ML19256G556
| ML19256G556 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/27/1979 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19256G557 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7912310458 | |
| Download: ML19256G556 (64) | |
Text
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General Offices: 212 West Machigan Avenue, Jackson, Michigan 49201
- Area Code S17 788-0550 s
December 27, 1979 Director, Nuclear Teactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission W::shington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - REQUIREMENTS RESULTING FROM REVIEW OF THREE MILE ISLAND 2 ACCIDENT:
ACTIONS TAKEN IN RESPONSE TO Consumers Power Company letter dated October 30, 1979 submitted implementation criteria and schedules for actioas to be taken at Big Rock Point in response to requirements resulting from NRC review of the subject accident. The commitments provided were subsequently discussed between Consumers Power Company and NRC personnel and were revised by Consumers Power Company letter dated December 17, 1979. The December 17, 1979 letter stated that our October 30, 1979 letter would be revised to reflect in more detail the cevisions co previously established commitments.
Forwarded as an attachment to this letter is a tabulation of the actions to be taken at Big Rock Point to address the subject requirements.
The requirements addressed in the attachment are those referenced or provided by NRC letter dated September 13, 1979.
The attached tabulation includes a detailed description of the specific actions takan to complete those actions which required completion by January 1, 1980.
Thir letter revises and supersedes our letter of October 30, 1979 on the same subject.
David P Hoffman (Signed';
David P Hoffman Nuclear Licensing Administrator 1661 001 CC JGKeppler, UShTC 7912310 g
1 IMPLEFENTATION CRITERIA FOR POST-TMI REQUIRE?ENTS AT BIG ROCK POINT 1.
NUREG 0578 Reauirement 2.1.1:
" Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves, and Pressurizer Level Indicators in PWRs."
Provide redundant emergency power for the minimum number of pressurizer heaters required to maintain natural circulation conditions in the event of loss of offsite power. Also provide emergency power to the control and motive power systr s for the power-operated relief valves and associated block valves and to the pressurizer level indication instrument channels.
Action To Be Taken at Big Rock Point The requirements related to pressurizer heaters and level indicators are not app'.icable to Big Rock Point. Big Rock Point does not have power-operated relief valves for primary system pressure control.
The plant is equipped with pilot-operated relief valves in the reactor depressurization system. The pilot valves are electrically operated. Each pilot-operated relief valve is in series with a normally closed, air-operated gat..
alve. The control power and actuation logic are designed such that a single failure cannot cause the system to fail to operate nor will any single failure cause the system to inadvertently operate. Each set of reactor depressurization system valves is powered from its own dedicated, uninterruptible power supply.
1661 002
2 Both the pilot-operated valve and the gate valve in each pair fail closed on loss of the uninterruptible power supply in its respective channel. Three out of the four RDS paths are adequate to perform the design depressurization function.
The air-operated gate valves are supplied from the plant air system through double check valves. There are low air pressure alarms on each RDS channel.
The gate valves, however, fail open on loss of air.
The gate valves will remain closed, by action of the check valves, for a sufficient time to manually attach a bottled air supply during a loss of air event. Even if this manual action is not taken, the series pilot-operated valve, which is unaffected by loss of air, will still block reactor blowdown.
If the RDS channel is actuated during a loss of instrument air event, the gate valves could not be reclosed. Under this condition, the electrically actuated, pilot-operated relief valve will still close to stop blowdown.
There would be no backup valve, however, should it also fail.
1661 003
3 2.
NUREG 0578 Requirement 2.1.2:
" Performance Testing for BWR and PWR Relief and Safety Valves."
Commit to provide performance verification by full scale prototypical tcsting for all relief and safety valves. Test conditions shall include two phase slug flow and subcooled liquid flow calculated to occur for design-basis transients and accidents.
Action To Be Taken at Big Rock Point Analyses of Big Rock Point transient response have consistently indicated that the safety valves would not be required to pass two phase flow for any design basis transient or accident.
In fact, operation of even a single loop of the emergency condenser serves to preclude opening of a safety valve for essentially all analyzed events. Safety valve cycling is, therefore, not a routine occurrence at Big Rock Point.
Consumers Power Company is a participating member of the BWR Owners Group.
The Group has recently established a program for testing BWR valves in response to this requirement. This program was discussed in a letter from the Owners Group chairman to D G Eisenhut (NRC) dated December 14, 1979.
- Briefly, this program involves testing those conditions which could result in single-phase liquid or two phase flow through the valves at low pressure (eg, as is experienced in an alternate shutdown mcde at some BWRs).
Installation of safety grade high-level trips for high pressure emergency core cooling systems (ECCS) and reactor feedwater systems is being considered as a means of addressing concerns about two-phase flow at high pressure. The Owners Group 1661 004
4 intends to complete its evaluation of such high-level trips by January 31, 1980.
Consumers Power Company is participatir.3 in the Owners Group Program.
In this respect, it should be noted that the s.1 ternate shutdown mode which can result in low pressure liquid flow through relief valves is not used at Big Rock Point.
Consumers Power Company has initiated preliminary design of a high steam drum level trip for the Big Rock Point feedwater system for use should this option be selected as the Owners Group means to address high pressure concerns. Big Rock Point has no high pressure ECCS.
If the Owners Group decides to recommend installation of such a trip rather than high pressure testing, insta.'lation will be complete by January 1, 1981; a design description will be submitted for NRC information in the event this decision is made.
In any event, Consumers Power Company intends to conform to BWR Owners Group recommendations in this area.
1661 005
5 3.
NUREG 0578 Requirement 2.1.3.a:
" Direct Indication of Power-Operated Relief Valve and Safety Valve Position for Pkm's and BWR's."
Provide in the control room either a reliable, direct position indication for the valves or a reliable flow indication device downstream of the valves.
Action To Be Taken at Big Rock Point Indication will be provided for each of the six safety valves. Acoustic sensors will be used for this purpose.
A single sensor will be associated with each valve. The installation will meet the same requirements as other engineered safety features (except redundancy). Equipment which can be made available to support rapid installation has not all been fully certified to post-accident environmental conditions. This certification will be completed by January 1, 1981.
Appendix A to this enclosure includes, for NRC information, a detailed description of the safety valve position indicator system to be installed.
This appendix includes the following documents prepared by the design contractor, Energy Incorporated: proj ect requirements, construction specifications for the portian of the system inside containment, construction specifications for the portion outside containment, and the test specification for the system.
Big Rock Point will be removed from service prior to January 1, 1980 to install the position indicator system described in Appendix A.
The system installation will be complete and the system will be operable prior to plant start-up.
1 6,6 1 0 0 6,
6 Backup methods for determining that a safety relief valve is open include a common drain header high temperature alarm, contaiment high pressure alarm, and indication and dewcell temperature recorder for high humidity and pipeway high temperature. These indirect methods of recognizing an open valve are discussed in Plant Operating Procedures.
Reactor depressurization system valves are presently equipped with direct position indication in the control room.
1661 007
7 4.
NUREG 0578 Requirement 2.1.3.b:
" Instrumentation for Detection of Inadequate Core Cooling in PWR's and BWR's."
