ML19210E093
| ML19210E093 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/27/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7911290300 | |
| Download: ML19210E093 (38) | |
Text
V Wisconsin Electnc m coum 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE. WI 53201 November 27, 1979 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.
S. NUCLEAR REGULATORY COMMISSION Washington, D.
C.
20555
Dear Mr. Denton:
DOCKETS 50-266 AND 50-301 IMPLEMENTATION OF NUREG-0578 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On November 5, 1979, we received your letter dated October 30, 1979, which discusses implementation schedules and provides clarifications of the NUREG-0578 Short Term Lessons Learned requirements.
Attached herewith is a detailed summary of our commit-ments and schedules for implementation of these short-term requirements.
We note that certain of the material provided in to your October 30 letter are modifications of the initial proposals transmitted with your letter of September 13.
Also, we understand your Staff did not have time to complete its review of our October 20 response thereto.
This reply incorporates the clarifications and cther matters requested in your October 30 letter.
Based on telephone conversations with your Staff on November 19, and our telegram of November 20, we have also included supplemental information to address and clarify specific items raised by your Staff.
We call your attention to two items where we are not in complete agreement with the Staff position.
These are items 2.2.1.b, Shift Technical Advisor, where we propose a duty and call system in which the Technical Advisor would be available in thirty minutes instead of ten minutes; and in the completion of the permanent technical support center where state of Wisconsin certification, building permit and physical construction require-ments may make completion by January 1, 1981 uncertain.
For further details in respect to these and the other items, we refer you to our October 20 response and to the attached senmary.
1431 32l 7 9112 9 o 3 cac>
~
Mr. Harold R.
Denton November 27, 1979 We trust that the Staff will complete its review of our October 20 initial response regarding implementation of the Lessons Learned items from NUREG-0578.
We have made a specific and detailed analysis of the Three Mile Island accident and its relation to our Point Beach facility.
Our commitments and schedules for implementation have been carefully and deliberately specified.
They represent our best realistic assessment of work required for accomplishment.
In some areas, we have offes;ed comments as to the purpose and need for the Staff recommended action.
We believe these comments are pertinent and of sufficient substance to have them considered by the Staff in its continuing review and analysis of lessons learned from Three Mile Island.
These additional comments do not indicate a lack of commitment to comply with the actions required or the implementation schedule, but are intended to provide constructive input to this ongoing analysis.
We have also called attention to the major design differ-ences between Point Beach and Three Mile Island as identified in our Task Force Report attached to our October 20, 1979 letter.
These differences significantly reduce the possibility of and consequence of a Three Mile Island-type accident at Point Beach and, of course, affect the nature of implementing the lesson learned requirements.
We are confident that the measures and schedules for compliance with NUREG-0578 requirements are satisfactory and realistic.
We believe that continued operation of Point Beach without these changes does not present any undue hazard to the health and safety of the public or to the plant employes charged with its safe operation and maintenance.
We would again offer to meet with you and members of your Staff to review any of these items further.
Very truly ycurs, Sol Burstein Exe utive Vice President Enclosures 1431 322
e 2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs Pressurizer Heater Power Supply A.
Position The current Point Beach Nuclear Plant (PBNP) pressurizer heater power supply, as described in detail in the October 20, 1979 submittal meets all of the criteria stated in the Staff position and clarification.
Original plant design includes appropriate equipment for this purpose.
B.
Discussion Procedures to specify the proper timing and loading of the pressurizer heaters will be completed in conjunction with the emergency procedure effort underway wiu.in the Westinghouse Owners' Group.
In establishing these procedures, careful consideration-will be given as to which ESF loads may be appropriately shed for a given situation such that the operator has clear criteria to prevent over-loading a diesel generator.
C.
Schedule The procedure effort will be completed by January 1, 1980.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators A.
Position The current PBNP pressurizer relief and block valves and pressurizer level indicators power supplies, as described in the October 20, 1979 submittal and detailed below, meet all of the criteria stated in the Staff position and clarification.
B.
Discussion
..s The attached .igure 2..l.1-1 and Tables 2.1.1-1 through -3 provide the pressurizer relief, solenoid and block valve configuration and power supply distribution for the pressurizer relief and block valves, air compressors, and pressurizer level instrumentation channels, respectively.
Both the red and blue instrument buses are vital instrument grade buses.
The yellow and white instrument buses are non-vital.
The white bus is supplied from the respective unit B02 AC/AC motor-generator set which serves to eliminate short term transients (power spikes) on the white bus.
ihb\\
2.1.1-1
f B.
