79 54 01/l Salem l & 2
,i
- 2.
Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
- 3.
Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.
- 4.
The pressurizer level indication instrument channels shall be powered 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.
RESPONSE
The Salem design is such that it has the capability to manually connect approximately 400 kW of pressurizer heaters from one backup group to the emergency power source.
This connection is accomplished by an installed manually operated interlocked transfer scheme between the pressurizer heaters and the "A" diesel generator.
An additional backup group of heaters, approximately 400 kW, is being provided with the capability to be connected in a similar manner to the "C" diesel generator to provide redundancy.
Procedures will be established to address the transfer to the emergency power source.
Motive and control power interfaces with the emergency power source will satisfy safety-grade requirements.
M P79 54 01/2 Salem 1 & 2
Analyses performed by Westinghouse indicates that approxi-mately 150 kW of pressurizer heaters are necessary to maintain natural circulation at hot standby conditions.
The pressurizer PORVs and their associated block valves are powered from the emergency power source.
Motive and control power interfaces with the emergency power source satisfy safety-grade requirements.
Pressurizer level indication instrument channels are powered from the vital instrument buses.
The modifications described above will be completed in accordance with the Category A implementation schedule.
M P79 54 01/3 Salem l & 2
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Performance Testing for BWR and PWR Relief and Safety Valves (Section 2.1.2)
NRC Position Pressurized water reactor and boiling water reactor li-censees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under ex-pected operating conditions for design basis transients and accidents.
The licensees and applicants shall determine the expected valve operating conditions through the use of anal-yses 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 safety and relief valves are maxi-mized.
Test pressures shall be the highest predicted by conventional safety analysis procedures.
Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and supports as well as the valves themselves.
Response
PSE&G is participating in generic industry wide qualification testing program for the conditions described above on the particular model Reactor Coolant System safety and relief valves used on the Salem Units.
This program is presently being defined under the guidance of the Electric Power Research Institute (EPRI) and the Westinghouse Owners Group.
The program, description and schedule will be submitted in accordance with the Category A implementation schedule.
M P79 54 01/4 Salem 1 & 2
Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and BWRs (Section 2.1.3.a)
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 indication of flow in the discharge pipe.
Response
Open and closed position indication for each PORV and PORV block valve is provided on the main control console.
In addition, an overhead annunciator alarm is provided in the Control Room to indicate whether any PORV is not fully closed.
The pressurizer safety valves will be provided with limit switches which will provide a "not fully closed" visual display and audible alarm in the Control Room.
This modification will be completed in accordance with the Category A implementation schedule.
M P79 54 01/5 Salem 1 & 2
Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs (Section 2.1.3.b)
NRC Position
- 1.
Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with cur-rently 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 form the basis for operator training and procedure develop-ment shall be provided pursuant to another short-term requirement, "Analysis of Off-Normal Conditions, Includ-ing Natural Circulation" (see Section 2.1.9).
In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of cool-ant saturation and condition.
Operator instruction as to use of this meter shall include consideration that is not to be used exclusive of other related plant para-meters.
- 2.
Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-inter-pret indication of inadequate core cooling. A descrip-tion of the functional design requirements for the system shall also be included.
A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a*
schedule for installing the equipment shall be provided.
Response
The existing instrumentation available in the Control Room is sufficient to recognize inadequate core cooling. The indications available for determination of core heat removal are:
- a.
RCS delta T less than full load delta T.
- b.
RCS or core exit thermocouple temperatures constant or decreasing.
M P79 54 01/6 Salem l & 2
Response (Cont'd)
- c.
Stearn generator pressure constant or decreasing at a rate equivalent to the rate of decrease of RCS tempera-tures while.maintaining steam generator level with continuous auxiliary feedwater.
A further guide for recognition of inadequate core cooling is the recent addition of a computer/CRT display for sub-cooling.
The significant parameters displayed at operator request are reactor coolant differential pressure (P actual P saturated) and differential temperature (T saturated -
T actual).
Alarms are set for pressure differential less than 200 psi and temperature differential less than S0°F.
The computer program is predicated on the hottest in-core thermocouple reading.
The CRT matrix of in-core thermo-couples will display the location of the hottest in-core thermocouple.
Operating procedures have been revised to address the use of this computer program to monitor the margin of subcooling in the Reactor Coolant System.