Perform analyses and implement procedures and training for prompt recognition of low reactor coolant level and inadequate core cooling using existing reactor instrumentation (flow, temperature, power, etc) or short-term modifications of existing instruments. Describe further measures and provide supporting analyses that will yield more direct indication of low reactor coolant level and inadequate core cooling such as reactor vessel water level instrumentation.
Action To Be Taken at Big Rock Point Consumers Power Comp iny is participating in the BWR Owners Group effort in this area. Analyses performed by General Electric Company for the Owners Group have indicated that adequate core tooling can be assured by verifying either a vessel water level above the core or rated flow in one of the low-pressure core spray systems. Big Rock Point is equipped with a reactor vessel level instrument and flow indication in each of the two core spray lines.
Procedural guidelines addressing the use of these instruments are under development by the Owners Group and will be implemented on the schedule agreed between the Owners Group and the NRC.
NRC letter dated July 17, 1979 (Enclosure 1, Attachment 3) requested information concerning the effects of post-accident conditions on reactor vessel level indication. This issue was addressed in General Electric report NED0-24708 submitted on behalf of the BWR Owners Group on August 17, 1979.
This area was also discussed
- .n a General Electric Company S r e In ion
8 Letter. Specific evaluation of this issue for Big Rock Point identified a concern with regard to the ability of the temperature-compensated level instruments to reliably perform their safety functions following a postulated accident; this concern was reported by Licensee Event Report (LER) 79-22 submitted September 5, 1979. The existing level instrumentation was modified during the recently completed outage to eliminate temperature compensation and alleviate the concerns discussed in LER 79-22.
Consumers Power Company letters dated October 23 and October 31, 1979 discussed the modifications in support of Technical Specifications changes which they necessitated; the revised Technical Specifications were issued as Amendment No 31 to License DPR-6 on November 2, 1979.
Due to the differences between reactor vessel / steam drum level instrumentation available at Big Rock Potr; and that in newer BWRs, Consumers Power Company is evaluating alternate level instrumentation designs. Energy Incorporated (EI) is currently engaged in evaluating varying designs for new level instruments.
A primary objective of new instrumentation is to provide accurate indication and trip signals during normal operation, anticipated transient and accident conditions, and post-accident conditions. To achieve this objective, the naw system design will be selected to further reduce the effects of post-accident containment heating on the reliability of indication. As a minimum, the new system will have a range equal to that of the present instrumentation.
In addition to replacing the existing instrumentation with a design of improved reliability, consideration is being given to installing a wide range level indication. This wide range instrument would provide additional information for the operator but would not provide trip signals for safety kh
9 systems (such would be provided by the replacements for the current instrumentation discussed above). Accordingly, the wide range instruments vould be of control grade quality.
EI will complete conceptual design of a new level instrumentation system incorporating the features discussed above in January 1980.
Several alternate designs are still under consideration at this time.
A detailed design description of the instrumentation selected will be submitted for NRC information by July 1, 1980. The new instrumentation will be installed by January 1, 1981.
1661 010
10 5.
NUREG 0578 Requirement 2.1.4:
" Containment Isolation Provisions for PWR's and BWR's."
Provide containment isolation on diverse signals in conformance with Section 6.2.4 of the Standard Review Plan, review isolation provisions for non-essential systems and revise as necessary, and modify containment isolation designs as necessary to eliminate the potential for inadvertent reopening upon reset of the isolation signal.
Action To Be Taken -t Big Rock Point Diversity currently exists in parameters sensed for containment isolation.
Either low reactor vessel water level or high containment pressure results in an isolation. Normally open nonessential lines which carry fluids outside containment are closed automatically on an isolation signal. Normally open lines carrying fluid into containment are equipped with check valves and can also be secured by manually operated gate valves or air-operated control valves.
Systems considered essential are as follows:
Post-Incident and Fire Water Supply System Post-Incident Backup System Ventilating Vacuum Breaker Sensing Line Core Spray Recirculation System Classification of systems as essential / nonessential will be reviewed as part of SEP.
Consumers Power Company letter dated May 4, 1979 provided our response to IE Bulletin 79-08.
This response stated, in part, that certain v 1 e lo ay
11 a containment isolation signal could automatically reopen if the actuating signal cleared; interim administrative actions to prevent this were described, and a commitment was made to modify the isolation logic to preclude such automatic reopening. The problem of inadvertent reopening of the isolation valves was found in four (4) control circuits which control nine (9) isolation valves. The automatic reopening of *he isolation valves will be eliminated through the addition of a seal-in relay and a push-button switch into each of the four (4) control circuits.
Both the seal-in relays (General Electric Type HMA relay) and the push-button switches (General Electric industrial miniatur' oil-tight push button) are of a quality equal to or greater than similar equipment presently in use in Big Rock Point safety systems. The seal-in relay will be installed in each isolation valve control circuit before the isolation valve hand switch and in parallel with the circuit. One normally open contact of the seal-in relay will be placed in series with the circuit and before the seal-in relay coil to act as a holding contact. This arrangement is shown on diagrammatically in Appendix B.
A push-button switch, with momentary contacts, will be installed so as to bypass the holding contacts. This will provide a means of energizing the seal-in relay coil and closing the holding contact which will, in turn, complete the circuit, provided that the reactor protection systems containment isolation signal is clear and the isolation valves hand switch is in the closed position.
If loss of a-c power should occur or reactor protection systems containment isolation signal should actuate, the seal-in relays coil will become de-energized, dropping out the contact.
(See Appendix B.)
Re-energizing of the circuit can only be accomplished through a direct operator action via the push-button switch, and only af ter the de-energizing signal is corrected or cl. red.
In 1661 012
12 conjunction with the seal-in relay and the push-button switch, an indicating lamp will be added in parallel aith each isolation valve control circuit to indicate when the holding contact is made up.
The system described above will be installed during the forthcoming outage discussed in response to Requirement 2.1.3.a above.
Installation will be completed before the first reactor start-up of 1980.
1661 013
13 6.
NUREG 0578 Requirement 2.1.5.a:
" Dedicated Penetrations for External Recom'siners or Post-Accident Purge Systems."
Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment isolation systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.
Action To Be Taken at Big Rock Point Big Rock Point does not use external recombiners or purging for combustible gas control. Accordingly, no action is needed.
166i U14
14 7.
NUREG 0578 Requirement 2.1.5.b:
"Inerting BbR Containments."
It shall be required that the Vermont Yankee and Hatch 2 Mark I Bh1 containments be inerted in a manner similar to other operating BWR plants.
Inerting shall also be required for near term OL licensing of Mark I and Mark II BkRs.
Action To Be Taken at Big Rock Point This requirement is not applicable.
1661 015
15 8.
NUREG 0578 Requirement 2.1.5.c:
" Capability To Install Hy(
r Recombiner at Each Light Water Nuclear Power Plant."
A minority of the Task Force recommended that all light water reacte u.nts have the capability to install recombiners within a few days following an accident. Procedures to accomplish such installation were to be developed.
Action To Be Taken at Big Rock Point Big Rock Point's licensing basis does not iuclude use of hydrogen recombiners.
No action will be taken.