Discussion (Continued)
Nos. 1 and 2 batteries are shared between units; the inverters are unique to each unit.
Also attached is Figure 2.1.1-2 (previously supplied to the NRC with other subnittals) which is a simplified electrical distribution diagram for the PBNP and combines Figures 8.2-1 and 8.2-2 of the Final Facility Description and Safety Analysis Report (FFDSAR).
Prefix numbers in this figure and the tables refer to unit designation (e.g.,
1(2)-PCV-431C indicates the same configuration applies to the valve for both Units 1 and 2).
The power operated relief valves, 1(2)PCV-430 and 1(2)PCV-431C, are air operated valves which close on loss of air pressure to the valve actuator diaphragm.
Instrument air system pressure is applied to the power operated relief valve actuator through solenoid valves 1(2) SOV-4 30 and 1(2) SOV-4 31C, respectively.
The solenoid valve is actuated by two pressurizer pressure bistable / relay combinations.
Both bistables must actuate to operate the solenoid valve allowing the relief valve to open.
If a pressurizer pressure channel fails low (Channel 429, 430, 431, or 449),
the power operated relief valve associated with that channel cannot open.
Instrument air is supplied through containment penetra-tions from either of two instrument air compressors, K2A and K2B.
This is backed up by the service air system which is supplied by either of two service air compressors (K3A and K3B).
Instrument air pressure at 90 psig is normally supplied by the instrument air compressors.
If the instrument air system falls to 80 psig, service air is then supplied to the instrument air system.
This is sufficient pressure to operate air operated valves supplied by the instrument air system.
The instrument air compressors are stripped on a safety injection coincident with a loss of AC.
They must be manually restarted.
The instrument air headers inside containment are isolated when a containment isolation signal is initiated for,qach unit.
In the case of automatic safety injection, this isolation remains until safety injection is reset and the instrument air system is unisolated.
The instrument air system may be unisolated before safety injection is reset as long as the control switch for the instrument air contain-ment isolation valve is manually held in the "open" position.
A backup pressurized nitrogen gas bottle is provided to supply nitrogen gas to operate the power operated relief valves in the event of instrument air system pressure loss when low temperature-overpressure protection is required during plant shutdown.
1431 324 2.1.1-2
e C.
' Schedule No additional action is required.
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FIGURE 2.1.1-1 PBNP PRESSURIZER POR AllD BLOCK VALVE SCHEMATIC
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TABLE 2.1.1-1 PBNP PRESSURIZER POWER OPERATED RELIEF AND BLOCK VALVE POWER SUPPLIES ITEM POWER SUPPLY VALVE APPLICABILITY Bistable / Relay 1(2)PCV 1(2)MOV 1(2)SO3 Combinations 430 431C 515 516 430 431C Channel 429 No. 1 battery / inverter for X
Backup the unit, red instrument to Ch.431 bds AC Channel 430 No. 1 battery / inverter from Unit 1 Unit 2, red instrument bus AC No. 2 battery / inverter from Unit 2 Unit 1, blue instrument bus AC Channel 431 No. 2 battery / inverter for X-the unit, blue instrument bus AC Channel 449 1(2) BO4 ESF bus, yellow Backup to X
instrument bus AC unit's Ch. 430 Motive Power 1(2)B03 ESF bus X
1(2)BO4 ESP bus X
Control Power es 1(2)B03 ESF bus X
LN 1(2) BO4 ESF bus X
125V DC
[]
No. 1 battery X
Co No. 2 battery X
t TABLE 2.1.1-2 PBNP AIR COMPRESSOR POWER' SUPPLIES AIR SUPPLY ENGINEERED S AFEGUARDS BUS COMPRESSOR SYSTEM 1B03 2B03 1904 2B04 K2A Instrument X
K2B Instrument X
K3A Service X
K3B Service X
1431 329
i TABLE 2.1.1-3 PBNP PRESSURIZER LEVEL INSTRUMENTATION POWER SUPPLIES PRESSU RE LEVEL CHANNEL POWER SUPPLY 1(2) L426 No. 1 battery / inverter for the unit, red instrument bus 1(2) L427 1(2)B02 AC/AC motor-generator set nonsafe-guards bus, white. instrument bus 1(2) L4 28 No. 2 battery / inverter for the unit, blue instrument bus 1(2) L4 3 3 No. 1 battery / inverter for the unit, red (Cold Calibration) instrument bus 1431 330
t 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves A.