Westinghouse, under the direction of the Westinghouse Owners Group, is performing further analyses to aid in selection of more direct indicators of inadequate core cooling, and to serve as a basis for augmented emergency procedures.
PSE&G is participating in the Westinghouse Owners Group and will fully assess the results of these analyses with respect to any recommended modifications, including reactor vessel water level indication.
Procedures will be written M P79 54 01/7 Salem 1 & 2
or revised, as necessary, to reflect any modifications or information that may be used to recognize inadequate core cooling.
M P79 54 01/8 Salem 1 & 2
Containment Isolation Provisions for PWRs and BWRs (Section 2. l. 4 )
NRC Position
- 1.
All containment 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 determined 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-evaluation to the NRC.
- 3.
All non-essential systems shall be automatically iso-lated 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.
Response
- 1.
The containment isolation system complies with the requirements for isolation initiation by diverse para-meters as described in Section 5.4 of the FSAR.
A num-ber of isolation signals are provided for valve closure.
Each signal is indicative of certain operating condi-tions and is generated by diverse input parameters.
The isolation signals and their input parameters are as follows:
Containment Isolation -
Phase A
- a.
Manual Actuation
- b.
Automatic Safety Injection Actuation Logic M P79 54 01/9 Salem l & 2
-,t
- c.
High Containment Pressure
- d.
Low Pressurizer Pressure
- e.
High Differential Pressure Between Steam Lines
- f.
High Steam Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
Containment Isolation -
Phase B
- a.
Manual Actuation
- b.
Automatic Actuation Logic
- c.
High-High Containment Pressure Containment Ventilation Isolation
- a.
Manual Actuation
- b.
Automatic Safety Injection Actuation Logic
- c.
High Containment Pressure
- d.
Low Pressurizer Pressure
- e.
High Differential Pressure Between Steam Lines
- f.
High Steam Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
- g.
High Containment Radiation - Particulate
- h.
High Containment Radiation -
- i.
High Containment Radiation -
Gaseous Main Steam Line Isolation
- a.
Manual Actuation
- b.
Automatic Actuation Logic
- c.
High-High Containment Pressure
- d.
High Steam Line Flow Coincident with Low Steam Line pressure or Low-Low Tavg.
M P79 54 01/10 Salem 1 & 2
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Fe~dwater Isolation
- a.
Manual Actuation
- b.
Automatic Actuation Logic
- c.
High Containment Pressure
- d.
Low Pressurizer Pressure
- e.
- High Differential Pressure Between Stearn Lines
- f.
High Stearn Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
- g.
High-High Stearn Generator Water Level
- h.
Reactor Trip Coincident with Low Tavg.
- 2.
The containment isolation system isolates those system which are not required for the mitigation of* accidents specified in Section 14 of the FSAR~
A review of Salem design has demonstrated conformance with these require-ments.
The valves and systems isolated by the various isolation signals are indicated in Table 5.4-1 and Figures 5.4-1 through 5.4-27 of the FSAR.
All lines penetrating the containment are shown in these figures along with their isolation provisions.
All non-essential systems are either automatically isolated upon a containment isola-tion signal, or provided with non-return check valves, or closed during power operation and under administrative control.
Essential systems are not isolated since they M P79 54 01/11 Salem 1 & 2
are required to perform functions needed to maintain the plant in a safe condition following an accident.
These essential systems are as follows:
Residual Heat Removal - part of Safety Injection Safety Injection Containment Fan Coolers -
Service Water Steam Supply to Auxiliary Feedwater Pump Turbine Main Steam Atmospheric Relief Auxiliary Feedwater Charging -
Portion for Safety Injection It is anticipated that additional review of isolation system design criteria will be undertaken by the Westinghouse Owners Group and that any applicable changes will be implemented.
- 3.
As stated previously, all non-essential systems are either isolated upon containment isolation signals, or provided with non-return check valves, or closed during power operation and under administrative control.
- 4.
A review of the containment isolation valve control systems has been performed to verify that the valves remain closed upon resetting of the isolation signal until the operator takes deliberate action to reposi-tion them.
As a result of the review, design changes M P79 54 01/12 Salem 1 & 2
have been initiated to modify the control circuitry in two areas.
