ML20027C911
| ML20027C911 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 08/16/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1997, NUDOCS 8210270417 | |
| Download: ML20027C911 (25) | |
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MINUTES OF THE ACRS SUBCOMMITTEE ON MIDLAND PLANT UNITS 1 & 2 JUNE 2,1982, WASHINGTON, D.C.
The ACRS Subcommittee on Midland Plant Units 1 and 2 held a meeting on June 2, 1982 in Room 1167 and 1130,1717 H St. NW., Washington, D.C.
The purpose of the meeting was to review the application of Consumers Power Company (CPCo) for a license to operate Midland Plant Units 1 and 2.
Approximately 30 minutes of this nieeting were closed to the public under exemption 3 of the Sunshine Act to discuss details of the Applicant's physical security program and systems.
Notice of this meeting was published in the Federal Register on Wednesday, May 26, 1982.
A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B.
The schedule for the meeting is included as Attachment C and a list of all reference mater-ial (including slides and documents provided to the Subcommittee at the meeting) l 1s included as Attachment D.
A complete set of handouts has been included in the ACRS Files.
There were no oral or written statements from the public.
The Designated Federal Employee for this meeting was Mr. David C. Fischer.
SUBCOMMITTEE CHAI_RMAN'S OPENING REMARKS Dr. Okrent opened the meeting with a statement on the purpose and goal of the meeting.
t STATUS OF THE NRC STAFF REVIEW Mr. Darl Hood (NRC/NRR/DL/LB4) indicated that the open items and licensing conditions remained unchanged from the discussion that was presented at the ACRS Subcoruittee meeting on Midland Plant Units 1 & 2 dated May 20-21, 1982.
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MINUTES / MIDLAND, JUNE 2,1982 QUESTIONS ON THE NRC STAFF'S SAFETY EVALUATION REPORT The Subcommittee members and consultants questioned the NRC Staff and Applicant on items in the operating license Safety Evaluation Report (0L SER).
Dr. Okrent asked the Staff what the open item related to Natural Circulation Cooldown Analysis entailed.
The Staff said that they wished to see the results of analyses and tests by the Applicant showing that the reactor system can be brought to cold shutown under all (including accident) conditions within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Dr. Okrent asked why the issue of potential loss of l
natural circulation capability in this type of reactor under certain small break loss-of-coolant accident (SBLOCA) conditions i
was not addressed in the SER or brought to the attention of the Subcommittee.
He suggested that the Staff " caucus and find out where they think this question of natural circulation in the presence of small LOCAs stands as a safety issue." The Staff indicated that it would.
Nothing further was said on this issue l
during the meeting.
Dr. Okrent asked if there was any problem with regard to diesel generator accessibility during the probable maximum flood (PMF).
l The Staff indicated that the PMF would not preclude access to the diesel generator building.
Dr. Moeller asked what impact flooding might have on the evacuation of the nearby chemical plant and on Midland's Emergency Plan. Mr.
l Slade, CPCo, said that due to the difference in elevation, the Dow Chemical Plant would have to shutdown and evacuate long before the nuclear power plants experience any flooding problem.
MINUTES / MIDLAND, JUNE 2,1982 Dr. Okrent asked the Staff how the turbine missile open item would be resolved.
Mr. Hood referred Dr. Okrent to the meeting handout, "NRC Staff Responses To Questions By The ACRS Subcommittee During The Meeting Of May 20-21, 1 982 On Midland Plant, Units 1 And 2."
Dr. Okrent asked about the accuracy of pump net positive suction head (NPSH) calculations.
He questioned whether credit can be taken for containment pressure when computing NPSH available, particularly when comparing NPSH available to NPSH required. Mr.
Tedesco (NRC/NRR/DL) said that the Staff will not allow credit to be taken for containment pressure.
Dr. Okrent asked the Staff how it determined the reliability of a plant's containment isolation system.
Mr. Tedesco indicated that the Staff has no firm requirements or criteric regarding containment isolation reliability.
He said that the quality of components in the system is checked.