Position As stated in the October 20, 1979 submittal, the Licensee is a member of the Westinghouse owners' C oup which is working in conjunction with the other PWR owners and EPRI to develop an appropriate industry testing program and schedule to demonstrate valve operability under various flow conditions.
We will continue our participa-tion in this program and will provide technical and financial support as required.
B.
Discussion We believe that our previous comments on the need for such testing are pertinent and of sufficient substance to merit NRC Staff consideration in its. ongoing review of lessons learned from Three Mile Island.
The current design basis and Code specifications in this area require redundant Code safety valves.
nonisolatable, on any pressurized system as a means of overpressure protection.
In addition, the discharge from the Code safety valves must be unrestricted in its capability to ultimately relieve to the atmosphere.
To reduce the number of challenges to the Code safety valves because they may not reclose and cannot be isolated, the design of Point baach and other PWR primary coolant systems also includes one or more power operated relief valves (PORV).
Automatic control systems are provided to operate the PORVs and provide relief of RCS pressure at a setpoint below that of the Code safety valves.
Manual actuation is also provided.
It is our understanding that, during power operation, this design has eliminated all challenges to the Code safety valves and that, in fact, no code safety valve has yet been required to open on a Westinghouse PWR plant.
Additionally, the basic design of the inverted U-tube steam generators with a large secondary side water inventory (thirty minutes or greater to dryout without AFW) and safety grade auxiliary feedwater system provide substantial post-accident heat removal capability.
This serves to limit the primary side heatup and pressure transient and severely restricts even the challenges to the PORVs.
Since the PORVs are expected to open, the design also provides block valves in series to insure isolation of the RCS should the PORVs fail open.
The operation of the PORV and block valve is ensured by separation of power supplies and actuation logic as described in 2.1.2-1
t B.
Discussion (Continued) item 2.1.1.
This design is a proven, accepted practice in pressurized systems, both nuclear and non-nuclear, and meets all Code and NRC licensing requirements.
The transient and accident analyses specifically exclude both the protection (relief) and control (relief-isolation) capabilities of the POI.V and block valves.
Instead, the overpressure conditions analyzed must use the Code safety valve setpoints.as the upper bounds of the calculations.
Conversely, the LOCA analyses must consider a range of breaks which cover an assumed PORV or safety valve opening.
Thus, the function of reclosure is strictly an operational benefit without unanalyzed safety concerns.
It should be noted that while the pressurizer relief tank (PRT) provides for the suppression of PORV or safety valve releases and eventual unrestricted release to containment via the tank rupture disk, containment pressure calculations must assume direct release to the containment for maxi-
~
mizing pressure.
Piping system integrity downstream of the safety valves is, thus, not necessary.
In light of these Code requirements, licensing and design bases, and analyses, it is unclear as to the objectives of the committed testing program required by vac NRC.
We know of no plan for a redetermination as to whether the Code safety valves are to provide control in addition to overpressure protection.
Studies of PORV-block valve configurations, piping, and supports are not safety concerns and should rather be part of long term operational studies.
We would be pleased to participate in any review of this matter.
C.
Schedule The Westinghouse PWR Owners' Group or industry (EPRI-NSAC) program description will be submitted by January 1,
- 1980, with a testing program to follow.
.s 1431 532
e 2.1.3.a Direct Indication of Power-Operated Relief Valves and Safety Valve Position for PWRs.and BWRs A.
Position The power operated relief valve indication, as described in the October 20, 1979 submittal, meets the NRC requirements for direct indication.
The stem-mounted limit switches from which the indication is derived are safety-grade, seismically qualified, and environmentally qualifed as discussed below.
Direct indication of position for the Code safety valves will be provided.
B.
Discussion The basic operational and safety requirement is that the operator be able to determine the condition of the plant and take appropriate action in response to an accident condition.
The Licensee continues to be committed to providing sufficient and necessary instrumentation in the control room such that the operator has this accident assessment capability.
The present RCS instrumentation provides the operator with redundant, safety grade instrumentation which permits the determination of the existence of a LOCA regardless of whether the LOCA is due to an open -Code safety or relief valve or valves.
Revised emergency procedures have been developed and the operations of the safety systems are based on plant parameters which are independent from any valvo position indication.
3hstrumentation is provided, as detailed in the October 20, 1979 submittal, so that the position of individual power operated relief ralves and Code safety valves can be determined.
Even if new redundant, seismically and environmentally qualified (safety grade) valve position indications were to be provided, the availability of this information to the operator would not change any action (s) to be taken during the course of an accident.
The current PORVs have direct. positive indication in the control room from seismically qualified, stem-mounted limit switches for both units.