The results of this review, including a description of the two areas where modifications were deemed warranted, were submitted on July 13, 1979 in response to IE Bulletins79-06A.
Implementation of the design changes will be completed in accordance with the Category A implementation schedule.
M P79 54 01/13 Salem 1 & 2
Dedicated Penetrations for External Recombiners or Post-Ac-cident Purge Systems (Section 2.1.5.a)
NRC Position Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmos-phere should provide containment isolation systems for ex-ternal recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single fail-ure requirements of General Design Criteria 54 and 56 of Ap-pendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner of purge system.
Response
Each Salem unit incorporates two redundant, physically separated, permanently installed electric hydrogen re-combiners, located inside the reactor containment, as describe in Section 14.3.6 of the FSAR.
Each recombiner is capable of maintaining post-accident hydrogen concentration in the containment below the lower limit of flammability in air of 4%, in accordance with the assumptions used in the FSAR.
M P79 54 01/14 Salem 1 & 2
Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant (Section 2.1.5.c)
NRC Position
- 1.
All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.
- 2.
The procedures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering shielding requirements and personnel exposure limitations as demonstrated to be necessary in the case of TMI-2.
Response
Post-accident hydrogen control capability is described in the response to Item 2.1.5.a.
Procedures for use of the hydrogen recombiners have been reviewed and revised as required in response to IE Bulletin 79-06A.
M P79 54 01/15 Salem 1 & 2
Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs (Section 2.1.6.a)
NRC Position Applicants and licensees shall immediately implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.
This program shall include the following:
- 1.
Immediate Leak Reduction
- a.
Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
- b.
Measure actual leakage rates with system in operation and report them to the NRC.
- 2.
Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels.
This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.
Response
The Salem design has been reviewed to ascertain capability for detection and contact of leakage in systems that would or could contain high post-accident radioactivity invento-ries.
A major portion of the fluid systems which would or could contain high post-accident radioactivity inventories are equipped with valve lantern leakoff connections that are hard-piped to the Liquid Radwaste System.
Additionally, the majority of relief valves in systems containing radioactive fluids outside containment are piped to the pressurizer relief tank inside tbe containment.
M P79 54 01/16 Salem 1 & 2
The Salem design conforms with the intent of this position in that leakoffs from valves larger than 3-inches and equipment leakoff s are all hard-piped to the Liquid Radwaste System.
The hard-piping, while restricting the ability to measure leakage, does reduce the amount of airborne activity that would result from contaminated leakage.
In order to ensure the continued integrity of these systems, PSE&G will continue to implement the provisions of ASME XI-1974.
This code requires service pressure vessel leak tests of Nuclear Class I Systems during each refueling outage, Nuclear Class II Systems every 10 years (these are performed every 3-1/3 years, however, as an extension of commitments in response to IE Bulletin 76-06) and Nuclear Class III Systems every 3-1/3 years.
Additionally, waste gas handling systems will be incorporated into an equivalent program.
M P79 54 01/17 Salem 1 & 2
Design Review of Plant Shielding of Spaces for Post-Accident Operations (Section 2.1.6.b)
NRC Position With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials.
The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.
Each licensee shall provide for adequate access to vital areas and 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.
Response
A radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials is being performed.
This review en-compasses vital areas and equipment and includes equipment qualification for post-accident radiation fields and access-ibility to personnel for operation and maintenance.
A preliminary design review will be completed and design modifications will be identified in accordance with the Category A implementation schedule.
M P79 54 01/18 Salem 1 & 2
Automatic Initiation of the Auxiliary Feedwater System for PWRs (Section 2.1.7.a)
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 loss of system function.
- 6.
The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.
- 7.
The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWs from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements.
Responses The Auxiliary Feedwater System is described in Section 10.2.1.2 of the FSAR.
The system is designed to Class IE criteria and is powered from the emergency power source.
M P79 54 01/19 Salem 1 & 2
Automatic initiation of the Auxiliary Feedwater System is provided by the following signals.
Motor Driven Pumps
- a.
- b.
Loss of Main Feed
- c.
Low-Low Level in One Steam Generator
- d.
Safeguards Sequence Signal Turbine Driven Pump
- a.
- b.
Low-Low Level in Two Steam Generators
- c.
