ML20027C927
| ML20027C927 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 08/16/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1991, NUDOCS 8210270433 | |
| Download: ML20027C927 (78) | |
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NUTES OF THE ACRS SUBCOMMITTEE ON MIDLAND PLANT UNITS 1 & 2 MIDLAND, MICHIGAN MAY 20 & 21, 1982 The ACRS Subcommittee on Midland Plant Units 1 and 2 held a meeting at the Holiday Inn,1500 W. Wackerly Road, Midland, MI. 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.
The meeting was entirely open to the public. Notice of this meeting was published in the Federal Register on Tuesday, May 4,1982.
A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B.
The sched-ule for the meeting is included as Attachment C, and a list of all reference material (including slides and documents provided to the Subcommittee at the meeting) is included as Attachment D.
A complete set of bandouts has been included in the ACRS Files. There was one oral statement made by a member of the public.
Three written statements, Attachment E, were received from members of the public (Dr. Charles Anderson, Ms. Mary Sinclair, and l
Ms.BarbareStamiris).
The Designated Federal Employee for this meeting was Mr. David C. Fischer.
SUBCOMMITTEE CHAIRMAN'S OPENING REMARKS Dr. Okrent opened the meeting with a statement on the purpose and goal of the meeting.
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o STAFF AND APPLICANT INTRODUCION AND OVERVIEW Mr. Hernan (NRR/DL) provided the Subcommittee with a brief history of the oot y
Midland operating license (OL) review. He discussed each of the 16 remain-o-
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ing open items.
The next action required to close out each open item was cn a.
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MIDLAND 1 & 2 5/20&21/82 identified along with a projected completion date for tnat action. The license conditions being imposed on the Applicant were similarly discussed.
Significant items wnicn remain unresolved include:
the potential effects of using natural gas onsite for auxiliary heating of tertiary steam in evaporators.
ongoing soils remedial actions, the need for a reactor vessel head vent, and turbine missiles.
Dr. Okrent questioned wnetner probabilistic criteria were being used to resolve the turbine missile open item.
lne Staff said tnat probability is an important factor in tne resolution of this issue but could not state its criteria. Dr. Okrent asked tne Staff to clarify the criteria it will use to resolve the turbine missile issue by the next ACRS meeting on Midland.
Dr. Okrent also asked if the Staff was using any special approacn in its review of the Midland emergency preparedness plan. The Staff acknowledged the proximity of the Midland plant to industrial and residential areas then said that the criteria would be the same but the implementation would have to account for the unique site features at Midland. Significant license conditions being imposed on the Applicant relate to:
the adequacy of the control plan for monitoring the water table establisned by the permanent site dewatering system.
the final design drawings and plans for a system to monitor inadequate core cooling (Mr. Tedesco indicated that a Staff position on what constitutes adequate, inadequate core cooling instrumentation is expected in the late sunr.er of 1982.).
the need to nave experienced (from the standpoint of operating B&W reactors) people on shift during the initial testing and start-up phase,
e MIDLAND 1 & 2 5/20&21/82 the development of emergency operating procedures (ATOG program),and the development of an adequate small break LOCA model.
During the discussion of the various license conditions Mr. Tedesco in-dicated that there is a safety-grade level control system on the Midland steam generators which protects against steam generator over-fill from
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both the main and auxiliary feedwater systems.
Dr. Okrent next asked the Staff to identify those safety issues which were the most difficult to resolve for Midland.
That is, issues for which the Staff position on what was adequate was not easy to find.
In response to this request the Staff identified:
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- the need for a third auxiliary feedwater pump to improve system reliability, the need for motor operators on both of the decay heat removal pump suction valves, the various soils issues, and the seismic resolution for the Midland site.
The first two of these issues have been resolved via licensee appeal to Dr. R. Mattson (D/DSI).
The first was resolved 50 that a third auxiliary feedwater pump will be installed at Midland.
The second issue was resolved so that motor operators are not required on their decay heat removal pump suction valves.
The basis for these Staff decisions relied heavily on
" engineering judgment" as opposed to quantitative analyses.
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MIDLAND 1 & 2 5/20&21/82 Dr. T. Sullivan, CPCo, indicated that the Staff did have a damage criteria related to turbine missiles in the range of ten to the minus seven per year for damage to safety-related equipment. He indicated tnat the area of controversy centered around the two times ten to the minus eighth probability of failure of the low pressure turbine disk which was assumed in a GE tur-bine analysis and used by CPCo in their analysis.
Dr. Okrent asked several questions related to his review of the Staff's SER on Midland. He questioned whether a thorough systematic study of internal floooing from all possible sources was made for the Midland plant.
NRC REGION III COMMENTS ON MIDLAND Mr. W. Little from the NRC's Region III Staff described significant con-struction QA and QC experiences at the Midland plant. He also made an assessment of CPCo's construction management.
Problems in the following areas which occurred prior to December 1979 were mentioned.
placement, sampling, and testing of concrete drawing control and procedures for control of design and procurement activities training, procedures, and inspections of cadwell activities QA program implementation relative to embedment containment tendon sheath omissions plant soil, foundations, and settlement He said that for eacn identified problem the Applicant was responsive in taking corrective action and in upgrading the QA/QC Program. Mr. Little identified two common underlyino causes for tnese deficiencies:
MIDLAND 1 & 2 5/20&21/82 Consumers Power Company tended to over-rely on Bechtel.
There was insensitivity on both the part of Bechtel and Consumers Power to recognize the significance of isolated event (i.e., failure to adequately evaluate possible generic application of these events).
Mr. Little identified a couple of quality-related problems which have been identified it Midland since December 1979.
One involved deficient anchor bolts for the reactor pressure vessel and a second involved procurement and installa-tion deficiencies with the. heating, ventilation,and air, conditioning ducts and supports within the plant.
These deficiencies led to an NRC inspection, an agreed upon work stoppage, and a civil penalty against CPCo. The NRC found that neither of these problems appeared to be indicative of a brcader breakdown in Midland's overall Quality Assurance Program. Mr. Little described several special NRC investigations which were conducted at Midland between December of 1980 and May 1981. He said that these special inspections revealed that:
the Licensee's corrective actions relative to the implementation of the soils remedial measures had been completed.
Consumers Power Company and Bechtel Corporation had reorganized to form an effectively integrated and coordinated construction and quality management team at the site.
the quality-related problems were generally isolated or limited to a specific area and not indicative of major programmatic weak-nesses in the implementation of their program.
Mr. Little indicated that these quality assurance audits were detailed and, thorough.
He said that the region's most experienced inspectors were used.
Dr. Okrent questioned the Staff on how it decides that it is appropriate to ascertainthingsingreaterdepthrelativekoQA. He asked specifically how this site
MIDLAND 1 & 2 5/2D&21/82 mignt be evaluated differently in light of the population within several miles of the plant.
He also wanted to know what kind of a prior record, on a technical basis, suggests that a more in-depth look be given at plant quality.
Tne Staff indicated that they evaluate those design activities which have historically been problems and tnat they cannot audit everytning at the plant site in detail.
They expect the Itcensee and applicants to conduct detailed reviews through their own Quality Assurance Program. The Staff indicated that CPCo's QA record to date nas been iess tnan average. Tney said that for sensitive sites such as Midland there is above-normal manage-ment participation in overviewing tne plants design and construction quality. While the Staff did mention tnat a probabilistic risk assessment I
(PRA) is being performed for Midland they acknowledged tnat they are not requiring that the PRA be done even though the population around Midland is large.
Next, Dr. Siess and Mr. Little discussed generically the purpose of a QA program and how the NRC Staff goes about measuring the effectiveness of a QA program. The Staf f, at the request of Dr. Moeller and Dr. Zudans, explained the effect of past and recent Atomic Safety and Licensing Board orders related to Midland remedial soils activities. Continuing with his statement, Mr. Little identified three areas of concern during the past year.
For each, he ide,ntified the specific problem and gave a status of the Staff and Licensee actions related to it. The areas discussed included:
remedial soils activities.
non-qualified cable pulling inspectors, and piping suspension system installation. problems.
MIDLAND 1 & 2 5/20821/82 He said that Region III is reevaluating the actions that the Licensee should take in each of these areas.
In his concluding comments, Mr. Little stated that Region III believes that the Midland construction management is staffed with competent people and that a program does exist such that the plant can be completed in accordance with design and regulatory requirements.
Dr. Okrent questioned the Staff on the need for detailed audits to assure that plant quality is adequate.
Mr. Tedesco said that the Staff is asking all near-term operating license (NTOL) applicants to provide the Staff with an evaluation of why they believe that their plant has been designed and built in accordance with their application. He added that this would involve a thorough look at their whole QA program and the experiences that I
they have had during constructien.
The Staff has not yet determined the scope of the design re-verification recuirement and whether it needs to be i
done by an independent organization. Mr. Tedesco said that the program for conducting design re-verification for 1982 NTOL plants is being developed entirely by the NRC Staff.
He said that a 1983 NTOL plant (e.g., Midland) may involve an INPO-developed program.
Recognizing that an independent design re-verification may be required, CPCo is currently having their architect-engineer, Bechtel, perform a design re-verification.
STATEMENT BY DR. CHARLES ANDERSON Ms. Mary Sinclair introduced Dr. Charles Anderson, "a highly competent expert witness and consultant". Dr. Anderson discussed cracks in the Midland Plant diesel generator building and service water pump structure.
He stated that he had toured five buildings'in a one and one-half hour I
MIDLAND 1 & 2 5/20&21/82 period the previous day and that he had not had ample opportunity to review the file. Using cardboard boxes, he demonstrated how a structure might lose its rigidity. He stated that the buildings have numerous, long cracks in random orientation running through the thickness of the buildings' walls.
He indicated that the buildings do not have their designed structural integrity because of these cracks.
According to Dr. Anderson the buildings no longer have and cannot be made to have a reasonable, structural integrity with a calculated factor of safety. He characterized the structures as pieces of concrete between cracks that are now fitted as blocks between other pieces and held together with a fabric of reinforcing steel. CPCo stated that they have evaluated the structures with and without the cracks, accounting for future settlement (short-and long-term settlement), and have found that neither structure's integrity is effected.
STATEMENT BY MS. BARBARA STAMIRIS Ms. Barbara Stamiris stated that she was shocked and appalled by the previous QA presentation by the Staff.
She asked that the ACRS be provided with a copy of NRC's Office of Inspection and Enforcement (I&E) Inspection Report 81-12 and the 1980 and 1981 Systematic Assessment of Licensee Performance (SALP) Reports. She said that CPCo did identify its own problems but was slow to correct those problems. She indicated that the generic implications of problems identified were not addressed (i.e.,
problems were looked at as isolated events).
She also said that the Staff lacked criteria for judging an applicants QA program. Consequently, outside observers have no way of determining the basis for a Staff judgment
MIDLAND 1 & 2 5/20&21/82 on the adequacy of an applciant's QA program. Dr. Okrent indicated that the generic implication of QA problems at Midland would be a topic of discussion at the ACRS full Committee meeting.
STATEMENT BY MS. MARY SINCLAIR Ms. Mary Sinclair summarized the written statment she provided to the Sub-committee.
Here written statement is included in Attachment E to the Minutes (see p. E-4 through E-6).
- L. Sinclair encouraged the Subcommittee to look at the cracks in the diesel generator building and service water 4 amp structure. Topics mentioned by Ms. Sinclair in her summary statement include: the leakage rate at Midland, the radiation effe, cts of the plant (particularly when replacing a steam generator), and the potential for passing radioactive materitl to the Dow Chemical Comcany via the process steam system.
INTRODUCTION BY THE APPLICANT At the suggestion of Mr. Cook, CPCo the Subcommittee did not hear a general presentations on the plant and site description, construction schedule, and projected schedule to commercial operations.
HUMAN FACTORS REVIEW 0F THE CONTROL ROOM Mr. R. Hamm, CPCo, briefly described some of the ongoing evaluations CPCo is doing in the area of control room design and human factors. He said that Midland's Bechtel-built control room has been evaluated by CPCo operations staff and by a human factors consultant. He described the preliminary control room design review which was performed.
This preliminary review
a MIDLAND 1 & 2 5/20&21/82 involved use of a control room mock-up, operator interviews, a checklist-type human factors review of components, and walk-throughs of emergency and routine operations on the mock-up by operators.
Problems identified during this review include:
th3re was no method for operators to determine how various functional groups were related; labels on panels were too small to read and contained too much information; alarms were not prioritized.
A control room enhancement program was subsequently initiated at Midland.
CPCo is looking at demarcation of the functional groups to better show the interrelationship between controls and indications.
CPCo is also looking at the use of mimics and the use of hierarchical labeling to improve the read-ability of labels.
Mr. Hamm described CPCo's ongoing task ana;ysis cf co1-trol room operators' functions.
The company is evaluating the equipment, l
l information, and personnel requirements of the control room.
He said that this evaluation would be based on the routine, off-normal, and emergency operations of the plant.
He related this to other human factors related l
l activities at Midland (e.g., control panel enhancements, abnormal transients operating guidelines (ATOG), safety parameter display systems, plant com-l l
puter upgrade, and shift manning).
Dr. Okrent questioned how the prelimi-l nary review finding of alarms not being prioritized was being handled.
He also asked the Staff if it was doing anything in this area.
CPCo stated that this review deficiency is still being evaluated. Mr. Hamm described several things which CPCo might do in this regard. Mr. Sullivan, CPCo said that the event sequence diagrams developed for the PRA have been fed into the AT0G program and will thus be incorporated into the overall task analysis and alarm prioritization decisions.
