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{{#Wiki_filter:.! (' ( M L A15' April 27, 1982 John C. Hoyle Advisory Committee Management Officer MEETING OF THE ACRS SUBCOM't!TTEE ON MIDLAND PLANT UNITS 1 AND 2, MAY 20 AND 21, 1982, MIDLAND, MI Attached is a proposed Federal Register notice regarding subject meeting. It nay be necessary to close portions of this meeting for the reason stated in the attached notice. Please publish this notice by Tuesday, May 4,1982. David Fischer Staff Engineer
Attachment:
Proposed FR Notice cc with
Attachment:
C. Siess, ACRS M. W. Libarkin, ACRS T. G. McCreless, ACRS J. C. McKinley, ACRS G. R. Quittschreiber, ACRS P. Davis, OGC J. T. Kopeck, PA E. Goodwin, NRR NRC Public Document Room 8510080448 850930 PDR FOIA g/g,g5-401 BRUNNER85-bO2 PDR \ N " FILE: Midland PLANT UNITS 1/2 8 'ssI l AC RS e mer)
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f , Federal Register Notice NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE ON MIDLAND PLANT UNITS 1 AND 2 Notice of Meeting g The ACRS Subcommittee on Midland Plant Units 1 and 2 will hold a meeting on May 20 and 21, 1982, at the HOLIDAY INN,1500 W. Wackerly Road, Midland, MI. The Subcommittee will review the application by Consumers Power Company for a license to operate Midland Plar.t Units 1 and 2. Notice of this meeting was published April 13. In accordance with the procedures outlined in the Federal Register on September 30, 1981 (46 FR 47903), oral or written statements may be pre-sented by members of the public, recordings will be pemitted only during those portions of the meeting when a transcript is being kept, and questions may be asked only by members of the Subcommittee, its consultants, and Staff. Persons desiring to make oral statements should notify the Designated Federal Employee as far in advance as practicable so that appropriate arrangements can be made to allow the necessary time during the meeting for such statements. The entire meeting will be open to public attendance except for those ! sessions which will be closed to. protect proprietary infomation (Sunshine ! Act Exemption 4). One or more closed sessions may be necessary to discuss l l such information. To the extent practicable, these closed sessions will be held so as to minimize inconvenience to members of the public in attendance. l The agenda for subject meeting shall be as follows: l Thursday, May 20, 1982 - 8:30 a.m. until 12:30 p.m. and 6:00 p.m. until 9:00 p.m. Friday, May 21, 1982 - 8:30 a.m. until the conclusion of business
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!4 l < During the initial portion of the meeting, the Subcommittee, along with any of its consultants who may be present, will exchange preliminary g views regarding matters to be considered during the balance of the meeting.
The Subcommittee will then hear presentations by and hold discussions with representatives of the Consumers Power Company, the NRC Staff, their consultants, and other interested persons regarding this review. Further information regarding tcpics to be discussed, whether the meeting has been cancelled or rescheduled, the Chairman's ruling on requests for the opportunity to present oral statements and the time allotted therefor can be obtained by a prepaid telephone call to the cognizant Designated Federal Em-ployee, Mr. David Fischer (telephone 202/634-1413) between 8:15 a.m. and 5:00 p.m., EST. I have determined, in accordance with Subsection 10(d) of the Federal Advisory Committee Act, that it may be necessary to close portions of this meeting to public attendance to protect proprietary information. The authority for such closure is Exemption (4) to the Sunshine Act, 5 U.S.C. 552b(c)(4). Date John C. Hoyle Advisory Comittee Management Officer
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- TM A S D. > 1 % L N R C. STA r C-Mtw3':'I : l RECEIVra 5/2/82 ADVISORY C0',T:TTEE ON REACTOR SAfEGUAES, U.S fi R.C JUN 141982 EU PM 7,8,9,101112,1 i 2i 3:4.5 6 A
NRC STAFF RESPONSES TO QUESTIONS BY THE ACRS SUBCOMMITTEE DURING MEETING OF MAY 20-21, 1982 ON MIDLAND PLANT, UNITS 1 AND 2
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la. What is the staff's criterion for turbine missiles? Answer The SRP Section 2.2.3 risk acceptance guidelines that are used for potential accident situations in the vicinity of the plant are and will continue to be used in determining the sufficiency of protection against turbine missiles. During the past several years the results of turbine inspections at operating nuclear facilities indicate that cracking to various degrees has occurred , at the inner radius of turbine disks, particularly those of Westinghouse design. Within t~n is time period, there has actually been a Westinghouse turbine disk failure at one facility - Yankee Atomic Electric Company. Furthermore, recent inspections of General Electric turbines have also resulted in the identification of disk bore cracks. In view of current experience and NRC safety objectives, the NRC staff intends to emphasize the turbine missile generation probability (i.e. turbine system j integrity) in its reviews of the turbine missile issue and eliminate the need for elaborate and somewhat ambiguous analyses of strike and damage probabilities given an assumed turbine failure rate. Although straightforward in principle, the latter calculations have to be based on detailed facility information and assumptions as to missile shape and size, missile energies, barrier penetration potential and ultimately to the likelihood of damaging a facility safety system. Generally, there are significant differences between licensees or applicants submittals and the final evaluation by the staff. Nevertheless, the staff concludes, based on our reviews of many facilities, that the probability of a turbine missile striking and damaging a safety system is in a relatively narrow range depending on turbine orientation. More refined , i analyses or additional calculations for other facilities are unlikely to change this conclusion. Therefore, expensive and time consuming strike probability analyses on the part of applicants / licensees and/or the NRC staff are judged to be unwarranted. This shift of emphasis requires all nuclear steam turbine manufacturers to develop volumetric (ultrasonic) examination techniques suitable for inservice inspection of turbine disks and shaft, and to prepare reports for NRC review which describe their methods for determining turbine missile generation i probabilities. These methods are to relate disk design, materials properties, and inservice volumetric inspection interval to the design werspeed missile generation probability, and to relate overspeed protection system characteristics, and stop and control valve design and inservice test interval to the destructive overspeed missile generation probability. i It should be noted that although evaluations of strike .and damage probabilities are not involved in following the proposed new procedures, the effect of these probabilities are taken into account in these procedures. The new procedures I 1 l
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are related to the NRC safety goal for turbine missiles (SRP Section 3.5.1.3) by taking the P2 P3 roughly in the range 10product (i.e.$
-4 to 10 - orf favorably oriented turbines and 10-Jthe to strike and 10-Z for unfavorably oriented turbines, for all plants in each category, and .
specifying degrees of unacceptable damage in terms of missile generation ' probability ranges and corresponding appropriate responses required of the applicant or licensee. lb. What is the status 07 the Midland Turbine Missile Protection Evaluation? Answer The applicant has made an evaluation of the turbine missile risk for Midland Plant Units 1 and 2. Based on their analysis, which uses General Electric calculated probabilities for the generatiog of missiles from design gnd destructive overspeed failure of 8.7 x 10- per year and 5.0 x 10- per y respectively, the probability of unacceptable damage for Unit 1 is 1.4 x 10 gar, per year and that for Unit 2 is 1.5 x 10-9 per year. However, based on the SRP Section 3.5.1.3 from design andrecommended missile generation destructive overspeed failure of probabiligies 6 x 10- for missiles per year and 4 x 10-5 per year, respectively, t Units 1 and 2 are about 1 x 10 ge probability per. of unacceptable These damage are two orders for both of magnitude above the NRC safety objective of 10-', year. per year. The applicant contends that their turbine inspection and test programs are either explicitly or implicitly incorpr rated in their evaluation and justify their use of the General Electric missile generation probabilities. It is the staff's position that the relevant General Electric analyses be submitted to the staff for review and acceptance in order to verify the adequacy of the applicant's turbine inspection and test programs. l l
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- 2. How does the staff define " adequate" core cooling?
Answer It is well established by calculations and experiments that adequate core cooling will occur after a reactor trip so long as a two-phase froth level (liquid level swollen by the presence of steam bubble) covers the reactor core. Thus, with the possible exception of brief intervals of complex cooling conditions associated with large break LOCAs, the existence of a collapsed liquid level above the core is evidence of sufficient coolant inventory to cover the core. The large break LOCA conditions are not a detriment to the dependability of vessel level information simply because the blowdown would be over too rapidly to pose a longstanding source of confusion. When reactor coolant pumps are running, adequate core cooling by pumped two-phase coolant will be maintained until depletion of coolant inventory well beyond the quantity required to cover the ccre after pumps have been shut off. Therefore, an indication of coolant inventory loss with pumps running is indicative of an approach to inadequate core cooling conditions. i See also the response to Question 3 a - d i l j
I 1 3a. What are the staff's criteria for direct measurement of bubbles in the vessel head? Answer The staff's requirements for ICC instrumentation, as defined in Item II.F.2 of NUREG-0737, are: (a) It must indicate the existence of ICC cause by various phenomena (i.e., high-void fraction-pumped flow as well as stagnant boil-off). (b) It must give advanced warning of the approach of ICC (i.e., inventory trending capabilility). (c) It must cover the full range from normal operation to complete core uncovery. 3b. What ways have other PWRS found for ICC? Answer Westinghouse's Reactor differential pressure, and Vessel Level Instrumentation CE's Reactor System (RVLIS) Vessel Level Measurement using(RVLMS) System using heated junction thermocouples (HJTC), have both been offered for inventory trending. Both systems, in conjunction with core exit thermocouples and subcooling margin monitors, appear to meet the staff's requirements for ICC instrumentation. 3c. Are these used for Midland? If not, why not? Answer They are not used for Midland. The applicant has proposed by FSAR Revision 38 to use a B&W Hot Leg Level Measurement System (HLLMS). Two trains of HLLMS are proposed for each of the Midland units to monitor the primary coolant level from the top of each hot leg. The proposed design is still in the preliminary engineering phase, and no detailed system description for HLLMS is provided in FSAR Revision 38. Unlike the Westinghouse RVLIS, the proposed HLLMS for Midland has no dp tap at the top or bottom of the vessel. Therefore, the proposed HLLMS would not indicate void formation in the reactor head until the vessel water level reaches the hot leg nozzle.
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( . 2-3d. What is the status of the staff position with regard to ICC instrumeritation requirements? Answer A briefing for the CRGR evaluation of TMI Action Plan II.F.2 requirements was given by the staff on March 24, 1982. As a result of the briefing, additional infomation addressing some open technical issues and a cost / benefit study for ICC instrumentation design requirements were requested. The staff expects to resolve those issues with CRGR in June 1982. The staff has also discussed several variations of the hot leg dp monitoring system with licensees and applicants for B&W reactors. However, detailed engineering descriptions and evaluations of the concepts have not been provided for staff review. Therefore, the discussion and preliminary evaluation of dp monitoring concepts is predicated on the following assumptions: (1) Proposed dp concepts can be shown to function in an acceptable manner with pumps tripped by calculations and testings. (2) Concepts which do not include dp across the core (vessel bottom tap) will not provide a reliable indicator for trending a loss of coolant inventory with the pumps running. (3) Concepts which do not include dp from the vessel head to hot leg will not provide indication of voiding in the reactor vessel head until the bubble extends to the top of the hot leg nozzle. (4) The detailed design of proposed systems will be accomplished in an acceptable manner with hardware which can be environmentally qualified. Our preliminary conclusions are that an acceptable dp monitoring system for B&W reactors must include the following: (1) A dp transmitter between the vessel head and the hot leg designed to indicate voiding in the vessel head and to track vessel level to within 5 feet of the top of the core (based on lower level of the hot leg nozzle in some reactors); (2) a dp transmitter from the top of the candy cane to a level in the hot leg which is sufficiently low to distinguish between the most severe overcooling transient and a loss of coolant inventory; and (3) a dp transmitter sensing pressure change from a tap at the bottom of the vessel and designed to trend voiding with the pumps running, or a pump current monitor for level trending with the pumps running. l l l l i
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- 4. Did the staff conduct a thorough review of internal flooding? <
Answer ( Yes. The staff reviewed internal flooding at the Midland Plant from sources inside and outside the containment. ' (a) Flooding Inside Containment Each Midland containment, including the reactor vessel cavity, is designed to direct all leakage to the containment sumps which are situated at the lowest point inside containment. As discussed in SER Section 5.2.5, the two separate, adjacent sumps are 70 inches deep and have a low level alarm at 18 inches, corresponding to a release of 1600 gallons. The sumps also have a rate of change alarm set at 3/4 inch per hour
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The design for containment water level monitoring after an accident is addressed in SER Section 6.2.8 (NUREG-0737 item II.F.1). The water level instrumentation at the Midland Plant has a range from the bottom of the sump to 10 feet (600,000 gallons) above the reactor building floor. The maximum calculated water level following a LOCA is calculated to be 9.5 feet above the reactor building floor (which is just below the bottom 4 of the reactor vessel). The sensitivity of the water level instrumentation is such that is can detect a 1 gpm leak in one hour. The staff reviewed leakage detection capabilities directly associated with service water to containment air coolers after the flooding incident at Indian Point, Unit 2. Each of the Midland containment air coolers is provided with a drain pan with a high flow alarm on the pan drain line. The high flow alarm annunciates and records leakage from the service water system into the containment. The flow into and out of each pair of air cooling units is also indicated in the control room. - The differential flow is recorded and alarmed in the control room. In summary, with the above sump design features at the Midland plant, the Staff concludes that a small leak in a system inside containment could be readily detected long before filling the sumps. (b) Flooding Outside Containment The results of the NRC staff's review regarding flooding outside containment (performed under various SRP sections) and its evaluation are provided in Sections 3.4.1, 3.6.1, 9.3.3 and 10.4.5 of the SER. Flooding sources included in the evaluations were piping and tanks, both seismic and nonseismic, which are the major contributors. i
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- 5. a) What are the staff's criteria for requiring a PRA?
b) How does the population distribution for the Midland site, especially within the first few miles, compare with other sites? Answer a) The staff presented a paper (SECY 81-20) to the Commission which proposed the PRA's or other types of special analyses be performed on a priority basis for high population density sites. The analysis in that paper which identified those sites is attached (Attachment 1). The staff recommended that PRA's be performed on an expedited basis for those sites in Groups IV and V, designated as "Above Average" and "Significantly Above Average", respectively, but that all sites eventually be included as part of the NREP program. As can be seen from the attached analysis, the Midland site falls in Group III, "Slightly Above Average", b) The 1970 residential population data for the Midland site for various distances out to 30 miles are shown together with similar data for several other sites on the accompanying Table 5-1. Indian Point, Zion and Limerick have been identified as the three sites which comprise Group V, designated "Significantly Above Average", in SECY 81-20. The Palisades site is representative of a typical or average site with regard to population. Also shown are the population values corresponding to 500 people per square mile, as given in Regulatory Guide 4.7. If at the time of CP review a site is projected to exceed these values at plant startup, then alternate sites having lower population densities should be considered. The general conclusions that can be gained form this table indicates that:
. within 3 to 5 miles from the reactor the Midland site is among the highest in population density and that these values are a!:0 in excess of Reg. Guide 4.7.
. beyond 10 miles from the reactor, the Midland site is close to average in population density.
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iI ' l Table 5-1 1970 RESIDENT POPULATION
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Dist., Miles Midland Indian Pt. Limerick Zion Palisades 500/MI2 , 750 480 790 51 1,570 0-1 64 4,394 4,900 6,900 320 6,280 0-2 9.300 20,000 19,000 19,000 1,800 14,140 0-3 24,973 35,000 52,000 33,000 3,700 25,130 0-4 40,223 53,000 67,000 46,000 5,800 39,270 0-5 48,500 220,000 150,000 190,000 30,000 157,000 0-10 72,700 890,000 780,000 530,000 130,000 628,000 0-20 304,750
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481,100 4,000,000 3,800,000 1,300,000 220,000 1,413,000
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Attachment 1 Prioritization of Sites with Regard to Population Density
- 1. Introduction ,
In comparing and evaluating the population around nuclear power reactor sites, ( the staff has long recognized that the population characteristics of a site, that is, its density and distribution, are a relatively crude measure of the consequences associated with the accidental release of radioactivity. The residual risk from an accident would depend not only upon the population den-sity of the site, but also upon many other factors, such as reactor design, onsite and offsite management and technical support resources, external hazards, liquid pathway considerations, meteorological conditions at the time of the accident, and effectiveness ar$f nature of public protective actions taken. In addition, the risk is not unifonn for all members of the population regard-less of distance from the site, but would be higher for those persons relatively close to the site, and would generally decrease with distance away from the site. An analysis has been carried out to obtain a first-order prioritization of i sites based upon population density and distribution. The discussion that follows outlines the rationale and methodology used and gives the results of this analysis. l
- 2. Methodology ,
l In carrying out this analysis, the following asstanptions and methodology were used: l _,. - . .- . __
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, (a) All sites where a reactor was either in operation, under construction, j or where a construction pemit was presently under active review were evaluated. This involved a total of 93 sites. .
(b) The population data used were taken from NUREG-0348, based on the 1970 census. The population data for the Femi site as reported in NUREG-0348 are in error and were corrected for this analysis by a special l computer ran of the 1970 c2nsus tape. (c) Although it is well-known that individuals closer to the reactor are at a higher level of risk, given an accident, than those more remotely located, the precise quantification of the variation of risk with distance is still somewhat uncertain. For the purpose of this analysis, the distance weighting given by the Site Population Factors (SPF), as given in WASH-1235, were used. Further, population bepnd 30 miles was neglected, i because the consequences at distances within 30 miles were considered to dominate any considerations of overall societal impact, and beyond 30 miles L the potential population exposure differences from site to site become less , i sharp. Preliminary analyses carried out by the staff have indicated that somewhat differing weighting schemes, or the factoring in of population out to 50 miles, does not change the resulting prioritization of sites to a significant degree. (d) The power level of the largest reactor at the site was multiplied by the
$PF value to account, in a first-order way, for the variation of reactor l fission product inventory from site to site. Only one reactor at a site was considered, even where multiple reactors exist or are contemplated,
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because the probability of an accident involving more than one reactor simultaneously was considered negligible. Although it can be argued that the population around a 4 reactor site is at a higher level of risk than those around a single reactor site, the prioritization of sites is , intended to give a measure of the relative consequences, given that an
' accident has occurred. The number of reactors at a site presumably effects only the probability of an accident. Also, it could be argued that a multi-reactor site would have some attributes that would reduce risk, compared to a single-reactor site, because of greater management i
and technical resources that can be applied to reducing either the likeli-hood or consequences of an accident. Using the above methodology, the reactor power level times the SPF value was calculated and tabulated for each of the 93 sites considered. The results are discussed below.
- 3. Results i
The reactor power level times SpF (P x SpF) was calculated for each of the 93 sites. The resulting values ranged from a high value of 2980 to a low
- - value of 6. The median value is 206; and the median site has a population of less than 100 persons per square mile, which is almost a factor of two i
1ess than the population of the average site. The sites are not listed in l numerical order, since this would imply a greater degree of precision than is warranted by the uncertainties in the analysis. Also, as pointed f out previously, the residual risk at a particular site cannot be measured in tems of consequences alone, since plant design and other factors are important contributors to risk. Therefore, we decided to place each site P
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into one of five groups or categories. The variation within a given group was selected to be sufficiently small so that each site within that group is considered to have about the same ranking. In selecting the groups we decided to use the median value and factor of two varia- : tion about the median to demarcate the " average" group boundaries. The , other groups were chosen as indicated below. Title Range Grouc No. I I Below Average PXSPF less than one-half the median value (PXSPF < 100) II . Average PXSPF between one-half and twice the median value (PXSPF from 100 to 400) III Slightly Above PISPF between twice and four Average times the median value (PXSPF from 400 to 800) Above Average PXSPF between four and eight IV times the median (PXSPF from 800 to 1600) I V Substantially Above PXSPF greater than eight times Average the median (PXSPF > 1600) l Within each group the sites have been listed in alphabetical order, as j shown in the following tables. Group V - Substantially Above Average
- 1. Indian Point
- 2. Limerick
- 3. Zion
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5-5 Group IV - Above Average 1
- 1. Bailly 5. Seabrook
- 2. Beaver Valley 6. Shoreham
. 3. Fenni 7. Three Mile Island ,
- 4. Millstone 8. Waterford Group III - Slightly Above Average
- 1. Byron 11. Peach Bottom
- 2. Catawba 12. Perkins
- 3. Cook 13. Pilgrim 4'. Cherokee 14. Perry
- 5. Erie 15. Salem
- 6. Forked River 16. Sequoyah
- 7. Haddam Neck 17. Susquehanna
- 8. Hope Creek 18. Rancho Seco i
. 9. McGuire 19. Turkey Point
- 10. Midland 20. Zimmer Grouc II - Average-
- 1. Arkansas 21. Palisades
- 2. Bellefonte 22. Phipps Bend
- 3. Black Fox -
- 23. Preirie Island
- 4. Braidwood 24. Quad Cities
- 5. Browns Ferry 25. River Bend
- 6. Calvert Cliffs 26. Robinson i
- 7. Clinton 27 . San Onofre
- 8. Brunswick 28. Shearon Harris
- 9. Davis-8 esse 29. Summer
- 10. Duane Arnold 30. Surry
- 11. Fort Calhoun 31. St. Lucie
- 12. Fitzpatrick 32. Skagit
- 13. Ginna 33. Trojan
- 14. Hartsville 34 Vogtle
- 15. LaSalle 35. Matts Bar
- 16. Maine Yankee 36. WPPSS 3/5
- 17. Marble Hill 37. Vermont Yankee
- 18. Nine Mile Point 38. Monticello
- 19. Oconee 39 Yellow Creek 20 Oyster Creek .
1Bailly and Millstone Unit 3 are the only plants in Group IV that are in the early stages of construction. l
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Group I - Below Average
- 1. Allens Creek 13. Kewaunee
- 2. Big Rock Point 14. Lacrosse *
- 3. Callaway 15. North Anna ;
4 Comanche Peak 16. Palo Verde
- 5. Cooper 17. Pebble Springs
- 6. Crystal River 18. Point Beach
- 7. Diablo Canyon 19. South Texas
- 8. Dresden 20. WPPSS 2
- 9. Farley 21. WPPSS 1/4
- 10. Ft. St. Vrain 22. Wolf Creek
- 11. Grand Gulf 23. Yankee Rowe
- 12. Hatch i
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( ( 6 .' Does the Staff have criteria on prioritization of alarms? Answer Staff provides guidelines on prioritization of alarms as presented in NUREG-0700,
" Guidelines for Control Room Design Reviews," Section 6.3 Annunciator Warning ,
Systems, Subsection 6.3.1, General System Characteristics. The specific guideline follows:
"6.3.1.4 Prioritization Because of the large number of annunciators typically found in control rooms and the likelihood that numerous alarms may come in concurrently, some logical prioritization should be applied such that operators can differentiate the most important or serious alarms from less important ones.
- a. Levels of Priority (1) Prioritization should be accomplished using a relatively small (2-4) number of priority levels.
(2) Prioritization should be based on a continuum of importance, severity, or need for operator action in one or more dimensions, e.g., likelihood of reactor trip, release of radiation. Exhibit 6.3-3 (see below) provides an example of prioritization based on three levels of prioritization. First Priority Alarms
. Plant shutdown (reactor trip, turbine trip).
. Radiation Release
. Plant conditions which, if not corrected inmediately, will result in autvaatic plant shutdown or radiation release, or will require manual plant shutdown.
Second Priority Alarms
. Technical specification violations (other than those associated with first-priority alarms) which if not corrected will require plant shutdown.
. Plant conditions which, if not corrected may lead to plant shutdown or radiation releases.
Third Priority Alarms
. Plant conditions representing problems (e.g., system degradation) which affect plant operability but which should not lead to plant shutdown, radiation release, or violation of technical specifications.
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- b. Priority Coding (1) Some method for coding the visual signals for the various priority levels should be employed. Acceptable methods for priority coding include color, position, shape, or symbolic coding.
l (2) Auditory signal coding for priority level is also appropriate. See Guideline 6.2.2.3 for recommended coding teheniques." Sumarizing the staff's guidelines, prioritization should be: j 1. Accomplished using a relatively small number of levels;
- 2. Based on a continuum of importance, severity or need for operator action; and
- 3. Presen^ed to the operator through use of coded signals.
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- 7. What are the staff's criteria or requirements for ICC controls under .,
conditions where the control room has been evacuated, e.g., due to 1 a fire? Answer There are no criteria for specific ICC controls on the alternate shutdown panel. However, the shutdown panel does record primary system pressure and temperature, f
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- 8. What is the basis for the Staff's finding that manual operation of the Decay Heat Removal valves is acceptable?
Answer Section 5.4.4 of the SER provides the Staff's conclusions regarding manual actions outside the control room necessary for achievement of cold shutdown at the Midland Plant. The Midland DHR system design requires local operator action to align the DHR suction valves from the reactor building sump to the RCS hot leg before the system can be brought into service. The staff's review of the Midland DHR design has been performed recognizing that manual action outside the control room in the absence of a postulated single failure is, in general, not consistent with RSB Branch Technical Position 5-1 for Class 2 plants (i.e., plants with cps docketed before January 1,1978 and OLs issuance scheduled on or after January 1,1979). Two of the more significant factors of the Staff's evaluation are; (1) the time available for the action, and (2) accessibility of the operator to the valve. Review of the latter consideration is continuing.
- a. Time Consideration The design of the DHR system requires manual operator action outside the room between 6 to 30 hours during a cooldown to align the DHR suction valves. Once the DHR system is aligned, it can be operated from the control room without further remote manual action. In view of the ample time available for operator action and the ability of the DHR system to be operated remotely once properly aligned, the staff concludes that the system meets the requirements of BTP RSB 5-1 and is therefore acceptable, subject to resolution of the accessibility item below.
- b. Accessibility The manual DHR valves are located in the lower level of the auxiliary building, six levels below the control room. The valves are equipped with reach-rods which pass through a concrete wall between the auxiliary building hallway and the room housing the manual valves to reduce the radiation exposure to the operator from radioactivity which might be contained in the DHR water. In the SER, we note that the applicant is required to provide an evaluation of the environment which might exist in the vicinity of the valve hand wheels and in the passages which must be traversed between the control room and the manual DHR valves. The evaluation should consider all potential accident conditions (e.g.,
fire, radiation leaks in systems contained in the auxiliary building, small break loss of coolant accidents that are subsequently isolated and require RHR cooling) which might necessitate that the plant be brought to cold shutdown.
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- 9. Will the staff require SG overfill protection on operating B&W 177 i plants?
Answer As discussed in the work scope and schedule for Task II.E.5.1 of NUREG-0660, the staff will determine in the review of the modifications proposed for CP holders if the proposed modifications warrant backfit for operating plants. The staff has not at this time imposed any requirement that operating plants install additional SG overfill protection. It is also not known what hardware changes would be required, if any, to provide protection similar to that at Midland. Cost benefit studies in this area have not been performed to date. This issue was discussed in Recommendation 2 of NUREG-0667, " Transient Response of B&W - Designed Reactors," May 1980. The conclusion was that provisions to throttle or trip the auxiliary feedwater system to avoid grossly overfilling the steam generators are subject to failures that could isolate the reactor from its heat sink. The net effect of this type of overfill protection may increase risk. Subsequent staff review (Mattson memo to Denton, 8/8/80) agreed with the NUREG -0667 recommendation. Some operating plants (Rancho Seco, Crystal River 3, ANO-1) have proposed installation of the SG 1evel protection in an effort to reduce plant sensitivity as part of the AFW upgrade and program required by Item II.E.1.1 of NUREG-0737. The need for protection against steam generator overfill resulting from main feedwater control system failure is also being reviewed as part of Unresolved Safety Issue A-47 effort on safety implications of control system failures. l
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- 10. Are the probabilities of occurrence expressed in NUREG-0654 for reactor events and alerts consistent with experience?
Answer i There are no probabilities given in the final revision of NUREG-0654, although there were some in the draft document which was issued for interim use and comment in January,1980. In that draft the frequency for " Notification of Unusual Events" was given as once or twice per year per operating unit. With approximately 72 operating units, this would translate into between 72 and 144 " Notification of Unusual Events" per year. In the time period 1/1/82 - 5/27/82, the NRC Operations Center has logged 108 " Notification of Unusual Events". However, these 108 events have not been analyzed to determine if they would meet the general criteria given in Revision 1 to NUREG-0654. This would be necessary before any firm reliance could be placed on the data because experience has shown that many events are over-classified by licensees and what they term a " Notification of an Unusual Event" is in reality a reportable occurrence under 10 CFR 50.72. i In the draft version of NUREG-0654 the frequency of occurrence of an " Alert" was given as once in 10 to 100 years per unit and the frequency of a " Site Area Emergency" was given as once in 100 to once in 5000 years per unit. In the same five month period described above, there has been one alert (which was subsequently upgraded to Site Area Emergency). This was the Ginna incident. _.c.....-. , 7_. . . . . . . . -
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- 11. What is the basis for the Midland DES statement that seismic events and other natural phenomena do not contribute significantly to risk?
Answer The basis is that the WASH-1400 estimate of the probability of severe release due to earthquakes is 10 -6t o 10 -8 / reactor year. This probability is small (at the higher end) compared to the sum of the release probabilities in the Midland DES of 4.8 x 10 -o / reactor year. It is only about one-tenth of the sum of the probabilities of the three sequences that release the largest fractions of core inventory (8 x 10 -6 / reactor year). The staff did not evaluate the probability of a severe release due to a seismic event at Midland, nor did it determine the probability of severe releases caused by in-plant events for the Midland design. Rather, the WASH-1400 results were used, rebaselined as described in the DES Appendix E. The staff will revise the FES section in which the level of significance will be indicated to be the uncertainty of the risks presented in the statement. Further, the section on uncertainty will be expanded and the bounds of the uncertainty given as over a factor of 10 but not so large as a factor of 100. I l l 1
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- 12. Why is liver the critical organ for the fish consumption pathway and the recreation pathway shown in DES Table C.6?
Answer The major contributors to adult dose for the fish consumption pathway are the Cesium 134 and 137 isotopes. The relative uptake by different human organs of the various radioisotopes in the fish are such that the adult liver dose is about one-third higher than the total body dose, and about twice the dose to the bone. The more significant recreational pathway is expcsure from contaminated sediments. Because this is an external exposure pathway. DES Table C.6 indicates the same dose rate to the total body and internal organs. " Liver" is only one of several organs involved and vill be deleted for the FES. 'I th et 3 e e e e-**g_
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- 13. Why is H-3 not listed in Table 5.4 of the Midland DES?
Answer Table 5.4 contains nuclides used in the calculation of health effects following severe accidents. The contribution of H-3 health effects following i a severe reactor accident is negligible compared to the contribution to health effects of the 54 nuclides in the Table. The selection criteria used in the Reactor Safety Study to reduce the hundreds of nuclides actually present in the plant to manageable proportions for calculations includes: half life, total content, relative dose contribution within a chemical group. The factors considered in the relative dose contribution included: radiation type and energy, daughters produced. Consideration of the mass of primary coolant, about 2 x 108 grams, and the concentration of about one micro-Ci/ gram shows that the total content of tritium in a PWR plant is between 200 and 300 Curies. It is a beta emitter with very low end point energy, about 0.02 MeV. 4
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- 14. Page 5-58 of the DES states that "a groundwater pathway for public radiation exposure and environmental contamination that would be associated with severe reactor accidents was identified in Section 5.9.3 Exposure Pathways." However, this pathway does not appear to !
be identified in Section 5.9.3. Is this an error? (TR 580) ' Answer The reference on DES page 5-58 to Section 5.9.3 will be deleted in the FES. As noted on page 5-58 the groundwater pathway from severe reactor accidents 1 are associated with soluble radionuclides which might be leached and trans-ported with groundwater to downgradient domestic wells used for drinking, or to surface water bodies used for drinking, aquatic food and recreation.
.-_. _ - . - - . . ~ . . . . . . _ -
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( .- : ( l l 15. What are the staff's criteria regarding draining and flushing of systems? Answer { ! There are no specific requirements stating that tanks or systems must be drained and/or flushed to reduce dose rates in the region prior to maintenance. Hcwever, Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" (Rev. 3), states that " accumulations of crud or other radioactive i material that cannot be avoided within components or systems can be reduced by providing features that will permit the recirculation or flushing of ! fluids with the capacity to remove the radioactive material through chemical , .or physical action." (Section C.2.f.(3)). The applicant's ALARA program should ! contain provisions for minimizing the amount of personnel time spent in ! radiation areas. i
! IN the FSAR, the applicant states that equipment or components requiring l personnel attention will be designed; 1) to provide for remote draining or flushing of equipment containing radioactive material, and 2) to minimize
, the buildup of radioactive material and facilitate flushing of crud traps. 1 Prior to performing maintenance work on valves located in high radiation areas, I the applicant will drain adjacent radioactive components to lower the area j dose rates. Pumps containing radioactive liquids will be drained prior to maintenance. Other components, such as filters, demineralizers and tanks, which have the potential for containing radioactive liquids, will be provided ! with drains or spray taps for flushing and/or draining purposes. t Although the frequency with which these components are drained and/or flushed is not within the scope of the staff review, Midlands ALARA program states that equipment general design considerations are directed toward minimizing i radiation levels proximate to equipment or components requiring personnel attention. One way to minimize equipment radiation levels is through equipment flushing and/or draining. l t i i i I i . . . . . . . . . . . _ , ~ , . ,..
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' { I, t 'l 3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545 j JUL
- 1 1967 Honorabic Glenn T. Seaborg Chairman J. S. Atomic Energy Comission Washington, D. C.
Subject:
REPORT ON OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3
Dear Dr. Scaborg:
At its eighty-sixth meeting, on June 8-10, 1967, and its eighty-seventh meetin3, on July 6-8, 1967, the Advisory Committee on Reactor Safeguards reviewed the proposal of the Duke Power Company to construct the Oconee Nuclear Station, Units 1, 2, and 3, at a site near Cicmson, South Carolina. This project uas reviewed by an ACRS Subcomittee on llay 2,1967, at the site and at Cicsson, and on liay 31 and June 23, 1967, in Washington, D. C. The Comittee had the benefit of discussions with representatives of the
;I - Duke Power Company and its consultants, The Babcock and Wilcox Company, Bechtel Corporation, and the AEC Regulatory Staff, and of the documents listed.
Each unit of the Oconee Station includes a pressurized-water reactor rated at 245211Ut. Each unit is to be provided with an emergency core cooling system (ECCS), including two core flooding tanks, three high-pressure in-jcction pumps, and three low-pressure injection and recirculation pumps. The applicant proposes not to operate a unit with a core flooding tank valved off. The Co::tsittee recommends that the Regulatory Staff review ' the detailed design of the ECCS and the analysis of its performance for the entire spectrum of break sizes, as soon as this information is avail-able. In this recpect:
- 1. The Regulatory Staff should review analyses of possible, effects, upon pressure-vessel integrity, arising from thermal shock induced by ECCS operation.<
- 2. The effcets of blowdown forces on core and other primary s system components should be analyzed more fully as de-tailed design pro,ceeds.*
. 3. Further evidence should be obtained to show that fuel-rod failure in loss-of-coolant accidents will not affect significantly the ability of the ECCS to prevent clad melting.*
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-2 JUL 11 1967 Honorable Glenn T. Seaborg 7
- 4. The applicant has proposed adding swing-check valves in the core barrel to ensure obtaining adequate height of v cooling water in the core under all circumstances of ECCS operation. This feature should be further reviewed to ensure that no new problems are introduced.
- 5. The applicant will explore further possibilities for improvement, particularly by diversification, of the instrumentation that initiates ECCS action.
" Emergency power sources for the ECCS and other safeguards are: (a) the other Oconee units (each unit can withstand and will be tested to with-stand instantaneous loss of load without a reactor trip or a turbine trip); (b) two hydrocicetric units at Keowce station less than one mile away, with independent overhead and underground transmission lines; and (c) a gas-turbine unit thirty miles away with independent transmission line, transformer, and switchyard -- all in addition to the usual multi-pie ties to the power transmission grid. The applicant stated that switching and sequencing of sources, buses, and loads would be such that no single failure would impair system availability. <<
l The applicant stated that the entire primary system of each unit, includ-ing the inside and outside of the reactor vessel, will be accessible for inspection over the life of the plant. The Committee continues to emphasize the importance of quality assurance in fabrication of the primary system as well as inspection during service life, and recommends that the applicant bsplement those L:provements in primary system quality that are practical with current technology.* The moderator coefficient of reactivity is calculated to be positive at the beginning of core life, for the first core. The applicant is making detailed studies of the effect of this coefficient on the course of postu-lated accidents; if necessary, the coefficient will be made more negative by the addition of solid poison shims to the core. Further evidence should be obtained concerning the ability of the fuel to withstand expected transients at the end of its anticipated lifetime.* The applicant is investigating further the stability margin for xenon oscillations. The containment structures are similar to those for the Turkey Point re-actors previously reviewed. Consideration should be given to improved
' inspection of welds in the steel liner of such containments, because an acceptance pressurization test does not stress the liner to postulated accident conditions.
( f tionorable Glenn T. Seaborg JUL 11 1967 > Power for the reactor protection systems and the safeguards protection systems for all three units is provided by a system of six batteries, static inverters, and six buses. The came batteries, via other inver- # ! ters and busca, provide, power to the control systems for all three units. The Committec urges the applicant to review the design of these systems with respect to independence of each unit from troubles in the others. The applicant proposes to construct a submerged earthen weir in the in - take canal to nacure a heat sink in the event Keowce Reservoir is drawn down excessively. The Committec believes that careful attention is' neces-sary in the design and construction of this weir to avoid hydraulic erosion and soil instability, particularly in case of rapid drawdown. The Advisory Caucittee on Reactor Safeguards believes that the items raen-tioned abovo can be resolved by the applicant and the Regulatory Staf f during construction of the reactors. On the basis of the foregoing com-monts, the Committee believes that the proposed Oconec Nuclear Station can be constructed with reasoncble assurance that it can be operated without undue risk to the health and safety of the public. Sincerely yours, i ORIGINAL SICBID B'l N. J. PALS.'af.*i] N. J. Palladino Chairman
*The committeo believes that these matters are8 si nificant for all large water-cooled power reactors, and warrant careful attention.
References:
1. Duke Power Company, Oconee Nuclear Station, Units 1 and 2 Preliminary Safety Analysis Report, Volumes I and II undated, received December 5, 1966.
- 2. Amen &nent No. 1, dated April 1, 1967.
- 3. Amendment No. 2, dated April 18, 1967.
- 4. Amen &nent No. 3, dated April 29, 1967.
- 5. Amendment No. 4, dated May 25, 1967. ,
- 6. Amendment No. 5, dated June 16,'1967. !
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' L 1 1 1967 Oconce Nuclear Station, Units 1, 2, and 3 -
- i ACRS OFFICE COPIES O!EY
- 1. DRL Preliminary Staff Analysis dtd 2/13/67 (OUO). ;
- 2. Inst. for Atmospheric Sciences consents dtd 1/26/67.
- 3. DnL ltr to Duke dtd 5/11/67.
- 4. Fish & Wildlife Service comments dtd 4/24/67.
f- 5. DRL Staff Analysis Report No.~ 1 dtd 5/24/67 (0U0).
- 6. j'!).nteractions. Among Units", undated, received 6/8/67.
- 7. Engineered Safeguards and Emergency Power", undated, received 6/8/67.
- 8. F&WS comments dtd 6/7/67.
- 9. DRL Staff Analysis Report No. 2 dtd 6/16/67 (0U0).
- 10. Inst. for Atmospheric Sciences comments dtd 6/9/67.
- 11. C&GS comments dtd 6/16/67.
- 12. GS comments dtd 6/19/67. .
- 13. Neumark & Hall comments dtd June 1967.
- 14. Hall ltr to !!crris dtd 6/23/67.
- 15. Addendum to DRL Rpt. No. 2 dtd 7/6/67 (0U0).
- 16. Newmark Associates comments dtd 7/6/67 on Submeaged Weir.
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o __-_._._-.-...m. t: . re n ...p . ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY CCMMISSION . WASHINGTON. D.C. s0HS July 11, 1967 vc ' Honorable Glenn T. Seaborg Chairr.on U. S. Atomic Energy Commission Washington, D. C.
Subject:
REPORT ON OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3
Dear Dr. Seaborg:
At its eighty-sixth meeting, on June 8-10, 1967, and its eighty-seventh meeting, on July 6-8, 1967, the Advisory Committee on, Reactor Safeguards reviewed the proposal of the Duke Power Company to construct the Oconee Nuclear Station, Units 1, 2, and 3, at a site near Clemson, South Carolina. This project was reviewed by an ACRS Subcommittee on May 2,1967, at the site and at Clemson, and on May 31 and June 23, 1967, in Washington, D. C. { The Committee had the benefit of discussions with representatives of the Duke Power Company and its consultants, The Babcock and Wilcox Company, Bechtel Corporation, and the AEC Regulatory Staff, and of the documents i listed. , Each unit of the Oconee Station includes a pressurized-water reactor rated l st 2452 MWt. Each unit,is to be provided with an emergency core cooling system (ECCS), including two core flooding tanks, three high-pressure in-jection pumps, and three low-pressure injec' tion and recirculation pumps. The applicant proposes not to operate a unit with a core flooding tank valved off. The. Committee recommends that the Regulatory Staff review the detailed design of the ECCS and the analysis of its performance for the entire spectrum of break sizes, as soon as this information is avail-able. In'this respect:
- 1. The Regulatory Staff should review analyses of possible
~
effects, upon pressure-vessel integrity, arising from thermal shock induced by ECCS operation.* 2 The effects of blowdown forces on' core and other primary f: system c'omponents should be analyzed more fully as de-tailed design proceeds.*
- 3. Further evidence should be ubtained to show that fuel-rod failure in loss-of-coolant accidents will not affect significantly the ability of the ECCS to prevent clad melting.* .
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in. c ' .a I ~ ,' bd' M July 11, 1967 > Hono-cbis Glenn T. Seaborg g e 4 l-i
- k. The applicant has proposed adding swing-check valves in l the core barrel to ensure obtaining adequate height of j
cooling water in the core under all circumstances of r l ECCS operation. This feature should be further reviewed ,~ i' to ensure that no new problems are introduced. l ? 5. The applicant will explore further possibilities for improvement, particularly by diversification, of the l instrumentation that initiates ECCS action. Emergency power sources for the ECCS and other safeguards are: (a) the other Oconee units (each unit can withstand and will be tested to with-l i stand instantaneous loss of load without a reactor trip or a turbine j trip); (b) two hydroelectric units at Keowee station less than one mile away, with independent overhead and underground transmission lines; and l
- (c) a gas-turbine unit thirty miles away with independent transmission j line, transformer, and switchyard -- all in addition to the usual multi-pie ties to the power transmission grid. The' applicant stated that' switchi,g and sequencing of sources, bus'es, and loads would be such that l
~ no sing e f ailure would impair system availability.
~
I The applicant stated that the entire primary system of each unit, includ-j ing the inside and outside of the reactor vessel, will be accessible for 4 inspection over the life of the plant. 4 ! The Comittee continues to emphasize the importance of quality assurance l in fabrication of the primary system as well as inspection during service i life, and recommends that the applicant implement those improvements in
. primary system quality that are practical with current technology.*
*l The moderator coefficient of reactivity is calculated to be positive at l
{ the beginning of core life, for the first core. The applicant is making
- 2. detailed studies of the effect of this coefficient on the course of postu-lated accidents; if necessary, the coefficient will be made more negative I by the addition of solid poison shims to the core.
L Further evidence should be obtained concerning the ability of the fuel to i withstand expected transients at the end of its anticipated lifetime.* . ' ihe applicant is investigating further the stability margin for xenon j i oscillations.
- The containment structures are similar to those for the Turkey Point re-actors previously reviewed. . Consideration should be given to improved inspection of welds in the steel liner of such containments, because an
! acceptance pressurization test does not stress the liner to postulated l , accident conditions. i - l ' i %- - d3 ' t: L
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<s4u}L J iy 11, 1967 Monorable'Glenn T. Seaborg .
Power for the reactor protection systems and the safeguards protection e systems for all three units is provided The same by a batteries, system ofvin sixother batteries, inver- : l static inverters, and six buses. ! ters and buses, provide power to the control systems for all three units. l3 The Committee urges the applicant to review the design of these systems with respect to independence of each unit from troubles in the others. The applicant proposes to construct a submerged earthen weir in the in-take canal to assure a heat sink in the event Kcowee Reservoir is drawn down excessively. The Committee believes that careful attention is neces-sary in the design and construction of.this weir to avoid hydraulic erosion and soil instability, particularly in case of rapid drawdown. The Advisory Committee on Reactor Safeguards believes that the items men- . tiened above can be resolved by the applicant and the On the basis Regulatory of the f6tegoingStaff com-during construction of the reactors. ments, the Co=mittee believes that the proposed Oconee Nuclear Station can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public. Sincerely yours, ( .
/s/
N. J. Palladino Chairman for all large
*The Committee believes that these matters are significant water-cooled power reactors, and warrant careful attention.
References:
- 1. Duke Poser Company, Oconee Nuclear Station, Units 1 and 2, Preliminary Safety Analysis Report, Volumes I and II, undated, received December 5, 1966.
- 2. A7endment No.1, dated April 1,1967
- 3. Amendment No. 2, dated April 18, 1967.
- 4. Amendment No. 3, dated April 29, 1967.
- 5. Amendment No. 4, dated thy 25, 1967.
- 6. Ame,ndment No. 5, dated June 16, 1967.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS , UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 2054S July 11, 1967 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission ' Washington, D. C.
Subject:
REPORT ON OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3
Dear Dr. Seaborg:
At its eighty-sixth meeting, on June 8-10, 1967, and its eighty-seventh
. meeting, on July 6-8, 1967, the Advisory Comittee on, Reactor Safeguards reviewed the proposal of the Duke Power Company to construct the Oconee Nuclear Station, Units 1, 2, and 3, at a site near Clemson, South Carolina.
This project was reviewed by an ACRS Subcommittee on May 2, 1967, at the site and at Clemson, and on May 31 and June 23, 1967, in Washington, D. C. The Committee had the benefit of discussions with representatives of the I Duke Power Company and its consultants, The Babcock and Wilcox Company,
! Bechtel Corporation, and the AEC Regulatory Staff, and of the documents listed.
) Each unit of the Oconee Station includes a pressurized-water reactor rated at 2452 MWt. Each unit,is to be provided with an emergency core cooling j system (ECCS), including two core flooding tanks, three high-pressure in- l' j jection pumps, and three low-pressure injec~ tion and recirculation pumps. The applicant proposes not to operate a unit with a core flooding tank
- valved off. The, Committee recommends that the Regulatory Staff review the detailed design of the ECCS and the analysis of its performance for the entire spectrum of break sizes, as soon as this information is avail-able. In this respect:
- 1. The Regulatory Staff should rev'iew analyses of possible
- effects, upon pressure-vessel integrity, arising from
- thermal shock induced by ECCS operation.* ,
- 2. The effects of blowdown forces on core and other primary system components should be analyzed more fully as de-tailed design proceeds.*
- 3. Further evidence should be obtained to show that fuel-rod failure in loss-of-coolant accidents will not af fect significantly the ability of the ECCS-to prevent clad melting.* -
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- ' Mdnorable Glenn T. Seaborg July 11, 1967
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- 4. The applicant has proposed adding swing-check valves in f the core barrel to ensure obtaining adequate height of I cooling water in the core under all circumstances of ECCS operation. This feature should be further reviewed to ensure that no new problems are introduced. .
- 5. The applicant will explore further possibilities for l-improvement, particularly by diversification, of the i instrumentation that initiates ECCS action.
- Emergency power sources for the ECCS and other safeguards are: (a) the other Oconee units (each unit can withstand and will be tested to with-stand instantaneous loss of load without a reactor trip or a turbine trip); (b) two hydroelectric units at Keowee station less than one mile away, with independent overhead and underground transmission lines; and (c) a gas-turbine unit thirty miles away with independent transmission
+ line, transformer, and switchyard -- all in addition to the usual multi-
~
pie ties to the power transmission grid., The applicant stated that' i switching and sequencing of sources, buses, and loads would ia such that no single failure would impair system availability. The applicant stated tha't the entire primary system of each unit, includ-l ing .the inside and outside of the reactor vessel, will be accessible for inspection.over the life of the plant. The Connittee continues to emphasize the importance of quality assurance in fabrication of the primary system as well as inspection during service life, and recommends that the applicant implement those improvements in primary system quality that are practical with current technology.* The moderator coefficient of reactivity is calculated to be positive at the beginning of core life, for the first core. The applicant is making detailed studies of the effect of this coefficient on the course of postu-lated accidents; if necessary, the coefficient will be made more negative by the addition of solid poison shims to the core. Further evidence should be obtained concerning the ability of the fuel to
' withstand expected transients at the end of its anticipated lifetime.*
The applicant is investigating further the stability margin for xenon oscillations.- The containment structures are similar to those for the Turkey Point re-
- actors previously reviewed. , Consideration should be given to improved inspection of welds in the steel liner of such containments, because an acceptance pressurization. test does not stress the liner to postulated
- accident conditions.
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J ty 11, 1967 Monorable Glenn T. Seaborg i Power for the reactor protection systems and the safeguards protection 6 systems for all three units is provided by a system of six batteries, l l static inverters, and six buses. The same batteries, via other inver- !: l ters and buses, provide power to the control systems for all three units. !' The Committee urges the applicant to review the design of these systems l with respect to independence of each unit from troubles in the others. t I i The applicant proposes to construct a submerged earthen weir in the in-take canal to assure a heat sink in the event Keowee Reservoir is drawn i down excessively. The Committee believes that careful attention is neces-sary in the design and construction of.:his weir to avoid hydraulic erosion j and soil' instability, particularly in case of rapid drawdown. l The Advisory Committee on Reactor Safeguards believes that the items men- . tioned above can be resolved by the applicant and the Regulatory Staf f j On the basis of the fdregoing com-during construction of the reactors.
]
ments, the Committee believes that the proposed Oconee Nuclear Station can be constructed with reasonable assurance that it can be operated l 1
- without undue risk to the health and safety of the public.
Sincerely yours, l 1 : 1 i .
' - /s/ -
N. J. Palladino , ' Chairman
*The Committee believes that these matters are significant for all large ! water-cooled power reactors, and warrant careful attention.
Re ferences_: i f Duke Power Company, Oconee Nuclear Station, Units 1 and 2, Preliminary 1. Safety Analysis Report, Volumes I and II, undated, received December 5, - 1966.
- 2. Amendment No.1, dated April 1,1967
- 3. Amendment No. 2, dated April 18, 1967.
4
- 4. Amendment No. 3, dated April 29, 1967. !
j 5. Amendment No. 4, dated May 25, 1967.
- 6. Amendment No. 5, dated June 16, 1967..
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y , - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20S45
- January 17, 1968 l
l Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C. 20545
Subject:
REPORT ON THREE MII.E ISIAND NUCLEAR STATION UNIT 1
Dear Dr. Seaborg:
At its ninety-third meeting, January 11-13, 1968, the Advl.sory Committee on Reactor Safeguards reviewed the proposal of the Metropolitan Edison Company to construct Three Mile Island Nuclear Station Unit 1. This project had been considered previously at Subcommittee meetings he'.d on January 4,1968, in Washington, D. C., and on October 19, 1967, in Hershey, Pa. During its review, the Connittee had the benefit of discussions with representatives and consultants of the Metropolitan Edison Company, the Babcock and Wilcox Company, Gilbert Associates, Inc. , and the AEC Regula-tory Staff. The Committee also had available the documents listed below. The station is located on Three Mile Island near the cast shore of the Susquehanna River in Dauphin County, Pennsylvania, about 10 miles south-east of Harrisburg. Unit 1 is a pressurized-water reactor plant, rated at 2452 MWt, and is similar in design to the units already approved for . , construction at the Duke Power Company's Oconee Nuclear Station. Flood protection is to be provided at the site by suitable carth dikes. Two natural-draf t cooling towers are to be used for condenser-water cooling. The emergency core cooling system (ECCS) includes two core flooding tanks, two indeper. dent low-pressure systems, and two independent high-pressure systems. Two separate systems are provided for containment cooling. One system consists of three fan-cooling units, and the other consists of two spray systems. The applicant , stated that suitabic and periodic component and integrated system tests will be performed on these engineered safety features. To further insure low containment leak rates, a fluid block system and a containment penetration pressurization system are to be provided. Operation of the ECCS ~ is, initiated automatically by redundant low-pressure signals from transducers actuated by pressure in the two primary loops. The Committee reconmends that in the interest of diversity ,another method, C O
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P [- l y Honorable Glenn T. Seaborg January 17, 1968 different in principle from the one proposed, should be added to initiate this function. The diversity thus achieved would enhance the probability that this vital function would be initiated in the un14.kely event it is needed. The output circuit of the proposed reactor protection system consists of a single d-c circuit (bus) fed from two station batteries. Both feeders must be interrupted to de-energize the bus and drop all rods. Failure to interrupt either feeder, or any other event that prc' vents de-energizing the single bus, will inhibit dropping all the rods. The Committee believes this system can and should be revised to correct the deficiency. The revised design should be provided for review prior to installation of the protection system. The applicant has proposed using certain signals from protection instru-ments for control purposes. The Committee believes that control and protection instrumentation should be separated to the fullest extent practicable, and recommends that the applicant explore further the possibility of making safety instrumentation more nearly independent of control functions, Consideration should be given to the development and utilization of instru-mentation for prompt detection of gross failuce of a fuel element. I The applicant described the research and development work planned to confirm the final design of the plant. The Committee continues to emphasize the importance of work to assure that fuel-rod failures in loss-of-coolant accidents will not affect significantly the ability of the ECCS to prevent elad melting. The applicant is continuing studies on the possible use of part-length rods for stabilizing potential xenon oscillations. Solid poison shims will be added to the fuel elements if necessary to make the moderator temperature coefficient more negative at the beginning of core life. The Regulatory Staf f should review the effects of blowdown forces on core internals and the development of appropriate load combinations and deforma-tion limits. The Regulatory Staf f should also revicu analyses of the possible effects upon pressure vessel integrity of thermal shock induced by ECCS operation. ; The applicant has proposed core barrel check valves between the hot Icg and the cold icg to insure proper operation of the ECCS under all circum-stances. Analytical studies indicate that vibrations will not unseat these valves during normal operation. This point should be verified experimentally. - C O f P Y O
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u O P. y . ( ( Honorable Glenn T. Scaborg January 17, 1968 The Advisory Committee on Reactor Safeguards believes that the various items mentioned can be resolved during construction and that the proposed reactor can be constructed at the Three Mile Island site with reasonable assurance that it can' be operated without undue risk to the health and safety of the public. Sincerely yours,
/s/ C. W. Zabel 2
Carroll W. Zabel Chairman
References:
- 1. Metropolitan Edison Company letter, dated May 1,1967; Application
. for Reactor Construction Permit and Operating License, Metropolitan j Edison Company, Three Mile Island Nucicar Station Unit 1; Preliminary ,
i Safety Analysis Report, Vols.1, 2, and 3. i 2. Metropolitan Edison Company letter, dated July 21, 1967; Amendment >
! No. I to application.
- 3. Metropolitan Edison Company letter, dated October 2, 1967; Amendment l No. 2 to application, including Supplement No. 1, Safety Analysis i Report, Vol. 4.
l 4. Metropolitan Edison Company 1cteer, dated November 6,1967; Amendment j No. 3 to application, including Supplement No. 2. ] 5. Metropolitan Edison Company letter, dated December 8, 1967; Amendment No. 4 to application, including Supplement No. 3.
- 6. Metropolitan Edison Company letter, dated December 22, 1967; Amendment No. S to application, including Supplement No. 4.
- 7. Ubtropolitan Edison Company letter, dated January 8,1968; Amendment No. 6 to application.
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Three Mile Island References January 17, 1968 i ACRS Office Copies Only:
- 1. Institute for Atmospheric Sciences comments, dated June 19, 1967.
- 2. DRL Report to ACRS, dated June 28, 1967.
- 3. Letter from Nathan M. Newmark, dated July 13, 1967.
- 4. DRL Ictter to Metropolitan Edison Company, dated August 25, 1967.
- 5. DRL 1ctter to Metropolitan Edison Company, dated September 21, 1967.
- 6. Fish and Wildlife Service letter, dated September 26, 1967.
4
- 7. Geological Survey letter, dated October 16, 1967.
- 8. Coast and Geodetic Survey letter, dated December 13, 1967, with i enclosure.
- 9. DRL Report to ACRS, dated December 22, 1967.
> 10. Newmark and Hall Report to AEC Regulatory Staff, dated December 1967.
- 11. Geological Survey letter, dated January 10, 1968, with attachraent.
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. y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WAS&HNGTON. D.C. 20545 January 17, 1968 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C. 20545
Subject:
REPORT ON THREE MIIE ISIAND NUCLEAR STATION UNIT 1
Dear Dr. Seaborg:
. At its ninety-third meeting, January 11-13, 1968, the Advisory Committee on Reactor Safeguards reviewed the proposal of the Metropolitan Edison Company to construct Three Mile Island Nuclear Station Unit 1. This
. project had been considered previously at Subcommittee meetings held on January 4,1968, in Washington, D. C.., and on October 19, 1967, in Hershey, Pa. During its review, the Committee had the benefit of discussions with l representatives and consultants of the Fktropolitan, Edison Company, the Babcock and Wilcox Company, Gilbert Associates , Inc. , and the AEC Regula-tory Staff. The Committee also had available the documents listed below.
The station is located on Three Mile Island near the east shore of the Susquehanna River in Dauphin County, Pennsylvania, about 10 miles south-east of Harrisburg. Unit 1 is a pressurized-water reactor plant, rated at 2452 MWt, and is similar in design to the units already approved for construction at the Duke Power Company's Oconee Nuclear Station. Flood protection is to be provided at the site by suitabic earth dikes. Two natural-draf t cooling towers are to be used for condenser-water cooling. The emergency core cooling system (ECCS) includes two core flooding tanks, two independent low-pressure systems, and two independent high-pressure systems. Two separate systems are provided for containment cooling. One system consists of three fan-cooling units, and the other consists of two spray systems. The applicant stated that suitable and periodic component and integrated system tests will be performed on these engineered safety features. To further insure low containment leak rates, a fluid block system and a containment penetration pressurization system are to be provided. Operation of the ECCS is initiated automatically by redundant low-pressure signals from transducers actuated by pressure in the two prinary loops. The Committee recommends that in the interest of diversity another method, C O P Y
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Honorable Glenn T. Seaborg January 17, 1968 different in principic from the one proposed, should be added to initiate this function. The diversity thus achieved would enhance the probability that this vital function would be initiated in the unlikely event it is needed. The output circuit of the proposed reactor protection system consists of a single d-c circuit (bus) fed from two station batteries. Both feeders must be interrupted to de-energize the bus and drop all rods. Failure to interrupt either feeder, or any other event that prevents de-energizing the single bus, will inhibit dropping all the rods. The Committee believes this system can and should be revised to correct the deficiency. The revised design should be provided for review prior to installation of , the protection system. . l The applicant has proposed using certain signals from protection instru- j ments for control purposes. The Committee believes that control and protection instrumentation should be separated to the fullest extent practicable, and recommends that the applicant explore further the ! possibility gf making safety instrumentation more nea' rly independent ! of ~ control functions. Consideration should be given to the development and utilization of instru- . mentation for prompt detection of gross failure of a fuci element. I, The applicant described the research and development work planned to confirm the final design of the plant. The Committec continues to emphasize the importance of work to assure that fuel-rod failures in loss-of-coolant . accidents will not affect significantly the ability of the ECCS to prevent clad melting. The applicant is continuing studies on the possible use of part-leagth rods for stabilizing potential x'enon oscillations. Solid poison shims will be added to the fuel elements if necessary to make the moderator temperature coef ficient more negative at the beginning of core life. The Regulatory Staff should review the effects of blowdown forces on core internals and the development of appropriate load combinations and deforru-tion limits. The Regulatory Staff should also review analyses of the possible effects upon pressure vessel integrity of thenaal shock induced by EC.CS operation. s The applicant has proposed core barrel check valves between the hot leg and the cold leg to insure proper operation of the ECCS under all circum-stances. Analytical studies indicate that vibrations will not unscat these valves during normal operation. This point should be verified experimentally.
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Honorabic Clenn T. Scaborg January 17, 1968 The Advisory Committee on Reactor Safeguards believes that the various items mentioned can be resolved during construction and that the proposed reactor can be constructed at the Three Mile Island site with reasonabic assurance that it can be operated without undue risk to the health and safety of the public. I Sincerely yours,
/s/ C. W. Zabel Carroll W. Zabel Chairman
References:
- 1. Metropolitan Edison Company letter, dated May 1,1967; Application for Reactor Construction Permit and Operating License, Metropolitan Edison Company, Threei M,le Island Nuclear Station Unit 1; Preliminary Safety Analysis Report, Vols.1, 2, and 3.
- 2. Hetropolitan Ediscn Company latter, dated July 21, 1967; Amendment l
No. I to application.
- 3. Metropolitan Edison Company 1ctter, dated October 2,1967; Amendment No. 2 to application, including Suppicment No.1, Safety Analysis Report , Vol. 4.
- 4. Nbtropolitan Edison Company letter, dated November 6,1967; Amendment i No. 3 to application, including Supplement No. 2.
- 5. Metropolitan Edison Company letter, dated Ddcember 8,1967; Amendment No. 4 to application, including Supplement No. 3.
- 6. Hetropolitan Edison Company letter, dated December 22, 1967; Amendment No. 5 to application, including Supplement No. 4.
- 7. Metropolitan Edison Company letter, dated January 8,1968; Amendment No. 6 to application.
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( Y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545 January 17, 1968 . l W. Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C. 20545
Subject:
REPORT ON THREE HIII ISLAND NUCLEAR STATION UNIT 1
Dear Dr. Seaborg:
At its ninety-third meeting, January 11-13, 1968, the Advisory Committee { on Reactor Safeguards reviewed the proposal of the Metropolitan Edison Company to construct Three Mile Island Nuclear Station Unit 1. This project had been considered previously at Subcommittee meetings held on January 4, 1968, in Washington, D. C., and on October 19, 1967, in Hershey, I Pc. During its review, the Committee had the benefit of discussions with representatives and consultants of the >ktropolitan, Edison Company, the Babcock and k*ilcox Company, Gilbert Associates , Inc. , and the AEC Regula-tory Staff. The Committee also had available the docunents listed below. The station is located on Three Mile Island near the east shore of the Susquehanna River in Dauphin County, Pennsylvania, about 10 miles south-east of Harrisburg. Unit 1 is a pressurized-water reactor plant, rated at 2452 MWt, and is similar in design to the units already approved for construction at the Duke Power Company's Oconee Nuclear Station. Flood protection is to be provided at the site by suitable earth dikes. Two natural-draf t cooling towers are to be used for condenser-water cooling. The emergency core cooling system (ECCS) includes two core flooding tanks, two independent low-pressure systems, and two independent high-pressure systems. Two separate systems are provided for containment cooling. One system consists of three fan-cooling units, and the other consists of two spray systems. The applicant stated that suitable and periodic component and integrated system tests will be performed on these engineered safety features. To further insure low containnent leak rates, a fluid block system and a containment penetration pressurization system are to be provided. Operation of the ECCS is initiated automatically by redundant low-pressure signals from transducers actuated by pressure in the two primary loops. The Committee recommends that in the interest of diversity another method, t C l O : t
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$ Honorable Glenn T. Seaborg January 17, 1968 different in principic from the one proposed, should be added to initiate this function. The diversity thus achieved would enhance the probability that this vital function would be initiated in the unlikely event it is needed. :
The output circuit of the proposed reactor protectiun system consists of a singic d-c circuit (bus) fed from two station batteries. Both fceders must be interrupted to de-energize the bus and drop all rods. Failure to th interrupt either feeder, or any other event that prevents de-energizing the single bus, will inhibit dropping all the rods. The Committee believes this system can and should be revised to correct the deficiency. The revised design should be provided for review prior to installation of the protcetion system. . The applicant has proposed using certain signals from protection instru-ments for control purposes. The Committee believes that control and protection instrumentation should be separated to the fullest extent practicable, and recommends that the applicant explore further the i possibility gf making safety . instrumentation mors nearly independent - of control functions. Consideration should be given to the development and utilization of instru- . ( mentation for prompt detection of gross failure of a fuel element. The applicant described the research and development work planned to confirm the final design of the plant. The Committec continues to emphasize the importance of work to assure that fuel-rod failures in loss-of-coolant accidents will not affect significantly the ability of the ECCS to prevent clad melting. The applicant is continuing studies on the possible use of part-length rods for stabilizing potential x'enon oscillations. Solid poison shims will be added to the fuel elements if necessary to make the moderator temperature coefficient more negative at the beginning of core life. The Regulatory Staff should review the effects of bloudown forces on core internals and the development of appropriate load combinations and deforma-tion limits. The Regulatory Staff should also review analyses of the possible effects upon pressure vessel integrity of thermal shock induced by ECCS operation. The applicant has proposed core barrel check valves between the hot Icg and the cold leg to insure proper cperation of the ECCS under all circum-stances. Analytical studies indicate that vibratfors will not unscat
, these valves during normal operation. This point should be verified experimentally, l C
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4 P Y . Honorabic Clenn T. Scaborg January 17, 1960 The Advisory Committee on Reactor Safeguards believes that the various ~ items mentioned can be resolved during construction and that the proposed I reactor can be constructed at the Three Mile Island site with reasonabic
- assurance that it can be operated without undue risk to the health and safety of the public.