The Staff's primary concern is with the containment isolation system configuration.
Mr. Lipinski, ACRS consultant, questioned the ability of the con-tainment purge and vent valves to operate under a differential pressure.
The Applicant indicated that these valves can close l
under the containment's 60 pound design pressure (ie, the valves can operate with a 60 pound differential pressure).
Dr. Okrent asked if either the Staff or Applicant had analyzed the failure modes of the Feed Only Good Generator (FOGG) system.
He questioned whether both steam generators could be inadvertently isolated via this system.
The Staff said that because of their
MINUTES / MIDLAND, JUNE 2,1982 familiarity with the system and the way the system is designed, they are confident that no single failure will cause both steam generators to be isolated simultaneously.
Dr.
Okrent indicated that the Staff might want to consider failure modes under situations other than just the usual single failure criterion.
Dr. Okrent asked if an analysis had been conducted to determine if steam line failures might cause dynamic forces on valves that are required to isolate the failed steam line.
Mr. T.
Ballweg, Bechtel, indicated that such an analysis has been con-ducted.
ITEMS FROM PREVIOUS ACRS LETTERS The Subcommittee reviewed and discussed the written response of the NRC Staff to concerns expressed in previous ACRS reports on Midland Plant, Units 1 and 2.
Dr. Okrent questioned the Staff on the status of equipment qualification at Midland.
He asked on what basis the ACRS should assume that this matter is or will be satisfactory for Midland.
The Staff outlined their equipment qualification review procedures and schedule. Mr. Hernan (NRC/NRR/DL/LB4) i said that the Applicant understands the Staff's requirements and has established an equipn'ent qualification program that the Staff finds acceptable. As a l
result of the program's suitability, the Staff has scheduled an equipment qualification audit.
o MINUTES / MIDLAND, JUNE 2,1982 METHODS TO REDUCE COMMON CAUSE FAILURE Mr. B. Harshe, CPCo, described Midland's program for reducing the potential for common cause failures.
He defined common cause failures as systems inter-actions.
He identified three different types of systems interaction:
spatial interactions, functional interactions, and human induced interactions.
Spatial interactions were described as the coupling of systems by virtue of their proximity to each other (ie., physical interaction).
Spatial interactions will be identified by plant walkdowns.
These walkdowns will look for things like the potential effects of Category II equipment being located above Category I equipments, the potential effects of flooding from various sources, and the potential effects of high energy line breaks.
In addition, spatial-type walkdowns will be done to check for the effects of thermal growth and stress, to evaluate the design adequacy of the fire protection system, and in conjunction with the turnover of each system from the construction crew to the operations crew.
Functional interactions can occur where systems are coupled together through a process.
Functional interactions are being identified by Bechtel's design con-trols, through the probabilistic risk assessment (PRA), through control systems failure analysis (eg., failure modes and effects analysis of the integrated control system), through preoperational testing, and through the review of oper-ating experience.
Human induced interactions are being identified in conjuntion with: operator training (ie., on the plant mock-up or plant-specific simulator when available), the detailed Control Room Design Review, and Plant operations.
The scope of Midland's program to identify adverse systems interr.ctions will include both safety-grade / safety-grade and nonsafety-grade / safety-grade types of interactions.
Mr. Harshe identified several plant modifications which l
l MINUTES / MIDLAND, JUNE 2,1982 have been made as a result of CPCo's search for adverse systems inter-actions.
Examples given included one resulting from the PRA and one from an evaluation of an operating experience at another plant.
Licensee event reports, INP0 documents, and information provided by B&W are all factored into CPCo's search for systems interactions.
Mr. Harshe indicated that while this CPCo program receives little Staff guidance (ie., via the Standard Review Plan or a NUREG) their program is comparable to what the Staff is asking to be done on Indian Point related to systems interactions.
INTEGRATED CONTROL SYSTEM Mr. R. Hamm, CPCo, outlined the functions, interfaces, and improvements that have been made in Midland's integrated control system (ICS). This B&W designed system coordinates the response of the reactor's steam gene-rators and and turbines.