MIDLAND 1 & 2 5/20821/82 AUXILIARY SHUTDOWN PANEL Mr. Hamm described CPCo's auxiliary shutdown panel. This panel provides those controls and indicators necessary to maintain the plant in hot standby.
It also provides some of the instrumentation and controls needed to bring the plant to cold shutdown. Midland's auxiliary shutdown panel is safety grade with the exception of the pressurizer spray and the letdown controls.
Mr. Hamm showed where the auxiliary shutdown panels are. located in the plant and listed the controls and indication available at the panel. He said that there are two indicators and controls for both trains of components that are there. He indicated that the panel is designed so that one of these trains would still be operable in the event of a fire in the control rocm.
INSTRUMENTATION TO DETECT INADEQUATE CORE C00 LINT Mr. Hamm stated that CPCo would use a subcooling monitoring system, a hot i
leg level monitoring system (HLLMS), and 24 safety-grade, core-exit thermo-couples to detect inadequate core cooling.
He described the HLLMS config-uration. Mr. Hamm also explained how steam voiding could be detected both when the voids are only in the head region and when they extend beyond the i
head region. His major point was that a void in the reactor coolant system would not result in a loss of natural circulation. Mr. Hamm explained that the highpoint vents for the Midland design will be off the top of the i
hot leg. No vent is planned for the reactor head. Gasses which do not condense as they bubble from the upper reactor head region to the top of the
MIDLAND 1 & 2 5/20821/82 hot leg could there be vented. Natural circulation allegedly would not be effected. The current Staff position is that Midland should have a head vent.
Mr. Gibson, CPCo, mentioned several methods for reducing the concen-trations of condensable and non-condensable gasses in the reactor coolant system. Mr. Hamm said that if a steam void extended beyond the head region it would immediately be detected by the HLLMS.
(The region from the top of the head down to the hot le,g where the bubble would start venting is approxi-mately 1400 cubic feet, approximately 12 percent of the total volume of the reactor coolant system.) To further detect the possibility of inadequate core cooling, CPCo has developed correlations of the co're exit thermocouples to the fuel clad temperature.
Mr. Davis questioned why none of the instru-l mentation to detect inadequate core cooling is on the auxiliary shutdown panel.
Dr. Okrent asked the Staff to address the adequacy of proposed methods to detect inadequate core cooling at the next full Committee meeting.
AC/DC POWER SYSTEM RELIABILITY AND STATION BLACK 0UT Dr. Okrent asked that the Applicant to prepare a discussion of those system functions lost (as a function of time) following a complete station black-out. He also asked the Applicant to compare its DC power system with that analyzed by the Staff in NRUEG-0666.
Dr. Okrent indicatd that these items would be discussed late in the meeting.
PROCESS STEAM SYSTEM Mr. J. Alderink, CPCo, outlined the basic system functions of the evaporator system and gave an overview of its operations.
The system will provide both 4
MIDLAND 1 & 2 5/20821/82 high and low pressure steam to Dow (four hundred thousand pounds / hour at 600 psi and 3.65 million pounds / hour at 180 psi).
The high pressure steam will come off inter-tied piping between either unit's main steam system. Low pressure steam comes either from the same inter-tied piping through a pressure reducing valve or from an extraction point on Unit l's high pressure turbine.
There are 11 low pressure evaporators,10 are required to produce rated low pressure 4
i process steam flow. There.are two high pressure evaporotors and only one is required to produce rated steam flow.
Mr. Alderink indicated that the pro-l cess steam system has essentially doubled the amount of steam piping in l
Midland's secondary system.
Mr. Mathis asked the Applicant if an analysis had been conducted to determine the potential effect of the tertiary system on power plants' primary and secondary systems.
The Applicant responded that the worst case single load that can be added on the tertiary side in a single activity equates to approximately three percent reactor power.
Mr. Alderink described the three control rooms which will be controlling the process steam system (i.e., the power plant main control room, the evaporator control room, and the Dow Management Control Center). The information flows and interaction between control rooms / centers to coordinate process steam system operators was discussed. Mr. D. Summers next outlined CPCo's process steam radioactivity monitoring program.
He said that the system evolved following a rather extensive analytical approach which included multiple calculations considering varying leak sizes (primary to secondary and secondary totertiary).
The Technical Specification limit of 25 gpm secondary to tertiary leakage is used as the system's design basis. The response times of the tertiary system are such that you would not anticipate seeing radioparticulate
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f MIDLAND 1 & 2 5/20821/82 saturation, even with a secondary to tertiary leak, for approximately 50 to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. He said that the tertiary process steam system will impose no special restrictions on primary coolant activity or primary to secondary leakage rate. Tne process steam monitoring program involves both on-line and off-line monitoring.
Tne on-line monitoring checks for gross gamma.
It has seven radiation monitors, three of which monitor the process steam on a continuous basis. There are spare monitors for dedicated confirmation and reliability. Tne of f-line monitoring checks for gross gamma, gross beta, and t ri ti um.
The off-line program is used to confirm on-line monitoring. A dual f requency monitoring schedule is employed, the frequency of off-line sampling being based on seconcary radioactivity. The radioactive sample action / alert levels were mentioned.
SITE TOUR t
The Subcommittee meeting recessed to tour the Midland Plant site. Tne l
meeting was reconvended at 6:00 p.m.
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l SEISMIC ANALYSIS OF MIDLAND Dr. Okrent opened the seismic discussion by stating tnat he is not only l
interested in what the safety shutdown earthquake should be for design I
purposes; he is also interested in the level of safety tnat is being l
achieved and whether this level of safety is adequate. He mentioned that l
the Committee has demonstrated an interest in assuring itself that there is an acceptably low level of risk introduced by the seismic component of total risk.
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MIDLAND 1 & 2 5/20&21/82 Consumers Power Company Overview:
Dr. T. Thiruvengadam presented a brief overview of the criteria to which plant structures and equipment were built.
He said that these criteria were established during the construction permit stage.
Dr. Thiruvengadam explained that the plant site lies in the geo-logical entity referred to as the Michigan Basin. CPCo considered this to be a tectonic province. Within the Michigan Basin and within 200 miles of the site the highest intensity earthquakes which have occurred historically
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have been epicentral intensity 6.
A conservative horizo'ntal ground motion acceleration of 0.12g was chosen to correspond to this intensity. The horizontal design spectra originally proposed was a Housner spectra modified in the period range of 0.2 seconds to 0.6 seconds (1.57 to S hertz). The Housner spectra was modified to compensate for differences between the Housner and the Newmark (Reg. Guide) spectra. [This frequency range is where the fundamental frequencies of the structures at Midland was expected to be.] The design spectra is applied at the various foundation eleva-tions of the seismic Category 1 structures. The dynamic analysis of seismic Category 1 structures are performed by developing mathematical models consisting of lumped masses and analyzed by a modal supposition method. The soil-structure interaction is accounted for by the use of impedence springs and dampers at the structure base, derived on the basis of elastic half-space theory. Dr. Thiruvengadam said that all of the Category I structures and systems at Midland have been designed to these criteria.
He said that due to the proposed foundation modifications and added improvements, these structures are being reanalyzed and reevaluated for seismic loadings. When the underpinning was being designed, the Sta'ff indicated that they did not
MIDLAND 1 & 2 5/20821/82 agree that the Michigan Basin was a tectonic province.
The Staff proposed two alternative solutions; CPCo could use either a 0.199 acceleration and a Reg. Guide 1.60 spectra or it could develop a site-specific response spectra from historical earthquake information.
The site-specific response spectra would be required to meet the following criteria.
It should have:
a magnitude in the range of 5.3 plus or minus 0.5 mbLg an epicentral distance less than 25 kilometers at soil sites, its spectra at the 84th percentile of the response spectra derived directly from real time histories, and its spectra applicable to the original ground surface (apprcximately elevation 600)
In addition, the Staff requested that the Applicant assess the soil amplifi-cation for structures founded near the top of the fill (elevation 634).
Dr. Thiruvengadam said that CPCo elected to develop two separate site-specific response spectra, one for the original ground surface and another for the top of the fill.
These spectra are going to be applied to buildings founded on top of the fill.
Dr.
Thiruvengadam indicated that CPCo has performed a probabilistic seismic hazard analyses.
He said that CPCo has reached agree-ment with the Staff to use the site-specific response spectra to perform a seismic margin review of plant structures, systems, and components that are required for safe shutdown.
The Staff has also asked CPCo to use the site-specific response spectra as a design basis for the soils remedial work (e.g., for the auxiliary building and service water pump underpinning and for the borated water storage tank foundation modifications).
MIDLAND 1 & 2 5/20821/82 Consumers Power Company on Seismic Input:
Mr. R. Holt of Weston Geophysical Corporation discussed Midland's site-specific response spectra.
He first outlined two different approaches to seismic design, one which results in a standard response spectra and the another which results in a site-specific spectra.
Both start with a thorough geological and seismological analysis of the region and site.
By the standard approach, the safe shutdown earth-quake is selected on the basis of intensity.
A peak hor 1zontal ground acceleration is then selected using some relationship of intensity to peak acceleration.
Finally, a standard spectral shape is set to that zero period acceleration.
This spectral shape is independent of the site.
(TheReg.
Guide 1.60 shape was developed from a number of magnitudes, ranging from about magnitude 5 to about magnitude 7; from a number of distances, ranging from zero to about 200 kilometers; and for a number of site conditions, ranging from poor soil to hard bedrock.) Mr. Holt explained that by the site-specific approach the SSE is based on magnitude (as opposed to intensity). He said that the advantage here is that the magnitude is independent of local geologic conditions.
Next the earthquake is defined with respect to magnitude, with respect to distance, and with respect to foundation conditions. Basically, a shear wave velocity profile is defined from the worldwide strong motion data set. A site-specific spectra can then be constructed from these accelerograms.
The spectra are based on some percentage (e.g., 50th or 84th percentile) of the strong motion data set.
Mr. Holt explained that there are three conditions under which one selects a safe shutdown earthquake: a capable fault, a tectonic province, and a tectonic structure. He said that the Midland site was selected on the basis of a tectonic province.
This criteria calls for
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1 MIDLAND 1 & 2 5/20&21/82 moving the largest earthquake to the site. The differences between the Staff and the Applicant related to the site-specific spectra hinge on the Applicant's use of the Michigan Basin as a tectonic province.
Use of the Michigan Basin determines whether certain earthquake records from the strong motion data set are or are not used to define the SSE (Midland used 22 sets of records in developing its site-specific spectra, the average magnitude was 5.3, and an average epicentral distance of 17.6 kilometers.)
Mr. Klimkiewicz, a seismologist at Weston Geophysical Corp., presented the results of a seismic hazard analysis (i.e., a determination of the probability of occurrence of the spectra) performed for the Midland P.lant. He first described the seismic hazard methodology. He listed the input requirements for the computer code used to calculate ground motion at the site (i.e., seismic source geometry, recurrence frequency of earthquakes in each source, and the ground motion attenuation model appropriate for the region).
He said that i
earthquake occurrence can be modelled as a poisson process. The analytical techique he described involved the solution of a " total probability theorem" integral. Mr. Klimkiewicz said that three alternative tectonic models were examined in calculating probabilities at the Midland site. One model used the Michigan Basin as a tectonic province, a second used the central stable region, and a third used a broad definition of the central stable region. At a given magnitude, the rates of seismic activity are approximately an order of magnitude lower in the Michigan Basin versus the interior arch structure.
At higher intensi-ties, the predominant source of seismic hazard at the site is the local occurrence of a mod.erate-site earthquake.
MIDLAND 1 & 2 5/20&21/82 NRC Staff on Seismic Input: Mr. J. Kimball (NRR/DE/SEB) discussed the issues that went into tne seismic portion of the operating license review for Midland.
He compared the Reg. Guide 1.60 spectra anchored at 0.12g with the Midland original design spectra and the Midland site-specific spectra.
He did this for botn tne top of the original soil and also for the top of the fill. He said that the Staff concluded that the original zero period acceleration (0.12;) and spectra (modified Housner) for Midland might not be an adequate representation of tne SSE. The Staff considered tne fact tnat tne Applicant had not used the Reg. Guide 1.60 response spectrum with a 0.129 peak acceleration. The Staff looked at more recent sites in the central United States and found tnat there have been a range of peak accelerations essentialy between
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0.12 and 0.199 The Staff also reviewed tne Paleozoic geology of the Michigan Basin as compared to the central stable region and concluded tnat the Michigan Basin co'uld not be separated out from the rest of the central stable region. Consequently, tne Staff gave the Applicant the choice of using either 1
a Reg. Guide 1.60 spectra with an 0.199 zero period acceleration or develop a site-specific spectra. Mr. Kimball said that the Staff also suggested that the Applicant pursue tne seismic nazard analysis, mainly because of a quick observation that around the Midland Plant there have been very few earthquakes (particularly within 200 miles of the plant site). Mr. Kimball pointed out that seismicity is not uniform in tne central stable region. He said that the Staff tries to take this into account as part of its review. After reviewing the Applicant's proposal and the strong motion data base, the Staff requested that CPCo perform a set of sensitivity tests on the site-specific spectra.
The Staff requested that CPCo include the sensitivity of additional records.