Sincerely yours,
/s/ C. W. Zabel Carroll W. Zabel Chairman
References:
- 1. Metropolitan Edison Company letter, dated May 1,1967; Application for Reactor Construction Permit and Operating License, Metropolitan Edison Company, Three Mile Island Nuclear Station Unit 1; Preliminary
'( Safety Analysis Report, Vols. 1, 2, and 3.
- 2. Metropolitan Edison Company letter, dated July 21, 1967; Amendment No. I to application.
- 3. Metropolitan Edison Company letter, dated October 2,1967; Amendment L No. 2 to application, including Supplement No.1, Safety Analysis Report , Vol. 4. !
- 4. Metropolitan Edison Company letter, dated November 6,1967; Amendment No. 3 to application, including Supplement No. 2,
- 5. Metropolitan Edison Company letter, dated December 8,1967; Amendment No. 4 to application, including Supplement No. 3.
- 6. Metropolitan Edison Company letter, dated December 22, 1967; Amendment No. 5 to application, including Supplement No. 4.
- 7. Metropolitan Edison Company letter, dated January 8,1968; Amendment No. 6 to application.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545 I Hay 15, 1968 Honorable Glenn T. Seaborg Chairman i U. S. Atomic Energy Commission ! Washington, D. C. 20545
Subject:
REPORT ON CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT
Dear Dr. Seaborg:
i At the special ACRS meeting on April 27, 1968, the Advisory Committee on Reactor Safeguards reviewed the proposal of the Florida Power Cor-poration to construct the Crystal Rive- Nuclear Generating Plant. This project had been previously considered at a Subcommittee meeting held on February 15, 1968, at the site. During its review, the Com-mittee had the benefit of discussions with representatives and consul-t tants of the Florida Power Corporation, the Babcock and Wilcox Company, Gilbert Associates, Inc., and the AEC Regulatory Staff. The Connittee also had available the documents listed below. The plant will be located on the Gulf of Mexico about 70 miles north of Tampa, Florida, and 7 1/2 miles northwest of the town of Crystal River. 3 The population, including Crystal River, within a ten mile radius of the plant is 3300. The site co:oprises 4738 acres, .on which the Florida Power Corporation operates a 387 MWe coal-fired Unit 1 and is building a coal-fired 510 MWe Unit 2. Unit 3 will use a pressurized water reac-tor, rated at 2452 MWt and 855 MWe. The program for foundation grouting and the prctection to be provided against flooding appear to be satisfactory as do other site-related factors. ., , The Conmittee believes that the proposed off-site power syt, tem should be modified to fulfill Criterion 39 so that no single failure will pre-vent the operation of minimum electricelly-powered safety features necessary to protect the core.
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< L Honorable Glenn T. Seaborg May 15, 1968 The proposed Unit 3 is similar to the Duke Power Company's Oconee Units (ACRS Report, July 11, 1967) and the Metropolitan Edison Company's Three Mile Island Unit (ACRS Report, January 17, 1968). The Committee con-tinues to call attention to matters that warrant careful consideration for all large, water-cooled, power reactors. These catters, stated in the Three Mile Island and Oconce reports, apply similarly to the Crystal River Unit 3.
The Advisory Committee on Reactor Safeguards believes that, if due con-sideration is given to the foregoing itccs, the proposed reactor can be constructed at the Crystal River site with reasonable assurance that it can be operated without undue risk to the health and safety of the public. Mr. Harold Etherington did not participate in the Committee's review of this project. Sincerely yours,
/s/
Carroll W. Zabel Chairman References cttached. e M e e
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t-( 7 , . Honortble Glenn T. Seaborg May 15, 1968 References - Crystal River
- 1. Letter from Florida Power Corporation, dated August 10, 1967; Application for License; Volumes 1, 2, 3 and 4 of Preliminary Safety Analysis Report, Crystal River Units 3 and 4 Nuclear Generating Plant
- 2. Letter from Florida Power Corporation, dated January 15, 1968; A=cndment No. 1 to License Application.
- 3. Letter from Florida Power Corporation, dated February 7, 1968; Amendment No. 2 to License Application; Supplement No. 1
.) , 4. Letter from Florida Power Corporation, dated March 1, 1968; I
- Amendment No. 3 to License Application; Supplement No. 2
- 5. Lette t from Florida Power Corporation, dated March 11, 1968;
' Amendment No. 4 to License Application; Supplement No. 3
- 6. Letter from Florida Power Corporation, dated April 1, 1968; g Amendment No. 5 to License Application 1
ACRS OFFICE COPIES ONLY
- 1. DRL Report to ACRS, dated September 28, 1967
- 2. ESSA Report, dated October 18, 1967
- 3. Fish and Wildlife Service Letter, dated February 12, 1968
- 4. Coastal Engineering Research Center Letter, dated February 26, 1968
- 5. Coast and Geodetic Survey Letter, dated March 15, 1968
- 6. DRL Report to ACRS, dated March IE, 1968
- 7. Newmark, Hall and Hendron Report, dated March, 1968
- 8. U. S. Geological Survey Letter, dated April 2, 1968
- 9. DRL Report to ACRS, dated Ap'ril 16, 1968
- 10. Newmark, Hall and Hendron Report, dated April, 1963 e
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON D.C. 20545 JUL 191968
- i 1:enorable Clenn T. Seaborg Chairman U. S. Atomic Energy Commission Wachington, D. C. 20545
Subject:
REPORT ON RANCHO SECO NUC1 EAR GENERATING STATION, UNIT NO. 1
Dear Dr. Seaborg:
During its ninety-ninth meeting, July 11-13, 1968, the Advisory Committee on Reactor Safeguards reviewed the proposal of the Sacrarnento Municipal Utility District to construct the Rancho Seco Nuclear Generating Station, Unit No. 1. This project had been considered previously during Subcom-inittee meetings on April 23, 1968, at the site, and on June 28, 1968, in Washington, D. C. In the course of its review, the Cormtittee had the benefit of discussions with representatives and consultants of the Secra-g mento Municipal Utility District, the Babcock and Wilcox Company,11echtel Corporation, and the AEC Regulatory Staff. The Comnittee also had avail-abic tha documents listed below. This 2452 me pressurized water reactor will be located about 25 miles southeast of Sacramento, California, in a sparsely populated area. This region of California is seismically relatively inactive; the largest earthquake of historic record in the vicinity of the site is of Intensity VI,1:odified Mercalli Q21) scale. The applicant has agreed to design for safe shutd:wn following an earthquake during which the maximum horizontal acceleration is 0.25 g Det VIII), cad the design will allow continued operation for an earthquake of about one-half of this accaleration. He plans to instan a strong motion accol'erograph. All vater needs for this plant will be supplied from the Folsom South Canal,.which will pcss witbin five miles of the site. Should completion of this const be delayed, a separate pipeline from Lake Natoma, about 20 miles north of the site will be constructed. An on-site reservoir will have a cepacity of 2500 acre-feet, sufficient for about 35 full power ! days of operation, and wasta heat will be discharged to the atmosphera , through use of cooling towers. The plant is unique in that the appli- , i cant proposes not to discharge liquid wastes to the environment. The applicant is studying methods to cope with possible build-up of tritium
. in the reactor coolant water.
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I ' Honorable Glenn T. Seaborg
- JUL 191968 The applicant has proposed using signals from the protection system for control and protection purposes. The Committee reiterates its belief that control and protection instrumentation should be as nearly Lndepen-dont of common failure modes as possible, so that the protection will not be impaired by the same failure that initiates a transient requiring protection. The applicant and the AEC Regulatory Staff should review the proposed design for common failure modes, taking into account the possi-bility of systematic, non-random, concurrent failures of redundant de-vices, not considered in the single-failure criterion. In cases where hypothesized control or override failure could lead to the need for ac-tion by interconnacted protection bastrumentation, separate protect ion instrumentation channals should be provided or some other design approach
- used to provide equivalent safety.
The Committee suggests that, in view of possible uncertainties in current predictive techniques, further analyses be made of the anticipated inte-grated fast flux at the pressure vessel wall, and that the adequacy of the proposed pressura vessel material surveillance progcam be resolved between the applicant and the Regulatory Staff during construction of the station. This reactor is similar to others designed by this vendor and reviewed previously (see, for example, the ACRS report on the Crystal River plant, I May 15, 1963). The Committee continues to call attention to matters that warrant careful consideration by the manufacturers of all large, water-cooled, power reactors. These matters, referred to in the above-mentioned report, apply similarly to the Rancho Seco project. The Advisory Committee on Reactor Safeguards believes that the items noted above can be resolved during construction, and thac the proposed plant can be built at the Rancho Seco site with reasonable assurance that it can be operated without undue risk to the health and safety of the public. l Sincerely yours.
. l Originsi signed by l Carroll W. Zabel '
Carroll W. Zabel Chairman References attached. t 9 k..' ! e 1
/ -
( Honorable Glenn T. Seaborg JUL 191968 References - Rancho Seco
- 1. License Application for Construction Permit. Sacramanto Municipal Utility District, deced Nove:uber,1967; Volumes I, II, III, IV of the Prelictinary Safety Analysis Report for Rancho Seco Nuclent Generating Station, Unit No. 1
- 2. Sacramento Municipal Utility District; Amendment No. 1, dated February 2, 1960
- 3. Sacrom-auto Municipal Utility District; Amendment No. 2, dated April 15, 1968
- 4. Sacramento !!unicipal Utility District; Amendment No. 3, dated May 30, 1968
- 5. Sacramento Municipal Utility District; Amendment No. 4, dated June 33, 1968 l
ACRS OFFICE COPIES ONLY
- 1. DRL Report to ACRS, dated December 14, 1967
- 2. B. M. Page Letter, dated May 27, 1968
- 3. Fish and Wildlife Service Letter, dated May 27, 1968
- 4. ESSA Report, dated May 27, 1968
- 5. J. T. Wilson Letter, dated May 28, 1968
- 6. M. P. White Letter, dated June 10, 1968 *
- 7. DRL Report to ACRS, dated June 24, 1968
- 8. K. V. Steinbrugge Letter, dated June 26, 1968
- 9. Neumark and Hall Draft Report, d,ated July,1968 (received July 8,1968)
- 10. Coast and Geodetic Survey Report, dated July ,5,1968 (received July 10, 1968.'
- 11. DRL Supplementary Report to ACRS, dated July 10,1968 (received July 10, 1961
- 12. Hydrology and Geology Report, undated, unsigned (received July 11,,1968)
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( ( ADVISORY COMMIT rEE ON REACTOR S. . EGUARDS
- UNITED STATES ATOMIC ENERGY COMMISSION
*- WASH INGTON. D.C. 20545 January 27,.1970 l
Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Causaission Washington, D. C. 20545 i
Subject:
REPORT ON PALISABES PLANT ; Dear Dr. Seabeegt At a Special Meeting, January 23-24, 1970, the Advisory Committee on Reactor Safeguards completed its review of the application by Consumers Fewer Company for authorisatioa to operate the Palisades Plant at power levels up to 2200 Wt. .This project was also considered at the 113th ACES meeting, September 4-6, 1969, the 115th ACES meeting, November 6-8, 1969, and the 116th ACRS meeting, December 11-13, 1969. Subcommittee meetings were held on July 31, 1969, at the site, and on October 29, 1969, December 3, 1969, and January 22, 1970, in Washington, D. C. During its review, the Committee had the benefit of discussions l with representatives of Censumers Power Campany, Combustion Engineering, Inc., Bechtel Corporation, the ABC Regulatory Staff, and their consultants. The Committee also had the benefit of the documents listed. The Committee
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reported to you en the construction of this plant in its letter dated January 18, 1967. fhe site for the Palisades Plant consists of 487 acre's on the eastern shore of Lake Michigan in Covert Township, approximately four and one-half miles south of South Raven, Michigan. The minisman exclusion radius for the alte is 2300 feet and the nearest population conter of more than 25,000 residente consists of the cities of Benton Harbor and St. Joseph, Michf.gan, which are appreuimatsly 16 miles south of the site. The nuclear steam sapply system for the Palisades Plant is the first of the Combustion Eagineering line currently licensei .for construction. A feature of the Felisadas reactor is the emission of the thermal shield. Studies were made by the applicant to show that omission s'f the shield would not adversely affect the flow characteristics within the rsector vessel or alter the thermal stresses in the wolls of the vessel in a manner detrimental to safe operation of the plant. Surveillance specimens in the vessel will he used to monitor the radiation das.ege during the 1tfe of the plant. If these specimens reveal changes that affect the safety of the plant, the reactor vessel will be annealed to reduce ee8 A ,M .'
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I ( ( Boaorable Clenn T. Seaborg January 27, 1970 L radiation damage effects. The results of annealing will be confirmed ; by tests on additional surveillance specimens provided for this purpose. Prior to accumulation of a peak fluence of 1019 nyt (/1 Mov) on the reactor vessel well, the Regulatory Staff should reevaluate the continued i suitability of the currently proposed startup, cooldown, and operating conditions. The secondary containment is a reinforced concrete structure consisting of a cylindrical portion prostressed in both the vertical and circumferential directions, a dome roof prostressed in three directions, and a flat non-prostressed base. Before operation, it will be pressurised and extensive measurements will be made of gross deformations and of strains in the liner, reinforcement, and concrete, and the pattern and size of cracks in the concrete will be observed and measured. The applicant has proposed suitable acceptance criteria for the pressure test, and the ACRS recomunends , that the Regulatory Staff review and assess the results of this test
. prior to operation at significant power.
The prestressing tendons in the containment consist of ninety, one-quarter-inch diameter wires. They are not grouted or bonded, and are protected from corrosion by grease pumped into the tendon sheaths. The applicant has proposed that selected tendons be inspected periodically for broken l wires, loss of prostress, and corrosion. If degradation is detected, the inspection can be extended to the remaining tendons, all of which are accessible. The applicant is performing studies to determine the appropriate number and interval for tendon inspection. This matter should be resolved in a manner satisfactory to the Regulatory Staff.- The core is calculated to have a slightly negative mod'erator coefficient at full power operation at beginning-of-life, but uncertainties in the calculations are such that the existence of a positive moderator coeffi-cient cannot be precluded. The applicant has stated that the moderator coefficient will not exceed +0.5 x lo-4A k/k/or at d. ginning-of-line, computed from start-up test data on a conservative basis. The applicant also plans to perform tests to verify that divergent asimuthat menon oscillations cannot occur in this reactor. The Canonittee recommends that the Regulatory Staff follow r.he measurements and analyses required to establish the value of the moderator coefficient. The meteorological observation program conducted at the sita subsequent to the Committee's report to you on January 16, 1967, indicated the l need for the addition of iodins: removal equipment to the containment for use in the unlikely event of a loss-of-coolant accident. The applicant proposed to inst.sil means for adding sodium hydroxide to the water in the containment spray system. However, because of uncertainties regarding
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the generation of hydrogen and the effects of other materials resulting ( ,
( ( Menorable filenn T. Seaborg January 27, 1970 i r from the reaction of this alkaline solution with the relatively large amounts of altsainum in the containment, this spray additive will not be used unless it can be shown by further studies that the use of sodissa hydroxide is clearly acceptable. In addition, the applicant will carry out studies of iodine rarnovel by borated water sprays without sodissa hydroxide. If the results of these studies are not acceptable, a different iodine removal system satisfactory to the Regulatory Staff will be installed at the first refueling outage. A report on the applicant's plans will be submitted to the AEC within six months following issuance of a provisional operation license. The Committee believes that this procedure is satisfactory for operation l at power levels not exceeding 2200 MWe. The applicant has stated that if fewer than four primary coolant pumps are operating, the reactor overpower trip settings will be reduced such that the safety of the reactor is assured in the absence of automatic changes in the thermal margin trip settings. The Comnittee believes that, for transients having a high probability of securrence, and for which action of a protective system or other engineered safety feature is vital to the public health and safety, j an exceedingly high probability of successful eetion is needed. Common failure modes must be considered in ascertaining an acceptable level of protection. Studies are to be made on further means of preventing
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comunon failure modes from negating scram action, and of design features to make tolerable the consequences of failure to scram during anticipated transients. The applicant should consider the results of such studies and incorporate appropriate provisions in the Palisades Plant. The Commaittee recommends that attention be given to the long-term l l ability of vital components, such as electrical equipment and cables, to withstand the environment of the containment in the unlikely event of a loes-of-coolant accident. This matter is applicable to all large, water-cooled power reactors. Continuing research and engineering studies are expected to lead to enhancement of the safety of water-cooled reactors in other areas than those mentioned: for example, by determination of the extent of the generation of hydrogen by radiolysis and from other sources, and development of means to contrcl the concentration of hydrogen in the containment, in the unlikely event of a loss-of-coolaint accident; by development of instrumentation for inservice monitoring of the pressure vessel and other parts of the primary system for vibration and detection i of locae parts in the system; and by evaluation of the consequences of water contamination by structural materials and coatings in a loss-of-coolant accident. As solutions to these problema develop and are evaluated l l 1
( Esmerable Glenn T. Seaborg Ja wary 2,', p by the Regulatory Staff, appropriate estion should be taken by the applicant 1 se a resseeable time seale. I t The Advisory Cenedttee on Roseter Safeguards believes that, if due regard is S iven to the items mentioned above, and subject to satisfactory templetion of construction and pre-operatiemal testing, there is reasonable assurance that the Palisados Plant can be operated at poner levels up to 2200 left without undue risk to the health and safety of the public. Steeerely yours, Original Signed by Joseph E. Hendrio Joseph M. Hendrie chairman
References:
- 1. Flaal Safety Analysis Report for the Palisades Plant
- 2. Amenamenta No. 9-19 to 11eense appliestion l
i l l
, Revised Page
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ADVI3ORY COMMITTEE ON REACTOR S... EGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASH INGTON. D.C. 10545 January 27, 1970 l l L Bonorable Olena T. Seaborg Chairman ej. U. S. Atomic Ener5y Commission Washington, D. C. 20545 Subjoet: REPORT ON PALISADg8 F1 ANT
Dear Dr. Seaborg:
At a Special Meeting, January 23-24, 1970, the Advisory Ceummittee on Reactor Safeguards completed its review of the application by Consumers Power Company for authorisation to operate the Palisades Plant at power levels up to 2200 Mit. .This project was also considered at the i 113th ACES asating, September 4-6, 1969, the 115th ACES meeting, November 6-8, 1969, and the 116th ACES meeting, December 11-13, 1969. Subcomunittee meetings were held on July 31, 1969, at the site, and on October 29, 1969, December 3, 1969, and January 22, 1970, in Washington, l D. C. During its review, the Coimaittee had the benefit of discussions l with representatives of Consumers Fever Campany, Combustion Engineering. l Inc., Bechtel Corporation, the AEC Regulatory Staff, and their consultants. I
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The Commaittee also had the benefit of the docssments listed. The Coenittee l reported to you en the construction of this plant in its letter dated l January 18, 1967. , i The site for the Palisades Plant consists of 487 acres on the eastern shore of Lake Michigan in Covert Township, approximately four and one-half miles south of South Haven, Michigan. The minimisa exclusion radius J for the site is 2300 feet and the nearest population center of more than 25,000 residents consists of the cities of Benton Barbor and St. Joseph, Michigan, which are approximately 16 miles south of the site. The nuclear steam supply system for the Palisades Plant is the first of the Combustion Engineering lina currently licensed fer construction. A-.fsature of the Falisadas reactor is the omiscion of the thermal shield. Studies worn made by the applicant to show that omission si the shield
. would not adversely affect the flow charsetarieties within t!w reactor vessel or alter the thermal stresses in the walls of the vessel in a
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manner. detrimental to safe operation of the plant. Surveillance specimens in the vessel will be used to monitor the radiation damage during the life of the plant. If these specimesa revest changes that affect the safety of the plant, the reactor vessel will be annealed to reduce
t Beaorable Glenn T. Seaborg January 27, 1970 radiation damage effects. The results of annealing will be confirmed by tests en additional surveillance specimens provided for this purpose. Prior to accumulation of a peak fluence of 1019 avt ( r 1 Mov) on the - reacter vessel wall, the Regulatory Staff should reevaluate the continued evitability of the currently proposed startup, cooldown, and operating I conditions, w ^ The secondary containment is a reinforced concrete structure consisting of a cylindrical portion prostressed in both the vertical and circumferential directions, a done roof prestressed in three directions, and a flat non-prestressed base. Before operation, it will be pressurised and extensive measurements will be made of gross deformations and of strains in the ; liner, reinforcement, and concrete, and the pattern and siae of cracks i in the concrete will be observed and measured. The applicant has proposed l suitable acceptance criteria for the pressure test, and the ACRS recommends < that the Regulatory Staff review and assess the results of this test i prior to operation at significant power. ! I The prestressing tendons in the containment consist of ninety, one-quarter-inch diameter wires. They are not grouted or bonded, and are protected , 4 from corrosion by grosse pumped into the tendon sheaths. The applicant ' l 5 has proposed that selected tendons he inspected periodically for broken i wires, loss of prostress, and corrosion. If degradation is detected, the inspection can be extended to the remaining tendons, all of which
! are accessible. The applicant is performing studies to determine the appropriate number and interval for tendon inspection. This matter should be resolved in a manner satisfactory to the Regulatory Staff.
The core is calculated to have a slightly negative moderator coefficient at full power operation at beginning-of-life, but uncertainties in the calculations are such that the existence of a positive moderator coeffi- , cient cannot be precluded. The applicant has stated that the moderator coefficient will not exceed +0.5 x 10-4/ 1 k/k/*F at beginning-of-life, computed from start-up test data on a conservative basis. The applicant also plans to perform tests to verify that divergent asinuthal menon oscillations cannot occur in this reactor. The Committee recommends that the Regulatory Staff follow the meseurements and analyses required to i establish the value of the moderstor coefficient. The meteorological observation program conducted at the site subsequent to the Committee's t. pert to you on January 18, 1967, indicated the need for the addition of iodina removal equipment to the containment for use in the unlikely event of a Icss-of-coolant accident. The applicant proposed to install means feq adding sodium hydroxide to the water in the containment spray system. However, because of uncertainties regarding ' the generation of hydrogen and the effects of other materials resulting
I l i Nonerable Glenn T. Seaborg January 27, 1970 from the reaction of this alkaline solution with the relatively large amounts of aluminum in the containment, this spray additive will not . be used unless it can be shown by further studies that the use of - sodium hydroxide is clearly acceptable. In addition, the applicant : will carry out studies of iodine removal by borated water sprays without soditan hydroxide. If the results of these studies are not acceptable, a difforent iodine removal system satisfactory to the Regulatory Staff will be installed at the first refueling outage. A report on the applicant's plans will be submitted to the ABC within six months following issuance of a provisional operation license. The Comittee believes that this procedure is satisfactory for operation at power levels not exceeding 2200 MWt. The applicant has stated that if fewer than four primary coolant pumps are operating, the reacter overpower trip settings will be reduced such that the safety of the reactor is assured in the absence of astomatic changes in the thermal margin trip settings. 4 The Committee believes that, for trans ents having a high probability of occurrence, and for which action o1 a protective system or other ( engineered safety feature is vital to the public health and safety, an exceedingly high probability of successful action is needed. Common failure modes must be considered in ascertaining an acceptable level of protection. Studies are to be made on further means of preventing 4
-' common failure modes from negating scram action, and of design features to make tolerable the consequences of failure to scram during anticipated transients. The applicant should consider the results of such studies and incorporate appropriate provisions in the Palisades Plant.
"he Consmittee recomunends that attention be given to the long-term ability of vital components, such as electrical equipment and cables.
- to withstand the environment of the containment in the unlikely event of a loes-of-coolant accident. This matter is applicable to all large. - water-cooled power reactors. Continuing research and engineering studies are expected to lead to enhancement of the safety of water-cooled reactors in other areas than those mentiened: for exa:sple, by determinatiot of the extent of the generation of hydrogen by radiolysis and from other sources, and development of means to control the concentration of hydrogen in the containment, in the unlikely event of a loss-of-coolant accident; by development of instrumentation for inservice monitoring of the pressure vessel and other parts of the primary system for vibration and detection of loose parts in the system; and by evaluation of the consequences ei vetor contamination by structural materials and coatings in a loss-of-coolant accident. As solutions to these problems develop and are evaluated i
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Janury 2,, p Romerable 81enn T. Seaborg by the Regulatory Staff, appropriate estion should be taken by the applicant se a reasonable time seale. i The Advisory Committee en Raaetor Safeguards believes that, if due regard is given to the items mentiemed above, ed subject to satisfactory esopletion of construction and pre-operatieaal testing, there is reasonable
.e assurance that the Palisades Plant can he operated at power levels up to 2200 we without undue risk to the health and safety of the public.
Sincerely yours, Original Sfsned by Joseph _M. Hendrio Joseph M. Eendrie Chairman Esferences:
- 1. Final Safety Analysis Report for the Palisades Plant j 2. Amendments Bo. 9-19 to license application l
l
. Revised Page
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.' > hi ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545 March 12, 1970
. 1 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C. 20545
Subject:
REPORT ON llUTCHINSON ISIAND PIANT UNIT NO. I
Dear Dr. Scaborg:
At its 119th. meeting, March 5-7, 1970, the Advisory Committee on Reactor Safeguards completed its review of the application of the Florida Power and Light , Company for authorization to construct a ' nuclear power plant at its Hutchinson Island site in St. I,ucie County, Florida. A Subcommittec visited the site on January 5, 1970; a second Subcommittee meeting was j held in Chicago on February 21, 1970. During its review, the Committec , had the benefit of discussions with the applicant, Combustion Engineering,
, l Inc. , Ebasco Services, Inc. , the AEC Regulatory Staf f, and their consult-ants. The Connnittee also had the benefit of the documents listed.
The Hutchinson Island Plant Unit No. I will be located on a tract of land of approximately 1100 acres, about half way betwccu Fort Pierce and Stuart on the cast coast of Florida. About 1000 people live within a five mile
- radius of the site. 'The nearest population cent er i s' Fort Pierec (popula-tion about 34,000), which is eight miles away.
The plant site on Hutchinsor. Island is underlain by sand to a depth of several hundrdd feet. To provide satisfactory bearing and' settlement characteristics and resistance to liquefaction, the first sixty feet of loose send is being removed and the excavation refilled to foundation depth with granular traterial compacted to a relative density of 85 per-cent. . Tlicproposedpressurizedwaterreactorhasadesignpowericvelof 2440 .W(t) and is similar to the previously reviewed Ma-inc Yankee and Calvert Cliffs reactors (ACRS reports dated July 19, 1968 and March 13, 1969). The containment system consists of a steel containment vessel enclosed within a reinforced concrete building, with the annular space maintained at a slight negative pressure and exhausted through filters. The applicant has stated that the containment and other structures and
- systems important to safety will be designed to meet the same tornado O
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( i r 2- Mar 12, 1970 (, Honorable Glenn T. Seaborg - I t I design criteria as have been used for other recently reviewed plants, and that protection of vital components will be provided against the probable ' maximum hurricane-induced flood and runup level as estimated by the Coastal Engineering Research Center. i The applicant stated that a dynamic scismic analysis will be perforiaed on ! the primary sy' stem. Several other matters related to seismic design, including the spectra to be used in the design of piping and equipment, and the design procedures to be used for various types of Class 1 piping, should be resolved in a manner satisfactory t0 the Regulatory Staf f. The applicant stated that the primary system will be designed so that annealing of the pressure vessel will be practical at a temperature of at least 650 F. Pump seal and other leakage from emergency core cooling (ECCS) equipmentof and lines outside the containment may ' lead to undesirable releases The radioactivity in the unlikely event of a loss-of-coolant accident. Committee recommends that the atmosphere around the ECCS lines and pumps outside the containmen't be vented through a charcoal filter system. j Further study is required with regard to potential releases of radio-activity in the unlikely event of gross damage to an irradiated subasr.embly during funi handling and the possibic need for a charcoal filtration sys-tem in the fuel handling building. This matter should be resolved in a manner satisfactory to the Regulatory Staff. All hot process lines penetrating the containment annulus will be designed in with a guard pipe to direct steam flow back to the primary containment the unlikely event of a rupture of the process pipe in the annulus region. In view of the importance of the guard pipes, the applicant will arrange for an independent review of the design. The applicant stated that he will install a concrete wall in the contain-ment penetration room to separate the cables and penetrations for redon-dant devices essential to safety. The Connittee believes that the separation of redundant elements in the penetration room and elsewhere requires further study, as to both criteria and design details. A suitabic preoperational vibration testing program should be employed for
- the primary system. Also, attention should be given to the development
- and utilization of instrumentation for in-service monitoring for excessive vibration or loose part s in the primary system.
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Mar 12, 1970 Honorable Glenn T. Seaborg Wh'en details of the planned loatis and ratings of the emergency diesel generators become availabic, the Regulatory Staff should assure itself sufficient that adequacy of design conservatism is realized and that testing and experience will be available prior to plant startup to prove the reliability of the emergency power system. The Cor:mittee reiterates its interest in active participation by appli-cants in overall quality assurance programs to better assure the con-struction of safe plants. In this regard, a greater icvel of dircet in the quality assurance program of the participation by the applicant Hutchinson Island Plant would be desirable. he
'Information on a number of items, identified in previous reports of t Committee, is to be provided by the applicant to the Regulatory Staf f during construction. These include: .
a) A study of means of preventing common failure modes from negating scram action and of design features to make tolcr-able the consequences of failure to scram during anticipated transients.
'l b) Review of development of systems to contro,1 the buildup ef including an appropriately hydrogen in the containment, conservativ'e estimate of possibic hydrogen sources, and of instrumentation to monitor the course of events in the un-likely event of a loss-of-coolant accident.
Other probicms related to large water reactors have been identified by the Regulatory Staf f and the ACRS and cited in previous ACRS reports. The Connittee fccis that resolution of these items should apply equally to the llutchinson Island Plant. The Conmittee b'clieves that the above items can be resolved during con-struction and that, if due consideration is given ta these items, the nuclear plant proposed for the lhttchinson Island site can be constructed with reasonabic assurance that it can be operated without undue risk to the health and safety of the public. Sincerely yours,
/s/
Joseph M. llendric Chairman References attached.
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f.{}f'. - l, Honorable Glenn T. Seaborg Mar 12, 1970 4 References - Hutchinson Island Plant Unit No. I
- 1. Hutchinson Island Plant Unit No.1, Preliminary Safety Analysis Report , I.
Volumes l'- 3.
- 2. Florida Power & Light Company letter, dated April 1,1969.
! 3. Amendments 1 - 8 to License Application. l l i 1 1 . I i . r I Revised Page
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9 dWi1] 1 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D C. 20545 March 12, 1970
. l
?z Honorabic Clenn T. Seaborg Chairman
, U. S. Atomic Energy Commission Washington, D. C. 20545
Subject:
REPORT ON llUTCHINSON IS1AND PIANT UNI'l NO.1
Dear Dr. Scaborg:
At its 119th meeting, March 5-7, 1970, the Advisory Committee on Reactor Safeguards completed its review of the application of the Florida Power and Light. Company for authorization to construct a' nuclear power plant at its Hutchinson Island site in St. Lucie County, Florida. A Subcommittee visited the site on January 5, 1970; a second subcommittee meeting vac held in Chicago on February 21, 1970. During its review, the Committee had the benefit of discussions with the applicant, Combustion Engineering, Inc. , Ibasco Servicer, Inc., the AEC Regulatory Staf f, and their consult-ants. The Connittee also had the benefit of the doemnents listed. The Hutchinson Island Plant Unit No. I will be located on a tract of land of approximately 1100 acres, about half way betuccu Fort Pierce and Stuart on the cast coast of Florida. About 1000 people live within a five mile radius of the site. ' The nearest populat. ion cent er is Fort Pierce (popula-tion abuut 34,000), which is cight miles away. The plant rite or. Hutchinson Island is underlain by sand to a depth of several hundre~d feet. To provide satisfactory bearing and settlement characteristics and resistance to liquefaction, the first sixty feet of loose send is being removed and the excavation refilled to foundation depth with granular material compacted to a relative density of 85 per-cent. TkeproposedpressurizedwaterreactorhasadesignpowerIcvelof 2440 MJ(t) and is similst to the previously reviewed Na.inc Yankec and Calvert Cliffs reactors (ACRS reports dated July 19, 1968 and March 13, 1969). 1he containment system consists of a steel containment vessel enclosed within a reinforced concrete building, with the annular space maintained at a slight negative pressure and exhausted through filters. The applicant has stated that the containment and ot her structures and ; systems important to safety will be designed to meet the same tornado l l
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.. . j'l ki. U N 2- Mar 12, 1970 ',g Honorable Glenn T. Seaborg -
I> i design criteria as have been used for other recently reviewed plants, and . that protection of vital components will be provided against the probahic 'I
- maximum hurricane-induced flood and runup level as estincted by the Coastal Enginecting Research Center.
y, The applicant stated that a dynamic scismic analysis will be performed on the primary system. Several other matters related to seismic design, including the spectra to be used in the design of piping and equipment, and the design procedures to be used for various types of Class 1 piping, should be resnived in a manner satisfactory to the Regulatory Staff. The applicant stated that the primary system will be designed so that annealing of the pressure vessel will be practical at a temperature of at least 650 F. Pump seol and other leakage from emergency core cooling (ECC .) equipment of and lines outside the containment may lead to undesirable releases The radioactivity in the unlikely event of a loss-of-coolant accident. i Committ ee recommends that the atmosphere around the ECCS lines and pumps outside the containment be vented through a charcoal filter system. Further study is rcquired with regard to potential releases of radio-activity in the unlikely event of gross damage to an irradiated suharsembly during fuel handlin;; and the possibic need for a charcoal filtration sys-tem in the fuel handling building. 1his matter should be resolved in a conner stai:f actory to the Regulatory Staf f. All hot process lines penetrating the containment annulus will be designed with a guard pipe to direct steam flow back to the primary containment in the unlikely event of a rupture of the process pipe in the annulus region. In vice of the importance of the guard pipes, the applicant will arrange for an independent review of the design. The appli: ant stated that he will install a concrete wall in the contain-ment renetration room to scporate the cables and penetrations for redon-dan,t devices essential to safety. The Conanitt ee believes that the se, aration p of redundant elements in the penetration roem and ciscwhere requires further study, as to both criteria and design details. A suitabic preoperational vibration testing program should be employed for j the primary system. Also, attention should he given to the development and utilization of instrumentation for in-service monitoring for excessive vibration or loose part s in the primary system. ( l i D I
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Mar 12, 1970 Honorable Glenn T. Seaborg When details of the planned loarls and ratings of the emergency diesel generators become availabic, the Regulatory Staff shouldsufficient assure itself
- that adequacy of design conservatism is realized and that i testing and experience will be available prior to plant startup to prove the reliability of the emergency power system.
T in active participation by appli-The Committee reiterates its interest cants in overall quality assurance programs to better assure the con-struction of safe plants. In this regard, a greater Icvc1 of direct in the quality assurance program of the part icipation by the applicant Hutchinson Island Plant would be desirable. Information on a number of items, identified in previous report 8 of the Committee, is to be provided by the applicant to the Regulatory Staf f during construction. These include: a) A study of means of preventing common failure modes from negating scram action and of design features to make tolcr-able the consequences of failure to scram during antteipated , I transients. b) Review of development of systems to contro,1 the buildup ef hydrogen in the containment, including an appropriately conservativ'e estimate of possibic hydrogen sources, and of instrumentation to monitor the course of events in the un-likely event of a loss-of-coolant accident. Other problems related to large water reactors have been identified by the Regulatory Staf f and the ACRS and cited in previous ACRS reports. The Connittee feels that resolution of these items should apply equally to the llutchinson island Plant. The Consnit tee b'clieves that the above items can be resolved during con-struction and that, if due consideration is given to these items, the nuclear plant proposed for the llutchinson Island site can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public. Sincerely yourr., Isl I Joseph M. llendric Chairman References attached.
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' Honorable Glenn T. Seaborg Mar 12, 1970 .
References - Hut chinson Island Plant Unit No. 1 _
- 1. Hutchinson Island Plant Unit No.1*, Preliminary Safety Analysis Report, S Volumes 1 - 3.
Y 2. Florida Power & Light Company letter, dated April 1,1969.
- 3. Amendments 1 - 8 to License' Application.
i I I I f l Revised Page-t e ' ,'l *l . l,8 a.Ii!i ..
i ( s 5795 N. River Road Freeland, MI 48623 May 29, 1982 Chairman Okrent Members ACRS
Dear Chairman Okrent:
Re: Midland Plant I am unable to attend the ACRS meetings in Washington this week, but I wish to submit the following statement regarding statements made at the May 20, 21st ACRS meeting in Midland, Michigan. Upon reviewing the Quality Assurance and outstanding soils issues in prepar-ing this statement, I am struck by the illogic and basic unfairness of dealing with operat orb license issues while such major questions as the soils issues remain unresolved. The ASLB hearing the soils case for the past year has not even heard the vast majority of the technical issues dealing with the soils remedial fixes, let alone offer their decision in this case. The one major area which has been covered in the hearing, quality assurance, is now about to be reopened as Mr. Keppler reconsiders his previous testimony on QA adequacy in the wake of the recent SALP report. In light of these facts, and considering Consumer's pattern of continuing inability or unwillingness to properly execute the sensitive soils work at hand, I urge this board to withhold any decisions leading toward the operators license until all the facts are in and the ASLB has had an opportunity to complete its soil settlement decision. Consumer's 1984 Dow steam contract deadline and Congressional pressures such as "the Bevill dates" for NRC licensing review have played a substantial role in bringing the operator's license review to this committee at this time. But Consumer's pressing licensing needs cannot be placed above the more important health and safety questions which must be resolved first. This committee's efforts to ascertain the underlying criteria on which the NRC has based their safety review, as pursued at the May 20 and 21st meeting are essential to the protection of public health and safety. The NRC's failure to demand rigorous adherence to objective criteria as seen in Quality Assurance, if repeated in technical areas, cannot provide assurance of a safe nuclear plant. The NRC summary of quality assurance issues amounted to a whitewash of a seriously deficient QA history as documented in public records. From the QA review I
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. Chairman Okrent Page 2 May 29, 1982 ,
j .- presented May 20th, the ACRS could not begin to understand or critique QA adequacy. l j 1 fear that the technical review summaries are similarly flawed, leaving the ACRS without the necessary data base for their assessment of NRC regulation. The ACRS - is to audit and evaluate the NRC review, but is relying on the NRC's own presenta-i tions of the facts--or those of the applicant, rather than objective sources. I hope this committee is able to overcome the obvious shortcomings of this system, and somehow achieve an objective and critical analysis of this nuclear < i plant. The public deserves no less. Thank you for your consideration of these i important matters, i
- Sincerely, M
' Barbara Stamiris 2 I l 1 I 4
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r . i INTERVENOR RESPONSE TO NRC QUALITY ASSURANCE ASSESSMENT AT 5/20/82 ACRS MEETING At the May 20, 1982 ACRS meeting on Midland, and in the SER, the NRC has con-cluded that Consumer's QA program has been and will be acceptable. I wish to document the public criticism I made of that assessment. i The QA history of this plant reveals major QA problems in virtually every construction area undertaken. Consumer's 54Ra.23 response lists the major construc-tion activities prior to 1977 as "aolls, rebar and embeds, concrete cadwelding, structural steel erection and liner plate erection." The NRC's October,1979, Chronological Listing of Major Events identifies problems in rebar and embeds in 1975-76, concrete and cadwelding in 1970 and 1974 (ALAB 106, LBP 74-71), and in tenden sheath omissions and liner plate bulge in 1977, and containment post ten-sioning errors in 1979. Since that time, electrical and piping have been the major construction areas in addition to soils remedial work. The NRC's recent 1980-81 SALP evaluation rates - these three areas negatively today. As early as 1972, the ALAB 106 Board found that "neither the applicant nor the architect-engineer has provided reasonable assurance that the QA program will be implemented properly." Suspension of the construction permits was considered then, and again in 1974 as a result of QA deficiencies. Yet the NRC and Consumer's have repeatedly defended inadequacies, cited imporvements, and given their reason-able assurance to licensing boards that proper QA would ensue in the future. Most recently a " reasonable assurance" conclusion was given by the NRC prior to the commencement of the soil settlement hearing intended to decide that very i question. It was based on the May 1981 team inspection and served as a condition for a stipulation agreement between the NRC and Consumers that the soils QA break-
.* ( 2
. down would not be litigated. As a result of the stipulation, the NRC drew a line between their consideration of pre and post December 6,1979: QA problems, as an NRC witness admitted (p. 3869). I found that QA problems occurring after Decem-ber 6,1979, were minimized, defended, or overlooked by the NRC, while previous problems were addressed straightforward 1y in the soil settlenent hearing.
During the May 1981 team inspection, NRC inspectors actually had copies of the proposed QA Stipulation with them during the inspection. The May 81-12 inspec-tion served as the basis for Mr. Keppler's reasonable assurance judgement and the subsequent QA Stipulation. This inspection and personal visit by Mr. Keppler, at the March 13, 1981, invitation of Consumer's James Cook represented Consumer's best- QA effort. It is a significant inspection because, stripped of its summary judgements and conclusary statements, it reveals surprisingly inadequate construc-tion and QA implementation on the part of Consumer's. Noncompliances and deviations were identified in eight of the eighteen areas inspected, particularly in the ongoing construction areas of soils, electrical, and piping, and an I.A.L. was a issued. The body of the May 81-12 inspection findings, as well as the two negative SALP reports for 1980 and 1981, contravene the NRC's prepared testimony of QA ade-quacy for- that same time period. Yet even if we were to accept the NRC?s June 1981 position that "as a result of revisions, improved implementation, and other 4 factors . . . (there is) reasonable assurance that QA and QC will be appropriately implemented 'with respect to future soils construction activities, including reme-dial actions," we once again have the passage of time to prove this assurance wrong. i , For as the most recent I&E reports (82-05 & 82-06) document, the soils reme-dial work has not been properly implemented. QA deviations, and misleading infor-mation in the remedial work have led to a consideration of escalated enforcement. And Director Keppler l's' currently reconsidering his QA testimony in light of these events C . 2 -ti.4 10._ wEluation.
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3 Yet in presenting the NRC assessment of QA to the ACRS the NRC minimized Mid-land's poor QA history saying "following each major problem period the licensee has been responsive." Mr. Little of Region III went on to define the NRC criteria for judging the QA program adequate as being that the licensee identifies his pro- , blems. However, the records of audits and inspections reveal that once problems are identified, adequate ccrrective or preventative actions do not follow. Mr. Little said that when the NRC conducted its follow up soil settlement investigations in December 1980, that the licensee had met its soils commitments, aside from some minor FSAR review and documentation problems. This statement overlooks and avoids the reality of Consumer's inadequate soil settlement responses. Specifically the December follow up inspection, 80-32, indicates that 40% of Consumer's soil settlement commitments from questions 1 & 23 "either have not been completed by CP Co. or the action taken was considered insufficient." The 80-32 FSAR re-review problems were extensive, and particularly significant because Consumer's own audits had revealed the problems several months earlier, but failed to correct them. The problems with relaxations of procedures or failure to fol-low approved procedures were paralled to the original FSAR review problems which are cited in the December 6th order. Yet only at NRC insistance were proper proce-dures implemented in what was termed the re re review. The 80-32 inspection resulted in three noncompliances involving inadequate design control, one involving inade-quate corrective actions (the same criteria cited in the December 6 order), and numerous unresolved items.
"Ihe second soils follow up inspection, 81-01, involving onsite implementation of soils commitments resulted in four noncompliances and a deviation. Again the significance of these soils problems comes from their remarkable similarity to the original soils errors of 1975-78, and als6 from the fact that once again they repre-sented a continuation of problems which Consumer's own audit findings had identified j i
six months earlier. .
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k 4 , Six CPC audit findings (M-01-11-0-01 to 06) of July 1980 involved soils test-ing errors ins (1) relative density; (2) test elevations; (3) inadequate inspec-tions; (4) "Q" placement of soils; (5) equipment verifications; and (6) inadequate geotechnical review. Consumer's considered, but did not initiate a soils stop-C work at this time (3167-70). And the soils testing errors of January 1981 (81-01) v which were taking place under the direction of an unqualified geotechnical engineer, i contrary to their NRC commitment, indicate that corrective actions a6ain did not r, follow the earlier identification of the problems. 3 Regarding this series of July 1980 audit findings, I believe the NRC was remiss in not considering them in their post-December 6,1979 testimony, despite the fact , 4 that I had personally presented these findings informally to NRC members in Jan- e uary 1981 and thereafter. In cross examining witnesses about the apparant contra-dictions between these audit findings and their testimony that soils problems were resolved in 1980, the significant similiarity of these problems to past prob-lems was denied (p. 2590). . The.May 81-12 inspection revealed similar audit deficiencies and failures Also to correct identified problems as indicated in the noncompliances cited.
' NRC inspectors identified five unresolved items regardin6 dewatering well plans (the only ongoing soils work) and found soils QA to be inadequately staffed (a repeat of the 81-01 deviation). As noted 'in the April 82 SALP report, "every inspec-tion involving regional based inspectors and addressing soils settlement issues has resulted in at least one significant item of noncompliance" despite the tremen-dous attention focused on soils.
In other statements before the ACRS, the NRC has credited Consumer's with undertaking a voluntary soils work-stop during the course of the hearing, since However, legally the soils prohibitions of the December 6 order were not binding. the evidentiary record of the soils hearing, and the ASLB April 30, 1982 memoran- l dum (p.13), confirms that certain soils work continued beyond this agreement with-u h.t.l.y f out NRC concurrence. f fI" N [' 3 h f ' - _ 1
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I i he NRC also granted specific concurrence for portions of the soils remedial In fact, the initial ,' work, which would have been prohibited by the December 6 order. j NRC testimony of the hearing (July 5,1981) was a complete reversal of j' 1 the licensing board's question and concern regarding the impact of ongoing soils f work. At the January 1981 prehearing conference, the board asked if there was a 4 any soils work that needed tobe stopped at the outset of the hearing because it threatened the soils problem remediation. De NRC "re-interpreted" this question to be Was there any soils work which needed to go forward because it threatened Consumer's construction schedule needs. The NRC answered this new question affirma-tively and proceeded to grant concurrence for 12 permanent dewatering wells on those grounds. The overwhelming preponderence of evidence represented in CPC and NRC repoets for 1980 and 1981 indicates that despite the programatic QA improvements, actual QA improvements, actual QA implementation has remained sorely deficient.
'Ihe Midland QA organization does not provide the required independence from cost / schedule responsibility because Vice president James Cook retains ultimate authority for the Midland project in both areas (p. 2054). This issue of QA cost /
schedule independence was criticized in Midland's ALAB 147 decision of September 18, 1973 yet is openly repeated in 1980-82. This summary of recent events just beB i ns to tell the story of QA inadequacy and NRC leniance at the Midland plant. The extent of design and construction defi-ciencies which has resulted at Midland over the years tends to be overshadowed by the predominant soil settlement problems, but is just as serious. f
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DP September 29, 1981 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION RECElVED II 0 BEFORETHEATOMICSAFETYANDLICENSINGB0A&.If5kh3,f3jjp AM 00T 61981 In the Matter of ) CONSUt1ERS POWER C0ftPANY ) Docket Nos. 50-3290-ObM81121 3:4 6
) . 50-330Ott-OLj (Midland Plant, Units 1 and 2) )
NRC STAFF BRIEF IN SUPPORT OF THE USE OF A SITE SPECIFIC RESPONSE SPECTRUM TO COMPLY WITH THE REQUIREMENTS OF 10 C.F.R. PART 100, APPENDIX A I. INTRODUCTION This brief explains how the use of a site-specific response spectrum coaplies with the requirements of Appendix A to 10 C.F.R. Part 100. It is in response to the concerns of the Licensing Board in its Memorandum of August 18, 1981 about the use of this approach. When the Staff in 1970 had reviewed the seismic criteria for eval-uating tne adequacy of the design of the Midland nuclear power plant, it detemined that a modified Housner response spectrum,1/ scaled or
" anchored" to a ground motion acceleration of 0.12g, defined the I
1/ The response spectrum originally approved in 1970 was a modified version of one published by Dr. S. W. Housner in 1959. (l,' -,'. 3 E) ! - -. N -
a , , . i, . 2 earthquake ground motion for which the plant should be designed.2/In the course of its subsequent review of the operating license application for this facility, the Staff questioned a) whether or not the standardized I modified Housner spectrum is an acceptably conservative representation of the ground motion of an earthquake represented by 0.12g acceleration, and b) whether or not 0.12g acceleration adequately represents the Safe Shutdown Earthquake for this site. The proper choice of a response spectrum is one of some immediacy. As the Board recognized in its May 5,1981 order: [w]e fail to see how we could meaningfully resolve the soils settlement questions--exceat to grant the full relief sought by the modification order l cessation of all soils related construction activities]--without ruling upon at least certain of the seismic parameters of the proposed remedial actions.3/ In its May 5 order the Board accepted a definition by the parties of how seismic issues would be addressed in this proceeding:
- 1. Seismic criteria, including an SSE, ground motion, and associated response spectra, would be established.
- 2. The analysis model for each structure as modified by the remedial action, which would include the basis on which the spring constants are to be derived, would be constructed.
2/- " Safety Evaluation by the Division of Reactor Licensing, U.S. Atomic Energy Commission, in the Matter of Consumers Power Co., liidland Plant Units 1 & 2, Occket Nos. 50-329 and 50-330," November 12, 1970, 6 3.3 and Appendix E. The choice of a maximum ground accel-eration value was derived from two earthquakes of intensity VI on - the Ibdified Mercalli scale. These events took place between 100 and.150 miles of the Midland site in 1883 and 1947. 3/ Prehearing Conference Orcer (Ruling upon Applicant's flotion to Defer Consider,ation of Seismic Issues Until the Operating Licensing Pro-ceeding and upon other matters), fiay 5,1981. M
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- 3. However, a judgment on whether or not the plant's --
structures as built conform to the final seismic criteria would await the later stages of the OL proceeding.4/ In its review of the operating license application, the Staff , suggested5/ that two approaches would be acceptable for characterizing the response spectrum associated with the Safe Shutdown Earthquake.5/ The first acceptable approach would be to determine the intensity of the Safe Shutdown Earthquake and describe its ground motion effects by using a standardized, site indeoendent response spectrum as defined in Reg. Guide 1.60. Although this spectrum is anchored to a ground acceleration value derived from the Safe Shutdown Earthquake's intensity, the shape of the spectrum is independent of the particular characteristics of any specific site. The second approach acceptable to the Staff would be to determine the nagnitude of the Safe Shutdown Earthquake and describe its ground
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motion effects by using a site-specific response spectrum. This spectrun would be derived from ground motion records of earthquakes which (a) were similar to the magnitude of the Midland site's Safe Shutdown Earthquake, 4/ ld at 11. 5/
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Letter from Robert L. Tedesco, Assistant Director for Licensing, NRC, to J. W. Cook, V.P. Consumers Power Company, dated October 4, 1980. , 6/ See 10 C.F.R. Part 100, Appendix A Sections III(c) and V(a)(1). W
l l (b) were recorded within 25 kilometers of the epicenter and (c) in geologic , conditions similar to the Midland site. II. ISSUES Concerned that this second approach, the site-specific response spectrum, represented a "probabilistic" approach inconsistent with the
" deterministic" approach to Appendix A 10 C.F.R. Part 100, found in Reg.
Guide 1.60, the Board requested briefs from Consumers Power Company ( Appli-cant) and the Staff, which analyze:
- 1. the compatibility of the site-specific response spectrum approach with the requirements of 10 C.F.R. Part 100, Appendix A, and -
- 2. the Commission's approval of, or other action, if any, with ,,
respect to any previous use of this approach. III. DISCUSSION A. The Use of a Site Specific Response Spectrum to define the Ground Motion Associated with the Safe Shutdown Earthquake satisfies all applicable requirements o' the Comission, including 10 C.F.R. Part 100, Aopendix A. To assure that a nuclear power plant can be operated without endangering the health and safety of the public, the U.S. Nuclear Regulatory Comission has required: (1) that a nuclear power plant be designed so that, if the level of earthquake ground shaking (the " vibra- , tory ground motion") associated with the Safe Shutdown Earthquake occurs at the site, the plant can be shut down in a safe and orderly manner D
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l . afterwards, without a loss of_ reactor coolant and without the offsite release of unacceptable levels of radiation, and (2) that a plant be designed so that, if the maximum level of earthquake ground shaking expected during the life of the plant (the vibratory ground motion asso- f ciated with the operating basis earthquake) occurs at the site, those features of the plant necessary for continued operation will remain functional without undue risk to the public. These two concepts are embodied in the Commission's regulations, in particular Appendix A to 10 C.F.R. Part 100.2/ Appendix A contains two principal requirecents:
- 1. an assessment of the controlling earthquakes for the site of a 4
nuclear power plant; f.e., identification of the safe shutdown earthquake and the operating basis earthquake, and l 2. a definition of the vibratory ground motion associated with these earthquakes, i.e., defining a response spectrum. t i 2/ General Design Criterion 2,10 C.F.R. Part 50, Appendix A requires that a nuclear power plant's structures, systems and components important to safety shall be designed to withstand the effects of earthquakes without loss of capability to perform that safety function. The plant's design basis must reflect appropriate considerations of the most severe earthquake reported in the region, the combined effects of normal, accidental and seismic conditions, and the impor-tance of the safety functions to be performed. 10 C.F.R. 9 100.10 requires the Commission to take into consideration the seismology and geology of a site in determining whether a site is an acceptable location for a nuclear power plant. How these standards are to be - met is found in Appendix A to 10 C.F.R. Part 100; this appendix identifies the seismic and geologic criteria for evaluating whether or r.ot the plant's site is acceptable under Part 100 and whether or not the design basis complies with General Design Criterion 2 under Part 50.
0
. _ _ (1) The Choice of a Controlling Earthquake Appendix A to 10 C.F.R. Part 100 requires the designation for the nuclear power plant site of a " safe shutdown earthquake," the earthquake I
which is " based upon an evaluation of the maximum earthquake potential considering the regional and local geology and seismology and specific characteristics of local subsurface material." Appendix A, Section III(c). The designation of a safe shutdown earthquake involves a multistep process, beginning with the collection of information on the seismic, geologic and engineering characteristics of the region and the site, ;L1., Section IV. Important components of this investigation include the listing of all historic earthquakes whicn affected or could reasonably be expected to have affected the site, Jd,,, Section IV(a)(5), the correlation of these earthquakes with tectonic structures or, if not reasonably correlated with such structures, with tectonic provinces, Jd . Section IV(a)(6), and the identification and assessment of any capable faults, Jd,., Section IV(a)(7) and (a)(8). On the basis of an evaluation of this information, a safe shutdown earthquake for the site is determined, jb!.. Section V(a)(1). Large his-toric earthquakes which can be correlated with tectonic structures are assumed to recur at that part of the structure nearest to the site, Jd., SectionV(a)(1)(1). Earthquakes in adjacent tectonic provinces, which are not associated with structures, are assumed to recur at the nearest edge of the province boundary, if that boundary is within 200 miles of l the site, Jb!.. Section V(a)(1)(iii) incorporating Section IV(a)(6). 6 -
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Earthquakes in the same tectonic provinceE are postulated to recur at the site, Id. Section V(a)(1)(ii). In essence, Appendix A requires the assumption that an historical earthquake associated with a particular 3 region or a particular structure will recur at the point in that region or on that structure closest to the plant site. If all these historic earthquakes were to recur as hypothesized, the one which would produce the maximum vibratory acce'leration at the site is designated the safe shutdown earthquake, unless geologic and seismic i l
; 8/ A tectonic province is defined as "a region of the North Anerican continent characterized by a relative consistency of the geologic l
structural features contained therein." Appendix A, Section III(a), 10 C.F.R. Part 100. Seismologic as well as geologic characteristics are used in defining the boundaries of each province, Public Service Co. of N.H. (Seabrook, Units 1 & 2) ALAB-422, 6 NRC 33, 61 (1977); Consolidated Edison Co. of NY (Indian Point-Units 1, 2 & 3), ALAB-436, 6 NRC 547, 562ff (1977). Both cases had further opinions in ALAB-561, 10 NRC 410 (1979); Seabrook was remanded to consider certain evidence concerning probabilistic methods for earthquake estimation in CLI-80-33, 12 NRC 295 (1980). The Standard Review Plan, Section 2.5.2, " Vibratory Ground Motion," NUREG-0800, indicates how seismologic data is to be analyzed in defining a tectonic province: - The applicant may choose to define tectonic provinces -l to correspond to subdivisions generally accepted in the literature. A subdivision of a tectonic province is accepted if it can be corroborated on the basis of detailed seismicity studies, tectonic flux measurements, contrasting structural fabric, different geologic history, differences in stress regime, etc. If detailed investigations reveal no significant differences between areas within a tectonic province, the areas ~ should be considered _to compose a single tectonic ~ province. The presentation should be augmented by a regional-scale map showing the tectonic provinces, j the earthquake epicenters, and the locations of geo-i logic structures and measurements used to define provinces. Acceptance of the proposed tectonic prov-inces is based on the staff's independent review of '. the seismicity, tectonic flux (Ref. 39), geologic I structure, and stress regime in the region of the site. I
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conditions indicate that an even more severe earthquake shoulo be chosen, Id. Section V(a)(1)(iv). (2) The Design of a Response Spectrum Appendix A to Part 100 then requires that the " vibratory ground i motion produced by the Safe Shutdown Earthquake shall be defined by response spectra corresponding to the maximum vibratory accelerations at the elevations of the foundations of the nuclear power plant structures...", Appendix A, Section VI(a). A response spectrum is a description for design purposes of vibratory ground motion associated with an earthquake or series of earthquakes.El This description is represented by a graph which plots vibratory ground acceleration, velocity and displ& Cement. Engineers use the resulting spectrum--rather than the intensity of the earthquake or only its peak acceleration--in the design of earthquake resistant structures. . . Although the regulations require that the spectrum represent an appropriately conservative description of the vibratory accelerations '1 associated with the safe shutdown earthquake throughout the frequency l 4 9/ As defined in Appendix A. Section 111(l), a response spectrum "is a plot of the maximum responses (acceleration, velocity or displacement) of a family of idealized single-degree-of-freedom damped oscillators . against natural frequencies (or periods) of the oscillators to a specified vibratory motion input'at their supports." M
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t range relevant to the design of a nuclear power plant,$E/ the regulations do not indicate how one is to design the required spectrum. What is clear from Appendix A, however, is that a response spectrum would be 0 inadequate in characterizing the safe shutdown earthquake, if its appli-cation ' ailed to account for the seismology and geology of the site and the surrounding region. As Appendix A indicates, the design basis for the maximun vibratory ground motion--a basis quintessential 1y derived from the response spectrum required for defining the SSE's vibratory ground motion, Jd., Section VI(a)--
"should be determined through evaluation of the seismology, geology, and the seismic and geologic history of the site and the surrounding region,"
see Id., Section V(a). In addition, the SSE's maximum vibratory accelerations-- which are defined by this response spectrum, Ijf., Section VI(a) _ "shall be determined taking into account the characteristics of the underlying soil material in transmitting the earthquake-induced motions," Jd., Sec-tionV(a)(1)(iv). What this means is that, before a response spectrum can be considered appropriate, it must account' for the characteristics of the geology underlying the site, the behavior of these underlying materials 10/ "The characteristics of the Safe Shutdown Earthquake shall be derived
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from more than one earthquake determined from paragraphs (a)(1)(1) through (iii) of this section, where necessary to assure that the
- maximum vibratory acceleration at the site throughout the frequency range of interest is included." Appendix A. Section V(a)(1) (iv).
It is well established that the " maximum vibratory acceleration" of Appendix A refers to maximum acceleration values having engineering significance, Pacific Gas & Electric Co. (Diablo Canyon, Units 1 and 2), ALAB-644, Slip op. at 67-72 (June 16, 1981), although the way in which isolated high frequency peaks without structural significance see Public Service are Co. identified has not been of New Hampshire definitively (Seabrook, Units 1addressed,
& 2), CLI-80!TI,12 NRC 295 (1980).