4kV Bus Undervoltage Manual initiation of the systems may be accomplished from either the Control Room, or locally at the pumps.
The system and its components are designed for single failure considerations and are testable.
M P79 54 01/20 Salem 1 & 2
Auxiliary Feedwater Flow Indication to Stearn Generators for PWRs (Section 2.1.7.b)
NRC Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:
- 1.
Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
- 2.
The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
Response
Safety-grade indication of auxiliary feedwater flow to each steam generator is provided in the control room.
These indicating channels are designed to the same criteria as the protection system indicators.
M P79 54 01/21 Salem 1 & 2
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Improved Post-Accident Sampling Capability (Section 2.1.8.a)
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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 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 features or shielding should be provided to meet the criteria.
A design and operational review of the radiological spectrum analysis facilities shall be perfomed to determine the capability to promptly quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radioisotopes that are indicators of the degree of core damage.
Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), 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 components in the auxiliary building and possible contamination and direct radiation from airborne effluents.
If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Procedures shall be 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 chloride sample analysis within a shift.
Response
A design and operational review of the containment atmos-phere sampling systems has been performed to determine the capability of personnel to obtain a sample under accident M P79 54 01/22 Salem 1 & 2
conditions within the time and exposure constraints identi-fied above.
This review has indicated that certain design modifications are warranted.
A description of these design modifications will be completed in accordance with the Category A implementation schedule and implemented by January 1, 1981.
A similar design and operational review of the Reactor Coolant Sampling System is in progress.
This review, as well as a description of design modifications, will be completed in accordance with the Category A implementation schedule and implemented by January 1, 1981.
A design and operational review of the radiological spectrum analysis equipment is being performed to determine the capability to quantify certain radioisotopes that are indicators of degree of core damage.
A review of the physical location of this equipment and its environmental conditions is being performed in conjunction with the radiation and shielding design review described in the response to Section 2.1.6.b.
These reviews will be completed in accordance with the Category A implementation schedule.
Procedures will be prepared, or revised, as necessary, to incorporate any design modifications that are made to improve post-accident sampling capability.
PSE&G is participating in the Westinghouse Owners Group for the development of procedures for analysis of samples.
M P79 54 01/23 Salem 1 & 2
Increased Range of Radiation Monitors (Section 2.1.8.b)
NRC Position The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instrumentation to Follow the Course ~f an Accident," 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 conditions 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 of 105 uCi/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 a minimum of 10-7 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 factor 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 the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
- 3.
In-containment radiation level monitors with a maximum range of 108 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.
M P79 54 01/24 Salem l & 2
Response
- 1.
The plant vent gaseous monitors have the following detection range capabilities:
Unit 1:
5x10-6 to 5x10-l uCi/cc Xe-133 Unit 2:
lx10-6 to lx102 uCi/cc Xe-133 Design changes have previously been initiated, and equipment purchased to upgrade the detection range capability of Unit 1 to that of Unit 2.
Further design modifications are presently being evaluated to provide the gaseous monitors with a detection range capability of lo-7 to 105 uCi/cc Xe -
133.
The modified system will utilize multiple monitors with the required overlap to meet the above criteria.
An alternate consideration is the use of a detector with a range of 104 uCi/cc if the containment exhaust is diluted by at least a factor of 10.
These modifications will be completed by January 1, 1981.
- 2.
The Salem design provides for iodine sampling by adsorption on charcoal cartridges, followed by onsite laboratory analysis.
- 3.
The containment high range monitors presently have the following maximum detection ranges:
unit 1:
104 R/hr.
unit 2:
107 R/hr.
M P79 54 01/25 Salem 1 & 2
One monitor is provided for each unit.
The Unit 2 monitor has undergone environmental qualification to demonstrate proper operation in an accident environment.
In addition, this monitor has been calibrated in a special test facility to verify proper readings in high radiation fields.
In order to meet the requirement for monitors with a range of 108 R/hr, we are investigating the possibility of shielding the existing Unit 2 monitor.
An alternate consideration is the use of the existing 107 R/hr (gamma) monitor.
An additional monitor with similar range capability will be installed to meet redundancy requirements.
Installation of new monitors for both units will be completed by January 1, 1981.