Consumers Power Company letter dated May 4, 1979, in response to IE Bulletin 79-08, reported the results of an analysis of the total amount of hydrogen which could be generated by a metal-water reaction involving all the cladding in the fueled region of the core. The maximum possible hydrogen concentration was incorrectly reported as 5.7 v/o; the correct value is 6.7 v/o. Despite this error, Consumers Power Company continues to consider that the hydrogen concentration which might be present inside containment is sufficiently small to preclude the need for hydrogen recombiners.
Conservative assumptions used in this analysis included the assumption that the reactor coolant system contained the maximum possible dissolved hydrogen (ie, was saturated) at the start of the accident.
1661 016
16 9.
NUREG 057d Requirement 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 program shall be implamented to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious accident. The program shall include (1) implementing all practical leak reduction methods, (2) measurement and report to NRC of actual leak rates and (3) preventive maintenance including at least integrated leak tests each refueling outage.
Action To Be Taken at Big Rock Point The core spray system would contain radioactivity af ter an accident. The piping and valves in this system which are outside containment will be observed for leakage when testing the core spray pumps.
In addition, core spray heat exchanger tube leakage will be calculated to ensure no tube degradation in excess of Technical Specifications limits.
The core spray pump operability test procedure has been modified to require monitoring for leakage and issuance of a maintenance order for repair if any leakage is observed.
This test is performed during each refueling outage. This test was performed October 18, 1979 with no observable leakage.
The radwaste system could contain radioactivity after an accident. However it would be unlikely that high levels of radioactivity would be found since double valve isolation is provided for all sources of such contamination. The primary concern for off-site exposures from radioactivity in this system would be release of radioactive gases from degasification; leakage testing would not 166,1 017
17 contribute to reduction of off-site doses from such degasification since all tanks in the system are open to the atmosphere. Thus, the radwaste system will not be tested for leakage in response to this requirement.
Big Rock Point design is such that other systems which might contain radioactivity after an accident are generally inside containment.
Systems penetrating containment and the tests performed on each penetration are tabulated in Appendix F.
1661 018
18 10.
NUREG 0578 Requirement 2.1.6.b:
" Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations."
Each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design rev4ew should identify the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded ly the radiation fields during post-accident operation of these systems.
Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facilitr.
Action To Be Taken at Big Rock Point Consumers Power Company has completed a design review of Big Rock Point shielding. This review has identified modifications which must be made in the short term to permit necessary post-accident actions to be performed while limiting personnel radiation exposures to 25 rem assuming existance of the radiation source term specified in NUREG 0578.
Consumers Power Company has also engaged a contractor, Catalytic Inc, who has completed a conceptual evaluation of options for providing the shielding necessary to limit personnel exposures to 10 CFR 20 limits following an 1661 019
19 accident with the specified radioactivity release. As anticipated, the modifications necessary to provide such shielding are significant; viable options would necessiate erecting a concrete silo around the containment sphere and/or adding concrete shielding up to several feet thick to existing structures. The addition of this amount of shielding would severely impact ongoing evaluations which are part of the Systematic Evaluation Program (SEP).
The installation of shielding required to fully address the concerns discussed in NUREG 0578 will, therefore, be incorporated into the SEP and will be accomplished following completion of this NRC program.
A detailed discussion of the design review and the need to incorporate this modification into the SEP is provided as Appendix C to this enclosure.
1661 020
20 11.
NUREG 0578 Requirement 2.1.7.a:
" Automatic Initiation of the Auxiliary Feedwater System for PWRs."
This requirement is not applicable to Big Rock Point.
12.
NUREG 0578 Requirement 2.1.7.b:
" Auxiliary Feed nter Flow Indication to Steam Generators for PWRs."
This requirement is not applicable to Big Rock Point.
1661 021
21 13.
NUREG 0578 Requirement 2.1.8.a:
" Improved Post-Accident Sampling Capability."
A design and operational review shall be performed to determine the capability of obtaining reactor coolant and containment atmosphere samples under accident conditions while limiting personnel radiation exposures to 10 CFR 20 limits.
Additional design features or shielding should be provided to meet these criteria if necessary.
A design and operational review shall be performed to determine the capability to promptly quantify certain radioisotopes indicative of core damage. Design modifications or equipment changes shall be made to facilitate this.
Capability to perform boron and chloride chemical analyses under accident conditions shall be provided.
Action To Be Taken at Big Rock Point Capability of obtaining reactor coolant and containment air samples without entering containment does not now exist. Thus, it would be impossible to obtain such samples under the accident conditions postulated. Provision of post-accident sampling capability is related to the shielding modifications under consideration in response to Requirement 2.1.6.b in that no adequately shielded location (other than the Operstional Support Center discussed in response tc Requirement 2.2.2.c below) exists in which to locate a sampling system. Accordingly, it is intended to delay installation of such a system so that it can be accomplished in conjunction with the shielding modifications of Requirement 2.1.6.b (after the SEP program).
Interim methods of quantifying 1661 022
22 release of radioactivity as a result of core damage have been developed for use until sampling capability is available and are described below.
1.
Interim Procedure for Sampling Core Spray Recirculation (Accidents With Limited Core Damage)
In the event of an accident which does not release a large portion of the core inventory of radioactivity, it would be pcssible to obtain samples of core spray recirculation water from the core spray heat exchanger.
Calculated radiation levels for various postulated accidents indicate that such samples could probably be obtained if 10% core damage occurred.
Radiation levels in the area of the core spray heat exchanger following a radioactivity release from the core of the magnitude specified in NUREG 0578 wu. id preclude entry to the area and would also make on-site analysis equipment unusable until the shielding modifications in response to 2.1.6.b above are completed.
A procedure has been writteu to obtain samples of core spray recirculation water. The procedure censists of a radiological survey checklist to assess radiological conditions in the analysis facilties, the walkway to the sample location, the sampling and dilution, and during analysis.
Consideration is given to the impact of handling highly radioactive san les.
The procedure also provides for involvement of different f
personnel in different aspects of the sampling and analysis to minimize individual exposures, and addresses other radiological concerns. The procedure requires final review of the Plant Health Physicist or his 1661 023
23 designee including a briefing of the personnel who are to perform the procedure before the activity commences.
Samples obtained using the procedure described above would be analyzed with existing Ge(Li) detector and multichannel analyzer.
The minimum training of personnel to be accomplished before start-up, following the upcoming outage discussed in response to Requirement 2.1.3.a above, will consist of a review of the procedure and a dry run of the actual sampling operation.
2.
Interim Quantification of Core Damage (Accidents With More Extensive Core Damage)
Quantification of core damage for accidents which preclude the sampling procedure of No 1 above will be accomplished using a shielded, collimated ionization chamber or GM probe. The detector will be located in the cable penetration room near the containment sphere and will have remote readout in the Operations Support Center.
(See 2.2.2.c below.) The monitor will have a range of 10 R/h to 10 R/h; the direct radiation from the containment at the monitor's location is estimated at 9 x 10 R/h for a 100% core melt accident (assumed instantaneous release at time of accident) and 2 x 10 R/h for release of gap activity. A graph of radiation !cvel versus time for varying degrees of core damage (based on calculated radiation levels from noble gases inside containment) will be provided for use in interpreting readings of this monitor.
1661 024
74 The monitor will be in place by January 1, 1980 and procedures for its use as discussed above will be completed before start-up from the upcoming outage discussed in response to Requirement 2.1.3.a above.