He showed and explained a simplified block diagram of the functions of Midland's ICS.
He described the interfaces that the ICS has with other plant systems.
He said that the ICS requires inputs from many other systems to operate correctly and to supply information to the control system.
The majority of the inputs to the ICS are from the non-nuclear instrumentation (NNI).
Midland's ICS differs from other B&W ICSS in that:
it incorporates an evaporator (process) steam demand system, the auxiliary feedwater (AFW) system at Midland is safety grade and controls of the AFW system are independent of the ICS, and Midland has indications in the control room which are independent of the NNI and ICS
MINUTES / MIDLAND, JUNE 2,1982 After reviewing B&W's failure modes and effects analysis of its standard ICS and reviewing ICS related events at Rancho Seco and Crystal River, CPCo decided to make several other modifications to its system to:
improve the external power supply reliability to both the NNI and ICS by providing them with redundant batteries,
- reduce the fuse sizes in the NNI and ICS to minimize the probability that local faults would result in complete loss of power to the system, incorporate features to ensure that the spray valve would not fail open and the pressurizer heaters would not fail on when power was lost, and provide annunciator alarms in the control room to alert the oper-t l
ator of a loss of control power to NNI and/or ICS.
CPCo is identifying in its emergency procedures those actions that the operators need to take in the event of a loss of power to the NNI/ICS.
These procedures will also identify the instrumentation on which the operator 1
can rely.
The alarm to tell the operator that he has lost power to the NNI/ICS will not be safety grade.
Mr. Hamm said that while the ICS has yet to be proved on Midland, the system should be tuned so that the plant can take i
a turbine trip from 100% power, run the reactor back to 15% power on the bypass through the condenser, and preclude a reactor trip.
In response to a question from Dr. Okrent, Mr. Hamm said that Midland has safety grade systems to preclude
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MINUTES / MIDLAND, JUNE 2,1982 steam generator overfill as well as to peclude steam generator underfill and depressurization.
Mr. Ballweg, Bechtel, indicated that Midland's steam generator overfill protection cost somewhere in the range of a quarter to a half of a million dollars per unit (total installed cost).
He said that only a small percentage of this was direct hardware cost; most of the cost was for engineering evaluation, for modifications to related interfacing systems, and for construction labor.
Mr. Hamm next described ongoing evaluations of the ICS, NNI, and Evaporator Steam Demand Development (ESDD).
The effects of other control systems have been included in these evaluations (if they share sensor inputs or instrument lines with ICS, NNI, or ESD0).
Events being considered include:
Loss of single sensor input,
- Break of instrument lines having more than one instrument with at least one input into the ICS, NNI, or ESDD,
- Failure of individual fuses or breakers, and
- Complete loss of power.
Mr. Hamm indicated that Midland's ICS and AFW System designs minimize the possibility of cold overpressurization at Midland. He said that Midland's PORV is independent of the NNI and is safety grade (At Crystal River the PORV was controlled by the NNI and failed open on loss of power to the NNI).
Dr. Okrent asked Mr. Hamm how ICS control room instrumentation failed. Mr.
Haan responded that the system was -10v to +10v and consequently failed midrange. He said it would be very expensive to modify the system so
MINUTES / MIDLAND, JUNE 2,1982 that ICS instrumentation failed high/ low on loss of system power.
He said that is why CPCo decided to install a loss of ICS power alarm in the control room.
He added that operators are specifically trained to recognize a loss of power to the ICS.
SEISMIC AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT IMPORTANT TO PLANT TAFTTy Mr. J. Zabritski, CPCo, discussed the seismic and environmental qualifi-I cation of equipment important to plant safety.
He mentioned the organi-l zations that are participating in Midland's equipment qualification programs.
l In discussing the elements of Midland's program to environmentally qualify equipment, Mr. Zabritski said that their program is developed in accordance with current criteria.
It addresses the qualification of electrical and
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mechanical equipment in both harsh and mild environments.
It accounts for l
resolution of discrepancies and deficiencies and contains corrective action plans which allow for retesting, shielding, moving, or replacing of equip-ment if it does not meet the requirements.