M P79 54 01/26 Salem l & 2
Improved In-Plant Iodine Instrumentation (Section 2.1.8.c)
NRC Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions.
Response
Sufficient instrumentation for iodine concentration monitoring throughout the plant under accident conditions will be provided in accordance with the Category A implementation schedule.
Additionally, procedures will be prepared and implemented for use of the iodine monitoring equipment and training will be provided for personnel assigned to monitor iodine concentrations.
M P79 54 01/27
-27 Salem l & 2
Analysis of Design and Off-Normal Transients and Accidents (Section 2.1.9)
NRC Position Analysis, procedures, and training addressing the following are required:
- 1.
Small break loss-of-coolant accidents;
- 2.
Inadequate core cooling; and
- 3.
Transients and accidents.
Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force.
These should be completed.
In addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K of 10 CFR Part 50.
In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:
- 1.
Low reactor coolant system inventory (two examples will be required -
LOCA with forced flow, LOCA without forced flow).
- 2.
Loss of natural circulation (due to loss of heat sink).
These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists.
The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included.
Each case should then be repeated taking credit for correct operator action.
These additional cases will provide the basis for developing appropriate emergency procedures.
These calculations should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b).
M P79 54 01/28 Salem 1 & 2
The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR.
The analyses shall include a single active failure for each system called upon to function for a particular event.
Consequential failures shall also be considered.
Failures of the operators to perform required control manipulations shall be given consideration for permutations of the analyses.
Operator actions that could cause the complete loss of function of a safety system shall also be considered.
At present, these analyses need not address passive failures or multiple system failures in the short term.
In the recent analysis of small break LOCAs, complete loss of auxiliary feedwater was considered.
The complete loss of auxiliary feedwater may be added to the failures being considered in the analysis of transients and accidents if it is concluded that more is needed in operator training beyond the short-term actions to upgrade auxiliary feedwater system reliability.
Similarly, in the long term, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-term analyses.
The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations could be used to provide important quantitative information not available from an event tree.
For example, failure to initiate high-pressure injection could lead to core uncovery for some
- transients, and a computer calculation could provide information on the amount of time available for corrective action.
Reactor simulators may provide some information in 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 operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.
The information derived from the preceding analyses shall be included in the plant emergency procedures and operator training.
It is expected that analyses performed by the NSSS vendors will be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant.
M P79 54 01/29
-29 Salem 1 & 2
In addition to the analyses performed by the reactor vendors, analyses of selected transients should be performed 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 LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures.
Instrumentation to Monitor Containment Conditions During the Course of an Accident Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain containment conditions during the course of an accident, the following requirements shall be implemented:
- 1. A continuous indication of containment pressure shall be provided in the control room.
Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus five psig for all containments.
- 2. A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room.
Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
- 3. A continuous indication of containment water level shall be provided in the control room for all plants.
A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump.
Also for PWRs, a wide range instrument shall be provided and cover the range from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity.
For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool.
The containment pressure, hydrogen concentration and wide range containment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testability.
The narrow range containment water level measurement instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.
M P79 54 01/30 Salem l & 2
Installation of Remotely Operated High Point Vents in the Reactor Coolant System 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 Appendix 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 provide the following information concerning the design and operation of these high point vents:
- 1.
A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe.
The results of the analyses should be demonstrated to be acceptable in accordaDce with the acceptance criteria of 10 CFR 50.46.
- 2.
Analyses demonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations 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.
Response
PSE&G is a member of the Westinghouse Owners Group and is actively supporting the generic analysis work described above.
This analysis work will be completed on a schedule compatible with the industry effort.
Emergency procedures will be prepared or revised, as appropriate, to incorporate the results of the analysis work performed.
M P79 54 01/31 Salem 1 & 2
Containment pressure indication will be modified to meet the requirements identified above by January 1, 1981.
Containment water level indication meeting the requirements identified above will be provided by January 1, 1981.
Containment hydrogen indication is presently installed in the Salem plant.
Calibration adjustments necessary to meet the requirements identified above willAb'e*-'C'OThp1eted by January 1, 1981.
Reactor Coolant System high point venting is being evaluated by the Westinghouse Owners Group.
The design of a RCS venting system will be submitted in accordance with the Category A implementation schedule.