Replacement of these switches will upgrade them so that they are also environmentally qualified.
This upgrading was completed during'the recent November 1979 refueling outage for Unit 1.
It is excected to be completed during the Spring 1980 refueling outage for Unit 2 scheduled to start in March.
There is also an alarm on high pressure at the setpoint at which the valves will open.
}43l
)
2.1.3.a-1
i C.
Schedule The replacement of limit switches on the Unit 2 power operated relief valves will be completed in April of 1980.
Implementation of direct position indication for Code safety valves will be completed by January 1, 1980, depending upon cquipment availability.
To accomplish this schedule for Code safety valves, acoustic devices to indicate valve operation or leakage are being procured.
It is possible that unit shutdowns will be required to permit installation, r
1431 334 2.1.3.a-2
f 2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs I.
Procedures and Description of Existing Instrumentation Analyses and procedure guidelines will be submitted to the Bulletins and Orders Task Force through the Westinghouse Owners' Group as detailed in the October 20, 1979 submittal.
II.
Subcooling Meter A.
Position The installation of RCS saturation condition monitoring and alarm will be provided by the Licensee as stated in the October 20, 1979 submittal.
B.
Discussion 1.
A two step approach was previously described utilizing the plant process computer initially and a separate system later.
The initial process computer system implementation has been completed and uses multiple incore thermocouples for temperature input and can use either the redundant safety grade pressurizer pressure signals or the wide range pressure signal as input for the pressure.
This system will be upgraded and include extended temperature range for the incore thermocouples.
Display of the subcooling meter information will be provided on operator demand and an alarm will sound should less than 50*F subcooling occur.
The current system is described in the attached tabulation.
2.
Implementation of the separate system for monitoring subcooling as an integral part of a larger diagnostic system is currently under review, with the aid of a consultant.
C.
Schedule 1.
The interim computerized system described above has aircady been incta_'ed.
2.
Implementation of a separate system will proceed according to the schedule presented in our letter of October 20, 1979.
1431 535 2.1.3.b-1
t III.
Design and Installation of New Instrumentation A.
Position 1.
The Westinghouse owners' Group program for analyses and procedures to address inadequate core cooling will provide the basis for any new instrumentation.
2.
Although the Owners' Group analysis has not yet been completed, we will provide reactor vessel level instrumentation as requested by NRC.
B.
Discussion The request for vessel level indication to cover the full range from normal water level to the bottom of the core will be included.
The additional features requested by the NRC are being included in the evaluation.
Means of reactor vessel level indication are being considered together with reactor vessel head vent arrangements.
C.
Schedule 1.
The schedule for the analyses and procedures by the Owners' Group is g3.ven in item 2.1.9 of the October 20, 1979 submittal and concludes by the end of the first quarter of 1980,.as agreed with the Bulletins and Orders Task Force.
2.
Ye will provide the conceptual design of reactor vessel level instrumentation by January 1, 1980.
Detailed design will be provided as soon as possible thereafter.
3.
A description of any other new instrumentation, functional design requirements and schedule for installation will be provided after completion of the Owners' Group analysis.
NRC proposal review is required before implementation per the NRC clarification schedule of October 30, 1979.
4.
Instrumentation requiring modification to the RCS and additional wiring to be pulled into containment would be implemented at the earliest possible scheduled refueling outage for the unit.
Present schedule for Unit 2 is Spring 1980 and for Unit 1 is Fall 1980.
Implementation of vessel level indication will be completed by January 1, 1981, assuming valve and instrument delivery schedules.
)kb\\
b 2.1.3.b-2
INFORMATION REQUIRED ON THE SUBC00LIftG METER Display Information Displayed (T-Tsat, Tsat, Press, etc.)