Inclusion of the Parkfield eartnquake records (3 seperate sets of records) had a significant effect on the site-specific spectre. While the Applicant felt that
MIDLAND 1 & 2 5/20&21/82 these records were inappropriate for use with a central U.S. site-specific spectra, the Staff took the position that they should be included. Mr.
Kimball explained that the Applicant was looking at the seismic nazards around tne Midland site in an attempt to quantify the low seismicity near Midland.
He said that tne Staff next asked the Applicant to run all the studies for Midland along with five other sites in tne central U.S.
Tne Staff looked at the intensity at different return perio'ds for Midland and compared them to those determined for the otner five sites. Tne results showed tne Staff that Midland is about 0.5 to 0.7 intensity units lower than tne otner five sites (return periods of one in 1,000 and one in 10,000 were used; the difference between Midland and the other five sites were then simply ratioed).
Tne intensity difference translates to about 0.25 to 0.35 magnitude units. Magitude 5.3 had been set up as tne target magni-tude for Midland.
So, the Staff used tne relative hazard to conclude that a 5.0 would have been a conservative target magnitude for tne site-specific spect ra. Mr. Kimball said tnat since this conservatism is larger in magni-tude than the effect of excluding the Parkfield records, the Staff has found i
l CPCo's site-specific spectra acceptable. The Staff did review computer analyses to ensure that the fill material would not amplify the seismic effect. Mr. Kimball showed the Subcommittee some of the results of these l
computer analyses (SHAKE RUNS).
MIDLAND 1 & 2 5/20821/82 Mr. L. Reiter (NRR/DE/GSB) discussed the use of probabilistic estimates to determine seismic hazards.
He mentioned several areas where probabilistic estimates are currently being used. These include:
defining the operating basis earthquake (OBE) determining the return period associated with the safe shutdown earthquake (SSE).
(Many studies show return periods on the order of 1,000 or 10,000 years.)
assessing the relative hazard associated with different levels of ground motion at the same site (e.g., Sequoyah review) calculating equivalent levels of hazard for one plant relative to other sites (e.g., systematic evaluation program reviews) placing deterministic estimates of fault offset in probabilistic perspective (e.g., determining if and how much a fault might rupture near the General Electric test reactor) helping to make a rational seismic decision for a
" tectonic province" that covers one-third or more of the U.S. and has widely varying levels of seismic activity (e.g., Midland OL review)
Mr. Reiter said that for each of these areas the Staff has emphasized relative rather than absolute levels of probability.
He indicated that the Staff has found relative estimates far more stable than absolute estimates and that these estimates are used to gain insight rather than using them directly.
Mr. Reiter mentioned two sensitivity studies related to seismic hazards: the Lawrence Livermore Laboratory TARA study and one conducted by Dr. D. Perkins.
He outlined the current approach for doing a seismic risk or hazard estimate (i.e., a method to determine the probability of exceedence/non-exceedence of 4
various acceleration levels).
In so doing, he showed some of the controlling
MIDLAND 1 & 2 5/20821/82 parameters and how these probabilistic studies may be applied to different return periods. He specifically addressed the effects of variations in input parameters and the resulting uncertainty in zonation, seismicity, upper magnitude cutoff, and ground motion. The Livermore study concluded that:
parametric variation had increasing effect as return period increased, with an extrapolation beyond the exisitng data you get very large differences from sm:ll variations in input parameters, and it is very important not to just look at the parameters by themselves but to see how they interact with each other. Mr. Reiter briefly discussed Dr. Perkins' study "Effect of Changing Return Periods on Probabilistic Ground Motion." He described some rough rules-of-thumbs proposed by Dr. Perkins.
Uncertainties in ground motion increases ground motion acceleration by 25-40% at short periods (hundreds of years or less) and by 100% at long periods (thousands of years or more);
To get equivalent changes at long periods would require l
larger changes in input parameters (tenfold increase in level of seismicity, twofold increase in relative size distributions, or an increase in magnitude from 5.0 to 8.5);
At long return periods uncertainty in ground motion swamps reasonable changes in seismicity parameters; For very long return periods the most important aspect in understanding attenuation is the configuration of the upper tail of the magnitude of the peak accelera-tion distribution (i.e., the type and shape of the distribution.
After providing the Subcommittee with some words of caution r3garding astimation of hazards, Mr. Reiter outlined the Staff's present course of action related to probabilistic estimates.
He said that the Staff recognizes
MIDLAND 1 & 2 5/20821/82 that probabilistic estimtes are powerful tools. The Staff will proceed slowly while expanding its use of probabilistics. The Staff will use probability to obtain relative as opposed to absolute insights into seismic hazards. Mr. Reiter said that reliance upon probabilistic estimates for very long return periods is not the way to alleviate concerns about earthquakes greater than the SSE.
He did, however, encourage research to facilitate increasing use of probabi-listic estimates.
Dr. Okrent expressed his interest in knowing what the Staff considered to be an acceptable level of risk due to earthquakes. He also commented that if what Dr. Perkins says is reasonably valid (that, at probabilities of 10 to 10-5 per year, it may be hard to judge too much from the last 100, 200 or 300 years of history) then the Staff and Applicant may be drawing more com-I fort than they should be from the greater or lesser seismicity in the regions of interest.
Mr. Holt of Weston Geophysical pointed out that earthquakes have a geologic cause called faulting. He said that to ignore the geologic cause and to go in somewhere and make a count and start statistical processes nay not be well based. He suggested that more attention should be given to the struc-tures, arches, etc., in the immediate area of interest. Dr. Trifunac questioned the Applicant's rationale for not including the Parkfield earthquake records in the development of the Midland site-specific spectra.
Dr. Okrent asked each of the seismologist who participated in the seismic '
l discussion to give an expected value for the earthquake acceleration (tied to Reg. Guide 1.60 spectra) which has a probability of exceedance of one in 1,000, one in 10,000, and one in 100,000 per year. He asked that the seismologist give answers which they have a 90 percent confidence in.
e MIDLAND 1 & 2 5/20821/82 Consumers Power Company on the' Seismic' Analysis of Structures and Equipmenti Dr. R. Kennedy, President of Structural Mechanics Associates (SMA) and con-sultant to CPCo, discussed the seismic reevaluation of the Midland facility.
He discussed the criteria that are being used in the seismic margin review and gave a sample of some of the preliminary results from that review. The seismic margin review is trying to determine if there are margins against code allowable stresses.
The seismic margin earthquake being used in the seismic margin review is based on a site-specific spectra.
The review includes both structures and equipment. A screening process will be used to select the structural elements, equipment, and systems to be reviewed.
In that the a
seismic margin review will allow for some deviation frc 'tha standard review plan, the review is being treated as a review and not as a new design criteria.
l Nevertheless, the review process is following formats and procedures used on I
other operating licenses and on the older SEP plants. Dr. Kennedy highlighted some of the differences between the seismic margin review and the FSAR design.
Differences exist in the areas of: seismic input, range of soil par:metsrs, parametric variations of relative soil stiffness under the auxiliary building electrical penetration areas, and in damping.
Dr. Kennedy elaborated on each of these differences.
The seismic margin review is being conducted using either the site-specific spectra or the 0.12g Housner spectra, whichever is more restrictive for the frequency (period) being evalurted. The following struc-tures ara being reviewed:
reactor building, control and auxiliary building, service water pump structure, diesel generator building, and borated water storage tank. Dr. Kennedy briefly described how the structural evaluations
MIDLAND 1 & 2 5/20&21/82 would be conducted. He indicated snat the dynamic models developed by Bechtel for the FSAR's structural analysis would be used in most cases. These models nave incorporated, where appropriate, the soil remedial work on the structures.
Dr. Kennedy said tnat SMA is independently generating new structural loads and new response spectra throughout tnese buildings for the seismic margin earthquake.
Seismic margins against code strengths will be calculated for a selected list of structurat elements and equipment due to these new loads.
Wherever the code margin comes out less than unity, a conservative failure analysis will be performed to show significant margins against failure.
Critical seismic Category 1 structural elements required,for safe snutdown will be subjected to the seismic margin eartnquake review.
In addition to these preselected elements, a careful review of in-structure shears and j
moments is being conducted to select other structural elements for review.
Dr. Kennedy explained a two-step process whereby a dynamic analysis of all buildings is done.
Tnen, after the dynamic analysis is completed, a detailed stress analysis of individual elements is performed. The dynamic analysis is being done to develop floor spectra in all buildings that contain Category 1 equipment.
Dr. Zudans questioned how Dr. Kennedy justified using the static de-tailed analysis on the basis of a very crude dynamic model (Stick models are being used in the dynamic analyses.).
Dr. Kennedy said that he believes this to be the most accurate way of doing seismic evaluations of facilities. He explained that there is more consistency between models made by difference people using what are commonly called stick models than there is by using the finite element models. He contends that stick modsis tend to agree more with the
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MIDLAND 1 & 2 ^
5/20821/82 median results and with actual responses than finite element models do.
Dr. Trifunac questioned the need to expend so much time and money reducing the uncertai ntie.t in the structural models when the uncertainties in the seismic input might be much more significant.
He suggested a more balanced a~pproa'ch to looking at the problem.
Dr. Okrent asked if the seismic margin review would allow for an assessment to be made of the potential for structural or equipment failure from an earthquake two or three times as large as the seismic margin earthquake.
Dr. Kennedy indicated that it would not. He said that to do a failure estimate on all of the strucures and equipment being reviewed would require approximately twice the work. Dr. Kennedy said that SMA is reviewing the seismic margin earthquake coupled wi'th other expected loads on the plant (i.e., dead load, expected operating load, and seismic margin earthquake). He outlined the range of soil properties being used in their analysis and said that they are taking the largest calculated responses from any of these soil profiles and using that in the review. He talked at length about structures damping, soil material damping, and the radiation of energy into the soil. Dr. Kennedy said that it is his general expectation that all the structures at Midland will show code margins in excess of unity.
In discussing equipment seismic response, Mr. Kennedy said that for about 25%
e of tne floor spectra (and in the range of 4-20 hz) the seismic margin earth-quake's in-structure spectra exceeds the design spectra by up to a fact of two.
Since most of the equipment is qualified for generic-type application, he suspects that most of the equipment (at least the passive equipment) will show code margins greater than unity.
He said that 75% of the cases show that the seismic margin spectra exceeds the design spectra by only a very small amount, and that is the vast majority of the horizontal spectra.
Dr. Kennedy
-aca e.
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MIDLAND 1 & 2 5/20&21/82 considering all of the seismic Category 1 components necessary for safe shutdown. They took a sampling of these critical components, considering the design seismic loads.
Those components in which the design seimsic load was a high percentage of the expected capacity were evaluated.
In addition, they took other components that were critical or potentially vulnerable to seismic loading based upon things that have failed in previous earthquakes and based upcn a welk-through of the plant.
(Equipment loation was considered in that
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the higher up in the plant equipment is, the higehr the seismic input.
So, more equipment from higher up in the plant was taken than from lower down).
This group of equipment will be subjected to the seismic margin earthquake floor spectra for passive system stress and limit loads.
The seismic limit loads will be compared with faulted condition allowables.
Where those code margins are less than unity, a failure margin analysis will be performed.
For active components, the seismic margin earthquake floor spectra will be compared to the spectra that the manufacturers have used to qualify the com-ponents. Again, where margins are less than unity, SMA will get back with the manufacturers and try to work things out. For piping, SMA used piping systems that had either high stresses, seismic induced stresses, or contained active valves. All of the pipe supports and all of the active valves in each of those piping systems are being reviewed.
Switch gear, motor control centers, and instrumentation and control equipment have been selected on the basis of criticality and location. Batteries, the diesel generators, cable trays, and HVAC ducting are also being considered.
l MIDLAND 1 & 2 5/20821/82 e
Mr. Knight commented that "if we successfully design by our present techniques for the type of earthquake that has been specified for this plant, we have a very high confidence that, even given a much more intense seismic event, we would, in fact, be able to shut the plant down and cool it down".
Dr. Okrent reiterated his desire to have the Staff identify the level of risk from earthquakes that the Staff does not want to exceed. He expressed concern that tha margins to failure that exist beyond the SSE are not quantified.
POTENTIAL FOR SOIL LIQUEFACTION AT THE MIDLAND SITE Dr. Thiruvengadam, CPCo addressed the potential for soil liquefaction at the Midland site.
He briefly described the permanent site dewatering system and its design basis.
The facilities requiring permanent site dewatering to protect against possible soil liquefaction include the diesel generator build-ing and the railroad bay area of the auxiliary building.
The loose granular backfill supporting these structures will not liquefy during an earthquake with a peak ground acceleration of 0.19g provided the ground water level in the backfill is maintained below elevation 610.
The dewatering system will maintain the water level under these structures at about elevation 595. Total failure of all pumping capacity in the system would still permit an ample 60 days to repair or reinstall the system before the water reaches elevation 610 in critical areas. Dr. Thiruvengadam demonstrated the availability of this 60 days by showing the Subcommittee the results of a full-scale dewatering and recharge test.
L' ell failure mechanisms and responses were shown.
The design of the dewatering system is based on the fact that the granular backfill material is hydraulically connected to the underlying natural sands and the fact that the
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MIDLAND 1 & 2 5/20821/82 cooling pond (at elevation 627) is the main source of recharge and seepage.
During plant operation the functioning of the dewatering system will be continuously monitored to ensure that the water level stays below elevation 610 and that soil particles (fines) removed are below acceptable levels.