It should be noted that this method would also provide backup information for accidents wherein lesser core damage occurs and the sampling procedure of No 1 above can be used.
3.
Long-Term Solution Consumers Power Company's preliminary evaluation of possible methods of providing the required capability in the long term centered
.nd techniques for in-line sample monitoring. This would eliminate the problems attendant upon handling of samples containing high levels of radioactivity which would be involved in laboratory analysis.
Dissolved boron is not used for reactivity control at Big Rock Point except in the liquid poison system; thus, sampling capability to analyze boron is not needed. Lake Michigan coolant used for heat rejection in the main condenser is extremely low in chlorides; thus, chloride analysis capability would not enhance post-accident diagnosis. The small size of Big Rock Point's core with respect to the large containment free volute results in essentially no concern due to hydrogen generation.
(See 2.1.5.c above.) Thus, hydrogen analysis capability is not needed.
Quantification of radionuclide content can be performed using an in-line monitor.
It is, therefore, concluded that capability to obtain coolant and containment atmosphere t.imples for laboratory analysis is not needed.
1661 025
25 The analysis system which appears, at this time, to best meet the needs of Big Rock Point involves a Ge(Li) (or intrinsic germanium) detector installed in a shielded assembly and monitoring radioactivity in a small diameter line passed before it.
Separate lines for reactor coolant and containment atmosphere sampling could be provided.
Both sample lines would be routed back to containment to preclude release of effluents as a result of sampling operations. Detailed design of such a system, or the location in which it can be installed, would be affected by resolution of the plant shielding concerns discussed in 2.1.6.b above. Thus, installation of this or other appropriate sampling capability will be accomplished in conjunction with the plant shielding modifications (at the end of SEP).
1661 026
26 14.
NUREG 0578 Requirement 2.1.8 d:
" Increased Range of Radiation Monitors."
Provide high range radiation monitors for noble gases in plant effluent lines and redundant high-range radiation monitors in the containment.
Provide capability of measuring and identifying radioiodine and particulate radioactive effluents under accident conditions.
Action To Be Taken t.t Big Rock Point The following will he provided:
Radioactive noble gas effluent monitors from ALARA ranges to 103,:Ci/cc.
Capability of radioiodine sampling of effluents followed by on-site analysis.
In containment radiation level monitors (minimum of 2) capable of 6
measuring radiation levels to a maximum of 10 rad /h.
The maximum ranges of the instrumentation to be provided were determined based on an analysis of the highest possible radionuclide release at Big Rock Point.
Big Rock Point's containment is large (approximately 2.66 x 10 10 3
cm ) while core size is considerably smaller than in newer plants. These factors result in maximum post-accident radiation levels and noble gas concentrations (assuming Regulatory Guide 1.3 release fractions) which can be monitored by instraments of the specified ranges.
1661 027
27 Instrumentation installed will meet the 11owing criteria:
1.
Two physically separate and redundant containment radiation level monitcrs will be provided. Radioactive noble gas effluent monitors will not be redundant.
2.
Seismic qualificacion per Regulatory Guide 1.97.
3.
Noninterruptible power supply.
4.
Conform to ANSI N320-1978 as applicable to fixed instrumentation.
5.
Conform to Quality Assurance requirements of ANSI N45.2-1971 and Consumers Power Company Quality Assurance Program as appropriate.
6.
Continuous display and recording.
7.
Testing per Regulatory Guide 1.118 except that electronic calibration will be used for the upper end of the containment radiation level monitors to preclude the high personnel radiation exposures which would be attendant upon use of a radioactive source of the strength necessary to accomplish direct calibration.
Interim methods of obtaining the required information and additional information regarding the long-range instrumentation to be provided is as fcllows:
1.
Interim Methods of Quantifying Gaseous Effluents Current stack monitor capabilities are limited to a release rate of approximately 40 Ci/s which represents three times the Regulatory Guide 1.3 1661 028
28 maximum hypothetical accident (WLA) source term with leakage at a rate of 0.5%
of gaseous inventory per day.
The current monitor samples a stream which would contain all leakage from turbine building sources, ir.cluding condenser off gas, a.
System and Method (1) A monitor with readout which is remote from the detection chamber is being dedicated to emergency use in quantifying high level releases. Dynamic range requirements of approximately 35 mR/h to 20 R/h have been determined for chosen sensor Ic:ation in order to quantify release rates from the level of current instruments and procedures to the level of 10,000 Ci/s. Sensitivity of each sensor to Xe-133 (81 kev) will be accounted for in the tables for dose rate-to-release rate conversion. The tables will provide conversion values as a function of time after shutdown such that the larger percentage contribution from Xe-133 at later times will be acknowledged. The installed instrument, these tables, and graphical displays of tabulated data, will be available for use with plant procedures prior to start-up from the upcoming outage discussed in 2.1.3.a above.
(2) Monitor location for interim high level stack release monitoring is adjacent to current stack gas sample and return lines. A lead shield for background reduction is provided. The monitor location has been chosen at a level of approximately 13 feet below grade in order tu eliminate direct shine from the unshielded containment structure. Measurements will be adjusted 1661 029
29 as necessary for indicated flow rate changes, in addition to the time-dependent conversion factors discussed in (1) above, in order to ensure measurements which are representative of stack emission rates.
No on-site monitoring technique appears feasible for determining activity release rate through the steam tunnel blowout panel or emergency condenser exhaust, should activity be released in that manner. These release paths are highly unlikely, however; a major failure concurrent with the LOCA would be required to permit any release via these paths. Off-site sampling will be used for these unlikely sources of activity as well as particulates and iodines from the plant stack. Description of the off-site techniques is given in Item 2 below.
(3) Radiation readings will be displayed in the Operations Support Center.
(See 2.2.2.c below.)
(4) Radiation level will be displayed continuously with range selection capability at the readout station.
(5) A special battery pack has been designed for this instrument to ensure minimum time of continuous operation of at least seven days. The battery pack will be available before start-up af ter the upcoming outage discussed in 2.1.3.a above.
1661 030
30 b.
Description of Procedures Procedures are now in draf t format and awaiting entry of the tabulated response conversion factors. All necessary procedures and aids described below will be complete by January 1, 1980.
The procedures, as now drafted, or currently available approved plant procedures, include specific instructions in each of the following areas:
(1) Minimization of personnel exposure.
(2) Calculational methods for determining release rates.
(3) Reporting of results (Operations Support Center communications).
(4)
Instrument calibration (current plant procedures).
2.
Interim Methods for Quantifying Radioiodine and Particulate Effluents Shielding design review performed in response to Requirement 2.1.6.b above has shown that the current stack monitor location would be inaccessible for recovery of particulate and iodine samples during the first five days after a Regulatory saide 1.3 Maximum Hypothetical Accident. Furthermore, a release rate of 10,000 Ci/s noble gas, with Regulatory Guide 1.3 nobte gas-to-iodine partitioning of 4 to 1, implies 2,500 Ci/s iodine, or 225 Ci of iodine plus 360 Ci of noble gas in the filter media af ter 15 minutes of sampling.
Provision of shie ding for stack access alone would not solve l
the problem of removing a stack sampler filter with activity this high in the sample media. Due to these personnel exposure concerns, emergency i661 031
31 procedures are being modified to specifically prohibit the removal of sample media under this type of accident condition.
