ML18081A876
| ML18081A876 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/04/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001090380 | |
| Download: ML18081A876 (99) | |
Text
Public Service Electric and Gas Company 80 Park Place Newark. N.J. 07101 Phone 2011430-7000 January 4, 1980 Director of Nuclear Reactor Regulation
- u. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. D. B. Vassallo, Acting Director Division of Project Management Gentlemen:
REQUEST FOR ADDITIONAL INFORMATION RESULTING FROM THE THREE MILE ISLAND 2 ACCIDENT NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-3ll Public Service Electric & Gas hereby submits its revised responses to the requests for additional information contained in-your letter of September 27, l979 and clarified by your letter of November 9, 1979.
This information updates and supersedes the information on Lessons Learned transmitted in to our letter of October 12, 1979.
Should you have any questions, please do not hesitate to contact us.
Enclosure The Energy People
.Very0ul: yo JZ~t471A R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction 8001090 9~2001 (400M) 9-77
i I*
Emergency Power Supply Requirements for the Pressurizer Heaters, *power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs (Section 2.1.1)
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 pressu*rizer 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 suprly capability.
- 2.
Procedures and training shall be established to make the operator aware of when and how the required pres-surizr 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.
M P79 54 01/1 Salem 1 & 2 J.~N 1 1980
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 manual~y 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.
An analysis performed by Westinghouse indicates that 150 kW of pressurizer heaters is needed to assure maintenance of natural circulation.
These backup heater groups will be manually set up such that only 150 kW can be supplied from each vital bus.
Each redundant heater group has access to only one Class lE division power supply.
Motive and control power interfaces with the vital buses will be through safety grade circuit breakers.
Emergency Procedure EI 4.9, "Blackout" addresses the transfer of the heaters to the vital buses.
The diesel generators are capable of supplying the 150 kW of pressurizer heaters concurrent with the equipment loads required for a LOCA.
The diesel loads during the injection M P79 54 01/2 Salem 1 & 2 JAN i
1~80
phase (with the inclusion of the pressurizer heater load) would be slightly above t~e 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating but well below the 30 minute rating.
Under blackout conditions, the diesel generators have sufficient c~pacity to supply the required equipment loads including the pressurizer heaters and meet the continuous diesel rating.
The connection of the pressurizer heaters to a vital bus is through a normally open Class lE circuit breaker.
This.cir-cuit breaker is mechanically key interlocked with the heaters' normal, non-vital power feed circuit breaker.
A manually operated disconnect sw.itch must also be* closed to make the connection.
In addition~ the backup group heaters are set up to supply only 150 kW.
Once the connection pathway is estab-lished manually, the final connection of the pressurizer heaters to the vital bus (open/close the Class lE circuit breaker) can be accomplished in the control room.
The setup of the vital bus feed to the pressurizer heaters can be com-pleted in a time frame consistent with maintenance of natural circulation.
The manual action of opening the non-vital pressurizer heater supply circuit breaker prior to closing the vital power supply circuit breaker by*mechanical key interlocks is necessary to eliminate any possibility of feeding other non-vital loads from the vital power supply.
M P79 54 01/3 Salem 1 & 2 JAN 1 1980
It is not necessary that certain equipment loads be shed in order to connect the pressurizer heaters to the vital buses.
As a precaution, statements will be added to the operating procedures alerting the operator to remain within the appro-priate diesel ratings.
The diesel-generator ratings are posted on the control console with the diesel watt meters marked with the 30 minute rating.
Connection of the pressurizer heaters to the vital buses does not require the reset of an SI signal.
The pressurizer heaters will not be automatically tripped from the vital buses upon a safety injection actuation signal.
This requirement is not applicable to the Salem design.
An event where the pressurizer heaters are on the vital buses when a LOCA occurs is a highly improbable occurrence.
Normal operation of the heaters does not and will not require their supply from the vital buses.
They will only be needed when normal power is lost (blackout).
A LOCA would have to occur following a blackout during which time it became necessary to operate the heaters to maintain natural circulation.
M P79 54 01/4
-3a-Salem 1 & 2 JAN 1 1980
Power Supply for Pressurizer Relief and Block Valves and Pr~ssurizer 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.
- 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 off.site 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 ~hen offsite power is not available.
RESPONSE
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 r~quirements.
The design of the pressurizer relief and block valve arrangement for both Salem units was predicated on ensuring the ability to relieve.
This concept resulted in providing two parallel relief paths which are completely independent and redundant.
Such a design concept is nec~ssary to pro-vide protection for ATWS conditions and low temperature overpressure transients.
M P79 54 01/5
-3b-Salem 1 & 2 JAN 1 1980
Incorporation of complete independence between the relief valve and block valve would negate the system's ability to meet the single-failure criterion for the events identified above.
The existing design, however, does incorporate the use of diverse power supplies for the PORV's and their asso-ciated block valves.
The relief valves are supplied by Class lE, 125VDC systems while the block valves use 230V and llSV vital AC.
There is no requirement for provisions to switch from normal power to emergency onsite* power for these devices since the normal power supply in. all cases is part of the onsite vital power system.
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 01/6
-3c-Salem 1 & 2 JIUi 1 1980
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 co~ditions 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
By letter dated December 17, 1979, the EPRI Safety and Analysis Task Force submitted its "Program Plan for the Performance Verification of PWR Safety/Relief Valves and Systems," dated December 13, 1979, to the NRC for review.
PSE&G considers the program to be responsive to the NRC's position.
The EPRI program plan provides for completion of the essential portions of the test program by July, 1981.
PSE&G will be participating in the EPRI program to the extent of providing program review and plant specific data as required.
M P79 54 01/7 Salem 1 & 2 1 1980
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
Each PORV is presently equipped with a lirnit switch to pro-vide an alarm in the Control Room if the PORV is not fully closed.
These switches will be replaced by seismically and environmentally qualified switches as soon as possible.
To provide positive indication of safety valve position, a limit switch will be mounted in each safety valve bonnet which will actuate a Control Room alarm if the valve is not fully closed.
This modification will be completed in ac-cordance with the Category A implementation schedule.
Although the switches being installed on the safety valves are qualified for both seismic and environmental condi tioris, an improved switch is expected to be available by April, 1980.
The improved. switch will be capable of indicating open, closed and an intermediate position.
Both of the above schemes utilize a single switch on each valve.
As discussed in the response to NRC Bulletin 79-06A, several reliable backup methods are available to detect an M P79 54 01/8
-s-Salem 1 & 2 JAN 1 1980
open valve which are addressed in Emergency Procedure EI 4.24, "Malfunction of Pressurizer Relief Valve."
- 1. Pressurizer pressure
- 2. Valve discharge piping temperature
- 3. PRT level, pressure and temperature
- 5. Control Room alarms for a*11 of the above indicators.
M P79 54 01/9
-Sa-Salem 1 & 2 JAN 1 1980
.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 consider.ation 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 01/10 Salem 1 & 2 JAN 1 1980
- c.
Steam 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 which are continuously displayed are reactor coolant differential pressure {P act-ual -
P saturated) and differential temperature (T saturated
- Tactual).
Alarms are set for pressure differential less than 200 psi and.temperature differential less than soop.
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.
Additional information is provided in Table 2.1.3.b-l.
Emergency Procedures, EI 4.4, "LOCA", and EI 4.6, "loss of Secondary Coolant," have been revised to address the use-of this computer program to monitor the margin of subcooling in the Reactor Coolant System.
PSE&G is a member of the Westinghouse Operating Plant Owners' Group.
Westinghou_se,.*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 emer~
gency procedures.
M P79 54 01/11 Salem 1 & 2 JAN
A preliminary report on inadequate core cooling was sub-mitted to the NRC on October 30, 1979 by the Owners' Group.
A more comprehensive report is scheduled for March 1, 1980.
Operating instructions from the preliminary report will not be incorporated into the existing station procedures.