He said that Midland's program is a 40 year (life of the plant) program.
Mr. Zabritski provided the Subcommittee with the status of the environment qualification review.
In discussing the l
elements of Midland's program to seismically qualify equipment, Mr. Zabritski said that their program is developed to evaluate all safety-related equipment.
The program is based on the seismic floor response spectra as revised in 1982 to account for structural changes which resulted from soils remedial activities and for model changes.
MINUTES / MIDLAND, JUNE 2,1982 Equipment is being requalified in accordance with the FSAR commitments.
In addition, all equipment is being evaluated against the current NRC seismic criteria.
The margin review program will be looking at equip-ment relative to the seismic site specific response spectra. Mr.
Zabritski gave a brief status of the seismic qualification review.
Dr.
Okrent questioned why equipment qualification is an open item as opposed te confirmatory for Midland.
The Staff said that this is because some of the Applicant's submittals are not yet in.
Dr. Okrent asked whether seismic qualification is done by analysis or test.
He asked, if a test is used, is the equipment tested to the design motion or is it tested beyond the design motion (eg., to failure).
Mr. Zabritski responded that some equipment, such as valves, are generically qualified.
He said other equipment is tested only to the specified floor spectra (ie., the floor spectra is a input to the test platform).
Mr. Jacobsen, CPCo, indicated that very few margins to failure are determined since few fragility tests are run.
Mr. Lipinski, ACRS consultant, expressed concern that the resonance frequencies of small component qualified by test are not determined and used as an input to the qualification test.
DECAY HEAT REMOVAL SYSTEM OPERATION Mr. L. Gibson, CPCo, described the various means of removing reactor decay l
heat at Midland following a shutdown or reactor trip.
He said that CPCo's original intention was to design their systems so that the hot shutdown condition could be achieved and maintained using only safety grade equipment.
l Midland's present design allows for taking the plant to cold shutdown using l
l
l MINUTES / MIDLAND, JUNE 2,1982 11 -
only safety grade equipment.
Heat rejection through the steam generator can be achieved via the: main steam isolation valves (to the condenser or modulating atmospheric dump valve), auxiliary feedwater system, power-operated relief valves (PORVs), or code safety valves.
Heat rejection can also be accomplished through the decay heat removal system or by feed and bleed operations.
The temperatures and pressures at which each of these methods can be used was described.
Simplified schematics of the auxiliary feedwater and decay heat removal systems were explained.
There are two decay heat removal loops at Midland; either can handle the system's design heat load.
CPCo has done an analysis which shows that feed and bleed operation using one high pressure coolant injection pump at the PORV setpoint is adequate to remove the decay heat generated in the primary coolant system.
BOLTING AND HIGH STRENGTH MATERIALS Mr. H. Slager, CPCo, outlined the bolting experiences at the Midland site.
He listed four groups of low-alloy quenched and tempered bolts (2000 series steel bolts) which have failed or been found deficient.
These bolts include:
- Unit I reactor vessel anchor bolts,
- Pipe whip restraint bolts,
- Steam generator anchor bolts, and
- Reactor coolant pump snubber anchor bolts.
As a result, CPCo has initiated a survey to look at a large portion of the safety-related bolting at the Midland site to assess, based on hardness testing,
l MINUTES / MIDLAND, JUNE 2,1982 how they expect these bolts to perform.
This survey is being conducted on a statistical basis using a 95% probability and 90% confidence level.
The bolts which were found to be bad were supposedly sampled as part of an ASTM procedure.
Dr. Okrent asked if CPCo had gone back to look at all ASTM bolt specifications to decide for themselves which they felt were adequate.
Mr. Slager indicated that CPCo had not.
He said that CPCo is more concerned with the adequacy of installed bolts.
Mr. Mathis questioned why CPCo's QA receipt inspection had not detected the deficient bolts prior to installation.
Mr. Slager summarized the Unit I reactor vessel anchor bolt problems.