Isopleth plots of dispersion coefficients for Pasquill meteorological Categories A through F, respectively, have been developed for both elevated (stack) and ground level (blowout panel) releases.
Portable instruments will be used to sample and conservatively quantify iodines adsorbed on silver zeolite cartridges. Particulates, which are less limiting in terms of radiation dose for public protective actions than iodine, will be sent to our Palisades Plant for analysis within an expected time frame of four to five hours. Silver zeolite cartridges also will be sent to Palisades for Ge(Li) spectral analysis in order to refine the field instrument measurements.
Procedures for accomplishing the above sampling will require the Pasquill stability category to be estimated using visual observation of cloud cover and average wind speed. This is based on Table 3.3 of " Meteorology and Atomic Energy, 1968." A monitoring team will be dispatched to tJa location estimated to have the highest ground level concentration. A radioiodine sample will be collected using a portable air sampler collecting a 10 to 20 cu ft air sample at a rate of approximately 1-2 cfm on a silver zeolite cartridge. The sample will be counted using a pancake probe and a rate meter. Assuming a 94% collection efficiency and a 25%
detector efficiency, the minimum concentration detectable is
-10 6 x 10 Ci/cc. This is sufficiently low to permit detection prior to reaching levels corresponding to a State of Michigan Nuclear Incident 1661 032
32 Class B.
The previously calculated X/Q can then be used to calculate an estimated release rate.
The procedures and equipment described above will be available for use before start-up from the upcoming outage.
3.
Nobel Gas, Particulate and Iodine Monitoring - Long-Term Methods Iodines and entrained noble gases would provide a source of approximately 800 curies in the charcoal adsorber for a 15-minute (3 cfm) sample following an accident with Regulatory Guide 1.3 releases at Big Rock Point. Only one commercially available sampling system (Science Applications /RadeCo) provides the automated (hands off) sampling and analysis which is required for personnel protection at these levels.
Automatic fresh air purge of the charcoal releases the noble gas so that a maximum of not more than approximately 300 curi.; of iodine would be present for automatic analysis by an intrinsic germanium spectrometer.
We are evaluating an SAI proposal for application at Big Rock Point. This or other similar monitors would require routine servicing (approximately once per week dewar fills) and must, therefore, be located in an accessible area. No location is available meeting these criteria after an accident until the shielding to be installed in response to Requirement 2.1.6.b above is available. This system or its equivalent will be installed, however, in conjunction with those shielding modifications (at the end of SEP). A detailed design description of the system to be provided will be submitted for NRC information by January 1,1981.
1661 033
33 4.
High Range Containment Monitors The lack of containre?t shielding at Big Rock allows accurate determination ot _ atation levels inside by means of external measurement.
Location of radiation monitors external to containment offers the advantage of a much less severe environment with subsequently higher reliability under accident conditions.
Consumers Power Company will install two instruments external to containment with capability of determining gamma dose rates within containment up to 1 x 106 R/h 5
(approximately 8 x 10 R/h at the monitor location).
These monitors would remain outside of containment but inside a future shield wall, if and when such shielding is constructed.
Installation of the monitors is expected to be complete by January 1, 1981. A detailed design description of the installation will be submitted for NRC review by March 16, 1980.
166,1 034
34 15.
NUREG 0578 Requirement 2.1.8.c:
" Improved In-Plant Iodine Instrumentation."
Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.
Action To Be Taken at Big Rock Point Airborne radioiodine monitoring with conventional charcoal canisters may produce overly conservative results under accident conditions due to interference from radioactive noble gases. Occupanty of various locations could be unnecessarily restricted if airborne radioactive iodine levels were incorrectly determined to be high.
1.
Short-Term Solution RadeCo Model GY-130 silver zeolite radioiodine sampling cartridges demonstrate an acceptable collection efficiency for inorganic as well as organic iodines (94-96% in the no6221 sampling flow rate range) but do not retain noble gases to any appreciable degree. Although these cartridges are too expensive ($45 each) for normal plant sampling, they are justifiable for emergency monitoring. Two hundred of these filters are being procured.
Respiratory protection is currently available and potassium iodide as a thyroid blocking agent will be available by January 15, 1980 to reduce thyroid burdens of radioactive iodine.
1661 035
35 a.
Procedural Method A 10-minute (10-20 cubic foot) air sample will be passed through a combination particulate filter and silver zeolite cartridge holder at a flow rate of I-2 cfm using the standard Big Rock Point air sample, the RadeCo Model H809V. The silver zeolite cartridge will then be counted using a standard frisker (Eberline RM-14 or Ludlum 177) equipped with a pancake GM probe. Collection efficiency of 94% and detector efficiency of 25% yield a minimum detectable activity of 6.0E-10 pCi/ml for the smallest sample size (10 cubic feet), which is a factor of 10 below the I-131 MPC.
Projected iodine levels in occupied areas in excess of 520 MPC hours (equivalent to 40 MPC hours for 13 weeks permitted by 10 CFR 20.103) will require respiratory protection to be worn or potassium iodide to be prescribed.
Evaluation of the type of protection to be used will be made on o case-by-case basis depending upon the individual action required in the airborne area.
The Company's general medical consultant has aeveloped a procedure for the use of potassium iodide in emergency situations.
Initial draft of this proceudre was submitted to the Company on December 13, 1979 and will be implemented by January 15, 1980.
b.
Equipment Impact (1) Without reuse, a quantity of 200 cartridges is adequate to permit sampling of the Control Room / Technical Support Center (TSC) and the Operations Support Center atmospheres every 15 minutes for 036
36 the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during an accident and then every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the next 10 days. With reuse of cartridges which have insignificant accumulations of radioactivity, sampling can be expected well in excess of this minimum frequency.
(2) Portable battery-operated air samplers and survey instruments are available for use during any period of power loss.
(3) Eight standard size (2" x 4" x 8" or 2" x 4" x 6") lead bricks will be stored in both the TSC and in the Operations Support Center such that they are available for the construction of counting shields for pancake GM probes in the event that background radiation levels preclude air sample cartridge counting without shielding. These lead bricks are available on sita.
c.
Training The sample procedure is similar to existing surveillance air sample procedures and utilizes existing air sampling equipment and counting instruments. Therefore, the minimum training for health physics personnel will consist of a required review of the procedure prior to plant start-up from the upcoming outage. A group discussior,on each shift, led by the responsible assistant s ervisor, will also be performed prior to plant start-up.
1661 037
37 2.
Longer Term Actions Orders are being placed for s eral additional a-c/ battery-operated friskers (Ludlum Model 177) to supplement the - si s t...
supply and ensure that adequate numbers of 'riskers wil' ou available for contamination monitoring and air sample counting dur an emergency.
o Five battery-operated ratemeters (Eberline PRM-6) will be purchased to permit long-term (500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> per battery charge) monitoring and air sample counting in the field during emergencies.
1661 038
38 16.
NUREG 0578 Requirement 2.1.9:
" Analysis of Design and Off-Normal Transients and Accidents."
Provide the analysis, emergency procedures, and training to a.
substantially improve operator performance during a small break loss-of-coolant accident.
b.
Provide the analysis, emergency procedures, and training needed to a..are that the reactor operator can recognize and respond to conditions of inadequate core cooling.