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during prior earthquakes and the engineering properties of these materials-- . including their characteristics in transmitting earthquake-induced motions. Id., Section V(a)(1)(iv) incorporating Section IV(a)(1)(a)(2) and (a)(4). At least two methodologies for designing and applying a response spectrum are consistent with the requirements of Appendix A: a site independent response spectrum, as defined in Regulatory Guide 1.60, and the site-specific response spectrum, as suggested by the Staff for use in this case. The site independent response spectrum defined in Reg. Guide 1.60 is a standardized spectrum that can be used at a wide range of sites to define the vibratory ground motion cf a large variety of earthquake inten-sities. The spectrum was derived from strong motion records of a large number of earthquakes of various magnitudes, recorded at various distances, and on varying site conditions. After the ground motion valu.es.of these records were normalized to the same acceleration, a spectral shape was derived representing the mean plus one standard deviation. After some smoothing of its shape, this response spectrum became the one associated with Reg. Guide 1.60. Because the standardized spectrum of Reg. Guide 1.60 was derived from earthquake recordings for a wide range of geologic and seismological conditions, the resulting spectral shape does not depend on the characteristics of any one site to which it is applied. When Reg. Guide 1.60 was promulgated in 1973, the number of strong motion records available for any single type of geologic and seismologic , conditions was relatively small. Thus, at that time any attempt to match specific site conditions with earthquake records would have failed for
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11 - __ lack of a statistically meaningful data set of records from analogous site conditions. To overcome this lack of data, the then available data - recorded in a' wide range of differing site conditions, was compiled, I and a shape enveloping this wide range of data was derived. The resulting overall shape, encompasing a wide range of soil conditions, may be used as a standardized spectrum, whose shape need not be nodified for the specific site conditions. When used to describe the vibratory ground motion for a specific site, the standardized spectrum is scaled to the ground acceleration level associated with the site's SSE. Because the standardized spectrum of Reg. Guide 1.60 has been designed to be applicable to a broad range of different site conditions, accounting for--or " enveloping"--a myriad of possible earthquake ground motions, it taay not be the most accurate representation of ground motion for any one
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site. When using this standardized spectrum, an applicant meets the requirements of Appendix A mandating a detailed accounting for site condi-tions, e.g., Appendix A, Section V(a)(iv), by showing that the site does not have any particular geological and seismological characteristics which indicate that the standardized spectra would not appropriately bound the SSE's ground motion. Because it must account for, or envelop, conditions for a myriad of sites, it is by definition a less accurate description of earthquake ground notion at any one site than a site-specific response spectrum would be. The site specific spectrum is designed to fit more precisely the - seismology and geology of the site and the surrounding region and that takes into account the engineering properties of the soil. As the 1 l l
). ; .,,.- I l }- - s , available ensemble of strong motion records has increased and as analy- , l tical techniques have improved, it has become technically possible to ! . design such a response spectrum that is specific to the site--and in i f the process to comply more closely with the mandate of'. Appendix A to ! account for site specific conditions. This development--the availability j j of means to portray accurately a spectrum representative of given site 4 f conditions--has led the staff to encourage for the last several years in i j appropriate ca'ses the use of a site specific response spectrum as an alternative to the approach in Reg. Guide 1.60.E l l The principle underlying the use of a site-specific response spectrum 4 is straightforward. Because earthquakes of similar magnitudes have been f found to have sinilar ground motion characteristics when recorded at ! similar distances from the epicenter and in similar soil conditions,E i . j an accurate representation of possible ground motion for an earthquake I of a postulated magnitude can be derived from analyzing an adequate j g The Standard Review plan for " seismic impact," Section 3.7.1, NUREG-75/087, states: 1 As noted in Regulatory Guide 1.60, there are site circumstan es where the design response spectra are more appropriately developed to suit the particular site characteristics. Design response spectra based upon site-dependent analysis must be derived considering ] in situ variable soil properties, a representative
- number of site earthqake records, vertical amplifica-1 tion, possible slanted soil layers, and the influence ;
j of any predominant soil layers. The finite element . i approach or equivalent should be used to consider j i variable soil properties and nonlinear stresT-strain i relations in the soil media. The procedures used to 1 obtain site-dependent design response spectra are j reviewed on a case-by-case basis. l g See Staff testimony. Kimball, p. 6ff. i J t i
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'6 set of recordings for similar magnitude earthquakes at similar sites elsewhere. To make this comparison, the data base for strong motion
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records is searched for all recordings of historical earthquakes of I similar magnitude to the chosen safe shutdown earthquake recorded close to the epicenter of the event,and recorded in similar geologic conditions. If the ensemble of recordings fitting these parameters is of sufficient size then the ground motion data for each of the records are plotted, and an idealized spectrum is drawn representing a mean-plus-one-standard-deviation. This idealized spectrum is the response spectrum specific to the site.
.Several reasons compel the conclusion that the use of a site-specific response spectrum is an appropriate and acceptable method of complying with Appendix A to 10 C.F.R. Part 100. First, the design and applica-tion of a site specific response spectrum is a more -accurate method for deriving at a site the ground motion associated with the Safe Shutdown Earthquake,ll/ JJ1.,SectionsV(a)(1)(iv)andVI(a)(1). Second, the design of the spectrum is based on an objective analysis of empirical records of earthquake ground motion, Jd,. Sections IV(a) and V(a)(1).
i Historical records for inclusion in designing the spectrum are chosen l based upon objective criteria, analytically related to the Safe Shutdown
- Earthquake under examination. Third, the design of a site specific spectrum 4
takes substantial account of the seismology and geology of the site, Jd., 13/ See Staff testimony. Kimball, p. 8. b% M
Section V(a), and of the characteristics of the underlying soil material in transmitting earthquake-induced motion. Ijf., Section V(a)(1)(iv). Only historical recordings made in substantially similar soil conditions are chosen for designing the site specific spectrum. In the final result, the design of a site specific spectrum is no more than the adjusting or tailo'Fing of a standardized response spectrum for the particular seismic and geologic characteristics of the selected site. (3) The approach used to review the seismic design adequacy of the plants in the Systematic Evaluation program The deterministic or empirical method for defining for the Midland site the specific response spectrum associated with the controlling earthquake differs materially from the methodology used in a probabil-istic analysis. A determinative response spectrum is derived directly 1 from a simple compilation of analogous historical records, whereas a probabilistic spectrum represents ground motion values associated with future earthquakes affecting the site which are predicted to occur at different magnitudes and at different distances from the site. Different spectra are computed, each with different relative probability that its values will be exceeded. The probabilistic response spectrum may or may not be a "more conservative" representation of ground motion than a response spectrum drawn directly from an appropriate set of historical records. Although denominated a " site specific spectrum" when so used, the probabilistic spectrum bears little resemblance to the deterministic , sits-specific response spectrum drawn for the fiidland site.
B. Site Specific Fesponse Spectra have been used for several years in Commission I.oceedings for evaluating the appropriateness of a nuclear power plant's design. In response to the second question of the Board concerning previous I use of site specific response spectra, counsel for the staff has learned the following from various members of the NRC technical staff.E The staff approved the use of a site specific response spectrum was in March of 1979 for the Tennessee Valley Authority's Sequoyah Nuclear Plant.E For Sequoyah, just as for Midland, the Staff in its review of the operating license application had questioned both the spectrum and the ground accel-eration value originally chosen. From thirteen sets of strong motion records from earthquakes of 5.3 to 6.2 magnitude, recorded within 27 kilo-meters of the event in geologic conditions similar to the rock underlying Sequoyah, a 84th percentile site specific response spectrum was calculated. In all material respects the procedure used in.Sequoyah was identical to tr st employed for designing the Midland response spectrum. Subsequently, the Advisory Committee on Reactor Safeguards reviewed the approach of the site specific response spectrum in depth and endorsed the use of thisprocedure.EI Subsequently, the Commission approved the issuance of, licenses for both Units 1 and 2 of the Sequoyah Plant. M/ Appropriate affidavits will be subsequently provided. M/ Safety Evaluation Report, Sequoyah Nuclear Power Plant, Untis 1 & 2, Tennessee Valley Authority, Docket Nos. 50-327 & 50-328, March 1979, NUREG-0011. Section 2.5.3. 16/ Letter from ACRS Chairman M. Carbon to Chairman J. Ahearne, " Interim . l Low Power Ooeration of Sequoyah Nuclear Power Plant, Unit 1," dated December 11, 1979; see attachment A.
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Another case in which the staff has utilized a site specific spectrum approach as part of its review process which has also been reviewed by ACRS was in the proceedings for San Onofre 2 & 3.11I During the OL review for the San Onofre plant, several methods were utilized to assess the adequacy of the seismic design response spectra approved during the construction permit stage. It should be noted, however, that the San Onofre site is very different from the Midland site in that it is located approximately 8 kilometers away from a capable fault. In the San Onofre review, strong motion recordings from locations with site con-ditions similar to San Onofre were utilized. In that case 56 recordings from 7 earthquakes of magnitude approximately equal to 6.5 were used. Since this site is in close proximity to a capable fault, special treatment of the data was required in order to be able to estimate response spectra in the nearfield. A site specific response spectrum for a magnitude 6.5 was construted. Since a magnitude 7.0 was esti-mated by the Staff, it was necessary to multiply the M=6.5 peak acceleration and spectrum by appropriate scaling factors to account for the higher magnitude. This is done because there are essentially too few records available at magnitude 7.0 to develop a site specific spectrum directly. Although seismological circumstances exist at San Onofre which require special consideration and modification of the deterministic site specific spectrum which was developed, what resulted was inherently 17/ Safety Evaluation Report (Geology & Seismology), San Onofre Nuclear Generating Station, Units 2 & 3. Southern California Edison Co., et al ., December 1980, NUREG-0712, Section 2.5.2. M
a site specific spectrum. This approach was reviewed by the ACRS in its overall assessment of the seismic design adequacy of San Onofre and was extensively discussed as part of the San Onofre evidentiary hearing in July,1981.30/ The Staff is in the final stages of approving site specific spectra, designed using a similar methodology to that employed for the Midland spectrum, for Detroit Edison's Enrico Fermi Plant 1E/ and for TVA's Watts Bar and Bellefonte plants. The Staff has also used site specific response spectra to evaluate the adequacy of the spectrum proposed by the Applicant.20/ On occasion, site specific spectra are drawn to estimate the long period (low fre-quency) effects on a site due to the occurrence of a large magnitude earthquake at long distances from that site. The site specific spectrum is drawn for this purpose in much thep same way as was the one for the Midland site. Relevant historical records, statistically recomputed to account for the attenuation, are plotted and a site specific spectrum drawn. The site specific spectrum fo this very large, distant event is compared with the design spectrum to sure that the latter appropriately accounts for a recurrence of the 18/ June 22-August 5,1981 evidentiary hearings, Docket Nos. 50-361-OL
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and 50-362-OL. , i
--'19/ Safety Evaluation Report, Detroit Edison Co. (Enrico Fermi Unit 2),
NUREG-9798, July 1981, Section 2.5.2. 20/ See NUREG-0098, Newmark and Hall, " Development of Criteria for
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Seismic Review of Selected Nuclear power Diants." 6
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i~ .. distant event and structures at the site in the appropriate ' frequency range are properly designed. Site specific spectra have been used in this way to evaluate design I spectra for example at the Grand GulfE! and Fermi sites.E IV. CONCLUSION The use of a site specific response spectrum complies with the requirements of 10 C.F.R. part 100, Appendix A. pectfully submitted,
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Ja. s H. Thessin N. h' & Counsel for NRC Staff Dated at Bethesda, Maryland this 29th day of September,1981. 21/ Safety Evaluation Report. Mississippi Power & Light Co. (Grand Gulf Units 1 & 2), NUREG-0831, September 1981, Section 2.6.2. M/ Op. cit. Section 2.5.2.
' - ATTACHt1ENT I Source: Sequoyah Nuclear Plant, Unit 1 & 2
- Safety Evaluation Report Supplement No.1 NUREG--00ll February, 1980 APPENDir O ADVISORY COMMITTEE ON REACTORS SAFECUAR05 -
GENERIC MATTER AND LETTER I UNITED STATES [ e NUCLEAR REGULATORY COMMISSION
! } ,~g a, s AQvisORY CCMMITTEE CN REACTOR $APEGUAROS wasms.stois. o. c. noiss i.D. . ,4. /,/
s.. .a December 11, 1979 i
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Honoeable John F. Ahearne Chairman U.S. Nuclear Regulatory Camelssion Washington, DC 20555
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Deer Dr. Ahearnes Curing its 236th sweting, December 6-4, 1979, the Comnittee considered a pecposal for interia, low powe operation of the Seqtmysh Nuclear Powr Plant, unit 1. At its 229th reeeting, May 10-12, 1979 aid also at its 228th meeting, April 5-7, 1979 the Corsaittee had considered aspects of the application of the 7emessee Vallwy Authority (hereinafter referred to as the Applicant) for authorization to operate h Sequoyah Nuclear Power Plant, Units 1 and 2. A tour of the facility was made by members of the Subcammittee on January 24, 1976 and the application was considered at subcomalttee meetings on March 12, 1979 and on November 5, 1979. During
- its review, the comittee had the benefit of discussions with representa-tiws and consultants of the Applicant, the Westinghouse T.loctric Corpore-tion, and the Nuclear Regulatory Commission (NRC) staff. W e Coneittee also had the benefit of the documents listed. Se Committee reported on the application for a construction permit for this plant on February 11, -
1970. The sequoyah Nuclear Power Plant is located on the weet bank of the , Tennessee River in Man 11 ton County in southeastern Tennessee approximately 17 miles northeast of the center of chattanooga, Tennessee. Construction ; on Unit 1 is essentially complete and construction of Unit 2 is about 904 complete. Each unit will ut111:e a four-loop pressurized water reactor nuclear steam supply system having a powr level of 3411 MWt and an ice * ' condenser systen enclosed within a free-standing steel containment wasel Witch is marroeded by a reinforced concrets shield building. De ice condenser system is similar to that used in the McGuire Nuclear Station and I the Donald C. Cook Nuclear Plant. De Applicant has modified the ice condenser syntes as a result of the operating owperience gained in the Donald C. Cook Nuclear Plant. Se Applicant and the NRC Staf f have mede plans to monitor the performance of the ice condenser contairments at the I Sequoyah Nuclear Powe Plant (Generic Itan 63 in the Report ACRS No. report, 7,' ' dated Status of Generic Itans Relating to Light Meter Raseters: March 21, 1979). Se Committee recomends that each plans be implemented.
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- December 11, 1979 Ronorable John F. Ahearne The sequoyah Nuclear Plant will utilize 17x17 fuel assemblies. A surveillance program has been developed by the NRC Staff to follow the behavior of these assemblies, and data are beirq obtained froen several plants now in operation in which such assemblies have been installed for test. Experience to date has been satisfactory. The Committee wishes to be kept informed of the results of the various 17x17 assembly inspections and test programs now under way.
De sequoyah site is considered by the NRC Staff to be within the Southern Valley and Ridge tectonic province. The maximurs historic eart!*;uake within this tectonic province is the 1897 Modified Mercalli Intensity (ftc) VIII earthquake in C11es County, Virginia. D.tring the construction per:mit review, the NRC Staff. concluded that a modified Housner respense spectrus anchored Since that time, at 0.18g was acceptable as the safe shutdown earthquake. the NRC Staff has adopted methods Wich Wuld characterize an MM VIII earthquake with the more conservative response spectrtan specified in Regulatory Guide 1.60 anchored at 0.25g. i " The Applicaat, in response to NRC Staff recomendations, has evaluated the Segacyah design using a site-specifle safe shutdown response spectrus I developed ft::rs Nor*.h American and Italian strong motion records of appro-priate eagnitude and epicentral distance and has compared the probability of the safe shutdown earthgake betrq exceeded at Segacyah to that at [ other Tennessee Valley Authority plants that meet the Standard Review l Plan. It has been concluded that the risk of exceeding the present design spectrus and the risk of exceeding the site-specific spectrt-t are corsparable o and that the probability of exceeding the safe shutdown earthquake is not i The NRC r appreciably different from that for other plants in this region. 6 Staff has reviewed the Applicant's evaluation and has concluded that the Sequoyah plant is adegate to withstand the effects of the safe shutdom F earthquke without loss of its capability to perfor: required safety functions, ne NRC Staff, to verify their judgments regarding structural and component design margins, has performed an audit of the design margins in representative critical sections of the reactor and auxiliary building structures and in representative components required for safe shutdown. The Corsnittee recorsnends that this program for 'the gantific tien of the seismic design nargin be continued and expanded to the extent necessary to ensure that all structures and equipsent necessary to accor plish safe shutdown do indeed have some margin. Similar recorrendations have been made by the Corvaittee for the North Anna Power Station, Units 1 and 2, and the cavis-Besse Unit 1 in its reports dated January 17, 1977 and January 14, 1979. 21s matter should be resolved on a schedule and in a manner satis-factory to the Staff. The Dnergency Core C$oling Systerns (ECCS) for the Sequoyah Nuclear Plant incorporate the Upper Head Injection (VHI) system. Se NRC Staf f has cornpleted its review of the Westinghouse Electric Corporation ECOS eval-uation model for plants equipped with UMI, and the Corwittee in its April 12, 1978 repstt on the Md3uire Nuclear Station has concurred with the l D2 I
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[* _ E Honorable Jonn F. Ahearne Decerber 11, 1979 Staff's conclusica.s. Ce NRC Staff has completed its review of tna ! application of this a;9 roved evaluation rodel to the Sequoyah Nuclear Plant and conears w:th the Applicant. ne co-mittaa hsr t een reviewing the circumstances relatirq to the recent accident, at the " nree Mile Island Nuclear Station Unit 2 and has e.ede recocmere'at.!cn* *-r 6-rovements in plant design and operatirq procedures which shoeld ha ersiderc.i for all pressurized water reactors. Se Carnittee is e o inuing its review of the i=p11 cations of this accident and expects t: ptride additional recoernendations. It is expected that thesa re:cc +nitt!cns will be considered and implemented as appropriate ty the NRC Sta'f. te Consittee wishes to be kept informed. 2e NRC Staff ras identified a number of outstanding issues, confire. story issues, and cs ssing conditions, not related to W.:-2 accident cor. sider-ations, which have not been specifically addressed in this report. Sese issues should be resolved in a ranner satisfactory to the NRC Staff. Various generie prabless are discussed in the Cornittee's reprt,
- Status of Generic Items Relating to Light-Rter Reactorst Repet No. 7,' dated March 21, 1979. Bose proble.s relevant to the segmyah Nuclear Plant should be dealt with by the NRC Staff arx! the Applicant as soluttens are found. Se relevant iter.s are: 54-60, 63-65, 69, 71, 72, 74, and 76.
We NRC Staf f has ret cocpleted its review of the Sequoyah Nuclear Power Plant application for a rermal operating license at full power, and various implications of the tree Mile Island accident on the Sequoyah Pla*.; ressin to be decided. Be ACRS has not cocpleted its own review in regard to these raatters. De Applicant has propsed a prograts of interim low pwr operation to provide improved operator training and the developnt of additional ex-per!. mental information on the behavior of a nuclear unit and its systems under transient conditiens. De Applicant has propsed a special test series which includes the following: ..
- 1. Natural circulation following a simulated reactor trip.
- 2. Natural circulation followirq a simulated loss of offsits .
Powr.
- 3. Natural circulation with loss of pressurizer heaters.
- 4. F.ffect of stears generator isolation on natural circulation.
- 5. Natural circulation at reduced pressure. .
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- 6. Cooldown capability of the chargirq and letdown systers.
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3 -- {, Honorable John F. Ahearne December 11, 1979
- 7. Seat terreval following a simulated loss of onsite and of fsite AC power.
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- 8. Establishment of natural circulation from stagnant flow conditions.
9,. Boron mixing and cooldown. The NRC Staff plans to review the proposed experimental program in detail to assure itself that all safety-related aspects are being dealt with appropriately. De Connittee wishes to be kept informed. The NRC Staff advised the Committee that it will require that 7/A's
- emergency procedures for Sequoyah be reviewed by Westinghouse. D e NRO Staf f also stated that an acceptable emergency plan will exist prior to reactor operation.
The Comittee believes that there is reasonable assurance that the Sequayah Nuclear Power Plant, Unit 1 can be operated on an interim basis up to power levels of about five percent of full powe without undue risk to the health and safety of the public. Subject to approval of the detailed test program by the NRO Staf f, the Cerraittee recornwnds approval of an interim lov ; ewer license for the purposes proposed. . Sincerely, N
- Max W. Carbon i
Clairman References I
- 1. Tennessee Valley Authority,
- Final Safety Vialysis Report, Sequoyan Nuclear Powr Plant,* Voleurs 1 to 13, and Amendments 1 to 61.
- 2. U.S. Nwlear Regulatory Comission, " Safety Evaluation Report Related t to the operation of Sequoyah Nuclear Plant Units 1 and 2,* NE-C311, March 1979.
- 3. Letter from L. M. Mills, T/A, to D. 8. Vassallo, NRC, dated October 31, 5
1979, containing revised responses to the Lessons Learned Require-ents.
- 4. IAtter, L. M. Mills, T/A, to L. S. Rubinstein, NRO, dated October 30, *
? 1979, containing responses to ACRS questions.
l 5. Letter from L. M. Mills, T/A, to L. S. Rubinstein, N*AO, dated October 23,
' 1979, containing information on natu.1 circulation in Sequoyan, Unit 1, and Diablo Canyon, Unit 1.
- 6. tatter from L. M. Mills, T/A, to D. 3. Vassallo, NRO, dated October 12, 1979, containtrq responses to ACRS recorwendattor.s.
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Nonorable John F. Ahearne - 5- December 11,1979
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- 7. tatter from L. M. Mills, TVA, to D. 8. Vassallo, NRC, dated Septat*.ber 7, .
1979, containing responses to the Short-Term Recorcendations of the Lessons
- f. earned Task Force. t
- 8. tatter from L. M. Mills, TVA, to D. B. Vassallo, WRC, dated July 12, 1979, I containing responses to NRC-IEE Bulletin 79-06A and ACRS recormendations.
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l u,_' x)*a. - p September 29, 1981 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 0 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of CONSUMERS POWER CCMPANY Occket Nos. 50-329 CM & OL 50-330 OM & OL (Midland Plant, Units 1 and 2) TESTIMONY OF JEFFREY K. KIMBALL
- 01. Please state ycur name and position with the huelear Regulatory Commission.
A1. My name is Jeffrey K. Kimball. I am a Seismologist / Geophysicist with the U.S. Nuclear Regulatory Comission assigned to the Seismology Section of the Geosciences Branch, Division of Engineering, - Office of Nuclear Reactor Regulation. Q2. Have you prepared a statement of professional qualifications? A2. Yes, a copy of this statement is Attachment 1. Q3. Please state the nature of the responsibilities that you have had with respect to the Midland Plant. A3. I was assigned as the seismologist reviewer for the Midland , plant in July 1980 and am presently responsible for the FSAR review and SER preparation as it related to seismology and geophysics. Q4 Please state the purpose of this testimony. . A4 The purpose of this testimony is to discuss the resolution of an open item involving the characterization of the Safe Shutdown Earthquake vibratory ground motion for the Midland site.
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Se, 7 ,' - - 1 . 2-During the construction permit review (Midland CP SER November < 1970) the Staff's seismological advisor, the Seismology Division of the National Ocean Survey (N05), had concluded that an acceleration of 0.12g I resulting from an intensity MMI=VI earthquake (maximum historical earthquake within 150 miles of the site) would be adequate for representing the ground motion from the Design Basis Earthquake now called the Safe Shutdown Earthquake (SSE). In its operating license review the Staff has followed the tectonic t
' province approach of Appendix A to 10 C.F.R. Part 100 to detemine the vibratory ground motion corresponding to the Safe Shutdown Earthquake.
As discussed below, in using the above approach the size of the controlling earthquake for this site has changed over that which was determined at the CP stage. .
. Second, during the operating license review the Staff has questioneo the characterization of the ground motion (the response spectrum) from the controlling earthquake. The Staff was concerned about the use of a modified Housner spectrum anchored at 0.12g. It was detemined that the design response spectrum was no longer a conservative representation of i the ground motion as used.
QS. How has this open item been resolved? AS. This open iten has been resolved using state-of-the-art
> seismological information and data analysis. The Staff has concluded that the site-specific spectrum proposed by the Applicant (as identified .
by the 84th percentile in figure 1) is appropriately conservative to l represent the free field Safe Shutdown Earthquake response spectrum (vibratory ground motion) for the original ground surface at the Midland i m a b e
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si te. The Staff has further concluded that the site-specific spectrum proposed by the Applicant (as identified by the 84th percentile in figure 6) is appropriately conservative to represent the Safe Shutdown i Earthquake response spectrum for the top of the plant fill. This conclusion is based on several elements and is discussed at length below. TECTONIC PROVINCE: From the standpoint of the surface geology, the Midland site lies within the Central Stable Region described by Eardley (1962). The Central Stable Region is a region of relative consistency of surface geologic structural features characterized by a series of arches, basins, and domes formed during the Paleozoic Era. King (1964,1969) describes the area as "platfom deposits on Precambrian foldbelts." The site is situated near the center of the Michigan Basin, a regional structural i, basin that underlies the southern peninsula of Michigan and parts of adjoining states. The Staff has recognized that the surface geology of the Central Stable Region does not explain the fact that different areas of this large region exhibit different levels of seismicity activity. This observation has recently been substantiated by the work of Barstow, g
- d. (1981), who developed an earthquake frequency map of the Central and Eastern United States. Their work shows that the Midland site region has had from 0-3 earthquakes per 11,689 square kilometers in the period 1800 to 1977, while other areas of the Central United States (CUS) have -
experienced from 4-32 events in the same time frame. Earthquakes typically occur at depths (below ground surface) of 5 to 20 kilometers in the CUS; therefore, the relevant explanation of the geologic mechanism O%
l causing earthquakes is to be found in the geologic structural features at these depths rather than those at the surface. Levels of seismic activity are the best available means of inferring the geologic 1 mechanisms causing earthquakes. In its initial operating license review, in order to define the seismotectonic region for the Midland site, the Staff asked the Applicant to analyze the level of seismic activity for the area within 200 miles of the site, when compared to other 200-mile areas of the Central Stable Region (Question 361.7 issued on February 14, 1979, Applicants' response dated September 28,1979), because the Staff was aware of the variation of earthquake potential in the CUS. This was an attempt to resolve the broad Central Stable Region (as defined by surface geology) int'o seismotectonic regions (the " tectonic province" defined in Appendix A to f 10 C.F.R. Part 100), which would more adequately explain the diversity of the underlying geologic structural features as implied by the widely divergent levels of seismic activity within the Central Stable Region. This approach is consistent with Standard Review Plan Section 2.5.2. In addition, the Applicant has more recently submitted a report entitled "Part III-Site Specific Response Spectra-Seismic Hazard Analysis, February 1981." As discussed in a subsequent section of this testimony, these results have been used by the Staff in the overall decision making process in arriving at the Safe Shutdown Earthquake response spectrum. CONTROLLING EARTHQUAKE: , In order to expedite our review, recognizing the added time needed to resolve the level of seismic activity near the Midland site, for de4 fining seismotectonic provinces, the Staff initially asked the O O _1 --
Applicant to assume that the Central Stable Region could not be subdivided. The largest historical earthquake, in tems of intensity, in the Central Stable Region which has not been associated with tectonic structure is the March 9,1937 Anna, Ohio event. The intensity of this earthquake was MMI=VII-VIII (Coffman and Von Hake,1973). In tems of magnitude, the Staff has observed that the 1937 Anna earthquake (Magnitude of 5.0-5.3, both instrumental and interpreted, Nuttli and Brill,1981, Nuttli and Herrmann,1978) along with other Central United . States events have similar magnitudes. These events include, for example, the July 27, 1980 Kentucky earthquake (Magnitude of 5.2 instrumental, Zimmer SSER,1981), the May 25, 1909 Northern Illinois earthquake (!1agnitude of 5.1, interpreted, Nuttli and Brill,1981) and the June 18, 1875 Anna Ohio earthquake (Magnitude of 5.3, interpreted. (\ Nuttli and Brill, 1981). The Staff and Applicant then utilized the above earthquake information in developing the site-specific spectra. This was done before sufficient information had been developed to resolve the issue of seismotectonic provinces. The resolution of this issue has resulted in a lower controlling earthquake than the Anna-type event. GROUND MOTION CHARACTERIZATION; RESPONSE SPECTRUM: In the OL review the Staff has considered two alternative procedures for characterizing the response spectrum from the Safe Shutdown Earthquake. These procesures have been discussed in detail in Staff SER's (Sequoyah, March 1979; Femi, June 1981). Both procedures are also . discussed in the Standard Review Plan, Section 2.5.2, which represents approaches and practices the Staff considers acceptable methods of l establishing conformance with the Nuclear Regulatory Commission ) l l
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regulations. The Staff outlined these methods to the Applicant by letter l (R. L. Tedesco to J. W. Cook dated October 14,1980). The first procedure involves using the intensity of the maximum historical earthquake, which has not been associated with a tectonic structure within the tectonic province of the site, and the reference acceleration determined by using the trend of the means relating peak acceleration to intensity (Trifunac and Brady,1975). The peak acceleration is then used as the high frequency anchor point for the spectrum of Regulatory Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear power plants." This procedure produces a site independent response spectrum (a standardized response spectrum only dependent on the reference acceleration). The second procedure involves determining site-specific earthquake
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response spectrum (SSRS). Typically this procedure involves the collection of acceleration time histories from similar magnitude earthquakes (similar to the safe shutdown earthquake for the site), recorded at appropriate distances, and similar site conditions to Midland, which is a soil site. This procedure produces a site dependent response spectrum (a non-standardized response spectrum dependent upon which acceleration time histories are chosen). Each of these three items (magnitude, distance, site conditions) is briefly described in the context of site-specific spectra, for the Midland site.
- 1. Magnitude - the Staff uses the maximum historical earthquake -
which has not been associated with a tectonic structure within the tectonic province of the site. Past Staff practice has been to utilize acceleration time histories from earthquakes whose i l l
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magnitudes are within about 2 0.5 magnitude units. For example, if I we assume a magnitude of 5.3, we would utilize records within the magnitude range of 4.8 to 5.8, i.e. , magnitude of 5.3 2 0.5. This i range is chosen to ensure that a large enough, yet reasonable, data sample is collected.
- 2. Distance - the Staff has suggested that acceleration time -
histories collected be less than 25 kilometers from the earthquake source, based upon three reasons. First, is that, within the meaning of Appendix A of 10 C.F.R. Part 100, the Safe Shutdown Earthquake is assumed to occur near the site. Second, is that the actual acceleration time histories recorded at distances less than about 25 kilometers which are used for the SSRS do not include attenuation differences (differences which exist betwe5n the eastern and western United States, see for example, Nuttli,1981; Chung and Bernreuter,1981). Third, is that Gupta and Nuttii (1976) have determined this distance to be the approximate distance (20-25 kilometers) to which maximum epicentral intensities are felt in the CUS, i.e., distance to which the ground motion is assumed to be roughly equivalent.
- 3. Site Conditions - these are determined by the subsurface geologic conditions at the siting location. In particular, both the physical layer properties of the soil or rock, (such as its shear velocity) and the overall layer characteristics (such as discontinuities in physical properties between layers) are taken into account.
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Both of the above methods (site independent and site dependent) have their merits. However, as discussed below, it is the Staff's position . that the site-specific spectra results are more realistic descriptions of I earthquake ground motion. The reasons for reaching this conclusion are that:
- 1. Because of the way the Staff uses site dependent and site independent procedures (as discussed previously) a distinction between magnitude and intensity is needed. Magnitude is a better ,
estimator of earthquake source strength than intensity. Intensity is a subjective description of the damage and felt effects to buildings and the general public. In some cases intensity assignments can be highly dependent on the age of buildings, local soil conditions or the distribution of population. Magnitude is an instrumental measurement. It can be measured over great distances
. with consistency, and is less dependent on local soil conditions.
It should be noted, however, that much of our historical information on earthquake occurrence is dependent upon intensity observations since seismic instruments were not present.
- 2. Regulatory Guide 1.60 standardized response spectrum was developed using earthquake acceleration time histories with a wide range of magnitudes, recorded epicentral distances, and site conditions (rock, soft soil, stiff soil). It is therefore, by )
definition, not specifically designed for use on any one type of , > site. It was specifically developed for use with differing reference peak accelerations (anchor points) to estimate different I earthquake conditions. The advantage of such a method is that it
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allows for some standardization and relative ease of use. However, disputes still arise over what the proper reference acceleration (i.e., "g" value) to use for specific design decisions.
- 3. Determination of a site-specific response spectrum from strong motion acceleration time histories of the appropriate size earthquake, distance and site conditions allows for the direct estimation of the response spectrum at all frequencies rather than the need to develop a reference acceleration. That is, a specific "g" value is not needed because the response spectrum is derived directly using the collection of acceleration time histories. Hays (1980), for example, s+ates that, although care should be exercised in developing a site dependent spectrum, its use is more appropriate than site independent procedures. "Unlike the site-independent s procedures, site-dependent procedures will generally produce ground motion parameters that correspond closely with those expected on the basis of the seismological and geological conditions at the site."
(Hays 1980). MIDLAND SITE-SPECIFIC SPECTRUM: The Applicants' consultant, Weston Geophysical Corporation, developed a site-specific spectrum for Midland based upon the assumption that the controlling earthquake was an Anna-type event. This was accomplished by searching the earthquake strong motion data base for accelerograms which fit the magnitude (5.3 .5), distance (less than - about 25 kilomet'ers), and site conditions (soil profile at the original ground surface with about 300 feet of stiff material overlain by 40 to 50 feet of softer material) for Midland. The Applicant met with the
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W Staff on December 5,1980, to discuss this work and in February 1981 docketed a report on the characteristics of theAt4hquake ground motion at the top of the glacial till, entitled " Site Specific Response Spectra: Part 1 Safe Shutdown Earthquake Original Ground Surface." As shown in - Table 2 of the Part I report the range of magnitudes and distances of acceleration records collected were magnitude 4.9 to 5.5, distance 7 to 33 kilometers. The acceleration time history records which were in the distance range 25 to 33 kilometers were included because these records increased the data set withcut affecting the response spectra (see section on site-specific spectra sensitivity tests). Results from Part I report are shown in Figure I which displays the median (50th) and median plus one standard deviation (84th) of the site-specific spectrum along with the original Midland design response spectrum (DBE). It is the Staff's position that the appropriate representation of the _ response spectra as derived directly from the real time histories in the 84th percentile (Sequoyah OL SER, March 1979; Femi OL SER, June 1981). The choice of the 84th percentile is based upon past practice. This level (median plus one standard deviation) is also the level used in deriving the Regulatory Guide 1.60 spectral shape, and is also the level of acceptability being expressed by the Staff in revisions of the Standard Review Plan dealing with the use of site-specific spectra. It is the Applicants' position (March 2,1981 letter from G. S. Keeley to H. R. Denton) that the 84th percentile level of the , site-specific response spectrum, as shown in Figure 1, is appropriately conservative for the Midland site. Compared to the original Midland design response spectrum (Modified Housner), the 84th percentile level of en a 4
lm the site-specific spectrum ranges from about 18 percent to 104 percent higher (in the- frequency range of 25 Hz to 5 Hz). The 84th percentile of Figure 1 is roughly equivalent to a Regulatory Guide 1.60 spectrum anchored at 0.12g (about Intensity of MMI-VII using Trifunac and Brady, 1975) in the frequency range of about 2 Hz to 25 Hz. SITE-SPECIFIC SPECTRA SENSITIVITY TESTS: After its initial review of Part 1, the Staff and Applicant met (April 16, 1981) to discuss the input parameters used in the development of the site-specific spectra (magnitude, distance, site conditions) and how sensitive the response spectra result was to these parameters (how the response spectrum changed as each input parameter was changed). The Staff had independently reviewed the available earthquake strong motion data base (Crouse, g al.,1980) and expressed concern that additional s acceleration time histories may need to be included in the data set - (acceleration time histories which also satisfied the input parameter wnich had been selected, magnitude of 5.3 .5, distance less than about 25 kilometers at soil sites), along with sensitivity tests on each input parameter. As an example, the Staff requested that the original data set used be tested for dependence on average distance. The Applicant's consultant took the original data and systematically restricted the distance of chosen strong motion records to 25, 20 and 15 kilometers and less. Results of this test showed that the 84th percentile level remained stable when excluding the acceleration time histories recorded - at distances between 25 and 33 kilometers. As the distance was further restricted (excluding time histories between 20- and 33-kilometer distance) the 84th percentile level, as expected, systematically
s _, increased, consistent with the observation that the earthquake source to recording distance is decreasing. Results of the all sensitivity tests were docketed in June 1981 as an addendum to Part 1 of the Site Specific Response Spectra. In reviewing this additional information the Staff concluded that, in general, the data set was not very sensitive to,small variations in input parameters and showed expected results when subjected to systematic parameter variations. However, the results to one of the Staff's requests was significant. This involved an additional set of strong motion records from the June 28, 1966, Parkfield, California, earthquake. In establishing a site specific spectrum for an Anna type event for Midland (which was the initial premise of the work on site specific spectrum) the Parkfield earthquake fit the magnitude, distance and site conditions for Midland. Figure 2 shows the median and the median plus one standard deviation of the site-specific spectra with and without the Parkfield records included in the data set along with the original design response spectrum. The Applicant and their consultant have taken the position that the Parkfield records are inappropriate for inclusion in the Midland data set for an Anna type event at Midland. They base their opinion on the following arguments.
- 1. They believe that an accelerogram showing characteristics related to capable faulting or the effects of proximity to a tectonic structure should not be used in the development of -
site-specific response spectra within the context of a tectonTc province approach. S
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- 2. They point out that the recordings are "near field" (close proximity to the energy release of the earthquake) which area
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conditions which would not occur at Midland.
- 3. They also state that the Parkfield event ruptured the surface (not a known characteristic of Central U.S. earthquakes) and that the strong motion recordings were influenced by the rupture.
4 Additionally they state that this earthquake was anomalous in the rupture process as follows; the rupture was possibly incoherent (not smooth but occurring in spurts possibly jumping over barriers), rupture propagation traveling faster than shear velocity of the medium causing higher amplitude accelerations and that the size of this event may be larger than the target magnitude of 5.3 : .5 for Midland. ( In a letter dated August 11, 1981, the Applicant stated that the evidence of very 1cw seismicity of the region (Midland site area) forms another basis for not including 1966 Parkfield earthquake accelerograms for use in the Midland site-specific spectra. They requested that the Staff fonnally review a report entitled "Part III, Seismic Hazard Analysis for the Midland Site," wtich they believe supported this conclusion. Our review on each of the above items is discussed below. SITE SPECIFIC SPECTRUM, PARXFIELD ACCELERATION TIME HISTORIES: The Staff disagrees with the Applicant on the relevance of the -
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Parkfield records, in developing a site specific spectrum for a Anna type event at Midland, for the following reasons:
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- 1. A voluminous amount of literature has been published concerning the Parkfield event. This literature points out the wide variation of expert opinion concerning the various anomalies the Applicant states surround the Parkfield acceleration tine histories. Experts do not uniformly agree on several of the Applictnts' assertions such as:
- a. What is the rupture length of the earthquake? A recent article by Lindh and Score (1981) states that the initial rupture may not have extended as far as other investigators have suggested. This raises questions on how close the recordings were to the epicenter or fault (to help determine if they were in the near field).
- b. Did the earthquake rupture -the surface? Archuleta and Day (1980) states that they see no compelling evidence of breaking the free-surface depth during the 1966 Parkfield earthquake.
In modeling the strong motion records they did not assume that the rupture broke the surface. Synthetic ground motion records were produced that were in overall agreement with the data.
- c. Is the Parkfield earthquake outside target magnitude of 5.3 2 0.5? The work of Kanamori and Jennings (1978) was used by the Applicant to argue that the size of this event exceeded the Staff's target magnitude of 5.3 2 0.5. However, the authors point out that the Richter attenuation rate, a critical .
assumption in determining magnitude, may be inappropriate for the Parkfield recordings, and that using the standard distance-amplitude relations may result in higher readings for M% O
magnitude at close-in stations (Magnitudes, averaging from 5.9 to 6.2) compared to the magnitudes determined by seismometers (Magnitude of 5.6) throughout California. Therefore, we find that Parkfield would properly fall within a magnitude range of 5.3 2 0.5.
- 2. The Parkfield earthquake is similar to other earthquakes (such as the Imperial Valley October 15, 1979 event) in terms of determining source characteristics (such as how coherent was the -
earthquake rupture, what was the stress drop, etc.). The Staff, in fact, used judgment in the initial discussions with the Applicant that station #2 (another record which could fit the Midland input parameter crite*ia) from the Parkfield event, should not be included because it may have been within 80 meters of the event rupture. (
'.. This judgement is supported by the following results numerical modeling of earthquake ground motion completed by Del Mar Technical Associates (1980), on the October 15, 1979 earthquake, has shown that fault rupturing to the surface only affected Station 6 (at about 1 kilometer distance), while Station 8 (at about 4 kilometers distance) was unaffected by the surface rupture. In this example they found that the shallow rupture is severly attenuated as a function of distance from the fault due to the increased amount of material attenuation in the shallow soil layers.
In developing site-specific spectra by using the strong motion 3. 1 data base one attempts to match as best as possible the potential ) conditions which could arise if an earthquake occurred. This involves accounting for uncertainty in the size of the controlling M L -_ _
s
... .. -. .= _ _ _ _ . _
earthquake, the source properties of the event (other than its magnitude) the distance between the event and the site, and any site conditions which may influence the ground motion. In particular, for the Central United States (CUS), the . largest uncertainty involves the earthquake source (stress drop, fault rupture length, fault displacement, rupture velocity). It is the Sta'ff's position that given this uncertainty in the possible earthquake source in the CUS that the source characteristics of the Parkfield event should be considered as being included within the range of this uncertainty. LOW SEISMICITY NEAR THE MIDLAND SITE TO RESOLVE SEISM 0 TECTONIC PROVINCE ISSUE: The Applicant also performed analysis of seismic hazard for the Midland Region (Part III). This work shows, annual exceedance e probabilities at different intensity levels; that is, the annual k' probability that a given intensity earthquake will occur at the Midland site. The key elements of the Applicants methodology to estimate the seismic hazard involves the selection of the earthquake occurrence model (including seismic source zones, rate of activity in each zone, and largest earthquake in each zone) and ground motion model (relating size of eachquake to ground shaking). An important part of completing a seismic hazard analysis involves the selection of an approach to incorporate the uncertainty of all input parameters into the analysis. Difficulty in accounting for this uncertainty is one of the reasons the Staff has used probability studies in only a very limited sense. The Staff prefers to compare the relative probability between a number of sites, rather than the absolute probability at a specific site. The D
- - - . .. . . - . . ~ . _ _ . . _ _ . ._ . _
. reason for this is that sensitivity tests indicate stability when estimating relative hazard probabilities (see Sequoyah SER, March 1979).
The Applicant has, in the Part III report, presented the seismic 8 hazard analysis for three alternate seismic source models. The first l model is based upon the results of Nuttli and Brill (1981) which associate seismic activity with Arches and Basins in the Central United States. The two source zones closest to the Midland site using Model 1 (shown in Figure 3) are the Michigan Basin and the Cincinnati Findlay-Kankakee Arch. Model 2 (shown in Figure 4) separates out the , Anna, Ohio and Attica-Niagara, New York areas from the Central Stable Region based upon historic earthquakes. Model 3 (shown in Figure 5) , treats the Central Stable Region as one unit, not taking into account the a location of historic earthquake activity. These three models represent f ( ., differing interpretations of where earthquakes can occur, and can be thought of as possible seismotectonic province models taking into account both the seismologic and geologic history of the Central United States. The results of the Staff's initial review of the Part III report, centered upon how these results could be used to choose one response spectrum (without Parkfield) over another (with Parkfield), for the Midland site. In particular the Applicant has presented the Staff with ! the results of additional work r6 quested by the Staff on Part III. One of the Staff's requests involved computing the seismic hazard (with the same input as used in Part III) for 5 additional sites in the Central . United States. These 5 sites were chosen by the Staff to represent (and encompass) the range of activity levels expected in the Central United States. The five sites chosen were Western New York, Northeastern Ohio, J h% ' i W
., .m. _ . . _ ,,
. . . ~ . . , . , , _ _ . . . . ._ .._m , _ , , , . , , , -
7: - . Northwestern Ohio, Northern Illinois-Indiana Border and Southeastern Wisconsin (sites are also shown on Figures 3-5). The results of the seismic hazard analysis for the 5 sites were given to the Staff on September IS,1981. These results were compatible with those given in the part III report; that is, the annual exceedance probabilities were tabulated at different intensity levels. In examining this information we find that, in general, the Midland site has lower expected intensities than the 5 other sites at all exceedance probabilities. This means that the Midland site is associated with lower seismic hazard than other parts of the Central Stable Region. The specific results are also model dependent (depending on how close each site is to each source zone). Table 1 shows the expected intensity for the 5 sites and Midland for all 3 models at 2 exceedance ( probability levels (10'3 and 10-4). - The next step is to choose which is the best model or assigning weights to each model. One choice is to weigh each model equally, stating that each is as defensible as the other. If this is done we find that Midland is on the average about 0.5 intensity units lower than the 5 typical CUS sites at the same level of seismic hazard. The applicants consultant puts more weight on Model 1 (50 percent) compared to Model 2 l (30 percent) and Model 3 (20 percent). Using these results, we find that Midland is on the average about 0.57 intensity units lower than the 5 typical CUS sites at the same level of seismic hazard. The Staff also . attempted to determine a preference between the 3 models. We consider the following: i
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- s.
- P-' ,
- 1. As part of the Systematic Evaluation Program a group of seismologists were asked to given their expert c Sinfon on input parameters for a seismic hazard analysis (TERA,1980). One of these I
parameters were the configuration of seismic source zones. Results compiled by TERA demonstrate that while there are varying opinions among experts on earthquake zonation in this region of the CUS, that nost expert choose a model similar to Model 2 in the Part II report (7 out of 10 experts seismic source model was similar to model 2).
- 2. The Applicant has also completed a study of historic intensity occurrences at Midland and the 5 sites. More properly this may be call 6d a psuedohistoric analysis because actual intensities have not been " recorded" at each site. Site intensities are estimated from empirical data relating site intensities to epicentral distance 'and
-( '
epicentral intensity from the actual earthquakes. The Staff has attempted to determine how each of the 3 models fit the pseudohistoric observed data. The expected number of earthquakes based upon Model 3 do not fit the pseudohistoric observed data very well and we have accordingly given it very little weight. This is not surprizing because model 3 (Central Stable Region as one unit) is based upon the surface geology not taking into account historical earthquake activity. Models 1 and 2 (based upon possible different seismotectonic provinces) both are more appropriate models and generally fit the pseudohistoric observed data, with Model 2 being slightly better than Model 1. The information cited above shows that Model 2 alone may give a good indication of the relative differences in seismic hazard for the CUS. l to
^^
_. . . . _ . . . ._::_J: _ _ . . . . - . i _. .
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i Using model 2, the Midland site is on the average about 0.70 intensity units lower than the 5 typical CUS sites at the same level of seismic hazard. Using the above information (weighing the 3 models equally, using the Applicants weight, and using model 2 alone) we find that Midland is about 0.50 to 0.70 intensity units lower than the 5 typical CUS sites at the same level of seismic hazard. Using a relationship between magnitude and intensity (Nuttli and Herrmann,1978), we find that Midland is about 0.25 to 0.35 magnitude units lower than the 5 typical CUS sites at the same level of seismic hazard. The magnitude difference of 0.25 to 0.35 is a key point. This result indicates to us that the classical tectonic province approach,
, defining the Central Stable Region from the surface geology is
\- inappropriate when attempting to achieve consistent descriptions for the SSE and vibratory ground motion for sites in the Central United States.
Since surface geologic conditions across this large province are relatively uniform, the Staff has concluded that the level and character of seismicity in given regions (areas) must assume a primary role in determining the SSE. The Staff concludes that, for the purposes of seismic design, the Midland site is in a different seismotectonic province (one requiring the use of magnitude 5.0 to define the controlling earthquake), compared to other areas within the Central Stable Region (those requiring the use of an Anna-type earthquake of magnitude 5.3 to define the controlling earthquake). A cagnitude of 5.0 corresponds to the largest historical earthquake which should be l associated with the seismotectonic province in which the Midland site Ob M
resides (such as the 1860 Minnesota event, MMI=VI-VII, Mbig=5.0, interpreted, Nuttli and Brill 1981, and the July 9,1975 Minnesota event, MMI=VI, Mbig=4.3, instrumental and interpreted, Nuttli and Brill 1981). Since the Magnitude of 5.3 was used for the davelopment of the SSRS, however, the Staff has attempted to quantify the average difference between Midland and the 5 typical sites as follows for the purpose of defining a spectrum for a 5.0 event. SAFE SHUTOCWN EARTHQUAKE RESPONSE SPECTRUM: The Staff has utilized four relationships which correlate peak ground motion values with magnitude or intensity. These four are Murphy and O'Brien (1978), Trifunac and Brady (1975), Herrman (1981) and ' Campbell (1981). Using these relationships we have datermined peak ground motion ratios assuning intensity differences of 0.50 to 0.70, and ( magnitude differences of 0.25 to 0.35. Based on these results, the l ground motion at the Midland site is about 30 to 70 percent lower (average if about 46 percent), than greund motion at the 5 typical CUS sites at the same level of seismic hazard. , Table 2 displays the ratio of the site specific spectrum with the Parkfield records to the site specific spectrum without the Parkfield records. (The latter is the one proposed by the Applicant and ultimatd accepted by tne Staff.) For frequencies between about 2.5 to 20.0 Hz the - difference is 10 to 30 percent, while for frequencies between 1.0 and 2.5 Hz the difference is between 30 and 45 percent. In reviewing Figure 2, - j however, both site specific spectra (84th percentile) are less than the original Midland design spectrum for frequencies below about 2 to 2.5 Hz. , Thus the differences between the two (with Parkfield versus without m a. e e --- ,-g - , -,v -n- w--,,--y- - - - - - - - ,-,,m---,- n, - - - , - - - -------.,,,n~,ng--- --- e
! Parkfield) site specific spectra is thus 10 to 30 percent where the site specific spectras exceed the design spectrum.
- 1. The results of the seismic hazard analysis have been used to quantify the differences in the response spectra between Midland and .
five other sites at the same level of seismic hazard. .
- 2. The Staff finds that the difference in ground motion (at the 1 same level of seismic hazard) at Midland is 30 to 70 percent (46 percent average) lower than the ground motion at the five typical CUS sites. This represents the approximate difference expected if a 4
magnitude of 5.0 had been used in the development of the site specific spectrra.
- 3. The Applicants proposed response spectrum is defined by the 84th percentile of figure 1 the site specific response spectrum
\ without the inclusion of the Parkfield records. This response spectrum is about 10 to 30 percent lower than the site specific spectrum with the inclusion of the Parkfield records. The seismic hazard results have demonstrated that if a magnitude of 5.0 had been
! used for developing a site specific spectrum for Midland, the resulting spectrum would have been about 30 to 70 percent (as discussed in #2 above) lower than the site specific spectrum with the inclusion of the Parkfield records. This demonstrates that the Applicants proposed site specific response spectrum is conservative compared to that expected from a magnitude 5.0 earthquake.
~~
I
- 4. The Applicants proposed response spectrum (84th percentile in
, figure 1) is roughly equivalent to a Regulatory Guide 1.50 response spectrum anchored at 0.129 (about intensity of MMI=VII using trends . i i d
;m of means in Trifunac and Brady,1975). No historic earthquake within about 200 miles (325,000 square kilometer area) of the site
~
have been either equal to or larger than intensity Ri!=VII. I Using the above information it is the Staff's position that the 84th percentile of the site specific spectrum as shown in figure 1 is appropriately conservative to represent the free field Safe Shutdown Earthquake response spectrum for the original ground surface at the Midland site, because it exceeds that expected from a magnitude 5.0 site specific response spectrum. RESPONSE SPECTRUM FOR THE TOP OF THE PLAriT FILL: Another topic which was identified in the October 14, 1980 letter to the Applicant required an assessment of soil amplification through the plant fill. In making this assessment the applicant has chosen two (< methods to arrive at a response spectrum at the top of the fill material. The first method involves developing a second site specific spectra matching the fill soil profile (ah additional 30 feet of low velocity material on top of the original ground surface). The second method
~
involves utilizing a one-dimensional wave propagation program (SHAKE) to assess the possible amplification at different frequencies. This testimony will discuss the first method chosen while the testimony of Dr. Paul Hadula will discuss the second method chosen. The results of the two methods provide an independent check of the other method. Both methods indicate that results (discussed below) are consistent.
~
The Applicants consultant presented a report entitled "Part II. Response Spectra Applicable For The Top of Fill Material". This study involved developing a second site specific spectra, again searching the-A e
/
L__
/
earthquake strong motion data base for accelerograms which fit the originally assumed magnitude (5.3 + .5), distance (less than about 25 kilometers), and site conditions (soil profile with about 300 feet of till material overlain by 70 to 80 feet of softer material) for Midland. Qualitatively the difference in the two site specife spectra (for original ground surface versus top of plant fill) involves matching the same controlling earthquake at two soil profiles, one of which has a thicker soft layer. As shown in Table 1 of the part 2 report the range of magnitudes and distances of the acceleration records chosen were magnitude of 4.9 to 5.6, distance of 6 to 31 kilometers. It is noted that 5 recording sites were used in both site specific spectra, reflecting the flexibility in matching the site conditions from strong motion recording sites to those at Midland. In addition, 8 sites which ( were used for the original ground surface were not used for the top of the fill, while 4 sites were added which were not used for the original ground surface. This is appropriate because of the different soil conditions at the top of the fill versus the top of the till. Results from part II are shown in Figure 6 which displays the median (50th) and median plus one standard deviation (84th) of the site specific spectra for the top of the fill material along with the Midland Modified Housner design response spectrum. Figure 7 displays the ratio of the two different site specific respcase spectrums at different frequencies. As one can see from these figures, the results of the site specific spectra . . demonstrates that amplification of' ground motion occurs for frequencies less than about 4 Hz. .This is qualitatively consistent with the fact w O
~ " that the fill spectra is matching a deeper soft layer, which tends to amplify the lower frequency motion (Seed, et al.1976). As with the Part I report the Staff independently reviewed the I strong motion data base to insure that the appropriate records were used in this analysis. We have also used the sensitivity results from Part I to get an idea of the effects that parameter variation may have on the data set. Independently, the Applicants consultants studied the potential for amplification using a one-dimensional wave propagation program (SHAKE). The Staff's consultant, Dr. Paul Hadala, of the Waterways Experiment Station, both reviewed the Applicant SHAKE analysis and conducted his own SHAKE analyses. As discussed by Dr. Hadala, these results provide additional
, information tnat the amplification as identified in Figure 7 using the
( site specific spectra is conservative. Using the above information it is the Staff's position that the amplification through the plant fill as identified in Figure 7 is appropriately conservative for the Midland site. The 84th percentile of the site specific spectra in Figure 6 represents the Safe Shutdown Earthquake response spectrum for the top of the fill material at the Midland site. CONCLUSION The 84th percentile of response spectra in Figures 1 and 6 appropriately characterize the Safe Shutdown Earthquake vibratory ground
~
motion or the Midland site. s MO C
- r. -
Attachment 1 JEFFREY K. KIMBALL' - - GEOSCIENCES BRANCH, P-314 . DIVISION OF ENGINEERING ~ U. S. NUCLEAR REAGULAT0rY CCMMISSION WASHINGTON, D.C. 20555 . I I am employed as a Seismologist / Geophysicist i Ny.nameis Jeffrey K. Kimball.revie.ter, Geosciences Branch, Divisien of E Reactor Regulation. . I received a S.S. degree in Oceanography from the. University of Michigan in 1977 and a M.S. degree in Geology frem the University of Michigan in 1979, with a specialty in seismology and geophysics. I have been empTcyed by NRC since May 1930 as a Seismologist / Geophysicist reviewer as applied to the evaluation of applications for construction h and cperation of nuclear facilities, and to determine the thcrou5 ness-of this information for defining the seismic hazard for which facilities must be designed. Since joining the Nuclear Regulatory Commission staff,'I have
' participated in the licensing activity for approximately ten sites. ,
Frem 1977 to 1950, I was a research assistant and teaching assistant at the Uaiversity of Michigan. My activity as a research assistant included seismic data ecmpilation studies for the U. S. Geological Survey and data analysis and cperati.cn of a nine station seismic network. My M.S. thesis
' work involved a study en surface wave dispersien of the Atlantic Ocean Basins and has been presented at national meetings of professional societies and Teaching assistant experience censisted J
published in a professiona'i journal.of helping teach both introductory Wyoming for two summers and an introductory geology laboratcry class at the
- University of Michigan. ,
I I am a member of the American Geophysical Union and the Seismological Scciety of America, and have co-authored 7 publications including abstracts of pre- - sentations to professional societies and NUREG documents. S S et e '
- . i
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'N.W. N.E. N.W. Ind-Ill SE Midland New York Ohio Ohio Border Wisconsin Model #1
-3 5.85 6.7 6.3 6.9 6.65 6.45 10
-4 6.90 7.7 7.3 7.95 7.75 7.40 10 Model d2 10
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-3 6.4 6.4 6.4 6.4 6.6 6.4
~4 7.5 7.5 7.6 7.5 10 7.5 7.5 TABLE (
Ratio of with Parkfield/Without Parkfield-Period (sec) .05 .08 .10 .20 .30 .40 .50 .80 1.0 Frequency (hz) 20.0 12.5 10.5 5.0 3.3 2.5 2.0 1.25 1.0 - ratio 1.14 1.11 1.11 1.23 1.28 1.32 1.39 1.35 1..; 5 i .. M
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_. __ _ _ . . .. . . . _ . . . _ . _ _ _ .. ~ _ _ _ . . . .
., ~: . * ,
i ' s . KIMBALL REFERENCES ; Archuleta, R.J., and S.M. Day,1980 Dynamic Rupture In A Layered Medium: The 1966 Parkfield Earthquake, Seismol. Soc. Amer. Bull, V 70, p. 671. Barstow, N.L., et al.,1981, An Approach To Seismic Zonation For Siting . Nuclear Electrid Fower Generating Facilities in the Eastern United States, Nuclear Regulatory Commission, NUREG/CR-1577. Campbell, K.W.,1981, A Ground Motion Model For The Central United States Based on Near-Source Acceler. tion Data,' Proceedings of the Conference on Earthquakes and Earthquake Engineering, Knoxville, Tennessee. Chung, D.H., and D.C. Bernreuter,1981, The Effect of Regional Variation of Seismic Wave Attentuation on the Strong Ground Motion from Earthquakes, U.S. Nuclear Regulatory Commission, NUREG/CR-1655. Coffman, J.L., and C. A. Von Hake,1973, Earthquake History of the United States, NOAA - U.S. Dept. of Commerce Publication 41-1. (' Crouse, C.B., et al.,1980, Compilation, Assessment and Expansion of the
. Strong EarthquaYe Ground Motion Data Base, U.S. Nuclear Regulatory Commission, NUREG/CR-1660.
Del Mar Technical Associates,1980, Simulation of Earthquake Ground Motion for San Onofre Nuclear Generating Station, Unit 1, Supplement No. III. Eardley, A.,1962, Structural Geology of North Anerica, Harper and Row, 743 pp. . Gupta, I.W. , and'0.W. Nutt11,1976, Spatial Attenuation of Intensities for Central U.S. Earthquakes, Seismol. Soc. Amer. Bull., 63, pp. 227-248. D s O m.* O e
_ _ _ . . . . . _ - . . . . - - - = = - - - - 5 .. Hays, W.W.,1980, Procedures for Estimating Earthquake Ground Motions, United States Geological Survey Professional Paper 1114. Herman, R.B.,1981, Personnel Communication. Kanamori, H., and P.C. Jennings,1978, Detemination of Local Mangitude, fic, Frcm Strong Motion Accelerograms, Seismol. Soc. Amer. Bull., V 68 p. 471. King, P.B.,1964, " Tectonic Map of North America," U.S. Geological Survey, , Dept. of the Interrior Publication. King, P.B.,1969, The Tectonics of North Merica - A Discussion to Accompany the Tectonic flap of North America, Scale 1:5,000,000 U.S. Geol. Survey Prof. Paper 628, Washington, D.C. , Lindh, A.G., and D.M. Boore,1981, Control of Rupture by Fault Geometry During the 1966 Parkfield Earthquake, Seismol. Soc. Amer. Bull., V 71,
- p. 95.
flurphy, J.R., and L. J. O'Brien,1978, Analysis of a Worldwide Strong Motion Data Sample to Develop an Improved Correlation Between Peak Acceleration, Seismic Intensity and Other Physical Parameters, U.S. Nuclear Regulatory Commission, NUREG-0402. (_- Nuttli, 0.W., and R.B. Hermann,1978, State-of-the-Art for Assessing
. Earthquake Hazards in the United States: Credible Earthquakes for the Central United States, Misc. paper S-73-1, Report No.12 U.S. Amy Waterways Experiment Station, Vicksburg, Miss., 100 p. .
Nuttli, 0.W., and K.G. Brill,1981, An Approach to Seismic Zonation for Siting Nuclear Electric Power Generating Facilities in the Eastern U.S., Part 2 U.S. Nuclear Regulatory Commission, NUREG/CR-1577. Nuttli, 0.W.,1981, Similarities and Differences Between Western and Eastern United States Earthquakes, and Their Consequences for Earthquake Engineering, Proceedings of the Conference on Earthquakes and Earthquake Engineering, Knoxville, Tennessee.
~
Seed, H.B., et al.,1976, Site-Dependent Spectra for Earthquake-Resistant Design, Seismol. Soc. Mer. Bull., V 66 p. 221. TERA Corporation,1980, Seismic Hazard Analysis, Solicitation of Expert - Opinion, NUREG/CR-1582, Vol. 3, 115 pp. Trifunac, M.D., and A.G. Brady,1975, On the Correlation of Seismic Intensity Scales with Peaks of Recorded Strong Ground Motion, Seismol. l Soc. Amer. , Bull . , v. 65. i b.
"O
~
.g i,.,'
f- , , U.S. Atomic Energy Commission,1970, Construction Permit Safety Evaluation for the Midland Nuclear Plant 1 and 2, Docket No. 50-330/331. U.S. Atomic Energy Commission, 1973, Design response spectra for seismic design of nuclear power plants, Regulatory Guide 1.60. i U.S. Nuclear Regulatory Commission,1975, Standard Review Plan, NUREG-CR 75/087. U.S. Nuclear Regulatory Commission,1975, Seismic and geologic siting ' criteria for nuclear power plants 10 C.F.R. Part 100, Appendix A. U.S. Nuclear Regulatory Commission,1979, Safety evaluation of the Sequoyah Nuclear Plant Units 1 and 2 Docket Nos.- 50-327 and 50-328. U.S. Nuclear Regulatory Commission,1981,' Safety Evaluation of the Enrico Fermi Atomic Plant No. 2. Docket No. 50-341. Weston Geophysical Corp., " Site Specific Response Spectra, Midland Plant - Units 1 and 2, Part I, Response Spectra - Safe Shutdown Earthquake, Original Ground Surface," Feb. 1981.. Weston Geophysical Corp., " Site Specific Response Spectra, Midland Plant - Units 1 and 2 Addendum to Part I, Response Spectra - Original Ground (-, Surface," June 1981. Weston Geophysical Corp., " Site Specific Response Spectra, iiidland Plant - Units 1 and 2, Part II, Response Spectra Applicable for the Top to Fill s Haterial at the Plant Site," April 1981. Weston Geophysical Corp., " Site Specific Response Spectra, Midland Plant - Units 1 and 2, Part III, Seismic Hazard Analysis" Feb.1981. 1 l 1
~
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. '/.I UNITED STATES NUCLEAR REGULATORY CoMMISSloN ,
W ASHINGTON, D. C. 20555 RE:E,v0 January 28, 1977 peg 7 FEE : R.1 10 25 ff/[f
./cr MV t Myer Bender, Chairman
.gh Advisory Committee on Reactor %S gg i'y.l j Safeguards r.EACICR St.?tGUARDS U. S. Nuclear Regulatory Commission Washington, DC 20555 RE: MIDLAND PLANT UNITS 1 AND 2
Dear Mr. Bender:
The Board has reviewed the reports in evidence in this case by the Advisory Committee on Reactor Safeguards (ACRS) (Staff Exhibits 1, 2 and 3) and has decided to return those
. responses to the ACRS for further elaboration. These responses were originally submitted as a result of the decision in Aeschliman vs. NRC F.2d , (DC Cir. 1976),
slip opinion at 21. The Board has received two responses, both dated November 18, 1976, one including a copy of some minutes of an ACRS meeting discussing Midland and the other i having no such enclosure. e have three areas of comment. I. The minutes mentioned contain references which we believe require further comment under the rules set forth in the Aeschliman~ case. Two of these are:1/
"c. Exclusion area and low population zone - the exclusion area extends 1100 meters from the proposed plant and includes a portion of the Dow plant, including 53 Dow employees; the low population zone extends to three miles and includes all of the Dow plant and part of the City of Midland. The site received a-34 index rating when compared to the hpothetical the maximumreference population site in the(considering Dow complex ).