The system has been designed to ensure its operation during various accident conditions (e.g., power outages, loss of wells, and pipe breaks). Dr. Zudans and Dr,. Scavuzzo expressed concern,that over the life of the plant, by removing sand from underneath these structures through the dewatering wells, CPCo might undennine the plant. Their concern was that dewatering might cause settlement and perhaps more damage than pockets of liquefaction.
Dr. W. Paris of Bechtel explained that the well screens have been designed to retain material and not allow fines into the well. He l
l said that the well discharge is being monitored to five micron particle size.
The maximum allowable fines in the pump discharge is ten parts per million (200 gpm aggregate flow from 20 wells). A 10 ppm limit was derived based on an allowable one cubic yard of solid material being removed over the life of each l
well. Mr. Paris said that if this limit is exceeded or is close to be exceeded, corrective action will be taken (the well will be grouted and replaced by a new well).
Dr. Zudans also questioned whether the hydrostatic gradients caused by dewatering might upset the service water pump structure. Dr. Thiruvengadam concluded the discussion of ifquefaction potential by addressing margins to safety.
He said that for a ground motion of 0.199 (Magnitude 6) there is a 1.5 safety factor.
For a 0.25g ground motion, there is a 1.1 safety factor.
(Liquefaction potentials 1
MIDLAND 1 & 2 5/20821/82 were computed by the simplified Seid method.) The Staff indicated agreement with the Applicant's liquefaction analyses.
In response to a request from Dr. Okrent, the Applicant, NRC Staff, and ACRS consultants each gave their estimates of that earthquake having a return frequency of a thousand years, ten thousand years, and a hundred thousand years.
Each person expressed his concern over the uncertainties involved in making such estimates.
Their responses are summarized in the table below.
Mr. Klimkiewicz (Weston Geophysical) 10~4 approx.
0.12g 10" 0.179 approx.
Mr. Kimball (NRC/NRR/DE/GEB) 10-4 approx.
0.12g 10~
0.16g to 0.20g Mr. Reiter (NRC/NRR/DE/GSB) 10~4 approx.
0.12g 10~
no answer 3
Dr. Pomeroy (ACRS Consultant) 10 4 intensity 8 +1 10-intensity 9 T2 10-5 intensity 1013 Dr. Trifunac (ACRS Consultant)*
0.05g 80% exceedance prob.
30 yr. return period 0.10g 40% exceedance prob.100 yr. return period 0.15g 15% exceedance prob. 300 yr. return period 0.20g 8% exceedance prob. 600 yr. return period 3r. Trifunac corrected an error in his calculations after the meeting. He amendmend his answer stating: "the exceedance probabilities for peak acceler-ations greater than 0.03g are smaller than 5% for 50 years of exposure and the return periods are longer than 1000 years."
s MIDLAND 1 & 2 5/20821/82 MIDLAND ORGANIZATION AND MANAGEMENT Mr. J. Cook, Vice President for Projects Engineering and Construction for Consumers Power Company, briefly outlined the company's corporate structure and the engineering and construction operation for which he is responsble.
He highlighted the various lines of communication within the corporate structure and the extensive nuclear experience which the company's top management has. Mr. Cook said that in relation to the Midland Plant, the company's activities can be divided into two line organizations; one is responsible for the engineering and construction activities while the other is responsible for operations. Mr. Cook indicated that since 1980, the company's overriding priority and its total resources ha"ve been devoted toward completing the Midland nuclehr power plant. He said that the Midland Project Office has a direct line of communications to the compnany's chief executive officer. He pointed out that the Midland project Quality Assurance Department is an integrated department that contains all of the project's quality assurance functions, including those of Bechtel. Mr. Cook explained that the Midland site organization contains both their operating and engineer-ing personnel.
This will provide the operations people with several years of hands-on test program experience prior to plant operations.
The site organt-tation will simply report to a different vice president after fuel load.
Mr. DeWitt will take over the direct single-point responsibility of the site and Mr. Cook will support him. Mr. Cook explained that Mr. DeWitt is cur-rently in the support mode but that he has direct funtional supervision of the recruitment of the plant operating staff, the training of that staff, 9
MIDLAND 1 & 2 5/20821/82 and the preparation of the plant operating and maintenance procedures.
Dr. Okrent acknowledged the need for experienced people to run a nuclear utility but questioned its sufficiency with regard to ensuring the public's health and safety. Dr. Okrent brought up CPCo's performance record at its Palisades nuclear power plant.
Several CPCo representatives agreed with Dr.
Okrent that experience alone is not enough.
They each indicated steps CPCo is taking to improve safe opefating and regulatory performance. Dr. Okrent also expressed some concern that the utility is not trying to get a considerable diversity of opinion when it addresses safety-related issues. CPCo contended that they have made every reasonable attempt to get technically competent people with diverse opinions to help them. They gave several examples to support this contention.
Mr. R. DeWitt, Vice President of Nuclear Operations, very briefly reviewed Consumers Power Company Corporate organization, Nuclear Operations Department organization Energy Supply organization, and his nuclear experience / background.
He mentioned planning to transition the Midland Plant over to the Operations Department.
Mr. F. Buckman, CPCo's Executive Director of Nuclear Activities, described Midland's Nuclear Operations Department.
He provided the Subcommittee with information on the expertence possessed by each of CPCo's plant superil-tendents. He outlined the major areas of responsibility for each of the departments within the Nuclear Operations Department and described the people
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!o MIDLAND 1 & 2 33 -
5/20821/82 heading them up.
He similarly described the functions of the departments within his Nuclear Activities Department. Mr. Buckman elaborated on the experience and education which Nuclear Activities Department managers have.
Finally, Mr. Buckman described CPCo's Nuclear Safety Board. He showed the membership of the board, their experience, and qualifications. The board is rcspensible for the oversight of CPCo's operating plants.
In January of 1983 the board will start reviewing activities (events and issues) related to Midland.
Three of the seven members of the board are from outside the operating organization even though they are CPCo employees.
At the request of Dr. Moeller, CPCo listed several of the Nuclear Safety Board's major ac-l l
complishments.
Neither the Safety Review Board nor the Nuclear Activities l
Department are directly involved with the review of licensee event reports l
from other plants. How the requirements for personnel training at Midland are identified was briefly discussed. Mr. Buckman said that, as Midland's PRA is developed, training in the area of risk asessment methodology will probably be required for Midland's licensed operators and plant management personnel.
Dr. J. Osterberg, ACRS consulant, questioned CPCo with regard to the need to have someone outside CPCo on the Nuclear Safety Board.
Dr. Osterberg indicated that this would give the board more objectivity.
Dr. Zudans noted the lack of someone from Midland's training staff on the Nuclear Safety Board.
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MIDLAND 1 & 2 5/20821/82 Mr. G. Slade, CPCo's Assistant Site Manager for the Midland Site Management Office, discussed the organization of the plant staff, human resources planning for the plant staff, and the qualification program for the plant.
He mentioned that the plant staff would receive support from others in the Nuclear Operations Department as well as from other areas of the company. The plant staff is comprised of approximately 750 permanent staff members (not including security personnel, the QA/AC organization, or the training staff).
Mr. Slade indicated that the plant operators, repairmen, technicians, and people on the plant technical staff are actively involved with the plant test program, thereby gaining valuable experience. With regard to plant staffing, Mr. Slade in-dicated that their hiring has been very close to the projections and plans established several years ago. He described the experience of the plant staff.
The organization of on-shift personnel was discussed.
Normally in the control room there will be a shift supervisor (one for both units) and four reactor operators (two for each unit). A plant supervisor (SRO qualified) and, a shift engineer (SRO trainee) will normally be in an office adjacer.t to the control center and available in the event of an emergency. CPCo has tried to go considerably beyond the minimum requirements for a Shift Technical Advisor (STA) in determining the requirements for its shift engineers. A simulator which simulates both units as well as the process steam system is expected to be available for use at Midland by June 1983.
Mr. Slade described the various training programs at Midland. He discussed reactor operator training programs 1
MIDLAND 1 & 2 5/20&21/82 and those for other than licensed operator.
The reactor operator training program has been tailored based on the experience of the individuals (i.e., the length of time each has had at other nuclear facilities).
CPCo is currently conducting a requalification program for all the operators who have already completed the formal training program.
Mr. Slade said that this would allow them to remain familiar with the fundamentals and update their knowledge with regard to field and design changes which ahve occurred.
- Finally, Mr. Slade described the qualifications of the training staff.
He said that the company is not only concerned about the technical competence of instructors; CPCo is placing emphasis on having instructors who know how to teach. CPCo has suplemented its training staff with contract instructors for the applicable portions of the Reactor Operator Training Program (e.g., with instructors who are currently licensed on B&W type facilities).
MIDLAND'S PROBABILISTIC RISK ASSESSMENT Dr. T. Sullivan, Manaer of Safety and Licensing for the Midland project, I
discussed Midland's ongoing probabilistic risk assessment (PRA).
Midland's licensing staff is working with Pickert, Lowe & Garrick and the Midland Site Organization (including STAS and the operating staff) to conduct the PRA.
The PRA was initiated in December 1980 and is 75% completed.
Final results of the PRA are expected in January 1983.
Dr. Sullivan discussed the l
objectives of the Midland PRA. He said that an o'bjective was to evaluate the sources of risk at Midland, paying particular attention to plant unique l
MIDLAND 1 & 2 5/20821/82 features sucn as the process steam system and the proximity to the Dow Chemical Plant. He said that the PRA will serve as a management tool to assess proposed plant improvements on an ongoing basis. The Midland PRA should improve CPCo's PRA capability as well as enhance its overall risk awareness.
Dr. Sullivan next outlined several unique features of the Midland PRA. The Midland PRA will make use of:
event sequence diagram [also used in Midland's emegency operating procedures (EOP) de.velopment program, ATOG],
a three-step plant model (auxiliary system matrix, regular main-line systems matrix, and a computational efficiency method whereby systems operated in the long term are placed in separate sub-trees).
and a system recovery analysis.
l In addition, the plant models are set up so that external initiators are propogated through the internal event's logic. Dr. Sullivan ex-plained that by using tnis method CPCo hopes to obtain an integrated set of plant damage states (frequencies or probabilities) combining external and internal events.
Dr. Sullivan highlighted several benefits which have resulted from Midland PRA efforts to date. Dr. Moeller asked the NRC Staff to verify a statement in Midland's Environmental Statement which said, in effect, that:
The radiological consequences of natural pheno.mna, such as seismic events, would not be different from those tnat have been treated, and In the Staff's judgment these events did not contribute significantly to risk.
Dr. Okrent expressed some disbelief that, at' 75% complete, CPCo has no preliminary numerical results from its PRA. He indicated that in view of the
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MIDLAND 1 & 2 5/20821/82 large population near the Midland Plant he considers these two reactors particularly sensitive. He said that he would like to know the results of the Midland PRA before the plant goes to full power, to see if they suggests a nytning.
Dr. Zudans expressed an interest in including the permanent site dewatering system in Midland's PRA (and not simply assume that the system works or doesn't work).
AUXILIARY FEEDWATER SYSTEM RELIABILITY Mr. L. Gibson, CPCo, discussed Auxiliary Feedwater (AFW) system reliability at Midland.
He briefly mentioned some of the things that have been done to the AFW system, from a design standpoint, since TMI. A simplified piping diagram of the system as it is currently configured was discussed. Tne system has two pumps (one steam driven and one motor driven), automatic switchover from condensate storage tank to service water on loss of suction supply, and either pump can feed eitner steam generator tnrougn a double cross-over. Mr. Gibson said that after extensive discussions with tne Staff, CPCo has agreed to install a third AFW pump by the second refueling outage. Tne third pump will l
be non-safety grade and capable of delivering 100% design AFW flow. Mr. Gibson provided tne Subcommittee with a table that showed AFW system conditional unavailabilities on demand for 2 vs 3 pump systems, conditional on loss of main feedwater and loss of all off-site power as indicated in the table. Unavailabilities of the system at the 15 minute point were also shown under the same conditions
MIDLAND 1 & 2 5/20&21/82 (the assumption or recovery did not include maintenance repair of the system).
The Midland AFW system met the Staff criteria for unavailability (in the range of ten to the minus fourth to ten to the minus fifth per demand) for the 3 pump system configuration but didn't quite meet it for the 2 pump configuration.
Midland AFW system reliabilities met the Staff criteria for both the 2 and 3 pump configurations with a 15 minute system recovery period.
Mr. P. Davis, ACRS Consultant, asked whether cross-connecting the AFW discharge headers for the two units had been considered.
This would allow AFW from one unit to be potentially available for use in the opposite unit.
CPCo representatives gave several reasons why this alternative was not chosen.
Dr. Okrent next questioned the basis for the Staff's AFW unavailability crieria.
Mr. Lobel (NRC/NRR/DSI/ASB) said that the Staff used a core melt frequency (six times ten to the minus six per reactor year) based on a table in Appendix 5 of WASH-1400 for a transient loss of all feedwater, then backed out to the current Standard Review Plan (SRP) criteria. Mr. Lobel indicated that the Standard Review Plan allows for other methods to achieve this criterion other than reliability of the AFW system (Compensatory factors such as other methods of accomplishing the safety functions of the AFW system or other reliable methods for cooling the reactor core during abnormal con-ditions may be considered to justify a larger unavailability of the AFW system.
ref. SRP Section 10.4.9).
Mr. Lobel said that this is the first system where a licensing requirement exists based on the reliability of the system.