The station procedures will be updated after completion of the final Owners' Group report will be assessed with respect to any recommende.d modifications, including reactor vessel water level indication.
M P79 54 01/12 Salem 1 & 2 J.4N
TABLE 2.1.3.b-l SUBCOOLING MEI'ER INFORMATION Display Information Displayed (T-Tsat, Tsat, Press, etc.)
Display Type (Analog, Digital, CRI')
Continuous or on Deman::!.
Single or Redundant Display I..oc::ation of Display Alanns (include setr:cints)
Overall uncertainty ( ° F, PSI)
Range of Display Qualifications (seismic, environmental, IEEE323) calculator Type (process canputer, dedicaterl digital or analcg' ca.le * )
If process canputer is userl specify avail-
. ability. ( % of time)
Single or rerlundant calculators Selection Logic (highest T., lo,.;est press)
Qualifications (seismic, environmental, IEEE323) calculational Technique (Stearn Tables, Functional Fit,. ranges)
Input Temperature (RI'D's or T/C's)
Temperature (number of sensors and locations)
Range of temperature sensors M P79 54 01/13
-8a-Note 6 None Process Computer 90%-95% (Estinated)
Single Note 7 None Steam Tables 32<°F<705 123'sia<3204 chranel/Alurrel T/C 65 Incore T/C's 30-2200°F Salem 1 & 2 JAN l_. 1980
TABLE 2.1. 3.b-1
( CONI'INUED}
Uncertainty* of temperature sensors ( °F at 1)
.+/-. 5°F Qualifications (seismic, environmental, IEEE323}
None Pressure (specify instrument used}
Barton 763 Pressure (number of sensors arrl locations}
2-#11 Hot Leg Range of Pressure sensors 0-3000 psig Uncertainty* of pressure sensors (PSI at l}
+150~ -300 psi (I..OCA Conditions)
Qualifications (seismic, environmental, IEEE323)
Qualified Backup Capability Availability of Temp & Press Availability of Steam Tables etc.
Trafuin:J of.operators
. Procedures Main Console Indica-tion Conversion Curves canpleted Canpleted
- Uncertainties are* not affecte:J. by differences in RCS flo.v conditions.
Thernocouples are located in h::>ttest regions and pressure rceasurement is.
irrleperrlent of flo,.,r conditions.
Notes
- 1.
Infonnaticn displayed:
(Tsat-Tact), (P-Psat).
Infonna.tion available en denand:
(Tsat-Tact), P-Psat), Psat, Pressure, Tenperature arrl locaticn of hottest in-core T/c.
- 2.
'!he c::ontim.ous infornation display is either an analo; recorder or a single CRI'.
The infornaticn available on derrarrl can be displaye:J on the CRT or trend typewriter.
M P79 54 01/14
-Sb-Salem 1 & 2 1 1980
Notes
,TABIE 2.1.3.b-l (CONTINUED)
- 3.
'Ihe CRI' :i,s iocated en the center-left p::>rtion of the main control console. 'lhe other displays are available at the operator's can-puter console.
- 4. Alanns:
(Tsat-Tact) - less than 50°F subcooling (P-Psat) - less 'tllan 200 psi Temp. - any T/C greater than 630°F If *these ala.nt'S occur, they will* re displayed even if the subcooling
- calculation prbgram has rot l:::een requirested by the q::erator.
- s. overall uncertainty is a factor of the uncertainty in the tempera-ture and i;ressure *rreasurements and the resulting p:::>tential error
'When usin; the steam tables. The uncertainties of.these devices are listed separately en the table.
- 6.
Anal03 Ranges:
(Tsat-Tact) 0-120°F Diff; (P-Psat) 0-1000 psi Diff CRI' Fange:
N/A
- 7.
The selecticn 103ic used is:
- highest in-core T/ C reading
- average of t\\'.O reactor coolant pressures.
The q::erator is provided with the capability to reject the selected T/C; the next, highest reading T/C will then be autaratically se-lected for the calculation.
A reasonability check of the t\\'.O pressure readings is perfonned by the o:rnputer. If the readings indicate that one of the rreasurernents is invalid, the canputer will reject the invalid reading.
M P79 54 01/15
-8c-Salem 1 & 2 J.'1i~
J. i:WU
Containment Isolation Provisions for PWRs and BWRs (Section 2*-~:-1
- 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 in~tiation of containment isolation.
- 2.
All-plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each syst.em determined* to be essential, shall identify each system determined to be non-essential, snall 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 del.iberate 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 M P79 54 01/16 Salem l & 2 J.~N 1 1980
- b.
High Containment Pressure
- c.
Low Pressurizer Pressure
- d.
High Differential Pressure Between Steam Lines
- e.
High Steam Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
Containment Isolation -
Phase B
- a.
Manual Actuation
- b.
High-High Containment Pressure Containment Ventilation Isolation Manual Actuation
- a.
b *
- c.
High Containment Pressure Low Pressurizer Pressure
- d.
High Differential Pressure Between Steam Lines
- e.
High Stearn Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
- f.
High Containment Radiation - Particulate
- g.
High Containment Radiation -
- h.
High Containment Radiation -
Gaseous Main Steam Line Isolation
- a.
Manual Actuation
- b.
High-High Containment Pressure
- c.
High Steam Line Flow Coincident with Low Steam Line pressure or Low-Low Tavg
- M P79 54 01/17 Salem 1 & 2
.JAN 1 1980
Feedwater Isolation
- a.
Manual Actuation
- b.
High Containment Pressure
- c.
Low Pressurizer Pressure
- d.
High Differential Pressure Between Steam Lines
- e.
High Stearn Line Flow Coincident with Low Steam Line Pressure or Low-Low Tavg.
- f.
High-High Stearn Generator Water Level
- g.
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 S.4-1 through 5.4-27 of the FSAR.
All lines penetrating the containment are shown*in these figures.along with their isolation provision~. 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 01/18 Salem 1 & 2 JAN 1 1980
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 01/19 Salem 1 & 2 JAN 1 1980
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 Bulletins 79~06A.
Implementation of the design changes will be completed in accordance with the
. Category A.. implementation schedule.
M P79 54 01/20 Salem 1 & 2 JAN 1 1980
Dedicated Penetrations for External Recombiners or Post-Ac-cident Purge Systems (Section 2.1.s.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 01/21 Salem l & 2 JAN 1 1980
Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant (Section 2.1.5.c)
NRC Pos*i tion
- 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.
The procedure for use of the hydrogen recombiners, OI II -
15.3.1, Hydrogen Recombiner -
Normal Operation", has been reviewed and revised as required in response to IE Bulletin 79-06A.
M P79 54 01/22 Salem 1 & 2 JAN 1 1980
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 pro-gram to reduce leakage from systems outside containment that would or could contain highly radioactive fluids dur-ing a serious transient or accident to as-low-as-practical levels.
This program shall include the following:
- l.
Immediate Leak Reduction 2 *
- 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 opera-tion and report them to the NRC.
Continuing Leak Reduction Establish and implement a program of preventive main-tenance 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
A review has been conducted on those systems outside of the Containment Building which may contain high post-accident radioactive fluid inventories to determine whether any modifications are necessary to ensure prevention of unplanned release of radioactivit~. This review has determined that the present design meets the intent of the NRC position because:
P79 130 12 Salem l & 2 JAN 1 1980
- 1.
All valves, 2" and larger, meeting the following re-quirements, are hard piped to the liquid radwaste system:
- a.
Operating fluid temperature above 212or,
- b.
Normally radioactive service.
- 2.
Those valves not hard piped to the liquid radwaste system which develop leakage and any other leakage from a system is identified during normal plant tours by operating personnel., Operatoring person-nel report any abnormalities such as system leakage to station management.
- 3.
Periodic testing to meet In-Service Inspection re-quirements provides an indication of system integrity.