He said that 3 of Unit l's bolts had failed within 8 months of preloading.
l The bolts were found to have failed due to stress corrosion cracking followed by fracture due to low toughness.
Hardness in the area of failures was as high as Rockwell C-48.
Corrective action to preclude the failure of more reactor vessel anchor bolts includes lowering the prestress from 92ksi to 6ksi on all the anchor bolts for both Unit 1 and Unit 2.
Since these bolts can no l
longer support all of the loads for which they were originally designed, 12 upper lateral supports are being added to the reactor vessel. These supports will stick out from the primary shield wall and come close to the re-l actor vessel during normal operation.
They will absorb LOCA loads in the event of a pipe break.
In addition, CPCo is limiting the accident loading on these bolts to 70 percent of the proof load.
Mr. Slager presented the results of several analyses which showed the adequacy of some of the remaining 2000 series bol t s.
He indicated that the overall bolt survey results are not yet available.
Mr. C.D. Sellers from the Staff acknowledged the adequacy of CPCo's program to correct bolting deficiengies.
e MINUTES / MIDLAND, JUNE 2,1982 FIRE PROTECTION Mr. R. Polich CPCo, outlined Midland's fire protection program and addressed several specific topics related to fire protection in general.
He said that the few areas of concern related to the fire protection open item will be suitably resolved in the near future.
He commented on the potential for flooding and wetting of critical components resulting from both system actuation and from breaks in the system.
He said that the plant has been de-signed to preclude such flooding and wetting.
Dr. Okrent questioned whether there is a systematic look at the combined effects of age and water outside the containment as well as inside the containment building.
The Staff, Applicant, and Bechtel all indicated that cabling is designed to allow for wetting without causing a malfunction and that the effects of low energy line breaks (including from the fire protection system) are evaluated in a systematic fashion.
Mr. Polich also talked about fire damper actuation.
He said that all fire dampers are rated for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> against fire.
Mr. Eberling of the Staff stated that there is, however, no requirement for the dampers to be gas-tight.
Fire dampers at Midland will be tested in place prior to plant operation.
Finally Mr. Polich addressed the potential for spurious equip-ment actuation.
He said that equipment which is required to achieve and maintain shutdown has been protected to assure operability of one train after any single exposure to fire.
This protection ensures that power will be avail-able to at least one train of the necessary equipment. Mr. Polich stressed that design features at Midland make a hot shutdown analysis unnecessary.
The Staff indicated that their review of Midland's fire protection program is l
ongoing.
They declined to comment on its adequacy.
MINUTES / MIDLAND, JUNE 2,1982 CONTROL ROOM HABITABILITY Mr. B. Harshe, CPCo, discussed Midland's control room habitability and specific hazards existing at the Midland plant that could affect control room habitability.
He explained that CPCo has done a systematic survey within a five-mile radius of the plant, looking for both toxics and explosive hazards.
CPCo also looked onsite for hazards (eg., fuels, lubri-cating oils, gases, liquid chemicals).
After CPCo did the onsite and offsite surveys it developed a list of concerns.
Based on this list CPCo performed a detailed hazard evaluation.
With the possible exception of the onsite natural gas (auxiliary heating for the tertiary steam evaporators) there were no explosive hazards.
Mr. Harshe explained that CPCo has there-fore concentrated its efforts at resolving toxic concerns.
The bulk of the toxic concerns are associated with the Dow Chemical ar:d Dow Corning plants.
Mr. Harshe indicated that Midland's primary protection against toxic substances from Dow lies in the communications from their normal response centers.
j Dedicated telephone lines connect Dow directly with the control room such that in the event of a problem, the plant can be notified.
Mr. Harshe said that backup radio communications between the facilities and the control rom also have been established.
He explained that a hazardous gas monitoring I
system (which is redundant and safety grade) monitors the inlet air to the control room.
This system monitors 24 gases that are known to be within five-miles of the plant and that could reach the control room in toxic con-centrations.