M P79 54 01/32 Salem 1 & 2
Shift Supervisor's Responsibilities (Section 2.2.1.a)
NRC Position
- 1. The highest level of corporate management of each licensee shall issue and periodically reissue a manage-ment directive that emphasizes the primary management re-sponsibility of the shift supervisor for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties.
- 2. Plant procedures shall be reviewed to assure that the duties, responsibilities, and autnor.Itf,;z'.f* the **:shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel.
Particular emphasis shall be placed on the following:
- a.
The responsibility and authority of the shift super-visor shall be to maintain 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 reinforced that the shift supervisor should not be-come totally involved in any single operation in times of emergency when multiple operations are re-quired in the control room.
- b.
The shift supervisor, until properly relieved, shall remain in the control room at all times during acci-dent situations to direct the activitiei 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 management function the shift supervisor is to provide for assuring safety.
M P 79 54. 01/33 Salem 1 & 2
- 4. The administrative duties of the shift supervisor shall b~ 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.
Response
A written directive describing and emphasizing the primary management responsibilities of Shift Supervisors and establishing their command duties was placed in effect September 12, 1979.
Using the guidance of this directive, plant procedures and the requalification program.will be revised as necessary in accordance with the Category A implementation schedule.
Shift administrative activities are in *the process of being reviewed to determine which duties should be delegated to personnel that are not on duty in the Control Room.
Duties which are found to detract from the Shift Supervisor's responsibility for safe operation of the plant will be reassigned in accordance with the Category A implementation schedule.
M P79 54 01/34 Salem 1 & 2
Shift.Technical Advisor (Section 2.2.1.b)
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 a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and acc.i r'len.t.s...... The shift tech-nical advisor shall also receive train+/-.nzj,~::,i:<ri.p;ian:t:*design and layout, 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.
Response
The position of Shift Technical Advisor will be established, with personnel assigned, in accordance with the Category A implementation schedule.
Personnel initially selected for the position of Shift Technical Advisor will have a bach-elors degree, or equivalent, in a scientific or.engineering discipline.
A training program is in the process of being developed which will cover plant design and layout, plant transient and accident analysis, plant transient and accident response and control room instrumentation and controls capabilities.
This training program will be placed in use during 1980 and by January 1, 1981, all personnel assigned at that time as Shift Technical Advisor will have completed this program.
M P79 54 01/35 Salem 1 & 2
- rt is our intent.in the long term to qualify the Senior Shift Supervisors to assume the on-shift functions of the Shift Technical Advisor.
The routine duties and assignments of the Shift Technical Advisor involving engineering evalua-tion of day-to-day plant operations from a safety point of view would be performed by an engineering support group.
M P79 54 01/36 Salem 1 & 2
Shift and Relief Turnover Procedures (Section 2.2.1.c)
NRC Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:
- 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 minimum, shall be included in the checklist:
- 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 compcinenets, the length of time in.the degraded mode s*hall 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 transients and accidents or initiate an operational transients (what to check and criteria for acceptable status shall be included on the checklist); and M P79 54 01/37 Salem 1 & 2
- 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) *.
Response
The Salem operating logs contain check lists which provide oncoming and offgoing shifts with a status of critical plant parameters.
In order to provide for a more formalized shift turnover, a program ha:s::.iJ~e:n established to ensure that the oncoming shift log and plant status review has been properly accomplished.
An *adequate evaluation system, which provides for a monthly management inspection to determine the quality of shift operations is already in use at Salem.
This inspection consists of verification of operator under-standing of equipment status and plant alarms, direct observation of the conduct of operations in the Control Room, and verification of tagging requests.
The tagging request verification is an independent check of the system lineup as modified by the tagging request and is a review of Control Room documentation of that lineup.
M P79 54 01/38 Salem 1 & 2
Co~,~rol *Room Access !ection 2. 2. 2 ~a)
NRC Position The licensee shall make provisions for limiting access to the control room to those 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 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 ~n 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 communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.
Response
The Control Room layout presently provides for a single point of access to both Unit 1 and 2 Control Rooms.
This vital area access point is card key controlled and monitored by closed circuit television.
Only those personnel who are required in the Control Room have unescorted access to the Control Room area.
Additional administrative controls will be established to further restrict access to the Control Rooms to only those personnel who can demonstrate an actual need to be there.