Provide the analysis, emergency procedures, and training to c.
substantially improve operator performance during transients and accidents, including events that are caused or worsened by inappropriate operator actions.
Action To Be Taken at Big Rock Point These analyses are being performed and emergency procedure guidelines are being developed as part of BWR Owners Group efforts.
These efforts are being conducted in accordance with schedules established in conjunction wi.h NRC Bulletins and Orders Task Force. Consumers Power Company is participating in this effort.
1661 039
39 17.
NUREG 0578 Requirement 2.2.1.a:
" Shift Supervisor's Responsibilities."
Review plant administrative and management procedures. Revise as necessary to assure that reactor operations command and control responsibilities and authority are properly defined.
Corporate management shall revise and promptly issue an operations policy directive that emphasizes the duties, responsibilities, authority and lines of command of the control room operators, the shift technical advisor, and the person responsible for reactor operations command in the control room (i.e., the senior reactor operator).
Action To Be Taken at Big Rock Point A management directive has been issued by the Vice President for Nuclear Operations. This directive clearly states that the on-duty shift supervisor has primary management responsibility for the safe operation of the plant; he is the only person authorized to direct licensed activities or licensed operators. The directive requires that the shift supervisor remains in the control room at all times during accident situations until properly relieved; the shift supervisor is permitted to be temporarily absent from the control room during normal operations in which case he must inform the #1 control room operator to assume command of the control room and he must announce this to all control room personnel.
Plant administrative pro edures include the requirements of the management directive discussed above. These aspects of shift supervisor responsibilities will be emphasized as part of annual certification training for shif t supervisors.
1661 040
40 Shift supervisor's duties have been reviewed. Duties which detract from or are subordinate to the primary responsibility for safe operation of the plant will be assumed by another individual. This will be accomplished prior to start-up from the upcoming outage discussed in 2.1.3.a above.
Shift supervisor duties will be reviewed on an annual basis.
1661 041
41 18.
NUREG 0578 Requirement 2.2.1.b"
" Shift Technical Aavisor."
Each licensee shall provide an on-shift technical advisor to the shift supe rviso r.
The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shif t technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.
Action To Be Taken at Big Rock Point Shift technical advisors (STA) having bachelor's degrees or equivalent and with specific training in plant response to off-normal events and in plant accident analysis will be on shif t by January 1, 1980. These shift technical advisors will perform both the technical assistance and operational assessment functions specified in NUREG 0578.
The STAS who will assume duties January 1, 1980 have received seven weeks of training in plant design and layout, plant administrative and emergency procedures, response and analysis of the plant to transients and accidents and other topics necessary for successful commencement of watch standing duties.
They will receive one week of simulator training in February during which period they will be replaced on shif t by plant engineering staff members.
1661 042
42 The operational assessment function to be performed by the STAS will consist of:
1.
Providing technical support to the shift supervisor in the form of engineering evaluations of plant conditions required for all maintenance and testing on safety systems. This includes return of safety systems to service.
2.
Reviewing valve lineups munthly and after all maintenance on safety systems.
3.
Reviewing all Technical Specifications surveillance tests af ter completion.
4.
Providing an engineering evaluation of the adequacy of 91 ant emergency and operating procedures.
5.
Providing engineering evaluation of the operating history of the plant (equipment failures, design problems, operations errors, etc) and Licensee Event Reports from other plants of similar design, with suitable dissemination of the results of such evaluations to other members of the plant staff.
6.
Providing engineering evaluation of the adequacy of Company policy for maintenance, testing, equipment procurement, etc.
7.
Providing engineering evaluation of continuing adequacy of plant operations and Quality Assurance.
1661 043
43 The shift technical advisors will maintain a log in which complete entries on Items 1-3 above will be made as the actions occur. Additionally, the STAS will perform formal watch turnovers and will use watch relief sheets which include the following information (minimum):
1.
Plant conditions at time of relief.
2.
Maintenance and testing in progress.
3.
Maintenance and testing planned.
The S'fAs will report to an individual in a management position on the plant staff. This supervisor will present monthly updates to the Plant Review Committee on STA activities. The STA supervisor will also submit directly to the Vice President for Nuclear Operations a monthly memorandum on STA acitivites including a listing of the items accomplished under the operational assessment function.
I66'1 044
44 19.
NUREG 0578 Requirement 2.2.1.c:
"Shif t and Relief Turnover Procedures."
Review and revise plant procedures as necessary to assure that a shift turnover checklist is provided and required to be completed and signed by the oncoming and offgoing individuals responsible for command operations in the control room.
This checklist shall include assurance that critical plant parameters are within limits, that all systems essential to accident / transient mitigation are available and properly aligned, and identification of systems and components in a degraded mode of operation permitted by Technical Specifications (including a specific comparison of the time in such mode and Technical Specifications requirements). Supplementary checklists and shift logs should be developed for auxiliary operators and technicians.
Action To Be Taken at Big Rock Point The individual items of clarification for this requirement p avided by NRC letter dated October 30, 1979 will be addressed as follows:
Item 1 "A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a minimum, shall be included in the checklist.
"a.
Assurance that critical plant parameters are within allos,able limits (parameters and allowable limits shall be listed on the checklist)."
1661 045
45 Action To Be Taken Plant parameters are currently recorded on the control room yellow log sheet.
The limits for the critical plant parameters are listed in the margin of the yellow log sheets. These limits will be included on a check sheet where the offgoing/ oncoming shift supervisors and control operators will document completion of their review.
"b.
Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console.
"(What to check and criteria for acceptable status shall be included on the checklist.)"
Action To Be Taken Switching and tagging orders, the section tag logbook, the status board and the control console are checked to determine which equipment is either not available or operable. The equipment / systems that are not available or operable are listed in the shif t supervisor's log and control room log and verified by both oncoming /offgoing shift supervisors and control operators at the time of shift turnover.
"c.
Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications.
For such systems and components, the length of time in the degraded mode shall be compared with the Technical 1661 046
46 Specifications action statement.
(This shall be recorded as a separate entry on the checklist.)"
Action To Be Taken Where a degraded mode of operation is permitted by the Technical Specifications, the equipment / system in the degraded mode and the Technical Specifications action statement time criteria is recorded and signed by both oncoming /offgoing shift supervisors and control operators in the shift supervisor's logbook and control room logbook at the time of shift turnover.
This action identifies to the shift supervisor and control operator the amount of time the equipment / system can be left in the degraded mode per the Technical Specifications.
Item 2 "Chechiists of logs shall be provided for completion by the offgoing and ongoing auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational r
transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist)."
Action To Be Taken The oncoming /offgoing auxiliary operators perform a review of the switching and tagging orders still open, the caution tag logbook, the control room console and the status board to determine which safety-related equipment is under maintenance or test conditions. Documentation of their review and a 1661 047
47 list of the equipment is recorded at the time of their shift turnover in the control room log.
Instrument and control technicians and health physics technicians normally work day shift and thus do not perform turnovers. No checklists will be used by these technicians.
Item 3 "A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system al.gnments)."
The plant Quality Assurance Department will perform a surveillance check after implementation of the above and semiannually thereafter.
I661 048
48 20.
