ML19257C279
| ML19257C279 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/18/1980 |
| From: | Fields H CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001250487 | |
| Download: ML19257C279 (25) | |
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General Offices: 212 West Michigan Avenue Jackson. Michigan 49201. Area Code 517 788-0550 January 18, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - REQUIREMENTS RESULTING FROM REVIEW OF TMI-2 ACCIDENT - ACTIONS TAKEN IN RESPONSE: TRANSMITTAL OF UPDATED PAGES Consumers Power Company letter dated December 27, 1979 described actions to be taken at Big Rock Point in response to requirements resulting from NRC review of the TMI-2 accident. Review of this letter has identified a need to make minor changes.
Transmitted herewith are revised pages (marked Rev 1) incorporating the necessary changes to reported actions. These pages should be substituted for pages of the same number in the enclosure to our December 27, 1979 letter.
Changes are identified by a vertical line in the margin. For your convenience, a list of effective pages is provided.
Howard Fields (Signed)
Howard Fields Nuclear Licensing Engineer CC JGKeppler, USNRC Att 24 Pages
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1809 MB 8001250+87
IMPLEMENTATION CRITERIA FOR POST-TMI REQUIREMENTS AT BIG ROCK POINT List of Effective Pages Page Number Revision 1
Original 2
1 3-4 Original 5
I 6-21 Original 22-24 1
25 Original 26 1
27 Original 28-30 1
31-32 Original 33-37 1
38 Original 39-40 1
41-44 Original 45 1
46-47 Original 48-49 1
50-55 Original 56 1
57 Original 58 1
59-62 Original
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e 1809 99
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, during a loss of air event. The series pilot-operated valve, which is unaffected by loss of air, will 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.
i There would be no backup valve, however, should it also fail.
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y %.1 1809 +90 Rev 1, 1/18/80
5 3.
NUREG 0578 Requirement 2.1.3.a:
" Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWR'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: project requirements, construction specifications for the portion 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 energized prior to plant start-up. The system will be operable, following final adjustment for correct bandwidth, when full power conditions are achieved.
809 E R L Rev 1, 1/18/80
22 release of radicactivity as a result of core damage have been developed for use until sangling 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 por.sible to obtain samples of core spray recirc41ation water from the core spray heat exchanger.
Calculated radiatien levels for various postulated accidents indicate that such samples could probably be obtained if less than 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 would 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 written to obtain samples of core spray recirculation water. The procedure consists 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 har.dling highly radioactive samples. The procedure also provides for involvement of different 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 1809 ' t'r2' Rev 1, 1/18/80
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 minimtm training of personnel accomplished before start-up, following the outar,e discussed in reipoase to Requirement 2.1.3.a.above, consisted of a review of the procedure cud 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 an ionization chamber or GM probe. The detector is located in the cable penetration room near the containment sphere and has remote readout in the Operations Support Center.
(See 2.2.2.c below.) The monitor has 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 104 R/h for a 100% core melt accident (assumed instantaneous release at time of accident) and 2 x 104 R/h for release of total core gap activity. A graph of radiation level versus time for 100%
fuel failure (based on calculated radiation levels from noble gases inside containment) is provided for use in interpreting readings of this monitor..
I I
6 1809 T35 Rev 1, 1/18/80
24 The monitor was in place and procedures for its use as discussed above l
were compleced before start-up from the cutage 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 I above can be used.
3.
Long-Term Solution
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Consumers Power Company's preliminary evaluation of possible methods of providing the required capability in the long term centered around 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 volume 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 samples for laboratory analysis is not needed..] [ b 1809 M4 Rev 1, 1/18/80
26 14.
NUREG 0578 Requirement 2.1.8.b:
"Incressed 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 raeasuring and identifying radioiodine and particulate radioactive effluents under accident conditions.
Action To Be Taken at Big Rock Point The following will be provided by January 1, 1981:
Radioactive noble gas effluent monitors from ALARA ranges to 10 pCi/cc.
Capability of radiciodine sampling of effluents followed by on-site analysis.
Containment radiation level monitors (minimum of 2) capable of measuring radiation levels to a maximum of 106 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 1010 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 instruments of the specified ranges.
h 1809 5 Rev 1, 1/18/80
28 maximum hypothetical accident (MHA) source term with leakage at a rate of 0.5% of gaseous inventory per day and 30,000 cfm stack flow. The current monitor samples a stream which,would contain all leakage from turbine building sources, including 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 100 R/h have been determined for chosen sensor location in order to quantify release rates from the level of current instruments and procedures to the level of 10,000 Ci/s. Sensitivity of the sensor to Xe-133 (81 kev) will be accounted for in the graph for dose rate-to-release rate conversion. 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, tables, and l
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 to eliminate direct shine from the unshielded containment structure. Measurements Rev 1, 1/18/80 1809 M
29 are based upon a stack design emission rate of 30,000 cfm.
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 are displayed in the Operations Support Center.
(See 2.2.2.c below.)
(4) Radiation level is displayed continuously with range selection l
capability at the readout station.
(5) Monitoring instrumentation is powered from the 120 V a-c emergency bus (3Y).
180915d Rev 1, 1/18/80
30 b.