Implementation of the provisions of ASME XI-1974 requires service pressure vessel leak tests of Nuclear Class I Systems every refueling outage, Nuclear Class II Systems every 10 years (these are performed every 3-1/3 years, however, as ari extension of commitments in response to IE Bulletin 76-06) and Nuclear Class III Systems every 3-1/3 years.
P79 130 13 Salem 1 & 2 JAN 1 1980
- 4.
The Ventilation System in the Auxiliary Building is designed such that gases emitted from system leakage will be carried from areas of lesser contamination to areas of higher contamination as described in FSAR Sections 9.10.l.2 and 9.10.3.
- 5.
All floor and equipment drains are piped to the liquid radwaste system.
To provide even greater assurance that those systems which contain radioactive fluid.are leak-tight, PSE&G is in the process of conducting a review per NRC IE Circular 79-21.
This Circular requests that as-built systems be reviewed for the potential of inadvertent release of radioactivity.
The gaseous radwaste system will be tested by performing a soap bubble test on 30% of all weld and mechanical joints every 3-1/3 years.
Quantitative leak rate testing on these systems is not
- practical.with the hard-pipe leakoff design~
No reasonable means of performing this type of test is available and therefore no quantitative leakage tests are expected to be performed.
P79 130 14
-l7a-Salem l & 2 "JAN j
19~0
The leakage reducti~n program includes periodic review of open work orders involving system and/or containment integ-rity by the Shift Technical Advisors an6 valve lineup veri-fication by both the STA's and the station QA Department
- P79 130 15
-l7b-Salem l & 2 J.~N 1 1980
NRC Position
- With-the assumption of a post-accident release of radio-activity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a radiation and shielding design re~iew of the spaces around systems that may, as a result of an accident, contain highly radio-active 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 per-sonnel occupancy may be*unduly limited or safety equipm~nt may be unduly degraded by the radiation fields during post-accident operations of these sy~tems.
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-acci-dent procedural controls.
The design review shall de-termine which types of corrective actions are needed for vital areas throughout the facility.
RESPONSE
The original design of the Salem Station included con-sideration of access requirements in many areas of the Auxiliary Building and Penetration Areas under post-accident conditions.
Source terms were based on.the guidance given in TID-14844.
A conservative approach was taken in developing the plant arrangement and shielding design from a normal operation occupancy standpoint.
Thus, even in those areas where specific consideration was not given to post-accident ~ccess, dose r~tes in those areas would be lower than otherwise expected bacause of additional shielding, separation of components and layout arrangement.
P79 129 32 Salem 1 & 2 J.~N _.1. 1o~n
The design review undertaken as required by NRC Position 2.1.6.b was based on a post-accident liquid sourc~ term of a release of 100% of the core noble gases, 50% of the halogens and 10% of the cesuims, strontiums and bariums.
The re-sult-ant source term is slightly more conservative than that pres~nted in Regulatory Guide 1.4 For airborne containment sources, 100% of the core noble gases and 25% of the halogens were assumed mixed in the containment atmosphere.
Our analyses are based on a radioactive decay time of one day after reactor trip.
Calculations are focused on areas in the Auxiliary Build-ing and Penetration Areas.
Dose rates in the containment have been calculated only where doses to certain equipment and instruments are of concern.
The systems and areas reviewed include:
RHR System Safety Injection CVCS Demineralizer Area Charging Pump Compartments Reactor Coolant Filter
- Seal Water Fiiter Area Chemistry Lab Primary Sample Lab Fuel Handling Building Spent Fuel Pool.Heat Exchanger Area Liquid Radwaste (review still in progress)
P79 129 33
-isa-Salem l
& 2
Accessibility to systems and areas:
Residual Heat Removal System - Elev. 45' and 55' Aux. Bldg.
l)
The RHR pump compartments on elevation 45' in the Auxiliary Building would have a general area dose rate with the compartments of approximately 30,000 R/hr.
- 2)
The dose rate in the adjacent RHR compartment will be approximately 30 mr/hr.
This compartment is access-ible while the other RHR system is operating~
- 3)
The dose rates on elevation 55' from the operating RHR system below are approximately 8 R/hr.
This dose rate will drop off by a factor of 2 after one week decay.
Lead sheet placed on the floor will further reduce the dose rate such that limited access is available to this area.
Permanent shielding in this area, is required on an exposed portion of 14" RHR suction pipe.
Six inches of lead will be installed to shield this pipe.
- 4)
Access to either of the RHR pump compartments can be accomplished by draining and flushing each respective system.
P79 129 34
-18b-Salem 1 & 2
Safety Injection System
- 1)
The Safety injection pump compartment is inaccess-ible while operating.
- 2)
Dose rates in adjacent areas, such as the Spent Fuel Pool Heat Exchanger area and Component Cooling.
Heat Exchanger compartments are approximately 60b mr/hr at contact with the pipe chase and pump compartment shield walls.
This dose rate drops off substantially several feet from the walls.
There is limited access to these areas and no additional permanent shielding is planned.
Charging Pump Compartments
- 1.
Dose rates in the vicinity of these pumps are estimated to be 5000 R/hr, thus precluding access while the pumps are operating.
_ 2)
The dose rate through the wall separating the pump compartments is approximately 5 R/hr.
- 3)
The dose rate outside the Charging Pump compartments is approximately 200 mr/hr; therefore, access to the components in the general area is available
- P79 129 35
-18c-Salem 1 & 2
. I fl ~1 1 1no"
- 4)
Charging Pum~ valve compartment dose rates will be unacceptably high due to short lengths of exposed pipe and valves.
Permanent lead shielding will be installed to reduce dose rates from the valves to levels below 200 mr/hr at the outer pump compart-ment walls.
This will provide accessibility.
Chemical and Volume Control - Demineralizer Area
- 1)
Dose rates from the demineralizers would not have a significant effect on access.
Resins are changed upon either high radiation level or high pressure drop.
- 2)
The dose rates from piping and valves located behind valve aisle shield walls would be the major source of radiation and result in levels of approximately l R/hr in the operating aisles.
This would be reduced by decay and will afford limited access to the area for valve operations.
\\
P79 129 36
-18d-Salem 1 *& 2 JAN 1 i9BO
Reactor Coolant and Seal Water Filters
- 1)
The dose rates from these filters do not present a problem, since the elements are replaced at pre-determined radiation levels rather than high.pressure drop.
Post-accident radiation levels in this area will not preclude access for filter changing.
Each filter is located in an individual shielded compartment.
Primary Sample Lab
- 1)
Dose rates from Primary Sample system tubing that would be used to draw a Reactor Coolant sample are calculated to be approximately 50 R/hr at contact and 2 R/hr irt the general area of the lab.
These dose rates would be higher at T = o.
Permanent shielding will be installed on those sample lines to be used for post-accident samples.
Counting Room
- 1)
Direct dose rates in the counting room are not significantly affected by accident radiation source terms due to the location of the counting room.
P79 129 37
-18e-Salem l & 2 JAN 1 1980
Fuel Handling Building l)
Dose rates in the Fuel Handling Building due to direct radiation from the Containment will not be significantly affected.
The only exception to this is streaming from the elevation 130' Containment Personnel Hatch and through the doorway into the Fuel Handling Building at elevation 130'.
This would be minimized by placing temporary block shielding in front of this doorway.
- 2)
The dose rates at the Spent Fuel Pool Heat Exchanger and Pump area in the Auxiliary Building are estimated to be approximately 600 mr/hr, thus affording limited access to this area.
Liquid Radwaste
- 1)
At present, the liquid radwaste system is being reviewed.
- 2)
Design parameters such as mode of operation, source terms, and access requir~ments are being developed.
- 3)
Included in the design review will be a consideration of the airborne contamination resulting from overflow-ing a Waste Hold-up Tank.
The Auxiliary Building Ventilation system is designed to preclude airborne contamination from spreading from one portion of the building to another.