1/ Others may exist. We presently focus on these because
, , , _ , of their relationship to current suspension hearings. j. p; l- ug-Fo / A-35-6o2- B, 33
~ ~ '
( (
~
Myer Bender January 28, 1977 "g. Other asoects ... the Committee mentioned but did not explore in any depth: the suit-ability of B&W reactors for marginal sites, t protection required against reactor vessel splits, cavity flooding systems, and the use i of process steam in products to be consumed by people." Neither the ACRS letter dated June 18, 1970, nor the one dated November 18, 1976, furnished to meet the require-ments of Aeschliman, mention these matters. We believe that the court, in the words that are set out in footnote 2 below requires that these matters, as well as any other
" matters of concern" (including any matters mentioned in furnished or unfurnished minutes) be treated fully by the i .
Committee. The significance of the rating system referred to in item
- (c) and the hypothetical reference site is not apparent nor are there explanatory references cited. Furthermore, i the Board does not understand what the ACRS means by "the i suitability of the B&W reactors for marginal sites" in item "g."
II. 4 We are concerned with the adequacy of some responses in i the November 18, 1976, letter to meet the Aeschliman test. To illustrate we set out the first of the eleven topics in the letter:
"1. Separation of protection and control instru-mentation - The Applicant proposed using l signals from protection instr *ments for con-trol purposes. The Committee believed that control and protection instrumentation should be separated to the fullest extent practicable, i aJ.d recommended that the Applicant explore further the possibility of making safety instrumentation more nearly independent of control functions. (Three Mlle Island,1/17/68) .
. 2/ "At a minimum, the ACRS report should have provided a short explanation understandable to the laymen of the additional matters of concern to the Committee and a cross-reference to previous reports in which those problems and the measures proposed to solve them were developed in more detail."
i i 4 -s, + - - - + -
( (
'M'er y Bender January 28, 1977 It is unclear to the Board what this paragraph means. The danger is not specified and it is unclear as to whether the I
" separation" mentioned refers to a physical separation of components or to the necessity for separate energy sources for signals and controls or to some other separation. No standard is set for the Applicant's (now Licensee's) con-formance. The referenced documentation (Three Mile Island, January 17, 1968) says no more. There is in that document a list of references (some marked ACRS Office Copies Only) which may clarify the matter. But no direction is given as to which of these references is relevant to the partic-ular subject.
This illustration is exemplary only and whether the same infirmity exists in other items is a problem we have not -- had the opportunity to address. We furnish this now so that the Committee is made aware of our concern and so that further elaboration is not delayed. III.
.'i The letter of the ACRS to Chairman Rowden, November 18, 1976, referred to other ACRS letters. Those letters con-tain items which have ambiguities similar to those dis-approved in Aeschliman. For example, the March 12, 1970 letter on Hutchinson Island stated:
"Other problems related to large water reactors have been identified by the Regulatory Staff, and the ACRS and cited in previous ACRS Reports" (p. 3).
Those items, we feel, need to be identified if they apply to Midland and if they do, to be described as the Court directed. See footnote 2 hereof. We write this under what we perceive to be out dut under the direction given in the Aeschliman case 3/ywith-out waiting to fully identify all of the possible areas 3/ A "sua sponte" request for elaboration.
- l
.i *t
.~
(
. .- (
Myer Bender January 28, 1977 + of concern relative to the November 18, 1976, letter. We do so because we are in the midst of suspension hearings and will need a resolution of this matter as soon as it i may reasonably be furnished. i R ectfully su i ted, Frederic J. ufal, airman Atomic Safety and Licensing Board
)
I l f f s I l l
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UNITED STATES NUCLE AR RE' ULATORY COMMISSION W ASHINGTON, D. C. 20555 _ , , , m . E . ,., . January 28, 1977 ,
, p p3- ! /.'.110 25 ,.-
h- ' Myer Bender, Chairman Advisory Committee on Reactor {S.,gy./ g , gg $ Safeguards Ct.cIct S.U EGO.*J. S U. S. Nuclear Regulatory Commission Washington, DC 20555 RE: MIDLAND PLANT UNITS 1 AND 2
Dear Mr. Bender:
The Board has reviewed the reports in evidence in this case by the Advisory Committee on Reactor Safeguards (ACRS) (Staff Exhibits 1, 2 and 3) and has decided to return those responses to the ACRS.for further elaboration. These responses were originally submitted as a result of the decision in Aeschliman vs. NRC F.2d. , (DC Cir. 1976), slip opinion at 21. The Board has received two responses, l both dated November 18, 1976, one including a copy of some minutes of an ACRS meeting discussing Midland and the other having no such enclosure. We have three areas of comment. I. The minutes mentioned coniain references which we believe require further comment under the rules set forth in the Aeschliman case. Two of these are:l_/ ,
"c. Exclusion area and low population zone - the exclusion area extends 1100 meters from the proposed plant and includes a portion of the Dow plant, includi'ng 53'Dow employees; the low population zone extends to three miles and includes all of the Dow plant and part of the City of Midland. The site received a-34 index rating when compared to the hypothetical referenceinsite the maximum population the(considering Dow complex ). . 1/
Others may exist. We presently focus on these because of their relationship to current susperision hearings. .. w w ,.e
i
' ~
b_ ( Myer Bender January 28, 1977
~.
l "g. Other aspects - ... the Committee mentioned i' but did not explore in any depth: the suit-ability of B&W reactors for marginal sites, 4 protection required against reactor vessel splits, cavity flooding systems, and the use i of process steam in products to be consumed : by people." Neither the ACRS letter dated June 18, 1970, nor the one
' dated November 18, 1976, furnished to meet the require-ments of Aeschliman, mention these matters. We believe that the court, in the words that are set out. in footnote 2 below requires that these matters, as well as any other
" matters of concern" (including an furnished or unfurnished minutes) y matters mentioned inbe treated full a Committee.
- j
- The significance of the rating syste'm referred to in item
- ., (c) and the hypothetical reference siteis not apparent
< nor are there explanatory references cited. Furthermore, l l the Board does not understand what the ACRS means by "the suitabilit y of the B&W reactors for marginal sites" in - item "g." 1 l 4 i II. We are concerned with the adequacy of some responses in the November 18, 1976, letter to meet the Aeschliman test. t To illustrate we set out the first of the eleven topics 4 in the letter: ! "1. Separation of protection and control instru-mentation - The Applicant proposed using j signals from protection instruments for con- ] trol ^ purposes. The Committee believed that control and protection instrumentation should
- be separated to the fullest extent practicable,
- and recommended that the Applicant explore further the-possibility of making safety instrumentation more nearly independent of
,~
control functions. (Three Mile Island, 1/17/68). ,
. r 2 / "At a minimum, the ACRS report should have provided a
, short explanation understandable to the laymen of the additional matters of concern to the Committee and a i cross-reference to previous reports in which those problems and the measures proposed to solve them were
- developed in more detail."
~
l ( Myer Bender January 28, 1977
~
It is uncicar to the Board what this paragraph means. The danger is not specified and it is unclear as to whether the i
" separation" mentioned refers to a physical separation of components or to the necessity for separate energy sources for signals and controls or to some other separation. No 1
. standard is set for the Applicant's (now Licensee's) con-formance. The referenced documentation (Three Mile Island, January 17, 1968) says no more. There is in that document a list of references (some marked ACRS Office Copies Only) which may clarify the matter. But no direction is given as to which of these references is relevant to the partic-ular subject.
This illustration is exemplary only and whether the same
. infirmity exists in other items is a problem we have not had the opportunity to address. We furnish this now so that the Committee is made dware of our concern and so that further elaboration is not delayed. ' ,,
i III. The letter of the ACRS to Chairman Rowden, November 18, 1976, referred to other ACRS letters.. Those letters con-tain items which have ambiguities similar to those dis-approved in Aeschliman. For example, the March 12,,1970 letter on llutchinson Island stated:
"Other problems related to large water reactors l have been identified by the Regulatory Staff, and the ACRS and cited in previous ACRS Reports" (P. 3).
Those items, we feel, need to be identified if they apply to Midland and if they do, to be described as the Court directed. See footnote 2 hereof. We write this under what we perceive to be out dut under the direction given in the Aeschliman case 3/ywith-out waiting to fully identify all of the possible~ areas .
. t 3/ A "sua sponte" request for elaboration.
.. . o Am _
Myer Bender - 4- January 28, 1977
~.
of concern relative to the November 18, 1976, letter. We do so because we are in the midst of suspension hearings j and will need a resolution.of this matter as soon as it may reasonably be furnished. l R cetfully sub i ted, Frederic J. Coufal, airman Atomic Safety and Licensing Board e S e e b 9 1 1
UNITED STATES Attrehment 1 NUCLEAR REGULATORY COMMISSION - -
- . . W ASHINDTON. D. C. 20555 tie l[ .'.'N January 28, 1977 l
'F" i 1'I IU 25 Myer Bender, Chairman Advisory Committee on Reactor i'J. .*i . . . . .T4 Safeguards ' j $((,.,. . ....U.f3 U. S. Nuclear Replatory Commission Washington, DC 20555 ..
g RE: MIDLAND PLANT UNITS 1 AND 2 .
Dear' Mr. Bender:
The Board has reviewed the reports in evidence in this case by the Advisory Committee on Reactor Safeguards (ACRS) (Staff Exhibits 1, 2 and 3) and has decided to return those responses to the ACRS for further elaboration. These responses were originally submitted as a result of the
, decision in Aeschliman vs. NRC. F.2d , (DC Cir. 1976),
slip opinion at 21. The Board has received two responses, both dated November 18, 1976, one including a copy of some minutes of an ACRS meeting discussing bEdland and the other having no such enclosure. We have three areas of comment. I. The minutes mentioned contain references which we believe require further comment under the rules set.forth in the Aeschliman case. Two of these are:1/
"c . " Exclusion area and lou conulation zone - the ,
exclusion area extends 1100 meters from the - proposed plant and includes a portion of the Dow plant, including 53 Dow employees; the low population zone extends to three miles and includes all of the Dow plant and part - of the City of Midland. The site received
- a-34 index rating when compared to the hypothetical referenceinsite the maximum population the(considering Dow complex ).
1/ Others may exist. We presently focus on these because of t. heir relationship to current suspension hearings. ... pyUToom s
%,, ']: .
- w. . ,,*
P . . . . . ...
) . . . ._
l
. ,.) -
M'er y Bender . January 28, 1977 "g. Other aspects - ... the Committee mentioned but did not explore in any depth: the suit-ability of B&W reactors for marginal sites, protection required against reactor vessel - - splits, cavity flooding systems, and the use i
- of process steam in products to be consumed by people." .,
v. Neither the ACRS letter dated June 18, 1970, nor the one dated November 18, 1976, furnished to meet the require-ments of Aeschliman, mention these matters. We believe, that the court, in the words that are set out in footnote 2 below requires that these matters, as well as any other
" matters of concern" (including an furnished or unfurnished minutes) y matters mentioned inbe treated fully b Committee. ,
The significance of the rating system referred to in item (c) and the hypothetical reference site is not ap,arent nor are there explanatory references cited. Furtaermore, i the Board does not understand what the ACRS means by "the suitability of the B&W' reactors for marginal sites" in item "g." II. We are concerned with the. adequacy of some responses in the November 18, 1976, letter to meet the Aeschlima~n test. To illustrate we set out the first of the eleven topics in the letter:
"1. Separation of protection and control instru-mentation - The Applicant proposed using
. signals from protection instruments for con-trol purposes. The Committee believed that control and protection instrumentation should be separated to the fullest extent practicable, and recommended that the Applicant explore further the possibility of making safety instrumentation more nearly independent of
. control functions. (Three Mile Island, 1/17/68).
~/
2 "At a minimum, the ACRS report should have provided a short explanation understandabic to the laymen of the ; additional matters of concern to the Committee and a cross-reference to previous reports in which those prob 1 cms and the measures proposed to solve them werc developed in more detail."
.)
}
Myer Bender
. January 28, 1977
- It is' unclear to the Board what this paragraph means. The -
danger is not specified and it is unclear as to whether the 4 " separation" mentioned refers to a physical separation of - components or to the necessity for separate energy sources " for signals and controls or to some other separation. No + 1
~
standard is set for the Applicant's (now Licensee's) con-l 9* formance. The referenced docume'ntation (Three Mile Island,
- , January 17, 1968) says no more. 'There is in that document a list of ' references (some markhd ACRS Office Copies Only) which may-clarify the matter. But no direction is given as to which of these references is relevant to-the partic-ular subject.
This illustration is exemplary only and whether the same infirmity exists in other items is a problem we have not
. had the opportunity to address. We furnish this now so that the Committee is made aware of our concern. and so that further elaboration is not delayed.
t III. The letter of the ACRS to Chairman Rowden, November 18, 1976, referred to other ACRS letters. Those letters con-tain items uhich have ambiguities similar to those dis- , approved in Aeschliman. For example, the March 12, 1970 letter on Hutchinson Island stated: i "Other problems related to large water reactors have been identified by the Regulatory Staff, , and the ACRS and cited in previous ACRS Reports" ; (p. 3)- ; Those items, we feel, need to be identified if they apply l to be described as the Court to Midland and if they do, 2 hereof. directed. See footnote 4
- We write this under what we perceive to be out dut !
under the direction given in the Aeschliman casc3/ywith-
- out waiting to fully identify all of the possibic areas I
3/. A "sua sponte" request for elaboration. l t
** .. .... ...,, , .= . j
. . . . _ . . . . , . _ . . _ _ _ . . .~, ,_ _ . . - . . , _ _., . _ _ . ~ _. _ .. _ . _- _ ._.. _ .. _.. _ - .._.__.. _.... _ . .
,_.)
. .uo -
O. Myer Bender , January 28, 1977 of concern relative to the November 18, 1976, letter. We do so because we are in the midst of suspension hearings
, and will need a resolution of this matter as soon as it '
may reasonably be furnished. 0 R ectfully sub i ted, vn . Frederic J. Coufal, 1 airman Atomic Safety and Licensing Bo'ard a t O e S 6 e e e O 6 e 4 e O
- - . . - ~ ~ . - - - - -
4 -
' '[
.- Attachment 2 (2) The Court concluded in the Aeschliman decision that: e i
The ACRS report in this case must be evaluated in light of the congressional purposes. While the reference 4 y to "other problems" identified in previous ACRS reports may have been adequate to give the Commission the bene - fit of ACRS members' technical expertise, it fell short of performing the other equally important task which ! Congress gave ACRS: informing the public of the haz-ards. At a minimum, the ACRS report should have pro-vided a short explanation, understandable to a layman, of the additional matters of concern to the committee, and a cross-reference to the previous reports in which those problems, arid the measures proposed to solve them, were developed in more detail. Otherwise, a concerned
, citizen would be unable to determine, as Congress in-tended, what other difficulties might be lurking in the proposed reactor design. Since the ACRS report on its face did not comply with the requirements of the statute, we believe the Licensing Board should have returned it sta sponte to ACRS for further elaboration of the cryptic reference to "other problems.""
Turning to the propriety of discovery directed to indi-vidual ACRS members and ACRS documents, we ' con-clude it'was not error to deny these requests. ACRS' unique role as an independent "part of the administra-tive procedures in chapter 10 of the act," supra, is suffi- - ciently analogous to that of an administrative decision-maker to bring into play the rule that the " mental proc-esses" of such a " collaborative instrumentalit[y] of jus-tice" are not ordinarily subject to probing. United States
- v. Morgan, 313 U.S. 400, 422 (19411. This rule is par-i .
ticularly apropos in light of ACRS's collegial composition
, such that no individual may speak for the group as a whole. Where an ACRS report on its face omits material i l l
"This is not to say that an ACRS report must cohtain de.
talled facton! (mdings of the kind necessaiy to aid judicial review. Under Commission rules, when ACRS conclusions are controverted. a factual record is compiled anew before the Lleensing Board. Sec 10 C.F.R., pt. 2, App. A, V(f)(1) (19'l0). -
I - t information, the appropriate course is not discovery but : to return it for supplementation. Cf. Dunlop v. Bachore. 3 ski, 421 U.S. 5G0, 574-75 & n.11 (1975i. We merely - - J hold here that neither the Atomic Energy Act nor general vf, principles of administrative law required the Commission
. to grant Saginaw's discovery requests."
On remand, the ACRS report should be returned to the ACRS for clarification of the ambiguities noted above. l "The case as presented calls upon the court to make no
- decision ivhether the Federal Advisory Committee Act, 5 U.S.C. App. I 110(b) (Supp. III,1973), entitics a party upon proper request to have access to data which were before the ACRS.
( a 6 I . f e b
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1 p 4 UNITED STATES NUCLEAR REGULATORY COMMISSION T . p
- WASHINGTON, D. C. 205$5 g ,g
- a 1
't. March 16, 1977
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Honorabl'e Marcus A. Rowden Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555 D
Subject:
ADDITIONAL REQUEST FOR INEORMATION FROM THE MIDIAND AS&LB
Dear Mr. Rowden:
The Comittee has received an additional request from the Atomic Safety and Licensing Board in the Midland case for further elaboration and
" treatment" of matters mentioned in the Comittee's Supplemental Report to you of November 18, 1976 and attachments thereto. That report, you may recall, was written in response to a previous request which followed directly from the decision in Aeschliman vs. NRC. A copy of the most recent AS&LB request, dated January 28, 1977, is attached.
Although the Comittee is willing to provide reasonable and necessary ( clarification of its recomendations and opinions, we believe that the Board in this case has misinterpreted the Aeschliman decision and has embarked on a course which, if pursued, could involve the Comittee in an unnecessary and potentially unending series of requests for clari-fication and elaboration of its reports, in connection with not only the Midland proceeding, but other proceedings as well. 'Ihe Board's "three areas of coment" are addressed below: I. The Board notes two specific paragraphs of interest to the Midland pro-ceeding in a set of ACRS meeting minutes (106th ACRS meeting held February 6-8, 1969) during which the Midland project was discussed, and the Board requests "further coment under the rules set forth in the Aeschliman case" regarding these two paragraphs "as well as any other
' matters of concern' (including any matters mentioned in furnished or unfurnished minutes)" and requests that these matters be treated fully by the comittee in accordance with the following excerpt from Aeschliman vs. NRC:
U *At a minimum, the ACRS report should have provided a short explanation, understandable to a layman, of the additional matters of concern to the Comittee, and a cross-reference to the previous reports in which those problems, and the measures proposed to solve them, were developed in more detail." C NTheHMwr } .-
Honorable Marcus A. Rowden March 16, 1977 l In the opinion of the Comittee, the Board has incorrectly concluded that . all topics discussed during an ACRS review, and recorded in the meeting I minutes, are " matters of concern" to the Comittee in the context of the Aeschliman decision. " Items of concern" to the ACRS at the completion of its review are identified in the Comittee's report and have been explained I
- in the Comittee's Supplemental Report of November 18, 1976, in language
* " understandable to the layman" as required by the Aeschliman decision.
l Many other item of interest are documented and discussed during the course of an ACRS review and are not identified as matters of concern in the ACRS report. Some of these items are considered satisfactory or are adequately resolved by amendment of the application or other means during the review process. Some represent points of general information, some represent matters that the Comittee explores on a generic basis. It should be noted that the Aeschliman decision did not address the con-tent of ACRS meeting minutes or other information available to or con-sidered by the Comittee but was limited (see Attachment 2) to those matters identified in ACRS reports as items of concern. To require that the ACRS address in its report every item discussed or considered during the course of a review is impractical and unnecessary. For exarple, the suitability of the Midland Plant for the proposed Midland site was discussed at length during six Subcomittee meetings held on January 22 and February 4,1969, and March 24, April 24, June 10, and Septenber 14, 1970, and at five full Comittee meetings held on February 6, 1969, and April 9, May 8, June 11-13, and September 17-19, 1970; appropriate safety features were included in the design for this reactor at this site. The minutes of these meetings have been in the public domain since 1974. II. , his section of the Board's request deals with the substance of the Comittee's Supplemental Report of Novenber 18, 1976 and requests that the Comittee further clarify one of its recomendations, specifically, ' that the Comittee specify the " danger" that is of concern if instru-mentation and control are not separated; further describe the type of separation required (e.g., physical or other); and specify a standard for conformance. , ne Board further notes that8 this illustration is only an exanple of an area where a problem may exist and further elaboration of other matters may also be required. . e e
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t Honorable Marcus A. Ibwden March 16, 1977 , The Comittee appreciates the Board's desire and interest in understanding the issues identified by the Cortnittee but does not agree with the method ._ - being used to develop this understanding. The Comittee's Supplemental t Report dated Novertber 18, 1976 did provide a brief description of the~ items
. considered to have been problems by the Comittee and specific cross ref-erences to other applicable cases, as required by the Court in Aeschliman
? vs. NRC. .
% e desire for additional clarification by the Board with respect to spe-cific questions of this nature is best served by:
- Examination of the record related to the Midland review and the review of other cases specifically cross-referenced by the Comittee.
- Discussion with the NRC Staff who participate in the Comittee's review process, are thoroughly familiar with the problems and issues involved, and are participants in the hearings.
t The example chosen by the Board is itself a case in point. The matter of separation of control and protection instrumentation relates to reducing the probability of failure due to a comon cause and is dealt with gener-ically by Section 7.3 of the NRC's Standard Review Plan, which provides guidance to Staff reviewers; the Comittee provided a specific reference, in its November 18, 1976 Supplemental Report, to the Three Mile Island Nuclear Station, Unit 1, in response to the Court's order to provide a
" cross-reference to the previous reports in which those problems and the measures proposed to solve them were developed in more detail." The July 11,1973. Safety Evaluation of the then Directorate of Licensing in the matter of Three Mile Island, Unit 1, deals directly with this ACRS concern in Section 7.5, " Separation of Control and Protection Systems" and the Comittee's August 14, 1973 report on operation of t ree Mile Island, Unit 1, indicates that this matter was no longer of concern for the Three Mile Island case. In the Midland case, the Comittee will review the adequacy of the final design as it exists at the time it re-views the Midland Plant for an operating license.
In general, we believe that examination of the implementation of the Com-mittee's advice and of any resulting changes in the application are best left to the NRC Staff which plays a direct role in the hearing, and that any evidence relating to such matters should be sought from them. Indeed, the Court in Aeschliman itself notes, "This is not to say that an ACRS report must contain detailed factual findings of the kind necessary to aid ! judicial review. Under Comission rules, when ACRS conclusions are oon- l troverted, a factual record is conpiled anew before the Licensing Board." l
U T March 16, 1977 Honorable Marcus A. Rowden . . He NRC Staff (previously, the AEC Regulatory Staff) has routinely ad- , dressed itself to the coments and recomendations in ACRS A typical reports is example forto - many years as part of the NRC hearing process.be 15, 1973.found in Supplem
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Evaluation for Three Mile Island, Unit 1, dated October w Chapter 4 of that document is addressed entirely to the issues raised in the ACRS report of August 14, 1973. III. This section of the Midland Board's most recent request points to per-ceived " ambiguities" resulting from an examination of several ACRS reports provided as references in the Comittee's Supplemental Report of November 18, 1976. W e Board notes that those references contain
" ambiguities" similar to the ones cited by the Court in AeschliJran and points, by way of example, to the Comittee's reference 12, to "other 1970. he g problems" in it's Hutchinson Island report of MarchBoard asks th identified and described as the Court directed.
18, 1976 was provided We Comittee's Supplemental Report of November as ordered by the Court to identify those "other problems" which had been considered applicable to the Midland Plant'at the time of the CP review and which were noted generically in the ACRS report of June 18, 18, 1976 1970. Any items not so identified in the Comittee's November report were not considered applicable to Midland during the CP . review. The Comittee will be in a position to update this list and address the current status of specific items when it has completed its review for an Operating License for the Midland Plant. W is review has not ~ yet been scheduled. In sumary, the Comittee believes that the response already provided in'its Supplemental Report of November 18, 1976, fully meets the re-quirements of the Aeschliman Court since: (1) he Court requested elaboration only of those items
- referred to in the Comittee's original report as "other problems" and no others.
(2) De Comittee's Sdpplanental Report of November 18, 1976, did provide a "short explanation understandable
- to a layman of the additional matters of concern to the Comnittee and a cross-reference to the previous reports in which those problems, and the measures proposed to solve them, were developed in more detail" as specifically directed by the Aeschliman decision.
o l Honorable Marcus A. Rowden March 16, 1977 -
~
(3) ne conmittee's Supplemental ' Report of Novenber 18, j 1976, fully identified all additional matters of '
- concern to the Comnittee during its CP review of -
the Midland Project. h e ACRS does not feel that any further clarification of its reports on Midland is necessary. Sincerely yours, 9n. M. Bender Chairman Attachments:
- 1. F. J. Coufal, Chairman, AS&LB letter to M. Bender, ACRS, dated January 28, 1977.
- 2. Excerpt from the decision in Aeschliman vs. NRC.
0 4 I O e oe eea
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i I
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I SUPPLEMENT NO. 2
.g TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF CONSUMERS POWER COMPANY MIDLAND PLANT UNITS 1 AND 2 DOCKET NOS. 50-329 AND 50-330 U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D. C. 20555
- l. .
i'. . TABLE OF CONTENTS j PAGE
1.0 INTRODUCTION
.. .............. ....................... ................. 1-1 1
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2.0 I DISCUS $10N OF THE ELEVEN ITEMS IN ADVISORY COMMITTEE ON f 8 REACTOR SAFEGUARDS LETTER REPORT DATED NOVEMBER 18, I 1976............... 2-1 l 2.1 1 Separa tion of Protection and Control Instrumentation. . . .. . . . .. . . .2-1 . 2.2 Vi bra tion a nd loo s e Pa r t s Mo ni t oring . . . . . . . . . . . . . . . . . . . . 2-2
........... [
2.3 ! 2.4 Po tential for Axial Xenon Osci11a tion. . . . . . . . . . . . 2-2 .................. I 2.5 The Behavior of Core-Barrel Check Valves in Nomal Condition....... 2-4 C The Potential Consequences of Fuel Handling Accidents........ ..... 2-4 9 2.6 2.7 The Ef fect on Blowdown Forces on Core Internal s. 2-5 . . . . . . . . . . . . .. . . . . Assurance That Loss-of-Coolant-Accident-Related Fuel Rod Failures Will hot Interfere With Emergency Core Cooling System....... ....... 2.8 25 The Effect on Pressure Vessel Integrity of Emergency Core Coolin 2.9 Sy s t em I nduced Th ern:a 1 S hoc k . . . . . . . . . . . . . . . ... . . . .2-6
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Environmental Qualifications of Vital Equirent in Containment..... 2- 7 ( 2.10 lautrumenta tion to Follow the Cour ses of an Accident. . 28 .............. 2.11 Improvad Quality Assurance and in-service Inspection r of e imary Systems..... ......
............... ............... ..... ....... 2- 9 3.0 ',.
CONCLUSIONS... ......... ............. ........................ ... 31 ..... 4.0 i REFERENCES........................................ ........ .. ............ 41 APPENDIX A Chronology of Milestones $1nce Issuance of the Safety Evaluation Report..............................
... ...................... . A-1 APPENDIK B Supplemental Report on Midland Plant. Units 1 and 2 dated Novertber 18, 1976........
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f 1
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O m _.s m -
1.0 INTRODUCTION
The Midland Plant Units 1 and 2 Safety Evaluation Report was issued on November 12 1970. On January 14, 1972 a Supplemental Safety Evaluation Report was issued : providing the staff's review and conclusions regarding the Midland Plant Urf ts 1 and 2 emergency core cooling system, radioactive waste treatment systen and Consumers Power Company financial qualifications. f l Construction permits CPPR-B1 and CPPR-82 authortring construction of the Midland Plant Units 1 and 2 were issued to the Consumers Power Company (Licensee) on o December 15, 1972. e d On July 21. 1976, the United States Court of Appeals for the Ofstrict of Columbia *
. Circuit in helsen Aesch11 man, et al. v. U.S. Nuclear Regulatory Commission. $47 F2d 622
-{
- among other matters rer,anded for clarification the June 18, 1970 report issued by k:
the Advisory Comittee on Reactor Safeguards (ACR$) for the Midland Plant Units 1 and 2. l 'l On August 16, 1976, the Commission reconvened an Atomic Safety and Licensing Board { (Beard) to censider whether the construction permits for the Midland Units 1 and 2 should be continued, modified or suspended in light of the issues remanded by the O.C. Circuit. In a letter dated October 14, 1976, the licensing Board returned the original Advisory Comittee on Reactor Safeguards (ACRS) report to the ACRS for clarification . in response to the Board's request, the ACR$ issued a " Supplemental Report on Midland l I Plant Units 1 and 2" dated November 18, 1976. This ACRS report is attached as Appendix B. Therein the ACR$ identified eleven items which were: } "...those items referred to in its paragraph on 'other ) problems related to large water reactors' which had i ) been previously' identified by the Regulatory Staff I and the ACRS.' and which the Corrnittee considered f . applicable to the Midland Plant." (Page?ofthe November 18, 1976 letter.) [ i The purpose of this Supplement No. 2 to the Safety Evaluation Report for the Midland Plant Units 1 and 2 is to provide an updated status and identify resolutions of the ;
)
eleven identified ACRS items for the Midland Plant, i. ! Appendfr A to this supplement provides an update of major milestones that have occurred for this facility since the issuance of the Safety Evaluation Report on November 12. 1970. l 1-1
l 8
- 2. 0 D15CU55!0N OF THE 11 ITEM 5 l
IN ADVISORY COWITTEE ON REACTOR SAFEcd!ARDS LETTER REPORT DATED NOVEMBER 18. 1976_
- i 2.1 Separation of Protection and Control Instrumentation i i
"1.
Separation of protection and control instrumentation - The Applicant proposed using signals from protection instruments for control purposes. The Comittee believed that control and protection instrumutation shoulde be separated to th ( fullest extent practicable, and recorrended that the Applicant explore f urther the i possibility of making saf'ety instrumentation more nearly independent functions. rol of cont (ThreeMlleIsland,1/17/68)." The physical independence design of the circuits and electric eprising or equipment cc } associated with the Class IE power system, the protection system, systems actua r j controlled by the protection system and auxiliary or supporting systems must assu that operability of the protection system and the systems e it actuat s t o perform their safety-related functions are not compromised by eany failure in th control I system or other nonprotection systems. l in order to show the resolution of the above Advisory egua Cormittee rds. on Reactori Saf (ACRS) item it is necessary to establish the chronology of events . [ The ACRS concern was first indicated in the Three Mile Island ACRS letter dated January 17 , Appendix 8 to this report) in the third paragraph of the second .1968.(See , page. On August 30
, 1968, the Institute of Electrical and Electronics Engineers (IEEE 279) stl -
issued and accepted for purposes of regulations applicableandard to was
' nuclear power plants.
- This standard sets forth the criteria and requirements for separati on of control and protection instrumentation used in our evaluation of the Midland esign. At Plant d }
the 122nd ACRS meeting held on June 11-13, 1970, the ACRS Comittee completed its , review of the application by the Constsners Power Company fo'r a permit to ! the Midland plant. Units 1 and 2. construct ' On June 18,1970, Units 1 and T. was issued by the ACRS wherein aration the concernof regarding septhe R protection and control instrumentation identified by the Three Mile Island ACRS 1etter was referenced.
- In the Safety Evaluation Report by the Atomic Energy uclear Comission (now the N Regulatory Comission) for the Midland Plant issued, on
' November 12 1970 in Section 8.1 " Instrumentation and Control" thg staff indicates that umers the appitcant (Cons Power Company) will design the instrumentation for protection ystems to and control s conform to IEEE-279 dated August 1968, and we concluded this was acceptable.
e
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e
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Tnerefore the separation of protection and control instrumentation design for the Midland Plant was resolved in the Staff's Safety Evaluation Report issueJ November 12
'{
1970 by the applicant's comitment to meet the requirements of the IEEE-279 Standard cated August 1968. In Section 8.4. " Installation Criteria." of our Safety Evaluation Report dated November 12. 1970. we indicated the applicant will develop more detailed criteria and procedures for installation of protection and eme,gency power systems ,as recomended by the ACRS. In Amendment 32 to the Preliminary Safety Analysis Report. the applicant, to comply with this recomendation, submitted the detail criteria and procedures for installation of the protection and emergency power system for the balance-of-plant. These detailed criteria and procedures are currently being reviewed by the staff and will be completed prior to installation of these systems. The detail criteria and procedures for installation of the protection and emerge 9cy power systems for the nuclear steam supply system will be submitted for staff review and approval
, prior to installation of these systems.
Acceptability of the final implementation of tLese requirements will be determined during the operating license review stage for the Midland facility. 2.2 Vibration and loose Parts Monitoring "2. Vibration and loose parts monitoring - The Comittee recomended that the Applicant study possible means of in-service monitoring for vibration or the presence of loose parts in the reactor pressure vessel as well as in other portions of the prie:ary system, and implement such means as found practical and appropriate. (Falisades.1/27/70)." The loose part monitoring system is designed to provide early warning of reactor components which have failed or are vibrating to such an extent that failure may be irr inen t. The staff Safety Evaluation Report on the Midland Plant issued on November 12, 1970.
' and the Supplemental Safety Evaluation issued on January 14. 1972, make no mention of a requirement for a loose parts and vibration monitor. However, in a response to the staff question 4.5 in Amendment No. 5 to the PSAR. the licensee comitted to install a loose parts monitor if a practical and reliable system were available. Such equipment as might ultimately be required can be in the nature of add-on equipment which could De added to a plant at any time. Loose part monitoring systems are currently available and we will resolve this matter during the operating license review stage for the Midland Plant Units 1 and 2.
Acceptability of the final implementation of this requirement will be determined during the operating license review for the Midland Plant. Units 1 and 2. 2.3 Potential for Axial Xenon Oscillations
-3. Potential for axial xenon oscillations - The Applicant was continuing studies on the possible use of part-length rods for stabilizing potential menon oscillations.
2-2 1 1 i i e
Solid poison shims were to be added to the fuel elements if necessary to make the moderator temperature coefficient more negative at the beginning of core life. (Three Mile Island.1/17/68)." This subject is addressed in the staff's Safety Evaluation Report in the Midland Plant issued on November 12. 1970. The staff noted that analyses at that time - indicated that the core would be stable to potential radial or azimuthal power oscillations due to menon, and that potential axial oscillations could be controlled by use of part-length control rods, i Tests of core stability were perfomed during start-up tests for the Oconee Unit I reactor, a sister or similar type unit to the Midland Plant reactors. A diagonal (combination of axial and azimuthal) oscillation was induced at 75 percent full power and the reactor response was monitored for 72 hours. The azimuthal component of the oscillation was damped, but the axial component was divergent. At 70 hours into the transient, the part-length rods were used to suppress the axial imbalance which was redaced to near zero where it was maintained. On the basis of this demonstration of azimuthal stability of the Oconee IJnit I reactor (essentially identical to the Midland Plant reactors) and the ability of the control system to suppress axial oscillations, we conclude that this concern is resolved for the Midland Plant. i The use of solid poison shim rods in the fuel element provides a means to assure that , the moderator temperature coefficient is only slightly positive or negative throughout the core life. Where some Technical Specifications allow a slightly positive coefficient. I the accident and stability analyses take this into account. Burnable poison previsions have been designed into the Midland Plant fuel to reduce excessive positive coefficients to allowable values. f t Acceptability of the final implementation of these requirements will be determined I during the operating license review for the Midlant Plant, f i 2.4 ! The Behavior of Core-Barrel Check Valves in Nomal Operation "4 The behavior of core-barrel check valves in nomal operation - The Applicant had ?
}
proposed core-barrel check valves between the hot leg and the cold Ieg to insure I proper operation of the ECCS under all circumstances. Analytical studies had indicated i that vibrations would not unseat these valves during nomal operation. The Committee desired r.q that this point be verified experimentally. (Three Mile Island.1/17/68)." This matter is of generic concern to nuclear steam supply systems designed by Babcock and Wilcox. Other reactor vendors do not use core-barrel vent valves.The concern of the Comittee was that there was a potential for the core-barrel check valves to open during normal operation allowing excessive core by-pass flow. , 2-3 ~ -o . .
For the Oconee units, which are of Babcock and Wilcox design and sister or similar type units to the Midland design, the staff initially imposed a 4.6 percent reactor coolant flow penalty in the thermal-hydraulic design analysis to provide conservatism due to the possibility of leakage through the vent valves during normal operation. By letter to the licensee of the Oconee Nuclear Station (Duke Power Company) dated January 30, 1976, the staff advised the Itcensee thct it had concluded that sufficient ' evidence had been provided by Babcock and Wilcox to assure that the core-barrel vent i valves would remain closed during normal operation. Accordingly, we advised the Ifeensee that the vent valve flow penalty could be eliminated provided the licensee l established appropriate surveillance requirements to demonstrate, at each refueling outage, that the vent valves are not stuck open and that they operate freely. The Oconee resolution of this matter is directly applicable to the Midland Plant design since the designs are identical and the matter is therefore satisfactorily resolved for the Midland Plant. Acceptability of the final implementation of the core-barrel check valves requirements will be determined during the operating Itcense review for the Midland Plant. 2.5 i The Potential Consequences of Fuel Handlino Accidents "S. The potential consequences of fuel handling accidents - The Cornittee believed l that further study was required with regard to potential releases of radioactivity in the unlikely event of gross damage to an irradiated subassembly during fuel handling f and the possible need for a charcoal filtration system in the fuel handling building. ; The Committee recommended that this matter be resolved in a manner satisfactory to the Regulatory Staff. (Hutchinson Island, 3/12/70)." This concern is resolved by General Design Criterion 61 of Appendix A to 10 CFR Part ; 50 which requires that fuel storage and handling systems be designed to assure l adequate safety under normal and postulated accident conditions. Regulatory Guide 1.13 " Spent Fuel Storage Facility Design Basis" describes a method acceptable to the staff for implementing General Design Criterion 61 j By letter to the Licensee dated September 29, 1976, the staff noted that the initial design of the Midland Pland did not include charcoal filters in the exhaust system for the spent fuel storage facility. However, the staff also noted that during discussions with the Licensee, the Licensee had agreed to install charcoal filters in conformance with Regulatory Guide 1.13. On the basis of this commitment by the i Licensee, the staff concluded that the design of the Midland Plant is in conformance j with Regulatory Guide 1.13 and is acceptable. [l Il Acceptability of the final implementation of the Regulatory Guide 1.13 requirements I will be determined during the operating licensee review for the Midland Plant. ! l t 24 . I
. i 2.6 The Effects of Blowdown Forces on Core Internals _
s' "6. The effects of blowdown forces on core internals - The Comittee recomended that the Regulatory Staff review the effects of blowdown forces on core internals and the develogrnent of appropriate load combinations and defomation Ifmits. (Three Mile Isla nd,1/17/68) ." ,
- h In the Safety Evaluation Report on the Midland Plaet dated November 12, 1970 the
- above ACR$ item was resolved to the staff's satisfaction as discussed in Section 5.4, '
' Reactor Vessel Internals," and Section 15.10. " Blowdown Forces on Core Internals."
l This matter is partially covered by Regulatory Guide 1.20. " Comprehensive Vibration . Assessment Program for Reactor Internals During Preoperational and Initial Startup } g Testing." By letter dated September 29, 1976, the staff informed the Licensee of our I conclusion that the Midland design was in full confomance to Regulatory Guide 1.20. I There have been recent additional concerns raised about the loads on reactors internals ' during a loss-of-coolant acef dert. The staff now is working with all the reactor vendors on this matter. The vendors, including Babcock and Wilcox, are developing themal- hydraulic codes that properly handle the loadings on the core internals during subcooled blowdown. We expect that versions of these codes acceptable to the staff will be available within about one year. To date, the preliminary indications are that the internals design of the Babcock and Wilcox reactors will withstand the - blowdown loads and is acceptable. In the event analyses indicate that the internals design is not acceptable design modifications may be required. Resolution will be made during the staff's operating license review for the Midland Plant Units 1 and 2. 2.7 Assurance That less-of-Coolant-Accident-Related Fuel Red Failures Will Not Interfere Wfth Emergency Core Cooline System "7 Assurance that LOCA-related fuel rod failures will not interfere with ECCS function - The Comittee desired to enphastre the importance of work to assure that fuel-rod failures in loss-of-coolant accidents will not affect significantly the ability of the ECCS to prevent clad melting. (Three Mile Island,1/17/68)." ' r This matter is considered resolved on the basis of the generic rulemaking hearing on Acceptance Criteria for Emergency Core Cooling Systen:s for Light-Water-Cooled Nuclear Power Reactors, RMS-50-1, which resulted in promulgation of regulations, specifically 10 CFR 50.46 and Appendix K to 10 CFR Part 50 to which all nuclear plants must comply." , The Midland Plant will be required to confom to these requirements which will assure i that fuel rod failure will not interfere with the amergency core cooling system function. Operating Plants of the Midland type are now meeting the requirements of Appendix K to 10 CFR Part 50 and 10 CFR 50.46. As indicated in our Supplemental Safety Evaluation Report dated January 14, 1972, the Midland Plant Unit 1 and 2 Emergency a Core Cooling Systems met the Comission's Interim Policy Statement and Acceptance Criteria. 1 2.$ , 1
. l 1
l l
During the operating license review the staff will require that the Midland Plant Units 1 and 2 emergency core cooling system meet the requirements set forth in . Appendix K to 10 CFR Part 50 and 10 CFR 50.46. The matter is therefore satisfac-torily resolved for the Midland Plant. 2.8 The Effect On pressure vessel Integrity of Emergency Core Cooling System Induced . Thermal Shock "8. The effect on pressure vessel integrity of ECCS induced thermal shock - The Committee reca, mended that the Regulatory Staff review analyses of possible effects, upon pressure-vessel integrity, arising from thermal shock induced by ECCS operation. (Oconee , 7/11/67) ." Regulatory Guide 1.2, " Thermal Shock to Reactor Pressure Vessels," covers current information on this subject. The ultimate position as to the significance of thermal shock requires input of fracture mechanics data on irradiated steels from the Heavy Section Steel Technology (HSST) program. The Nuclear Regulatory Commission's confirmatory safety assessment of hot reactor pressure vessels subjected to thermal shock effects due to injection of cold emergency core cooling system water following a loss-of-coolant accident is continuing and the third thermal shock experiment has been conducted at the Oak Ridge National Laboratory. In this experiment, a het 21-inch diameter steel cylinder with a deliberate flaw was quickly cooled with a wate*-$1cchol mixture at -10'F, thereby producing a severe i thermal shock. preliminary studies of the flaw indicate that it grew uniformly in depth, as predicted. These results, plus those from the first two tests, have provided important assurance of the validity of the thermal shock analysis methodology to predict crack initiation and extension in reactor presure vessels under LOCA-type conditions. The work done to date at ORNL, plus results from the haval Research Lab on the beneficial effects of warm prestressing, demonstrate that flawed, irradiated reactor vessels subjected to thermal shock from LOCA-ECCS water, will not fall catastrophically. Two tests on warm prestressing are scheduled for FY 77, to provide the final verifica-tion r.ecessary for routine application of the methodology by the licensing staff. Additional testing of 39-inch cylinders may also be conducted provided liquid nitrogen can be used as the coolant. In a letter to the Applicant dated Septenber 24, 1976, the staff concluded that the Midland design conforms to Regulatory Guide 1.2. Pending results from the HSST program, which is designed to confirm the validity of the analytical design model for irradiated pressure vessels, conformance to Regulatory Guide 1.2 and design, of vessels in accordance with the ASME code and subsequent adherence to guidelines for I 2-6 I l l
o ____..# t
.a surveillance of radiation damage and nil-ductility transition temperature changes resulting there from are acceptable to the staff as proper assurance against pressure vessel failure.
i Should the pressure vessel surveillance program for the Midland Plant Units 1 and 2 . Indicate that greater than anticipated irradiation damage is occurring to a Midland
- }
reactor pressure vessel, the Licensee will be required to anneal the vessel to ' restore the toughness properties to acceptable values. 2.9 Environmental Qualificatiers of Vital Equipment in Containment "9. Environmental qualification of vital equipment in containment - The Committee recommended that attention be given to the long. term ability of vital components, such as electrical equipment and cables, to withstand the environment of the contain-ment in the unlikely event of a loss-of-coolant accident. (Palisades. 1/27/70)." The concern regarding this matter is verification that systems and components located in the containment, and required to function during and following a loss-of-coolant cu ' accident. can withstand the temperature, pressure, humidity and radiation conditions which ceuld occur in the containment. The qualification requirements of critical components are now covered by Regulatory Guides 1.40. 1.63. 1,73 and 1.89 and by IEEE Standards 382-1972. 393-1974. 317-1972, and 323-1974 which provide acceptable methods to meet these qualification requirements. ' i The staff review of the Midland Plant regarding qualification of vital equipment in containment is not complete. Completion most likely will not occur until the staff review of the operating license application for the plant. However, since this matter deals exclusively with components, rather than structures, continued construction of the plant would not preclude possible upgrading of components, if required, during the operating license review. Due to the advanced state of construction and procurement on the Midland facility. . complete cumpliance with the above guides and standards will not be required on all
- components and equipment. however. exceptions will be required to be justified during the operating license review.
Regulatory Guide 1.40 *0ua11fication Tests of Continuous-Duty Motors Installed IEside the Containment of Water-Cooled Nuclear Power Plants" was reviewed by the staff and the Licensee. By letter dated September 29, 1976, we informed the Licensee that the staff had concluded that the Midland design was in full conformance to this Regulatory Guide. I Regulatory Guide 1.63 " Electrical Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants" endorses IEEE Standard 317-1972. By letter to the Licensee dated September 29. 1976, the staff informed the Licensee that additional 2-7 s.
- 9
infomation would be required regarding the ability of penetrations to withstand, without loss of mechanical integrity, the maximum possible fault current vs time conditions (position C.1 of the Guide). Regulatory Guide 1.73 "Quelification Tests of Electric valve Operators Installed Inside the Containment of Nuclear Power Plants" endorses IEEE Standard 382-1972. By letter to the Licensee dated Septembe* 29, 1976, the staf f infomed the Licensee that , implementation of this guide is acceptable. Regulatory Guide 1.89 " Qualification of Class 1E Equirent for Nuclear Power Plants" endorses IEEE Standard 323-1974 This guide was issued in November of 1974 and it notes that the staff may reevaluate the plant design on a case-by-case basis to assure that acceptable methods for qualifiestion of Class lE equirent have been specified in purchase orders executed after November 15. 1974 The degree of confomance of the Midland design to the guidelines of this Regulatory Guide has not yet been evaluated by the staff. Such evaluation will occur during the staff review of the operating license application. IEEE Standard 383-1978 pertains to the type testing of cables, splices and connections for nuclea, power plants. It is a sub-element of IEEE Standard 323-1974, which is endorsed by Regulatory Guide 1.89. Acceptability of the final implementation of the guides and standards requirements for environmental qualification of vital equipment will be detemined during the operating license review for the Midland Plant. 2.10 Instrumentation to Follow the Course of an Accident "10. Instrumentation to follow the course of an accic'ent - This item related to the development of systems to control the buildup of hydrogen in the contairvnent, and of instrumentation to monitor the course of events in the unlikely event of a loss-of-coolant accident. (Hutchinson Island. 3/12/70)." During and following a loss-of-coolant accident, hydrogen is generated by radiolysis and water-metal reactions, a system such as a hydrogen recombiner is required to assure the hydrogen concentration within the containment remains below the flamability limit. General Design Criterion 41 requires that systems to control hydrogen, oxygen and other substances which may be released into the reactor containment be provided as necessary to control their concentrations following postulated accident to assure that containment integrity is maintained. Regulatory Guide 1.7 (Safety Guide 7)
" Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident". describes a method acceptable to the staff for implementing General Design Criterion 41. The issuance of the regulation 10 CFR Part 50 Appendix A which provides the General Design Criteria for light-water reactors occurred after the Midland ACRS 2-8 .
i 1 i
letters were issued. These criterte were not issued as a Commission regulation at the t time the Midland application was under review but were used to evaluate the Midland design. j In a letter to the licensee dated September 29, 1976, the staff noted that the Licensee has committed to comply with the design guidance and assumptions for analysis contained in Regulatory Guide 1.7 as supplemented by Standard Review Plan Section 6.2.5 and
- Branch Tc:hnical Position CSB 6-2. " Control of Combustible Gas Concentrations for t Containment Following a LOCA." The staff found this design approach to be acceptable, but noted that we will review the combustible gas contro) system design and supporting
. analyses in conjunctic ' with the appiteation for an operating license.
Acceptability of the final implementation of hydrogen control system requirements will be determined during the operating license review for the Midland Plant. The matter of instrumentation to follow the course of an accident still is carried by the Connittee in the " Resolution Pending" category of concerra. Regulatory Guide 1.97 " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," was distributed for connent in December
- 1975.
Comnents now have been received, the guide is being revised as deemed appropriate by the staff and by the Committee, and the present schedule calls for publication in 1977. The Licensee will be required to meet the requirements of the Regulatory Guide ' 1.97 or provide staff approved alternatives. Acceptability of the final implementation i of the Regulatory Guide 1.97 requirements will be determined during the operating license review for the Midland Plant. ' I 2.11 Improved Quality Assurance and In-Service Inspection of Prinary System "11. Improved quality assurance and in-service inspection cf primary system - The ' Connittee continued to emphasize the importance of quality assurance in fabrication of the primary system as well as inspection during service life, and recon 7 ended that the Applicant implement those improvements in quality practical with current technology. (Oconee,7/11/67)." This concern is satisfactorily covered by Appendix B to 10 CFR Part 50 which specifies
- the requirements for a quality assurance program for design, construction and opera-tion of a plant and Regulatory Guides 1.28, 1.30, 1.37, 1.38, 1.39, 1.58, 1.64, 1.74, 1.88 and 1.94 which describe procedures for implementing the requirements of Appendix B.'~ lhe quality assurance program for the Midland Plant meets these requirements.
~
During a recent review by the staff to determine the extent of conformance of the Midland Plant Units 1 and 2 to these various Regulatory Cuides, the Licensee elected to upgrade the quality assurance program to meet the requirements indicated above. In a letter to the Licensee dated September 24, 1976, the staff reported that it had reviewed the revised quality assurance program description submitted by the Licensee in March of 1976, which incorporates Consumers Power Company Topical Report CPC-1, Bechtel Topical Report BQ-TOP-1, Revision lA dated May 1,1975, and Babcock and 2-9
- Dilcox Topical Report BAW-10096A, Revision 3, of Consumers Power Company, the Bechtel Corporation, and the Babcock and Wilcox Company. They replace the quality assurance program described in the Preliminary Safety Analysis Report for the Midland Plant.
We therefore consider the quality assurance program for the Midland Plant to comply with Appendix B of 10 CFR Part 50 and is acceptable. lhe Safety Evaluation Report for the Midland Plant, dated November 12, 1970, states on page 25 that in-service inspection will comply with the draft ASME tode for the In- - Service Inspection of Nuclear Heactor Coolant Systems (N-45) which is equivalent to ;I Section XI of the ASME Boiler and Pressure Vessel Code. By letter to the Licensee ) dated September 24, 1976, the staff concluded that the degree of conformance to I kegulatory Guide 1.65, " Materials and Inspection for Reactor Yessels Closure Studs," is acceptable. Recently the requirements of the ASME Code Section XI was incorporated into the regulation 10 CFR Part 50.55 and the Midland Plant will be required to meet the ASME code Section XI requirements or justify and request relief for any nonconformance, the matter of in-service inspection, therefore, is adequately resolved for the Midland Plant and, the quality assurance program is acceptable as noted above. Acceptcility of the final implementation of the requirements of the ASME Code Section XI will be determined during the operating license review for the Midland Plant. N ha 2-10 ig
a um
3.0 CONCLUSION
S Based on our review of the eleven items referred to in the Advisory Comittee on - Reactor Safeguards letter dated Jane 18, 1970 we find that our conclusions stated f I
'in Section 19.0 of our Safety Evaluation Report dated November m
- 12. 1970 are unchanged.
As discussed above, the ACRS has clarified the problems identified in their Midland letter and this report provides the status and resolution by the staff for the Midland Plant for the eleven identified items. E I l t r l 1 i lY
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4.0 REFERENCES
i The following references were used to provide the basis for acceptance and resolution of the eleven matters in Section 2.0 of this report and are available to the public at the Public Document Room,1717 H St. NW Washington DC. NI (1) ACRS report of April 16,1976. " Status of Generic Items Relating to Light Water Reactors: Report No. 4". (2) Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (3) Institute of Electrical and Electronic Engineers Standard (IEEE-279) Criteria for Protection Systers for Nuclear Power Generating Stations (4) Safety Evaluation Report on Midland Plant Units No. I and 2, Docket Nos. 50-329 and 50-330 dated hovember 12, 1970 (5) Supplemental Safety Evaluation Report sn Midland Plant Units ho. I and 2, dated A nuary 14, 1972 (6) Letter from huclear Regulatory Corr:ission to Duke Power Comp 3.: Guted January 30, 1976, with attachment entitied
- Report Evaluation: B&W Operating 1 Experience of Reactor Internals Vent Valves" (7) Regulatory Guide 1.13
" Spent Fuel Storage Facility Design Basis".
(8) Regulatory Gaide 1.20 .
" Comprehensive Vibration Assessment Frogram for Reactor Internals During Preoperational and Initial Startup Testing" j (9) I Acceptance Criteri* for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors RMS-50-1 {
(10) Appendix K to 10 CFR Part 50. Emergency Core Cooling Systems Evaluation Models j (11). Regulatory Guide 1.2 - Thermal Shock to Reactor Pressure Vessels l
, (12)
Regulatory Guide 1.40 - Qualification Tests of Continuous-Duty Motor Installed Inside
,: - the Containment of Water-Cooled Nuclear Power Plants (13)
Regulatory Guide 1.63 - Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants e 4-1 1\ i
*11 1'
l Ii 11
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o 1 - l l - 1 * (14) Regulatory Guide 1.73 - Qualification Tests of Eiectric Valve Operators Installed Inside the Containment of Nuclear Power Plants p (15) Regulatory Guide 1.89 - Qualification of Class IE Equipment for Nuclear Power Plants l (16) letters from Nuclear Regulatory Comission to Consumers Power Company dated ' September 29, 1976 entitled " Midland Plant Units 1 and 2 - Regulatory Guide Review". h l (17) Regulatory Guide 1.7 - Control of Combustible Gas Concentrations in Containment Following a loss-of-Coolant Accident (18) Standard Review Plan - NUREG 75/087 (19) Regulatory Guide 1.97 " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions and Following an Accident" under review (20) Regulatory Guide 1.28 - Quality Assurance Program Requirements (Design and Construction) (21) Regulatory Guide 1.30 - Quality Assurance Requirements for the Installation. Inspection, and Testing of Instrumentation and Electrical Equipment (22) Regulatory Guide 1.37 - Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (23) Regulatory Guide 1.38 - Quality Assurance Requirements for Packaging, Shipping. Receiving Storage, and Handling of items for Water-Cooled Nuclear Power Plants (24) Regulatory Guide 1.39 - Housekeeping Requirements for Water-Cooled Nuclear Power Plants (25) Regulatory Guide 1.58 - Qualification of Nuclear Power Plant Inspect 4on, Examination and Testing Personnel (26) Regulatory Guide 1.64 - Quality Assurance Requirements for Design Nuclear Power Plants (27) Regulatory Guide 1.88 - Collection, Storage, and Maintenance of Nuclear Power Plants Quality Assurance Records (28) wgulatory Guide 1.94 - Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants YI 4-2
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(29) Appendix B to 10 CFR Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Puel Reprocessing Plants ]. I 8
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r
, APPENDIX A CHRONOLOGY OF MILESTONES SINCE ISSUANCE 4 0F THE SAFETY EVALUATION REPORT I !- Safety Evaluation Report issued November 12, 1970 First prehearing conference held to outline Noverber 17, 1970 hearing agenda.
In compliance with AEC environmental November 25, 1970 regulations, Consumers Power files responses to agency coments. Construction permit hearing begins and December 1,1970 adjourns af ter nine hour session. k1 l AEC issues " final" regulations to December 4,1970 implement NEPA. I Delay in resumption of hearing January 7, 1970 requested by attorney for Sagi'naw Intervenors. I I Consumers Power submits additional January 19, 1971 environmental data to AEC. l - Prehearing conference held. January 21, 1971 Prehearing conference scheduled for January 26, 1971 January 30 cancelled at the request
! of attorney for Saginaw Intervenors.
l AEC issues draft environmental statement February 5,1971 on Midland plant for coment by federal, state and local agencies and the public. Scheduled May 17 resumption of hearing May 3, 1971
, cancelled at the request of attorney for Saginaw Intervenors, l
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U.S. Environmental .'rotection Agency May 20,1971 issues evaluation of environmental effects of Midland plant, concluding the site is " suitable for the facility as planned." Prehearing conference orders resumption June 7,1971 , of hearing after request for postponement i by attorney for Saginaw Intervenors. Construction pemit hearing resumes. June 21, 1971 Hearing adjourns for board members to June 26,1971 attend conference on emergency core cooling systems. AEC issues irterim standards for June 29, 1971 ' perfomance of emergency core cooling systems. Hearing resumes. July 7, 1971 Hearing adjourns as previously scheduled. July 23, 1971 Calvert Cliffs-NEpA court decision announced. AEC ordered to revise its environmental regulations. Saginaw Intervenors request manufacture August 5,1971 of Midland plant components be prohibited. Mapleton Intervenors request application August 13, 1971 i or construction pemit be dismissed. AEC issues new environmental regulations September 9,1971 to comply with Calvert Cliffs decision. Consumers Power .*ubmits 1100 page October 20, 1971 supplemental environmental report in compliance with new AEC environmental regulations. Computer analysis of Midland plant's Noventer 1,1971 emergency core cooling systr is submitted to AEC. A-2
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, Prehearing conference held to formulate November 23,1971 guidelines on environmental issues.
AEC determines Hidland emergency core December 18, 1971 cooling system meets interim criteria. Draft detailed environmental statement December 18, 1971 issued by AEC. Concludes environmental .
- benefits of Midland plant outweigh the costs.
Supplemental Safety Evaluation Report issued. January 14, 1972 National hearing on emergency core cooling January 28,1972 system regulations begins in Washington, D.C. Final detailed environmental statement issued March 31, 1972 by AEC. Concludes environmental benefits to be derived from plant outweigh the adverse effects. prehearing conference to finalize hearing April 28,1972 agenda held. Attorneys for Saginaw Intervenors and Mapleton Intervenors request for indefinite postponement of the hearing is denied. Construction permit hearing resumes. May 17, 1972 Construction permit hearing ends. June 15, 1972 Atomic Safety and Licensing Board issues initial December 14. 1972 decision authorizing construction permits. Construction permits issued by Atomic Energy December 15, 1972 Comnission. Exceptions to initial decision filed by January 1973 . Mapleton and Saginaw intervenors. Atomic Safety and Licensing Appeal Board issues March 26, 1973 preliminary order imposing additional quality 5 assurance reporting requirements on construction. i I r Atomic Safety and Licensing Appeal Soard affirms May 18, 1973 initial decision authorizing issuance of con- f struction permits. , l t A-3
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, Atomic Energy Comnission issues amendment to May 23,1973 construction permits incorporating quality assurance reporting requirements.
- Construction of plant resumes. June 15, 1973 Mapleton Intervenors petition US Court July 15,1973 of Appeals for review of AEC ASLB and "
ASLAB decisions approving issuance of construction permits. Saginaw Intervenors petition US Court of August 6,1973 Appeals for review of ASLAB decision approvirg issuance of construction permits. Saginaw Intervenors ask Appeal Board to August 21, 1973 revoke construction permits. Consumers Power and Dow authorized by August 23, 1973 US Court of Appeals to intervene Mapleton appeal. Saginaw Intervenors' motion to revoke September 18, 1973 construction permits denied by ASLAB. AEC inspection questions adequacy of procedure November 6-B. 1973 used to maintain cadwelding work on containment building foundation. Company voluntarily halts cadwelding operations until question can be resolved. Saginaw intervenors file notion to reopen November 20, 1973 license hearing to consider energy con-servation issues which they claim were not adequately investigated. AEC Director of Regulation orders Consumers December 3,1973 Power to show cause why construction should not be suspended pending a showing that the Company is in full compliance with AEC quality assurance regulations. Order also continues suspension of cadwelding work. I AEC inspection determines that revised cadweld December 6-7, 1973 measuring procedures are adequate and all cadwelds have been properly made. A-4 L
u ._ AEC Directer of Regulation modifies show cause December 17. 1973 order to permit resumption of cadwelding operations. I Saginaw Intervenors file " emergency petition" December 18, 1973 asking the AEC to void the action of the Director ; of Regulation which pemitted resumption of cadwelding. s I Saginaw Intervenors file petition with AEC to December 18,1973 revoke Midland plant construction pemits. AEC denies emergency petition, thus permitting December 20. 1973 cadwelding operations to be resumed. Consumers Power files response to show cause order December 24, 1973 detailing the adequacy of its quallity assurance j program. Company also filed motion to dismiss show cause order and requests a public hearing on i its Quality assurance program if motion is denied. Saginaw Intervenors file request for public December 24. 1973 hearing in connection with the show cause cader. AEC issues order. 1) denying Saginaw Inter, enc.rs' January 21, 1974 December'18 petition to revoke the construction $ permits. 2) denying the Company's motion to dsmiss show cause order, and 3) orde aing a public hearing on the show cause order and appointing a hearing board, i } Atortic Safetv and Licensing Appeal Board for the January 22. 1974 Midland construction pemf*, proceeding issues ) memorandum disqualifying themselves from any participation in the show cause hearing. i Saginaw Intervenors petition AEC to order a hearing January 23. 1974 for reassessment of the Midland cost-benefit analysis, I citing the increased cost of the plant for support. 1 AEC denies Saginaw Intervnors November 20 motion January 24, 1974 to reopen construction pemit hearing to consider energy conservation . issues. I AEC denies Saginaw Intervenors petition for re- February 5.1974 , l assessment of the Midland cost-benefit analysis. A-5 eL
t . Prehearing conference on show cause proceeding held. March 28, 1974 Second prehearing conference on show cause proceeding. May 30, 1974 .
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AEC denies Saginaw Intervenors request to pay for July 10, 1974 1awyers and expert witness fees. Show cause hearing starts in Midland. July 16,1974 - 1 AEC concludes show cause hearings in Midland. July 18, 1974 AEC reports "We find that Consumers Power Company is Septerter 13, 1974 financially qualified to continue construction of Midland plant since it has reasonable assurance of obtaining the necessary construction funds." Atomic Safety and Licensing Board issues findings September 25, 1974 3 rom its show cause hearing. The report concludes:
- 1) Consumers Power Company is implementing its quality assurance program in compliance with AEC regulations, 2) There is reasonable assurance that such implementation will continue through the construction process. 3) Construction permits should not be suspended, modified or revoked.
First of two steam generators arrives at plant site. October 31, 1974 In service dates of the two units delayed by one November 14, h '4 year each, to 1980 and 1981 respectively. Unit 2: Fuel load Nov.1980; comercial operation 3/81; Unit 1 Fuel load 11/81; Comercial operation 3/82. AEC Safety and Licensing Board hears oral argument November 18, 1974 in Chicago on Saginaw Intervenors motion that show cause hearing record be reopened. Small fire caused by hot slag from a welder's torch November 21, 1974 damaged some electrical cables in the reactor building and caused damage to a small area of liner plate. Damage estimated at about $10,000. Oral argument in U.S. District Court of Appeals November 27, 1974 in Washington, D.C., on petition by Saginaw Intervenors and Mapleton Intervenors to reopen construction permit hearing. The two groups A-6 m i ! 1 r ___ J
i e
- allege the AEC did not adhere to provisions of the Atomic Energy Act and National Environmental l Protection Act in granting the Midland construction ',
permits. Letter from applicant providing schedule for July 3,1975 implementation of regulatory guides with regard to operating license review for - the Midland plant. i Letter from Consumer Power Company concerning in July 21, 1975 depth review of Regulatory Guide Implementation on the Midland project. Sumary of Meeting on Implementation of Cuality August 4,1975 Assurance Regulatory Guides. Letters from applicant transmitting responses to August 19, 1975 regulatory Guides 1.10, 1.61, 1.15, 1.18,, 1.19,
- 1. 27, 1. 35, 1. 5 5, 1. 57, 1. 59, 1. 60, 1. 90 a nd 1. 92.