M P79 54 01/39 Salem 1 & 2
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These additional controls will be instituted in accordance with the Category A implementation schedule.
Responsibilities and lines of authority in the Control Room are addressed in the response to Section 2. 2.1. a.
M P79 54 01/40 Salem 1 & 2
Onsite Technical Support Center (Section 2.2~2.b)
NRC Position Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable 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 location of the technical support center.
A complete set of as-built drawings and other records, as described in ANSI N45.2.9-1974, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions.
These documents shall include, but not be limited to, general arrangement drawings, P&IDs, piping system isometrics, electrical schematics, and photographs of components installed without layout specifications (e.g., field-run ~iping and instrument tubing).
Response
A temporary onsite Technical Support Center (TSC) will be established in accordance with the Category A implementation schedule.
Design information for a permanent onsite Tech-nical Support Center will be available by January 1, 1980.
The temporary TSC will be located in our Clean Facilities Building which is located adjacent to No. 1 Unit and is equally accessible by personnel to the Control Rooms for both No. 1 and 2 Units.
The Clean Facilities Building is within the plant security boundary.
Direct telephone com-munication will be provided between the TSC and each Control Room and the onsite Operational Support Center.
In addi-tion, telephone communications will be available with appro-priate offsite agencies and emergency operations centers.
M P79 54 01/41 Salem 1 & 2
The room utilized for the TSC will contain at least 500 sq.
ft. and will be in close proximity to the existing Technical Document Room and the existing off ice provided for NRC per-sonnel.
Documents described in ANSI N45.2.9-1974 are stored within the Technical Document Room.
We are investigating means of providing the TSC with plant parameters independent of telephone conversations.
M P79 54 01/42 Salem 1 & 2
Onsite Operational Support Center (Section 2.2.2.c)
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 communication and management.
Response
The enclosed area between the Unit 1 and Unit 2 Control Rooms has been designated as the Onsite Operational Support Center.
This area is separate from each Control Room and is the place to which operations support personnel report in an emergency situation.
Communications with each Control Room are provided at this location.
M P79 54 01/43 Salem 1 & 2
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i ATTACHMENT 2 EMERGENCY PREPAREDNESS Salem 1 & 2
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EMERGENCY PREPAREDNESS NRC Position
- 1.
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.
Response
The NRC Staff on Page 13-3.of Supplement No *. 3 to the Safety Evaluation Report for Salem Nuclear Generating Station Unit 2 (NUREG-0492, December 29, 1978) concluded that " *** the emer-gency plans for the Salem Nuclear Generating Station are accep-table and that they meet or exceed the minimum requirements of 10CFR Part 50, Appendix E, conform to the provisions of Regula-tory Guide 1.101 (Revision 1) and that*they provide reasonable assurance that appropriate measures can and will be taken in the event of an emergency to protect public health and safety and prevent damage to property."
It is our conclusion that the Salem Emergency Plan conforms to the requirements of Regulatory Guide 1.101.
Furthermore, the emergency procedures at the Salem Station provide for action level criteria based on plant parameters such as containment radiation levels, containment pressure, and containment temperature.
EMOl/l Salem 1 and 2
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- .J EMERGENCY Pl~EPAREDNESS NRC Position
- 2.
Assure the implementation of the related recommendations of the Lessons Learned Task Force involving instrumen-tation to follow the course of an accident and relate the information provided by this instrumentation to the emergency plan action levels.
This will include instru-mentation for post accident sampling, high range radio-activity monitors, and improved in-plant radioiodine instrumentation.
The implementation of the Lessons Learned Task Force's recommendations on instrumentation for detection of inadequate core cooling will also be factored into the emergency plan action level criteria.
Response
Post-accident sampling, high range monitors and instrumenta-tion for detection of inadequate core cooling are addressed in our responses to the NRC Position Statem~nts 2.1.8.a, 2.1.8.b, 2.1.8.c, and 2.1.3.b of NUREG-0578.
A procedure will be developed to provide for means of esti-mating radiation releases from the facility prior to the
'I period when higher range radiation monitors are available and in service.
This procedure will become part of the station emergency procedures and will be completed in accordance with the Category A implementation schedule.
EMOl/2 Salem 1 and 2
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..J EMERGENCY PREPAREDNESS NRC Position
- 3.