T, Tsat, Press as selected Display Type (Analog, Digital, ERT)
Digital Continuous or on Demand On Demand Single or Redundant Display Single Location of Display Control Room Reactor Operators Console Alarm typewriter, 50 F setpoint Alarms (include setpoints)
Overall uncertainty (*F, PSI)
(+_5 F, +15 psi g)
O to 9999.9'F Range of Display Qualifications (seismic, environmental, IEEE323)
Original Plant Design Calculator Type (process computer, dedicated digital, or analog calculator)
PRODAC 250 Computer If process computer is used specify availability 100% (Monitoring Computer
(% of time)
Only)
Single or redundant ScuMtors Single Selection Logic (higries: T., icwest press)
Single Pressure Channel -
Wide Range Average Incore Thermocouple Qualifications (seismic, environmental, IEEE323)
Original Plant Design Calculational Technique (Steam Tables, Functional Steam Tables Fit, Ranges)
Input Temeerature (RTDs or T/Cs)
Incore Thermocouples Temperature (nurter of sensors and locations)
All Available Incore Thermocouples - Max. 39 Range of Temperature Sensors 0 - 2000 F 1431 M__/,
2,1.3.b-3
-1 g
[
l Uncertainty
- of Temperature Sensors (*F at 1) iS*F 9 620*F (175% x temp)
Qualifications (Seismic, Environmental, IEEE323)
Original Plant Design Foxboro Model 611 Pressure (specify instrument used)
Pressure Emitter 1 Sensor Reactor Coolant Pressure (nunter of sensors and locations)
Hot Leg 0 - 3000 psig Range o Pressure Sensors 0.5% of full scale Uncertainty
- of pressure sensors (PSI at 1)
(115 psig)
Qualifications (Seismic, Environmental, IEEE323)
Original Plant Design Backuo Capability, Located in Control Room -
Availability of Teco and Press Wide Range Pressure Incore Thernoccuple Readout Located in C)ntrol Room Availability of Steam Tables, etc.
Yes Training of Operators E0P 1, 2 and 3 revisions Procedures include subcooling implications
.+
- Uncertainties must address conditions of forced flow and natural circulation 1431 338
2.1.4 Containment Isolation Provisions for PWRs and BWRs A.
Position All of the NRC position and clarification items are satis-fled by the current Point Beach Nuclear Plant containment isolation system actuation and reset features.
B.
Discussion As stated in the previous submittal on October 20, 1979, the required information pertaining to the containment isolation system utilized at the Point Beach Nuclear Plant has been described in detail to the NRC Staff in Wisconsin Electric Power Company's response to Item 9 of Bulletin 79-06A dated April 27, 1979.
That submittal demonstrates that all of the NRC position and clarification items are satisfied by the current Point Beach Nuclear Plant containment isolation system actuation and reset features.
Specifically, resetting bf. the containment isolation signal will not automatically result in the reopening of any containment isol'ation valves.
Deliberate operator action is required to reopen any isolated valves.
C.
Schedule The evaluations committed to be provided are currently in progress and will be completed by January 1, 1980.
These include the identification of essential and nonessential systems and system isolation status (normally isolated durinJ operation, isalated on containment isolation, and not isolated).
A 2.1.4-1
2.1. 5. a)
Post-Accident Hydrogen Control Systems for PWR and 2.1.5.c)
BWR Containments A.
Position Items 2.1.5.a and 2.1.5.c do not apply to Point Beach Nuclear Plant as related to hydrogen recombiners since recombiners are not a part of the design or licensing basis for the plant.
C.
Schedule The purge system design review calculation committed to in item 2.1.5.a in the October 20, 1979 submittal has been made and verifie.s the original design with respect to the flow capacity of the purge system.
This corapletes the required action for these items.
1431 540 2.1.5.a-c-1
e 2.1.6.a Integrity of Systems outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs.and BWRs A.
Position We will implement an immediate and continuing leak identification and reduction program and provide a summary description consistent with the NRC clarification.
B.
Discussion All identified systems will be included in this program.
The methods, criteria, and provisions for testing of gaseous piping systems are being evaluated.
A consultant has been retained to perform this task consistent with the committed schedule.
C.
Schedule The identification of systems and summary description will be provided by January 1, 1980.
Initial testing of liquid and gaseous systems will be performed by January 1, 1980.
The complet8mm program will be implemented by January 1, 1981.
1431 341 2.1.6.a-1
t 2.1.6.b Plant Shielding Review A.
Position We will perform a shielding analysis for Point Beach Nuclear Plant and provide shielding or procedural modification where necessary.
B.
Discussion 1.
The postulated source term is extremely conservative and appears to us to be unrealistic.
We will, however, perform the analyses assuming this non-mechanistic source term.
2.
The "<l5 mR/hr" criterion,we believe,is not intended for cases where continuous occupancy does not involve the same person for 12 hrs / day for 30 days.
However, the analysis will identify those cases, if any, where credit is taken for occupancy by different persons for areas otherwise assumed to require continuous occupancy.
Although not apparent from the statement of the criterion, we are making appropriate allowance for radioactive
- decay, e.g., an average dose rate of 15 mrem /hr when averaged over an assumed period of 30 days following an accident.
C.
Schedule.
1.
Initial analysis will be complete by January 1, 1980, but detailed engineering design for required modifications may not be complete before July 1, 1980.