He later said that the Staff is not really striving for an absolute number on the l
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MIDLAND 1 & 2 5/20821/82 reliability of the system.
The Staff's AFW reliability criteria represents what the Staff thinks can be reasonably achieved by design.
In cases where an applicant is close to the upper limit of the criterion, Mr. Lobel indicated that the Staff would look at other related factors before accepting the system (e.g., grid stability, steam generator dryout time). Ms. Adensam (NRC/NRR/DL) pointed out that Unresolved Safety Issues, A-45 covers overall plant shutdown decay heat removal capability including the AFW system. She said that having the applicant address the reliability of the AFW system pro-vides a basis, in part, for allowing the applicant to operate until this Unresolved Safety Issue gets resolved.
EMERGENCY OPERATING PROCEDURES Mr. W. Hall, CPCo. discussed Midland's Emergency Operating Procedures (EOPs).
He first mentioned that he has been working with INP0 in an industry effort to develop an Emergency Operating Procedures Writer's Guide.
In addition, Mr. Hall is the Chairman of the B&W Owners Group Subcommittee which is responsible for developing B&W guidelines for writing E0Ps.
The B&W abnormal transient operating l
guidelines (ATOG) is designed to simplify the operating problem of identifying and treating abnormal transients. ATOG promotes symptom-oriented procedures, procedures that treat primary safety objectives and are independent of ini-tiating events.
Three symptoms of primary interest to the PWR operator were identified and selected for the use in these guidelines. They are inadequate primary inventory subcooling, excessive primary to secondary heat transfer, and inadequate primary to secondary heat transfer.
Treating these symptoms will allow returning the plant to a stable condition. Mr. Hall outlined the methodology used to develop the E0Ps.
The AT0G guidelines consist of two parts. The first part
[
MIDLAND 1 & 2 5/20&21/82 is procedural guidance.
The second part explains the design basis for the use of the procedures and is meant primarily as a training aid.
He ex-plained several methods by which ATOG was validated (eg., tested on simu-lators, backchecked against event trees, walked through by plant operators).
Mr. Hall highlighted the plan which will be used to implement ATOG.
The ATOG guidelines developed by B&W will be used as an input to the human factors task analysis done as part of the Control Room Design Review. The resulting task analysis report, along with ATOG, and the plant specific writers guide will be used to develop the plant specific procedures that will be used. The first draft E0Ps will be reviewed by plant operators and Plant Operations personnel. Validation will be conducted by Plant Operat, ions and human factors evaluators.
Finally, Mr. Hall discussed recent and proposed addtions to AT0G. Dr. Okrent noted that by the ATOG guidelines rapid operator action was needed to preclude a steam generator overfill event.
He l
was told that while Midland does, most operating B&W 177 plants do not have automatic feedwater overfill protection (in which case the operator must terminate the transient). Dr. Okrent requested that the Staff provide the Sub-l committee with a status of steam generator overfill protection on older plants at the next ACRS Subcommittee meeting on Midland. Whether or not the ATOG procedures are unambiguous to the operator was briefly discussed. This brief discussion primarily concerned high pressure coolant injection (HPCI) operation and the potention for pressurized thermal shock at Midland. Mr. J. Kelly, I
Babcox & Wilcox, assured Dr. Okrent that the AT0G guidelines are unambiguous I
with regard to HPCI operation. Dr. Okrent then asked to what extent Midland's E0Ps would depend on the correctness of the instrumentation given to the operator
MIDLAND 1 & 2 5/20&21/82 via the instrumentation in the control room. Mr. Hall responded that operators are trained to use Class IE instrumentation as well as other non-IE instrumentation and readouts in the control room.
He said operators are trained to make best use of the instrumentation available to them.
(Operators don't rely solely on the AT0G display for decision-making).
Dr.
Okrent was told that training is not currently conducted which involves scenarios with a loss of all non-IE instrumentation in the control room plus some other multiple events.
Dr. Zudans expressed some concern at ths. high fluence and NDT shift (160 F) which Midland's Unit I reactor pressure vessel will experience during the first 4 years of commercial operation.
AC/DC SYSTEM RELIABILITY Mr. B. Harshe, CPCo, outlined Midland's AC and DC power systems. He showed Midlands two (ie., redundant) off-site AC power sources.
He attested to the systems reliability since the plant can be supplied from either of two dif-ferent directions.
He said that if both Midland Units 1 and 2 were to simultan-l eously drop off the line that off-site AC power would not be interrupted to the site.
Mr. Harshe next described Midland's DC power systems.
The system is comprised of two separate DC buses each with its own battery and two battery chargers.
The buses are not interconnected.
Each bus supplies 120 VAC through
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redundant inverters.
Mr. Harshe compared the Midland DC system to the stan-dardized NUREG-0666 system.
He said that their system is better in that Midland utilizes two battery chargers per bus as opposed to one.
By not having the buses interconnected Midland has eliminated a significant cause of common mode failure.
In addition, Mr. Harshe indicated that the Midland system utilizes an improved alarm system and has expanded the system testing requirements.
Mr.
Epler, ACRS Consultant, complemented CPCo on several of Midland's DC system design characteristics.
MIDLAND 1 & 2 5/20&21/82 STATION BLACK 0UT Mr. B. Harshe, CPCo, began his discussion of station blackout by defining it as a loss of off-site power coupled with the failure of both generators to start or assume load.
He indicated that this event is not part of the design basis but added that CPCo is preparing for such an occurrance. Mr. Harshe said that procedures will be developed at the plant to provide guidance to the operators should a station blackout occur.
He outlined the guidance which will be given to the operators.
Mr. Harshe then said that assuming DC power is available and that the steam driven auxiliary feedwater pump operated properly the plant would have in excess of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (battery capacity) before serious problems developed.
Mr. L. Gibson (CPCo) said that if there was no active core cooling via high pressure coolant injection or auxiliary feedwater there would be in excess of 60 minutes before a degraded core cooling situation existed.
Mr. Kindinger, CPCo, indicated that the two controlling factors (related to time available subsequent to a station blackout) are reactor coolant pump seal leakage rate and battery life.
He said that they both appear to be competing and would lead to core melt in approximately the same amount of time.
He estimated about six hours to core melt based on conservative calculations (ie., not assuming extraordinary actions to shed load off the batteries and using worst case assumptions about what coolant pump seal leakage rate would be).
Mr. Lipinski encouraged CPCo to pre-plan load shedding and other corrective / mitigating actions prior to a station blackout.
i
MIDLAND 1 8 2 5/20821/82 HIGH POINT VENTS Mr. L. Gibson presented the Subcommittee with a discussion of Midland's high point vents.
CPCo's approach has been to provide reactor coolant system vents at the top of the hot legs and at the top of the pressurizer.
These vents are solenoid-operated. There are two valves in series at each vent location. The vents vent up into the atmosphere of the reactor bulding. Mr. Gibson showed, a diagram showing the system configuration.
He explained that any steam or gas which accumulated in the reactor vessel head would be vented into the hot leg before it could affect core cooling. He traced the natural circulation flowpath; from the core, out to the hot leg, down through the steam generator, back into the cold leg, and into the core. He said that this flowpath would not be interrupted by a bubble in the head. He pointed out that the hot leg vents allow for venting and filling of the natural circulation flowpath. Mr. Gibson acknowledged the t
existance of manual control rod drive mechanisms vents atop the core. Mr.
l J. Taylor from Babcock & Wilcox, said that the vents on the top of the indi-l vidual control rod drives are very, very small, and would not be suited for a 2,000 pound connection to remotely operated valves. He said that providing for a suitable head vent would involve drilling and tapping into one of the flanges on a control rod drive mechanism. Dr. Zudans, ACRS Consultant, expressed an interest in having experimental evidence that a bubble in the head could not interfere with the natural circulation flowpath. He also expressed concern that plant operators might not be able to detect voiding in the reactor vessel head region. The technical feasibility of installing a heated junction thermocouple
MIDLAN 1 & 2 5/2D/21/82 in the head to detet voiding was acknowledged by CPCo. However, they in-dicated that there is currently no penetration to install such a device.
Dr. Okrent left tne meeting.
Dr. Moeller assumed tne role of Subcommittee Chai rma n.
EMERGENCY PLANNING Mr. G. Slade, CPCo, discussed Midland's Emergency Plan.
He concentrated on three unique features of the plant and now each relates to the Emergency Plan.
Tne plant is located in the State of Micnigan Tne plant is located witnin the city limits of Midland Tne plant is located adjacent to a major chemical manufacturing f aci lity, Mr. Slade indicated that Micnigan has a State Emergency Plan. Tnat plan is required to be exercised routinely for four other nuclear sites (tnree different utilites). The state will participate in at least one full scale exercise a year. The state will also take part, to some limited extent, in tne annual full scale exercise conducted at each of tne other nuclear sites. Mr. Slade briefly described now the Midland Nuclear Plant Emergency Plan is coordinated with the plan for Dow Chemical Company. Ms. B. Hinton, CPCo, stated that the majority of Dow workers would be considered as members of the public as far as radiation protection is concerned. However, she indicated that certain processes on tne Dow property would require some attention and time to snut down.
Tnose workers required to stay at the Dow site will be afforded personnel monitoring devices i
f
MIDLAND 1 8 2 5/20/21/82 and a location to go where they will be protected in the event a plume passes directly over their area.
With regard to the proximity of the Midland nuclear power plant to the city of Midland, a detailed evacuation study was conducted.
Mr. Slade said that the study concludes that any combination of sectors that need to be evacuated within the 10 mile primary planning zone can be evacuated out to 10 miles within a period of four hours. Dr. Moeller expressed some concern that evacuation, if needed, be done in an orderly fashion and not under duress or under a panic condition.
CPCo will be installing a warning system around the Midland Plant site.
Lieutenant Tyler of the Michigan State Police described the State's role in Midland's Emergency Plan. He discussed the radiation monitoring devices used by the State.
He also described the training programs that the State Police have established to upgrade and educate state and local agencies about emergency planning and response.
Dr. Moeller asked the Staff to see if the numbers in Midlands Emergency Plan (taken out of NUREG-0654) related to the return frequencies of unusual events, alerts, site emergencies, and general emergencies were consistent with experience (ie., unusual event-several per year, alert - every 10 to 100 years, site emergency - once every 100 to 500 years, and general emergency once every 5,000 years).
MIDLAND 1 & 2 5/20/21/82 ENVIRONMENTAL ISSUES Mr. D. Sommers hignlighted several controversial or outstanding environ-mental issues on the Midland Project.
Issues discussed relating to the National Pollution Discharge Elimination System (NPDES) include:
control of total dissolved solids in the Tittabawassee River, thermal effects of discharges to tne Tittabawassee River, and the assimilative capacity for ammonia in the ri-ver.
Issues discussed relating to tne Draft Environmental statement (DES) include:
tne potential for fogging and icing, and tne potential for increased death due to disease, and starvation of water fowl in the cooling pond.
CPCo indicated tnat all of these issues should be resolved without difficulty.
POTENTIAL FOR GROUND WATER CONTAMINATION AT THE MIDLAND SITE Mr. D. Sommers, CPCo, discussed the potential for ground water contamination at the Midland site.
He said that the potential for ground water contamination is minimal because:
there are no routine discharges of radioactivity to the cooling pond, there is a cofferdam around the tank farm to minimize the spread of a spill.
the permenant site dewatering system discharges to the cooling pond, which would act as a hold-up tank if contamination was introduced into the dewatering system, and the power block and the tanks lie within the dike structure (thereby lying essentially in a big clay dish so tnat any spill remains confined).
MIDLAND 1 & 2 5/20&21/82 Dr. Moeller asked the NRC Staff to tell the Subcommittee at its next meeting why tritium was not included on the DES list of radionuclides in the core.
RADIATION PROTECTION Mr. Beckman, CPCo, made a brief presentation on the Midland Plant Radiation Safety Program.
He said that this program is based on the "as low as reasonably achievable" (ALARA) concept.
The Midland Plant program receives written guidance through the company's corporate offices in the form of Corporate Radiation Safety Standards and a Radiation Safety Plan. The plant safety and radiation safety organization have established an ALARA coordinator and annual ALARA goals.
The plant ALARA goals are approved by the plant manager and the Vice President for Nuclear Operations.
The company has established a cost guideline of $5,000 per occupational man-rem, for use in evaluating new equipment and modification cost.
Mr. Beckman said that a computerized tracking system has been developed to track the radiation exposure of individuals by work and job function, by specific task, and by component. He also outlined a procedure for reviewing and approving design changes and procedures as a function of their dose to workers.
These ALARA reviews will be done prior to completion of an activity and again after the work is done.
Finally, Mr. Beckman outlined both the normal radiation dose projection and accident dose assessment methods which will be used at the plant.
Dr. Moeller urged CPCo to keep the source term as low as possible by keeping the plant clean and to keep in mind potential mitigative actions that could be used to minimize personnel exposures.
MIDLAND 1 & 2 5/20821/82 Dr. Moeller said that, at the next Subcommittee meeting, he would like to know how many piping systams containing radioactive materials at Midland have the provision for draining and flushing to prevent crud build-up.
He also expressed an interest in finding out how frequently these systems will be drained and flushed, what the criteria will be for draining and flushing them, and if these systems can be decontaminated.
Dr. Moeller indicated that the ACRS Midland Plant Subcommittee would like to hear about the following topics at its afternoon meeting on June 2,1982:
Control Room Habitability Items on the tentative schedule for the May 20-21, 1982 meeting which were not covered.