Amendment 30 to Preliminary Safety Analysis Report September 3, 1975 containing design imformation regarding pipe lines from ultimate heat sink to service water pump structure. Letter from applicant transmitting response to September 9, 1975 Regulatory Guides 1.2, 1.14, 1.31, 1.34, 1.36, 1.43, 1.44, 1.50, 1.65, 1.66, 1.71. e l Sumary of Meeting on (Structural) Regulatory Guides September 30, 1975 1 { Letter to applicant from NRC indicating analytical October 9, 1975 ! procedures and criteria described in Amendment 30 are acceptable. Letter from applicant addressing implementation of : October 10, 1975 ' Regulatory Guides 1.20, 1.26, 1.29, 1.46, 1.67, 1.72 and,1.40 for the Midland Project. Letter from applicant responding to Regulatory Guide October 15, 1975 as 1.28, 1.30, 1.37, 1.38, 1.39, 1.58, 1.64, 1.74, 1.88 and l'.9'4.' h i A-7 b
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. ,l Letter from applicant responding to their positions November 11, 1975 regarding Regulatory Guides 1.1, 1.4, 1.7, 1.13, 1.25,1.42,1.49,1.52,1.54 and 1.70 for the #l I Midland Project.
Letter to applicant from NRC advising of a potential November 14, 1975 safety problem regardirg design of pressure vessel _ support systems. I' Additional Information Request from NRC to cpplicant November IS, 1975 on implementation of Regulatory Guides. Sumary of Meeting on Regulatory Guides (Electrical) November 21, 1975 Su m ary of Meeting on Regulatory Guides (Quality November 25, 1975 Assurance) Sumary of Meeting on Regulatory Guide (Mechanical) November 26, 1975 Letter from applicant providing information regarding December ll,1975 design of pressure vessel support system. Sumary of Meeting on Regulatory Guides (Quality December 24, 1975 Assurance) Sumary of Meeting on Regulatory Guides January 6,1976 Amendment 31 to Preliminary Safety Analysis Report January 9, 1976 providing updated information relating to maximun flooding conditions for the Midland Project. Sumary of Meeting held on Format Content and Schedule January 23, 1976 For the Final Safety Analysis Report. Additional Information requested by NRC regarding January 26, 1975 implementation of Regulatory Guides 1.26,1.29 and 1.94 for the Midland Plant. i Letter from applicant providing additional information February 5,1976 on implementation of Regulatory Guides concerning quality group, seismic classification and concrete placement. I Letter to applicant from NRC addressing Appendix 1 to February 23, 1976 10 CFR Part 50 requirements. A-8
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,a Sunrnary of Meeting to Discuss Criteria to be July 6. 1976 Used for Analysis of Breaks in High-Energy Lines.
' Letter to applicant from NRC regarding acceptance of September 24, 1976 '
Regulatory Guides for Midland Plant. , 'l Letter to applicant from NRC regarding acceptance of September 29, 1976 g Regulatory Guides for Midland plant. f Letter to applicant from NRC providing guidance regarding September 30, 1976 I 1 information required to evaluate fire protection systems. I Surr,ary of Meeting on Regulatory Guide Positions October 20, 1976 Start of Atomic Safety and Licensing Board hearing to November 30. 1976 decide if Midland Plant construction permits should be modified, suspended, or continued. 1 i, . if i A-9 es e t
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- I fl}$)__ l
<,[?> uc % UNIT E D ST AT E S
. 8' f "I'r. NUCLEAR REGULATORY COf.*f.ilSSION 32 I REGION 111
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l 7ss noostvrLT noAD cLEN ELLYN.1LUNots 00137 . (
'JAN 12 mat 4*
Docket No. 5 e , Docket No. 5 -330 q c3 y? { fW Consumers Power Company e < f f - ATTN: Mr. James W. Cook fgg y g jggj, $ Vice President Midland Project F}N "
- E & y *^'0 1945 West Parnall Road Jackson, MI 49201 h ;f t ',
Gentlemen: This refers to a special announced inspection conducted by
- Messrs. E. J. Gallagher and R. B. Landsman of this of fice and Mr. J. W. Gilray of the Office of Nuclear Reactor Regulation, Quality Assurance Branch on December 8-11, 1980, of activities at the Midland Nuclear Power Plant, Units 1 and 2, authorized by NRC Construction Permits No. CPPR-81 and No. CPPR-82 and to the discussion of our findings with you and others of your staff at the conclusion of the inspection.
The enclosed copy of our inspection report identifies areas examined during the inspection. The inspaction consisted of a review of the Consumers Power Company response and imolementation of corrective actions regarding the 10 CFR 50.54(f), Quastion 1 of NRC letter dated March 21, 1979 and Question 23, request for ad3itional information dated September 11, 1979. During this inspection, certain of your activities appeared to be in non-compliance with NRC requirements, as described in the enclosed Appendix A, and a written response is required. In addition to the above, the unresolved items described in Paragraph 3(c) and 3(d) requires your attention. Picase provide a written response to each individual part of the unresolved items for our review along with your response to the identified items of noncompliance. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter, the enclosures, 1 and your response to this letter will be placed in the NRC's Public Document 1 Room, except as follows. If the enclosures contain information that you or l l ( I i-, , , 29 t/
.: . . ., u u v , r g Q
A*TTACS MDJT 12,
-n
. Consumers Power Company I ' AN 1 2 I? 81 i your contractors believe to be proprietary, you must apply in writing to
. this office, within twenty-five days of the date of this letter, to withhold such inf6rmation from public disclosure. The application must include a full statement of the reasons for which the information is considered proprietary, and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application. I Ve will gladly discuss any questions you have concerning this inspection.
Sincerely,
' Cw
.v.:sh5.-
Q[JamesG.Kepher)b* Q Director
Enclosures:
( l. Appendix A, Notice of Violation
- 2. 1E Inspection Reports No. 50-329/80-32 and No. 50-330/80-33 cc w/encls:
(. Cent ral Files Reproduction Unit NRC 20b PDR Local PDR NSIC TIC Ronald Callen, Michigan Public Service Commission Myron M. Cherry
e
. Appendix A I
NOTICE OF VIOLATION Consumers Power Co. Docket No. 50-329 Docket No. 50-330 , As a result of the inspection conducted on December 8 - 11, 1980, and in accordance with the Interim Enforcement Policy, 45 FR 66754 (October 7, 1980), the following violations were identified:
- 1. 10 CFR 50, Appendix B, Criterion XVI states, in part, that " Measures shall be established to assure ,that conditions adverse to quality such as. . . deficiencies. . .a re promptly . . . corrected. The measures shall assure that the cause...is determined and corrective action taken to preclude repetition."
Consumers Power Co. QA Program, Policy No. 16, corrective action states, l in part, that " corrective action is that action taken to correct and pre-clude recurrence of significant recurrence of significant conditions adverse to the quality of items... Conditions or trends observed or identified which are adverse to quality are considered for corrective action..." ( Tle "FSAR Re-review Procedure" instructions for Block 8 requires that ,
"t he engineering design documents against which the FSAR review package 1 is to be reviewed are listed by the primary review engineer."
CPC0 Audit No. M-01-53-0 states, in part, "the following significant items were revealed by this audit...in many instances not all of the design documents were listed as required by the instructions for per-forming the re-review." Contrary to the above, CPC0 did not initiate preventive action to pre-clude repetition of not identifying design documents for the remaining re-review packages as evidenced by the inspectors review of other FSAR re-review packages which did not include all of the design documents. In addition, interviews with some of the primary reviewers indicated that they were not reviewing the FSAR for technical accuracy against l references at the end of the FSAR chapter as required by the procedure. Based on the above, the adequacy of the FSAR re-review is in question. This is a Severity Level IV violation (Supplement 11).
- 2. 10 CFR 50, Appendix B, Criteria III, states, in part, that " Measures shall be established t.o assure that. .. design bases ...are correctly 4 translated into specifications...and for the identification and control s l
of design interfaces...these measures shall include the establishment ; of procedures...for the review of documents involving design interf aces'." t 1 l
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,y
{
. Appendix A .
f Consumers Power Co. QA Program, Policy No. 3 states, in part that "Each group... performing detailed design translates the applicable regulatory
. requ,irements... design criteria into design documents, such as specifica-tions... pro.cedures. The design organization... establishes and controls the interface with other design organizations.
I
- a. Bechtel EDPI 4.25-1, Section 6.1, states, in part "Each originating design group shali. maintain a log of all documents which are routed to personnel external to the design group. These logs shall be re-tained.. .providing visibility of the projects design interface control.
Contrary to the above, Bechtel Civil Project Engineering group did not maintain a coordination log of specification and specification change notices as evidenced by our review of soils related specifica-tions C-211 and C-210.
- b. ANSI N45.2.11, Paragraph 4.1 requires that applicable design inputs are correctly translated into specifications drawings, procedures or
- l. instructions. In addition, Paragraph 7.0 requires that documents including changes are reviewed for adequacy.
Consumers Power Co.'s 50.54(f) response, Page I-17, Paragraph 4(a) required that specification change notice (SCN)-9004'be issued to
- require a laboratory compaction test to be performed for each field SCN-9004 was initiated on 4/13/79.
( density test. Contrary to the above, Revision 16, dated 8/24/79, to the present Revision 20 of specification C-208 did not correctly translate SCN-9004 as a requirement into the specification. Revision 16 permitted laboratory density tests to be performed at a frequency as determined by the geotechnical engineer rather than for each field density test performed.
- c. ANSI N45.2.11, Paragraph 8.2 requires that design changes be reviewed and approved by the same groups or organizations which reviewed and approved the original design documents. l Consumers Power Co. So.54(f) response, Page 23-11 committed to evise existing design control measures and require design interfaces on design changes. EDPI 4.25.1, Revision 7 added Section 4.2 which states, "It is the responsibility of the originator of a design change to effect ;
coordination of the change with all groups which reviewed and/or used I the original or subsequent revisions of that design document." Contrary to the above, Revision 8 to EDPI 4.25.1 permits the group supervisor to waive the design interface requirement by adding to Section 4.2, "as determined by the group supervisor of the discipline which originated the document." Revision 8 does not establish adequate measures as required by ANSI N45.2.11 or as committed per 50.54(f) response. [ l l j
Appendix A - 3-This is a Severity Level IV violation (Supplement II).
. Pursuant ,to the provisions of 10 CFR 2.201, you are required to submit to this office within twenty-five days of the date of this Notice a written statement or explanation in reply, including for each item of noncompliance: (1) correc-tive action taken and the results achieved; (2) corrective action to be taken I to avoid further noncompliance; and (3) the date when full compliance will be achieved. Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, this response shall be submitted under oath or affirmation.
Dated 3 . ,u .,, r2 r Hr 0 . .. s % lbs./( c_ 7 () James G. l(eppler V I
. '6irector
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND EhTORCEMENT REGION III Reports No.* 50-329/80-32; 50-330/80-33 Docket Nos. 50-329; 50-330 Licenses No. CPPR-81; CPPR-82 , Licensee: Consumers Power Company 1945 West Parnall Road Jackson, MI 49201 Facility Name: Midland Nuclear Power Plant, Units 1 and 2 Inspection At: Bechtel Power Co., Ann Arbor, Michigan Inspection Conducted: December 8-11, 1980
\ Inspectors: E . Gal g r, Region III b R. B Landsman, Region III b bIl J. Gi y h Qua ity Assurance Branch /'7'f/
Reviewed By: R. C.q'nop, Chiefic -de p 7_ y/ Projects Section No. 1 Approved By: G. Fiorelli, Chief Reactor Construction and Engineering Support Branch Inspection Summary Inspection on December 8-11, 1980 (Reports No. 50-329/80-32; 50-330/80-33) Areas Inspected: Consumers Power Company response and implementation of corrective actions regarding the 10 CFR 50.54(f) request of Question 1 of NRC letter dated March 21, 1979 and Question 23, request for additional information dated September 11, 1979. The inspection involved 106 inspector-hours at the Bechtel Ann Arbor office by three h7C staff. In addition, approximately 120 hours of review of the licensee response was performed prior to the inspection. Results: Two items of noncompliance were identified in the above areas inspected - Severity Level IV, Inadequate Design Control with three examples; Severity Level IV, Inadequate Corrective Action; and Unresolved Items identified in Paragraph 3(c) and 3(d). k' mn -
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- DETAILS
( Exit Meeting Attendees at Ann Arbor, Michigan, December 11, 1980 Nuclear Regulatory Commission E. J. Gallagher, Civil Engineer Inspector, IE: Region III R. B. Landsman, Civil Engineer Inspector, IE: Region III : J. W. Gilray, Quality Assurance Branch, NRR Consumers Power Company J. W. Cook, Vice President, Projects, Engineering and Construction B. W. Marguglio, Director, Environmental Services and Quality Assurance W. R. Bird, Quality Assurance Manager, Midland Project D. M. Turnbull, Site Quality Assurance Superintendent G. R. Eagle, Supervising Quality Assurance Engineer G. S. Keeley, Midland Project Manager
- G. E. Clyde, Licensing Engineer H. P. Leonard, Section Head, Quality Assurance Engineer k D. E. Horn, Group Civil Supervisor, Quality Assurance Engineer Bechtel, Ann Arbor Office J. Rutgers, Midland Project Manager J. Milandin, Manager of Quality Assurance
(' L. A. Dreisbach, Assistant Project Manager V. J. Manta, Project Quality Engineer N. Swanberg, Assistant Project Engineer G. L. Richardson, Quality Assurance Manager, Midland Project D. F. Lewis, Licensing Engineer
. R. E. Sevo, Quality Assurance Engineer A. E. Bico, Quality Assurance Engineer R. L. Rixford, Quality Assurance Engineer J. R. McBride, Quality Engineer R. C. Hollar, Quality Engineer
- 1. Background Meetings were held on February 23, 1979 and March 5, 1979 at the NRC Region III office in Glen Ellyn, Illinois to discuss the circumstances associated with the settlement of the diesel generator building at the Midland facility. This discussion was part of the investigation conducted by Region III as documented in NRC Investigation Report No. 50-329/78-20; 50-330/78-20, dated March 22, 1979. Representatives of the NRC staff from headquarters attended the meeting on March 5, 1979. The staff stated that it's concern was not limited to the narrow scope of the settlement of the diesel generator building, but extended to various buildings, utilities and other structures located in and on the plant area fill. In addition, the staff expressed concern with the Consumers Power Company Quality Assurance Progria.
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i I Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, and Section 50.54(f) of 10 CFR Part 50, additional in- ; formation was requested regarding the adequacy of the fill and the 1 '{ quality assurance program for the Midland site in order for the Commission to determine whether enforcement action such as license , modification, suspension or revocation should be taken. Question 1 of the"50.54(f) letter dated March 21, 1979 requested information ; regarding the quality assurance program. On April 24, 1979, l Consumers Power Company submitted the initial response to the 50.54(f) , request, Questions 1 through 22. As a result of the NRC staff review of Question 1, the NRC concluded that the information provided was not sufficient for a complete review. Subsequently, on September 11, 1979 I the NRC issued a request for additional quality assurance information l (Question 23). On November 13, 1979, Consumers Power Company submitted 1 revision 4 to the 50.54(f) responses which included response to Question
- 23. As a result of the Region III investigation report and CPC0 responses, the NRC issued an Order modifying construction Permits No. CPPR-81 and No. CPPR-82, dated December 6, 1979. The latest revision to Consumers l Power Company response to the 50.54(f) request is revision 10, dated l November-21, 1980. i k 2. Purpose of Inspection l The inspection was conducted at the Bechtel Power Company Ann Arbor, l Michigan offices on December 8-11, 1980 to verify implementation of the specific commitments and action items reflected in Consumers Power Company response to 10 CFR 50.54(f) Questions 1 and 23 with the exception
( of those areas where completion of commitments has not been satisfied as of this time. l l l The inspection was divided into the following areas: I
- a. A review of CPCo response to Question 1, Part (a) and Question 23,
! Part (1) regarding the identification of the specific quality assur;:nce deficiencies that contributed to the soils problem, including the root cause of the deficiency, remedial action in the soils area, the programmatic and generic corrective actions , as committed to in the response. J l l
- b. A review of CPCo response to Question 1, Part (b) and Question 23, f Part (2) regarding the provisions to be implemented to preclude areas of contradictions between the PSAR, FSAR and design docu-ments.
- c. A review of CPCo response to Question 1, Part (c) and Question 23, Part (3) regarding the programmatic and generic corrective actions to provide confidence that quality assurance deficiencies do not .
(or will not) exist in other areas. 1 The following sections of this report discuss the results of the review of the above areas of CPCo response to Questions 1 and 23. l . l l l l l ,
d Review of Question 1, Part (a) and Question 23, Part (1) 3. ( The identification of quality assurance deficiencies that contributed to the soils problem was discussed in Question 1, Part (a) and Question 23, Part (1). Concumers Power Company identified the root cause'bf the deficiencies, the remedial measures in the soils area,
, and the programmatic and generic corrective action to preclude further i recurrence of the deficiencies. CPCo complied a list of specific :
action items that would have to be accomplished in order to satisfy the commitments made in response to Questions 1 and 23 of the 50.54(f)
- request.
Attachment No. I provides an action item tracking system which includes i the action item description and reference and the status and documenta-tion verified by the NRC during this inspection. Those action items for which CPCo commitments have been accomplished are identified as being " closed"; items identified as "open" either have not been completed by CPCo or the action taken was considered insufficient. I Question 1 provided 26 action items of which the NRC verified 18 had been satisfactorily accomplished while 8 remain open. Question 23 provided 57 action items of which 34 were determined to be satisfac-torily accomplished while 23 remain open. The following are NRC findings regading the implementation of certain j ' CPCo commitments. I a. i Action items 23-5 and 23-38 as identified in Attachment No. 1 provided commitments to examine current procedures and practices for the preparation and control of the FSAR in view of past experiences. CPCo committed to procedural changes to existing engineering department procedures. Seven Bechtel procedures were examined and revised to clarify design control procedures for the FSAR. -Engineering Department Procedure Instruction (EDPI) 4.25.1, Design Interface Control, was revised by Revision 7 by including section 4.2 which states, "It is the responsibility of the originator of a design document ; change to effect coordination of the change with all groups which reviewed and/or used the original or subsequent revisions of that design document." Subsequently, Revision 8 to EDPI 4.25.1, changes the above by adding to the end of the. statement, "as determined by the group , supervisor of the discipline which originated the document." The j , originator of Revision 8 stated that the intent was that only technical changes have to be interfaced while editorial changes , would not necessarily require this interface control. The pro- I
-cedural change, however, does not reflect the intent and permits 4
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the group supervisor to waive interface control for any changes as evidenced by inspection finding in Paragraph 3(b) of this . ,[ report. The engineering procedures EDPI 4.25.1 does not satisfy CPCo commitment made to the NRC in response to Question 23, subsection 3.3, page 11 and identified as action item 23-5 of Attachment No. 1. This failure to provide adegaate design interface control is , considered contrary to 10 CFR 50, Appendix B, Criterion III as described in the Notice of Violation. (50-329/80-22-01; 50-330/ 80-33-01).
- b. Engineering Department Procedure Instruction, EDPI 4.25.1, Section 6.1 requires that, "each originating design group shall maintain a log of all documents which are routed to personnel external to the design group. These logs shall be retained . . .
providing visibility of the projects design interface control." It was determined based on a review of specification C-208, Revision 20, Materials Testing Services, Section 9, Soils Testing and C-211, Revision 12, Technical Specification for Backfill, that the civil project engineering group is not maintaining a complete (- coordination log of specifications and specification change notices. Interviews with cognizant Bechtel personnel indicated that it is up to the originator of the document to transmit the design docu-ment to the coordinator clerk to log it in as being interfaced with the appropriate groups. It was determined from reviewing the 3 [
\ interface log that the originator of the documents are not aware of this requirement and documents are not being interfaced with other design groups as required by the procedure. In addition, Regulatory Guide 1.64, Quality Assurance Requirements for the Design of Nuclear Power Plants and ANSI N.45.2.11-1974, Section 10 requires design interface records to be maintained.
This failure to maintain design interface and coordinator control is considered contrary to 10 CFR 50, Appendix B, Criterion III as described in the Notice of Violation. (50-329/80-32-02; 50-330/ r 80-33-02). )
- c. Specification C-208, Revision 10, Section 9 regarding soil testing requirements was reviewed for technical content. It was determined that the specification was not adequate as written. The following j specific findings were identified.
(1) CPCo was identified in Question 1, Appendix I, Page I-13, Paragraph A.4(a) that the subcontractors test procedures fer soil testing service were inadequate; specifically, U. S. Testing procedures did not provide for developing and updating a family of proctor curves used to compare in place field density tests to maximum laboratory standards. CPCo committed to the remedial action on Page I-17, Paragraph 4(a) ( . l
i l I i which states, " Selection of proctor curves will no longer be a problem because each field density test will be accompanied *
- ( by a separate laboratory standard compaction test which will provide a direct comparison." It was also stated that SCN-9004, dated April 13, 1979 was issued to require the above.
It was determined that SCN-9004 was issued as committed; however, during Revision 16, dated August 24, 1979, of g specification C-208, the civil project engineer failed to include the above requirement and instead revised Table 9-1 to permit the frequency of the laboratory test to be "as directed by the on-site geotechnical engineer" rather than
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for each field density test. This does not comply with the commitment made in 50.54(f) response to Question 1. This occurred because adequate design interface controls had not been implemented as required by ANSI N 45.2.11. There was no evidence that the geotechnical group had reviewed or approved the revision to the specification. This failure to provide adequate design interface control is I considered contrary to 10 CFR 50, Appendix B, Criterion III as described in the Notice of Violation. (50-329/80-32-03; 50-330/ 80-33-03). (2) Specification C-208, Section 9.1.1 should be reworded to remove confusion which exists about the word " compaction". ( This section should read: Modified proctor tests on cohesive material shall be performed in accordance with ASTM D 1557, Method D. (3) Section 9.1.3 (first paragraph) does not specifically indicate how ASTM D 1566 has been modified by USBR DES E-24. In addition, why does the specification prohibit the use of the nuclear density device for measuring in-place field density? This device is ac industry accepted method with a standsrd ASTM designation. 4 (4) Section 9.1.3 (second paragraph) assumes a specific gravity of 2.75. The actual specific gravity should be known and used as is the industry practice. (5) Section 9.1.3(c) should also include: if the results still plot to the right of the ZAV curve the test should be rejected and a new density test performed. (6) Section 9.1.3(d) uses the phrase 101% compaction. This should read 101% of maximum proctor density. This section also permits the on-site geotechnical engineer "to evaluate" the results of tests that exceed 101% proctor density for cohesive material and ( a
l l i l 105% for cohesionless material. This section should include j
, the qualitative acceptance criteria and/or instructions to be !
} .used for the basis of this evaluation. The above items 3(c) 2, 3, 4, 5 and 6 are considered unresolved
** items pending a review of CPCo response to each item. (50-329/
0-32-04; 50-330/80-33-04).
- d. Specification C-211, Revision 12 regarding backfill work activities o was reviewed for technical content. It was determined that the specification was not adequate as written. The following specific items were identified.
(1) Section 8.1 does not specify the type of material to be used beneath Category I, safety related structures. This should be ir.cluded in this specification. (2) Section 8.1.1 does not specify the type of material to be used around pipes and duct banks. The specification should specify or. refer to appropriate instructions. I (3) Section 8.3.2 (third paragraph) states, "the uncompacted lift thickness of the backfill material shall be determined by the on-site geotechnical soils engineer . .. " The on-site soils engineer should not have to determine the lift thickness when Attachment No. I to specification C-211 specifies the requirement for each type of equipment based ( on equipment qualification tests. (4) Section 8.5.2 permits the use of rubber-tired rollers to compact structural backfill and sand. Attachment No. I to specification C-211 does not indicate rubber-tired rollers
. as having been qualified and rubber-tired rollers should not be used to compact structural backfill and sand.
The above items 3(d) 1, 2, 3 and 4 are considered unresolved items pending a review of CPCo respor,se to each item. (50-329/80-32-05; 4 50-330/80-33-05).
- 4. Review of Question 1, Part (b) and Question 23, Part (2)
The provisions and the procedures to be implemented to preclude conflicts between PSAR, FSAR and design documents was discussed in response to Question 1, Part (b) and Question 23, Part (2). Consumers Power Company included in their response a procedure entitled, "FSAR Rereview Procedure"
,to be implemented to accomplish this commitment.
Action items 23-1, 23-44 and 23-44(a) as identified in Attachment No. 1 provided the commitments to be implemented to assure FSAR accuracy. The following are the NRC findings regarding the implementation of these commitments. l
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l It was deterr.ined that, in general, consultant reports were not attached l to the FSAR. However, the complete text of a consultant report prepared , .f, by Weston Geophysical Engineering Company was found as an attachment to the FSAR and included in the FSAR, as Appendix 2C. -Therefore, the CPCo response which states, " Consultant reports were not attached to the FSAR, but pdttions of consultant reports were extracted and incorporated into the FSAR text itself" (re: Question 23, Page 23-7) is not correct. CPCo also stated that the FSAR was rereviewed against design documents such as consultant reports for conflicts. It was determined that verification of portions of consultant reports incorporated into the FSAR have been adequately reflected in design documents has not been satisfactorily accomplished. FSAR Rereview Pro-cedure, Revision 1, dated March 13,_1980, Subsection 2.1.3 states that each FSAR section should be carefully reviewed against design documents
. . .'as a minimum, the following should be checked . . . references at the end of the FSAR chapter. The procedure also requires in Item 8 that engineering design documents against which the FSAR review package is to be reviewed are to be listed by the primary review engineer in .
g Block 8 of the FSAR rereview form. A review of FSAR packages Nos. 9474, 9473, 9472, 9471, 9096, 9097 and 9098 indicates that no design documents other than a few drawings were identified and listed. Numerous reports were referenced throughout the FSAR text of these sections, however, they were not recorded as required in Block 8 as being reviewed for consistency with the FSAR text. An interview with a Bechtel cognizant primary review engineer indi-cated that he physically checked the references to make sure that they agreed with the FSAR text. Subsequently, after the NRC inspector found an apparent discrepancy between the FSAR text and one of the references, the Bechtel reviewer indicated that he did not check the text of the references, but merely checked the reference for consistency of subject matter, i.e., title vs. sentence content not technical substance vs. FSAR statements. Another cognizant Bechtel primary review engineer indicated he could not check references in his section because he was not qualified to review the technical matter in this area. He indicated that
- i. he relied on the Bechtel Geotech group (the interface reviewer in Block 11) to verify the references. Discussions with a Geiotech reviewer indicated he did check reports for consistency with the FSAR, but did not list them in Block 8 as required. ,
After this was determined, the inspector was informed that a CPCo j interim audit No. M-01-53-0, dated March 1980, identified the same ' problem concerning the lack of identifying design documents in Block 8 of the FSAR review form. At this time approximately 600 of
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a total of 900 FSAR rereview packages had been completed. However, no corrective action was taken. CPCo final audit of this activity, audit No. M-03-202-0, dated November 1980, once again identified an unresolved item, URI-3, regarding this same problem. The FSAR rereview is cow complete and the unresolved item was pending resolution as of the date of this inspection. ( Cognizant individuals indicated that one of the reasons why documents ,
, were not listed in Block 8 was because there was not sufficient space.
( ' An interview with the preparer of the FSAR rereview document indicated that the intent of Block 8, and it's instructions, was to list all of the design documents to which the FSAR section was reviewed against in order lo assure there were no more conflicts between design documents and the FSAR text. I Based on the above, it was determined the CPCo failed to provide adequate corrective action with regard to the identified audit results. This is considered contrary to 10 CFR 50, Appendix B, Criterion XVI, as described in the Notice of Violation. (50-329/80-32-06; 50-330/80-33-06). Due to this finding, CPCo implementation of the specific commitment as discussed in response to Question 23, Part (2) has not been accom-plished and the adequacy of the FSAR rereview which has been completed is questionable.
- 5. Review of Question 1, Part (c) and Question 23, Part (3) g-CPCo and Bechtel have performed a detailed re-review of specifications, installations, and construction inspection plans, procurement documents, inspection and test procedures, including the results of inspections and tests to determine the completeness and accuracy of documents and the acceptability of hardware. In this regard, the I&E inspection activities involved a review and evaluation of activities associated with the above re-review actions and included discunnions with main
(, participants in the re-review effort. The following is a summary of this inspection.
- a. CPCo and Bechtel were able to demonstrate that an extensive re-review of specification, inspections and test procedures, and documents associated with procurements were conducted with mean-ingful results. The documents were evaluated by CPCo and Bechtel to assure that the necessary tolerance call outs and quality requirements were specified; that the qualification requirements i were adequately called out and met; that there were sufficient specificity provided in the documents; and that there were the necessary inspection requirements specified. In addition, the completed documentation was evaluated to detennine that technical I and quality requirements were met in an acceptable manner. l l
Areas that were found deficient resulted in revision and improve-ment to procedural controls and specifications. Hardware suspected of not meeting quality requirements were re-evaluated by engineering and quality assurance to determine their accept, repair, or reject status. Throughout this particular I&E inspection effort, specifications, procedures, and instructions were reviewed and a determination made that revisions and improvements were accomplished. (
- b. The improved trend analysis and corrective action program estab-lished by CPCo and Bechtel was evaluated and found acceptable. -
( It is expected that this program will prove effective in detecting major weaknesses in the early stages such that meaningful, prompt corrective actions can be initiated during the design and chnstruction phase,
- c. The " flag program," which provides assurance that problems, similar ;
to those experienced with reactor vessels holddown anchor bolts, do not exist in other similar procurement actions where in process source inspection activities are involved, was evaluated. Purchase orders and receiving documentation were reviewed by Bechtel to determine that critical design.and specification requirements were properly carried out and where questions were raised concerning product function, a " flag" was identified to the concern requiring further evaluation, discussions, and resolution by engineering and quality assurance. Evidence showed this activity to be productive and in accordance with documented instructions.
- d. The 1978 and 1980 independent audit results performed by the Management Analysis Corporation on CPCo and Bechtel were evaluated
( and found in accordance with program requirements. Overall, the personnel contacted conveyed their QA knowledge and their sincerity and dedication towards performing the activities described above. However, as a result of the findings identified during this inspection, it is clear that more emphasis must be (' placed on the attention to detail in the preparation and review of documents. In order to accomp1'ish this, upper management must play a more active role in conveying this principle to the working staff and observing attitudes and activities to assure QA principles and attention to detail are being properly carried out. Unresolved Item Unresolved items disclosed during the inspection are discussed in Para-graph 3(c) and 3(d) of the report. Exit Meeting The inspector met with licensee and contractor representatives at the conclusion of the inspection on December 11, 1980 and summarized the inspection scope and findings. The items of noncompliance identified during-the inspection were discussed in detail. The licensee acknowl-edged the inspection results.
Attachment:
Attachment No. 1 (.
- n 3 t 1 of 21 ATTACHMENT h. R1 in/81 ACTION ITEMS PROGRAMMATIC AND GENEftIC CORRECTIVE ACTIONS COMMITTED'TO IN T!!E RESPONSE TO QUESTION 1, PART (a)
AND IN T!!E RESPONSE TO QUESTION 23, PAllTS (1) AND (2)
' Action Action Item Description Actions Verified Item (Status) buring NRC Inspection Number and Reference 23-1 Consultant reports other than Dames & Moore were considered in accordance with the guidelines provided in NRC Regulatory Guide 1.70, Itevision
- 2. Consultant reports were not attached to the FSAR, but portions of consultant reports were extracted and incorporated into the FSA't text itself. Those portions incorporated into the FSAR become commitments. Therefore, disposition of recommendations in consulting reports has been adequately accounted for in the prepara-tionofthePSAp.
Verificationthhtthoseportionsofconsultant (open) Refer to Action Item 44 for Review of reports determined to be commitments and incor- FSAR Re-review porated into the PSAR have been adequately reflected in project design documents' is being acccmplished via the FSAR rereview program described in the_ response to Question 23, Part (2). __________________________ The two Bechtel OA audit findings reported in (Closed) Reviewed quality assurance audit 4.0-23-1(a) special 1, "SAR change control", & audit findings and our April 24, 1979, response (Paragraph D.1, A-34 & A-35. The audit was performed to assure 1 - 11 Page I-8) have been closed out. The results that there is a system to assure design changes of this audit are being utilized in the FSAR are reflected in the FSAR. Audit findings ident-
. Control system Study committed to in Subsection ified casen where design changes were not reflected 3.3 of this response to Part (1),
in the FSAR. Corrective action resulted in a review of all design requirement verification (Question 1, Appendix I, Section D.1, Page I-8 checklists (DRVCL's) for groups identified with Question 23, Subsection 3.1, Page 7) Problems. This review is documented in QE monitoring report DRVC-8.
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1/5/81 l l Action Action Item l Item Description . I Number and Reference * *
. (Status) 23- 2 On April 3, 1979, Midland Project Engineering (closed) Reviewed & verified memos & letters Group Supervisors in all disciplines were instructing proj. engr. field eng, & QC of pro-reinstructed that the only procedurally cedure for impicmenting clarifjcation or change correct methods of implementing specification to approved drawings or specifications:
changes are through the use of specification (1) Bechtel memo to QCE's, dtd 5/30/79. revisions or Specification Change Notices. This (2) Bechtel memo to Field Engr's, dtd 3/28/79. was followed by an interoffice memorandum from (3) CPCo letter to Bechtel, dtd 3/12/79. l the Project Engineer to all Engineering Group (4) Bechtel memo to Proj. Engr, dtd 3/21/79. l Supervisors on April 12, 1979. (5) Bechtel memo to Group Suprv, dtd 3/12/79. (6) Bechtel letter to CPCo, dtd 6/5/79. (Question 23, Subsection 3.2, Page 8; and Subsection 3.9, Page 24) 23-3 Engineering Departr'ent Project Instruction (Closed) Reviewed & verified EDPI 4.49.1, Rev. 4, and 4.49.1 was revised in Revision 2 to state, " specification change notice # to include require-1 - 12 "Under no circumstances will interoffice :,ent that 10M's, memo's , telex's, TWS's, etc. can memo'.anda, memoranda, telexes, 'NXs , etc be used to change the requirements of a not be used to change spec. requirements. A spec. spec ific a tion. " change notice must be issued in order to change spec. requirements.
' Question 1, Appendix I, Section 0.2.d, Page I-8 Question 23, subsection 3.2, Page 9, and Subsection 3.9, Page 24)
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~,.- - c 3 of 23 1/5/81 ActDan Action Item
- Ittm Description Ny be r and Reference (Status) 23-4 A review of interof fice memoranda, memoranda, telexes, 'IWXs, and other correspondence relating '
to specifications for construction and selected *- procurements of 0-listed items will be initiated. The purpose of the review will be to identify any clarifications which might reasonably have been interpreted as modifying a specification requirement and for which the specification itself was not formally changed. An evaluation will be trade to determine the ef feet on the technical acceptability, safety implications of the potential specifica tion modifica tion, and any work that has been or may be affected. If it is determined that the interpretation may have af fected any completed work or future work, a formal change will be issued and remedial action necessary for product quality will be taken in accordance with approved procedures. The foregoing procedure will be folldwed for all (Closed) Verified Bechtel memo dated 12/20/79 specifications applying ~to construction of (File 0455) which.provides the procedure for Q-Lis ted items, review of all (100%) Q-listed construction type For specifications concerning the procurement spec's. and sampling plan procedure for procurement type spec's, of 0-Listed items, the foregoing procedure will be implemented on a random sampling basis. The sample size has been established and the specification selection has been made. 4 (21) Review and acceptance criteria for the specifi- (0 pen) ca tions have been defined. Review criteria has been established (see above action item 4); acceptance criteria was not (47) The review of construction and selected defined. Audit report M01-200-0 also identified procurement specifications is scheduled to be this as an unresolved item. completed by April 1, 1981. (0 pen) File had no review data for construction type or procurement type spec's.
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Act i'J r. Actio. ttm I/5/81 Item Description - Number and Reference (Status) . (47) If the acceptance criteria are not met, the (0 pen) Preliminary indication per Bechtel (c nt'd) review will be expanded to include other Representative indicated that the review will specifications for 0-listed items. At that be required to be expanded to include other time, a revised completion date will be spec's than sampling plan identified. established. 8 (Question 23, subsection 3.2, Page 9, and Subsection 3.9, Page 25) 23-5 A study was completed which examined current (23-38) procedures and practices for the preparation and control of the PSAR in view of these experiences. Procedural changes have been initiated by the revision of or addition to the Engineering Department Procedures.
. (Question 23, subsection 3.3, Page 11) 23-6 An' interoffice memorandum dated April 12, 1979, (closed) Reviewed & verified inter-office memo was issued by Geotechnical Services to alert from S. Blue to Geotech personnel, dated 4/12/79 personnel of the need to revise or annotate which requires that changes in design be reflected calculations to reflect current design status. . in the original calculations & to reflect proper interdepartmental coordination has been achieved.
(Question 23, Subsection 3.4, Page 13) s 23-7 Field Instruction FIC 1.100, "O-Listed Soils (closed) Reviewed- 6 verified, field instruction Placement Job Responsibilities Matrix," has been FIC1.100, Rev. 3, dated 8/15/80 to include daily prepared and establishes responsibilities for job responsibilities of the onsite geotechnical performing soils placement and compaction. engineer. (Question 23, subsection 3.6, Page 18; Subsection 3.7, Page 20; and Subsection 3.11, Page 30) e e,
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,/ t 5 of 21 1/5/81 Action Action Item Item Description ~
Number and Reference (Status) e 23-7A Review Field Procedure FPG-3.000 to ensure (Closed) Verified that FPG-3,000, Rev. O," Job-and clarity and completeness responsibilities for field engineers"was reviewed 1 - 17 as a result of this review FIC 1.100," Job responsi (Question 1, Appendix I, Section 0.2, Page I-ll) bilities for the onsite geotechnical engineer"was established. ,, 23-8 Construction specifications, instructions, and and procedures were reviewed to identify any other (open) CPCo commitment not completed. 1 - 16 equipment requiring qualification which had not yet been qualified. No such equipment was identified. (Question 1, Appendix I, Section D.1, Page I-ll-Question 23, subsection 3.6, Page 18) 23-9 A dimensional tolerance study wac completed (Closed) Verified that dimensional tolerance using the reactor building spray pump and study was performed on the reactor building spray ancillary system as the study mechanism. pump system. (Question 1, Appendix I, Section D.2.b, Page I-8) 23-10 Engineering reviewed specifications not previously (closodi verified that a review of spec's A-17, and reviewed for the specificity or tolerance studies. C-67, M-342, C-208, C-231 & A-41 was performed 1-5 for specificity & tolerances. Revisions were (Question 1, Appendix I, Section D. 2.c, Page I-8 ) made to spec. as needed. 23-11 A specific review of the FSAR and specification (cInsedi Verified a review of FSAR & Spec 1 requirements for qualification of electrical requirements for the qualification of electrical and mechanical components has been made as part and mechanical components has been performed
& documented in CPC0 letter to NRC, Region III of the corrective action relating to CPCo's 50.55(e) report on component qualification.
d ted December 5, 1980, as required by 50.55(e) reporting requirements.
.(Question 1, Appendix I, Section D.2.e, Page I-8)
(open) CPCo commitment not completed, 23-12 Quality Assurance will schedule yearly audits of the design calculational process for techniques and actual analysis in each of the design disci-plines. (Question ', Appendix I, Section D.4, Page I-8)
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. sneet f 21 1/5/81 Action Action Item .
Item Description , Number and Reference (Status) ' 23-1 3 Audits of ITT Grinnell hanger design and CPCo (Closed) Verified that audit OT-ITT Grinnel relay setting calculation have been conducted. (April 5, 1979) and audit of electrical and I&C calculations (June 26, 1979) was performed. (Question 1, Appendix I, Section D.4, Page I-8) 23- 14 Bechtel Project Engineering will review design (closed) Reviewed file No. 54601-54618 and drawings for cases where ducts penetrate (calc #41-1) dated 9/5/78 which identifies each 1 - 10 vertically through foundations. The possibility
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duct bank in the plant and interface with any of the duct being enlarged over the design buildings. Results of study were documented in requirements and the effect this enlargement memo from L. Curtis to R. Rixford dated 5/27/80 may have upon the structure's behavior will be which indicates no other safety-related structure evaluated by Jurm 1, 1979. Proper remedial except D. G. Bldg was effected by an interface measures will be Laken if the inves tiga tion with duct banks. Provisions were made to allow shows potential problems. independent vertical movement between the diesel
- . generator b1dg and duct banks.
(Question 1, Appendix I, Section C.S.b, Page I-7) 23-1 5 An in-depth audit of U.S. Testing operations, (Closed) Reviewed and verified audit 25-2-7 of and covering testing and implementation of their U.S. Testing Company was performed on April 25-26c 1 - 20 QA program will be conducted in late April or 1979. early May 1979, by Dechtel Project QA and Engineering. ., (Question 1, Appendix I,' Section C.4.b, Page I-18; Section D.3.c, Page I-18) 23-1 6 An in-depth training session will be given to (open) See review of Action Item 23-17 and Midland QA Engineers covering the settlement 1 - 25 problem and methods to identify similar conditions in the future.
, (Question 1, Appendix I, Section D.1.b, Page I-22) w P _- __
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-Action Action Item
- Item Description Number and Itef erence (Status) 23 An in-depth training session will be given to (open) Reviewed IOM dated July 27, 1979 and and all CPCo and Dechtel OA Engineers and Auditors June 4, 1979 documenting training to CPCo and 1 - 25 to increase their awareness of the settlement Bechtel QA personnel on Midlahd plant fill problem and to discuss auditing and monitoring experiences. The file.does not contain docu-techniques to increase audit ef fectiveness. mentation of the contents or detail of the training nor any material handed out to parti-(Ouestion 1, Appendix I, .Section D.2, Page I-22) cipants for their future reference.
23-18 An in-depth review of the Bechtel trend (open) CPCo commitment not completed. and program data will be undertaken by Dechtel OA 1 - 24 management to ensure the identification of any other similar areas that were not analyzed in sufficient depth.in the past reviews. (Question 1, Appendix I, Section D.l.a, Page I-22) . 23-19 Quality Control Instructions have been evaluated and (Closed) verified the QCI's were reviewed and to ensure that the documentation characteristics items requiring further action and resolution 1 - 21 which are to be inspected (i.e., surveillance and identified (See Action Item 23-19A). 4 1 - 22 review callouts) are clearly specified. 23-19A (This action modified to include necessary revi- (0 pen) Completion of required changes to QCI's and sion to OCIs resulting from evaluation of surveil-per Action Item 23-19 have not been completed, f 1 - 21A lance and review callouts.) (Ouestion 1, Appendix I, Section D.3.a, Page I-18
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and Section D.1, Page I-18) ' ! 23-20 Field Instruction 1.100 has been supplemented (Open) CPCo commitment not completed. Records by establishing requirements ft: demonstrating identifying equip. capability not documented in equipment capability, including responsibility Action Item file, for equipment. approval, and providing records
' identifying this capabili ty.
(Question 23, Subsection 3.6, Page 18) 23-21 See Action Item Number 4 (21) (open) Acceptance criteris not defined (See Action Item 4 for review).
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1/5/81 l Acticn Action item , { Item - Description Number and Reference (Status) ;
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,. 23-22 Guidelines for surveillance of testing operations (Closed) Responsibilities for on-site have been developed and included in Field In- Geotechnical Engineer have been established .per structions for the onsite Soils Engineer. FIC 1.100, Rev. 3 which include requirements. 4 Engineering /Geotechnical Services has developed ,, the guidelines. ,, (Question 23, Subsection 3.10, Page 27) 23-23 Engineering has revised Engineering Depart- (Closed) Verified EDP 4.22 has been revised by i and ment Procedure 4.22 to clarify that Engineering issuance of MED 4.22, Rev. 6 to include Regulatory i 1-3 personnel preparing the PSAR will follow the Guide 1.70 which requires consultant reports to requirements of Regulatory Guide 1.70, Revision 2, be referred with specific commitments included in
" Standard Format and Content of Safety Analysis text of the FSAR.
Reports for Nuclear Power Plants" (September 1975) . Specifically, Regulatory Guide 1.70 (Pages iv and 1 v of the Introduction) . requires that-such consul- + tant reports only be referenced with the applicable commitments and supporting informa-tion included in the test (third paragraph, Page v). Such a requirenent precludes repetition . of this circumstance. - (Question 23, Subsection 3.1, Page 7 and Subsection 3.3d, Page 46) 23-24 To preclude any future inconsistencies between (Closed) Verified EDP 4.1.1, Rev. 2, Preparation j the FSAR and specifications, Engineering Depart- of the design requirement verification checklist',' 4 ment Project Instruction 4.1.1 has been revised Para. 3.1 requires the discipline engineer who to state that all specification changes, rather originates a design change document to fill out 4 than just " major changes," will be reviewed for a DRVC as the change is developed. The DRVC 4 consistency with the FSAR. include verification of consistency with the FSAR for design changes. l, (Question 23, Subsection 3.3, Page 11) 2
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Action Item Item Description Number and Reference (Status) 23-25 Quality Assurance has issued a Nuclear Ouality (Closed) Quality Assurance.nolicy, Section II, l Assurance Manual amendment to clarify the No. 2," design control proceduresT Para. 3.1.4, requi rement that procedures include measures for Rev. 28 states, engin*eering department procedure qualifying equipment under specified conditi'sns, shall include criteria for specifyin; equip. oualification requirements. Also construction (Question 23, Subsection 3.6, Page 18) qua.lity program, Section IV, No. I, Rev." 2B ! 23-26 In view of Action Item 6, Geotechnical Services Para. 3.2.3(P) requires instructions for quali-fications of equip. l has revised Procedure FP-6437 to require that l calculations be annotated to reflect current (Closed) Reviewed and verified procedure design status. FP-6437-A2 was issued (See ref. Ictter from (Question 23, Subsection 3.4 / Page 13) S. Blue to R. Rixford dated 4/10/80). 23-27 Engineering Department Procedure 4.37 has also been revised to require that calculations be a (Closed) Verified procedure MED 4.37, Rev. 11, annotated to reflect current design status. Design Calculation" and EDPI 4.25.1, Rev. 7,
" Design Interf ace Control"was issued to require tha (Question 23, Subsection 3.4, Page 13) originator of a design change to notify all groups which used the original design document and to 23-28 Civil / Structural Design Criteria 72202C-501 check the latest design info prior to revising has been modified to contain the requirements calculations.
that a duct bank penetration shall be designeg (closed) Verified civil design criteria C-501, to eliminate the possibility of the nonspecific size duct interacting with the structures. Rev. II, Para. 6.6 has been added which states,
,,All interfaces between b1dg's or foundations and (Ouestion 23, subsection 3.5, Page 15) duct banks designed after Jan. 1, 1980 shall be included on civil design drawings and shall indi-23-29 The civil standard detail drawings have been cate clearances or const. restrictions as requirei revised to include a detail showing horizontal to account for differential settlement, seismic and vertical clearance requirements for duct movement, etc.
bank penetrations. The detail addresses any mud ma t restrictions. (Closed) Verified civil standards and misc. concrete details, sheet 2, dwg C-141, Rev. 6 (Question 23, Subsection 3.5, Page 15) detail 12 provides duct bank clearance criteria. l l
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Action Action Item Item Description Number and iteference (Status) 23-30 Engineering clarified specifications and (39) (closed) Verified spec C-211T Rev. 12, Construction prepared procedures (governing Para. 8.5.1 (compact Lon equipment) now requires the soils compaction equipment) to implement proposed compaction equipment to be qualified to the requirements of the Nuclear Quality Assurance demonstrate compaction can be achieved at a Manual as stated in Action Item 25. specified 11ft thickness, number of passes, speed (Question 23, Subsection 3.6, Page 18) of equipment and frequency of vibration for vibrating equip. 23-31 Design documents, instructions, and procedures for those activities requiring inprocess controls (open) CPCo commitment not completed. have been reviewed to assess the adequacy of existing procedural controls and technical direction. Engineering review has been completed. (Ques tion 1, Appendix I, Section D.2, Page I-11; and Question 23, Subsection 3.7, Page 20; and Subsection 3.11, Page 30) 23-32 Guidelines for surveillance of testing operations (Closed) See Action Item 23-22. have been developed and included in field Instruc-tions for the onsite Soils Engineer. Engineering / Geotechnical Services has developed the guidelines and Field Engineering has prepared the instructions. (Question 23, subsection 3.10, Page 27) 23-33 The Quality Assurance audit and monitoring program will be revised to emphasize and increase attention (open) CPco commitment not completed. to the need for evaluating policy and procedural adequacy and assessment of product quality. A specialized audit training program will be developed and implemented to ensure guidance for this revised approach. (Question 23, Subsection 3.13, Page 35)
A A -le e -- . .m bhret 11 o 1/5/81' O Action Action Item Item Description Number and Reference (Status) 23-34 Control Document SP/ PSP G-6.1 has been revised (closed) Verified Procedure.Q-6.1, Rev. 5 bas and to provide requirements for inspection planning been revised to inclu,d,e requirements for planning, 1 - 23 specificity and for the utilization of scientific specificity (Para. 3.3.2) and utilization of sampling rather than percentage campling. scientific sampling (Para. 3.3.3.a.8). This deleted surveillance type inspection and now (Question 1, Appendix I, Section D.5.f, Page I-20; and equires inspection by witness or test. Question 23, Subsection 3.8, Page 22; Subsection 3.9, Page 24; Subsection 4.2.2, Pa;e 59) f( 23-35 Control Documents SF/ PSP G-3.2, " Control of (closed) verified G-3.2, Rev. 6, Control of Nonconforming Items," and Nonconforming Items'and QAPP C-101"QA Trend 23-36 OADP C-101, " Project Quality Assurance Trend Analysis"have been modified to provide for and Analysis" have been revised to provide an identifying repetitive nonconforming conditions. 1 - 24 improved definition of implementing require- Interviewed Mr. T. K. Subramanian. 1 - 25 ments for identifying repetitive nonconforming condi tions . (Ouestion 23, subsection 3.12, Page 33) 23-37 Consistent with the intent of Action item Numbers (open) CPCo commitment not completed. 35 and 36, Quality Assurance will review noncon-formance reports which were open as of November 13, 1979, or became open prior to implementation of the improved Project Quality Assurance Trend Analysis program as stated in Action Item 36. This review will be to identify any repetitive nonconforming condi tions perta tning to product type or activity, or pertaining to nonconformance cause. (Question 23, Subsection 3.12, Page 33)
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' 1/5/81 Action Action Item
- Item Description Number and Reference (Status) 23-38 A study was completed which examined current (23-5) procedures and practices for the preparation and (open) See Action Item 23-5 ,
control of the FSAR in view of these experiences. Procedural changes have been initiated by the revision of or addi tion to the Engineering Department Procedures. (Ouestion 23, Subsection 3.3, Page 11) 23-39 Engineering clarified specifications and (Closed) Verified FIC 1.100, Rev. 3 requires (30) Construction prepared procedures (governin9 the soils compaction equipment) to implement on-site geotechnical engineer to ensure compaction equipment is qualified and listed in the spec and the requirements of the Nuclear Ouality Assurance can deliver required degree of compaction. Manual as stated in Action Item 25. (Question 23, subsection 3.6, Page 18) 23-40 Design documents, ins tructions, and procedures for those activities requiring inprocess controls (Open) CPCo commitment not completed. will be reviewed to assess the adequacy of existing procedural controls and technical (31) direct ion. Engineering. review has been com-olete4, and Field Engineering and quality cont.ol review is scheduled for completion by February 27, 1981. (Ouestion 1, Appendix I, Section D.2, Page I-11; Ouestion 23, Subsection 3.7, Page 20, and Subsection 3.11, Page 30)
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1/5/81 Action Action Item . Item Osscription Number and Reference (Status) 23-41 OCIs in use will be reviewed to ascertain that (Open) cPco commitment not eompleted. provisions have been included consistent with the revised control document, SF/ PSP G-6.1, ,,
" Quality Control Inspection Plans."
l (Question 1, Appendix I, Section D.1, Page I-18; Question 23, Subsection 3.8, Page 22; and Subsection 3.9, Page 24) 23-42 Design documents, instructions, and procedures for those activities requiring inprocess controls (Open) CPCo commitment not completed. will be reviewed to assess the adequacy of l existing procedural controls .and technical (31) direction. Engineering review has been completed, (40) and Field Engineering and quality control review is scheduled for completion by February 27, 1981. Any revisions required will be completed by April 17, 1981. (Ocestion 1, Appendix I, Section D.2., Page I-11; Ouestion 23, Subsection 3.7, Page 20; and Subsection 3.11, Page 30) 23-43 The impact of Action Item 41 on completed work (Open) CPCo commitment not completed. will be evaluated, and appropriate actions will be taken as necessary. (Question 23, Subsection 3.8, Page 22; and Subsection 3.9, Page 25) 23-44 FSAR sections have been rereviewed as discussed (Open) 9 re-review packages were reviewed in the Response to Question 23, Part (2). by the NRC. Not all of the design documents were (Question 23, Subsection 3.1, Page 7; listed in Block 8 of required form per procedure Subsection 3.3, Page 11; for performing review issued 3/13/80. This was Subsection 3.2, Page 41; and identified as an item of noncompliance as dis-Section 4.0, Page 47) cussed in Paragraph 4 of this report. 1
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.et 14 of 21 1/5/81 Action e Action Item Item Description Number and Iteference 23 44A The audit committed to in our response to (open) (1) CPCo Audit not completed & (2)
Ouestion 1, Part b, and described in Part (2), Existing Audit findings (Mdi-53-0) not satis-Section 5.0 was conducted once during the factorily resolve; i.e., inadequate corrective course of the PSAR rereview (commencing March 17, action. This item has been identified as an 1980) and again after completion of the rereview item of noncompliance as discussed in para-(commencing November 3, 1980). graph 4 of this report. (Question 23, Part (2), Section 5.0, Page 48)' l I l l w -
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1, I l .. 1 Action Action Item
- l. Item Description Number #e and Reference 23- 45 U.S. Testing was required to demonstrate to (0 pen) CPCo commitment not completed.
. cognizant Engineering Representatives that ,, l testing procedures, equipment, and personnel used for quality verification testing (for other than NDE and soils) were capable of providing accurate test results in accordance with the requirements ot applicable design documents. (Ques tion 1, Appendix I, F7ction D.3.b, Page I-18;
- Question 23, Sebsection 3.10, Page 27; and Subsection 3.11, Page 31) 1 .-
23-4 6 A sampling of U.S. Testing's test reports (for (0 pen) CPCo commitment not completed. other than NDE and soils) were reviewed by coonizant Engineering Representatives to ascertain l that results evidence conformance to testing requirements and design document limits. l (Question 23, subsection 3.10, Page 28.; and Subsection 3.11, Page 31) 23d7 See Action Item Number 4 (47) (0 pen) CPCo commitment not completed. 23-48 CPCo performs overinspection for soils (Closed) Verified CPCo overinspection plan, placement, utilizing a specific overinspection 01-C-3A, Rev. 1 for soil placement and reviewed plan. completed overinspection results performed on (Question 1, Appendix I, Section C.2.b, Page I-ll* weekly basis.This overinspection program is an ongoing activity by Midland QA group. Section C l.c, Page I-16) 23-49 CPCo performs overinspection of the U.S. (Closed) Verified CPCo overinspection plan Testing soils testing activities and reports, Ol-C-4A, Revision 3, for soil testing and ! utilizing a specific overinspection plan. review completed overinspections performed on U. S. Testing. (Question 1, Appendix I, Section C.3.c, Page I-17)
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n ^ - Shcet 16 ca 21 1/5/81 Action Action Item Item Description Number and Reference 23-50 CPCo Project Management and QA review field (closed) Verified CPCo reviews of field pro-procedures (new and revised) and CPCo OA reviews cedures and quality controi' instruction in OCIs (new and revised) in line with Dechtel before addition to Bechtel ptfor to release. release. (Question 1, Appendix I, Section D.S.b, Page I-19) 23-51 In 1978, CPCo implemented an overinspection plan (Closed) Verified CPCo has overinspection plans struction and the Bechtel inspection process, in the civil, electrical, mechanical, and with the exception of civil activities. Re- welding /NDE work activities. inforcing steel and embeds were covered in the overinspection. (Question 1, Appendix I, Section D.5.c, Page I-19) 23-52 CPCo reviews onsite subcontractor QA manuals (Closed) Verified CPCo reviews subcontractor and covers their work in the audit process. QA manuals and audits subcontractor work. (Question 1, Appendix I, Section D.5.d, Page I-19) 23 53 An ongoing ef fort is improving the " surveillance" (Closed) Verified that SF/ PSP G-6.1, Rev. 5 mode called for in the OCIs by causing more " procedure for Quality Control inspection plans' specific accountability as to what character- have deleted surveillance method and new requirec istics are inspected on what specific hardware ments direct inspection by witness or test to and in some cases changing " surveillance" to be performed by Quality Control surveillance
" i ns pe c t io n . - method has been deleted in para 3.3.3.a.3 of G-6.1, Rev. 5.
(Question 1, Appendix I, Section D.S.e, Page I-19) 1 o
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Shcet 17 o. - 1/5/81 Action Action Item Item Description Number and Reference (Status) 1-1 Perform a final review & update of PSAR commitment list. (Open) Action item not reviewed by NRC during the inspection. 1-2 Review sections of FSAR determined to be inactive. (0 pen). See Action Item 23-44 for NRC review & results. 1-3 Review EDP 4.22. (closed) See Action Item 23-23 for NRC review. 1-4 Audit Action Items 1-3 (0 pen) See Action Item 23-44A for NRC review & results. 1-5 Review specifications not included in specificity (Closed) See Action Item 23-10 for NRC review study initially, and results. 1-6 Dames and Moore Report was reviewed and recommendations (Open) File indicated review was complete, identified and dispositioned. however, no details of the recommendations identified or the dispositions were available. (Question 23, Subsection 3.1, Page 23-6) (Question 1. Apx, I, Page I-6, Para C.I.(b) 1-7 Complete review of pertinent portions of FSAR sections (Closed), Verified FSAR, Revision 18 to have (1) inconsistency between FSAR 3.8.5.5 2.5 and 3.8. corrected: and 2.5.4, Figure 2.5 48, settlement values, (2) Inconsistencies between FSAR subection 2.5.4. and Table 2.5-9 and Table 2.5-14 regarding soil type 3.8.5 have been corrected via FSt.R Amendment 18 supporting structures from clay to (Zone 2) (Feb 28, 1979) the same revision also corrected ramdom fill, (3) Table 2.5-16, index of compressi-inconsistency between 2.5.4 and drawing C-45. bility factors to be determined from fill studies. (4) Table 2.5-21 compaction requirements. (Question 23, subsection 3.3, Page 23-11) Reviews of Section 2.5.4. are on " Hold" until (Question 1. Apx I, Page I-6, Para 3) resolution of soils issue.NRC office of NRR Ceotechnical Branch will review FSAR section when final.
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Sheet 18 of ;[ Action -Action Item 1/5/81 Item Description Number and Reference
- 1. - 8 Correct Settlement Calculations (closed) verified settlement calculations have-been made subsequent to surcharge operations (RE:
' Settlement calculations will be revised after calculation No. S-105 File 8230, dated February 14, completion of diesel generator building sur- 1980), results of these calculations have been charge operations, included in response to question 27 of 50.54(f) requests. Review of this response and results (Question 23, Subsection 3.4, Para 23-13) of calculations are being made by NRC office of NRR Geotechnical Branch.
(Question _1, Apx I, Page I-6, Para C.4.a) 1-9 Schedule audits of the Geotechnical Section on a six (Closed) Review audits of Bechtel Geotechnical month basis, dated February 26-28, 1979, and August 29-31, 1979, and February 26-28, 1980. A recent Bechtel QA audit of Bechtel Geotech Section was conducted in February 1979. Additional audits will Audits are scheduled for every six months, be performed in this area on a six month cycle until completion of soil work. (Question 1, Apx I, Page I-7, Para C.4.c) 1 --10 Review drawings for possible effect of vertical duct (Closed) See Action Item 23-14 for NRC review. bank restrictions. 1 -- 11 Complete actions in response to DRUCL audit. (Closed) See Action Item 23-1 for NRC review. 1 - 12 Revise EDP 4-49 to-incorporate clarifications and (Closed) See Action Item 23-3 for NRC review. instructors for use of specification change notices. 1 - 13 Schedule audits of each design discipline calculations (open) CPCo commitment not completed. on a yearly basis. 1 - 14 Re-evaluate construction equipment used for ccmpaction. (Closed) Verified 50.54(f) submittal, " Report on Test Fill Program" which provides documentation for Compaction equipment currently in use has been qualified qualification of compaction equipment currently in and construction notified of parameters governing use of use, Spec. C-211, attachment 1, provides a list of equipment. equipment to be used and compaction requirements. (Question 23 Subsection 3.6., Page 23-18) (Question 1 Apx I, Page I-11, Para C.1)
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Sheet 19 t 1/5/81 Action Action Item Item Description Number and Reference 1 - 15 Assign Field Soils Engineer and Soils Engineer from (Closed) Verified Spec. C-211 Para 8.3.5. requires design section. soil work to be performed under direction of One full time and one part time onsite Geotechnical Soils Engineer has been assigned. (Question 23, Subsection 3.7., Page 23-20) (Question 1, Apx I, Page I-11, Para C.2.a) 1 - 16 Review construction specifications and procedures to (0 pen) See Action Item 23-8 , CPCo Commitment identify equipment requiring qualifications. not completed. 1 - 17 Review field procedure FPC-3.00 to ensure clarify and (Closed) See Action Item 23-7a for NRC review. completeness. 1 - 18 PQCI 1.02 has been revised to incorporate the specific (Closed) Verified PQCI 1.02 (Rev. 5) has been characteristics to be verified by Quality Control. revised to include specific characteristics to be inspected. (Question 23, Subsection 3.8, Page 23-22) (Qunstion 1, Apx I, Page I-16, Para C.1.a) . Project Quality Control Instructions C-1.02 was revised (closed) Verified C-1.02, Rev. 5 requires compaction to include verification of use of qualified equipment & equipment to be qualified and will adequately com-compliance with qualified procedures, pact the material being placed and provides for a daily soil placement report. (Question 23, Subsection 3.6, Page 23-18) (Question 1, Apx I, Page I-16, Para C.I.a) (Question 1. Apx I, Page I-17, Para C.4.a)
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..- sh, 20 of 21 1/5/81 Action Action Item Item Description Number and Reference 1 - 18 PQCI 1.02 was revised to provide specific inspection (closed) Verified PQCI 1.02 (Rev. 5) Para 2.3 Cont' requirements for verifying soil moisture contents, has been revised to provide inspection of moisture rather than surveillance, testing. **
(Question'23, Subsection 3.9, Para 23-24) 1 - 19 Complete in depth' review of soil test results (Closed) Reviewed and verified report entitled, Geotechnical Services has completed an investigation " Review of U. S. Testing Field and Laboratory which includes an in depth review of testing performed Construction Test Data on Soil Uses as Fill", by U. S. Testing and. reported test-results. dated July, 1979 was performed. (Question 23. Subsection 3.10, Page 23-27) (Question 1 Apx I, Page I-17, Para C.3.a) An in depth soils investigation program provides verifi- (Closed) Verified that berings test pits, cation of the acceptability of the soils or identified laboratory tests, analysis of past test results any nonconformances requiring further remedial action. and plots of all tess have been performed as part of the investigation of the subsurface materials. (Question 23, Subsection 3.8, Page 23-23) This information has been submitted to the NRC and' is currently under review by NRC office of NRR, (Question 1, Apx I, Page I-17, Para 3.a) Geotechnical Branch. 1 - 20 Perform in depth audits of U. S. Testing. (Closed) See Action Item 23-15 for NRC review. 1 - 21 Review of QCI's for surveillance call outs. (Closed) See Action Item 23-19 for NRC review. 1 - 21A Modify.QCI's Based on Item 1-21. (Open) CPCo commitment not completed. 1 - 22 Evaluate documentation call outs on QCI's (closed) See Action Item 23-19 for NRC review. 1 Incorporate scientific sampling plans for (Closcil See Action Item 23-34 for NRC review. inspection. 1 --24 Complete in depth review of Bechtel trend program. (Closed) See Action items 23-35 and 23-36 for NRC review.