2.
System identification has already begun.
A consultant has been retained, and calculational work is in progress.
Steps of the analysis are described in our submittal of October 20, 1979.
3.
If shielding modifications are to be made, every effort will be made to complete such modifications by January 1, 1981.
The nature and extent of any modification is, of course, presently unknown.
We will keep the NRC Staff informed of progress in this area, i4 1
542
t 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs A.
Position A separate response prepared in reply to the September 21, 1979 letter from Mr.
D.
Eisenhut to Mr. Sol Burstein entitled, "NRC Requirements for Auxiliary Feedwater Systems at Point Beach" satis ~ies this item.
C.
Schedule The submittal, dated October 29, 1979, from Mr. Sol Burstein to Mr.
D. Eisenhut addressed all of the requirements not covered in the October 20, 1979 submittal and completes this item.
) b, b \\
2.1.7.a-1
i 2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generators for PWRs A.
Position The flow indicution instrumentation and schedule remain as stated in the October 20, 1979 submittal.
C.
Schedule Control grade incication has been prorided.
It is anticipated that. based on availability of safety grade equipment, this item will be completed by January 1, 1981'.
2.1.7.b-1
2.1.8.a Post-Accident Sampling A.
Position We agree to perform a sampling analysis and to make plant modifications where required.
B.
Discussion 1.
The postulated source term, e.g., 100% o.? noble gases in liquids and 100% of noble gases in gases, is extremely conservative'and, we believe, is unrealistic.
2.
Chloride analysis capability is neither appropriate nor necessary for determination of plant status or for plant-follow during post-accident conditions for Point Beach Nuclear Plant, which is located on a body of freshwater.
However, we will provide the ability to determine chloride as NRC requires.
3.
We not.e that oH and hydrogen determinations are new additions in the NRC letter of October 30, 1979.
We will incorporate these into our analysis.
4 The specific one-hour requirement for both sampling and analytical determinations may be difficult to meet.
We are making every effort to meet these criteria; we will advise the NRC if any difficulties arise in this area.
5.
We believe that nearby offsite analysis (<5 miles) should be entirely satisfactory in lieu of "onsite analysis" in the event normal plant laboratory facilities become uninhabitable.
6.
A containment under negative pressure has no potential for leaking and, therefore, has no impact on public health and safety.
Nonetheless, we will provide the capability to sample the containment under both positive and negative pressure conditions.
.s 7.
We ha.a not been able to verify that a " factor of 2 error" can be met in a one hour analysis for a high level sample.
We will provide state of the art uncertainty if the factor of 2 proves unobtain-able due to dilution errors, dead time, gain shifts, or other technical constraints.
1431 545 2.1.8.a-1
C.
Schedule 1.
This analysis will be conducted along with that for item 2.1.6.b.
As described in the response to that item, initial analysis will be completed by January 1, 1980.
2.
A consultant has been retained for this work, and calculational work is in ogress.
3.
If sampling modifications are required at the plant, work will be scheduled to be completed by January 1, 1981.
We will keep the NRC Staff informed of progress in this area.
1431 346 2.1.8.a-2
2.1.8.b High Range Radiation Monitors A.
Position We will provide increased range radiation monitoring instrumentation to the extent of commercial availability.
B.
Discussion 1.
As described in our letter of October 20, 1979, and as further discussed in the Wisconsin Electric Power Company Three Mile Island Accident Review Task Force Report transmitted therewith, we believe in-containment monitors are not necessary, since there is no purpose for such information from the viewpoint of either public health and safety or operation and recovery.
In-containment monitors are of questionable reliability and accuracy.
Containment sampling is believed to be far more practical and reliable for these purposes.
- However, we will provide these devices as directed by NRC.
2.
We believe that nearby offsite analysis (<5 miles) should be entirely satisfactory in lieu of "onsite analysis" in the event normal plant laboratory facilities become uninhabitable.
3.
We note that readouts of the high range effluent monitors are required in both the technical support center and the emergen cy operations center.
The NRC has only described the concept of an offsite emergency operations center and that it be provided with appropriate communications and facilities for assisting agenc es.
We are unaware of further i
requirements, except for the effluent monitor readout.
There may be other readouts appropriate to the technical support center or the offsite emergency operations center.
We will seek clarification of these requirements.
However, we will meet the requirements prevailing at the time of the develop-ment or construction of the permanent offsite emergency operations center.
C.
Schedule 1.
The analysis for interim procedures regarding high range effluent determination will be performed along with that for 2.1.6.b and 2.1.8.a.