Plant Security The meeting was adjourned at 4:15 p.m. on Friday, May 21, 1982.
i NOTE: A complete transcript of the meeting is on file at the NRC Public l
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)S54-2345.
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O proWding this service.To this urast be A,g:rterofCyy:Wat Employee. Mr. Pen! Boehnert tialepheme 202/204-32s7) between 3:15 aJa. and I,
added the basic
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5.100 p.m., a.a.t.
g appDestion for ugistretion Ice whids I have dettrunned. ts ecoordance with M *P" # C"8"'"
subsection 10(d) of the Federal Advisory special handling service is regnested I""""*'""""'s Committee Act, that ft any be neonesary L
and gramtei H the request for epedal
- o close some portions of this meetagto bandling Is 3 anted the $1m00 fee is not i
refundable.%e fee may be charged to a -
Protect proprietary tuformation.De p
deposit account estabbshed in the authonty for such closure is Exemption NUCt EAR REGUt.ATORY (4) to the Sanshine Act.5 U.S.C.
Copyright Office. U the deposit account COMMISSJON 8521(c14).
i f
contains insu!$cient funds to cover the Daise Aprazr.asas.
L total spedal hancing fee. ce if the AMsory Mee on Reactor k
remitter does not maintsin a deposit Safeguards: Subcommrttee en lohs c usyla, account. th e 1otal special handling fee Advanced Reactors; Meeting AdvisoryCarowmeM. - " Ogame-p may be paid elther in person at the
.ps:w an-masre m aessame Public Information O!5ce or !! may be Advanced Reactors will bold a moiting
~ * * * ' " ' * * * '
- remitted by med.,5uch payment must be M
W Mb W a,etional Labo stom ding 20s. Romn. Advisory Committee on Reactor to cash or tn the norm of a certiSad h
check cashier's check, or money ordar C-234. Argarme.II.He Sobcounnittee Safeguarda, C.Wn9ttee on Midland L.
made pmyable to the Eagister of will contmue discussion of a report an Plant UrWta 1 snd t Moenre t
gP Liquid Metal Fast Breeder Reactor
%e ACRS Subcommittee onMidland f
- 3. Fifing Requests For Spedal HandDng (LMF3R) eafety pb30eophy. Notice of
'I this meeting was published April 13.
Plant Units 1 and 2 wiD hold a meetung Requests for spacia! handling sney be la accordance with the procedures on May 20 and 21.19tt'. at the made in person on the form availsble in outlined in the Federal Register on HOLIDAYINN.1500W.Wackedy
- Road. Midland.ML %e Shi-the Pobbe Information O!Sce af the September 30.1981 (46 E 47903) oral or Copyright Office. Rm 1.M-401. (ames wntten statements may be presented by wiH nWew the application by Madasan Memorial Building library of members of the public, recordings wHj Consumers Power Coarpeny for a Congnsa. The Office will a!so consider be permitted ordy during those portions I' cense to operate Midland PIant Units -
requests by maa providing the Spedal of the meeting when a transcript is being and 2. Noboe of this assetsag was Handling Request forto is used or a kept and questions may be asked only Published April 22.
In accordance with the m,---
cover letter is submitted containing by icembers of the Subcommittee,its outlined in the Todaral Registae on answers to the foUowing quertions that consu!! ants and Staff.Penons desir!ng F
are required to be answcred in the to make oral etatements should netify September 30.1981 (46 FR 47903), oral oc specialbandling form;*'Whyis there an the Designated Federal Employee es far written statementa may be presented by members of the public,recordingswSt J
urgent need for special handlin3. Ifit is in advance se practicable so that be permitted only durms tbooe portions 7
I bes:ause oflitigation is thelitigation appropriate arrangements can be made
/
actual or respective? Are you or your to aI!ow the cecessary time during the of the muting wtes a transaipt is being ma may be as och ctent the plaintiff or defendant'.What meeting for such statements.
fep an n
g are the names of the parties and the c nsultants, and StafL Persons desiring name of the court where the actionla pu e att ndan ce for th e to make oral statreents abould noti (y sessions during which the Subcomndttee the Design ated Federal Employee as far 7
pending or expectedT-It is also finds it neensary to discuss proprietary in advance as practicable so that necessary to-certify that the answers to information. One or more dosed appropriate arrangements can be made these questions are correct to the best of sessions may b necessary to din'uss to a!!ow the necessary time during the yourInowledge. A mailed request far auch information. (Sucshine Act s'
special handling should be sent to:.
Exemption 4). To the extent pescticable, meeting for such statementa.
IJbrary of Corggres s. Departroe,1 DS, these closed sessions will be held so as ne ectare muting w11D be to Washingtort D C 20540. Attention:
Io nunimize Inconsenience to membats pMc aMendance except for sessims which wiH be closed in protect p
' Acquisinone and processing Division of the public in attendance.
proprietary information (Sunshine Act Olka.
The agenda for subject c:eeting shall Exemption 4). One or more closed be as foUows:
De outside of the envelope and the sessions may be necessary to discans letter inside should c!eerly indicate that Afond yandTuesday MayIf andJA JM-such information. To the extent
" del #A# ##"C#"# A88 '[O"'O8 practicable, these dosed ses sions wE
(
lt is a request for special handling.
- j be held so as to minmfre inconventemos
%e request for special handIIng of a The Subcommittee,andtts consu! tams wEl rethtration must be accormpartled by a d;souss possib!e design constderations.
to ciembea3 of the puMc in attendanna.
comp!eted applica tion, the required Issues, or criteria for future commerds!
%e agenda for subject meeting shall
- dranced teactam, be as foUows
)
deposit copies, phonorecords, or
,Further Inforrnation reguding topics 7hader.May 2a, rm-&30 e.m se ta ress l
Identifying c aterial. and the S130 00 fee
($CO 00 special handling fee and $10.00 to be discussed, whether the meetL9g PA CMdWPm antilmpau, reg 4tration fee). A minimum of five has been cancelled or reecheduled. the Ehd87 Af'I #12*--83888 ontf 88
- "*as e#ussasar worling days must be allowed for Chairman's ruling on requests for the processing a c!airn under spedal opportunity to prescot oral statements During the initial portion of the handEng procedures, and the time aEotted thenfor canbe meeting. the Subcommittee, along wfth
^
m
Federal Re:Ister / Vol. 47. No. 86 / Tuesday. May 4,1982] Notices 29230 o
cny ofits consultants who may be tee entire meeting wG! be open to Unit No.11ocated in Calvert County, present, wiU exchange prehminary pablic attendance except for those Maryland.
views regarding matters to be sessions which will be closed to protect ne amendment would revise the considered dunng the balance of the propnetary infermation (Sunshine Act Safety I'mits and Umiting Conditions m:eting.
Exemption 4). One or more dosed for Operation. contained in the The Subcommittee will then hear sessions may be necessary to discuss Appendix A Technical Specifications, prasentations by and hold discussions such information.To the extent for Calvert Cliffs Unit No.1in with representatives of the Consumers practicable, these closed sessions wEl accordance with the licensee's Power Company, the NRC StaH..their be held so as to minimize inconvenience app!! cation dated February 17.1982.
consultants, and other in'e'ested to members of the public in attendance.
Dese changes are primarily the result ptrions regarding this review.
The agenda for suoject meeting shall of new safety analysis methods applied Further information regarding topics be as foUows:
to the analysis for Cycle 6 operation of I be discussed.whether the meeting M'e6esday. May sa iss?-4m a.a until s:
Ca' vert Cliffs Unit 1.The analysis h:s been cancelled or rescheduled. the.
Noon presented provides for an increased Chairman's ruling on requests for the TAursdcy. May an iss'--am e.m. until the operating period between refuelings.
tpportunity to present oral statements condusion of business Prior to issuance of the proposed and the time aUotted therefor can be During the initial portion of the license amendment, the Commissior.
abtained by a prepaid telephone call to meeting, the Subcommittee, along with will have made fhviings required by the the cognizant Designated Federal any ofits consultants who may be Atomic Energy Act ofl954, as amended E-.p y w n *l Fischer (telephone present may exchange preliminary (the Act) and the Commission's t
202/634-1413) betv een 8:15 a.m. and views regarding matters to be regulations.
c nsidered dunna the balance of the By June 3.1982, the !!censee may file a heve de rmined in accordance with
~
requeat for a heating with respect to meeteg.
subsection 10(d)of the Federal Advisory The Subcommittee will then hear issuance of the amendment to the Committee Act. that it may be necessary presentations by and hold discussions subject facility operating license and ts close portaons of this meeting to with representatives of the NRC Sta5.
any person whose interest may be public attendance to protect proprietary their consultants, and other interested affected by this proceeding and who informauon. The authority for such persons regarding this review.
wishes to participate as a party in the closure is Exemption (4) to the Sunshine Further information regarding topica proceeding must fUe a written petition Act. 5 U.S.C. 552b(c)(4).
to be discussed whether the meeting for leave to Intervene. Requests for a Dated. Apn!27.tsa2.
has been canceHed or rescheduled. the hearing and petitions for leave to John C Hoyle.
Chairman's ruling on requests for the latervene shaU be fUedin accordance A drisory Comminee Mancrement oficer.
opportunity to present oral statements with the Commission *a " Rules of gru o si.mm ru.m.aa ns )
and the time aUotted therefor can be Practice for Domestic Ucensing an a.c caos ris.ei.e obtained by a prepaid telephone call to Proceedings"in to Cm Part 2. If a the cognannt Designsted Federal request for a hearing or petition for Employeee. Dr. Richard Savio or Sta5 leave to intervene is fued by the above Advisory Committee on Reactor Enyneer.Mr. Anthony Cappucci date, the Commission or an Atomic Safeguards, Subcommittee on (te ephone 202/634-3287] between 8:15 Safety and Ucensing Board designated Guahfication Program for Safety a.m. and 5:00 p.m. d.s.t.
by the Commission or by the Chairman Rslated Equipment; Westing I bave determined in accordance with of the Atomic Safety and Ucensing The ACRS Subcommittee on subsection 10(d) of the Federal Advisory Board Panel, wiU rule on the request Qualification Program for Safety Related Committee Act, that it may be necessary and/or petition and the Secretary or the designated Atomic Safety and Ucensing Equipment wiU hold a meeting on May to close portions of this meeting to 19 and 20.1982, at the AMFAC Hotel, public attendance to protect proprietary Board wiU issue a nudce of hearing or Valle Grande Room. 2910 Yale Blvd.,
information.The authority for such an appropriate order..
Albuquerque.NM.De Subcommittee closure is Exemption (4) to the Sunshine As required by to CR 2.714. a will discuss three major program areas Act. 5 U.S.C. 552b(c)(4).
petition for leas e to intervene sha!! set conducted by Sandia I.aboratory and Dsted. Apr0 2s 198L forth with particularity the interest of continue its redew of equipment John C Hoyle, the petitioner In the proceeding, and how that interest may be affected by the qualification requirements. Notice of A drisory Committee. Manegement CWicar.
this meeting was published April 13.
results of the proceeding.ne petition A es.ime r'w wm In accordance with the procedures should specificaUy explain the reasons o ene,7,m why Intervention should be permitted outlined in the Federal Register on September 30,1981 (46 m 47903). oral or with particular reference to the written statements may be presented by [ Docket No. 50-317]
foUowing factors:(1)The nature of the members of the ublic. recordings will petitioner's right under the Act to be Battimore Gas & Efectric Co4 made a party to the proceeding:(2) the be permitted o durirg those portions Consideration of Issuance of nature and extent of the petitioner's of the meeting w en a transcript is being 5ept, and questions may be asked only Amendment to Facility Operating property, fina ncial. or other Intere st la UC'"S*
the proceeding: and (3) the possible' by members of the Subcommittee.its consultants, and Staff. Persons desiring De United States Nuclear Regulatory effect of any order which may be to reake oral statements should notify Commission (the Commission)is entered la the proceeding on the the Designated Federal Emfoyee as far considering issuance of an amendment petitioner's interest.ne petition should in advance as practicable so that to Facility Operating Ucense No. DPR-alsoidentify the specific aspect (s) of the cpprepriate arrangementi can be made 53 issued to Baltimore Ces and Electric subject matter of the proceeding as to to allow the necessary time during the Company (the Ucensee), for operation of which petitioner wishes to intervene.
meetica for such statements.
the Calvert Cliffa Nuclear Power Plant.