NUREG 0578 Requirement 2.2.2.a:
" Control Room Access."
Provisions shall be made for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators),
to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel.
Procedures shall be developed to establish the authority and responsibility of the person in charge of the control room to 14 ait access and to establish a clear line of authority and responsibility in the control room in the event of an emergency. The line or succession for the person in charge of the conttel room shall be established and limited to persons possessing a current senior reactor operator's license.
The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.
Action To Be Taken at Big Rock Point The Administrative Procedures, Chapter 4, clearly state that the shift supervisor is authorized to refuse entry or to direct personnel to leave the control room if their presence either interferes with operations or may compromise plant safety. The shif t supervisor has absolute authority over all control room activities.
During routine operations, the shift supervisor vill make periodic rounds and inspection of the plant systems and equipment. When the shift supervisor is absent from the control room, the control room operator No 1 automatically assumes direct control of the control room activities.
1661 049
49 During accident conditions, the shift supervisor remains in the control room at all times to direct the activities of the control room operators.
He may be relieved by another qualified shift supervisor as directed by the Operations and Maintenance superintendent, or by the Operations supervisor. A proper watch relief must be effected prior to his leaving the control room.
The Administrative Procedures, Chapter 4, and the Site Emergency Plan require the shift supervisor (or site emergency director if on site) to limit access into the control room during emergency or accident conditions to those personnel responsible for the direct plant operation and to those required to support plant operation during the emergency conditions.
1661 050
50 21.
NUREG 0578 Requirement 2.2.2.b:
"Onsite Technical Support Center."
A separate technical support center shall be provided for use by plant management, technical, and engineering support personnel.
In an emergency, this center shall be used for assessment of plant status and potential off-site impact in support of the control room command and control function. The center should also be used in conjunction with implementation of on-site and off-site emergency plans, including communications with an offsite emergency response center.
Provide at the on-site technical support center the as-built drawings of general plant arrangements and piping, instrumentation, and electrical systems.
Photographs of as-built system layouts and locations may be an acceptable method of satisfying some of these needs.
Action To Be Taken at Big Rock Point A temporary Technical Support Center (TSC) has been established and is located in the area of the shift supervisor and assistant plant superintendent's office. This area is just outside the control room.
Ine temporary T5C is shicidcl in a.uanner stinilos we che control room and, thus, is expected to be tenable during post-accident periods. The location adjacent to the control room permits direct observation of all control console instruments (except area radiation monitor readouts) through large viewing windows; Big Rock Point's control room is smaller than that of newer, larger plants and, thus, direct observation pcrmits adequate information transfer.
Area monitor readouts can be provided from the control room to the TSC by telephone upon request.
1661 051
51 The temporary TSC has on file controlled copies of:
a.
Site Emergency Plan b.
Site Fire Plan c.
Piping and Instrument Drawings d.
General Arrangement Drawings (Selected) e.
Electrical Schematics f.
Plant Administrative Procedures g.
Plant Operations Procedures h.
Plant Emergency Operations Procedures
- i. Operating License and Technical Specifications
- j. Plant Technical Data Book The center also has access to piping system isometric drawings and the remainder of the General Arrangement Drawings.
Emergency lighting is available in the TSC. Available communications links include a dedicated telephone line to NRC headquarters, a " health physics network" dedicated phone line to various NRC offices, telephone communications to the control room equipped with " executive override" ensuring priority use, and dedicated telephone lines to the near-site emergency operations center at our Boyne City, Michigan service center.
Plant Emergency Procedures specify the staffing requirements for the TSC.
Short-term management direction is provided by the plant superintendent and Operations and Maintenance superintendent. Technical support in the TSC is provided by the technical superintendent, technical engineer, and plant health 1661 052
52 physicist. The emergency plan separately requires the reactor engineer to report to the control room, adjacent to the TSC. Additional technical support can be provided by calling back additional members of the plant staff as aecessary.
Portable instruments are provided within the "SC and/or control room to monitor both direct radiation and airborne radioactive contaminants. An area monitor is located within the TSC and will provide a warning should radiation levels increase to levels of concern.
If average radiation levels in the southern extremity of the TSC should exceed 100 mR/h (this portion of the TSC has lesser shielding than others), that portion of the center will be set off bounds.
If radiation levels should dictate complete evacuation of the TSC, the Site Emergency Director will relocate to the control room and the technical staff will proceed, at his direction, to the Operations Support Center or the near-site Emergency Operations Center.
Provision of a permanent TSC meeting all requirements of NUREG 0578 is not possible at this time since no adequately shielded location exists on site pending completion of shielding modifications in response to 2.1.6.b above. A permanent TSC will be provided in conjunction with these shielding modifications. Plans for the permanent TSC are discussed in more detail in Appendix E.
I661 053
53 22.
NUREG 0578 Requirement 2.2.2.c:
"Onsite Operational Support Center."
Each operating nuclear power plant should establish and maintain a separate on-site Operational Support Center outside the control room.
In the event of an emergency, shif t support personnel (e.g., auxiliary operators and technicians) other than those required and allowed in the control room shall report to this center for further orders and assignment.
Action To Be Taken at Big Rock Point The air compressor room has been designated as the Operational Support Center.
This space is separate frc? the control room and is shielded in a manner similar to the control room. Communication to the control room is available via the plant telephone system. The emergency plan has been revised to reflect use of this area as an Operational Support Center. Consideration will be given to collocating the Operational Support Center with the permanent Technical Support Center when designing the new structure discussed in response to Requirement 2.2.2.b above.
1661 054
54 23.
NUREG 0578 Requirement 2.2.3:
" Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability."
A rulemaking proceeding will be instituted to require that each NRC nuclear power plant license contain a provision requiring immediate plant shutdown in the event of complete loss of some safety function. A special report would have to be approved by the Director of Nuclear Reactor Regulation before subsequent start-up would be permitted.
Action To Be Taken at Big Rock Point NRC letter dated September 13, 1979 stated that no action need be taken at this time with respect to this requirement. Nevertheless, Censumers i ver Company wishes to state, for the record, that we consider the specific requirement recommended in NUREG 0578 inappropriate. Consumers Power Company agrees that complete loss of a safety function would be a serious occurrence and should be treated as such. However, means exist for ensuring that such occurrences receive the attention they deserve other than requiring a punitive plant shutdown before investigation of circumstances. Consumers Power Company intends to participate in the forthcoming rulemaking process either individually or in conjunction with other parties.
1661 055
55 24.
ACRS Requirement: " Instrumentation To Monitor Containment Conditions During the Course of an Accident."
Instrumentation shall be provided to monitor the following parameters af ter an accident:
1.
Containment Pressure. Measurement and indication capability shall extend from minus 5 psig to four times containment design pressure for steel C o n*wainmcLts.
2.
Hydrogen Concentration. Measurament capability shall be provided from 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
3.
Containment Water Level. Capability shall include that level equivalent to 500,000 gallons for PWRs. For BWRs, levels from the bottom of the suppression pool to 5' above normal water level shall be included.
Instrumentation shall include continuous indication and recording and shall meet the design and qualification provisions of Regulatory Guide 1.97 including redundancy and testability.
Action To Be Taken at Big Rock Point The required containment pressure and hydrogen concentration monitoring capabilities will be provided.