P79 129 38
-18f-Salem l & 2
.1n tJ 1 1QHO
Other Considerations:
Local and Pottable Shielding Temporary shielding such as lead bricks, lead blankets and lead sheet is available at the station for use where small quantities of shielding may be required to shield local hot spots.
It is intended, however, to install permanent shielding where possible to reduce the amount of temporary shielding required.
Source Terms The source terms used for this study were based on one-day decay.
The calculated dose rates would be reduced by a factor of approximately 25 at 30 days after the start of an accident.
For systems such as primary sampling, where access is required at one hour after an accident, dose rates would be a factor of 10 higher than the one day delay results.
In this instance, more shielding may be added as necessary.
P79 129 32/39
-18g-Salem l & 2
.IAN 1 198C
Automatic Initiation of the Auxiliar 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:
- l.
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.
- s.
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 feedwat.er 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 syste~ is designed to Class IE criteria and is powered from the emergency power source.
M P79 54 01/26 Salem 1 & 2 JAN 1 1980
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 Stearn 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 design~d for single failure considerations and are testable.
M P79 54 01/27 Salem 1 & 2 JAN 1 1980
Auxiliary Feedwater Flow Indication to Steam Generators for PWRs (Section t.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 requi~ements 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.
One flow instrument for each.
steam generator is provided.
In addition, three level instruments are provided for each steam generator.
The instruments are all powered from the vital buses, seismically qualified with environmental qualification for the level instruments which are located inside the containment.
M P79 54 01/28 Salem 1 & 2 JAN 1 1980
Assurance of sufficient water being provided to the steam generators is of primary concern.
This is accomplished by control of valve demand with steam generator level indica-tion.
Present indication of pump operation, valve demand/
position, auxiliary feedwater flow (one/steam generator),
auxiliary f eedwater discharge pressure and steam generator level (three/steam generator) is adequate to meet the information requirements necessary to assure appropriate operator action.
All of the above equipment is powered from vital buses, and is considered adquate to meet short.and long term requirements.
M P79 54 01/29
-2la-Salem 1 & 2 1.. *......
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 provid~d to meet the
.criteria.
A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radioisotope~ 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 m~lting). The initial reactor doolant 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 sh~ll 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.
P79 131 01/10 Salem l & 2
Response
A design and operational review of the containment atmosphere and reactor coolant sampling systems has been performed to determine the capability of personnel to pro~~tly and safely obtain a sample under post-accident conditions within the time and exposure constraints identified above.
A piping Arrangement Drawing, an Instrument Schematic and Controls Logic Diagrams are attached.
(Drawing Nos. SK-12879, 207510-8~9491, 248251-B-9803 through 248254-B-9803).
The containment components of the Radiation Monitoring System are utilized to acquire a containment air grab sample.
This portion of the system was designed for normal operating conditions.
Therefore, under the relatively higher pressures and temperatures which could be expected during an accident, the existing air sampling pump would fail and a representative sample of containment atmosphere would thus be unobtainable.
Acquisition of a post-LOCA sample is further precluded by the fact that the grab sample location point is in the electrical penetration area, an area which becomes inacce~sible during an accident due to radiation streaming through the surrounding penetrations.
In addition, the Radiation Monitoring System containment isolation valves close upon a containment isolation signal.
P79 131 02 Salem 1 & 2 1." :'";,.,
j_ i :.:; ~ ~:.
The Reactor Coolant Sampling System utilizes Reactor Coolant System pressure to acquire a sample.
Should an unpressurized condition exist after an accident, the head available is insufficient to draw a sample to the primary sampling laboratory.
Also, the post-accident containment flood level is such that the sample system components inside the containment isolation valves would be under water.
As with containment air sampling, the Reactor Coolant Sampling System containment isolation valves close upon a containment isolation signal.
Proposed Modifications As a result of this review, modifications will be made to enable containment atmosphere and reactor coolant sampling to be performed in an expeditious (within one hour) and safe (within allowable dose criteria to personnel) manner in the event of an accident.
New design features will be added for use during accident conditions only while continuing to use the existing features for normal operations.
The new features will be designed for a containment environment of 50 psig and 350°F and reactor coolant conditions of 2485 psig and 650°F.
P79 131 03
-23a-Salem 1 & 2
"* *,.. ; \\
These safety related additions will be designed to Nuclear Class II, Seismic Category I criteria.
The post-LOCA containment atmosphere sampling system will consist of two independent, electrically separated loops for each unit while the post-LOCA Reactor Coolant Sampling System will have two electrically separated lines and equipment for each unit up to the primary sampling laboratory at which time they will be tied into the existing sample lines and equipment.
The design will maintain physical and electrical separation as much as possible throughout the systems.
In the post-LOCA containment air sampling system design, each unit will have redundant air supply and return lines.
Th~ existing inside containment supply and return lines from the Radiation Monitoring System will be utilized by teeing upstream of the inside containment isolation valves (IVC7,9,ll,13 and 2VC7,9,ll,13).
The samples will thus be drawn from and returned to elevation 145' inside containment.
Upon teeing into the two supply and two retu~n lines, each new pair of supply and return lines will be run through separate electrical penetrations.
In the post-LOCA Reactor Coolant Sampling ~ystem design, new sampling lines will tee into the existing lines off the #11 and #13, (#21, and #23) hot legs, upstream of P79 131 04
-23b-Salem 1 & 2
valves 11 and 13SS32 (21 and 23SS32).
The new lines will be run through separate electrical penetrations.
Each of the new sampling lines inside containment will be provided with a 150-foot delay coil which will allow for decay of. some of the short-lived isotopes prior to the sample reaching the primary sampling laboratory.
Ea~h of the six new lines (four containment air and two reactor coolant) will have normally closed/fail tlosed, air-operated isolation valves inside and outside containment.
To meet channel separation criteria, the the isolation valves will have different vital power supplies to their solenoid coils.
The outside containment isolation valves will be supplied with redundant control air lines.
Backup pressurized air accumulators will be provided for the inside containment isolation valves.
Each accumulator wlll be sized for approximately 1000 cycles of operatidn (mor~ than a one-month supply, based on the conservative assumption that a sample will be taken once an hbur dµring the initial m~nth following an accident).
All air accumulators and isolation valves inside containment will be mounted on platforms above the. containment flood level.
For post-LOCA containment air sampling, there will b~ one sampling location in each unit, each with the ability to P79 131 05
-23c-Salem 1 & 2 JAN 1 1980
draw a sample from either unit.
The Unit 1 location will be in the Auxiliary Building on elevation 84' in the spent fuel pool heat exchanger compartment while the Unit 2 location will be in the Auxiliary Building on elevation 100'.
The Unit 1 lo~ation will have two radioactive gas processing pumps for drawing Unit 1 samples, two sample
.stations for acquiring Unit l samples, two sample stations for acquiring Unit 2 *amples, a panel for Unit 1, and a panel for Unit 2.
The Unit 2 location will have the same items with the exception that the two pumps will be used for drawing Unit 2 sa~p1Qs~ The 0.5 cfm pumps will be of
(
l the dual containment ':::...____:..less steel bellows type.
The dual containment feature contains an inter-barrier leak test port which will provide for an early indication of degradation.
Chilled water will be provided to the pumps to cool the motors.
The pumps will have the same vital.
power supply as the containment isolation valves in ea~h respective loop~
Check valves in the pump discharge line will prevent ~ontainment air from reaching the pump through the return line to the containment.
Each of the sample stations will be provided with permanent in-line, stainless steel sample vessels.
The panels will be provided with valve position indication along with valve and pump controls.
.In addition, the panels will have P79 131 06
-23d-Salem 1 & 2 111 ll
-I 1(10"
phone jacks for communication with the Control Room.
Hoods will be provided at the sample stations to capture any gases released when obaining the sample and exhaust them to the plant ventilation system.
Area radiation monitors will be installed in both of the sampling locations with Control Room indication.