Any person who hae filed a petition for
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'n. TENTATIVE SCHEDULE FOR THE MAY 20 & 21, 1982 ACRS SUBCOMMITTEE MEETING ON MIDLAND UNITS 1 & 2 HOLIDAY INN, 1500 WACKERLY RD., MIDLAND, MI THURSDAY, MAY 20, 1982 APPROXIMATE TINE TOPIC SPEAKER 8:30 A.M. I. CHAIRMAN'S OPENING STATEMENT D. Okrent 8:45 A.M. II., INTRODUCTION BY NRC STAFF NRC Staff A. Overview of OL Review 1. Status of review (including open items a licensing conditions) 2. Comparison of Midland with other similar PWRs reviewed by the NRC Staff 3. Summary of principal review issues & status of comments on the TMI Action Plan (including a summary of the safety issues in this review which the NRC Staff believes were the most difficult to resolve) B. Discussion C. Comments by Applicant 9:30 A.M. III. QUALITY OF DESIGN AND CONTRUCTION A. I&E Report on Significant Plant Experiences (including a discussion on Construction QA/QC and an IAE Assessment of Construction Management) B. Discussion 10:30 A.M. BREAK l
o TENTATIVE SCHEDULE MIDLAND MAY 20 & 21, 1982 THURSDAY, MAY 20, 1982 (CONTINUED) APPROXIMATE TIME TOPIC SPEAKER 10:40 A.M. IV. INTRODUCTION BY THE APPLICANT A. General B. Plant and Site Description I. Construction Schedule, Projected Schedule to Commercial Operations V. GENERAL TOPICS FOR DISCUSSION 11:00 A.M. A. Human Factors Review of the Control Room 11:10 A.M. B. Alternate Shutdown Panel 11:20 A.M. C. Instrumentation 1. To detect inadequate core cooling 2. Reactor vessel level indication 11:30 A.M. D. AC/DC Systems Reliability 1. Offsite power 2. Emergency diesel power distribution scheme 3. Station blackout 4. DC system reliability 12:00 E. Process Steam 1. System description 2. System operation 3. Radioactivity monitoring 12:15 P.M. RECESS / LUNCH 1:15 P.M. - 5:00 P.M. SITE VISIT
TENTATIVE SCHEDULE MIDLAND MAY 20 8 21, 1982 THURSDAY, MAY 20, 1982 (CONTINUED) APPROXIMATE TIME TOPIC SPEAKER 6:00 P.M. F. Seismic
- 1. Criteria to which plant and equipment were built (Seismic Input Criteria)
- 2. Site-specific spectra
- 3. Probabilistic estimates of increasingly severe earthquakes
- 4. Reevaluation of plant components and structures necessary for safe shutdown and decay heat removal.(including margins available)
- 5. Need for dewatering to reduce the probability for liquifaction due to an earthquake (including margins available)
- 6. Other factors relevant to issue 8:00 P.M.
G. Probabilistic Risk Assessment l l 1. Preliminary findings to date 8:15 P.M. H. Auxiliatry Feedwater System Reliability 9:00 P.M. I. Methods to Reduce Common Cause Failure 1. Systems interaction studies 2. Separation of control and protection instrumentation 3. Changes in design resulting from previous experience 9:30 P.M. J. Other l 10:00 P.M. RECESS l
r ,O TENTATIVE SCHEDULE MIDLAND MAY 20 8 21, 1982 FRIDAY, MAY 21, 1982 APPROXIMATE TIME TOPIC SPEAKER V. GENERAL TOPICS FOR DISCUSSION (CONT'D) 8:30 A.M. K. Organization and Management 1. Corporate & operational management
- a. Experience levels
- b. Staff buildup - (percentage of engineering and management buildup)
- c. Utility offsite and onsite technical capability organization (Compliance with NUREG-0731, " Management Structure and Technical Resources")
2. Operations organizatien
- a. Plant and industry operating experience
- b. Uses of contractor personnel
- c. Criterion used to determine the oper-ating organizations' ability to assume control of plant from startup groups and contractors 3.
Nuclear plant engineering organization 4. Nuclear services organization 5. Nuclear safety review groups j l 6. Quality assurance organization 9:15 A.M. L. Operations Staffing and Training 1. Plant management chain of command 2. Current status of operations staffing (percentage complete and experience level) 3. Selection of for operators and main-tenance personnel 4. Plant communications (normal operations, maintenance, refueling, outages, and emergencies (How are routine and emer-
'i - TENTATIVE SCHEDULE MIDLAND MAY 20 & 21, 1982 FRIDAY, MAY 21, 1982 (CONTINUED) APPROXIMATE TIME TOPIC SPEAKER 5. Plant safety review committee 6. Uses of contractor personnel 7. Uses of STA from utility perspective 8. Feedback to operators, STAS, and others of plant and industry ex-perience 9. Training for operators and maintenance personnel
- 10. Uses of simulators in training program l
11. Training for serious accidents, DBA and beyond DBA t 10:00 A.M. M. Emergency Operating Procedures 1. ATOG 2. Reliability of instrumentation during abnormal plant conditions 3. E0P for case of severe earthquake 10:30 A.M. BREAK 10:40 A.M. N. DHR System Operations 10:50 A.M. O. System Highpoint Vents 11:00 A.M. P. Bolting and Other High Strength Material 1. RPV bolts 11:10 A.M. Q. Fire Protection 1. Potential problems from spurious actuations, I flooding or wetting, damper actuations
l % = 4 TENTATIVE SCHEDULE MIDLAND
- MAY 20 8 21, 1982 FRIDAY, MAY 21, 1982 (CONTINUED)
APPROXIMATE TIME TOPIC SPEAKER 11:20 A.M. R. Integrated Control System 1. What has been done to upgrade the ICS since TMI changes brought About by B&W vs. CPC 2. Failure modes & effect analysis 11:40 A.M. S. Seismic and Environmenta1 ' Qualification of Equipment Important to Plant Safety 12:00 noon LUNCH 1:00 P.M. T. Items from previous ACRS letters NRC Staff 1:20 P.M. U. Habitability 1. Control room 2. Emergency response facilities 1:40 P.M. V. Emergency Planning 1. Meteorological monitoring program 2. Evacuation plan (including impact on DOW Chemical Plant) 1:50 P.M. W. Radiation Protesction Program 1. ALARA 2. Dose assessment procedures 2:00 P.M. X. Other 2:30 P.M. <;. Y. Status of ACRS review D. Okrent 3:00 P.M. BREAK (CHANGE OF CHAIRMAN) 3:10 P.M. Z. Environmental Monitoring (Including Cooling pond, Tittabawasse River and Indigeneous Wildlife) 3:20 P.M. AA. Potential for Ground Water Contamination at the Midland Site m F% DDRm
=*- e ATTACHMENT D REFERENCE DOCUMENTS FOR THE MAY 20-21, 1982 ACRS SUBCOMMITTEE ON MIDLAND PLANT UNITS 1 & 2 1. Consumers Power Co., " Midland Plant Units 1 and 2 - Final Safety Analysis Report" including Amendments 1-43 2. U.S. Nuclear Regulatory Comission, " Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2," NUREG-0793, dated May 1982 3. Ltr. from J. Hendrie, Advisory Committee on Reactor Safeguards (ACRS to G. Seaborg, Chaiman, US Atomic Energy Commission (AEC),
Subject:
Report on Midland Plant Units 1&2, dated June 18, 1970 4. Ltr. from J. Hendrie, ACRS to G. Seaborg AEC,
Subject:
Supplemental Report on Midland Plant Units 1&2, dated Sept. 23, 1970. 5. Ltr. from D. Moeller, ACRS, to M. Rowder, US Nuclear Regulatory Commission (NRC),
Subject:
Supplemental Report on Midland Plant Units 182, dated Nov.18,1976 6. Ltr. from M. Bender, ACRS, to M. Rowder, NRC,
Subject:
Additional Request for Info. From the AS&LB, dated March 16, 1977 7. NRC Supplement No. 2 to the Construction Permit Safety Evaluation of the Midland Plant Units 1 and 2 dated July 1977 8. Memo for ACRS Members From C.P. Siess,
Subject:
Report of Ad Hoc Subcommittee on Foundation Problems and Remedial Actions at Midland Plant Units 1&2 (Cfficial Use Only) 9. Ltr. from R. Foster, ACRS consultant, to D. Moeller, A.CRS,
Subject:
Coments on Midland DES and Emergency Plans, dated April 29, 1982
- 10. Ltr. from P. Davis, ACRS Consultant, to D. Okrent, ACRS,
Subject:
Evaluation of Auxiliary Feedwater System Reliability at the Midland Plants, dated April 26,1982
- 11. Ltr. from P. Pomeroy, ACRS Consultant, to D. Okrent, ACRS,
Subject:
Midland Site Specific Response Spectra, dated April 28, 1982
- 12. Ltr. from J. Hickman, ACRS Consultant, to D. Okrent, ACRS,
Subject:
Midland Plant Auxiliary Feedwater System Design, dated May 7,1982
- 13. Ltr. from M. Sinclair to C. Stess, ACRS,
Subject:
Midland Soil Settlement, dated April 26, 1982 14 Memo from A. Marchese NRC/ DST to 0. Parr, B. Sheron, A. Thadani,
Subject:
Appendix C SER Input for USI A-45 to Meet ALAB-444 Requirements dated April 29,1982 Staff, to
- 15. Memo from D. Fischer, ACRS/ Members,
Subject:
Midland Plant Safety Review Groups, dated May 17, 1982
- 16. Memo from R. Minogue, NRC/RES, to H. Denton, NRC/NRR,
Subject:
Research information Ltr., No.130 " Experimental Verification of Computer Simulations of Piping Response Using Heissdampfreaktor Seismic Tests", dated March 16, 1982
- 17. Extract from DP-1623 Workshop on Seismic Performance of Underground Facilities dated Feb. 11-13, 1981
Subject:
Effect of Changing Return Periods on Probablistic Ground Motion by David M. Perkins, US Geological Survey, Denver.CO. 1 e
DOCUMENTS AND SLIDES PROVIDED AT THE MEETING 1. Written statement by Dr. C. Anderson to ACRS Midland Plant Subcomittee dated May 20-21,1982 (See Attach. E) 2. Written statement by Ms. M. Sinclair to ACRS Midland Plant Subcommittee dated May 20-21,1982 (See Attach.E) 3. Ltr. from B. Stamiris to D. Fischser and ACRS Mbrs.,
Subject:
Soil Settlement and QA Issues, dated May 20, 1982 (See Attach. E) 4. Slides used by R. Hernan (NRR/DL/LB4), Project Manager's Introduction / Overview (5 slides) 5. Slides used by R. Ham, CPCo, Human Factors Review of the Control Room (4 slides) 6. Slides used by R. Hamm, CPCo, Alternate Shutdown Panel (3 slides) 7. Slides used by R. Hamm, CPCo. Instrumentation to Detect Inadequate Core Cooling (3 slides) 8. Slides used by J. Alderink, CPCo, Process Steam System (4 slides) 9. Slides used by D. Somer, CPCo., Process Steam Radiation Monitoring System (3 slides)
- 10. Slides used by T. Thiruvergadam, CPCo, Overview of the, Seismic Criteria to which the Plant was Built (,4 slides)
- 11. Slides used by R. Holt, Western Geophysical Corp., Site Specific Response Spectra (5 slides)
- 12. Slides used by G. Klimktewicz, Weston Geophysical Corp., Seismic Hazard Analysis (20 slides)
- 13. Slides used by J. Kimball (NRR/DE/SEB), NRC Staff Seismic Evaluation of Midland (16 slides)
- 14. Slides used by L. Reiter (NRC/DE/GSB), The Use of Probability in Detemining Seismic Risk (17 slides)
- 15. Slides used by R. Kennedy, Structural Mechanics Associates, Seismic Re-evaluation of Midland (55 slides)
- 16. Slides used by T. Thiruvergadam, CPCo. Pennenant Site Dewatering (3 slides) 17 Slides used by Mr. J. Cook, CPCo. Consumers Power Company Organization and Management (7 slides)
- 18. Slides used by Mr. R. DeWitt, CPCo, Consumers Power Company Organization and Nuclear Operations (4 slides)
- 19. Slides used by Mr. F. Buckman, CPCo, Nuclear Operations Department (14 slides)
- 20. Slides used by Mr. G. Slade, CPCo, Plant Operations Staffing and Training (16 slides) 21. Slides used by Dr.T. Sullivan, CPCo, Midland Probabilistic Risk Assessment (3 slides)
- 22. Slides used by Mr. L. Gibson, CPCo, Auxiliary Feedwater System Reliability (4 slides)
- 23. Slides used by Mr. W. Hall, CPCo, Emergency Operating Procedures (11 slides)
- 24. Slides used by Mr. B. Harshe, CPCo, AC/DC System Reliability (7 slides) 25 Slides used by Mr. B. Harshe, CPCo, Station Blackout (3 slides)
- 26. Si f des used by Mr. L. Gibson, CPCo, High Point Vents (2 slides)
- 27. Slides used by Mr. D. Sommers, CPCo, Environmental Issues (1 slide )
- 28. Slides used by Mr. D. Sommers, CPCo, Potential for Ground Water Contamination at the Midland Site (3 slides)
- 29. Slides used by Mr. W. Beckman, CPCo, Radiation Protection Program (5 slides)
- 30. NRC Staff response to ACRS Items of November 18, 1976 Supplemental Midland Report m_
S'TATEMENT BY DR. CHARLES ANDERSON,P.E., E before the Advisorv Committee on Reactor Safeguards May 20-21,1982 Soil / Structure Interaction Problems
- 1. Soil Settlement Fodndation settlement consists primarily of two parts, primary and secondary settlement. Primary settlement is usually used for design in this country and secondary is ignored. Just the opposite is the practice in Europe.