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e - . Si 21 of 21 1/br e n 1 Action Action Item . Item Description Number and Reference i 1 - 25 Conduct QA Training. (0 pen) See Action Items 23-16 and 23-17 for NRC review. (Untus.bered) Selection of protor curves will no longer be a problem (open) A review of this commitment resulted in an item of' noncompliance as discussed in paragraph because each field density test will be accompanied by a separate laboratory standard which will provide a 3.(c) of this report. 1 direct comparison. This was directed by a letter to U. S. Testing and reflected in specification change notice i C-208-9004, dated April 13, 1979. I (Question 23, subsection 3.10, Page 23-27) 1 (Question 1, Apx I, Page I-17, Para C.4a) Specifications ver revised to provide more definition (closed) Verified spec. C-211, Rev. 12, Para 8.4 l- (Unnumbered) (moisture control) has been revised to provide requirement for soil moisture testing. specific requirements for moisture. testing. (Question 1, Apx I, Page 1-16, Para C.2.a.) Spec. C-210 and 211 were revised to incorporate inter- (open) Interpretations had not been identified or (Unaumbered) pretations that affected specification requirements. evidence of being incorporated into specifications. (Question 23, Subsection 3.2., Page 23-8) i (Question 1, Apx I, Page I-6, Page C.I.a) [ 3 (Unnumbered) The requirements for the control of testing were (Closed) Reviewed and verified Spec. C-208, Rev. 20 j 20. Para 9.1.3. to require all field devsity tests adjusted, requiring the testing subcontractor to check all field density tests for cohesive material to be checked to the zero-air voids curve.
- against the zero-air-voids curve.
(Question 23, Subsection 3 10, Page 23-27) 4 l: PQCI SC-1.05 was revised to add more stringent require- (Closed) Verified PQCI 1.05, Rev. 11 was revised (Unnumbered) to include requirements for inspecting in process ments for in process inspections of U. S. Testing, testing activities. l (Question 23 Subsection 3.10, Page 23-27) ' i
PDi 5 p cacg[o , UNITED STATES 8 .c 7, NUCLEAR REGULATORY COMMISSION
$ ". ,I REGION lli
$ t".- 799 ROOSEVELT MoAD o*g CLEN ELLYN. ILLINOls 6o137
[ t, FEB 21981
. v t Docket No. 50-329 Docket No. 50-330 - - t.g
. p-Consumers Power Company , c r. .-G, ATTN: Mr. James W. Cook '
Vice President, Midland Project \., , " 3.i..y%[. 1945 West Parnall Road ' Jackson, MI 49201 ' .k.k (. . ' ' Gentlemen: This refers to the routine inspection conducted by Messrs. E. J. Gallagher, (' R. B. Landsman, and R. Sutphin of this office on January 7-9, 1981 of activities at the Midland Nuclear Power Plant, Units 1 and 2 authorized by NRC Construction Permits No. CPPR-81 and No. CPPR-82 and to the discussion of our findings with Mr. W. R. Bird at the conclusion of the inspection. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective ( examination of procedures and representative records, observations, and interviews with personnel. During this inspection, certain of your activities appeared to be in non-compliance with NRC requirements, as described in the enclosed Appendix A, and a written response is required. Certain other activities, set forth in Appendix B to this letter, appear to be a deviation from commitments which you have made in previous correspondence with the Commission. Please advise us in writing within twenty-five days of L the date of this letter of the corrective action you have taken or plan to take, showing the estimated date of completion with regard to this deviation. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter, the enclosures, and your response to this letter will be placed in the NRC's Public Document Room, except as follows. If the enclosures contain information that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty-five days of the date of this letter, to withhold f id I,n.-v .y .r.V .V 4 % d , ag -
7 2 Ib'I Consumers Power Company 2-( such information from public disclosure. The application must include a full statement of the reasons for which the information is considered proprietary, + and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application. We will gladly discuss any questions you have concerning this inspection. Sincerely,
'ab James G. Keppler Director
Enclosures:
(- l. Appendix A, Notice of Violation
- 2. Appendix B, Notice of Deviation
- 3. IE Inspection Reports No. 50-329/81-01 and
( No. 50-330/81-01 cc w/encls: Central Files Reproduction Unit NRC 20b PDR
. Local PDR t
NSIC TIC Ronald Callen, Michigan Public Service Commission Myron M. Cherry, Chicago (
_. .. . . .- .. ~ - -. - - . . - , i l l 1 l Appendix A . t ( NOTICE OF VIOI.ATION e l ^ Consumers Power Company Docket No. 50-329 g' Docket No. 50-330 1 As a result of the inspection conducted on January 7-9, 1981, and in accordance with the Interim Enforcement Policy, 45 FR 66754 (October 7, 1980), the following violations were identified.
- 1. 10 CFR 50, Appendix B, Criterion V, requires in part " Activities affecting
. quality shall be prescribed by documented instructions, procedures, or drawings". J CP QA Program Policy No. 5 states in part, " Prior to performing... inspection [ on a safety related item, suppliers are required to develop written procedures for . . . perioraing required inspection and tests. These procedures reference applicable drawings, specifications, codes and standards. CPCo QA Departments j review field... inspection procedures prior to implementation". 1 Contrary to the above, the inspector determined that U. S. Testing Company [ [\ has not established test procedures for soils work activities. The specifi- i cation for testing, C-208, references ASTM standards for performing specific < j tests, but does not include procedural controls or instructions for implement-ing the tests. This is Severity Level V violation.
- 2. 10 CFR 50, Appendix B, Criterion VI requires in part, " Measures shall be
( established to control the issuance of documents . . ." CP QA Program Policy No. 6 states in part, " Documents which prescribe activities affecting quality . . . are. . . controlled according to written i procedures. . . The document control system provedures for: Identifying r i the proper documents to be used in performing a quality related. . . activity; establishing current and updated distribution lists". Contrary to the above, the inspector determined that U. S. Testing Company t test result forms are not controlled. The proper documents to be used for a 4
. specific test are not defined. There is no distribution list for the forms.
The latest revision of the forms are not controlled. This is a Severity Level V violation.
- 3. 10 CFR 50, Appendix B, Criterion XVII, requires in part, " Sufficient records shall be maintained to furnish evidence of activities af fecting quality".
ANSI N45.2.9, Section 3.2.1. requires in part, " Quality assurance records l j (- shall be considered valid only if stamped initialed, signed, or otherwise ' authenticated and dated by authorized personnel". i L5,fa s a n & ^ i
-~ w-
)
Ag'
Apprndix A . CP QA Program Policy No. 17 states in part, " Compile records as specified ( in applicable procedures, codes . .. Bechtel Field Instruction FIC-1.100, Appendix A under Daily FER, paragraph No. It states in part, " Review and initial all acceptable test report sheets from U.S. Testing . . . Contrary to the above, the inspector observed that numerous U.S. Testing , report sheets were rubber stamped with the name of the onsite geotech engineer and not initialed and dated as required. In addition, there were no procedural controls for the use of the signature stamp. This is a Severtly Level VI violation. Pursuant to the provisions of 10 CFR 2.201, you are required to submit to this office within twenty-five days of the date of this Notice a written statement or explanation in reply, including for each item of noncompliance: (1) corrective action taken and the results achieved; (2) corrective action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved. ( Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, this response shall be submitted under oath or affirmation. l (* Dated Fn k ._ _a sqgj u ~L }b er he _ _
-~
i( 0 ' ames G. KeppleP U l Director (9 i J
I l l 1 l Appendix B !. ( NOTICE OF DEVIATION l . Consumers Power Company Docket No. 50-329 Docket No. 50-330 , ! As a result of the inspection conducted on January 7-9, 1981, the following was cited as a deviation. 10 CFR 50, Section 50.54(f) response from CPCo to Question 23, Subsection 3.7 states in part, "One full time and one part time onsite geotechnical engineer i have been assigned. These engineers provide technical direction and monitoring of the process." Contrary to the above it was determined that the assigned engineering technician does not satisfy the commitment made in 10 CFR 50.54(f) submitted to provide an I onsite geotechnical engineer and to implement the duties and responsibilities of FIC 1.100, Appendix A " Duties & Responsibilities of the Onsite Geotechnical Engineer". l s i i i (
,A s;;e 26 v =- -
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT \ REGION III Report No. 50-329/81-01; 50-330/81-01 Docket No. 50-329; 50-330 License No. CPPR-81; CPPR-82 Licensee: Consumers Power Company 1945 Parnall Road Jackson, MI 49201 Facility Name: Midland Nuclear Power Plant, Units 1 and 2 Inspection At: Midland Site Inspection Conducted: January 7-9, 1981 Inspectors: E. J. Gallagher O -
/'/d-O/
R. B. Landsman I
/-2.f - f /
V R. N. Sutphin, Jr. M, / - 2 (5 - f/ Reviewed By: fC$W R. C. Knop, Chief, Projection Section I I/ Approved By: . F' elli,k / 7 [ Reactor Construction and / / ( Engineering Support Branch Inspection Summary Inspection on January 7-9, 1981 (Report No. 50-329/81-01; 50-330/81-01) Areas Inspected: Consumers Power Company response and implementation of corrective actions regarding the 10 CFR 50.54(f) request of question 1 of NRC letter dated March 21, 1979 and question 23, request for additional information dated September 11, 1979; procedures, quality records and observation of work related soils work activity; quality assurance organization status, construction schedule; and status of personnel air locks. The inspection involved a total of 60 inspector-hours on site by three NRC inspectors. I. n/
. _~ -_ _
i Results: Three items of noncompliance and one deviation were identified in the above inspected areas - Severity Level IV, Inadequate Procedures; Severity Level -
< IV - Inadequate document control; Severity VI - Inadequate Quality Assurance Records; Deviation from commitment to provide Geotechnical Engineer.
i u b. t i J i t g 4 ( . l 1 - i 4 r
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1 4 2-
1 DETAILS ( Persons Contacted Consumers Power Company i D. Miller, Site Manager ;
*W. Bird, Quality Assurance Manager
*T. Cooke, Project Superintendent
*R. Davis, Quality Engineer
*R. Wheeler, Staff Engineer
*D. Horn, Civil Group Supervisor
*H. Leonard, Section Head, Quality Engineering
*D. Turnbull, Site Quality Assurance Superintendent i D. Keating, Section Head IE and TV l
i Bechtel Power Company 4
*J. Russel, Assistant Project Field Quality Control Engineer
*E. Smith, Project Field Quality Control Engineer
*P. Corcoran, Resident Assistant Project Engineer
*P. Goguen, Lead Civil Field Engineer
*J. Betts, Assistant Field Project Engineer
*M. Deitrich, Project Quality Assurance Engineer
*L. Snyder, Resident Quality Engineer
( U. S. Testing Company J. Speltz, Lab. Manager , The inspectors also contacted other licensee and contractor personnel during the course of the inspection. {
- Denotes those in attendance at the exit meeting on January 9, 1981.
( Licensee Action on Previously Identified Items 4 (Closed) CPCo Action Item: S100D, NRC Inspector E. W. K. Lee's concern regarding
" unreadable documents" in one of the document packages for a shop weld, as ex-1 pressed by the inspector during his October 9-10, 1979 inspection, and item "D" j of J. L. Corley October 12, 1979, Midland Memo No. 344FQA79. CPCo determined that i documents of concern applied to spool piece 2CCA-61-5611-2-6 wherein parts of two i
radiograph reports were not legible due to copy machine problems. Legible copies
- were secured for this file. In addition Bechtel QC reviewed 20 other M-104-A data j packages and found 9 out of 298 pages in similar condition. Legible copies were j nade of all discrepant pages and filed, as recorded in L. A. Dreisbach Memo
. LAD 168S of August 8,1980, to J. L. Corley. This item of NRC concern is closed. t ( .
4 (Closed) 329/79-12-01; 330/79-12-01 " Work Prints in use were not current revision", CPCo Action Item S500. CPCo advised inspector that incorrect issue (Rev. 11) of [ print was removed and correct issue (Rev. 12) provided to the work area at the time of the original inspection. Current issue of the drawing is Revision 14, issued September 14, 1979, title of the drawing in Question is " Decay Heat Removal, Core Flooding System Unit 2 Hangers, Location and Identification". Three audits have been performed since the original inspection. One was reported in May 1979, ' one reported September 9, 1980, and another reported on November 24, 1980, the ; results of which indicate a significant improvement in the revision control of work prints. Based on this corrective action and commitments by CPCo to maintain a continuing program of audits in this area of activity, the referenced item of noncompliance is closed. (Closed) Unresolved Item 329/80-11-03; 330/80-12-03, "Soubber Missing Required Spacer". Two snubbers were found to have one missing spacer washer each, however, they had not gone through the regular inspection and acceptance by Quality Control at the time 24 pipe supports in various stages of inspection and installation were observed by the NRC inspector. As a result of the express concern of the NRC inspector training was planned and conducted on June 10, 1980, construction corrected the condition of the two snubbers, and documented the inspection on a QC-Gl-1 form, I and activity 3.1.c in the P-2.10 PQCI has been added to verify future configurations and orientations are correct, per M. A. Dietrich Memo No. LADl?54 of November 12, 1980. Based on this documented corrective action, this unresolved item is con-sidered closed. (Closed) CPCo Action Item: S479, NRC inspector R. N. Sutphins concern regarding lack of update of EDPIS per comments in NRC reports No. 50-329/80-30; 50-330/80-31, (- Page 11. The inspector checked the two items in the Bechtel Power Corporation Engineering Department procedures manual and found them to be in proper order. i This concern is closed. However, additional technical review of the Engineering f Department procedures manual will be conducted at subsequent inspections. i 4 i 1 (
Section I Prepared by R. N. Sutphin
, Reviewed by R. C. Knop, Chief Projects Section 1
- 1. Functional or Program Areas Inspected
- a. CPCo Quality Assurance Organization The inspector reviewed the organization chart of the combined CPCo -
Bechtel Quality Assurance Organization, issue date January 1, 1981. Mr. M. A. Dietrich is (Acting) PQAE replacing Mr. L. A. Dreisbach who has been reassigned. Mr. D. M. Turnbull had reported to the site in the position of Site Project QA Superintendent. The combined Midland project Quality Assurance organization now has 40 persons assigned compared to 36 as of mid October 1980. The supervisor of the adminis-tration Group reporting to the site project Quality Assurance Superin-4 tendent will be announced in January 1981. Additional information was requested on the person who has been selected to fill an open position as Civil Quality Control Engineer in the Bechtel Quality Control Organi-zation.
- b. Construction Schedule
( The inspector checked the status of construction and construction schedules for the overall project, and received a copy of the January 8, 1981 CPCo report memo to h7C on the current yellow book schedule.
- c. Onsite Design Activity l The inspector checked the status of the ongoing onsite design activity and continued his review of the engineering department procedures manual.
The inspector toured the site to observe the status of the work. ( No items of noncompliance or deviations were identified.
- 2. Other Inspection Areas
- a. 50.55(c) Personnel Air Locks The inspector checked on the status of the activity to resolve the remaining open questions on the personnel air locks. The manufacturer is working on the completion of the as-built record, revised drawings, and updated stress report. Bechtel will review these item when they are completed. The manufacturer will advise if any further repair or rework is recommend. This item will remain open and additional review will be conducted at subsequent inspections.
n Section II { Prepared by E. J. Gallagher R. B. Landsman Reviewed by R. C. Knop, Chief Projects Section 1 i
- 1. Review of Onsite Soils Works Activities As a followup to h7C Inspection Report No. 50-329/80-32; No. 50-330/80-33 the region III inspectors performed an inspection of the current onsite soils work activities to verify whether adequate corrective actions have been implemented as described in Consumers Power Company response to questions 1 and 23 of 10 CFR 50.54(f) submittals. The following are the specific findings.
- a. Procedural Controls for Soils Work I
It was determined that U. S. Testing Company (UST) have not established written procedures for implementing the requirements of Testing Specification C-208. This specification references numerous ASTM standards for performing specific tests but does not include procedural control or instructions for the implementation of such tests. ( (1) While observing a laboratory relative density test (ASTM 2049) it was observed that the variable rheostat on the testing apparatus was set at maximum setting. The lab technician stated that ASTM D2049 requires the setting of the machine at maximum amplitude. It was determined that UST did not previously determine the rheostat setting that produced the maximum density for the
,. material being used onsite. It was assumed by UST that maximum
. setting produced maximum density. Relative density tests are used to assure that the inplace field density meets the specification requirements.
i Corps of Engineers Manual EM 1110-2-1906 dated November 30, 1970, Appendix XII, Page XII-8 states the following:
"It has been determined that for a particular vibrating table, mold, and surcharge assembly, the maximum dry density of a speci-men may be obtained at a displacement amplitude (rheostat setting) less than the maximum amplitude of which the apparatus is capable; i.e. dry density may increase with increase in rheostat setting to a setting, beyond which the dry density decreases, therefore each laboratory should determine for it's apparatus the rheostat setting at which maximum density is produced and use this setting for sub-sequent maximum density testing."
Footnote on Page X11-8 states: ( 6-
i "It may be desirable to redetermine the optimum rheostat setting at the inception of testing for each major project." U. S. Testing had not determined this setting nor did a procedural
,. control exist for the determination of the rheostat setting.
(2) While observing limited field soils work being performed at the metering pits south of the essential service water intake structure i at elevation 630' it was determined that samples used to perform relative density tests have been taken after the material has been compacted. These samples should be taken prior to compaction since grain size and gradations can be altered during compaction. The relative density test should be performed on as received material used prior to compaction. Grain size is one of the important char-acteristics of how soil behaves. The inspector determined from a review of the available grain size analysis that there appears to be a gradation change of the material comparing before and after compaction. A procedural control specifying where and when to taken soil samples should have been established. UST does not have procedural instruc-tions specifying the field technique where and when to take samples for density tests. (3) It was determined from discussions with the cognizant UST personnel that.they have been performing in place density tests "at the direction ( of the onsite geotechnical engineer." However, there are no pro-cedural instructions as to what depth below the lift being compacted the test should be performed. A review of the density test reports indicate that they are not correlating the density test depth to the lift being compacted. Based on the above, it has been determined that CPCo is in noncompliance with 10 CFR 50, Criterion V (Procedures) in that adequate laboratory and field test procedures have not been established for the control of soil testing activities. (50-325/81-01-01; 50-330/81-01-01), i CPCo response to 50.54(f) question 23, subsection 3.11, page 23-31 states that "U. S. Testing was required to demonstrate to cognizant engineering representatives that testing procedures, equipment, and personnel used for quality verification testing were capable of pro-viding accurate test results. . . " This commitment has not been satisfied based on the above findings. Subsequent to the inspection CPCo informed the RIII offices on January 16, 1981, by telephone that U. S. Testing Corporation would develop and issue implementing procedures for soil work activities. These procedures will be reviewed during subsequent inspections. (.
- b. Document Control For Soils Work It was determined that U. S. Testing was using uncontrolled forms to record quality control test results. A binder was observed in the U.S.
Testing lab which contained QC forms used to record test results. On the inside cover it stated that the index does not reflect the latest revision of each form. The cognizant lab personnel were not able to demonstrate that the latest revision of QC test forms were being used since there were no document control provisions established to control I these forms. An undated U.S. Testing inter-office memo was presented to the hTC inspector as the procedure to follow when receiving revised forms. It states in part, " log into controlled forms index". The inspector requested such a form index but did not receive it. There was no documentation onsite as to what forms are to be used for what test as well as what are the latest revisions of the forms. Based on the above, it was determined that CPCo is in noncompliance with 10 CFR 50, Appendix B, Criterion VI, (Document Control) in that measures have not been established to control the issuance of documents which affect quality activities. (50-329/81-01-02; 50-330/81-01-02) Subsequent to the inspection CPCo informed the Rill office that quality control verification forms would be controlled by the implementing pro-cedures. This will be verified during subsequent h7C inspections.
- c. Soils Test Records (1) Quality assurance records for backfill work activities were reviewed for completeness and compliance with licensee specifications, pro-cedures, and commitments.
Bechtel field instruction FIC 1.100. Appendix A, duties and respons-
- ibilities of the onsite geotechnical engineer, Paragraph 18, requires
( that the onsite geotechnical engineer review and initial all accept-able UST test report forms. ANSI N45.2.9 (Quality Assurance Records), Section 3.2.1, requires that " quality assurance records shall be considered valid only if stamped, initialed, signed, or otherwise authenticated and dated by authorized personnel". Numerous UST density test reports were rubber stamped by the geotechnical engineer, however, none were dated. In addition no procedural controls were established for the use or control of the rubber signature stamp of the geotechnical engineer. Based on the above it was determined that CPCo is in noncompliance with 10 CFR 50, Appendix B, Criterion XVII (Quality Assurance Records) in that the soil test reports are not initialed or dated and there were no established controls on the use of a rubber signature stamp. (50-329/81-01-03; 50-330/81-01-03) (
.g.
Subsequent to the inspection CPCo informed the RIII office that soil test reports would be initialed and dated as required. This ( will be verified during subsequent inspections. (2) Specification C-208, Section 9.1.3(d) requires the geotechnical . engineer to review and evaluate test results when densities exceed certain values. From discussions with the previous geotechnical I engineer, it was determined that the evaluation consisted only of a check of the numerical calculations for numerical errors. If the calculations were correct the disposition was "use as is", this review does not meet the requirement to evaluate test results. Subsequent to the inspection CPCo informed the RIII office that documented evaluations of the above would be performed. This is an unresolved item pending review of the evaluation (50-329/ 81-01-04; 50-330/81-01-04),
- d. Review of Nonconformance Reports The NRC inspector requested all nonconformance reports regarding soil g work activities since March of 1979. The following reports were made s
available. Bechtel NRC No. Date Description of NRC Status 2294 6/23/79 Failing density tests Closed ( 2307 2350 6/25/79 7/16/79 Failing density tests Failing gradation test Closed Closed 2492 8/30/79 Qual of compaction equip. Closed 3041 6/25/80 Failing density tests re-Opened 3159 10/07/80 Failing gradation test Closed 3165 10/09/80 Lift thickness exceeded Closed k CPCo NRC No. Date Description of NRC Status M-01-4-0-005 1/18/80 No spec. requirement for {s backfil around piping Open M-01-2-9-060 7/19/80 Spec & purchase order for sand gradation not the same. Closed M-01-9-0-038 5/15/80 Final report on qualifica-tion of compaction equip-ment. Open The above closed NCR's were determined to be adequately resolved, those open or reopened will be reviewed during subsequent inspections. ( -
- e. Qualifications of Onsite Geotechnical Engineer CPCo response to 50.54(f) question 23, subsection 3.7, page 23-20, states that, "one full time and one part time onsite geotechnical soils engineer have been assigned." The inspector requested the qualifications of the onsite geotechnical engineer. A resume was presented to the inspector as representing the assigned individual to implement the commitment in order to preclude future soils problems. This engineer is to provide I the technical direction and monitoring of the entire earthwork process.
The resume that was presented was of an " Engineering Technician" with no previous formal education in engineering or geotechnical engineering. The engineering technician had nominally 15 years of field and laboratory testing of soils. This information was discussed with representatives of the NRC geotech-nical branch. It was determined that CFCo committed to provide techn-ical direction from a geotechnical engineer capable of being recognized and licensed by a state board of registration of professional engineering or equivalent. In view of the fact that adequate technical direction had not been pro-vided per the commitment by CPCo in the 50.54(f) response it has been determined that CPC0 is in deviation from a NRC commitment as described in Appendix B of the transmittal letter of this report. (50-329/81-01-05; 50-330/81-01-05) Subsequent to the inspection, CPCo informed the RIII office that a geotechnical engineer would be onsite beginning January 19, 1981 and that job descriptions and qualifications for the geotechnical engineer for the speciality remedial work to follow would be developed. This action will be verified during subsequent inspection. ( 2. Borated Water Storage Tank Reanalysis During the inspection, the licensee informed the inspector that the prelimin-ary structural reanalysis of the borated water storage tank ring foundation ( yielded results that indicated the foundation to be overstressed. The in-spector inquired to the quality assurance group if this condition had been considered for reportability. After this inquiry, the Manager of Quality Assurance produced the CPCo form entitled, " Safety concern and reportability evaluation" for the BWST ring foundation. This document indicates that a finite element analysis results in forces and moments in excess of FSAR allowables. The reportability evaluation indicated the event was "not reportable" with further evaluation necessary. The planned actions included retaining a consultant to review the results obtained by the analysis and/or perform an independent check and excavate and inspect the foundation for signs of overstressing (i.e., cracking). (
On January 23, 1981, CEPo informed the RIII office the excavation around the Unit 1 tank was complete and that cracks were observed in the areas of over-( - stress. CPCo identified this item as a 50.55(e), significant construction deficiency. This item is under review by the NRC staff. 4 , The inspector informed the Region III office and NRR project manager on January 9,1981. The licensee's evaluations of the matter are included as 1 Attachment No. 1 of this report. Unresolved Items - Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items or items of noncompliance or deviations. Unresolved items disclosed during this inspection are discussed in Section II, Paragraph 1.c.(2). Exit Meeting The inspectors' met with licensee and contractor representatives at the conclusion of the inspection on January 9,1981 and summarized the inspection scope and
- findings. The licensee acknowledged the inspection findings. Subsequent to the exit meeting the inspectors and RIII management telephoned on January 14, 1981 the Quality Assurance Manager in order to verify what corrective action would be taken based on the inspection findings. On January 16, 1981, CPCo informed 3 the RIII office of the actions to be taken which are contained in the report Section II, Paragraph la, Ib, Ic, and le.
Attachment:
Attachment No. 1 (
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4 I F ( . I G ___ - - , . - . . - ..,-.. - . - - _ _ - . , - - , ,. - _ . - - - , , . - - - - - - - ~ ~ - . , . . . - , , - -,- - - -.
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$gm SAFETY CONCERN AND "l,#'ch'E"Mo?
* " ^ " "
U"l#?O REPORTABILITY EVALUATION ~'"7c^E .
- 5. now wAS C0aCEw IDEarIrIED, WEN, wERE2
( As a result of the 50.54(f) com:nitments to do a TO MANAGER-MPQ/ structural reanalysis of Category 1 Structures (See Items 1. R OM: R L Rixford 14-7 and 48-2), the BWST ring foundation was reanalyzed ORG G A M @ - % and values wire obtained which were inconsistent with SCPI NO: 5 previous values, and inconsistent with FSAR requirements. FILE NO: 15.1 The results of the analysis were obtained 1-4-81 and dis- DATE RECEIVED: cussed in a 1-5-81 CPCo/Bechtel meeting. The Project 11anager attended this meeting and subsequently briefed M 2 A PW ? ' the !!anager of Quality Assurance. WENT YES 6:30 N/A BY w0M? N/A
- 3. IS NRC AWARE OF THIS?
. YES H I:0 (CONTINUE ON NEXT PAGE) B OM? N/A
$. BRIEF DESCRIPTION OF CO::CER*! - SYSTEM, COMFOUENT, ACTIVITY, POSSIBLE SAFETY I!@ACT -
(ATTACH SUPPCRTIliG DOCU ENTS). The BWST ring foundation was analyzed for several loading combinations including the. dead f load plus live load which was determined to be the most severe. The analysis was first performed using the eethods uf BC-TOP-4A ,Rev. 3 (this method uses springs for soil / structure interaction during a scismic en nt), but gave displacement values inconsistent with anticipated and measured values. The analysis was then done using a finite element techn13 ue which gave consistent displacement values but forces and moments in excess of FSAR allowables. The values obtained fron the reanalyses which have been done indicate ( , an overstressing and, hence, a potential for failure of the foundation of the Category I BWST. (CO !TI!!UE O!! UEXT FAGI)
- 6. IKCIATI REFCFTABILITY EVAU!ATION: 7. ORGA:31ZATIC:I REFO : sib 11 FO? FURTHER
- a. O REPCRTABLE - GO TO 13 EVALUATION:
b.O POTE:iTICLY REPOSTABLE - GO TO 13 Bechtel Engineering - Civil c.[3 N:TI REPC.R! ELE, FURTHER EVALUATION 8. FINA1 REPORTABILITY EVA* UATION (, 1, 0 NOT REFOPTABLE (IT 6.c. CHECCD):
- a. StootTABLE D. !!CT PIPORTABLE 9 QA AFF307AL OF EVALUNI10:1 e g[
i
\
OF BIACVS 1 TO 7: MANAC ER - !?QA k~7/f/DATE
- 10. JUSTIFICATICN OF IVA1A ATION - ( ATTACH SUFF0!CI!!G DOC UICTS)
The first reanalysis give displacement values which vere inconsistent with measured settlement and anticipated. values. This cast doubt upon the spring values used in the analysis. The subsequint finite element analysis gase displaceme,nt values which were con sistent with the other values available for comparison, but gave forces.and moments which exceeded the FSAR allosables by an amount sufficient to warrant an additional check on these values also. Two actions planned to check these values aret
- 1. Retain a consuleant to review the results obtained by analyses done, and/or do an independent check.
2 Excavate and innpoct the foundation for signs of overstressing (i.e., crackin:). It was considered prenature to judge this a reportable condition prior to confirmction of
, the values obtain d by the finite element analysis. (C0!!TINUC 0:: !! EXT PA",")
k .. EVALUATOTU" #*NTITnC/.*>AT": 12. FJ;lAk QA AlPl10 VAL - l'AMACER lfrQA/DAT):: Q /~J'-JV ,
- 13. lihC 1:0r: FICA 110:l liOV7 DATE: . illC; It:ntvicmi. tur:nrni
,FEB 0 0 21 m TuracE:
' ymma I N'TI
. dAFETY CONCERN AND "f4'cE!lfu's'c"If i
REPORTABILITY. EVALUATION OUALITY ASSURANCE EEPARTMElif scar no. 5 QA70-0 PAGE 2
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I 3
- 5. CONTI!.UED I
- 10. CONTINUED e . .
ll . MIllIMUM DISTRIBtTION: 4
- 15. ADDITIOlAI. DISTRIBUTION:
VICE PRESIDE!iT - PE&C ltPQAD - DQAE Supervisor VICE l'RECIDEI;T - IIIDLAND PROJECT . DIllhCTOR - EiVIR0!!!!E ITAL SERVICES & QA 141DI.MID SI'IE !'A' LACER ( SITE QA StirEBII.'TF.!iDEllT ~ MMIMER - FAFETY & LICENSI!!O MIDIA'lD FILE fl0 15 1 .
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April 30, 19S1 Mr J G Keppler, Regional Director Office of Inspectica and Enforcement US Nuclear Regulatcry Comission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 MIDLAND PROJECT - DOCKET NO 50-329 AND NO $0-330 FILE 0.4 LTI 70*01 SERIAI 11980 Reference Letter JGKeppler to JVCook concerning the March 13, 1981 meeting f in Glen Ellyn, IL, dated April 2, 1981 a The referenced letter documents the meeting of March 13, 1981 at which Consumers Power made presentations concerning the Midland Project organization and the Midland Project Quality Assurance Program. In that meeting, Consumers Power made a commitment to provide additional written infomation on the
- Midland Project Quality Assurance Program. In partial fulfillment of that commitment, this letter transmits an " Executive Sur. mary" entitled " Midland Project Quality Assurance Program Update." The appendices referenced in the Executive Su.uary will be transmitted under separate cover. It is anticipated .
that they will be transmitted by May 22, 1981. k JVC/VRB/mo Enclosure 1 Midland Project Quality Assurance Program Update, Executive Summary, dated April 1981. CC JWGilray, NRR, Quality Ass,urance Branch RJCook, USNKC Resident Inspector, Midland Nuclear Plant (1) 1 oc0481-0312a102 g g 'f D f hh /\'11Adl4#I
BCC JLBacon, M-1085A WRBird, P-14-418A JEBrunnt.r, P-24-513 MADietrich, Midland-QA WDGreenwell, Bechtel-QA GSKeeley, P-14-113B l BWarguglio, JSC-220A DBMiller, Midland MIMiller, IL&B JARutgers, Bechtel DMTurnbull, Midland-QA k I i l l r oc0481-0312a102
f MIDLAND PRO.IECT QUALITY ASSURANCE PROGRAM UPDATE Executive Summary 0 1.0 PURPOSE During a meeting with the NRC in Glen Ellyn, Illinois on March 13, 1981, Consumers Power Company (CP Co) made a presentation in which certain improvements to the Midland Quality Assurance Program were described. This Hid1tnd Project Quality Assurance Program Update provides a summary of the aforementioned presentation and addresses topics previously identified in joint NRC/CP Co management meetings. 2.0 SCOPE _ Each is:provement to the Quality Assurance Program is presented with the following information: the background leading to the improvement; a description of the improvement; and future benefits expected from the implementation of the improvement. ( To11owing is a list of the titles of specific Quality Assurance Program improvements. The improvement titles are grouped according to the most applicable criterion of Appendix B,10 CFR 50. The list also cites the section of this Update in which a description of the improvement is given. Section In Which Applicable 10 CTR 50 Quality Assurance The Improvement Appendix B Criterion Program Improvement Is Described I. Organization CP Co Quality Assurance 3.1 Department (QAD) CP Co Midland Project Office 3.2 Midland Project Quality 3.3 Assurance Department (MPQAD) Onsite Project Engineering 3.4 II. QA Program CP Co Interdepartmental QA 3.5 Program Procedures CP Co QAD/MPQAD Departmental 3.6 Procedures Quality Tracking and Statusing 3.7 rp0381-0241a112
2 Section In Which Applicable 10 CFR 50 Quality Assurance The Improvement
- Appendix B Criterion Program Improvement Is Described Supplier Deviation 3.8 Disposition Requests Field Purchase Orders 3.9 i
New Reg Guide Implementation 3.10 III. Design Control Equipment Qualification Rereview 3.11 Specificity Reviews 3.12 VII. Control of Procurement Supplier 3.13 Purchased Quality Material, Equipment and Services Quality Verification 3.14 Documentation Rereview
" Flags" Review 3.15 g CP Co Quality Assurance 3.16 for New Work IX. Control of Control of Cable 3.17 Special Process Pulling X. Inspection Inspection 3.18 XVI. Corrective Quality Trend Analysis 3.19 Action XVIII. Audits Audits 3.20 The improvements to the Program are addressed in the following sections at an executive summary level. The appendices provide a more detailed description of each improvement and sometimes also provide supporting exhibits which constitute objective evidence of the implementation of the improvement.
3.0 IMPROVEMINTS 3.1 CP CO QUALITY ASSURANCE DEPARTMENT In 1976, CP Co management established a goal to enlarge and strengthen its Quality Assurance Department (QAD) and increase the Company's direct involvement in quality assurance for the Midland Proj ect. Several organizational changes were implemented to achieve this goal. rp0381-0241a112
_ _ - - . ._. .- = - - _ _ __. _- __ 3 A new Director of the CP Co QAD (Projects, Engineering and Construction) was hired in January 1977 as the result of a national j search for an experienced quality assurance professional. Soon , afterward, an analysis of the existing QAD organization resulted in an internal reorganization and the addition of several external personnel, the net effect having been a significant increase in the
- professionalism of the QAD.
The new QA Director reorganized the QAD in 1977 with two Section i Heads reporting to him for Midland quality assurance activities. One Section Head was located at the site and was responsible primarily for hardware inspection, examination and test verification (IE&TV); the other Section Head was located in the General Office and was responsible primarily for quality assurance engineering (QAE). This reorganization resulted in direct contact between the QA Director and IE&TV Section Head, thus enabling the QA Director's closer involvement with the site, greater ease of escalating and resolving site quality problems and greater ease of communicating quality improvements to the site. An individual with both quality assurance and nuclear design experience was assigned as QAI Section Head and was directly involved in evaluating the adequacy of proposed resolution of quality problems and in the quality aspects of the design phase of the project. A separate section for quality audits was established, also reporting to the QA Director. Along with the reorganization of the QAD came a large increase in ( the size of the QAD's staff assigned to the Midland Project. The number of CP Co quality assurance professionals increased from 9 to 22 in 1977 and further increased to 26 by 1979. These increases enabled more concentrated and expanded CP Co overinspection of the site work and increased CP Co's involvement in preventive and corrective actions. These improvements remained in effect when the present Midland Project Quality Assurance Department was organized in 1980. (See Section 3.3.) 3.2 CP CO MIDLAND PROJECT OFFICE In March 1980, the CP Co Midland Project Office was developed and implemented to increase CP Co's involvement and control of the Proj ect , to make the Project organization as self-sufficient as possible within CP Co and to provide impetus to the resolution and closure of open items and project decision-making needs in general. The CP Co Project Office is headed by a Vice President assisted by the Project Manager. These two individuals directly supervise all phases of the conduct of the project. Reporting to the Project Office are six Department Managers who have responsibility for safety and licensing, design production, administration, quality assurance, site operations (construction and operations) and cost I and schedule. These departments are staffed with personnel who have extensive nuclear project experience and proven track records. rp0381-0241a112 t
4 In addition, the size of the CP Co Midland Project staff has steadily increased to help to assure the attainment of the , objectives noted above. Appendix A provides a summary description ( . of the current Midland Project organization. , Correspondingly, the Bechtel project organization has been s'trengthened by the addition of several key persons to support the j Bechtel Project Manager and has been restructured to directly 1 interface with the CP Co organization. Appendix B provides more i information on this improvement.
^
Overall, the entire Midland Project team has been expanded and strengthened. There is increased CP Co and Bechtel awareness and emphasis on quality. The remaining items described in this update are examples of this emphasis that has been evolving over the past several years in both program content and in selection of personnel for the leadership roles on the project. 3.3 MIDLAND PROJECT QUALITY ASSURANCE DEPARTMEhT As part of the March 1980 CP Co reorganization, the Midland Project Quality Assurance Department (MPQAD) was also formed. The MPQAD Manager reports, in his line operating role to the CP Co Vice President in charge of the Midland Project Office. He receives QA policy direction from the Director of Environmental Services and Quality Assurance (ES&QA), who sets all the Company's quality assurance policy for projects, engineering and construction (, activities. Midland quality policies and procedures are approved by the Director of ES&QA prior to their implementation. MPQAD consists of all of the CP Co QA resources, formerly contained in the QAD, who were directly charged with implementing the Midland QA program. Currently, the MPQAD staff totals 46, including 15 Bechtel personnel as described in the following paragraph. On August 15, 1980, as another step in the reorganization, the Bechtel Project Quality Assurance organization was integrated into the MPQAD. This was a positive step toward meeting the overall goal of increasing CP Co's control of the Project. This also provided single point accountability for the implementation of the Project Quality Assurance Program and improved the utilization of all the available quality 1ssurance resources in meeting the commitments of both the CP Co and_Bechtel Topical Reports. Appropriate changes to the Project Quality Assurance Program were implemented concurrent with the integration of the Bechtel Midland Quality Assurance Organization into the MPQAD. In most cases, CP Co employees hold the supervisory positions reporting to the MPQAD Manager, who is also a CP Co employee. Direct communication between MPQAD and other departments within^either CP Co or Bechtel is assured by established organizational interfaces. Appendix C provides a detailed description of the MPQAD, including the special role of the Bechtel Project Quality Assuranc'e Engineer. 4 As a' result of the integration, the MPQAD is in a " primary" rather
;than " overview" role. This results in MPQAD's more timely and
' complete involvement in both preventive and corrective activities.
^
rp0381-0241a112
.n . - .
~
5 The singular quality assurance entity (MPQAD) has had the effect of
, promoting Project interests.
3.4 ONSITE PROJECT ENGINEERING (nother organizational improvement in the quality effort not torma11y associated with the quality assurance program is the utilization of a large project engineering group located at the site. Onsite (resident) Engineers initially were assigned to the i Midland site in 1976 to enhance the coordination between Project Engineering (Home Office) and Construction (Site). In 1979, a separate group of Onsite Engineers was assigned to the site to perform design activities. The Onsite Project Engineers performing the coordination activity help to assure the understanding of design documents; expedite the correction of design and construction problems; expedite the processing for Field Change Requests, Field Change Notices, Design Change Notices and Nonconformance Reports; and approve construction activities, as required. Since 1976, the number of Onsite Engineers performing this activity has increased to 40. An Onsite Quality Engineer also was assigned to the group in 1979. Currently, approximately 170 additional Onsite Engineers perform certain design activities which are best perforced with a continuing knowledge of construction progress. This onsite design minimizes design interference and discipline interface problems, while simultaneously affording greater construction flexibility. Appendix D provides a detailed description of Onsite Engineering activities. 3.5 CP CO INTERDEPARTMENTAL QA PROGRAM PROCEDURES New CP Co interdepartmental Quality Assurance Program Procedures (Volume II Procedures) were prepared by a Management Task Force in 1979 to cover new requirements, to provide flexibility for our primary involvement, to improve technical content and to improve interface definition within CP Co. New areas covered were Definitions; Turnover to Projects, Engineering and Construction; Manufacturer's Notices; and Stop Work Orders. Later, additional Procedures were prepared to cover Turnover from Projects. Engineering and Construction to Nuclear Operations and Safety Concerns and Reportability Evaluation. Improved specificity of requirements and interfaces and improved flexibility for CP Co Quality Assurance participation on either a primary or overview basis were provided in these Procedures. The management participation in the Task Force strengthened the already strong quality assurance understanding and attitude on the Midland Project. 3.6 CP CO QAD/MPQAD DEPARTMENTAL PROCEDURES I A complete revision to the CP Co Quality Assurance Department Procedures was made during 1979 to be consistent with and to supplement the Volume II Procedures described in Section 3.5, and rp0381-0241a112
6 to provide technical improvements, greater procedural specificity, and added opportunity for CP Co Quality Assurance involvement in . and control of site quality. Twenty-eight of the Procedures were (. revised from existing documents and thirteen Procedures were new. When MPQAD was formed in 1980, these Procedures formed the basis for the MPQAD Procedures. Appendix E provides a list of the
' subjects of these MPQAD Procedures.
3.7 i QUALITY TRACKING AND STATUSING I A computerized tracking system was implemented to provide manage-
- ment with a tool giving visibility to and accountability for the open quality-related action items (this being necessary to assure a disciplined approach to the completion of these items). For each action item entered, the system identifies the organization responsible for the action, the schedule for completion of the action, the status of the action, and the MPQAD staff member who is responsible for follow-up to assure completion of the action and closure of the item.
The Bechtel Quality Assurance organization implemented this system in the last quarter of 1979, but the system is now being admin-istered by the MPQAD. The system has been improved to provide more specificity regarding the types and levels of actions being tracked. Further improvement is being made to provide management with a prioritized, truncated list of actions for each responsible organization. The tracking system enables management attention to be focused on the most significant actions and on the total number of actions for which each organization is responsible. As a result, the number of old outstanding actions has decreased markedly, while the total number of outstanding actions has increased due to the system being made more comprehensive as noted earlier. Appendix F provides a detailed descripton of the strategy and goals of the action item tracking system, the results achieved up through November 1980, and examples of instructions and reports provided by the system. The Project's management team is also placing continued emphasis on reducing the number of all . types of open quality indicators (ie, various types of conconformances, as distinguished from the quality action items discussed above.) To facilitate this emphasis, another system was implemented in the last quarter of 1979 to measure the level and aging of the open quality indicators. Using ! this system, management has reduced the average age of open ! indicators and significantly reduced the number of open indicators. ) In the period of November 1979 to January 1981, the number of open Bechtel nonconformance and deviation reports was reduced by almost two-thirds. Appendix G shows this graphically. A parallel effort has reduced the number of open and outstanding Quality Control Inspection Records (QCIRs). In the 14-month period ( ending January 1980, such QCIRs were reduced from over 22,000 to less than 16,000, representing an improvement in the packaging cf rp0381-0241a112
7 the inspections and in the timeliness of their completion. The l total number of closed QCIRs, representing completed and accepted y [ work, is over 70,000. l 3.8 SUPPLIER DEVIATION DISPOSITION REQUESTS MPQAD performs an in-line review of Bechtel's Nonconformance Reports to assure the adequacy of the dispositioning and closure process. Consistent with this, since August 1980, MPQAD has been I reviewing and approving the disposition and closure process for Supplier Deviation Disposition Requests (SDDRs) on an in-line basis. Previous to this, approval was required of only the Bechtel Engineering and Procurement organizations with "information only" copies provided to the Bechtel and CP Co Quality Assurance organizations. The MPQAD in-line review provides a timely assessment of the discipline applied to the dispositioning process. In addition, the review provides direct feedback to MPQAD as to a given supplier's ability to achieve requirements. The final benefit is that it provides an opportunity for MPQAD to assess and enhance, as necessary, the quality requirements for future orders and to eliminate the root causes of SDDRs. 3.9 FIEI.D PURCHASE ORDERS Historically, Bechtel's Quality Control organization had been l reviewing and approving Field Purchase ' Orders (Pos), primarily to assure that the design and quality criteria previously established by Project Engineering were translated accurately into the Pos. In September 1980, MPQAD replaced Bechtel's Quality Control as the reviewer of field Pos. (This responsibility change is consistent with MPQAD's review and approval of Pos originated at Ann Arbor.) The scope and purpose of the MPQAD review is broader than the Bechtel Quality Control review. MPQAD also assures the technical adequacy of the PO quality assurance requirements, adjusting them as appropriate, to fit current conditionr. 3.10 NEW REG GUIDE IMPLEMENTATION In November 1976, Bechtel Quality Assurance Program for the Midland Project was revised to voluntarily commit the Project to the below listed ANSI Standards and Regulatory Guides (only those marked with an asterisk being a carry over from the PSAR). j ANSI Standard Regulatory Guide-Revision Date ; l
*N45.2-1971 1.28 - June 7, 1972 '
" Quality Assurance Program Requirements for Nuclear Facilities" N45.2.4-1972 1.30 - August 11, 1972
, " Installation, Inspection t
and Testing Requirements for Instrumentation and rp0381-0241a112
- a _ . . _ .
. -- ,_ - . _ - _ . . . - . = . - - - .
8 Electric Equipment During the Construction of Nuclear . [ Power Generating Stations" N45.2.1-1972 1.37 - March 16, 1973 "fleaningofFluidSystems and Associated Components During the Construction Phase , of Nuclear Power Plants" i N45.2.2-1972 1.38 - March 16, 1973 "Psckaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants During the Construction Phase N45.2.3-1973 1.39 - March 16, 1973
" Housekeeping During the Construction Phase of Nuclear Power Plants N101.4-1972 1.54 - June 1973
" Quality Assurance for Protective Coatings Applied to Nuclear Facilities N/A 1.55 - June 1973 N45.2.6-1973 1.58 - August 1973
" Qualifications of Inspection, Examination and Testing Personnel for Nuclear Power Plants" Nw5.2.11-1974 1.64 - Rev. 1, Feb. 1973
" Quality Assurance Requirements for the Design of Nuclear Power Plants.
N45.2.10-1973 1.74 - February 1974
" Quality Assurance Terms and Definitions" N45.2.9-1974 1.88 - August 1974
" Requirements for Collection, Storage and Maintenance of Quality Assurance Records l for Nuclear Power Plants"-
N45.2.5-1974 1.94 - April 1975 ( " Supplementary Quality Assurance Requirements rp0381-0241a112
9 for Installation, Inspection, and Testing of Structural ' f- Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants"
.=
N45.2.8-Draft 3, Rev 4 N/A
" Supplementary Quality g Assurance Requirements
- for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants."
N45.2.12-Draft 4, Rev 1 N/A
" Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants" N45.2.13-Draft 3, Rev 3 N/A l' Quality Assurance requirements for Control of Procurement of Items and Services for Nuclear Power Plants" Examples of implementing procedures that were either originated or revised in response to these QA Program improvements were:
MED 2.13 " Project Engineering Team Organization Responsibilities" EDPI 4.55.1 " Project Material Requisitions, Midland Proj ect" TPG-4.00 " Storage and Storage Maintenance of Equipment and Materials" FPG-7.000 " Housekeeping and Cleanliness Control During Construction"
. PSP-G-7.1 " Documentation, Records and Correspondence Control" 3.11 EQUIPMENT QUALIFICATION REREVIEW The equipment qualification rereview was initiated to assure that equipment qualification tests are consistent with FSAR commitments.
The need for the rereview was identified as a result of two initial actions that were taken concurrently - the issuance of a Problem Alert by Bechtel's San Francisco Power Division and the completion of a special review of the qualification of selected cable by CP ( Co's Quality Assurance Engineering Section, rp0381-0241a112
13 All equipment requiring qualification are being rereviewed. For each equipment, the rereview encompasses a comparison of FSAR requirements, IEEE Standard requirements, and procurement spec-I ification requirements to assure their consistency and adequacy. That completed, a comparison is then made between those require-ments and the actual test procedures and test reports provided by the equipment suppliers. CP Co issued a 50.55(e) Report based on the initial rereview results. This report contributed substan-tially to alert industry of the generic problem of qualification ; inconsistencies and inadequacies. The 50.55(e) Report and the Corrective Action Program preceded by three months the h*RC Bulletin (79-01) which required a review of equipment qualification documentation nearly identical to what was being performed for the Midland Project. The Bechtel Engineering Department Procedures have been improved and specific training has been provided to Engineering and Quality Engineering personnel to help preclude equipment qualification problems for new purchases. The systematic, proceduralized rereview activity is coupled with the statusing and tracking of open corrective action items. Corrective action documentation is also provided as auditable assurance of the qualification of Midland equipment. To date, Foxboro transmitters purchased under Specification 7220-J-204 have been the only hardware items judged unqualifiable. Appendix H provides a detailed description of the equipment qual-ification rereview. 3.12 SPECIFICITY REVIEWS In 1977, CP Co's Quality Assurance Engineering Section initiated a review of specifications to determine the need for their increased specificity, clarity of references to codes and standards, and clarity of wording, supportive of construction and inspection activities. Forty-nine design specifications for fabrication and installation were reviewed. The 49 specifications represented all the active field-oriented specifications; active being defined as significant remaining work to be accomplished to these specifications. This review and the Bechtel disposition of Quality Assurance Engi-neering's ecmments resulted in the revision of 12 specifications for tolerancing and wording improvements; through the comment resolution process, an increased design personnel awareness of the need for specificity in the preparation of future design documents; and an increased confidence in the understandability of the exist-ing design specifications for construction. Also in 1977, the CP Co Quality Assurance Engineering Section undertook a review of the dimensional tolerances for a portion of the Reactor Building Spray System (RBSS) while Bechtel's Engineer-ing Department conducted a parallel review. The object of the parallel reviews was to enable independent assessments and then to ( combine the results for resolution. The purpose of the reviews was to provide confidence that the drawings and specifications contain ! rp0381-0241a112 I l .-- - --
11 the specificity necessary for successful installation and inspec-tion. Forty design documents were reviewed, including drawings for - ( the RBSS installation (typical of drawings for other safety-related installations) and specifications generic to the installation of all safety-related systems. This review confirmed that dimensional toleranes were generally available to install safety-related systems. Improvements were made to seven generic design documents to clarify dimensional tolerances. Again, the review and comment resolution process increased Bechtel Engineering's awareness of the need for specificity and provided additional confidence that tolerance spec-ificity would be incorporated in future design documents. Appendix I provides a more detailed discussion of the dimensional tolerance review. In 1978, a review was conducted of 91 Bechtel Field Change Requests (FCRs) to assess whether Field and Design Engineering had been responsive to the need for specificity in design documents. This review verified that the specificity message was understood and was being addressed. Appendix J provides a more detailed description of the FCR review. Specifications and drawings are subject to a continuing review through the overinspection process which from a hardware orientation viewpoint evaluates the installation and inspection processes required by the design documents. Adequacy of tolerancing and acceptance criteria is specifically addressed in ( the overinspection process. Revisions to specifications are subject to MPQAD review as well as the corresponding changes to the Bechtel quality control instructions. 3.13 PROCUREMENT SUPPLIER QUALITY Over the life of the Midland Project, significant improvements have been made to the overall Bechtel Quality Assurance Program for procurement. Appendix K provides a detailed description of the Bechtel Power Corporation Procurement Supplier Quality Department organization and activities, including special activities which were implemented specifically for the Midland Project. Three of the more significant programmatic improvements are discussed below. Quality Assurance organizations are participating as part of the team to assess and qualify suppliers for the Midland Project. Supplier Quality Representatives are utilized as part of the team to qualify suppliers via the commodity audit at the time of the initial purchase and to perform subsequent supplier audits. (See Section 3.20.) The Bechtel Supplier Quality Group for the Midland Project utilizes the new Supplier Information System and Evaluated Supplier Listing, published by the San Francisco Power Division, as inputs to the establishment of specific procurement quality requirements. CP Co Quality Assurance personnel have performed supplier audits, or in conjunction with Bechtel Supplier Quality, have participated in supplier audits. It is a CP Co Quality ( Assurance Program commitment to do a minimum of 10 supplier audits each year for the Midland Project. rp0381-0241a112
12 These improvements in supplier evaluations have provided increased confidence in a supplier's capability to understand and meet , procurement requirements and have resulted in improved technical ( capability of the audit teams. As a means of facilitating the identification of significant Eharacteristics for inspecting, the Supplier Quality Department has been participating, along with Quality Engineering and Quality Assurance (now MPQAD), in reviewing procurement specifications and I in preparing Procurement Inspection Plans. An MPQAD contractual clause was originated and implemented and a Bechtel procedure was revised to require that applicable inspection witness and hold points be specified in suppliers' inspection planning documents. These improvements have resulted in increased assurance that the requirements are understood by the suppliers and inspection agencies.
~
Quality program verification, which is a form of a mini-audit, has been implemented on the Midland Project to provide a more timely assurance that a supplier's quality assurance program is being effectively implemented. (Again, s : Section 3.20.) Project Engineering, in conjunction with Quality Assurance (now MPQAD), provide the direction for tbe specific program implementation verifications which are made by Bechtel Supplier Quality Representatives. The net result is improved timeliness of verification (progressive) to supplement annual audits. Another benefit is that the Supplier Quality Representatives' capabilities have been improved through their training and participation in ( program evaluation (as contrasted to their being limited to performing only inspection). 3.14 QUALITY VERIFICATION DOCUMENTATION In February 1978, CP Co Quality Assurance engaged Science Applications Incorporated to perform an audit of the B&W (NSSS supplier) quality verification documentation. The results of this audit indicated that a complete rereview of this documentation was necessary, and in conjunction with B&W, CP Co Quality Assurance established the requirements by which to accomplish the rereview. This rereview has been completed, the discrepancies have been dispositioned and corrected, as necessary, and the effectiveness of the process has been verified through additional audits and summary reviews by MPQAD of all quality verification documentation. Confidence has been established that the documentation supports l hardware quality and is ready for turnover to CP Co. . In 1979, a rereview was started of supplier-originated, quality verification documents for Bechtel procured items. The purpose'of the rereview was to provide additional assurance of hardware quality by assuring the adequacy of the supplier quality verification documentation - adequacy with respect to documentation availability, traceability, legibility, and technical content. Supplier quality verification documenation received since July 1978 is subject to a 100% review for adequacy, but documentation k received prior to that time is subject to the rereview on a systematic sampling basis. When the adequacy of a supplier's rp0381-0241a112
13 quality verification documentation is judged to be "indeterminant" from the sampling, 100% of that supplier's quality verification * { documentation is subjected to the rereview. All discrepancies are dispositioned and corrected, as necessary. At the end of February 1981, the rereview was approximately 64% complete with 2,050 purchase order packages dispositioned by the Material Review Board. Appendix L provides a more detailed discussion of this documentation rereview. g 3.15 " FLAGS" REVIEW
- The purpose of the " flags" review is to identify " flags" which may indicate possible product quality concerns in the procurement packages and associated documentation. A " flag" is an adverse condition for which the available documentation does not provide evidence of adequate disposition and/or resolution of the condition. The " flags" review was developed in response to the problems encountered with the Unit I reactor vessel anchor bolts.
Procedures require the " flags" review to be accomplished on a disciplined basis by experienced Quality personnel who have been
. trained specifically for this task. The scope of the review includes Field P0s for which the procurement was made without source inspection and Field and Ann Arbor P0s which, on a judgmental basis, were considered to have higher probability of containing a " flag."
At the end of February 1981, the review is just starting to ( complete a significant fraction of the planned investigation. Twenty " flags" had been identified which require further resolution and disposition although no serious hardware concern has been positively identified. The resolution of the " flags" provides greater confidence in the quality of the procured materials and items. , Appendix M provides a more complete description of the " flags" review, including the procedures and examples of the results of the review process. 3.16 CP CO QUALITY ASSURANCE FOR NEW WORK Selected major procurements were processed through the CP Co
, Quality Assurance Program rather than the Bechtel Quality Assurance Program in order to provide CP Co with direct control of new work.
For the NSSS erection and preservice examination procurements, the CP Co Quality Assurance Department was established as the " primary" Quality Assurance organization rather than an " overview" organization. These jobs are each more than 50% complete. For these jobs, both the execution of the Quality Assurance Program and the suppliers' performance are considered above average based on the low number and lack of significance of the noncompliances. It is anticipated that additional future site work will also be [ executed wholly utilizing the CP Co Quality Assurance Program. l 3.17 CONTROL OF CABLE PULLING rp0381-0241a112
. _ ____ _ . _ . _ _ - . _ _ _ _ _ .4 14 To avoid damage to electrical cables during installation (pulling),
i a computer program for cable pulling force calculations was used as , (f a control mechanism. Based on Field Engineering and Quality Control inputs, among others, this program computes the anticipated pull forces based on field conditions before the actual pulling ' .pecurs. The program considers the frictional forces imparted where one or more bends are involved. Appendix N provides the methodology and equations necessary to develop this computatien. i Appendix N also provides an actual computer printout and schematic drawing for an actual cable run. Construction and quality utilizes the output from this program. The results are used in an assessment of quality attributes by Quality Control personnel prior to every Class IE pull. > r i Success has been achieved in adhering to allowable pulling tensions. This is evidenced by the absence of CP Co Nonconformance Reports and NRC concerns relative to this activity. 3.18 INSPECTION i' Improvements in this area were made by refining the requirements for both CP Co and Bechtel inspector qualification, instituting and increasing the CP Co overinspection activity and refining the ' Bechtel Project Quality Control Instructions (PQCIs). MPQAD personnel who perform inspection and Bechtel Quality Control inspection personnel are certified to requirements which exceed the i ( ANSI N45.2.6 requirements. ANSI N45.2.6 requires only that inspectors be certified on a discipline-by-discipline basis (eg, ' civil, electrical) whereas MPQAD Level II personnel are certified to each specific Inspection Plan that is used on a repetitive basis and Bechtel Level I and II personnel are certified to individual i PQCIs. The Ann Arbor Power Division uses discipline-certified Level III personnel for training and certification. These i improvements in assuring the qualifications of inspection personnel have increased inspection effectiveness. Appendices 0 and P provide further details of these improvements. ! Requirements for the certification of MPQAD nondestructive examination (NDE) personnel meet or exceed SNT-TC-1A criteria as well as ASME Section III and XI criteria for training, experience and visual acuity. , The CP Co overinspection activity was implemented to provide a measure of the supplier's " primary" inspection effectiveness and to provide increased confidence in the quality of the hardware. '. Reinforcing steel and embed overinspection commenced in 1976 and 1977. The overinspection activity was expanded in 1979 to cover
- all discipline activities at the site. The overinspection activity is performed such as to place frontend emphasis on new work and potential problem areas. Appendix P provides the details of the overinspection activities.
k rp0381-0241a112
. _ _ __ ~~
- . . e l 1
I 15 i l
. i A'special plan was implemented for overinspecting Bechtel on-site i radiography (RT) .on a sampling basis and for overinspecting the .
f NSSS erection RT on a 100*. basis. A review program for vendor
- i radiography is also being utilized. Appendix Q provides the details of the RT overinspection.
In 1980, 223 mechanical, 102 civil, 151 electrical and 116
, welding /NDE (excluding RT) overinspections were conducted. Each of i these overinspections corresponds to a work package involving 3
- numerous characteristics and may cover several Bechtel Quality
( Control Inspection Records (QCIRs). When there is a sample size of I a thousend or more and the lot size is at least ten times the j sample size, then the percentage of defects found in the sample closely approaches the percentage of defects that exists in the lot
, as.a whole. The number of overinspections being conducted along l with the many individual characteristics each looks at when compared to the nual.er of primary inspections and the corresponding , multitude of characteristics fit the large sample / lot criteria. ,
The number of individual deficient characteristics found compared to the total number of cha:acteristics looked at during i overinspection substantitiates a general conclusion that the *
- completed construction which has been accepted by the primary inspection agency is in conformance to the design documents.
The overinspection activity provides a timely identification of nonconforming conditions and corrective action in both the construction and inspection processes. Overinspections are scheduled to provide a close review of new activities and any areas where problems have been experienced. This additional inspection l( ~ j layer provides an increase in hardware quality through the identification and correction of specific nonconformances and i l process corrective action and through the verification of the overall inspection effectiveness of the primary inspection agencies. j Bechtel PQCIs were also improved to assure that characteristics important to safety are inspected and to provide increased
- accountability for the required inspections.
Fifty-four PQCIs active in 1977 were reviewed for specificity by j the CP Co Quality Assurance Department. The resulting improvements involved providing clarification of the inspection code callout (Visual, Measure and V&M) - ie, the method of inspection to be used; providing additional detail and clarified instructions by including the " inspection method" in the Instruction document (ie, I the PQCI) rather than in the record document; assuring that important characteristics were covered; providing greater ! specificity as to the meaning of the " surveillance" (S) and !
" review" (R) inspection techniques; and revising the PQCIs to
, eliminate the use of " surveillance" in any final inspection activity. Currently " surveillance" is being replaced with witness , or hold points as the Bechtel in process inspection technique. ! 3.19 QUALITY TREND ANALYSIS !k i
~
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.e<- _. , , , . . . .,.n.--,--m-.,,., . - - , , , , - , . , . - - . , _ . . , . . , - . , , , - - - . . , , , .
16 Trend analysis gives visibility to nonconformances in a given area which are increasing in number or which are remaining at an undersirable high level. It also provides an impetus to the timely ( correction of the root causes of these nonconformances. Appendix R provides a history of the improvements in trend analysis since its initial implementation in 1974. Currently quality indicators are categorized by 15 performance areas and by 4 nonconformance or deficiency codes. There are i separate performance areas for site subcontractors (eg, Zack and B&W). For each performance area, Nonconformance Reports, Quality Audit Findings, Deficiency Reports, Quality Action Requests and NRC Items of Noncompliance are entered into one of the four deficiency codes. Totals are obtained by area and by code and reviewed by MPQAD with special emphasis on detecting indications of any specific process Ieing out of control, and with special emphasis, as well, on detecting gross patterns across all areas and codes. Both a micro and macro approach are utilized for the analysis of the data. The MPQAD Manager is required to make and document a specific
, review of each Monthly Trend Report. If the trend data for a given month exceeds the 4-month trailing average for a specific area, an assessment is required of the need to stop work in that area. The Monthly Trend Report is distributed to the Project's management team.