In acquiring.a pos-t-LOCA. containment air sample, the sampling location in the affected unit will be utilized.
To prevent flow to the sampling location in the other unit normally closed solenoid valves will be provided in the supply and return lines.
If the primary sampling location is unavailable, these normally closed valves will be opened to allow the sample to be taken in the other unit *
. After obt~ining a containment air sample, a 100 psi nitrogen purge will be used to purge any airborne particulate in the tubing back into containment.
A check valve in th~ nitrogert line will prevent contamination of the nitrtigen supply.
For reactor coolant sampling, the sampling locatiorts for both units will be in the primary sampling laboratory located in the Auxiliary Building of Unit 1 on elevation 110'.
The existing equipment in the primary sampling laboratory will be used to process and analyze the sample, the equipment having been designed to handle fluids to P79 131 07
-23e-Salem l & 2 JAN
- l. 1~80
2485 psi and 650°F.
At the lab, new sample lines will tee into the existing reactor coolant sample lines.
Sample pressure and temperature will be reduced to within
.reasonable limits prior to reaching the sample vessel or sample sink.
The existing hoods at the sample sink are connected to the building exhaust system *. New panels, one for each unit, will be installed in the lab.
The new panels will provide valve position indications along with valve and pump controls.
In addition, the panels will have phone jacks so that Control Room communication can be maintained while personnel are acquiring samples.
The area radiation monitors in the lab provide Control Room indication and alarms.
Each unit will be supplied with tw~ reactor coolant sampling pumps 16cated in the boric acid evaporator compartment for each unit.
The 0.25 gpm pumps will be provided with the same vital power supply as their corresponding Containment isolation valves.
These pumps will be used when the reactor coolant pumps are not operating and the system pressure is insufficient to provide a sample to the lab~
A senso~ will be installed on the s~ction side of the pump.
When the containment isolation valv~s are opened a timer will start to ensure enough time expires to allow the sample to reach the P79 131 08
-23f-Salem 1 & 2
_ I I\\ ~I
pumps.
If pressure is sensed, the pumps will be.
bypassed.
After the sample is taken, the pump will be shut off and containment isolation reestablished.
The lines will be flushed with primary water to the drain header.
To minimize radiation effects, tubing will be routed th~ough normally or potentially radioactive compartments wherever p6ssible.
Additionally, high energy break analyses will be performed.
Tubing in the containment, electrical and mechanical penetration areas, and the pipe alley will not be shielded since these areas are already inaccessible in a post-LOCA condition.
A small portion of the tubing will be routed through the switch gear rooms.
Radiation shielding and tube rupture protection will be provided in the switchgear rooms.
- Molded lead shielding will be provided to house the tubing runs* through the Spent Fuel Pool Heat Exchanger Compartments, Safety Injection Comp~rtments, Unit 2 Component Cooling Compartment, Unit 2 Sampling location roo~ oi elevation 100', boric acid evaporator
_compartments, and the primary sampling laboratory.
All P79 131 09
-23g-Salem l & 2
tubing which must c~oss from one unit to the other will also be provided with molded lead shielding and will pass through a penetration on elevation 114' of the Auxiliary Building.
Supports for all the tubing runs will be designed to Seismic Category I criteria.
Sample vessels will be w~apped in lead.
Also, _to reduce contact exposure to extremities, all manual valves will be provided with long extension stems.
A study of the primary sampling laboratory has been performed to further assess additional shielding provisions which may be required to reduce background levels of radiation and exposure to personnel to as low as reasonably achievable.
P79 131 01/10
-23h-Salem l & 2 JAN 1 1980
I I
i~-
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 of 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 radio iodines. 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 01/32 Salem 1 & 2
Response
- 1.
The plant vent gaseous monitors have the following
~etection range capabilitie~:
Unit 1:
Sxlo-6 to Sx10-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 01/33 Salem 1 & 2
_J.4N 1 10on.
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 01/34 Salem 1 & 2
- " l~ -
Improved In-Piant 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.
The capability exists to accurately detect the presence of iodine in areas of the plant that may be occupied during an accident.
This capability utilizes portable air samplers with charcoal collection cartridges and a 365 kev peak for I-131.
These units are part o~ the emergency equipment specified in the Emergency Plan.
Operating procedures are available and training is conducted in accordance with the Emergency Plan.
The capability to purge a charcoal cartridge with clean air or nitrogen and.to remove the cartridge to a low background area for further analysis will be established by January 1, 1981.
M P79 54 01/35 Salem 1 & 2 JAN 1 1980
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 co~e cooling is approached as well as the period of time during which inadequate core cooling exists.
The calc.ulations 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 01/36 Salem 1 & 2
.J~ N. 1 1QP.()
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 01/37
-29 Salem 1 & 2
.tO.N 1 1980
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 comp*arisons 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.
M P79 54 01/38 Salem 1 & 2 JAN 1 1980
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 EI 4.4, "LOCA", and EI 4.17, "Leakage Greater than charging Flow," (in effect, inadequate core cooling) will incorporate the results of the analysis work performed.
Operating personnel have been advised of small break LOCA procedural changes and will receive the appropriate training for inadequate core cooling emergency procedures upon completion of the generic analysis work in early 1980.
Analyses of small break loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core coolingi and analysis of transient a~d accident scenarios including operator actions not previously analyzed are being performed on a generic basis by the Westinghouse Owner's Group, of which PSE&G is a member.
The small break analyses have been completed and were reported in WCAP-9600, which was* submitted to the Bulletins and Orders Task Force by the Owners' *Group on June 29, 1979.
Incorporated in that report were guidelines that were developed as a result of small break analyses.
These guidelines have been reviewed
- nd approved by the B&O Task Force and have been presented to the owners' Group utility representatives in a seminar held on October 16-19, 1979.
Following this seminar, each M P79 54 01/39 Salem 1 & 2 JAN 11980
utility has developed plant specific procedures and trained their.personnel on the new procedures.
Revised procedures and training are in place in accordance with the requirement in Enclosure 6 to Mr. Eisenhut's letter of September 13, 1979, and Enclosure 2 to Mr. Denton's le~ter of October 30, 1979.
The work required to address the other two areas--.inadequate core cooling and other transient and accident scenarios--has been performed in conjunction with schedules and require-ments established by the Bulletins and Orders Task Force.
Analysis related to the definition of inadequate core cooling and guidelines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation and for restoring core cooling following a small break LOCA were submitted on October 31, 1979.
This analysis is a less detailed analysis than was originally proposed, and will be followed up with a more extensive and detailed analysis which will be available during the first quarter of 1980.
The guidelines and training will be in place by December 31, 1979, as required by the B&O Task Force.
With respect to other transient accidents contained in the Salem FSAR, the Westinghouse Owners' Group has performed an evaluation of the actions which occur during an event by constructing sequence of event trees for each of the non-LOCA and LOCA transients.
From these event trees a list MP79 54 01/40 31a Salem 1 & 2 JAN 1 1960
of decision points for operator action has been prepared, along with a list of information available to the operator at each decision point.
Following this, criteria have been set for credible misoperation, and time available for operator decisions have been qualitatively assessed.
The information developed was then used to test Abnormal and Emergency Operating Procedures against the event sequences and determine if inadequacies exist in the AOPs and EOPs.
The results of this study will be provided to the Bulletins and Orders Task Force by March 31, 1980.
The Owners' Group has also provided test predictions of the LOFT L3-l nuclear small break experiment.
This analysis was provided on December 15, 1979, in accordance with the schedule established mutually with the Bulletins and Orders Task Force.
M P79 54 01/41 3lb Salem 1 & 2 JAN 1 1980
Instrumentation to Monito'r 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 sh~ll 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
Response
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 will be completed by January 1, 1981.
I A~ -.
C:"' 1 om 1
$:._. ?
Installation of Remotely Operated High Point Vents in the.