Instrumentation will meet the requirements of Regulatory Guide 1.97, Revision 1 (which references Regulatory Guide 1.89, Revision 0).
Containment pressure monitoring capability meeting the specified range will be provided by January 1, 1981 in addition to existing 1661 056
36 instrumentation. Hydrogen monitoring will be provided by January 1,1981.
Detailed design descriptions will be submitted for NRC information by July 1, 1980.
Big Rock Point does not have a suppression pool. The requirement applicable to monitoring PWR containment water level is, therefore, more appropriate for Big Rock roint than the requirement for BWRs. Big Rock Point's ECCS system is designed to accomplish long-term cooling by removing water from the containment, passing it through a heat exchanger, and returning it via the core sprays. Switchover to this recirculating mode from the initial lineup involving direct injection of Lake Michigan water via low pressure core sprays is accomplished manually based on containment water level.
Containment water level instrumentation, therefore, was a part of original plant design.
Continuous water level measurement exists from the level of the core spray recirculation pump suction strainers (574 feet) to a level approximately 6 to 9 feet above the maximum water level at which switchover to recirculation mode is to occur (maximum indication at 596 feet compared to the lowest elevation inside containment of approximately 570 feet). This system will be modified to provide recording of the measurements. Additional containment water level indication is provided by four float-type level switches (elevations 574, 579, 587, and 595 feet) represented by console-mounted indicator lights in close proximity to the continuous indication readout. A redundant continuous measurement system meeting the requirements of Regulatory Guide 1.97, Revision 1, will be installed. The redundant instrument reading will be recorded on a recorder separate from that to be used for the existing instrumentation. The new redundant instrument will be installed by January 1, 1981.
1661 057
57 25.
Director, NRR Requirement:
" Remotely Operated High Point Vents in the Reactor Coolant System."
Reactor coolant system high point vents, remotely operable from the control room shall be provided. Design shall conform to 10 CFR 50, Appendix A, General Design Criteria (later clarified by NRC Staff in a topical meeting held October 11, 1979 not to require redundant, single failure proof valving provided a failure would not result in a leak in excess of makeup capability).
Each licensee shall provide:
1.
A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe. The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.
2.
Analyses demonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentration limits in containment as described in 10 CFR 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Section 6.2.5.
3.
Procedural guidelines for the operators' use of the vents. The information available to the operator for initiating or terminating vent usage shall be discussed.
1661 058
58 Action To Be Taken at Big Rock Point The high point of the Big Rock Point reactor coolant system is the two tube bundles of the emergency condenser. A continuously open line vents the reactor vessel head to the steam drum and, thus, no isolated " pocket" of gas could exist in the reactor vessel. The required venting capability will be provided by installation of remotely operable vent valves on each of the emergency condenser tube bundles. Vent valves will be configured such that a single failure will neither create a reactor coolant system leak nor prevent venting of at least one tube bundle. A description of the proposed vent system design is provided as Appendix D to this enclosure for NRC review.
Failure of a vent line resulting in a LOCA would be the same as a small steam line break. Thus, previous analyses are adequate to describe this situation and no new LOCA analysis is needed.
Results of an analysis of hydrogen generation following an accident were reported in Consumers Power Company letter dated May 4, 1979 as modified in the response to Requirement 2.1.5.c above. This analysis assumed all hydrogen was released into containment as soon as it was generated.
In view of the low maximum value for hydrogen concentration demonstrated by this conservative analysis, no additional analyses based on operation of the emergency condenser tube bundle vent valves need be performed.
Procedures governing use of the proposed vent system vill be developed.
1661 059
59 26.
Near Term Requirements for Improving Emergency Preparedness (1) Upgrade licensee emergency plans to satisfy Regulatory Guide 1.101, with special attention to the development of uniform action level criteria based on plant parameters.
(2) Assure the implementation of the related recommendations of the Lessons Learned Task Force involving instrumentation to follow the course of an accident and relate the information provided by this instrumentation to the emergency plan action levels. This will include instrumentation for post-accident sampling, high range radioactivity monitors, and improved in plant radioiodine instrumentation. The implementation of the Lessons Learned Task Force's recommendations on instrumentation for detection of inadequate core cooling will also be factored into the emergency plan action level criteria.
(3) Determine that an emergency operations center for Federal, State and local personnel has been established with suitable communications to the plant, and that upgrading of the facility in accordance with the Lessons Learned Task Force's recommendation for an in plant Technical Support Center is underway.
(4) Assure that improved licensee off-site monitoring capabilities (including additional thermoluminescent dosimeters or the equivalent) have been provided for all sites.
i661 060
60 (5) Assess the relationship of State / local plans to the licensees' and Federal plans so as to assure the capability to take appropriate emergency actions. Assure that this capability will be extended to a distance of ten miles. This item will be performed in conjunction with the Office of State Programs and the Office of Inspection and Enforcement.
(6) Require test exercises of approved emergency plans (Federal, State, local and licensees), review plans for such exercises, and participate in a limited number of joint exercises. Tests of licensee plans will be required to be conducted as soon as practical for all facilities and before reactor start-up for new licensees.
Exercises of State plans will be performed in conjunction with the concurrence reviews of the Office of State Programs. As a preliminary planning basis, assume that joint test exercises involving Federal, State, local and licensees will be conducted at the rate of about ten per year, which would result in all sites being exercised once each five years. Revised planning guidance may result from the ongoing rulemaking.
Action To Be Taken at Big Rock Point 1.
The plant emergency plan has been upgraded to meet the requirements of Regulatory Guide 1.101 and has been submitted in draft form for NRC review.
2.
Actions in response to NUREG 0578 requirements are discussed separately above.
1661 061
61 3.
The Emergency Operations Center has been established at Consumers Power Company's Boyne City Service Center. Communications' links to the Technical Support Center were discussed in response to Requirement 2.2.2.b above. Alternate locations for the Emergency Operations Center are the State Police Post, Petoskey, MI, and the Charlevoix County Sheriff's Office.
4.
Provisions for plume location and monitoring, including the use of isopleths, are described in response to NUREG 0578, Item 2.1.8.b, Response Section 2, " Interim Methods for Quantifying Radioiodine and Particulate Effluents." In addition, we have reviewed the advisability of additional environmental thermoluminescent dosimeter (TLD) stations with analytical backup from our central TLD Laboratory in Jackson, Michigan. We are proceeding with plans for installation of additional TLD monitoring stations in each overland sector. Two rings of dosimeters will be provided. The rings will be approximately one to two miles and four to five miles from the plant, respectively.
Each station will be provided with a package of three emergency environmental TLDs. During emergency conditions, the top TLD will be removed from each location at daily intervals. A three-day supply of dosimeters will be maintained in a shielded off-site location to allow a new dosimeter to be placed at the bottom of the package.
5.
State and local emergency plans have been completed and submitted to NRC Region III. They will be reviewed against current criteria by July 30, 1661 062
62 1980 and against the upgraded criteria of Regulatory Guide 1.101 by January 1, 1981.
6.
Individual tests of plant and State emergency plans will be accomplished by July 30, 1980. A joint test exercise will be held within five years.
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APPEIDIX A DIRECT INDICATION OF SIEAM DRUM SAFEJ VALVE POSITION 1661 064