These two parts of total settlement are often roughly about the same amount. It appears that-in this case especially the Diesel Generator Building (DGB)- secondary settlement has not been considered. I believe settlement for the DGB is not yet completed, but will continue for some years causing further ( stress and cracking to the building. 2. Structural Integrity The DGB is excessively cracked from differential settlement. This crack-ing will continue for some years. The structure consists of roof girders l spanning to bearing walls. These bearing walls provide a box structuret ) however, in the case of the DGB nnd the Service Water Building, the classical box structure has been modified in the roof, in the floor, and especially in some walls where the box stiffness of these walls has been greatly reduced by large structural holes and cracks. The stiffness of the structure has been largely lost leaving the building without the degree of rigidity that certain representatives cf Consumers Power Company have told me is there. I assure you that it is not. Both buildings lack structural integrity today. The settling, the cracking and the further loss of structural integrity can only continue. l u 4 rs6tJ Dr. Charles Ande: son, P.E., Esquire A A Engineering 9213 Bois Avenue Vienna Virginia B.C. E., University of Minn.,,C)vil F
- eering, Professional Qualifications:
M.S.C.E., University of Minn., PhD., University of Minn., 6$ - ~ ince ring, B.S., University of d'ew Mexico, Geology, M.S.C.E., University ofbTe'Alco, Soils Fngineering, Post Doctoral Studies, University of Calif. at Berkeley, Geo-technical Fngineering, J.D., Georgetown Law School, Law, Own Consulting Firm A A Engineering, 9213 Bois Avenue, Vienna, Virginia Twenty years mn#4Aingrefera sf sonstruction, author of 6 books on
STATEMENT BY DR. CHARLES ANDERSON, P.E., ES Before the Advisory Committee on Reactor Safeguards May 20-21,1982 Addendum e 1, May 20,1982 Consumers' 11:00 a.m., May 20,1962 statement, "Dr. Anderson mistook the second Door as the roof,"is not factually correct nor is it relevant to the contention "... stiffness of the structure has been largely lost..."in the first paragraph of section " Structural Integrity". The " box stiffness" referred to by me on May 20, 1952 as being reduced is in fact due, in part, to the horizontal structural plates at the top of the building. It is my contention that Consumers' statement is either incompetent or intentionally malicious. t Dr. Charles Anderson, P.E., Esquire A A Engineering 9213 Bois Avenue Vienna, Virginia ~
e STATEh!ENT BY DR. CHARLES ANDERSON, P.E., ESQ. Before the Advisory Committee on Reactor Safeguards May 20-21,1962 Addendum
- 2, May 21,1982 After 3 days of being jerked around and lied to by Consumers Power Co.,
having been denied requested access to the Midland site, having had Consumers distort my May 20, 1952 statements before the ACRS, realizing fully Consumers general lack of good faith in their desperate situation, it is time to draw the bottom line on the Diesel Generator Building: Consumers erred in the placement of the fill. Dr. Peck erred in the pre-loading of the fill with the DGB already constructed. As a result, the structure has cracked ex!ensively, and yet secondary settlement has not been arrested. The building has lost its essential structural integrity and is now exposed to continued settlement and additional structural cracking. The function of the building apparently cannot be changed and is critical to the site. The' building cannot reasonably be made to structurally satisfy its critical function considering the troublesome site history. . The building can only be replaced to resolve the problem of public s afety. f Ji Dr. Charles Anderso., P.E., Esquire A A Engineering 9213 Bois Avenue Vienna, Virginin
STATEMENT BY MARY SINCLAIR before the Advisory Committee on Reactor Safeguards May 20, 21,1982 In the first summary statement that the Advisory Committee on Reactor Safeguards (ACRS) prepared for the Midland nuclear plants dated February 6-8, 1969, a number of significant issues were raised about which I believe the people of this area need explanation and some knowledge of their disposition. This memo-randum was supressed at the time and was never produced during the construction license hearings when citizens asked for information under discovery. The Committee expressed a great deal of concern about..the high population within four miles of the plant site, as well as the minimal engineering safe-guards that were proposed in the application and the use ofless than conserva- 'tive assumptions in the dose calculations. The Committee recommended that: (1) The facility should be equipped with adequate engineered safety features and protective systems; (2) the facility should be designed cufficiently conservatively - particularly in respect to determination of exclusion area and low population zone; assurance should be provided of low potential doses at short distances from the reactor in the un-likely event of a serious accident; evaluation made of the number and location 1 of people who could be safely and quigkly evacuated in such an event; and, use of conservative assumptions, for example, for those related to meteor 61ogy, in dose' calculations; (3) the facility should be designed, constructed and utilized sufficiently conservatively; and (4) the facility should be provided with thoroughly structured, effective emergency plans, including evacuation plans. l Other important issues that were a part of this original letter included the statement that the containment of these reactors was designed with a leakage rate that was greater than for most other reactors of this type. The Committee expressed some questions as to the suitability of B&W reactors for marginal sites of this type. They were concerned ab'out the prote,ction required against reactor vessel splits and cavity flooding systems. This ACRS Committee was the first to question the use of process steam in products to be consumed by people. We believe that the population here is entitled to a clear and complete ex-planation of how these original questions have been met and resolved since we were prevented, because of supression of this document, from examining these l Issues in the construction license hearing. Also, in 1969 when these reactors were being designed, the ACRS stated in a letter to Dr. Glen Seaborg who was then chairman of the ACRS that there was urgent need for additional research and development in a number of signift-cant areas for the kind oflarge-s! zed reactors similiar to the Midland nuclear i plant that were being planned for construction at that time.
? 4 Page Two Statement before the ACRS May 20, 21,1982 These areas included, gaining an understanding of modes and mechanisms of fuel fa!!ure, the possible propagation of fuel failure, the generation oflocally high pressures if hot fuel and coolant are mixed and gaining an understanding of the various mechanisms of potential isnportance in describing the course of events following partial or large scale core melting. It is important for this Committee to describe for the public of this area to what extent all these areas, which urgently needed research attention according to the ACRS,have been met at the Midland nuclear plant. We do know now that 10 years after these recommendations were made that the Three Mile Island accident demonstrated dramatically the fact that little was known by the Nuclear Regulatory Commission about the course of events following partial or large scale core melting. We need to know how many of these*a'r'e'as have been similarly neglected. Since these plants are similar to the one at the Three Mile Island accident, the public needs assurance that all the lessons learned at Three Mile Island have been carefully evaluated and incorporated in the nnal construction of these plants. The kinds of design deficiencies of the B&W plants th'at were demonstrated by the Three Mile Island accident must be shown to have adequate and conserva-tive engineered compensations or this, Committee should lower power output. . c cm m esa The electrical qualification deficiencies that were found in the numerous critical safety systems at Three Mile Island should all be reviewed and resolved at this plant. These include systems relating to: Core Flood, Containment Spray, Emergency Core Cooling, Auxiliary Feedwater, Nuclear Service Water. l Containment Isolation, Decay Heat Removal, and Containment Cooling. l The consequences of replacing the steam generators,which is considered inevitable during the lifetime of pressurized water reactors, should be evaluated for the high radiation dose such an operation can inflict on nearby populations and workers within the Dow plant who are actually within the exclusion area. This ACRS meeting has been scheduled while the soil settlement hearings l are still undezway. In fact, the licensing board has halted construction in some areas because even the proper proceduies for quality assurance for underpinning remedial work has not been agreed upon. By meeting at this time for a Snal safety review, this ACRS Committee is prejudicing the findings of the licensing board's review on what is regarded as one of the most serious safety and quality control problems in the nuclear industry in the country. ,_m_. _ _.-
4 1 e Page Three Statement before the ACRS May 20 -21,1982 Finally, it tests the limits of my credibility to believe that this Committee has been able to read, assimilate and evaluate the signincance of the 800-page SER Report that was issued only one week ago as the anal safety evaluation for these plants, ,y u. L.w a' t.6O Mary P.ginclair, M.S. 4 e
4 ~ 5795 North River Road Freeland, NE 48623 May 20, 1982 D. C. Fischer Members ACRS re: Midland Huclear Plant Dear Mr. Chairman ~ As an intervenor in the ongoing OM-OL hearing, I have studied the soil settlement and QA issues over the last two years regarding the Midland Naclear Plant. I was unable to attend the ACRS meeting in k'ashington April 29, 1982, but I have reviewed Consumer's and the NRC soils input to that meeting. Based on their presentations, I have a brief statement to make covering some important issues which I believe have been misrepresented i or overlooked regarding soil settlement. (The page and exhibit numbers l re ferenced. : therein are from the OM-OL hearing transcript. I will gladly provide the documents themselves, if so directed by this cons; mittee for their review.) I respectfully request the opportunity to si=ilary critique state-ments made by the NRC and Consumer's Power Co. at the ACRS meetings. today and tomorrow. I would like to be able to present my position in writing, or otherwise before this com=ittee at the conclusion of these meetings. Sincerely, f LL lML [t rt; M) Barbara Stamiris
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V u e i l l INTERVE:;0R RESPONSE TO CONSUMERS FOM COMPANY'S k-19-82
SUMMARY
OF SOILS RELATED ISSUIS FOR ACBS According to Cons =er's su==ary.. soils related proble=s were initially iden-tified in July 3 1978,when excessive settlement of the Diesel Generator hilding (DG3)wasnoted. Yet the evidentiary record of the soils settlement hearing shows that the soils related proble=s were identified in late 1977, prior to the October 1977 co=ence=ent of the in3. Censu=er's own March 15,1982, findings (p.221)ad=itthat,"evidenceexisted in 1977, which if given different weight, would have revealed the plant wide soils conditions in time to have prevented the proble=s which now confront us." The soils evidence available prior to DG3 construction consisted of: 1)theAugust 1977 Ad=inistration hilding grade bea= settlement, 2) knowledge of site wide U. S. Testing failures (p. 1312-131L) and 3) F77-32 Audit findings revealing Q soil place =ent deficiencies as a result of a review of some 2000 soils records. (seeattach=ents). Consu=ers not only chose to proceed in the face of this infor-mation in 1977, but continues to defend these actions and decisions in 1982. bring the spring resu=ption of construction at the DG3, scribe settlement =arkers were placed fro = March to May,1978!. According to the NRC inspector, construe tion survey crews were unable to close a traverse while placing some of the upper building elevations, and this caused them to look to the settlement mon $toring program (p.2375). The settlecent monitoring reading of July,1978, confirmea the excessive settlement. The building was about 60% complete at this time, but construe tion continued until late August when ultimate lifetime settle ent values of the FSAR were exceeded and the settle =ent was first written up as a non-conformance (I&E 78-20). o
v 4 7.~ - s 2 Although Consu=er's discussed the Ad=inistration Building settle =ent problem with its soil consultants in the fall of 1978 (p. h075), they withheld this infor-mationfrom'theNRCduringtheirsoilsinvestigations(p.2595)despitetheiden-tical soil specifications for the two areas (ILE 78-20). Consuier's states that "after a th= rough review of the options available, the applicant elected to institute a surcharge leading progra=, which subsequently was started in January 1979." Yet Consu=er's own docu=ents reveal that surcharge instrunentation was purchased in October 1978. 'me instrmentation was installed, - eenstruction of the DG3 resu=ed, electrical duct ba.ks were released, and sand surcharge application began, in Nove=ber 1978. (Starfris exhibits 8, 7, & 30). 'Ihese activities and other docu=ented state =ents confir= that the only alternative option (according to CPC consultants 11-7-78), 'Re= oval and Replace =ent,' was never seriously considered in 1978. In fact Consu=er's 1979 reply to NRCs 50 5hr question 21 defends their choice of the Preload option over Re= oval and Replace =ent as "the least costly feasable alternative for corrective action. Also construction of the structure can continue while the surcharge load is being applied. Thus, this alternative will minimize the i= pact on the construction schedule." As a part of this sa=e response, Con-su=er's believed that the preload option would avoid costly devatering. Eindsight confir=s that the very costs and delays Consu=er's sought to avoid in 1977 sad in 1978 have =ultiplied as the issue of the IG3 integrity re=ains unre-solved in 1982. Unfortunately Consu=er's cannot learn from these past mistakes, because they do not consider them to be =istakes, as Consu=er's present Q. A. =anage=ent ' defends and justifies these decisions as valid in 1982. l In 1979 and 1980, having full knowledge of soil settle =ent proble=s, the push l to proceed with construction continued. Bechtel's own Kepler-Tregoe Analysis of 1979 identified the Ek'ST and Auxiliary Building R.E. A. as having soils deficiencies . s.-
v ( 3 and s' (Stamiris exhibit 29). But these deficiencies were judged insignificant, o construction began on the questionable soils. But The l'oading of the BWST's was undertaken in 1980 as a soils proof test. when overstress and cracking resulted in 1981, the problem was called a design y (NRC's J. Kane and Consumer's own consultant R. Kennedy, deficiency by Consumer's. however have called the problem soils related.) Further unconservative decisions took place regarding soil settlement matters in 1978 and 1979 as numerous consultant suggestions were not followed or were changed regarding such things as: 1) fhilure to break up the :Inud=at at the DGB prior to preload, 2) failu're to grout gaps prior to release of electrical duct banks, 3) failure to cut underlying condensate line until stress resulted, and
- 4) f ailure to delay filling of cooling pond till af ter surcharge was complete, 12-10-81 finding p. 50-66).
which affected piezemetric measurements.(Stamiris The measurement of the DGB surcharge effect was further compromised when Consumer's chose to eliminate what it termed "anomolous' Sondex data in 1979 (p. l 3455-7, Stamiris exhibit 14). Although Consumer's consultants reco== ended labora-tory tests at the DGB to justify bearing capacity (which was not expected to be at issue), they advised against the use of laboratory tests (borings) as early as November,1978 to confirm preload results, because of their potential for unpre-t i dictable results. Unconservative soils work decisions and a tendency to push ahead are still evident today in the soils remedial work at the Auxiliary Building as evidenced in I&E Reports 82-05 and 82-06, as well as the negative S.A.L.F. assessment rega ing soils work present April 20,1982, by the NBC. ._}}