Reactor Coolant System Each applicant and litensee shall install reactor coolant system and reactor v~ssei head high poiht vents remotely operated from the control room.
Since these vents form a part of the reactor coolant pressure boundary, the desig~
of the vents shall conform to the requirements of Appendix A to 10 CFR ~art SO General Design Criteria.
In particular, these vents sh~ll 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 actua-tion.
Each applicant and licensee shall provide the following in-formation 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 ~ith 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 accept-ance criteria of 10 CFR 50.46.
- 2.
Analyses demonstrating that the direct venting of no.n-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.S.
- 3.
Procedural guidelines for the operators' use of the vents.
The information available to the ~perator for ir:ii.tiating or terminating vent usage shall be discussed.
Response
Reactor vessel remote venting capability ha~ been engineered and designed in accordance with NRC requirements.
The design features and safety considerations of the modif ica-tions involved are described below.
P79 130 17
-32a-Salem 1&2 I/HI
DESIGN FEATURES The basic design consists of extending the existing 3/4" Reactor Vessel head vent piping to the Pressurizer Relief Tank (PRT) via a redund~nt ~rouping af solenoid ~perated vent valves.
Figures HPV-1 and HPV-2 show the schematic and the physical arrangement of the modifications for Salem 1 (Salem 2 is of similar design).
As shown on Figure HPV-1, the design has the potential of either venting to the containment or to the PRT.
This remote manual vent ca~ be actuated f ~om the Control Room utilizing a key lock switch and would have power removed during normal operation.
The head vent is engineered and designed for safety grade requirements to satisfy the single failure criterion of IEEE-279.
The solenoid operated vent valves are powered from two redundant vital DC buses.
Open/close indications for the solenoid valv~s will be provided in the Control Room, both visual and audible.
P79 130 18
-32b-Salem 1&2 I II
.I
The head vent is sized to be capable of venting the gases of half the RCS volume in less than l hour.
The entire head vent piping ia 3/4" Sch. 160 S.S. Type 316 Seismic Category I piping.
Break flanges (male and female) are provided in the new vent piping to allow reactor vessel head removal.
The piping is supported to prevent pipe whip and jet impingment forces from degrading *other safety related functions.
The allowable loading on the existing vent nozzle on the reactor vessel will not be exceeded.
Operating procedures for the use of this remote head vent capability will be developed.
SAFETY CONSIDERATIONS (l)
The reactor head vent is connected to the reactor vessel through a 3/4" nozzle and piping.
A break in the vent line is considered as an infrequent fault and is covered in FSAR Section 14.2.l as a loss of reactor coolant accident resulting from a small bore pipe P79 130 19
-32c-Salem 1&2
._ f It l I c....,.,...
I break.
The analysis presented in the FSAR shows that the high head portion of the Emergency Core Cooling
- System together wi*th the accumulators provide sufficient core flooding to keep the peak clad temperature below the limits set forth in 10CFRS0.46.
The analysis presented in the FSAR for the small bore pipe break accident is based on a break size of 3" to 6" diameter.
Westinghouse report WCAP-9600, *small Break Accidents for Westinghouse NSSS Systems", June, 1979 provides analyses for break sizes of less than 3" as well as up to 6~. This report revalidates the FSAR analysis for the small bore pipe break accident discussed above.*
The 3/4" reactor head vent piping falls into the category of the break si~e 3/8" < diamet~r < l" a$
analyzed by WCAP-9600.
For this size break~ the report concludes that the core remains-covered throughout the transient with minimum safety injection, provided the safety injection flow is not *interrupted.
The report also establishes,that the syitem stabilizes at a P79 130 20
-32d-Salem 1&2
pressure (where safety injection flow matches break flow) well above the accumulator set pressure.
No clad heat-up is. expected for the head vent size break because no core uncovering is expected during such a transient.
In view of the above discussion, the present envelope of accident analyses remains valid and no new analysis for a loss of coolant accident initiated by a head vent pipe break is required.
(2)
Total hydrogen accumulated from all sources inside the containment was restudied.
The hydrogen concentration is well below the limit of 4 v/o (i.e., 4 percentage hydrogen concentration by volume) as required by the Regulatory Guide 1.7 *with the operation of one of the two hydrogen recombiners.
If, after a LOCA, the reactor vessel was vented to the contain~ent atmosphere the additional expected P79 130 21
-32e-Salem 1&2 JAN 1 198()
hydrogen released would not exceed the capability of the installed hydrogen recombiners in maintaining the hydrogen concentration below the defined limits of detonation.
P79 130 22
-32f-Salem 1&2 J.4N 1 rnRn ___ _
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FIGURE l
'-*~--------------
C.L.cAN
~li...I1"1.fE Bu l: 1..PJ: N C':i" LOCATION OF TECHNICAL SUPPORT CENTER IN RELATION TO THE CONTROL ROOM 42b -
Salem 1 & 2 I II l..i
'91'\\ t'\\"
A.
(Cont'd) in the building's mechanical equipment room adjacent to the TSC.
Light*
ing will be Rrovided by temporary fixtures, until completion of modifica-tion to a pennanent TSC.
The TSC will be upgraded during 1980 to satisfy the long term require-ments identified for January 1, 1981.
The modifications will include:
- 1. The Technical Support Center will be divided into four functional areas as shown in Figure 2.
- a.
The operational area with work facilities for 25 people, 1600 ft2*
- b.
An open lecture area with chairs for thirty persons, approximately 350 ft 2*
- c.
d
- An enclosed conference room approximately 300 ft2.
A bunk room for 10 persons, approximately 250 ft2*
. 2.
Installation of a dedicated HVAC*system, serving the TSC and selected adjacent locker rooms and toilet facilities. This system will utilize HEPA and charcoal filtration and will be isolable from outside air by dampers.
The existing HVAC system will also be modified to permit* isolation of the remainder of the building from outside air.
Isolation will occur automatically via an intake duct isolation signal derived from a radiation monitor insta-led in the intake duct.
Isolation may also be initiated via a manual control switch in the TSC or by manual operation of the dampers.
- 3. Radiation shielding will be installed in the operational area, open lecture area, and conference room to meet habitability requirements specified in GDC 19 and SRP 6-4. This will.primarily consist of installation of block walls inside of the prefabricated steel skin of the TSC area, and installation of shielding on the roof.
Equivalent facilities will be provided if further investigation shows that the above modifications are not feasible.
42c -
Salem 1 & 2
t:, t.JC: J :r:
urm: rr
- cr:z:r I
- o
... r-'L_;'.
. t.,...;_'--
FIGURE 2
...*.,. I ARTIST'S CONCEPT OF TECHNICAL SUPPORT CENTER ~ 1981 42d -
Salem l & 2 llUI _ -1 110f\\
A.
(Cont'd)
- 4. The present power supplies for the Clean Facilities Building are two 480.V and two 240V feeds from the No. 1 Unit Group Susses.
These feeds will constitute the normal power supply for the TSC dedicated HVAC sys tern, lighting, and equipment.
Emergency power to the TSC will be provided by a dedicated diesel generator which will start automatically on loss of normal AC power to the TSC, and may be manually started from the TSC.
The diesel generator wi 11 be provided with s uff i ci ent fuel to operate for twenty days.
If the diesel generator is started by a loss of nonnal AC power, it will automatically assume the TSC loads.
If the diesel generator is started manually by personnel in the TSC, and normal power is available, the control circuits will pennit, at the discretion of the TSC personnel, (1) the diesel generator to assume the TSC loads (although normal po\\'1er is available) or (2) the diesel generator to idle and transfer automatically should nonnal AC power be lost.
42e -
Salem 1 & 2
B.
ENGINEERING/MANAGEMENT SUPPORT AND STAFFING Activation of the Onsite Technical Support Center
- Activation of the Technical Support Center (TSC) will corrmence in accor-dance with the "Alert" level, defined in the "Draft Emergency Action Level Guidelines, NUREG-0610 11 dated September, 1979. The "Alert" level is described as an event in progress or having occurred which involves an actual or potential substantial degradation of the level of safety of the plant.