This initial analysis is to be completed by January 1, 1980.
A consultant has been retained and analytical work is in progress.
2.1.8.b-1
}43
f C.
Schedule (Continued) 2.
We expect to install permanent effluent instrumen-tation by January 1, 1981, subject to equipment deliveries.
~
3.
The required in-containment monitors will be ordered as soon as specification and procurement activity can be completed, with intent of meeting the January 1, 1981 installation date.
4 Because of the limited number of suppliers and the relatively large number of nuclear plants, equipment deliveries.may be extended which could affect installation dates.
We will proceed on an expeditious schedule and will keep NRC informed of our progress.
1A31 348 2.1.8.b-2
4 2.'l.8.c Improved In-Plant Iodine Instrumentation under Accident Conditions A.
Position We will reexamine iodine sampling procedures at Point Beach Nuclear Plant as described in our letter of October 20, 1979.
C.
Schedule 1.
As described '.n our letter of October 20, 1979, an agreement between the Kewaunee Nuclear Plant and Point Beach Nuclear Plant is expected to be implemented by January 1, 1980, to provide backup analytical laboratory services for both facilities.
2.
Review of existing plant methods and procedures is underway and will be completed by January 1, 1980.
1431 349 2.1.8.c-1
~
-m
2.'l.9 Transient and Accident Analysis A.
Position Our position, as detailed in the October 20, 1979 submittal, meets all of the requirements as discussed in NUREG-0578.
The Westinghouse Owners' Group and the Bulletins and Orders Task Force are establishing the scope and schedule for additional analyses.
We support these efforts as members of the Owners ' Group.
C.
Schedule We will meet the earfy 1980 completion commitment through the Owners' Group sponsored analyses.
1431 550 t
2.1.9-T&AA
2.1.9 Containment Pressure, Water Level and Hydrogen A.
Position The commitment to provide the required instrumentation, as stated in the October 20, 1979 submittal, remains unchanged.
B.
Discussion Relative to the implementation of containment water level, the equivalent capacity of the wide range smE level instrument for Point Beach Nuclear Plant will be in excess of 350,000 gallons.
The sources are identified as follows:
Source Maximum Volume (Gallons)
Refueling Water Storage Tank 285,000 Boric Acid Storage 5,000 Accumulators (2) 17,00's Reactor Coolant System 37,000 (above the nozzles on the vessel)
Containment Spray Additive Tank 3,000 TOTAL 347,000 gallons Since the sump is the lower portion of the containment building proper, only a wide range instrument is presently contemplated.
The current level indication is discreet at 3',
S',
7',
and 9' which exceeds the range of interest (7 feet s350,000 gallons).
By limiting the initial range setting to 7' for this instrument and allowing upward recalibration following an accident, maximum accuracy will be maintained over the range of most interest.
C.
Schedule A consultant has been retained to assist in meeting the scheduled date of January 1, 1981, for completion.
1431 351 2.1.9-ACRS
2.1.9 Reactor Coolant System Venting A.
Position Reactor coolant system venting capability beyond that available with the current penetrations to the Point Beach Nuclear Plant primary system will be provided.
Implementation of reactor vessel venting modifications to the plant will occur after the required proposal review by the NRC.
B.
Discussion The Wisconsin Electric Three Mile Island Accident Review Task Force discusses Reactor Coolant System Venting at Section 5.2.1.
The reactor vessel venting arrangement is being developed along with vessel level indication provisions and will consider venting to the RCS vent tank or to containment or both.
C.
Schedule The conceptual design for a RCS venting system will be submitted to the NRC by January 1, 1?80.
January 1, 1981 will be the target dane for completion of required plant modifications consistent with procure-ment of equipment.
Plant shutdowns will, of course, be required for any vessel head penetration or vessel instrument modifications.
1431 352 2.1.9-VENT
2.2 1.a Shift Supervitor's Responsibilities A.
Position As stated in the October. 20, 1979 submiotal, the required review of the responsibilities and authority of the Shift Supervisor will be conducted.
B.
Discussion The review, while emphasizing the aspects of Item 2 of the NRC-position, has not done this to the exclusion of other items of the position. This is clearly stated in the submittal for Items 3 (training programs) and 4 (administrative duties).
Relative to Item 1, the highest level of corporate management at Wisconsin Electric Power Company which is responsible for defining and assigning the Shift Supervisor duties is the Manager, Nuclear Operations, Point Bach Nuclear Plant. We believe that this level of management has both the required direct knowledge and credibility to adequately inform the Shift Supervisors of their responsibilities and duties and the capability to monitor the resulting performance of these personnel.