Description of Procedures All necessary procedures and aids described below were complete prior to start-up from the outage discussed in 2.1.3.a above. The 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 Guide 1.3 Maximum Hypothetical Accident. Furthermore, a release rate of 10,000 Ci/s noble gas, with Regulatory Guide 1.3 noble 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 after 15 minutes of sampling. Provision of shielding for stack access alone would not solve the problem of removing a stack sampler filter with activity this high in the sample media. Due to these personnel exposure concerns, emergency
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l80 W S Rev 1, 1/18/80
31 procedures have been modified to specifically prohibit the removal of l
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 require the Pasquill l
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 the 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 cf 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 6 x 10-10 pCi/cc. This is sufficiently low to permit detection prior to reaching levels corresponding to a State of Michigan Nuclear Incident N
jgQ}
Rev 1, 1/18/80
32 Class B.
The previously calculated X/Q can then be used to calculate an estimated release rate.
The procedures and equipment described above were available for use before start-up.
3.
Noble 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 curies 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 modiacations (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.
1809 Rev 1, 1/1e/80
33 4.
High Range Containment Monitors The lack of containment shielding at Big Rock allows accurate determina-tionofradiationlevelsinsidebymeansof$xternalmeasurement.
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 10 R/h 4
(approximately 9 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.
Rev 1, 1/18/80
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l809 16T
34 15.
NUREG 0578 Requirement 2.1.8.c:
" Improved In-Plant Iodine Instrumentation."
Each licensee shall provide equipment and associated training and precedures 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 radiciodine monitoring with conventional charcoal canisters may produce overly conservative results under accident conditions due to interference from radioactive noble gases. Occupancy 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 normal 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 have been procured.
Respiratory protection is currently available and potassium iodide as a thyroid blocking agent will be available (estimated shipping date 2/1/80) to reduce thyroid burdens of radioactive iodine.
1 /%
80O/
lJL Rev 1, 1/18/80
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, Ludlum 177 or equivalent) 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 a case-by-case basis depending upon the individual action required in the airborne area. The Company's general medical consultant has developed a procedure for the use of potassium iodide in emergency situations.
Initial draft of this procedure was submitted to the Company on December 13, 1979 and will be implemented by January 15, 1980.
b.
Equipment Impact (1) Without teuse, a quantity of approximately 200 cartridges is adequate to permit sampling of the Control Room / Technical Support Center (TSC) and the Operations Support Center atmospheres as 1809 163-I W Rev 1, 1/18/80
36 conditions dictate for 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.
1 (2) Portable air samplers powered by 120 V a-c emergency bus and survey instruments are available for use.
(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 site.
c.
Training The sample procedure is similar to existing surveillance air sample procedures and utilizes existing air sampling equipment and counting instruments. Training for health physics personnel has consisted of a required review of the procedure. A group discussion on each shift, led by the responsible assistant supervisor, also has been performed.
h 1809 +6T Rev 1, 1/18/80
37 2.
Longer Term Actions Orders are being placed for several additional a-c/ battery-operated friskers (Ludlum Model 177 or equivalent ratemeters) to supplement the existing supply and ensure that adequate numbers of friskers will be available for contamination monitoring and air sample counting during an emergency.
Battery-operated ratemeters (Eberline PRM-6 or equivalent) will be purchased to permit monit.rting and air sample counting in the field during emergencies.
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1809 t65 Rev 1, 1/18/80
39
- 17. NUREG 0578 Requirement 2.2.1.a:
" Shift Supervisor's hesponsibilities."
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 issueu by the Vice President for Nuclear Operations. This directive clearly sr.ates 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 emergency situations until properly relieved; during normal operations, the shift supervisor's duty station is the shift supervisor's office and/or control room. During the shift supervisor's absence from the control room, he informs the #1 control room operator to assume conunand.
Plant administrative procedures 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 shift
,lh supervisors.
1809 it6 Rev 1, 1/18/80
i 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 by the Vice President-Nuclear Operations.
180916$
Rev 1, 1/18/80
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 oncoming /offgoing 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 caution 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 shift supervisor's log and control room log and verified by bot.h oncoming /offgoing shift supervisors and control operators at the time of shif t turnover.
"c.
Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications.
For such systems and componer'.',. che length of time in the degraded mode shall be <> pa s ith the Technical Rev 1, 1/18/80 1809 4M
48 20.
NUREG 0578 Requirement 2.2.2.a:
" Control Rooc Access."
Provisions chall 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 advisora 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 limit access and to establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license.
The plan shall clearly define the lines of 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 shift supervisor has absolute authority over all control room activities.
During routine operations, the shift supervisor will make periodic rounds and inspection of tne plant systems and equipment. When the shift superviso-is absent from the control room, the control room _ operator No 1 is assigned direct control of the control room activities.
k 1809~itO tu' Rev 1, 1/18/80
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.
In certain rare instances, the shift supervisor may leave the control room during emergency situations provided the shift technical advisor remains in the control room during his absence. Prior to his leaving the control room, the shift supervisor will inform the No 1 control operator to assume command of 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.
G.
1809 tMT Rev 1, 1/18/80
56 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 Point 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 instrumenta-tion, 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 indi-cation 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 by January 1, 1981. 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, Revi-sion 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.
1809 ~R Rev 1, 1/18/80
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 by January 1, 1981. 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 liue 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 will be developed.
x13 l809 M
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Rev 1, 1/18/80