More significant events calling for the declaration of a "Site Emergency" or "General Emergency", would also call for activating the Technical Support Center.
The Emergency Duty Officer (EDO) has the responsibility to classify an event.
In his.absence the Senior Shift Supervisor shall make an initial classification and notify the EOO.
Staffing of the Onsite Technical Support Center The Senior Shift Supervisor notifies the Emergency Duty Officer (EDO) of an event. The EDO will be responsible for seeing that the appropriate personnel are notified as shown in the following figure:
SENIOR SHIFT SUPERVISOR I
HEADQUARTERS EMERGENCY DUTY ELECTRIC PRODUCT.l"IIO""""'N___
OFFICER STATION MANAGER EMERGENCY PLAN I
LOCAL STATE AND SUPPORT PERSONNEL*----------------. FEDERAL AG ENC I ES 42f -
Salem 1 & 2
. I l\\ ~'
1 1 o or'-
B.
(Cont'd)
In addition to staffing the initial survey requirements of the Station's Emergency Plan implementing procedures, the Electric Production Depart-ment wil 1 staff the TSC as required to provide cormJunfcations with survey teams and the Control Room, radiation plume plotting, data logging and
.miscellaneous activities such as personnel accountability and callout of additional personneJ.
Personnel to perfonn the above functions wil 1 be assembled in the Technical Support Center from existing operating and support staff, to comply with the requirements of the "Emergency Planning Acceptance Criteria 11 for licensed nuclear power plants.
The Emergency Duty Officer will be available by.telephone to provide guidance to the TSC staff.
The Emergency Duty Officer will be provided with a vehicle equipped with two-way corrmunica-tions equipment to provide for continuous communications with the TSC during the ED0 1 s transit to the station.
The Emergency Duty Officer will be at the TSC within 90 minutes of notification by the Senior Shift Supervisor.
In addition to the above, the following station*personnel will be available for duty at the Technical Support Center within four hours:
- 1.
Back-up for the Emergency *Duty Officer.
- 2.
A licensed Senior Reactor Operator for liaison with the technical support team from Headquarters.
Headquarters Support for the Onsite
- Technical Support Cent~r The Newark Headquarters support for the Technical Support Center wil 1 be activated upon a call for assistance from the Electric Production Department to the Engineering and Construction Department.. Each Depart-ment will activate its plan for providing support by communicating as shown in the following figure.
42g -
Salem 1 & 2
B.
(Cont'd)
ELECTRIC PRODUCTION DEPARTMENT (EPD)
ENGINEERING DEPARTMENT
- 1 ENG. & CONST. DEPT.
SUPPORT PERSONNEL FOR SITE DUTY ENG. & CONST. DEPT.
SUPPORT PERSONNEL FOR HEADQUARTERS DUTY EPD SUPPORT PERSONNEL FOR HEADQUARTERS DUTY The initial emergency response and staffing of the Onsite Technical Support.center and the Newark Headquarters by Engineering and Construction Department personnel wi 11 be as described below.
Long-tenn recovery actions and Headquarters support functions may be relocated as necessary.
L Engineering Department Management Contact Responsible for establishing initial contact with Engineering and Construction Department personnel to start the mobiliza-tion and.assembly of people at assigned locations.
2.. Site Engineering and Construction Department Personnel Upon arrival at the site the assigned personnel will report to the TSC.
The TSC will be under the direction of the Emergency Duty Officer.
The personnel reporting to the site will assist, under the direction of an assigned leader, in analyzing plant conditions and recommending actions to be taken.
Personnel experienced in the mechanical, electrical. controls and radiological disciplines will be available.
The above personnel will be available at the site within four hours of notification.
42h Salem 1 & 2
._ 1 __
B.
(Cont'd)
- 3.
Newark *Engineering and Construction Department Headquarters Personnel The personnel notified will report to a designated area of the Corporate Headquarters within two hours after notification.
Under the direction of an assigned leader, they will assist the Site Engineering support team 1n analyzing plant conditions and*
recommending actions to be taken.
The team wil 1 consist of ptrsonnel experienced in the mechanical, electrical and controls disciplines. Design personnel will be available as required to support the above disciplines.
Staffing of the Newark Headquarters by Electric Production Departnent personnel will be as follOtt1s:
- 1.
Electric Production Depa rtrnent Management Responsib 1 e for establishing contact with Engineering. Departnent Management and selected EPD personnel for mobilization and assembly at assigned locations.
- 2.
Newark Electric Production Departnent Personnel The EPD personnel notified will report to a designated area of the corporate headquarters, within two hours after notification.
Under the direction. of an assigned leader, they will assist in analyzing plant conditions and recorrmending actions to be taken.
The team will consist of personnel with experience and responsi~
bilities in disciplines such as ~lant operations, core analysis and safety.
42i -
Salem 1 & 2
c~
COMMUNICATIONS The TSC will be provided with fifteen telephones for general co111T1unica-tions, which will be a combination of outside lines and extensions on the Salem station telephone system. Additional telephones, delineated below, will be provided for special functions.
- 1.
The four dedicated telephone lines for (a) New Jersey State Police, (b) Lower Alloways Creek Township Municipal Building, (c) NAWAS, (d) NRC (Bethesda) which presently are located in the Senior Shift Supervisor's Office will be "bridged" to the TSC, so that calls may be initiated or received from either location.
- 2. Dedicated telephone lines will be installed between the TSC and (a) the Unit 1 Control Room, (b) the Unit 2 Control Room, and (c) the Senior Shift Supervisor's Office. The telephone in the Senior Shift Supervisor's Office will serve as corrrnunications to the Onsite Operational Support Center. *A fourth dedicated telephone line will be installed from the Headquarters Response Area to the TSC, and bridged to the Control Room.
- 3. *Five telephone extensions will be provided for data retrieval, as fol lows:
Dial-Up telephone modan (device to facilitate transmittal of digital data over a telephone line) for Unit 1 process computer, located at computer.
Dial-Up t~lephone modem for Unit 2 process computer, located at computer.
Dial-Up telephone modem for Unit 2 radiation monitoring syste~
canputer, located at computer.
Two modems in the TSC for use with typewriter tenninals for data output.
42i -
Salem 1 _&* 2
C.
(Cont'd)
- 3.
(Cont'd)
The use of telephone extensions for the aforementioned modems will pennit two typewriter terminals in the TSC to be connected to any two of the three computers.
Computer security will be assured aaninistratively by providing a disconnect switch at each computer modem.
The switches must be closed by station operating personnel to enable communication with the computer.
- The two security and emergency plan transmitters presently installed in the Senior Shift Supervisor's Office will be provided with remote con-
~rollers, located in the TSC..
- Two paging. handsets, connected to the station paging system, will be installed in the TSC.
The communications equipment d.escribed above except those marked with an asterisk will be installed for use by January 1, 1980.
Improvements in data transmission will be made as required to accc:mnodate the January 1, 1981 data acquisition plan, as outlined in Section E and for corrmunica-tion with offsite locations.
The installation of equipment marked with an asterisk is scheduled for completion before January 30, 1980.
42k -
Salem 1 & 2
- 0.
RADIATION MONITORING Radiation monitors for both direct and airborne radioactive contaminants wi11 be available in the TSC by January 1, 1980.
An area monitor, with a range of 0.1 mr to 10,000 mr and a continuous air monitor, calibrated to 1-131 will be located in the TSC.
Visual and audible alanns wi11 be set at levels prescribed by station radiation protection personnel.
Action 1 evel s to define requirements for protective measures {such as using breathing apparatus and potassium iodide tablets or evacuation to the control room) will be delineated in the Station Emergency Plan Implementation Manual *. Nonnally these decisions will be made by the Emergency Duty Officer based on several factors during the accident, such.
as location of plume, iodine dose levels, and area and ventilation system radiation levels. Bio-pacs and a supply of 300 mg potassium iodide tablets will be located in the TSC..