The hazardous gas monitoring system is capable of detecting i
j toxic gases and automatically isolating the control room such that the control room remains less than the toxic level. Mr. Harshe said that when l
MINUTES / MIDLAND, JUNE 2,1982 the control room is isolated it can b e pressurized for at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> by bottled air.
When the pressurization runs out air c within the low leakage control room an be recirculated apparatus is available for all of the indi idIf needed, self co (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> available).
uals in the control room v
The air intake to the control room is smoke and radiation detectors which can auto equipped with In addition, the emergency core coolingmatically isolate t room.
control room in the event of an automatic ECCS isystem (ECCS) isol nitiation.
outlined the habitability features of the Mr. Harshe next The emergency off-site facility (EOF) is lvarious emerg lities.
miles from the plant and has no special habit bi ocated 18 nautical ations support center (OSC) located on site lity features.
The oper-a features but has two fallback position, has no special habitability control room for meteorological reasons)s (in opposite directio (TSC) is shielded to the same conditions a The technical support ce The TSC has a radiation monitor in its ms the c doses.
Dr. Moeller asked CPCo what assurances it h anually controlled air intake.
heating and ventilation (HVAC) systems ad that its control room's would work.
program audits and its extensive test progr CPCo stated that its QA am will verify the performance of the HVAC systems.
The HVAC test program will be performed b (Eastern).
Dr. Moeller asked how the locations of Midl y a contractor E
air intakes were determined.
and's two control room based on judgement and experienceDr. Sullivan, CPCo, said that L
y much potential for toxic pollution of the nearbFinally, Dr. Moeller asked Chemical facilities.
y river or cooling pond by Dow i
He was told that catch basins preclude th chemicals to the river and that the cooling pond i e spill of Dow s monitored bLCPR&
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MINUTES / MIDLAND, JUNE 2,1982 DISCUSSION OF QUESTIONS REMAINING FROM THE MAY 20-21, 1982 SUBCOMMITTEE MEETING The Subcommittee discussed with the NRC Staff, the Applicant, and their consultants responses to questions raised by the ACRS Subcommittee on Midland Plant Units 1 and 2 during its May 20-21, 1982.
The Staff provided the Subcommittee with written "NRC Staff Responses To Questions By The ACRS Subcommittee During The Meeting Of May 20-21,1982 On Midland Plant, Units 1 and 2" (see meeting transcript or handouts).
In general the Subcommittee was satisfied with the Staff's responses.
Dr.
Moeller clarified that the probabilities of occurrence for reactor events and alerts, as expressed in the Staff's written response and in draft NUREG-0654, are "per reactor per year." Dr. Okrent questioned the Staff on the criteria it is using to resolve the turbine missile open item.
The Staff said that it was aviaiting receipt of a General Electric report for review.
The Applicant is using a turbine missile generation probability (on the order of ten to the minus nine per year) from this report.
The Staff, which used to assume that the probability of generating the missile was ten to the minus four per year, is now taking a close look at missile generation probabilities when it reviews an Applicant's calculations.
The Staff's is still using a damage criterion of ten to the minus seven per year.
The Applicant also responded to several questions asked during the May 20-21 Subcommittee meeting.
Mr. Harshe, CPCo, indicated that the two off-site power supply lines could not interfere with each other if one of the supporting towers was to fall over.
Mr. Slade, CPCo, described the Applicant's comprehensive program aimed at controlling occupational exposure. As part of this discussion, he identified those radioactive piping systems at Midland that have provisions to be drained and/or flushed (see slide in meeting transcript or handouts).
)
MINUTES / MIDLAND, JUNE 2,1982 DISCUSSION OF THE MIDLAND PORTION OF THE FULL COMMITTEE MEETING The Subcommittee chairman outlined some of the topics for discussion at the June 4,1982 ACRS Full Committee meeting. He said that topics to be discussed include:
Subcommittee chairman's report, Comments from members of the public,
- Summary of the status of the review by the NRC Staff.
- Quality Contr ol issues (whether the experience at Midland indicates a need for some broader review of the quality control).
- Midland's seismic design,
- Inadequate core cooling instrumentation, and
- Bolting Dr. Okrent then listed several topics that he would suggest that the full Committee might discuss.