The safe operation of the plant under all conditions has long been the Wisconsin Electric Power Company's primary policy as established by management and carried out by all operating personnel.
This is evidenced by the plant's record and the high standards of performance which have been maintained during its entire ooerating history. The existing pre-TMI plant policies and procedures have resulted in periodic reviews of the authority and responsibilities of key plant personnel, including Shift Supervisors, and will continue to do so. The NRC is appraised of these changes as previously stated.
We believe that these policies, procedures, and performance warrant consideration by the NRC as meeting the intent and implementation directives of Item 1 and its clarification.
We propose to continue Point Beach Nuclear Plant operations with this policy.
We will, in addition, issue a special directive, signed by the Executive Vice President responsible, which clearly reaffirms the policy, recognizes specific review responsibilities, c.nd reiterates the safety objective.
C.
Schedule The review, required revisions, and safety directive will be completed by January 1, 1980.
1431 353
- 2. 2.1,. b.
Position The implementation of this item both in intent and schedule remain unchanged from the October 20, 1979 submittal.
B.
Discussion Justification was provided with that submittal for the stated position. The increased commitments for additional analyses, procedures, training, and plant instrumentation will further enhance the adequacy of the Duty and Call Technical Advisor concept.
In fact, the increase in control room personnel capabilities, instrumentation, and control functions provides the technical capabilities in operational personnel and significantly reduces the need for such a position.
We request the NRC review our Duty and Call Technical Advisor proposed alternative to a shift position.
C.
Schedule Implementation of the D'uty and Call Technical Advisor position remains as stated in the October 20, 1979 submittal.
Personnel will be assigned by January 1, 1980 and meet the short-term qualifications with f-
- qualification to be achieved by January 1, 1981.
1431
$54
= - -. - -.- -.. - -.. - -
2.2.L.c Shift and Relief Turnover Procedures I
A.
Position The shift and relief turnover procedure revisions remain as stated in our submittal of October 20, 1979.
C.
Schedule This item will be completed by January 1,1980.
e 1431 355
2.2.,2.a Control Room Access A.
Position We currently limit and will continue to limit control room access at Point Beach Nuclear Plant.
The recognition by NRC of the desirability of such limi-tation is commendable.
B.
Discussion Current policy, as described in our letter of October 20, 1979, limits NRC access to the control room to one individual.
It is our understanding that the NRC Staff does not agree with this access control but has not specified the number of NRC personnel for which access is required.
As soon as NRC advises us of the number of NRC personnel for whon NRC anticipates access, we will revise our policy accordingly.
C.
Schedule 1.
Policy restric' ting access to the control room has been formalized by special memorandum of October 8, 1979, as described in our letter of October 20, 1979.
2, Revisions to this policy will be issued immediately upon receipt of MRC instructions regarding number of NRC personnel to be granted access.
1431 356
2.2.2.b On-Site Technical Support Center A.
Position We will establish an interim and a permanent Technical Support Center.
B.
Discussion Designation of the Point Beach control room as a backup to the interim Technical Support Center may not be appropriate, particularly in view of its size and configuration.
We are planning to designate our reactor engineering office and computer room above the control room as the interim alternate.
C.
Schedule 1.
Establishment of a permanent Technical Support Center will require construction.
Such construction may require prior Public Service Commission of Wisconsin authorization and certain building and other permits.
We are proceeding as rapidly as possible on this item but we believe that a comple-tion date of January 1, 1981, may be unrealistic.
2.
Also, we are consolidating a number of separate efforts to achieve a coordinated and well balanced enhancement of our facilities.
These efforts include the provision of the Technical Support Center, the offsite Emergency Operations Center, additional space for our training program, and permanent office space for resident NRC inspectors.
At the same time, we are attempting to prcvide a coordinated approach to all the "short-term lessons learned" instrumentation implementation.
These plans are in the conceptual stage at present.
3.
Conceptual planning for a permanent Technical Support Center has already begun.
It is anticipated that detailed engineering will begin by January 1, 1980, and regulatory and construction requirement will help define completion schedules shortly thereafter.
i431 557 2.2.2.b-1
2.2.2.c' Operational Support Center A.
Position An interim and a permanent Operational Support Center will be designated.
B.
Discussion Refer to October 20, 1979 letter for further details.
~
C.
Schedule Permanent designati~on of an Operational Support Center is related to other emergency centers.
See response to 2.2.2.b.
d 1431 358 s
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