421 -
Salem 1 & 2
-~
E.
PLANT INFORMATION DISPLAY Display of plant parameter information in the TSC to be available by January 30, 1980 will consist of data links to each unit's plant computer and the Unit #2 Radiation Monitoring System (RMS) computer. *Data pre-sentation will consist of a slave CRT which will display in the TSC, any infonnation requested by the operators in the plant control* room.
In addition to this CRT display, a typewriter terminal will be available in the TSC which will have the capability to access any of the plant data stored in the computer. A pre~selected number of key plant parameters will be "trended" upon request.
Attachment E.l is a compilation of the pre-selected parameters included to be trended on the typewriter terminal in the TSC.
This list can be modified fran the typewriter tenninal.. This action is independent of the Control Room operator.
The type of data to be available by January, 1981, is under study. Attach-ment E.2 indicates a preliminary.assessment of data to be available.
42m -
Salem 1 & 2
ATIACliMENT E.1 TSC DATA AVAILABILITY - 1980 CRT TYPEWRITER SYSTEM/PARAMETER DJ SPLAY TE~INAL A.
Core
- 1. Average va 1 ue of co.re exit thenoocoup1 es x
- 2. Control rod position x
- 3.
Individual core thenoocouples x
- 4. Source-range flux x
- 5.
Intenne*di ate-range fiux x
B.
- l. Hot and cold leg temperatures x
- 2. Average temperature x
- 3.
W1 de-range pressure x
- 4. Pressurizer pressure x
- 5.
Pressurizer level x
- 6.
Loop fl°"
x
- 7. Degree of subcoo1ing (PSAT/TSAT)
- X
- 8. Reactor coolant pump status x
- 9. Pressurizer Relief Tank Level x
Temp x
Pressure x
- 10. Reactor coolant activity x
- c.
Containment
- 1.
Pressure x
- 2. Temperature x
- 3. High-range radiation x
- 4. Fan-cooler unit status x
42n -
Salem 1 & 2
CRT TYPEWRITER SYSTEMf PARAMEiER DISPLAY TE~INAL D.
Steam and Feed-Kater Systems
- 1. Steam generator outlet pressure x
- 2. Steam generator wide-range level x
- 3.
Steam generator narrow-range level x
- 4. Steam generator blo.idown radiation x
- 5.
Main feedwater flow x
- 6.
Ma in steam fl ow x
- 7.
Condenser 11r removal radioactivity x
- a. Auxiliary feedwater pllTlp status x
E.
Auxiliary Systems
- l. High-pressure injection flow x
2.
ECCS PLlllPS status x
- 3.
RHR flCIPr' (hot-legs) x
- 4.
RHR heat exchanger outlet temp.
x
- 5.
Component cooling water temp.
x
- 6.
Component cooling water flow x
- 7. Service water temp.
x
- a. Boric acid charging flow x
- 9. letdown flow x
- 10.
Component cooling pump status x
- 11. Service water pump status x
F.
Power Supplies
- 1. Status of cl ass 1E supplies x
- 2. Status of non-class lE supplies x
G.
Radiation Monitoring (Unit 11)*
- 1.
- Various area 11oni tors x
- 2. Cont.a irrnent gas, iodine, particulate x
- Unit 12 RMS infonnation will be available from the Unit 12 Control Room.
Activities &re 1n progress to provide this data in the TSC.
420 -
Salem i. & 2 I Ill *..
SYSTHV PARAMtTER H. Met20ro1ogy
- 1. Wind direction and speed
- 2.
Vertical temp. difference I. Ventilation Systems
- 1.
- Various iiea temperatures 1n Auxiliary Building J.
- Additional Special CRT Displays
- 1. Alann review
- 2. Control rod profile
- 3. *Reactor profile
- 4.
Coolant system water inventory
- 5.
RCS leakage
- 6.
Reactor subcooling K. Additional Special Typewriter Reports
- 1. Flux map
- 2.
In-core thermocouple map 42p -
CRT DISPLAY x
x x
TYPEWRITER TERM I HAL Salem 1 & 2 r II ! '
.J
- e
!JJf}.::jt!EHT E. 2 TSC DATA AVAILABILITY - 1981 (HOT£: All parameters to be 1ndicated or displayed upon demand. Asterisk (*)
denotes recorder to be provf ded. )
SYSTEM/PARAMETER A. Core
- l. Average value of core thennocouples
- 2. Control rod position
- 3. Core thennocouples
- 4. Source range flux.
- 5.
Intennediate range flux
- l. Hot and cold leg temperature
- Z. Average temperature
- 3. Wide-range pressure
- 4. Pressurizer level and pressure
- 5.
RCS loop flOiii'S
- 6. Degree of sub-cooling (PSAT/TSAT)
- 7. Reactor coolant pump status
- 8. Pressur;zer relief tank temp, level,
- *9 *. RCS activity pressure
- 10. Pressurizer safety/relief valves position
- c.
Containment
- l. Pressure
- 2.
Tempera tu re
- 3.
Hydrogen concentration
- 4. Sump water level
- 5.
High-.range radiation
- 6.
Plant vent monitor
- 7.
Isolation valve status
- 8.
Fan-cooler status 42q Salem 1 & 2
. SYSTEM/PARAMETER (Cont'd)
D.
Steam and Feedwater Systems E.
- 1. Steam genera tor pressure
- 2. Steam generator 1eve1 (narrow and wide ranges) eJ. Steam generator b1owdown radiation
- 4. Auxiliary feedwater flow and PllllP status S.
Mafo feedwater flow
- 6. Main steam flow
- 7.
Aux. feedwater,,storage tank level
- 8. Condenser air removal llOnitor
- 9. Hotwell level Auxiliary Systems
- 1. High-pressure injection flow
- 2.
ECCS pump status (spray, SI, charging,
- 3.
- RHR flaw (hot and cold 1egs)
- 4.
.RHR heat exchanger outlet temp.
- 5.
Component cooling water temp.
- 6.
Component cooling water flow
- 7.
Service water temp
'*8. Boric acid charging flow
- 9.
Letdown flow
- 10. Canponent cooling pump status
- 11. Service water pump status
- 12. Spray additive tank 1 evel
- 13.
RWST level
- 14. Accumulator levels
- 15.
Volume control tank level
- 16. Boric acid tank. level
- 17. Boron injection tank level
- 18. Component cooling surge tank level
- 19. Essential valve positions
.)
42r -
RHR)
Salem 1 & 2
. I l'I "'
SYSTEM/PARA"1ETER (Cont'd)
F. Power Supplies
- l. Status of c1oss 1E power supplies
- 2. Status of non-class lE power supplies
- 3. Diesel auxiliaries
- 6. Radiation Moni tcr1ng
- l *. Various area monitors
- 2. Contairrnent gas, iodine, particulate
- 3. Plant vent gas. iodine, particulate
- 4. Expanded display of monitor alanns, locations and trends H.
Meteorology
- l. Wind direction and speed
- *2. Vertical temp. di ff ere nee I. Ventilation Systems
- 1. Various area temperatures in Auxiliary Building
- z.
TSC envirorrnent control system status 42s -
Salem 1 & 2 1"~
1 1qac
F.
ACCIDENT ASSESSMENT FROM CONTROL ROOM The Station Emergency. Plan Implementation Manual provides for performing assessment functions from the Control Room or nearby Senior Shift Supervisor's Office during accident conditions should the TSC become un-inhabitab 1 e.
G.
LONG RANGE PLANS The long range plans for the TSC have been identified in the previous sections.
42t -
Salem 1 & 2
.,, ll
..f
.41\\1\\"'
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 01/52 Salem 1 & 2 JAN
- 1 1980