This list included:
fire protection; systems interaction; integrated control system; high copper content welds in Unit I's pressure vessel; the process steam system; anticipated transients without scra n; AC/DC system reliability; probabilistic risk assessment; auxiliary feed-water reliability; organization, management, and training; e.nergency operating procedures; and turbine missiles.
PLANT SECURITY The Subcommittee went into closed session (Sunshine Act exemption 3) for approxi-mately 30 minutes to discuss the Applicant's physical security program and systems.
NOTE:
A complete transcript of the meet ng is on file at the NRC Pubite Document Room at 1717 H St. NW., Washington, D.C. or can be obtained from Alderson Reporters, 300 7th St. SW, Washington, D.C.
(202)554-2345 i
V 23062' rederal Register / Vol. 47. No.102 / Wednesday. Mry 26. 1982 / Notic:s amendment? !f so, what are the Company for a license to operate section 108 of the Copyright Act appears problems, and how could they best be Midland Plant Units 1 and 2.
in all prior notices requesting comments, resolved? If not,what changes should be In accordance with the procedures most recently at pages 79202 through outlined in the Federal Register on 79204 of the Federal Register for Friday, made in the law?
September 30,1981 (46 FR 47903), oral or November 28.1980 (Vol. 45. No. 231) and
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abihty oflibraries to reproduce phoncrecords and audiovisual works As,ociate Registerof copyrights.
persons regarding this review.
Further information regarding topics dealing with news In response to this Approved:
to be discussed.whether the meeting question describe any problems which Daniel l. Boorstin.
has been cancelled or rescheduled, the has e been encountered as the result of The Librian of Consress.
Chairman's ruling on requests for the the narrow er esemptions for nonprmt Int on a um mse a sm) opportunity to present oral statements materials under section 108.
- *
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_ and the time allotted therefor can be
- 6. liow has e the CO.VIU guidelines worked in practice? Ifow should obtained by a prepaid telephone call to the cognizant Designated Federal riod als more than the ) ears old be NUCLEAR REGULATORY Employ ee. Efr. David Fischer (telephone COMMISSION 20$MM Ween W am ad 7 Yhet is your opinion of the
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8 *. Advisory Committee on Reactor 5.00 p m.. EST.
relationship betw een section 107 (" fair use") and section 108 (" reproduction by4 Safeguards Subcommittee on Midiand t have determined. in accordance with hbraries and archives")?
Plant Units 1 and 2; Meeting Subsection 10(dj of the Federal 8 ffow should foreign copyrighted The ACRS Subcommittee on Efid!and Advisory Commfttee Act, that it may be necessary to close portions of this works ar<d requests from foreign Plant Units 1 and 2 will hold a meeting hbraries be treated under section 108 on June 2.1982. Room 1046,1717 H meeting to public attendance to protect and,in practice,how are they treated Street.NW Washington DC.ne proprietary information.He authority now?
Subcommittee will continue to review for such closure is Exemption (4) to 11 e
- 9. If problems do exist, can they be Sunshine Act. 5 U.S C. 552b(c)(4).
tesoh ed without resort to legislative the application by Consumers Power ATTAtut%t4T A Em' -
A1TACl4MEhm B JUNE 2, 1982 IE ING DATE: _
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\\.t TENTATIVE SCHEDULE FOR THE JJNE 2, 1982 ACRS SUBCOM.MITTEE MEETING ON MIDINO LNITS 1 & 2 ROCM 1046 (TENTATIVE) 1717 H ST. W, nASHIN3 TON, D.C.
Approximate Tine
'Ibpic Speaker 4: 00 p.m.
I.
CHAIR *.AN'S OPENING STATE *ENT D. Okrent 4:10 p.m.
II.
STATUS OF THE NRC STAFF REVIEW R. Hernan (Changes in o Mn ifats, If tensing conditions, etc. since the last Subcom-ittee meeting) 4: 20 p.m.
III.
CUESTIONS ON TrE NRC STAFFS SER D. Okrent 4: 50 p.m, IV.