The present improved trend program is responsive to the need to have a management system which identifies adverse quality trends. 3.20 ALTITS The Midland Project audit activities cover five areas: Bechtel's audits of its suppliers; Bechtel's monitoring of its own activities; Bechtel's management audits; Y_PQAD's audits and CP Co's
" corporate" audits. There have been improvements in all of these areas. The improvements in Bechtel's audits of its suppliers were described earlier in Section 3.13. The improvements in the other four areas are described below.
Bechtel's Quality Assurance monitoring activities began in August 1977 to provide more timely and less formal assessment of procedural adequacy and implementation of repetitive design, construction, and inspection activities. The monitoring activities utilize basic audit elements such as planning, checklists, auditor qualification, reported results, and a closed-loop system for obtaining corrective action. The monitoring activities complement the formal audit and overinspection activities. Since there are more monitoring activities than formal audit activities accomplished in a given period, monitoring supplements the confidence gained through audit in the activities affecting quality. Appendix S provides a detailed description of the monitoring activities. ( The need to increase the frequency of Bechtel's management audits was recognized. The frequency of these audits for the Midland rp0381-0241a112
- . ~ . - - _ . _
e . 17 Project has been increased from once to twice a year. The scope of the management audits is shifting to include auditing for technical compliance as well as for programmatic compliance. To achieve -( this, the programmatic requirements checklists have been supplemented with checklists for technical requirements relating to salculations, design documents and hardware. In addition, Eechnical specialists are included on the audit teams. Appendix T provides additional details pertaining to the management audit activities. Both the MPQAD and CP Co " corporate" audit activities were l'aproved by formal qualification and certification of auditors and lead auditors to ANSI N45.2.23 requirements (with the one exception of not requiring a fixed number of audits per year). These audit activities provide an assessment of the adequacy of the Quality Assurance Program, as well as its implementation and cover all phases of the Project from design through preoperational testing and final turnover to Operations.
4.0 CONCLUSION
The Quality Assurance Program improvements summarized above demonstrate the high level of effort in the Midland Project to comply with the requirements of 10 CFR 50, Appendix B; national nuclear quality assurance standards; and corresponding NRC Regulatory Guides. These improvements also demonstrate CP Co management's willingness to make large up-front investments for quality assurance; willingness to accept changes in the i[ Quality Assurance Program; willingnesss to be kept informed about quality
\_ assurance; to make timely decisions on quality assurance matters; to promote quality assurance throughout the organization and, very importantly, willingness to interact responsibly with the NRC - all excellent indicators of CP Co management's positive attitude about quality assurance for the Midland Project.
WRB/BWM 4/29/81 rp0381-0241all2
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,,, 7, UNITED STATES . . , . ..
NUCLEAR REGULATORY COMMISSION 2 -
.I REGION lil o 8 799 ROOSEVELT ROAD
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( "JUL' ! O 198P
- - - = . :- ,. ;, .
e . 4,B /,'t c.s y s. Docket No. 50-329 m C < ' ' ' d g,Q
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Docket Nd"50-330 ,--- (. Z ,,'s. g 'f' J.p ' ,#c[r
- f. *
- Consumers Power Company ;
ATTN- Mr. James W. Cook # g,, s
,/O Vice President ;A ,
Midland Project /-".:.s\g, 1945 West Parnall Road Jackson, MI 49201 Gentlemen: This refers to the special team inspection conducted by Mr. C. C. Williams and other members of this office on May 18-22, 1981, of activities at the Midland Nuclear Plant, Units 1 and 2, authorized by NRC Construction Permits No. CPPR-81 and No. CPPR-82 and to the discussion of our findings with you and others of your staff at the conclusion of the inspection.
- This was an indepth inspection to examine the implementation status and effectiveness of the current QA Program, to determine whether previously
(<~ identified quality assurance problems were sufficiently precluded from occurrance in other areas, and to ascertain whether management involvement in the QA Program was sufficient and effective. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of pro-cedures and representative records, observations, and futerviews with personnel. During this inspection, certain of your activities appeared to be in non-compliance with NRC requirements, as specified in enclosed Appead c A. A written response, submitted under oath or affirmation, is required. Although eight items of noncompliance were identified during this inspec-tion, it is our judgement that the scope and depth of this NRC inspection was such that the identified noncompliances do not contravene our conclusion that Consumers Power Company has established an effective organization for the management of construction and implementation of quality assurance at the site. The noncompliances associated with the piping support area of the inspection were considered the most adverse findings, and precipitated the issuance of an Immediate Action Letter (IAL) dated May 22, 1981. (Attached as Exhibit A to this report.) s ( // a -OMPe46464 010710 PDR ADOCK 05000329
- Q PDR
Appendix A
- NOTICE OF VIOLATION Consumess Power Company Docket No. 50-329 Docket No. 50-330 As a result of the inspection conducted on May 18-22, 1981, and in ac-cordance with the Interin Enforcement Policy, 45 FR 66754 (October 7, 1980), the following violations were identified
- 1. 10 CFR 50, Appendix B, Criterion XVI states, in part, that " Measures shall be established to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective materials and equipment, and nonconformances are promptly identified and corrected...the identification of the significant condition adverse to quality, the cause of the condition, and the corrective ,
action taken shall be documented and reported to appropriate levels of management." Consumers Power Company Program Policy No. 16 Revision 9, Para-graph 1.0 states, in part, " corrective action is that action taken to correct and preclude recurrence of significant conditions adverse to the quality of items." Consumers Power Company Quality Assurance Procedure M-2, dated March 2, 1981 requires the Midland Project Quality Assurance trend analysis be implemented. Specifically, for each perfomance area identify trends requiring corrective action, determine the sources of these trends and obtain appropriate corrective action commitments. Corrective action commitments are the responsibility of the " Appropriate Individuals." Contrary to the above, a review of Monthly Trend Analysis Reports and related documentation covering the period July 17, 1980 - March 31, 1981 revealed that appropriate site managers have not routinely established comprehensive corrective actions in response to the identification of adverse quality trends. Moreover, evaluations of adverse trends have not routinely identified the root causes of nonconformances. For example, 22 instances of construction personnel bypassing QC hold points were included in monthly trend analysis, but an adequate analysis to identify the root cause of these occurrences was not performed, (329/81 '.2-04; 330/81-12-04) This is a Severity Level IV violation (Supplement II). n'^'C'0'72 810710 PDR ADOCK 05000329 ' Q PDR
Appendix A l The Consumers Power Company Quality Assurance Program Policy No. 5, ( Revision 9 states, in part, " Instructions for controlling and per-
, forming activities affecting quality of equipment or operations
! during the design, construction...pbases of nuclear power plants, 6 suth as... construction, installation...are documented in instruc-
! tions...and other forms of documents," and the responsible CP depart:nents shall "also verify through audits that the required .
Instructions...are implemented." ' i Contrary to the above, seven large bore pipe restraints, supports, and anchors were not installed in accordance with design drawing . and specification requirements. (329/81-12-11; 330/81-12-12) ' This is a Severity Level V violation (Supplement II). l l
- 5. 10 CFR 50, Appendix B, Criterion X states, in part, "A program for inspection of activities affecting quality shall be established i and executed by or for the organization performing the activity to
{ verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity." The Consumers Power Company Quality Assurance Program Policy No.10,
- Revision 8 states, in part, " Inspection and surveillance are performed j to assure that activities affecting quality comply with... design documents."
1
!\. Contrary to the above, licensee construction quality control inspectors inspected and accepted six of seven large bore pipe restraints, supports, and anchors that, in fact, had not been installed in accordance with !
design drawings and specifications as determined by the NRC inspector. (329/81-12-12; 330-81-12-13) - e This is a Severity Level V violation (Supplement II). i l 6. 10 CFR 50, Appendix B,' Criterion III states, in part, "the design control j measures shall provide for verifying or checking the adequacy of design, l
- such as by the performance of design reviews... Design control measures shall be applied to items such as... stress analysis..."
lS 4 The Consumers Power Company Quality Assurance Program Policy No. 3, Revision 9 states, in part, "The design organization identifies the l applicable regulatory requirements, design bases, codes and standards; develop the design and specify the design interfaces; perform design i verification and prepare design documents." ) Contrary to the above, several of the small bore pipe and piping i ! suspension systes designs performed at the site had not been prepared, f reviewed and approved in accordance with established design control j procedures. (329/81-12-13; 330/81-12-14) ] , This is a Severity Level IV violation (Supplement II). k ) o
. - . .- _- _. =_ . - - - . - .
Appendix A { taken to avoid further noncompliance; and (3) the date when full compliance will be achieved. Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, this response shall be submitted under oa,th or affirmabion. Consideration may be given to extending your response time for good cause shown. i
, )
~ ~.
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,9%! w. t > \1 r A e _ -
I Dates Jumes G. Keppler JJ - Director j i e ( 4 7 i
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III ( Reportsfo. 50-329/81-12; 50-330/81-12 Docket Nos. 50-329; 50-330 Licenses No. CPPR-81; CPPR-82 . Licensee: Consumers Power Company 1945 West Parnall Road Jackson, MI 49201 Facility Name: Midland Nuclear Plant, Units 1 and 2 Inspection At: Midland Site, Midland, MI Inspection Conducted: May 18-22, 1981 Inspectors: d C.v?:2tk C. C. Williams, Team Leader .[. ~M~ R. J. Co k 'N' ( T . aa W (IE Headquarters r14
@ C ' M l L % n f r~
R. N. Gardner 7 - /6 ~ o. , [
#$ a,,,g;._ }l C. E. Jones '
k'.$ ZH 4V % 7 'IU - 2/ R. B'. Landsman 0 C SGb'( t L+s-~~* R. S. Love f f. 7 ' /c -- E i M. <!ly f. r ,' E. R. Schw & etbinz s / I. T. Yin h* 71 l tc .'7/ Other Accompanying Personnel: J. G. Keppler (May 21-22, 1981) R. C. Knop (May 21-22, 1981) k 0 ; ^, 2 0 ^, ^ ^ ^ 810710 PDR ADOCIO5000329 Q PDR 19g-
4 h DETAILS i(
- Persons Contacted
; consumers Power Company
*D. B. Miller, Site Manager ,
i *J. W. Cook, Vice President, RE&C t l *G. S. Keeley, Project Manager 5 j *J. Wood, Section Head, QAS } *D. R. Keating, Section Head, IE&TV
*H. P. Leonard, Section Head, QAE ;
*K. E. Marbaugh, Operaticas Quality Assurance Superintendent
, *R. E. McCue, Technical Supertindent
, *N. Ramanujas, Senior Geotechnical Engineer i
*T. C. Cooke, Project Superintendent
*D. E. Horn, Civil Group Supervisor, QAE
*G. B. Slade, Assistant Site Manager
- l *D. W. Turnbull, MPQA, Site Superintendnet ;
*L. R. Howell, MPQA ;
j *M. J. Schaeffer, MPQAD, Electrical Engineer, Supervisor ;
*W. R. Bird, MPQAD, Manager j *E. Jones, Electrical Group Supervisor, IE&TV, MPQAD j S. Love, Sub Contracts Engineer R. Whitaker, QA Engineering Supervisor !
/ R. Sevo, MPQA l J. Decker, NDE/ Welding Supervisor, MPQAD j(- I - Bechtel Power Corporation I t ) *J. A. Rutgers, Project Manager
*A. J. Boos, Assistant Project Manager t j, *W. D. Greenwell, AAPD Manager of Quality Assurance ,
*M. A. Dietrich, PQAE i
*L. H. Curtis, Project Engineer j *L. A. Driesbach, Assistant to the Project Manager J *D. L. Daniels, Chief Construction Quality Control Engineer
- *E. Smith, PTQCE -
i *L. Davis , Site Manager { L. Snyder, Resident Quality Engineer l E. Urbanawiz, QC Engineer i F. Almeida, Small Bore Resident Piping Design Engineer l l R. Myers, Large Bore Resident Piping Design Engineer ; l F. G. Young, Small Pipe and Hanger Group Supervisor ' 4 W. J. Creel, Senior QC Engineer P. Corcoran, Resident Assistant Project Engineer i A. McClure, Quality Assurance Engineer , l J. Hockwater, Civil Resident Engineer ! 4 T. K. Subramanian, QA Program Engineer i C. Webb, Tech Aid Corporation Assigned to Bechtel Project Resident Engineer
'k
) l !
the erf tria for writing change notices. Inter-Office Memo (IOM) 0-2707 was issued to document and reiterate the recommended training in the proper use of Specification G34(Q). Resident - f Engineers have been assigned to the Field Engineering Office to provide closer supervision on proposed design changes. This ites is closed.
- d. (Closed) Unresolved Ites (50-329/79-25-01; 50-330/79-25-01):
Cable separation violations in Motor Control Centers (MCC) 0B65 and OB66. Nonconformance Report No. 2765 was closed on May 12, 1980 and Bechtel Quality Action Request No. SD-293 was closed on May 14, 1980. These two documents pertained to the cable separation violations in the two MCC's. The inspector observed that the cable separation in the aforementioned MCC's conformed to the requirements of Bechtel drawing E47(Q). This ites is closed.
- e. (Closed) Noncompliance Item (50-329/80-01-01; 50-330/80-01-01):
Failure to have test procedures for soils work activities. The inspector reviewed QCP-10, Revision 1, dated March 16, 1981 and determined the following: (1) With respect to the first issue, Section 4.1 states that the vibrator control shall be at maximum control dial setting. However, the procedure should state to determine at what setting gives maximum density for each soil type to be used. The licensee agreed to revise this section accordingly. (- (2) With respect to the second issue, the inspector reviewed laboratory gradation data performed on material tested before and after compaction and determined that the change in gradation is insignificant. (3) With respect to the third issue, the inspector reviewed SCN 7220-C-211(Q) - 11002 dated May 12, 1981 which added the density testing depths to the specification. This item is closed. -
- f. (Closed) Unresolved Item (50-329/80-01-02; 50-330/80-01-02):
Failure to have soils laboratory forms under complete document control. The inspector reviewed QCP-10, Section 5.0 which was added to address documentation and distribution of soils labora-tory forms. The inspector also reviewed new Procedure QCP-14, Revision 0, dated February 12, 1981 which addresses the QC procedure for use of these forms. Th!- item is closed. 3 (Closed) Unresolved Ites (50-329/80-01-03; 50-330-80-01-03): Failure to have explicit instructions for the onsite geotechnical engineer's review of test results. The inspector reviewed new Procedure EDPI 2.14.7, Revision 0, dated May 14, 1981. Exhibit D to the procedure indicates how the onsite geotechnical engineer is to perform his review and document his review. This item is closed. {
i l
- 2. 10 CFR 50.55(e) Reportable Items
)
- a. (Closed) 10 CFR 50.55(e) Reportable Ites (329/78-13-EE; ;
( 330/78-13-EE): Undersized Wire Installed in the Control Room Makeup Filter Drain Heater Units, Nos. OVM-78A, 0VM78B, OVM94A, and OVM94B. The final report on this deficiency was received by the RIII office (Howe 78-79) on March 9, 1979. This letter stated, in part, that the undersized wiring would be re-placed by Bechtel under the supervision of Mine Safety Appliances (MSA), the heater unit vendor. The report also stated that the progress and closure of this operation (wire replacement) would be tracked through Nonconformance Report (NCR) No. 1733. The undersized wire was replaced and the installation inspected by Bechtel Quality Control. NCR No.1733 was closed on February 11, 1980. During an over-inspection by CPCo, Quality Assurance (QA) on October 20, 1980, it was noted that type RHH wire had been used to replace the undersizes wire rather than type TA or SIS wire which is specified in Specification 7220-M-150(Q), Revision 4. NCR M-01-4-0-067 was initiated to document these observations. i on December 5, 1980, MSA initiated Supplier Deviation Disposition Request (SDDR) No. 7. Bechtel SDDR No. 1967, requested type RHH wire be approved as equal to type TA or SIS wire. Bechtel Engi-neering approved the above referenced SDDR on January 11, 1981. This item is considered closed,
- b. (0 pen) 10 CFR 50.55(e) Reportable Ites (329/78-12-EE; 330/78-12-EE): Inadequate Crimping in Vendor Supplied
, Electrical Penetrations. The final report on the inadequate crimping of cable / wire terminal lugs in the inboard terminal boxes of Amphenol Sams/ Bunker Ramo supplied penetrations was received by the Region III office (Howe-153-79, dated May 25, 1979). CPCo prepared NCR No. M-01-4-8-107 and Bechtel prepared Management Corrective Action Report (MCAR) No. 26 for tracking this deficiency. The type of discrepancies noted on the MCAR were: wire not fully penetrating the lug barrel; crimps not tightly made; barrel of lug collapsed . preventing full wire compression and connections loose on the terminal block. An attachment to the final report states, in part, "During the April 30, 1979 through May 3,1979 inspection, all unsatisfactory terminations were reworked and passed further inspections and pull tests." In addition, all terminations were checked for: (1) Proper type of lug; (2) Proper lug indentation; (3) Tightness to terminal blocks. Bechtel Field Inspection Report (14 pages) docu-ments the rework and acceptance of the inspections and pull tests described above. CPCo Project Inspection Plan and Report No. 001, dated April 16, 1981, documents MPGAD's overinspection and accept-ance of the terminations in the inboard terminal boxes of the 26 Class 1E Electrical Penetrations. NCR No. M-01-4-8-107 was closed on June 8, 1979 and MCAR No. 26 was closed on June 26, 1979. (
Section I
- Prepared By: E. R. Schweibinz
[ Reviewed By: C. C. Williams
- 1. Scope of Persons Interviewed and Areas Reviewed The following Consumers Power and Bechtel personnel were interviewed. I I
Consumers Power Company (CPCo) , . Vice President, Midland Project l Site Manager Construction Superintendent Environmental Services and Quality Assurance Manager Site Quality Assurance Manager Site Quality Assurance Superintendent Bechtel l
! Project Manager I Site Manager Project Quality Assurance Manager Project Field Quality Control Engineer ; Lead Pipe / Mechanical Quality Control Engineer Several QC Inspectors The majority of the above personnel were interviewed separately by 4
( , a two man team from Region III. This team consisted of a reactor inspector and a section chief. In addition, the team met with several !
- of the individuals collectively. These interviews were made to assess
- the capability, attitudes, and functional adequacy of the personnel and to verify adequate and effective management involvement in the imple- ,
l mentation of the site quality assurance program relative to its status, 6 problem solving methodology, and the adequacy of resource support. No items of noncompliance were identified. i 2. Problem Areas Identified .
- a. Site construction and quality management personnel are not sufficiently sensitive to symptoms of inadequacy identified by their program and other sources as evidenced by the follow-ing summary of findings in other sections of this report:
1 (1) The licensee is not routinely making comprehensive
- evaluations of root causes of problems.
(2) When problems are identified in an area, the licensee continued working in that area and did not always expedite effective corrective actions.
- b. The Region III inspectors identi'le? a need to be more specific in the administration and organizational relationships of the
; i 1
SECTION II Prepared By: R. J. Cook C. E. Jones l Reviewed By: D. C. Boyd, Chief Reactor Projects Section IA
- 1. Objectives of the Inspection ,
The inspectors objective was to verify that current Quality Assurance Program description and implementation met requirements of 10 CFR Part j 50, Appendix B, and other licensee commitments. The critical elements of the objective were accomplished as follows:
- a. Verify that changes in QA Programs and Organization effective August 1980 and (reported to the NRC on March 13, 1981) are in place and adequate.
1
- b. Assess / evaluate the magnitude of previously reported breakdowns in Quality Assurance, t
{ c. Verify adequate and effective management involvement in the I j implementation of the site QA Program. l
- 2. General Areas Inspected In general the inspectors reviewed selected examples of the following
' - documents compiled by Consumers Power Company, Bechtel, and Babcock and Wilcox: ) a. Audit Finding Reports i 4 b. Nonconformance Reports (NCR's) and Nonconformance Report I.ogs j c. Quality Action Requests (QAR) i d. Corrective Action Requests (CAR) . l In addition the inspectors reviewed documents selected at random to l examine for corrective action, review and approval by authorized Quality
- Assurance and Engineering management, referal to Engineering Design and i the timeliness of clearing the probles.
i No items of noncompliance were identified.
- 3. Review of NSSS Nonconformance Reports (NCRs)
During the team inspection period, the Resident Inspector examined non-
, conformance reports issued by the NSSS contractor, B&W, and transmitted I to the NRC by virtue of the requirements of ALAB 106. Approximately 15 ,
(
, . - . m- - -
s-,,n---, , . . , - _ --.a,_p.c., ,--.n n - n ,. - - - p- . , -- .e.,
f Based on the above considerations, this item is considered an unresolved matter pending further review by the NRC. (329/81-12-02; 330/81-12-02) {
- b. Core Sapport Assembly Guide Block Positionina and Weldina During the team inspection, the Resident Inspector inquired as to the status of NCR's which might have been generated as a result of welding the core support assembly (CSA) and the subsequent motion g of these guide blocks. The Resident Inspector was aware that movement of nominally 0.030 to 0.040 inch had occurred between each pair of blocks during the welding operation. The motion was shared by each block of a given pair. Prior to welding, each block is fitted with an interference fit 1 5/8 inch diameter pin which engages the guide block and the core support assembly barrel.
The licensee stated that no NCR's had been initiated with regard to the motion of the guide blocks because the procedure referenced, PCA-58, Guide Block Positioning, required B&W to report any devia-tion to their Nuclear Service group when the expected gap exceeds the criteria. The site B&W representatives did report the final position of the guide blocks after welding. Consumers Power Company has issued a letter dated May 17, 1981, requesting information from B&W pertaining to the stresses induced when the guide blocks moved relative to the pin during the welding operation. This iten is considered unresolved pending further evaluation of the engineering data associated with notion of the guide blocks. (329/81-12-03; 330/81-12-03) k' No items of noncompliance were identified.
- 4. Review of Consumers Power Company Nonconformance Reports (NCR's)
During the team inspection period, the Resident Inspector examined in detail approximately 15 Nonconformance and/or Audit Findings Reports which were generated by the licensee (Consumers Power Company) and transmitted to the NRC by virtue of AI.AB 106. From this sampling, the trend which was noted was that the disposition to these NCR's generated by Consumers Power Company appeared to be rigorous and the action justifiable. No items of noncompliance were identified.
- 5. Review of Bechtel Corporation Nonconformance Reports (NCR's) .
During the team inspection, the Resident Inspector reviewed approxi-mately 75 Nonconformance reports initiated by Bechtel Power Corporation and transmitted to the NRC by virtue of ALAB 106. No discrepancies were noted in the disposition of these NCR's and there appeared to be a justification for these NCR's with a "use-as-is" disposition. No items of noncompliance were identified. ( a reduction in processing time of approximately five and a half months. The NCR's were audited for disposition since the processing time had been reduced and many NCR's were dispositioned by the Field Engineer ( to "use-as-is". Those selected for review appeared to be properly dispositioned in accordance with approved project procedures. In general the NCR's prepared after August 10, 1980 indicated more care in their analysis, documentation, and a noticable improvement in the timeliness of the NCR processing. No items of noncompliance were identified. t
- 7. Conclusions The inspectors reached the following conclusions during the review of plant documentation records and discussions with personnel from Consumers Power Company and major contractors onsite. The conclusions are as follows:
- a. Questions were raised regarding the dispositioning of specific B&W NCR's observed during the review of those initated during the selected time periods. Verbal Response from B&W and Bechtel resolved these questions. Other questions discussed in Para-graphs 3.a and 3.b of this section remain to be answered.
- b. Nonconformance Report resolution time was reduced from an average of eight months to two and a half months during the time of the past 12 to 18 months. Those processed presently are more complete than the earlier examples selected.
- c. Technical evaluatnions appeared to be adequate. In general evaluations of NCR's dispositioned "use-as-is" were reviewed with special emphasis and observed to meet requirements.
- d. NCR's processed recently are more comprehensively responsive to the project's governing procedures.
No items of noncompliance were identified. ( of the complex remedial activities, additional qualified staff will be available to participate in these activities, f No items of noncompliance were identified.
- 2. Trend Analysis and Evaluation A review of the Midland Quality Assurance trend analysis for the period of July 1980 through March 1981 was performed to verify that the requirements of Consumers Power Company Procedure M-2, Revision 1, dated March 2, 1981 have been implemented.
The Trend Analysis Procedure M-2 defines a trend as follows: A single or multiple occurrence of the magnitude defined in 10 CFR 50.55(e). A single or set of circumstances which warrant actions beyond the normal quality program to reverse a situation that is adverse to quality. When the current sonth's data exceed the four month trailing average of the data for the individual performance area. The procedure required for any of the trends identified above that a summary of corrective actions taken or the rationale for no cor-rective action be included with the trend report. It further specifies that MIQA personnel shall obtain appropriate corrective action commitments from the appropriate individual. (
- a. The following specific findings were made as a result of the trend analysis review
(1) Monthly Trend Analysis Report, July 17, 1980 to August 20, 1980 indicates a negative trend in the Mechanical area (Chart C) which shows an increase in deficiencies from approximately 12 to 75. The evaluation states, "It is therefore recocunended that subject supervision be given a review of this report and instructions and indoctrination in the improve:sents of such deficiencies." Letter dated September 8, 1980 required corrective action by Bechtel Power Company Site Manager to preclude recurrence. No response from the Site Manager nor corrective action docu-ment could be located and it was concluded by the licensees representative and the inspector that it had not been written. Chart C2 showed an increase in deficiencies from j two to 60. Therefore, no evaluation regarding the cause ' of drawing and specification tolerances being exceeded in , I the mechanical area was made. (2) Monthly Trend Analysis Report, August 21, 1980 to September 17, 1980 continued to indicate a negative trend on Chart C2 (Mechanical drawings and specification tolerances exceeded). The evaluation simply stated, "The quantity is expected to 1 l l l
effective." Since QAR F-033 action was not taken until March 31, 1981, it would seem unlikely that this action caused the reduction in deficiencies between March 1, 1981 I and March 31, 1981. i b., Based on the above review of trend analysis reports, the following
. has been concluded:
(1) Adverse trends have been identified without' a(equate re-sponse or corrective action from appropriate site managers. (2) Evaluations by QA have not been adequate and have not identified the " root cause" of the increases in deficiencies. (3) Routinely, increases in adverse trends are attributed to increases in production and inspection activity while decreases are attributed to corrective action. However, the trend reports do not substantiate these conclusions, and do not identify the real underlying causes (i.e., inadequate training, instructions, directions, etc.). (4) There was no evidence of stop-work consideration by the QA manager even with substantial increases in the occurrence of deficiencies in the electrical and mechanical work areas. Based on the foregoing, Consumers has not implemented the trend analysis program as required by Procedure M-2 in that appropriate corrective action commitments were not established by the appro-( priate individuals, resulting in failure to take comprehensive corrective action. This failure to take adequate and effective corrective actions as a result of the trend analysis indications, is an item of noncompliance, contrary to 10 CFR 50, Appendix E, Criterion XVI. (329/81-12-04; 330/81-12-04) i After the above findings were brought to the attention of the Consumers site QA superintendent, it was ascertained that the trend analysis program has been the subject of review. The site QA superintendent produced a memo dated May 19, 1981 which identified further weaknesses. These included the description of the trend categories, judgement in assigning trend codes and the variety of evaluations of the monthly trends. This memo proposed a revision to MPQAD Procedure M-2. Except as noted above no items of noncompliance were identified.
- 3. Nonconformance Report Reviews The inspector reviewed all civil NCR's closed by QAE during 1981.
These NCR's were opened between May 30, 1980 and April 24, 1981. The closed NCR's were dispositioned appropriately except forx22 repetitive NCR's regarding construction personnel passing QC in- , spection hold points for concrete expansion anchors. EighteenQC ' inspection hold points were " passed" during the month of March 1981. It was subsequently learned by the inspector that these 18 NCR's ( j - -
- b. "Preplacement, Placement and Curing Inspection of Grouting and Drypacking (Baseplates, Column Bases and Equipment Bases)"
No. 2 dated January 26, 1981 thru No. 11 dated April 27, 1981. I
- c. "Preplacement, ?lacement and Curing Inspection of Grouting and y Drypa: king (Dowels and/or Anchor Bolts)" No. I dated January 26, 1981-thru No. 10 dated May 5, 1981.
The overinspection plans reviewed covered their subject manner , comprehensively and were being implemented adequately. No itses of noncompliance were identified.
- 6. Permanent Dewatering System CPCo plans to install 20 of the permanent dewatering wells by the 1 service water structure to be used temporarily for construction dewatering of the remedial fix on the service water structure. The preliminary drawings and specifications were reviewed. The following concerns were discussed with the licensee:
I
- a. It was indicated that the wells are to penetrate five feet into the underlying till (clay) layer. However, the drawings are
- 2. unclear in this area. The licensee agreed to add this on the drawing.
- b. Supplemental borings are to be drilled at every fourth well to verify the aquifer grain size and the required length of well i
( screen. However, there was no indication in the specification to allow the well design (i.e., the slot size of the screen and its length) to be altered by the new borings. The licensee agreed to include this provision in the specification.
- c. The PVC plastic well casing is not classified as safety-related; however, the licensee agreed to include the casing on the Project Quality Control Instruction to verify that the proper material is being installed in the well.
- d. The drawings indicate a five foot blank piece of casing on the lower end of the well below the screen. The design of this was questioned in that the well could pull the water table farther down if the screen extended all the way to the bottom of the well.
The licensee agreed to review this matter,
- e. The drilling operation did not address the fact that the hole should be kept full of water to diminish the possibility of hole blow-in below the water table. The licensee agreed to evaluate this concern.
The above five items rema'n i open pending the licensee's response. (329/81-12-05; 330/81-12-05) . No items of noncompliance were identified. (
. _ ___ _ _ _ .- . _ _-_-- ._ _ _ _ _ _ ~
to review these plans to ensure that quality items are covered com-prehensively. The inspector reviewed MPQAD Procedure E-2M, dated March 2, 1981, which delineates how to perform the review, and found I that it was being utilized accordingly by MPQAE to review the PQCI's for quality items. No-items of noncompliance were identified. 1 d (- , f 9
I-1 (2) Tool No. BPC 2716; type MR 8-4 Date Certified: 2/5/81 [ Recertification Due Date: 8/5/81 4 (3) Toc 1 No. BPC 2671; type MR 8-4 I
. Date Certified: 12/2/80 Recertification Due Date: 6/2/81 The pertinent calibration records for the aforementioned crimping ; tools were reviewed and found to be clear, retrievable and well I
- maintained. Personnel involved in the calibration process were !
interviewed and found to have a good knowledge of the requirements L for calibrating such tools. Each crimping tool is checked monthly
- and recertified every six months.
No items of noncompliance were identified.
- d. The Region III inspectors observed completed and inprocess 1 Class IE 600 Volt cable terminations in the control room, service water pump house and in the general plant area. Ter-i minations were observed in the following panels and cabinets:
OC20(75); IC24(50); IC11(30); 1Y32(14); 1B23(25); OC180(100); { 2B64(10); IB64(10); OB64(15); and IB56(40). (The number in
- parentheses indicates the approximate number of terminations checked in that panel / cabinet). The following cable scheme numbers were selected at random for a follow-up review of the cable pull cards, QC inspection records and termination landing I
.[ points as compared with drawing E900, Revision 49: IBB5606C, i \ IBB5621F, IBB5631E, IAB6302G, OAY3301A, OAY3303A, OBV041D and OEW21K. No items of noncompliance were identified. j 2. Qualification of QC Inspectors - Electrical During a review of Consumers Power Company (CPCo) initiated Non-conformance Reports (NCR), Quality Action Requests (QAR) and Audit Finding Reports (AFR), it was noted that MPQAD was identifying numerous j noncompliances in items that had been previously inspected and accepted 4 by Bechtel Quality Control inspectors. As a sample, the following documents were selected for follow-up: AFR No, M-02-01-1-06 dated January 27, 1981; QAR No. F-028 dated February 19, 1981; NCR No. M-01-9-1-014 dated February 27, 1981; NCR No. M-01-9-1-016 dated March 24, 1981; NCR No. M-01-9-1-026 dated April 21, 1981 and NCR j No. M-01-9-1-045 dated May 6, 1981. The Region III inspectors requested that the Bechtel Project Quality Control Engineer (PQCE) provide the names and records of the QC personnel involved with the aforementioned nonconformance reports.
- The personnel qualification and training records of three QC inspectors J were reviewed and compared to the requirements of Regulatory Guide 1.58 l
j and ANSI N45.2.6. Following is a summary of the personnel records re- .' ' viewed: 4
-- __ _ _ , _ . . _ _ . _ _ _ _ . _ . _ . . - _ _ . . , . _ . . _ _ _ _ _ . _ _ _ . _ , . _ . ~ - - _ _ . - _ , . _ _ . _
4/7/81 Certified Level I to PQCI E-5.0 " Cable Termina-tion." Two (2) hours of documented training. 4/30/81 Certified Level I to PQCI E-2.0 " Installation of Cable Tray and Wireway". Two (2) hours of docu-mented training. . 5/1/81 Certified Level I to PQCI E-2.1 " Tray Supports." Six (6) hours of documented training. > Inspector "C" No previous QA/QC experience Education: BSEE 9/80 (No transcript) Experience: Miscellaneous parttime work 12/1/80 Date reported on board s
~
12/23/80 Certified Level I to PQCI E-4.0 " Installation of Electric Cable". Five (5) hours of documented training. 12/23/80 Certified Level I to PQCI E-5.0 " Cable Terminations." Six (6) hours of documented training. 3/26/81 Certified Level I to PQCI C-1.50 " Installation and Testing of Expansion Anchors". Two (2) hours of documented training. 5/15/81 Certified Level I to PQCI E-2.0 " Installation of Cable Tray and Wireway". Two (2) hours of docu-mented training. Discussions with the licensee's contractor (Bechtel) PQCE indicated that all QC inspectors are certified on the basis of an oral examina-tion plus observations of the individual in the field. This type of examination does not provide for an after-the-fact evaluation of the inspector's knowledge or the thoroughness of the examination. The Region III inspectors informed the licensee that while it was fully recognized that the requirements for education and experience are not absolute, the intent of Regulatory Guide 1.58 and ANSI N45.2.6 is that the individual have the required education and prior related experience in quality assurance, including testing and/or inspection of equivalent construction and installation activities, or documented objective evidence (i.e., procedures and record of written tekts) demonstrating that the individual indeed does have " comparable".ce " equivalent" competence to that which would be gained from having the required education and experience. (
- a. QAR No. F-032, dated March 25, 1981, identified that the Electrical Construction Quality Trend Graph B-2 for the period of January 22, 1981 thru February 17, 1981 showed an increase in deficiencies over .
l those of previous periods. The indicated cause for this increase was construction not assuring completion of and/or not installing the items per drawings and specifications prior to reporting the
" item complete. Construction was requested to take corrective ac-tion to reduce and/or eliminate these deficiencies in the future and to provide MPQAD with a response that states the corrective '
action to be taken. Examples of items identified were: (1) Threads not coated. i (2) Unapproved coatings. (3) Uninsulated conduit bushings. (4) Anchor bolt problems. (5) Too many bends between pull points. (6) Exceeding the maximum cable pull tensions. The reported action taken was to instruct construction to make a closer inspection of raceway prior to sign-off and reporting the item complete. A contributing factor identified by construction was the increase in production by a factor of two. QAR F-032 was closed on April 13, 1981. The Region III inspectors noted that the B2 Graph for the period of March 1, 1981 thru March 31, 1981 showed a decrease in the number of deficiencies.
- b. QAR No. F-033, dated March 25, 1981, identified that the Electrical Construction Quality Trend Graph B3 for the period of Janurary 22, 1981 thru February 17, 1981 showed an increase in deficiencies over those of previous periods. The indicated cause for this increase was Field Engineering not assuring the completion of work prior to reporting the item ready for final inspection. Field Engineering was requested to take corrective action to reduce and/or eliminate these deficiencies in the future and to provide MPQAD with a re-sponse that states the corrective action to be taken. Examples of areas identified were:
(1) Anchor bolts. (2) Supports. (3) Coating of welds. (4) Separation. (5) Cable splices. (6) Cable tie downs. The reported action taken was to instruct Field Engineering to make a closer inspection of items prior to sign-off and turnover to Quality Control for acceptance inspection. A contributing factor identified was the increase in production by a factor of two. QAR F-033 was closed on April 13, 1981. (
SECTION V [ Prepared By: I. T. Yin Reviewed By: D. H. Danielson, Chief
, Materials and Processes Section
- 1. Review of Procedures and Specifications In conjunction with observation of large bore pipe system installations and inspection of small bore piping design activities at the site on May 18-21, 1981, the inspector reviewed the following Bechtel procedures and specifications, and had no adverse comments:
. QCI C-1.50, " Installation and Testing of Expansion Anchors,"
Revision 7, dated July 29, 1980.
. QCI P-2.10, " Pipe (Component) Supports Installation,"
Revision 6, dated April 21, 1981.
. QCI P-2.00, " Pipe (Component) Supports Final Setting,"
Revision 5, dated April 13, 1981.
. Bechtel Engineering De'partment Project Instruction EDPI-4.46.9,
" Project Engineering Review of Field Mark-up Working Prints (Redlines)," Revision 0, issued on November 7, 1980. '
Bechtel Technical Specification 7220-M-366(Q), " Field Fabrication ( . of ACME Section III Pipe Supports, Hangers, and Restraints for 243 Inch and I.arger Piping in a Nuclear Power Plant," Revision 3, dated May 13, 1980.
. Bechtel Technical Specification 7220-C-305(Q), " Design, Furnishing, Installation and Testing of Expansion Type Concrete Anchors," '
Revision 13, dated December 30, 1980.
. Bechtel Technical Specification 7220-M-326(Q), " Installation, Inspection, and Documentation of ASME Section III Pipe Supports, Hangers, and Restraints for Piping in a Nuclear Power Plant,"
Revision 6, dated February 6,1981.
. Bechtel Technical Specification 7220-M-343(Q), " Design, Documenta-tion, and Field Fabrication of ASME Section III Pipe Supports, Hangers, and Restraints for Pipe 2 Inch or Smaller," Revision 6, dated November 24, 1980.
. Bechtel Engineering Department Procedure, EDP-4.37, " Design Calculations," Revision 2, dated May 27, 1976.
. Bechtel Manager of Engineering Directive, MED-4.37-0, " Design Calculations," Revision 15, dated January 21, 1981.
l 1
- d. Risid Frame Restraint FSK-M-1FCB-46-1-H1 This restraint was installed in the Auxiliary Building, F1. El. -
[ 568, Spray Pump and Decay Heat Removal Pump Roon No. 27. The clearance between the 3/4" pipe and the restraint was measured
.-to be 5/32", which exceedes the Bechtel Standard Drawing FSK-M-PGS-104(Q) and Bechtel Specification 7220-M-326(Q) re-quirements. The maximum acceptable gap should be 1/8". The installation was QC inspected and accepted on May 19, 1980. ,
- e. Risid Frame Restraint 18-1HCB-2-M13 This restraint was installed in the Auxiliary Building, F1. El.
568, Reactor Building Spray Pump and Decay Heat Removal Pump Room No. 27. Clearances on one of the restraint contact locations was measured to be from 1/16" to more than 3/8". By calculation, the fabrication angle exceeded the 2' established in Bechtel Specification 7220-M-366(Q), Paragraph 5.4.1, which states that " Dimensional tolerances apply to fabrication of component pipe supports where the tolerances are not explicitly stated. The angles, formed or torch cut, should be 12*." The installation was QC inspected and accepted on May 5, 1980. Furthermore, since portions of the clearance was 3/8" or more . and exceeded the Bechtel Specification 7220-N-325(Q) tolerance, the applicability of Bechtel Specification 7220-M-366(Q) was questionable. This is an unresolved ites. (329/81-12-10; 330/81-12-11)
- f. Slidina Stanchien Assembly 2HBC-124-H7 This assembly was installed in the Auxiliary Building, Fl. El.
584, Decay Heat Removal Exchanger Roon No. _125. Fair sized , gaps covering large areas were observed between the concrete wall surface and the base plate. The condition was contrary to Bechtel Specification 7220-M-326(Q) Paragraph 5.11.1 require-ments, which state that, "The clearance between the concrete walls and the structural attachment plates should not exceed
# 1/16 inch over a maximum of 20% of the bearing area; otherwise grouting is required to ensure proper bearing." The assembly was QC inspected and accepted on September 5,1980.
- g. Risid Frame Assembly 12-2HBC-124-H5R This assembly was installed in the same area as Item f above.
Holes were drilled within the shear cone areas of the installed concrete expansion anchor bolts. The distance was measured to be 5 inches from the center of the 14" bolt to the edge of one of the holes. This condition is in violation of Bechtel Specification 7220-C-305(Q), Table B-3 which requires that the distance for the ik" dia, bolt under these conditions should not be less than 74 inches. The assembly was QC inspected and accepted on March 18, 1980. (2) FSK-1-GCB-36-2, Revision 2 Piping stress calculations including summary sheets were ( not included in the design package. (3) FSK-MO-2HCB-136-2 The preliminary stress calculation package dated November 6, 1980, contained sufficient stress summaries, references, , and design basis documentation. (4) FSK-M-OHCC-58-3, Revision 3 Piping stress calculations including summary sheets were not included in the design package. (5) FSK-M-2KBC-138-1 The preliminary stress calculation package dated November 6, 1980, contained sufficient stress summaries, references, and design basis documentation. In discussion with the Small Pipe and Hanger Group Supervisor, the inspector was told that the stress calculations will be performed after the " stress walkdown" approximately ninety days prior to the system turnover for startup testing. The inspector 4 stated that failure to document stress calculations prior to !( issuance of drawings for construction is in nonconformance with ( Bechtel EDP-4.37, Revision 2, Paragraphs 7.5 and 8.3, which state that, " Calculations shall be checked and approved, in accordance with these procedures, prior to issuing drawings for construction.... Exceptions to this requirement shall be approved by the Project Engineer," and "To ensure follow-up and finalization of incomplete work, preliminary calculations tentatively committed to final design work are filed, after review, in a separate binder entitled, " Committed Preliminary Design Calculations (CPDC)." This is an ites of noncompliance, contrary to 10 CFR 50, Appendix B, Criterion III. (329/81-12-13; 330/81-12-14) . On May 21, 1981, the licensee informed the inspector that, as of that date, 1363 isometric drawings had been issued for construction. The total number of stress calculations involved was 924. Among these, 174 were considered to meet the CPDC status and 750 lacked sufficient stress analysis documentation.
- b. Document Control During the above design control review on May 19 and 20, 1981, the following document control deficiencies were identified at the Small Bore Piping Design Center:
(1) An out-of-date copy of Bechtel Specification 7220-N-343(Q),
, Revision 3, dated January 18, 1979, for field design of 2"
I i i (3) If the cut-off portions, including hangers, are to be installed in a different system, instructions on how to transfer installation or QC records (travelers, FCR's, [ . DCN's, NCR's, etc) into the new piping systes record files.
, (4) QC inspection records will not be assembled until just before systes turnover. What measures will be taken to ensure
, effective QA audit and surveillance under these conditions? ! This is considered an unresolved ites. (329/81-12-15; 330/81-12-16) j 4. Audits of Site Small Bore Pipina Desian Activities 1 j On May 20-21, 1981, the inspector reviewed the following licensee and i Bechtel QA audits and review of small bore pipe design activities at the site: i . CP Audit Report No. M-01-24-0, performed on September 24 - j October 13, 1980. Audit areas included the small pipe and '* support design process including review and approval, document
- control, and personnel training. Nine findings were identified.
f . CP Audit Report, No. M-01-17-1, performed on April 8 - 10, 1981. ] Audit areas included staff implementation of EDP's and control of i Red lined Drawings. One finding was identified. l . Bechtel audit report of audits performed on December 11, 1979 in the areas of support design. Eleven findings were identified. j;[ o ]
. Bechtel audit report of audits performed on July 8, 1980 on stress ; calculations. No deficiencies were identified. The report stated I that, "The stress calculations were found to be in accordance with j , standard engineering practice."
- j . Bechtel QA Management Audit performed on August 25 - 29, 1980, at i the site and at the Ann Arbor office in the areas of piping and
! pipe supports. Audits in small pipe design included Red Line drawing control and pipe hanger calculations. ! . Bechtel QA Audit from May 18, 1981 to May 22, 1981. No dis-j crepancies were identified in the hanger calculations. ! Subsequent to the audit report review and discussions with the respon-t sible CP and Bechtel staff, the inspector concluded that there were inadequate audits and surveillances of the site small bore pipe and j hanger design activities. The determination was based on: 4 l a. Piping stress analysis was not audited by_CP. Where the piping ! stress analysis was audited by Bechtel QA, the MED 4.37-1, " Design l Calculations - Piping Stress Analysis Instructions", Revision 2, , dated October 16, 1979 requirements were misinterpreted. The - l Bechtel small pipe design staff and QA staff interpreted Para- ! graph 9.0 of MED 4.37-1, which states, "the period following [ reconciliation of all as-built piping drawings with the stress l { i i i
' ? . .* .~ :
NUCLLAH HLUUL At us:Y s ur..:..:: :.s v.. GEGloriIll
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??? nociiryti T C(A*j Exhibit "A"
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- CLEN E LLYN. LLit42ts 60137
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May 22, 1981 Docket No. 50-329 Luuk c No.*50-33G Censu=ers Power Cocipany ATTS: Mr. Ja=es W. Cook Vice President - Midland Projec't 1945 k'est Parnall Road Jackson, MI 49201 Centlemen: Based on Discussions between Mr. J. W. Cook and Mr. R. C. Knop on May - we understand that you will not issue fabrication and' construction drawings . for the installation of the safety related sos 11 bore pipe and piping 'suspen-sion systems until steps one through four below have been complaced and audited.
- 1. ED '4.37-0 will be revised to include requirements that the specific revision number of the specification or procedure, of which the calcu-lation was based on, is identified in the calculation package. (Note:
This action was cocipleted on 5/21/81 by issuance of Revision 16 of ED 4.37-0) .
- 2. Conduct document control review to ensure that all the applicable up-to-date specifications and procedures are in place in the work locations.
- 3. Conduct training on ED 4.37-1 (Design Calculations), the importance of following QA procedures in general, and use of specificatien M 343 for all personnel within the small bore piping design group perfom-ing stress analysis for safety-related piping.
- 4. Establish plans and schedules to review all small bore piping isomet-rics that have been issued without supporting calculations properly packaged to the revised ED 4.37-1 requirements.
- 5. Perform the reviews identified in item 4, above, to accomplish the following
- a. Bring the calculation documentation up to the level required by ED 4.37-0, Rev.16.
- b. Ensure that the calculations are technically adequate.
In conducting those reviews, the highest priority shall be given to i
- 1. .% A yP @MkW . .- .,
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i C i BASIS FOR THE REJECTION OF THE 1966 PARKFIELD EARTHQUAKE ACCELEROGRAMS FOR USE IN MIDLAND PLANT SITE SPECIFIC SPECTRA { ] c Prepared for a s CONSUMERS POWER COMPANY
' i t
July 1981 ( L_ b I = d
+q<.,p -
Wes fon Geo p ,yCORPORA'.aON sical I' lY
I W l 4 Weston Geophysical CompOaAT40se ' July 21, 1981 ; Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201 Attention: Dr. Thiru Thiruvengadam
Subject:
Basis for the Rejection of the 1966 Parkfield Earthquake Accelerograms for Use in Midland Plant Site , Specific Spectra.
- Gentlemen:
The enclosed report together with comments by Dr. Otto W. Nuttli address the above subject matter. Sincerely, WESTON GEOPHYSICAL CORPORATION Richard J. Holt l RJH eag Enclosure 1 I i I Post Office Som $50. Westboro, Massachusetts 01581. (617) 366 9191 l t
I BASIS FOR THE REJECTION OF THE 1966 PARKFIELD EARTHQUAKE ACCELEROGRAMS FOR USE IN MIDLAND PLANT SITE SPECIFIC SPECTRA i l Prepared for CONSUMERS POWER COMPANY e July 1981 g Weston GeophyCORPORATION sical I e
.-,,..,_..____,,..m._.. . _ _ _- _ - _ _ _ . _ _ . _ , , . . - _ ,
_ _ . _ .~ - _ _ _ _ _ . _ _ _ _ = _ . . _ - -_ . _ _ _ _ _ _ _ _ _ . _ _ . - _ - _ . _ .__ _ . i.. .
.. - . \
I 1 Table of Contents o i i I i l La,gg i j
1.0 INTRODUCTION
1 1 1.1 Objective 2 , i i i 1.2 Background 2 : ( 1 2.0 THE PARKFIELD EARTHQUAKE 3 t ] 2.1 General Background 3 I f 2.2 Specific Arguments Against the Inclusion i of the Parkfield Nearfield Strong Motion j Recordings in the Midland Site Specific j Response Spectra , 5 t i 2.2.1 Surface Rupture 5 i l > i 2.2.2 Supersonic Incoherent Rupture 5 1 Sources of Conservatism j 2.2.3 8 4 . i l 2.2.3.1 Large Mangitude Range 8 3.0 CONCLUDING REMARKS 9 j j ! REFERENCES 11
) APPENDIX I i
i FIGURES t . [ I ( l i l , i i f - r f i . 1 Weston Geophyscal 1 i il
... . . . . - . . . . . . . - . . . - .... ... ---.. . ~ - - .
,? .
List of Figures Figure No. Title 1 Response Spectra for the Original Ground Surface at the Midland Nuclear Plant With and Without Parkfield (54 of Critical Damping) .
~
2 Map of the Fault Trace and Aftershock Epicenters of the Parkfield Earthquake of 1966 (Modified from Papageorgiou, 1981). 3 Response Spectra for the Original Ground Surface at the Midland Nuclear Plant Compared to Parkfield Accelerogram 3034 (54 of Critical Damping). t
. ._~ -_
1.0 INTRODUCTION
This report is in response to discussions with the NRC staff concerning the use of accelerograms from the Parkfield California Earthquake of 1966. The discussions were concerned with the appropriateness (or non-appropriateness) of these accelerograms for use in the site specific spectra developed for the Midland site. Included with this report is a letter by Dr. Otto Nuttli which addresses the same issue (see Appendix 1). Site specific spectra are developed by selecting accele-rograms from the world-wide data set that most closely match the safe-shutdown earthquake magnitude, distance (from the site), and local geological conditions. In the case of the Parkfield earthquake, the distances to the accelerometer stations from the epicenter appear to fall in the correct range. The local geologic conditions, as determined by shear-wave velocities, are acceptable. Some published magnitudes for this earthquake appear to be appropriate for use at Midland; however, as it will be discussed, individual station magnitudes determined at close distances would make it inappropriate. In addition, the earthquake effects, charac-terized in previous reports and in this one as "near field," make it inappropriate for use at Midland. As previously mentioned in the addendum to Part 1, Response Spectra - Original Ground Surface, in applying
" Appendix A", one or more of three geologic conditions govern f
Weston Geophyucol r__ x _-______________~_-_=________ __ - - -~ ~ - - - ~ -- -
1, - i .
! the selection of a Safe Shutdown Earthquakes a " capable fault",
a " tectonic structure", and/or a " tectonic province". The safe shutdown earthquake for Midland was based on a tectonic province only. This means that a maximum observed earthquake (or higher) in the province, is assumed to occur at the site where no capable fault or tectonic structure has been identified. It is believed that an accelerogram showing ; characteristics related to capable faulting or the effects of , proximity to a tectonic structure should not be used in the development of site specific response spectra within the context of a " tectonic province" approach, unless unusual . j tectonic circumstances exist at/near the site. 1.1 obiective The objective of this report is to demonstrate that accelerograms resulting from the Parkfield California earthquake of June 28, 1966 should not be included in a data set used to model a 5.3 mb earthquake for the ! Midland Nuclear Power Plant. i ! 1.2 Background In the original Midland submissions Parts I and II the Parkfield earthquake accelerograms that met the magnitude-distance-geologic foundation conditions criteria . were not included because of the anomalous nature of the
, event.
t Wtlion Geophylacol _____3-
. . - ~ ,
e, e .* In an addendum to Part I, at the request of NRC, sensitivity tests were performed including these records. The results showed that the anomalous Parkfield records , are much higher than the 44 components originally considered appropriate to model a magnitude 5.3 m b earthquake, and that their inclusion raises the 84th percentile by a significant amount (see Figure 1). In the next section (2.0), reasons for not including the near-field Parkfield records to model the Midland site earthquake potential are documented in detail. 2.0 THE PARKFIELD EARTHOUAKE 2.1 General Background A voluminous amount of literature has been published concerning the Parkfield event because of the extensive nearfield instrumental coverage and its unusual features, such as its dislocation and surface rupture. Some of the major references are: McEvilly et al. (1967) , Eaton (1967), Filson and McEvilly (1967), Aki (1968), Haskell (1969), Tsai and Aki (1969) , Scholtz et al. (1969), Eaton et al. (1970) , Aki (1972) , Tsai and Patton (1973) , Murray (1973), Trifunac and Udwadia (1974), Anderson (1974), Lindh and Boore (1974), Kawasaki (1975), Levy and Mal (1976), Archuleta and Day (1977), Wiggins et al, (1977), Hartzell et al. (1978), Kanamof i and Jennings (1978), Aki (1979), Wu (1968), Housner and Trifunac (1967), apd Papageorgiou (1981) . Wepon Geoonyucol
* ' "*~~'~ ' ~ ' ~ ~ ' ' " ~ ~ ~
-_M1--------------__-----.--.__:~~^-------T--___ _ _ - -l 1~ _
1 .. .
.. i
. I I
4 ! I
'i Some of the important parameters determined by '
j various authors and agencies exhibit a range of values: 1 Magnitude: 5.3 (USGS - Filson and McEvilly, 1967), , 5.8 ab (WU,1968) , 5.5 ML (BRK) , 56ML (PAS), j 6.4 Ms (WU,1968) . i i j Rupture velocity: 2.2 to 2.5 km/sec most often quoted ( Aki,1979) , some estimates above 3.0 km/sec 1 (Anderson, 1974). ; Rupture Length / Segmentation: Main fracture zone ' about 37 km (Brown and Wedder, 1967). Segmentation , lengths vary with interpretation. h l Fault Dislocation: 60 cm (Aki, 1968): surface i rupture actually smaller. ! J
! Depth: Quotes vary from about 3 km (Aki, 1968) to 15 i j km (based on af tershocks) depending upon the branch l of the rupture considered. '
T This earthquake occurred in one of the most ! i k j seismically active zones of the San Andreas Fault i j (McEvilly et al. ,1967) . It caused surface rupture and very high recorded accelerations in the nearfield. Many j studies have shown that the dislocation was characterized by incoherent starting, stopping, and jumping over L 4 barriers. I i 2.2 Specific Arguments Against the Inclusion of the Parkfield Nearfield Strong Motion Recordings in the
; Midland Gite Specific Response Spectra 2.2.1 Surface Rupture i h j l Extensive surface! cracking was observed after the Parkfield earthquake. Aki (1968) stated'that the j i
4 surface was decoupled by a thin layer of about 100 m l . Weston Geophywcol i i
[ below which a calculated 60 cm dislocation probably took place. This is larger than the one observed at the surface. In the central United States, surface i rupture is not expected to result from magnitudes less than 6.5 ab (Nuttli, personal communication 1981); undoubtedly, the hypothetical 5.3 magnitude assumed to occur "at the site" would not be . accompanied by surface rupture. Thus, the Parkfield 4 earthquake is not characteristic of a central United States scenario. 2.2.2 Supersonic Incoherent Rupture Murray (1973) and many other authors (Aki,1968; Haskell,1969; and Papageorgiou,1981) have addressed the affect of incoherent rupture or dislocation ou seismic radiation. When a fault rupture occurs, the
, crack propagation can be smooth, or it stops or ; . starts, or actually jumps over a barrier. Barriers i
may be classified into two types: (Aki, 1979) geometrical barriers such as a bend in a fault; , 4 barriers made by inhomogeneities such as evidenced by i velocity anomalies. Quoting Aki (1979), these I j barriers (both types) can act "...not only as a stopper of rupture but also as an initiator of rupture, as well as stress concentrator". j I WtWon GeophyWCol
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_ ______.___ _ _~ + . . ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _
- -~.,... - - - . - -.-
i.. ., .- i l> . 4 For Parkfield, such a jump (discontinuity) did l occur. The fault trace jumped from the east side of i i the Cholame Valley to the west side, very close to i i accelerograph Station 2 (See Figure 2) . The vertical ! 1 ! I component of the seismogram at Station 2 shows strong I ! high frequency waves which have been generated by the jump (Papageorgiou,1981) . Anderson also suggested that the rupture propagated through a region with ! j larger than average irregularities and possibly even stopped, and restarted on a separate plane, resulting i j in higher amplitudes of the high frequency F-waves. L Anderson (1974) showed that, although it could i j possibly be the arrival from the S-wave out of the i epicentral region, the high frequency phase arrival ; i ! j at Station 5 also corresponds to the time P-waves are [ l l arriving from the rupture front in the area of the l - jump. ( A consistent model of crack propagation to ! explain the strong motion radiation from Parkfield j , was developed by Murray (1973) . He found that the l rupture velocity for this event was supersonic, that I 1
; is, the crack propagation velocity was greater than i
the shear velocity of the medium. From his complex j history of the rupture he concludes that the l acceleration maxima were caused by the formation of -
- Mach waves generated by. stopping and starting. He-WrWon GeophyWCol ,
I '..- ... ...-.. .. . .. . . . . . . . . . . . - - , . . t
__ _ -_ _=____________ _ _ _______ __ _ - _- _ _ _ _______ futhermore notes that "...We may then expect that sufficiently close to the fault, the attenuation will approximate that due to an infinite source. At a i distance x from the fault that is sufficiently large compared to the vertical extent of the source, the amplitude should decay approximately as x"1 In fact, the transverse acceleration " jump" amplitudes (when the accelerations attain maximum values) at Stations 5 and Temblor (5.20 and 6.45 km from the fault) are the same as at Station 2. At Station 8 (9.23 km), the amplitude has fallen by 33 percent, and at Station 12 (14.7 km) by 87 percent." From the above discussion it is clear that the high values of acceleration recorded at the close stations were caused by an extremely unusual element in the dislocation history and that Station 5 and Temblor recorded nearly unattenuated Mach waves. Consequently, the usage of these abnormal accelerograms would introduce additional conservatism for the modeling of a 5.3 m b event at Midland. 2.2.3 Sources of conservatism 2.2.3.1 Large Magnitude Range Imposing a large magnitude range, 5.3 mb t0.5,sor 5.4 M gio.5, already skews the average and 84th percentJ1e of a site specific response spectrum towards the Wepon Geoony> col
.. - .. .. . . _ . . .. ~ . . . - , . - - . .
level of the higher magnitude. This influence was discussed in the Midland Addendum to Part I. If we add anomalously high recordings to a set of appropriate data, the resulting response spectra will not correspond to the targeted magnitude, 5.3 m b. As it can be seen in Figure 3, some of the Parkfield records are much higher than the 84th percentile spectrum. In fact, these records are higher than any . of the data set collected to represent the 5.3 m b spectrum. Further substantiation is found in a paper by Kanamori and Jennings (1978), where the acceleration time histories were run through a Wood Anderson simulator and M g's (local magnituden) were directly calculated from the strong motion records. This was done for two distances: 1) the closest approach of the fault, and 2) the distance to the zones of aftershocks. Their results for Parkfield are reproduced below: , I Wepon Geophywol
_ . - . y .. e
* . s Station Ref M3 Cholame 62 5033 6.75, 6.35 Cholame 65 5034 6.35, 5.95 6.25, 5.90 ,
Cholame 98 5035 5.75, 5.35 6.1, 5.7 Cholame 012 5036 5.75, 5.35 5.9, 5.55 Temblor 5037 6.4, 5.7 6.7, 6.0 It should be emphasized that these values are meant to be M 3 's. It is clear from the above that for an a =5.3, b M g=5.3-5.4 data set, only 5035 and 5036 should be given marginal consideration for inclusion. The inclusion of Cholame 5, with an estimated Mg=6.35 or 5.95 would not characterize the m b=5.3 (M 3: 4.9 to 5.5) at Midland. 3.0 CONCLUDING REMARKS '
. The Parkfield earthquake caused surface rupture and experienced supersonic dislocation velocities complicated by the presence of barriers generating Mach waves. These phenomena combined to generate extremely high accelera-tions in the nearfield. These high accelerations have been shown to be typical of magnitudes greater than 6.0.
Indeed, this earthquake has been modeled as a series of multiple shocks (Wu, 1968). i
. WrWon Geophyscal
i It is not correct to bias the 84th percentile of the 5.3 magnitude data set, carefully developed for Midland, by adding these anomalous records. If they are included and the 84th percentile spectrum is used, the results will substantially deviate from the defined potential at the site. Since this defined potential was determined using i the tectonic province approach, which of itself introduces substantial conservatism, the level of the resulting ! spectrum would be unreasonable for the tectonic i environment of the Midland plant. i I ) i i 1 i I t
- WeWon Geophywcol
.. .. . . - . , . . . . ..v. . - - -
9 RETERENCES Aki, K., 1968, Seismic Displacement Near A Fault, Journal of Geophysical Research, V. 73, p. 5359-5376. i Aki, K., 1979, Characterization Of Barriers On An Earthquake Fault, Journal of Geophysical Research, V. 84, p. 6140-6148 Anderson, J.G., 1974, A Dislocation Model For The Parkfield Earthquake, Bulletin of the Seismological Society of America, v. 64, p. 671-686. Archuleta, R., and Day, S.M., 1977, Near-Field Particle Motion Resulting From A Propagating Stress-Relaxation Over A Fault Embedded Within A Layered Medium (Abstract), American Geophysical Union Transaction, V. 58, p. 445. Bouchon, M., 1979a, Predictability Of Ground Displacement And
, Velocity At Proximity Of An Earthquake Fault: An Example: The Parkfield Earthquake of 1966, Journal of Geophysical Research, V. 84, p. 6149-6156.
Das, S., and Aki, K., 1977, Fault Planes With Barriers: A Versatile Earthquake Model, Journal of Geophysical Research, V. 82, p. 5648-5670. Eaton, J.P., O'Neill, M.E., Murdock, J.N., 1970, Aftershocks Of The 1966 Parkfield-Cholame, California Earthquake: A Detailed Study, Bulletin of the seismological Society of America, V. 60, p. 1151-1197. Filson, J. , McEnvilly, T.V. ,1967, Love Wave Spectra And The Mechanism Of The 1966 Parkfield Sequence, Bulletin of the Seismological Society of America, V. 57, p. 1245-1257. Haskell, N.A., 1969, Elastic Displacements In The Near-Field Of A Propagating Fault, Bulletin of the Seismological Society of America, V. 59, p. 865-908. Housner, G.W., and Trifunac, M.D., 1967, Analysis Of Accelero-grams-Parkfield Earthquake, Bulletin of the Seismological Society of America, V. 57, No. 6, p. 1193-1220. Kanamori, H., and Jennings, P.C., 1978, Determination On Local Magnitude, ML From Strong-Motion Accelerograms, Bulletin of the seismological Society of America, V. 68, No. 2,
- p. 471-485.
Kawasaki, I., 1975, On The Dynamical Process Of The Pa'rkfield Earthquake of June 28, 1966, Journal of Physics of the - Earth, V. 23, p. 127-144. . WrWon Geophyscal
. . . . . . - - . . . . . . . . .. ._. _ .. -.- - . ...- - . .~.- .
a - z .. .- . - .-. - - i , i Levy, N.A., Mal, A.K., 1976, Calculation Of Ground Motion In A Three-Dimensional Model Of The 1966 Parkfield Earthquake, Bulletin of the Seismological Society of i
' America, V. 66, p. 405-423.