Determine that an emergency operations center for Federal, State and local personnel has been established with suit-able communications to the plant, and that upgrading of the facility in accordance with the Lessons Learned Task Force's recommendation for an in-plant technical support center is underway.
Response
An Emergency Operations Center for the New Jersey Department of Environmental Protection, the lead state agency for coor-dinating responses to emergency situations, has been estab-lished at the Municipal Building in Lower Alloways Creek Town-ship, approximately five miles from the Station.
A communica-tions system has been established.
An.alternate location, in the event that this forward command post could not be utilized, would be at the Salem County Courthouse in the city of Salem, New Jersey, approximately eight miles from the Station.
The State of Delaware has established an Emergency Operations Center in Delaware City, Delaware, approximately eight miles from the station site.
The Delaware Division of Emergency Planning and Operations is headquartered in this facility and will direct any needed emergency action in the State of Dela-ware.
Direct communications with the station have been estab-lished at this facility.
An alternate headquarters has been proposed for Georgetown, Sussex Co., Delaware.
Development of an on-site technical support center is addressed in response to NRC Position Statement 2.2.2.b of NUREG-0578.
BM02/2 1 Salem 1 and 2
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- 1)
EMERGENCY PREPAREDNESS NRC Position
- 4.
Assure that improved licensee offsite monitoring capabilities (including additional thermoluminescent dosimeters or the equivalent) have been provided for all sites.
Response
PSE&G has had an extensi*-ve o.1:.E:si,.te. radiological monitoring program in effect for many years around the Salem Nuclear Generating Station site.
This program is defined as part of the Environmental Technical Specifications, Appendix B to the Salem No. 1 Operating License.
It is our understanding that this program, which includes the extensive use of TLD's, meets all current NRC Regulatory requirements for offsite monitoring programs.
PSE&G also works closely with the State of New Jersey, which conducts an independent radiological monitoring program in the vicinity of the Salem Station.
BM02/2 2 Salem 1 and 2 J
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EMERGENCY PREPAREDNESS NRC Position.
- 5.
Assess the relationship of State/local plans to the licensee's and Federal plans so as to assure the capa-bility to take appropriate emergency actions.
Assure that this capability will be extended to a distance of 10 miles.
This item will be performed in* conjunction with the Off ice of State Programs and the Office of Inspection and Enforcement.
Response
... ; ~ ~... :
It is our understanding that the emergency plans prepared by the States of New Jersey and Delaware meet all current cri-teria that have been issued by Federal agencies relating to emergency planning.
As such, we are in compliance with the Category Al implementation schedule for meeting current criteria.
As indicated in the response to item I.herein, the NRC staff has concluded that the Salem emergency plan meets the require-ments of Appendix E to 10 CFR 50 and Regulatory Guide 1.101.
It is our intent to work closely with the States of New Jersey and Delaware in upgrading the emergency plans to satisfy upgraded criteria requirements by January 1, 1981.
BM02/2 3 Salem 1 and 2
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EMERGENCY PREPAREDNESS NRC Position
- 6.
Require test exercises of approved emergendy plans {Federal, State,* local and licensees), review plans for such exer-cises, and participate in a limited number of joint exer-cises.
Tests of licensee plans will be required to be con-ducted as soon as practical for all facilities and before reactor startup for new licensees.
Exercises of State plijns will be performed in conjunction with the concurrence reviews of the Office of State 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 wo~ld result in all sites being exercised once each five years.
Revised planning guidance may result from the ongoing rulemaking.
Response
PSE&G has conducted annual exercises of the Salem emergency plan since 1976.
These exercises have involved testing the Station emergency procedures and the emergency plans developed by the States of. New Jersey and Delaware, which included exten-sive State and local involvement.
These exercises have been observed by representatives of various Federal agencies.
These exercises have continually demonstrated that a viable emergency.
plan exists and that we are in compliance with the Category Al implementation schedule for tests of licensee and State emer-gency plans.
Additionally, since the annual exercises involve extensive State and local participation and Federal observation, we believe that only a relatively minor expansion of the current exercise proce-dure would be required to more directly involve Federal agencies and satisfy the requirement for a joint exercise each five years.
BM02/2 4 Salem 1 and 2