ITEMS FRM PREVIOUS ACRS LETTERS D. Hood 5: 00 p.m.
V.
GENERAL TOPICS FOR DISCUSSION CPCo.
A.
Methods to Redre Com.on Cause Failure 1.
Systems interaction sttr$les 2.
Changes in design resulting from previous experience 5: 30 p.m.
B.
Integrated Control System R. M. Hamm 1.
What has been done to up-grade the ICS since 'IMI, changes broteht about by BrW vs. CPCo.?
l 2.
Failure modes & effect analysis 6: 00 p.m.
C.
Seismic and Envirowital Cualifi-J.J. Zabritski cation of Equipnent Im;ortant to Plant Safety 6: 20 p.m.
D.
DiR System Operation L.S. Gibson 6: 30 p.m.
E.
Bolting and Other High Strength H. W. Slager
.v terial a
ATTAUI ME NT C.
1 c-1
TEdTATIVE SCHEJULE
- 'IDLAdb UdITS 1 & 2.
1 Approx. Time Topics Spea,k,el 6:40 p.m.
F.
Fire Protection R. A. Polich 1.
Potential problems for spurious actuations, flooding or wetting, deper actuation 6: 55 p.m.
G.
Habitability B. L.*hrfu 1.
Control room 2.
Dnergency respanse fa:Ilities 7:15 p.m.
H.
Other 7: 30 p.m.
VI.
DISCUSSION OF QUESTIONS RD4AINING fry.
NRC Staff /CPCo.
>%Y 20-21 SUS 7#.ITTEE MEETING A.
Questions on the Draft Envirort-Inental Statement 1.
Trititan activity in the core.
2.
Consistency of event probabi-lities with experience.
j 3.
Staff Position on overfill Pro-tection for BW SGs.
C.
Others 7:45 p.m.
VII.
INDUS'IRIAL SECURITY CPCo.
8:15 p.m.
VI II. DISCUSSION OF Td! MIDLA.C PORTION OF D. Okrent
'IHE FULL CCMITTEE MEETING 8: 30 p.m.
ANOURN f2 -Si
4 ATTACHMENT D REFERENCE DOCUMENTS FOR THE JUNE 2,1982 ACRS SUBCOMMITTEE ON RIDLAND PLKNT UNITS 1 & 2 1.
Status of open items and licensing conditions relating to the operation of Midland Plant, Units 1 and 2 as of May 28, 1982 (same as that provided by the NRC Staff to the ACRS Subcommittee on Midland Plant Units 1 and 2 on May 20,1982. See handouts from May 20,1982 meeting for copy.)
2.
NRC Staff response to concerns expressed in previous ACRS reports on Midland Plant Units 1 and 2.
(see meeting handouts).
DOCUMENTS AND SLIDES PROVIDED AT THE MEETING 1.
NRC Staff responses to questions by the ACRS Subcommittee during meeting of May 20-21,1982 on Midland Plant, Units 1 and 2, dated June 2, 1982.
2.
NRC Staff response to concerns expressed in previous ACRS reports on Midland Plant Wn its 1 & 2 3.
Slides used by Mr. B. Harshe, CPCo, Methods to Reduce Common Cause Failures (6 sli des) 4.
Slides used by Mr. R. Hamm. CPCo, Integrated Control System (4 slides) 5.
Slides used by Mr. J. Zabritski, CPCo, Seismic and Environmental Qualification of Equipment Important to Plant Safety (5 slides).
6.
Slides used by Mr. L. Gibson, CPCo, Decay Heat Removal System Operation (3 slides) 7.
Slides used by H. Slager, CPCc,"Jaliing and Other High Strength Material (2 slides) 8.
Slide used by Mr. R. Polich, CPCo, Fire Protection (1 slide) l 9.
Slides used by Mr. B. Harshe, CPCo, Control Room Habitability (3 slides) l
- 10. Slides used by Mr. G. Slade, CPCo, Provisions-for Draining and Flushing of Radioactive Systems (1 slide).
l l
l