Lindh, A., Boore, D.M., 1974, The Relation Of The Parkfield Foreshocks To The Initiation And Extent Of Rupture ( Abstract) , Earthquake Notes, V. -45, p. 54. 1 McEnvilly, T.V., Bakun, W.H., and Casaday, K.B., 1967, The i Parkfield, California Earthquakes Of 1966, Bulletin of the
; seismological Society of America, V. 57, p. 1221-1244.
Papageorg' iou, A.S., 1981, On An Earthquake Source Model Of Inhomogeneous Faulting And Its Applications To Earthquake Engineering, MIT Research Report R81-10, No. 696, for National Science Foundation Grant No. PFR-7827068. Trifunac, M.D., and Udwadia, F.E., 1974, Parkfield, California, j Earthquake Of June 27, 1966: A Three-Dimensional Moving 4 Dislocation, Bulletin of the Seismological Society of j America, V. 64, p. 511-533. 4 Weston Geophysical Research, 1981, Part II, Response Spectra Applicable For The Top Of Fill Material At The Plant Site, In Site Specific Response Spectra Midland Plant - Units 1 and 23 .?repared for Consumers Power Company. Weston Geophysical Research, 1981, Addendum to Part I, Response Spectra - Original Ground Surface, In Site Specific Response Spectra Midland. Plant - Units 1 and 2: Prepared j for Consumers Power Company. Wiggins, R.A., Sweet, J., and Frazier, G.A., 1977, The
- Parkfield Earthquake of 1966 - A Study Of Dislocation I Parameters Based On A Complete Modeling Of Elastic Wave <
Propagation (Abstract), American Geophysical Union Trancaction, V. 58, p. 1193. 4 ) e a t ? e, ..
I 1 APPENDIX I i J t > J
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ff*/*/f' 6 58-312 4 July 13, 1981 g Mr. Richard J. Holt, President Weston Geophysical Corporation P.O. Box 550 Westboro, MA 01581
Dear Mr. Holt:
I am replying to your request to comment upon the appropriateness of using response spectra from the 1966 Parkfield earthquake in establishing a set of Site Specific Response Spectra for the Midland (Michigan) plant. The mainshock of the Parkfield earthquake sequence was anomalous in a number of ways, which you pointed out in your response to NRC, dated July 1981. Principal among them are the surface rupture and the
- large amylitude of high-frequency waves associated with incoherent supersonic rupture across seismic barriers. These are reflected in the high Hg values obtained by Kanamori and Jennings (BSSA, pp. 471-485, 1978) when they used the acceleration time histories to produce simulated Wood-Anderson seismograms, and obtained M3 values greater than 6 from the Cholane no. 2, Cholane no. 5'and Temblor strong-motion time histories.
From more distant Wood-Anderson seismograph records _ the ML value assigned to the aarthquake was 5.3. Thus the Kanamori and Jennings values show both that the large acceleration observed at the above-mentioned three near-field stations results from localized features, and that it attenuated much more rapidly than the motion produced by the principal part of the rupture process. The only characteristic of central United States earthquakes that is known to produce large amplitude, rapidly attenuating high frequency
- waves is very shallow focal depth. Table 23 of NUREG/CR-1577 lists all the known central United States earthquakes of this type. of the 59 events listed, the maximum mb value was 4.3. (The 1966 and 1967 Attica, N.Y. earthquakes had mb values of 4.6 and 4.4 and focal depths of 2 and 3 km, respectively.) These ab values are a measure of the far-field
! ground motion, as the Mt = 5.3 value was for the 1966 Parkfield earth- l
- quake. I believe we have a sufficient sample of very shallow central United States earthquakes to conclude conservatively that their ab value will not exceed 4. 8. Bob Herrmann and I (manuscript in preparation) found that mb(EUS) =Mn(WUS), using waves of approximately 1-Hz frequency, i.e. the amplitudes of 1-Hz waves excited by earthquakes in the two regions are the same when the magnitudes are numerically equal. There-fore, I _would not expect a
- very shallow central United Stat'es earthquake wesion seophysicai
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Mr. Richard J. Holt ' July 13, 1981 pg. 2 to produce near-field ground motion greater than that of an M 4.8 California earthquake, which would be log-1 0.5, or 0.32 timeh =that of the 1966 Parkfield earthquake. t It.is generally accepted that central and eastern United States earthquakes have not produced surface rupture, with the exception of the great earthquakes of the 1811-1812 New Madrid series. For the various reasons given above, it is my personal opinion that the -Parkfield earthquake accelerograms, at near-field stations should not be included in the set of spectia used to obtain a site-4 specific response spectrum for Midland, Michigan. Sincerely yours, 0% LLl.]R Otto W. Nuttli 1 . a wesion seconyi,ca
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10-8 10-' Ioo go' PERIOD (SEC) . RESPONSE SPECTRA FOR THE ORIGINAL GROUND SURFACE AT THE MIDLAND NUCLEAR PLANT WITH AND WITHOUT PARKFIELD (5% OF CRITICAL DAMPING) FIGURE 1 m- - mme oe o +e eem e.oome e - _ m-
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.Parkfield earthquake of 1966, reproduced from Eaton et. al.
(1970). Both the fault trace and the fault plane at depth, identified by the aftershock zone, jumps from one side of the Cholame Valley to the other. Two lines were drawn by Aki (1979a), fitting the two zones of aftershock epicenters. (Modified from Papageorgiou,1981). FIGURE 2 i e r
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10-' goo io' PERIOD (SEC) RESPONSE SPECTRA FOR THE ORIGINAL GROUND SURFACE , AT THE MIDLAND NUCLEAR PLANT COMPARED TO
- 1 PARKFIELD ACCELEROGRAM B034 (5% OF CRITICAL DAW ING) ~
FIGURE 3 weston Geophyscal 9
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UNITED STATES OF AMERICA !? D I NUCLEAR REGULATORY COMMISSION h , , e.. I 1981 * ,_~- Mq c : ., . ., O TESTIMONY OF RICHARD J. HOLT '\
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s MIDLAND SITE SPECIFIC RESPONSE SPECTRA My name is Richard J. Holt. I am President of Weston Geophysical Corporation, a geological, geophysical and seismological consulting firm, located in Westboro, Massachusetts. As the attached resume shows, I have an M.S. in geophysics from Boston College and over 25 years professional experience as a geophysicist and seismologist. At the request of Consumers Power Company (Applicant), Weston Geophysical has r constructed the Midland site specific response spectra described in this testimony. This work was either done by me or under my direct supervision. By reason of this personal involvement as well as the education and professional experience outlined in my resume, I believe I am qualified to testify with respect to the matters discussed below. I. INTRODUCTION
~
The purpose of this testimony is to respond to the Licensing Boards inquiry in its PREHEARING CONFERENCE ORDER dated May 5, 1981 concerning the establishment of a Safe
. Shutdown Earthquake * ("SSE") for the Midland site,
- As explained in Applicant's brief, use of this term does not imply substitution of a new design. basis for structures at the Midland plant.
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corresponding vibratory ground motion, and associated response spec tr a. As stated in more detail in the testimony below, Site Specific Response Spectra ("SSRS") representing the vibratory ground motion produced by t.he SSE have been developed for the 1 original ground surface (Holt Exhibit 1) and for the top of the fill material at the Midland site (Holt Exhibit 2) in accordance with the parameters dictated to Applicant by the NRC
, Staf f in the October 14, 1980 letter from Robert L. Tedesco (NRC) to J. W. Cook of Consumers Power Company (Holt Exhibit 3). My testimony also makes it clear that these parameters substantially overstate the appropriate Seismic hazard at the Midland site, and therefore the SSRS presented in Holt
,, Exhibits 1 and 2 are unnecessarily conservative.
II. EXPLANATION OF HOW MIDLAND SITE SPECIFIC SPECTRA WERE DERIVED 10 CFR Part 100, Appendix A, describes procedures for determining, for NRC Licensing purposes,' the quantitative vibratory ground motion at a site due to earthquakes. These procedures involve two basic decisions: first, the selection of the size and location of a Safe Shutdown Earthquake which represents a conservative estimate of the source of vibratory ground motion at the site; second, the construction of a response spectrum representing the resulting ground motion at the site. Appendix A is fairly explicit on the procedures to _ be used in selecting the size and location of the SSE. However the regulation does not specify how the resulting ground motion at the site should be determined. I
.* , n .
s 3-0- In 10 CFR Part 10 0, Appendix A, there are three methods by which one selects a safe shutdown earthquake; the safe shutdown earthquake is defined on the basis of a capable fault, a tectonic structure, and/or a tectonic province. 10 CFR B Part 100, Appendix A, Section V(a) . In the case of Midland the F tectonic province approach was used because there are no discernable tectonic structures within 200 miles of the site,
,, no capable faults have been identified near the Midland site, and generally capable faulting is not a problem in the eastern United States. Appendix A requires that the Safe Shutdown Earthquake intensity is, at a minimum, equal to the maximum historic earthquake intensity experienced within the tectonic province in which the site is located. 10 CFR Part 100, Appendix A, Section V(a) (iv). Thus the definition of the tectonic province in which the site is located plays a crucial role in determining the size of the Safe Shutdown Earthquake which t'1e plant must be able to withstand. Applicant and the NRC Staff disagree over the appropriate tectonic province to use for Midland, as explained in Part III of this testimony.
The Staff's choice of the Central Stable Region as the tectonic province for Midland leads to identification of an SSE similar , to the type earthquake which occurred in Anna, Ohio in 1937. The NRC staff characterizes this type of earthquake as a Magnitude 5.3 or a Modified Mercalli Intensity VII-VIII.1/ The SSRS shown in Holt Exhibits 1 and 2 have been developed based on this SSE, even though Applicant believes it is-too large.
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The vibratory ground motion produced by Safe Shutdown Earthquake for a nuclear power plant must be defined by response spectra. 10 CFR Part 100, Appendix A, Section V(a) (1) (iv). These response spectra relate the response of the I foundations of the nuclear power plant s tructures to the vibratory ground motion considering such foundations to be single degree of freedom damped oscillators and neglecting soil-structure interaction effects. Appendix A states that "In view of the limited data available on vibratory ground motions of strong earthquakes, it usually will be appropriate that the response spectra be smoothed design spectra developed from a series of response spectra related to the vibratory motions of more than one earthquake." 10 CFR Part 10 0, Appendix A, Section VI(a)- (1) . Appendix A does not dictate a specific method for developing such smoothed response spectra, except tha t the size of the earthquake and the specific site 1/ See Holt Exhibit 3. An earthquake's size is defined by intensity or magnitude. Both terms are used in Appendix A and in Holt Exhibit 3. Intensity is a measure of the earthquake's size, based on a standard (Modified Mercalli) scale of felt effects as reported by persons witnessing the event. See Holt Exhibit 4. However, intensity is greatly qualitative, and depends on local soils conditions, type of buildings, as well as observations of people. Magnitude is a better measure of earthquake size since it is instrumentally defined or related. However, for most historical earthquakes, intensity is generally referenced because instrumental records are unavailable. The relationship between intensity and magnitude, particularly for the central and eastern United States, is fairly well developed (Nuttli, Bollinger, Griffiths, 1979). s e
,e., s.
f foundation and geologic conditions must be considered. 10 CFR Part 100, Appendix A, Sections IV (a) (1) , (3 ) , (4 ) , (5) and V (a) (1) (iv) . I Common practice during the last part of 1970's for licensing reviews of nuclear power plants has been to represent SSE vibratory ground motion by scaling a standardized response spectral shape (Regulatory Guide 1.60) to an assumed zero _ period (that is, frequency > 33 Hz) acceleration level corresponding to the Safe Shutdown Earthquake. The assumed zero period acceleration level for the SSE has been selected by using an earthquake intensity-peak _ ground acceleration relationship such as that proposed by Trifunac and Brady, 1975. Thus one of the alternatives offered by the NRC Staff in the October 14, 1980- Tedesco letter (Holt Exhibit 3) is to represent the vibratory ground motion resulting from an Anna, Ohio type earthquake by using a Regulatory Guide 1.60 spectral shape anchored to 'an assumed zero period acceleration of O.19 g.1/
,, The standardized response spectra in Regulatory Guide 1.60 were developed using a wide range of earthquake magnitudes at -
distances ranging from a few to over one hundred kilometers. In addition, the. strong motion recordings (accelerograms) used 2/ It should be pointed out that the zero period or peak
~ ground acceleration, apart - from its use to scale or
" anchor" the Regulatory Guide 1.60 spectral shape, is not itself a very important value to engineers in determining the safety of nuclear power plant structures. .
~ _. _ 5 4 to develop the spectra were recorded on widely varying local site geologic conditions. The resulting spectral shape of Regulatory Guide 1.60 is site indeoendent: when it is scaled to an assumed zero period' ground acceleration value (0.19 g) representative of an Intensity VII-VIII or Magnitude 5.3 earthquake, it generally defines a level of ground motion in excess of that which the ' site would experience due to the occurrence of such an earthquake.
. At the time' that Regulatory Guide 1.60 was constructed the number of strong motion records in existence was relatively small. There were not enough strong motion records to construct different response spectra for sites with different foundation conditions. In recent years a large number of strong motion recordings have become available from earthquakes of various magnitudes at different distances and for a variety of foundation conditions. At the present time, response spectra corresponding to specific site foundation conditions can be constructed for most sites.
Basically the site specific method involves constructing response spectra from records of ground motion recorded by accelerometers located at sites similar to Midland from earthquakes similar in magnitude to the proposed Safe Shutdown Earthquake. Thus the important criteria for selecting these earthquake records are the range of magnitudes for the earthquakes, the distance from the epicenter to the recording
. .. . . _ . . . . . . .- ~ ----... - - - - - - - -- - - -
as Se station, and the recording station geology and foundation conditions. The October 14, 1980 Tedesco letter (Holt Exhibit 3) dictates the use of a magnitude range of 5.3+.5, epicentral distances of less than 25 kilometers, and recording ins truments on soil. The magnitude of 5.3 is meant to correspond to an Anna, Ohio type earthquake. Thus the NRC Staf f has specified use of the same Safe Shutdown Earthquake for both alternatives stated in the October 14, 1981 Tedesco letter (Holt Exhibi t 3) . The distance from the recording station to the earthquake is selected to approximate the distance from the site of the nuclear plant to the earthquake. In the case of the tectonic province the highest intensity of the earthquake is assumed at , the site [10 C.F.R., Part 100, Appendix A, Section V(a) (1) (ii)] with its corresponding ground motion. Since the epicentral' intensity occurs over an area, a range of distances for the magnitude occurrence is selected. Because capable faulting is generally not a problem in the eastern United States and has specifically been eliminated by thorough investigation of the Midland site, the selection of the strong motion records should be such that the "near field" effects of capable faulting are eliminated or minimized. This is accomplished through the selection of appropriate distances and, if possible, the most representative earthquakes. I I
. . _ _ . . . . . , . . . . . . - _ . . - - - - -- - - - - - ' ~ ~ ~ -
I s s The last criterion to be satisfied is the match of geologic materials underlying the site to those underlying the , strong motion recording station. In the last few years measurements of seismic compressional and shear wave velocities have been made at many of the recording sites so that a direct quantitative comparison can be made to the Midland site. 4 Holt Exhibit 1 shows the Site Specific Response Spectrum ^ which has been developed for the original ground surface at the Midland site using the magnitude, distance and site foundation materials criteria discussed above. A detailed explanation of l how this spectrum was developed is found in the report entitled
" Site Specific Response Spectra Midland Plant Units 1 and 2 -
Part I Response Spectra - Safe Shutdown Earthquake Original Ground Surface" which is attached .to my testimony as Holt Exhibit 5. The NRC Staf f had expressed concern that the SSRS shown in Figure 11 of Holt Exhibit 5 falls off sharply at in 1 the long period region. Since there are no means of addressing this concern within the bounds of the available data set2 /, and since the Midland FSAR design spectrum exceeds the SSRS at those long periods, the 84th percentile response spectrum was arbitrarily raised to the FSAR design spectrum level in this 1/ Such long period ground motion would come from extremely
. large earthquakes at great distances. There is a paucity 1
of strong motion records for such earthquakes. i
.. . . . . . _. . . . ..~ . . . . . . . - . .
9_ long period region. The final SSRS for the Midland site at the original ground surface shown in Holt Exhibit 1 therefore, differs from the spectrum shown in Figure 11 of Holt Exhibit 5. I Subsequent to the submission of Holt Exhibit 5, the United States Nuclear Regulatory Commission staff requested further information which is the subject of a report entitled " Site Specific Response Spectra Midland Plant Units 1 and 2 Addendum to Part I Response Spectra Original Ground Surface." (Holt Exhibit 6) . A further question from the NRC Staff led to the submission of a report entitled, " Basis For the Rejection of the 1966 Parkfield Earthquake For Use in Midland Plant Site Specific Spectra" (Holt Exhibit 7) . 10 CFR Part 100 Appendix A, Section (V) (a) (1) (iv) states that response spect'ra shall developed for the SSE at "each of the various foundation locations of the nuclear power plant". Most of the Seismic Category I structures at Midland are or will be following proposed modifications founded on the original ground (glacial till). The Diesel Generator Building
' however is founded on plant fill. It is possible to take this into account directly in the site specific response spectra methodology by matching foundation conditions under the Diesel Generator Building to similar foundatin conditons at accelerometer stations recording strong motion from Anna, Ohio type earthquakes (M = 5.3 +0. 5) . This has been done, as documented in a report entitled, " Site Specific Response J
._ _ . _ .. .. -~
Spectra Midland Plant Units 1 and 2 - Part II - Response Spectra Applicable for the Top of the Fill Material at the Plant Site", which is attached to my testimony as Holt Exhibit 8. For convenience, the SSRS for the top of the fill I material at Midland is also presented in Holt Exhibit 2. It would also be possible to derive a response spectrum at the top of the fill by multiplying the SSRS for the original ground surface (Holt Exhibit 1) by appropriate frequency-dependent amplification factors which would account for the 30 feet of compacted fill underneath the Diesel Generator Building. In fact such an approach has been explored by E. VanMarcke, E. Kausel and E. Samaris of the Massachusetts Institute of Technology and the results of their work are reported in Appendix B of Holt Exhibit 8. The response spectrum derived using these amplification factors predicts less ground motion at the top of the fill than does the SSRS shown. in Holt Exhibit 2, which was derived directly using strong motion records from recording stations with foundations similar to those under the Diesel Generator Building. Applicant will use the higher spectrum shown in Holt Exhibit 2. For the reasons stated, the SSRS in Holt Exhibits 1 and 2 are a more realistic representation than the Regulatory Guide 1.60 spectral shape anchored at 0.19 g for the ground motion which would result-from the occurrence of an Anna, Ohio type
. earthquake at the Midland site.
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r III. THE SITE SPECIFIC RESPONSE SPECTRA DEVELOPED FOR MIDLAND , USING THE STAFF' S CRITERI A ARE VERY CONSERVATIVE The Site Specific Response Spectra presented in Holt Exhibits 1 and 2 were properly constructed in accordance with the parameters dictated by the Staff to Applicant in the : October 14, 1981 Tedesco letter. However, I do not agree that these parameters are appropriate. To the contrary, in my i opinion the Site Specific Response Spectra constructed using ' the NRC Staff 's parameters are too high and overstate the ; seismic hazard at the Midland site. ! A. The NRC Staf f 's Tectonic Province Is Too Large ' 10 CFR Part 100 defines the term tectonic province as follows: A " tectonic province" is a region of the North American continent characterized by a relative ! consistency _ of _ the geologic structural features contained therein. From a geological standpoint, it is clear that the larger earthquakes of interest in siting nuclear power plants occur in r the crystalline basement rock. ( Therefore, it is the geology of ; the crystalline basement rock which is important for purposes ' of Appendix A. The Applicant proposed the use of the Michigan Basin as a ' tectonic province in the Midland FSAR. The Michigan Basin, shown in Holt Exhibit 9, is-a saucer-like tectonic unit nearly ! 200 miles in diameter whose crystalline basement rock is 'about
' . . . _ _ _ _ _ . t. . _ _ _ . . _ _ . . _ . . 6,000-8,000 feet deeper than the basement rock arches which surround it. Such relief between the basement rock in the Michigan Basin and the basement rock in the surrounding arches indicates that the two areas have different geological structural features. Moreover, as Holt Exhibit 9 also shows, the arches on the southern end of the Michigan Basin have been areas where earthquakes of approximate magnitude 5 and intensities VII or slightly larger have occurred. Within the basin the largest magnitude has been 4.5 and the maximum earthquake intensity has been VI. The Midland site, which is centered near the middle of the Michigan Basin, has experienced a maximum intensity of IV to V in historical times based on a conservative attenuation estimate. The largest earthquake within 200 miles of the site is VI. The nearest earthquake to the site which exceeded VI is the Anna, Ohio Earthquake of 1937 (205 miles away) with an estimated magnitude of 5.0 and an intensity of VII, or VII to VIII. The Anna, Ohio earthquake was spatially located in the Cincinnati and Finlay arch system. Thus the relatively lower seismicity of the Michigan Basin confirms that it should be treated as a separate tectonic province. In the October 14, 1980 Tedesco letter the NPC staff requested that the Applicant use the Central Stable Region as a tectonic province and stated that the Safe Shutdown Earthquake
v_. ._ . .. .. .-. .- - - - - . -- -- - - - -- - -- should be defined as the occurrence of an " Anna type earthquake" with magnitude 5.3 earthquake at the site.1/ The Central Stable Region is a very large region of the ; United States defined not on the characteristics of the crystalline basement rock which underlies it, but on the sedimentary rock strata which overlie the basement. These rocks are called sedimentary because they were formed from sediments deposited in a great inland sea about 200 to 600 million years ago. While the extent of the Central Stable Region is not clear, it would include most of the Central United States from Ohic to the Rocky Mountain front and from the Canadian shield to the Mississippi embayment (roughly the 38 th parallel). Since large earthquakes will originate in the crystalline basement rock, defining the tectonic province based on the presence of a veneer of sedimentary rock is unreasonable. The NRC staff has stated to the Commission that: The concept of tectonic province was developed to provide an appropriate design basis for earthquakes whose cause is indeterminate. Th e staff interprets this concept as employed in Appendix A to imply regions of uniform earthquake hazard. d/ Holt Exhibit 3. A recent authoritative study performed for the NRC indicates that the magnitude of the 1937 Anna, Ohio earthquake was 5.0, not 5.3. Nuttli and Brill,1980
" Earthquake Source Zones in the Central United States Determined From Historical Seismicity" (preprint). Thus the reference in the October 14, 1980 Tedesco letter to an " Anna type earthquake" with magnitude 5.3 is ambiguous.
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- Commission Information Report, SECY-79-300, " IDENTIFICATION OF i [
ISSUES PERTAINING TO SEISMIC AND GEOLOGIC SITING REGULATION, [ POLICY, AND PRACTICE FOR NUCLEAR POWER PLANTS" (April 29, i ' I F 1979), Enclosure A at p. 7. While I do not believe that tectonic provinces should be defined solely on the basis of i .. j historical seismicity or a probabilistic analysis of such seismicity, seismicity and analysis of seismicity can be used l to test the validity of a defined tectonic province. It is f
; clear that the Central Stable Region does not represent a ;
i ! region of uniform seismicity. For example, referring to Holt Exhibit 9, the Michigan Basin is an area of comparatively low j seismicity. Most of the larger historical earthquakes (MMI VI) ! shown on Holt Exhibit 9 appear id clusters geographically, ! which makes one suspect the presence of tectonic structures at i j - those locations. Formal probabilistic analysis of the . !' historical data set for the earthquakes represented in Holt { Exhibit 9 confirms that the Midland site is in an area of ; i d relatively low seismic hazard as compared to other sites surrounding the Michigan Basin in the Central Stable Region. This formal probabilistic analysis is summarized in Holt t : Exhibit 10. { Although I strongly believe that there are several , a definable tectonic provinces and structures within the Central Stable Region, we have, in accordance with the October 14, 1980 l Tedesco letter (Holt Exhibit 3) used an earthquake of magnitude 5.3 at the Midland site for the purpose of i j constructing Site Specific Response Spectra.-
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B. - Applicant 's Selection of Earthquakes for Determination of Site Specific Spectra is Appropriate In selecting strong motion records to develop site specific spectra, three basic criteria are used, a defined magnitude range, a range of epicentral distances and a shear wave velocity range, to simulate a Safe Shutdown Earthquake occurrence at the site. For the Midland site the strong motion
- records selected satisfied all three conditions as documented in Holt Exhibit 5. The NRC staff raised a number of questions concerning the choice of certain. accelerometer stations and ea r thquak e s. The answers to these were addressed in the Addendum to Part I, Holt Exhibit 6.
Of particular concern with respect to the selection of the data set representing the Safe Shutdown Earthquake were
- accelerometer recordings near fault ruptures. These conditions were discussed in the Addendum to Part I (Holt Exhibit 6) and in particular Obe reasons for not including one anomalous set of records in the data set were given on pages 3-5 and in a subsequent report entitled " Basis for the Rejection of the 1966 Parkfield Earthquake Accelerograms for Use in the Midland Plant Specific Spectra" (Holt Exhibit 7) . Briefly, it is inappropriate, for purposes of developing Site Specific Response Spectra for a site such as Midland which is not close to any capable faults or tectonic structures, to select earthquake records such as those recorded from the 1966 Parkfield, California earthquake. These records are influenced D
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16 - i by the proximity of the accelerograph stations to a capable ; i j fault and therefore are referred to as "near field". (In this I case, the accelerograph stations were approximately 5 miles { from the fault line) . Moreover, the 1966 Parkfield earthquake I records are influenced not only by the observed rupture along l ] the fault but also reflect an incoherent supersonic rupture a
; across seismic barriers which occurred during that earthquake, i
j These characteristics can not reasonably be expected to occur 1 { in any earthquake in the Central Stable Region or the Michigan i Basin. This conclusion is confirmed by Dr. Otto Nuttli, an j eminent seismologist whose personal knowledge of the Central i j l United States is unexcelled, who is on record as stating: ; j "{I]t is my personal opinion that the Parkfield earthquake 4 l accelerograms at the near-field stations should not be included l in the set of spectra used to obtain a site specific response spectrum for Midland, Michigan." See Holt Exhibit 7. ' { 1 I would add that the use of near-field' records in l constructing response spectra is specifically addressed in 10 ! 1 CRF Part 100, Appendix A, Section V(a) (1) (iv): ,
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"[I]n the case where a causative fault is near the site l I
the effect of proximity of an earthquake on the spectral : characteristics of the safe shutdown earthquake shall be ! l- , taken into account." I - 1 ' j Conversely, for a site which does not require the '
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j designing for an earthquake resulting from a " causative fault l 4 near_the site," such near-field records should be excluded when I constructing the site specific response spectra. 1, i i 4 4 0
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.I C. The NRC Staf f 's Suggested Spectral Level is Arbitrary once appropriate ground motion records are collected they must be combined to form a single smoothed response spectrum.
10 CFR Part 100, Appendix A, Section VI(a) . Appendix A does 3 not state how this is to be done. Constructing smoothed response spectra combining the records of many different earthquakes statistically gives a probabilistic result. A given percentile response spectrum is a probabilistic level of ' ground motion. That is, the percentile or " spectral level" indicates the probability that ground motion will be within that response spectrum, assuming the postulated earthquake occurs. l The NRC Staff have indicated that the individual records should be combined at the 84th percentile (Holt Exhibit 3). This is judgmental and to a degree arbitrary. Statistics does i not require this result. If one is logically to establish an appropriate spectral level for earthquake ground motion for a given site, then there are two probabilistic factors to be considered; first, the probability of the occurrence of the earthquaker second, given ' that the earthquake has occurred, the probability that a
~
certain ground motion (amplitude and frequency) will occur. When these two probabilities are considered together the dominant factor is the occurrence of the earthquake. Considering the Midland site as compared to sites immediately
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- outside the Michigan Basin, the largest historical earthquake experienced at the Midland site in 180 years is two or three intensities lower. If we were to consider an intensity- ;
acceleration relationship such as developed by Trifunac & Brady (1975), this would represent a factor of two in acceleration value. Ground motion spectra for maximum historical 9 earthquakes inside and just outside the Michigan Basin if adjusted to the same spectral level, would also vary by about a factor of two. The use of the different percentile levels in constructing response spectra is one way of accounting for differences in seismicity and seismic hazard among different sites within the Central Stable Region. It is clearly unreasonable to use the same response spectral level for a site in 'the middle of the Michigan Basin as for sites outside the Michigan Basin. The ef fect of this would be to insist that the Midland Plant should be designed with greater seismic margins even though it is located in an area of lesser risk. In fact, based on a reasonable assessment of the variable seismicity within the Central Stable Aegion, a mean centered value would be a more appropriate level for the Midland site than the 84th percentile. However, a preferable approach would be to recognize that the Central Stable Region is not a tectonic province. . O e
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, IV.
SUMMARY
In summary of this testimony and the reports submitted in support of the site specific spectra for the original ground surface and the top of fill material, the following points are i noted:
- 1. The tectonic province selected by the staff to include
- the Midland site is huge and should, within the context of Appendix A, be subdivided. This is a
particularly true in Midland where the largest earthquake within 200 miles of the site, roughly the outline of the Applicant 's Michigan Basin tectonic , province, is intensity VI. l 2. The selection of a magnitude 5.3 as representative of the Safe Shutdown Earthquake for the purposes of defining the response spectra is 8 tenths (.8) of a magnitude unit higher than the estimated maximum historical earthquake (4.5) within 200 miles of the site and 3 tenths (.3) of a unit higher than the
- correct maximum magnitude 5.0 of the 1937 Anna, Ohio earthquake. Since the magnitude scale is logarithmic 1
this represents a substantial increase. i
- 3. The magnitude range selected in developing the SSRS (4.8 to 5.8) includes earthquakes of intensity IX, "
three intensities larger than the largest historical earthquakes within 200 miles of the site. The site specific response spectra therefore have been
~ .e , e, developed including records of earthquakes which have produced disasterous effects of intensity IX (see intensity scale attached as Holt Exhibit 4) . t
- 4. Considering both historical and potential earthquake intensity occurrences, the Midland site inside the Michigan Basin compared to sites on the arch systems south of the Michigan Basin would be two intensity units less. Based on this the Staff's recommendation that the SSRS be constructed at the 84th percentile level is quite high.
- 5. With respect to the spectra developed for the top of fill material, all of the above conservatisms apply and in addition, the spectra developed using actual data from similar soil conditions are higher than those which would be recommended using theoretical considerations (Holt Exhibit 8) .
When all factors are considered the spectra developed for the Midland site are unnecessarily conservative. e
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< HOLT, RICRARD J.
J' Education': 1 Bachelor of Science - Mathematics '
' Boston College, 1954 Master of Science - Geophysics Boston College, 1956 Teaching Fellow Harvard University, l'956-1957
Background:
1961-present President, Weston Geophysical Corporation.
; 1957-presenc Vice President and Senior Consulting Geophysicist, Weston Geophysical Engineers. Inc.
1957-1970 Lecturer in theoretical and applied seismology in the graduate department of geophysics, Boston , College. 1954- 1957 Geophysicist, Cahagan Geophysical Surveys Division; supervision, direction.and consultation
- on geophysical projects.
} Registrations: State of California, Registered Geophysicist, Certificate No. GP 521 Scace of Maine, Registered Geologist, certificate No. 116 Societies:
~
Acoustical Society of America American Geophysical Union
~ Association of Professional Geological Scientists Boston Society of Civil Engineers European Association of Exploration Geophysicists Seismological Society of America Society of Exploration Geophysicists Professional Experience:
Complaced more than 250 projects involving seismicity, tectonic I structural elements, ground motion prediction, and foundation conditions during twenty-five years of experience for nuclear, fossil
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% t HOLT, RICHARD J. (Cont'd) x Holt, R. J. , " Characteristics and Properties of Extreme Seismic t
, Events," Ai. Force Cambridge Research Laboratories, Office of Aerospace Research, 1966. : ,
Murphy, V. J. and R. J. Holt, " Seismic Velocities and Elastic Moduli Measurements, Mount Holyoke Range, Massachusetts, U.S.A. " Proceedings of the First Congress of International Society of Rock Mechanics, Lisbon. Portugal, 1966. Murphy, V. J. and R. J. Holt, " Engineering Seismology Applications in Deep Weathered Rock Areas " First Pan American Conference on Soil Mechanics and Foundations Engineering, Mexico, O.F. , September 1959 In addition to the above publications, 250 individual geophysical research or engineering reports for numerous private clients have been prepared directly as well as several hundred.more that have involved direct consultation. 4 9 4 s em
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1 1 l HOLT EXHIBITS 1, 2, 3, 4, 9,.and 10 e e 4
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PROPOSED MIDLAtiD SITE SPECIFIC RESPONSE SPECTRUM FOR ORIG!!iAL GROUT 1D SURFACE (MODIFIED AT L0tlGER PERIODS) 5% CRITICALLY DAMPED FIGURE 1.2 Holt Exhibit 1 i
./ o UNITED STATES i
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4,! b.;v,;t"i./,' i NUCLEAR REGULATORY COMMISSION waswmcrew. o.c.:esss
.? e 001 14 ISSO .
Docket Nos.: 50-329/330 CH, OL
. I Mr. J. W. Cook Vice President Consuwers Power Company -
1945 West Parnall Road Jackson, Michigan 49201 *
Dear Mr. Cook:
SUBJECT:
SEISMOLOGICAL INPUT FOR THE MIDLAND SITE-Cne of 2he open items associated with our radiological safety review of your application for operating licenses for Midland Plant, Units 1 and 2, and
. identified in our letter of March 30, 1979, is the establishecnt of acceptable seismological input parameters. Resciution of this open item is also necessary for approval of the remedial actions associated with the soils settlement
matter which was the subject of the December 6,1979 Order on Modification of Construction Permits.
As noted in your response to our previous requests 361.2, 361.4, 351.5 and 361.7, you consider the Michigan Basin to be a distinct tectonic province for the purpose of evaluating site seismic design input, whereas during the Midland OL review, the staff has found insufficient support that the Central Stable Region can be subdivided into separate tectonic provinces. Your approach using historic seismicity in the litchigan Basin resulted in a Safe Shutdown Earth-quake (SSE) characterized by Modified Mercalli Intensity (MMI) of VI, and a
, Modified Housner response spectra anchored at 0.12g. Discussed below is the staff's current view as to two acceptable approaches, either of which specifies
'the controlling earthquake from the Contral Stable Region and which also re-
, , quires consideration of soil amplification.
. . The controlling earthqual:e we would currently require to be used in detan.;in-ing the SSE for the Midland site is similar to that which occurred in Anna, Ohio in March 1937, and has a body wave magnitude of 5.3 M , and a Mil of VII-VIII. .
,_ .Nuttli, (State-of-the-Art for Assassing Earthquake Hands in the United States, Report 12, Credible Earthquakes for the Central U. S.: Misc. Paper S-73-1 U. S. .
.- Army Engineering Waterways Experiment Station,1978) using an, alternative method has also suggested this magnitude as the " maximum" when using residual l
' events (t' hose remaining after seismic zones such as Anna, Wabash Valley, etc.
u are removed) for the Central United States. It is .important to note that the
" July 29, 1980 Kentucky earthquake had a magnitude of 5.1-5.4 f(9 and occurred in a " residual area".
~ '
The following alternatives of characteri:ing the SSE would be acceptable to
.the staff and are consistent with the staff's Standard Review Plan (SRP)
Section 2.5.2: I , .
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Bolt Exhibit 3
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4 o?* 'O '> &Rh* 0 4 9 SITE SPECIFIC RESPONSE SPECTRA p* 7 ,-
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MIDLAND DESIGN
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SPECTRUM (SSE) 1
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84THPERCE;iTILESITESPECIFICRESP0tiSESPECTbM FOR THE TCP OF FILL .vATERIAL AND DESIGil SPECTRLN FOR THE MIDLAND tiUCLEAR PO'4ER PLAfiT - 5% CRITICALLY DAMPED - FIGURE 7 Holt Exhibit 2
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n- = _ .____.- . Mr. J. W. Cook . The Anna, Ohio earthquake of March 9,1937 is the largest historic earth- - quake in the Central Stable Region tectonic province. This earthquake had a MMI of VII-VIII and should be assumed to occur near the site , (Appendix A to 10 CFR Part 100, SRP Section 2.5.2). Using this inten-sity one acceptable approach would be based upon the standardized responso spectra of Regulatory Guide 1.60 anchored at 0.199 as determined by the trend of the means of the intensity acceleration values in Trifunac and Brady (Seismological Society of America Bull., V. 65,1975). , An alternative method of describing the SSE and response spectra result-ing from an " Anna" type earthquake assurred to occur near the site involves using the magnitude. As was indicated during the recent OL review en Sequoyah, magnitude may be a more realistic estimate of earthquake size than intensity. Therefore a description of the SSE can also be obtained by
. collecting representative real time histories for a magnituoe of -
5.3 + .5 M , epicentral distances less than 25 kilometers at soil sites. Such a colN8 tion has been mada by Lawrence Livermore Laboratory (LLL, Draft, Seismic Hazard Analysis: Site Specific Response Spectrc Results, August 23, IM) but it would be beneficial if you update this data set as appropriate. It is the staff's position that the representation appropri-ate for use in establishing the SSE is the Both per:entile of the response
, . spectra as derived directly from the real time nistories, i- The input for the ccmoarative analysis of your present response spect-a (Modified Housner) and Regulatory Guide 1.60 both anchored at 0.12g was at the foundaticn level. It is our conclusion that the appropriete location for
. . vibratory ground motion input for your Midland site be at the top of the natural glacial till (essentially the original reigional ground surface). Above this till is a thin sand layer which is highly variable in density and the ccm-
. pacted fill that was placed to raise plant grade Therefore either of cur above acceptable approcches will also require an assessment of soil amplifica-
. - - tion from the tiil surface. .
We are available to meet with you at your earliest opportunity to discuss the -
. . above approach in order that acceptable data and raethods of describing vibra-tory ground motion can be utilized for the Midland site.
. Contact our project manager. Darl Hood, if you wish 'to arrange s'uch a meeting or desire clarification of this letter.
Sincerely, h k1E4.40 ' Robert L. Tedesco Assistant Ofrector for Licensing Division of Licensing -
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cc: See next page , e * *
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. ~ l cc: Michael I. Miller, Esq. Isham, Lir.coln & *eale Mr. Don van Farowe, Chief
- Suite 4200 Division cf Radioissical f.'esitn i Department cf Public Ecal:::
1 First National Plaza P. O. Ecx 33035 Chicago, Illinois 60603 Lansing Michigan 4E909 - Judd L. Bacen, Esq. William J. Scanlon, Esq. Managing Attorney 2034 Pauline Boulevarc Consumers Pcteer Cc pany Ann Arbor, Michigan 4S103 212 West Micnigan Avenue Jackson, Michigan 49201 U. S. Nuclear Regulatcry.Cermissicn Resident Inspectors Cffice Mr. Paul A. Perry, Secretary
^
Route 7 Consun.ers Pctce Cc:pany Midland, Michigan 42540 212 Fest Michigan Avenue Jackson, Michigan 49201 Es. Barbara Stamiris 5795 N. P.tver Kyron M. Cherry, Esq.' Freeland, Michigan 43623 1 IER Plaza Chicago, Illincis 60611 Ms. Sharon K. Warran
, - Ns. Mary Sincl~ air 636 Hillcrest Midland, Michigan 45640 5711'Summerset Drive .
Midland, Michigan 48640 Frank J. Kalley, Esq. . Attorney Gar.eral
* ~
State of Nichigan Enviror.:cr.tal . Protection Civisica .
. . 720 Law Suf1 ding
- Lansing, Michigan 48913 .
Mr. Wendell Karshall '
Route 10 . Midland, Michigan 48640 Mr.' Steve Gacier 2120 Carter Avenue
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. St. Paul, Mincasot: 55108 .
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cc: Commander, Naval Surface heapen: Center ATTN: P. C. Huang ' G-402 White Oak Stiver Spring, Maryland 20910 ' Mr. L. J. Auge, Manager
, . Facility Design Engineering Energy Techno!cgy Engineering Center P. O. Box 1449 Canoga Park, California 91304 Mr. William Lawhead U. S. Ccrps of Engineers .
NCEEC - T 7th Ficor 477 Michigan Aver.ce Detroit, Michigan 46225 Charles Bechheeffer, Esq. Atomic Safety 2 Licensir.g Board U. S. Nuclear Regula: cry Cot ission Washingten, C. C. 20555 Mr. Gustave A. Licenterger Atomic Safety a Licensing Ecard U. S. Nuclear P.4ge atory Commissicn Washington, D. C. 20555 Dr. Frederick P. Ccuan Apt. B-125 6125 N. Verde Trai! Boca Raton, Florica 33433 l se. e 0.0 h a to t- *
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6 The Modified Mercalli Intensity Scalel l Wood, Harry O. and Frank Neumann, 1931, Modified Mercalli Intensity Scale of 1931, Bulletin of the Seismological Society of America, Vol. 21, No. 4. 9
.. Holt Edibit 4
[_ . _ . . . _ . . - . . ._ .__. .. . . . _ . . . . . _ . . . _ . . . - i I Not felt except by a very few under especially favorable circumstances. (I Rossi-Forel Scale.) II. Felt only by a few persons at rest, especially on upper floors of buildings. Delicately suspended objects may t swing. (I to III Rossi-Forel Scale.) III. Felt quite noticeably indoors, especially on upper floors of buildings, but many people do not recognize it as an
; earthquake. Standing motorcars may rock slightly.
Vibration like passing truck. Duration estimated. (III Rossi-Forel Scale.) - IV. During the day felt indoors by many, outdoors by few. At night some awakened. Dishes, windows, and doors disturbed; walls make creaking sound. Sensation like i heavy truck striking building. Standing motorcars rocked noticeably. (IV to V Rossi-Forel Scale.) V. Felt by nearly everyone; many awakened. Some dishes, windows, etc., broken; a few instances of cracked plaster; unstable objects overturned. Disturbance of trees, poles, and other tall objects sometimes noticed. I t Pendulum clocks may stop. (V to VI Rossi-Forel Scale) VI. Felt by all; many frightened and run outdoors. Some heavy furniture moved; a few instances of fallen plaster or damaged chimneys. Damage slight. (VI to VII Rossi-Forel Scale.) VII. Everybody runs outdoors. Damage negligible in buildings of good design and construction; slight to moderate in
?
well built ordinary structures; considerable in poorly built or badly designed structures. Some chimneys broken. Noticed by persons driving motorcars. (VIII- Rossi-Forel Scale.) , VIII. Damage slight in specially designed structures; considerable in ordinary substantial buildings, with 3 partial collapse; great in poorly built structures. Panel walls thrown out of frame structures. Fall of chimneys, factory stacks, columns, monuments, walls. Heavy furniture overturned. Sand and mud ejected in small amounts. Changes in well water. Persons driving motorcars disturbed. (VIII+ to IX Rossi-Forel Scale.)
, IX. Damage considerable in specially designed structures; well-designed frame structures thrown out of plumb; great in substantial buildings, with partial collapse.
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}. ! g. ' I - l~ Buildings shifted off foundations. Ground cracked . I l conspicuously. Underground pipes broken. (IX+ Rossi-Forel Scale. ) X. Some well-built wooden structures destroyed; most masonry and frame structures destroyed with foundations; ground badly cracked. Rails bent. Landslides considerable from ! i l.- river banks and steep slopes. Shifted sand and mud. I: i Water splashed (slopped) over banks. (X Rossi-Forel Scale.) I j XI. Few, if any (masonry), structures remain standing. Bridges destroyed. Broad fissures in ground. Underground pipelines completely out of service. Earth i slumps and land slips in soft ground. Rails bent greatly. XII.. Damage total. Waves seen on ground surfaces. Lines of sight and level distorted. Objects thrown upward into the air. 5
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I HOLT EXHIBITS 5, 6, 7, and 8 ARE BOUND SEPARATELY i
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6: SEISMICITY OF NORTHEASTERN AND NORTH CENTRAL . UNITED STATES AND ADJACENT CANADA , FIGURE 1 [" -["
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Holt Exhibit 9 , ..
,f!1
. t
SUMMARY
OF THE APPLICANT'S POSITION WITH RESPECT
. TO THE MIDLAND SITE SPECIFIC SPECTRA INTRODUCTION In developing site specific spectra for the Midland Plant a number of reports have been submitted to the U. S. Nuclear Regulatory Commission and a number of meetings have been held.
As suggested by the NRC staf f in our most recent meeting on September 16, 1981, this letter is written to briefly summarize
, the applicant's position.
The reports consist of: Site Specific Response Spectra, Midland Plant - Units 1 and 2: Part I - Response Spectra - Safe Shutdown Earthquake Original Ground Surf ace; Site Specific Response Spectra, Midland Plant - Units 1 and 2: Addendum to Part I Response Spectra - original Ground Surface; Site Specific Response Spectra, Midland Plant - Units 1 and 2: Part II - Response Spectra Applicable for the Top of Fill Material at the Plant Site; Site Specific Response Spectra, Midland Plant - Units 1 and 2: Part III - Seismic Hazard Analysis; and Basis for the Rejection of the 1966 Parkfield Earthquake Accelerograms for Use in Midland Plant Site Specific Spectra. Holt Exhibit 10
. . - _ . . . - . . . . _ . . _ . . ..._ _. ... . . . . . _ = . . . . . .-. ._.
TECTONIC PROVINCE The applicant proposed the use of the Michigan Basin as a j tectonic province in the Midland FSAR. The Michigan Basin is a saucer-like tectonic unit with shallow dips whose crystalline basement rock is nearly 6,000-8,000 feet deeper than the basement rock arches which surround it. These arches have been
, areas where earthquakes of approximate magnitude 5 and intensities VII or slightly larger have occurred. Within the basin the largest magnitude has been 4.5 and the maximum ... .
earthquake intensity has been VI. The Midland site which is centered near the middle of the Michigan Basin has experienced a maximum intensity of IV to V in historical times based on a conservative attenuation estimate. The largest earthquake within 200 miles of the site is VI. The nearest earthquake to the site which exceeded VI is the Anna, Ohio Earthquake of 1937 (205 miles away) with an estimated magnitude of 5.0 and an intensity of VII, or VII to VIII. The Anna, Ohio earthquake is spacially located on the Cincinnati-Findlay Arch system. The NRC staff has requested that the applicant use the Central Stable Region as a tectonic province and to consider a magnitude 5.3 m b earthquake at the site. The Central Stable Region is a very large region of the United States defined not on the characteristics of the . crystalline basement rock which underlies it, but on the sedimentary rock strata which overlie the basement and its was
il _.__ _,_._,
. . . . . ~ . . . - - - - . . .
--~~7-~.~ - I 6
1
SUMMARY
OF THE APPLICANT'S POSITION WITH RESPECT
} TO THE MIDLAND SITE SPECIFIC SPECTRA ;
1 ; I INTRODUCTION i 4 I l , In developing site specific spectra for the Midland Plant a - f number of reports have been submitted to the U. S. Nuclear f
- Regulatory Commission and a number of meetings have been held.
[ As suggested by the NRC staff in our most recent meeting on i t i September 16, 1981, this letter is written to briefly summarize { } the applicant's position. 4 i The reports consist of: l i i ) - Site Specific Response Spectra, Midland Plant - Units 2 l 1 and 2: Part I - Response Spectra - Safe Shutdown Earthquake Original Ground Surf ace; i Site Specific Response Spectra, Midland Plant - Units 3
; 1 and 2: Addendum to Part I Response Spectra -
l Original Ground Surface; Site Specific Response Spectra, Midland Plant - Units i i 1 and 2: Part II - Response Spectra Applicable for 1 5 i the Top of Fill Material at the Plant Site; l . Site Specific Response Spectra, Midland Plant - Units : l 1 and 2: Part III - Seismic Hazard Analysis; and ' f . j - Basis for the Rejection of the 1966 Parkfield ! } Earthquake Accelerograms for Use in Midland Plant i : j Site Specific Spectra. } i i I l j i' Holt Exhibit 10
- TECTONIC PROVINCE The applicant proposed the use of the Michigan Basin as a tectonic province in the Midland FSAR. The Michigan Basin is a saucer-like tectonic unit with shallow dips whose crystalline basement rock is nearly 6,000-8,000 feet deeper than the basement rock arches which surround it. These arches have been areas where earthquakes of approximate magnitude 5 and intensities VII or slightly larger have occurred. Within the basin the largest magnitude has been 4.5 and the maximum ear.thquake intensity has been VI. The Midland site which is centered near the middle of the Michigan Basin has experienced a maximum intensity of IV to V in historical times based on a conservative attenuation estimate. The largest earthquake within 200 miles of the site is VI. The nearest earthquake to the site which exceeded VI is the Anna, Ohio Earthquake of 1937 (205 miles away) with an estimated magnitude of 5.0 and an intensity of VII, or VII to VIII. The Anna, Ohio earthquake is spacially located on the Cincinnati-Findlay Arch system. ,,
The NRC staf f has requested that the applicant "se the Central Stable Region as a tectonic province and to consider a magnitude 5.3 m b earthquake at the site. The Central Stable Region is a very large region of the United States defined not on the characteristics of the crystalline basement rock which underlies it, but on the
- sedimentary rock strata which overlie the basement and its
-.n .
U geomorphic features. Undoubtly since larger earthquakes will originate in the crystalline basement rock its structure which is revealed at least in part by its subsurface relief is important in the definition cf a tectonic province. Although it is strongly believed that there are several definable tectonic provinces and structures within the Central Stable Region, we have, in accordance with the staff's requirement taken an earthquake of magn'itude 5.3 at the Midland site. SELECTION OF EARTHQUAKES FOR DETERMINATION OF SITE SPECIFIC SPECTRA In selecting strong motion records to develop site specific spectra, three basic elements are used, a defined magnitude, distance and shear wave velocity range to simulate a Safe Shutdown Earthquake occu;rence at the site. For the Midland site (Part I) all three conditions were satisfied and have been documented. The NRC staf f raised a number of questions concerning the choice of certain accelerometer stations and earthquakes. The answers to these were addressed in the Addendum to Part I listed above. Of particular concern with respect to the selection of the data set representing the Safe Shutdown Earthquake were accelerometer recordings near f ault ruptures. These conditions a were discussed in the Addendum to Part I and in particular the
~
reasons for not including them in the data set were given on 4 m
. i
_ _ . . . . . _ . . _ _ . . .._. _ _ _ . . - . . - . . ~ . - - - --- - Earthquakes are caused by fault ruptures which may or may not break ground surface. Almost all of the strong motion l records which are used for developing site specific spectra are
, from active tectonic areas where surface faulting (capable faulting) may exist; such faulting may or may not be involved with any given earthquake. In some instances surface faulting is observable and well known; in other instances it is partially obscured and for many earthquakes the faulting is confined to depth and the surface is not ruptured. This latter case is characteristic of Eastern United States earthquakes.
The set of strong motion records for a given magnitude range may contain data from earthquakes associated with surface rupture. When this is known and the record (s) have anomalous characteristics due to capable faulting, they should be eliminated from a data set selected to represent a tectonic province unless the site is considered as near a tectonic structure or suspected tectonic structure (within a province). SEISMIC HAZARD ANALYSIS Introduction The results of the Seismic Hazard Analysis (Part III) and in particular those presented to NRC on September 16, 1981, indicate that the Midland site at a given probability level of occurrence would have a full intensity value less compared to other sites in the Central stable Region. This means that if the maximum earthquake at Anna, Ohio at a given probability
e level has a potential intensity of VII-VIII (0.2 g), then Midland would have an intensity of VI to VII (0.1 g). The i spectral level for the Midland site should be effectively lower than for other areas of the Central Stable Region. These relative probabilities confirm that the Midland site is not near any important seismic sources, tectonic structures or of course capable faults which are demonstraole geologically and seismologically. As previously stated, seismic design criteria, i.e. design response spectrum levels, for the Central United States region are derived on the basis of the Central Stable Region. The design earthquake for this broad region is defined to be the
" Anna, Ohio type" event characterized by a magnitude of approximately 5.3 and an intensity of VII-VIII. By postulating this broad Central Staole Region (Figure 1), all sites contained are thus assigned equivalent design criteria and therefore are implicitly assigned equivalent estimates of seismic hazard.
From the standpoint of the known distribution of historic seismic activity in the Central United Statos, the concept of equivalent hazard and the use of equivalent design criteria at all sites in the Central Stable Region are not supported. A demonstration of the non-uniform historical seismic exposure at six sites in the Central Stable Region is shown in Figure 2. The figure is a plot of the estimated seismic e.<posure at six
~
7_ sites in terms of felt intensities from the known historical earthquakes located in the broad region of Eastern North I America (30*-50*N: 60*-100'W). This figure demonstrates a
, significant difference in the historical seismic exposure of the Central Stable Region sites. Site 6, which is the Midland site, is located in the Michigan Basin, an historically aseismic portion of the Central Stable Region, and is demonstrated by the figure to have experienced maximum
, intensities ranging to two intensity units lower than other Central Stable Region sites. Also, the number of exceedances of various intensities is lower at the Midland site than at the other Central Stable Region sites, by f actors ranging to 3.5.
Formal seismic hazard calculations for the same six Central Stable Region sites, for several alternate seismotectonic
. models of the central region, also demonstrate the significantly varying seismic hazard among Central Stable Region sites, with the Midland site having the relatively lowest seismic hazard. Figure 3 shows the seismic hazard results for the assumption of the broad Central Stable Region.
. . These results indicate equivalent hazard at all sites examined, except for slightly increased hazard at the site (40.5'N, 88.0*W) nearest to the New Madrid seismic zone. These results, however, do not compare well with the historical seismic exposure of these sites (Figure 1) which illustrates a significantly larger variation.
, . _ . . . . . . . . . . . ...~- --... . - . - - - - - --- - -- - 1' t Figures 2 and 3 present the seismic hazard results for ' tectonic models which define limited seismic source regions - i within the broad Central Stable Region, at locations that have
,., experienced significant historical activity or conversely i
.. absence of activity. For example, the results shown in
' l Figure 4 are based on a tectonic model (Figure 5) that includes 1 the Michigan Basin Tectonic Province (lower historic activity) i = and the Cincinnati Arch-Findlay and Waverly Arch geologic
, structure (greater historic activity) contained within the !
i Central Stable Region. The results for this model illustrate a d significant variation in seismic exposure at the six sites, the i i . pattern of which closely resembles the historical exposure of i these sites shown in Figure 1. .For instance, the probabilistic I hazard and the historical exposure is lowest at Site 6, the t ) Midland site, and greatest at Sites 1, 3, and 5. , 1 Similarly, a tectonic model that incorporates the hazard results for tectonic structures within the Central stable j Region represented by the regions of historical maximum ,, t - activity at Anna, Ohio and Attica, New York (Figure 6), more , I closely models the historically observed exposure than does the broad Central Stable Region model. These results, shown in Figure 7, again demonstrate the lowest hazard at Midland and j the greatest hazard at Sites 1, 3, and 5, which parallels the ] - pattern of historic exposure. i r } i f
_9 The historical exposure and also the probabilistic seismic hazard results demonstrate a nonuniformity of seismic ground motion potential within the Central Stable Region. The
, relative differences range to two intensity units at the examined Central Stable Region sites at some given annual probability level and to between one to two orders of magnitude variations in annual probability at some given seismic intensity.
,, From both the historical and the probabilistic analyses, the Midland site is demonstrated to have the relatively lowest seismic exposure and hazard in comparison to other Central Stable Region sites.
I This lower seismic hazard at the Midland site is considered L in the recommendation of seismic design criteria. An 84 DU percentile level has been suggested for averaging strong motion
. . data. Yet, in numerous applications of NRC criteria, whenever the Regulatory Guide 1.60 spectral shape is used at some anchoring acceleration level, the 84th percentile is not rigorously applied across the entire frequency band. The ultimate level used in a specific design should take into consideration the seismic hazard at the site. Although the design spectrum recommended for Midland in the part I submittal th is the 84 percentile SSRS, for magnitude 5.3, a strong argument could be made for accepting a lower percentile level based on actual seismicity and predicted ground motions.
u_________________._________.____.______ _ _ . . - _ _ _ _ _ _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _
u .. ______ _ -- - _ - - - - - - - - - - - - - - This recommended spectrum lies between the 72nd to 76 th percentile of the suite of records including the Parkfield recordings (Figure 8). From the standpoint of the relative seismic hazard differences which suggest a reduction of 1 to 2 intensity levels at the Midland site from other central Stable Region sites, which can be translated into a factor of 2 to 4 reduction in peak horizontal ground acceleration, the difference between the 84th and 72 nd to 76 th 1,y,13, which is about 30%, is not significant. It is our continued conviction that the spectra given for the Midland Plant design as presented in the Part I and II , reports cited above, is a conservative assessment within the context of Appendix A. e* 9 i 9 I g a N be
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N l September 29, 1981 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of CONSUMERS POWER COMPANY Docket Nos. 50-329 OM & OL 50-330 OM & OL (Midland Plant, Units 1 and 2) ! TESTIMONY OF DR. PAUL F. HADALA WITH RESPECT TO THE STUDY OF AMPLIFICATION OF EARTHQUAKE INDUCED GROUND MOTIONS AND THE STABILITY OF THE COOLING POND DIKE SLOPES UNDER EARTHQUAKE LOADING Q1. Please state your name and position. j A1. My name is Paul F. Hadala. Since January 1980, I have served i as Assistant Chief of the Geotechnical Laboratory of the U.S. Army i Engineer Waterways Experiment Station in Vicksburg, Mississippi. Q2. Have you prepared a statement of professional qualifications? A2. Yes. A copy of this statement is given as Attachment 1. Q3. Please state the duration and nature of your responsibilities with respect to the Midland Plant Units 1 and 2. A3. I began work, on a part-time basis, on the review of geotechnical earthquake engineering issues at the Midland plant in February 1980 under the fiscal sponsorship of the U.S. Army Engineer District, Detroit. I visited the site on 27-28 February 1980 and inspected those areas discussed in this testimony and have reviewed all the sections of the FSAR and " Responses to NRC Questions Regarding Plant Fill" relevant to geotechnical earthquake engineering. I have authored 1 1
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', 1 two memoranda and two letter reports summarizing my findings. They are listed as References 2 end-5 in Attachment 2.
Q4. Please state the purpose of this testimony. i A4. My testimony is in two parts. The first concerns an analytical study of amplification and deamplification of earthquake induced ground motions at the Midland Michigan Nuclear Power Plant Site that I performed to assist in the evaluation of the deteministic site-specific response spectra for the top of the fill (See Reference 1). The second part concerns the earthquake safety of the baffle dike, perimeter dike, and emergency cooling pond slopes (see Attachment 6 for location), as they affect Category I return piping leading from the Service Water Intake Structure to the Emergency Cooling Water Reservoir. QS. What studies have you conpleted concerning amplification or deamplification of gro'und motion for the Midland project? A5. I was initially requested by the NRC Staff through the Detroit District, Corps of Engineers, to review Appendix B of the Applicant's study, which is entitled " Site Specific Response Spectra, Midland Plant Units 1 and 2, Part II Response Spectra Applicable For The Top Of The Fill Material At The Plant Site" (Reference 1) and later was asked to make further analytical studies of site amplification in the same subject area as Appendix B. My findings are contained in a Memorandum for Record, "Effect of Plant Fill on Seismic Grour;d Hotion Environment at the Midland Michigan Nuclear Power Plant" (Attachment 3). The objective of my investigation was to determine, via theoretical-stress wave propagation analyses, the effect of the addition of the plant I fill in its present condition on the characteristics of earthquake ground i i i k
'. !I motion at the top of the fill. In its study for establishing ground motion at top of the fill (Reference 1), the Applicant developed a deterministic site-specific response spectrum for the top of the fill which took into account the potential for amplification of the spectrum at the top of fill due to a thicker soil layer overlying bedrock. In Appendix B of this study, the Applicant conducted a series of one-dimensional vertical shear stress wave propagation analyses to determine, for this particular analytical model and the selected range of input parameters, what the effect of the addition of the plant fill would be on the ground motion environment at the new ground surface, the top of the plant fill.
My review of the Applicant's study (Appendix B) indicated that the Applicant had used two factors not representative of actual conditions. First, the stiffness of the fiil and the top 50 feet of the till as actually used in the calculation were substantially-softer than indicated by the range of the field shear wave velocity data. Second, all the calculations were perfomed with the El Centro record scaled to 0.12g. This record was obtained on deep alluvium in a Richter Magnitude 6.7 event. However, the record was used as if it were a rock outcrop record. In conducting one-dimensional stress wave propagation analysis, accelerograms should be used as outcrop motions only for those layers whose stiffnesses are approximately consistent with those of the site where the accelerogram was recorded. As a result, I recommended that additional calculations be perfomed to (1) use a range of site , stiffnesses bracketing the field measurements, (2) investigate a range of different accelerograms scaled to 0.12g and (3) investigate the effects l
__l ___ . ; ~ 7 '~::. 2 : r ~ ~ ~ '_ of variation in the layer used as outcrop (that is, in simple terms, the point in the site profile where the input motion is applied in the calculation). Some of these calculations were perfomed by the Applicant and I perfomed 13 myself. One of my calculations duplicated one of the 1 Applicant's calculations so I am confident that wa both were perfoming the mechanics of the calculations correctly. The analytical method used by the Applicant to detemine the effect of the addition of the plant fill on the ground motion environment at the top of the plant fill is the same one I chose to use. The method used is the SHAKE Computer Code which is described in Reference 7. It solves, in closed . form, the problem of one-dimensional vertical propagation of shear wave in linear viscoelastic layered media. The nonlinear character of earth materials are accounted for through the use of the equivalent linear method in which engineering properties of a layer in the next iteration are adjusted to be physically compatible with the strain levels experienced in that layer in the last iteration. This process converges to a compatible set of strains and stresses in a few iterations. To use this method the horizontal ground motion in any direction must be known or assumed at some point in the layered soil profile. The horizontal motions in this same direction at all other points including the ground surface can then be computed if the stress-strain properties, mass density and thickness of each layer of soil above bedrock are known. The method is widely used to study the effects of variations in the soil profile on earthquake ground motion environments as indicated by Professor John t.ysmer in his state-of-the-art address at the 1978 ASCE Earthquake Engineering and Soil Dynamics Specialty Conference. I can
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also testify to this based on my own knowledge and my own professional s \cener experience aspChief of the Earthquake Engineering and Geophysics Division at the U.S. Army Engineer Waterways Experiment Station. The most obvious limitations of the SHAKE Code is its one-dimensional nature. Earthquake ground motion are three dimensional in character. However, horizontal motions are more severe than vertical in most cases and it is the horizontal component that is calculated with SHAKE. The one-dinensional assumption becomes more nearly correct when the horizontal dimensions of the site greatly exceed the vertical ones. The plant fill is nominally 30 feet thick. The minimum horizontal dimension, the width of the plant fill is 1200 feet. In this ' case, the use of SHAKE to study the effects of the addition of plant fill or variations in its soil properties in terms of a ratio or percentage change from the original ground surface motion, as was done for the Midland project is acceptable and entirely in line with the current state of the art in the analysis of local site effects on earthquake ground motions. . I found, as a result of my analysis of the 13 SHAKE code calculations which I performed, that the effect of adding the fill on the ground motion environment can be different depending on the stiffness of the fill (upper or lower bound), the accelerogram chosen and the layer in the soil profile which is chosen as the outcrop. The effect of using a stiffer and more realistic range of site properties was to increase the top of fill shock spectrum amplitude in the low frequency range. The effect of using a more realistic outcrop selection and records from more realistic magnitude earthquakes was to I s
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decrease the response spectrum amplitude. This was detemined by comparing the results for the El Centro record used as a bedrock outcrop (unrealistic) and as a till outcrop (more realistic) and by comparing the results of the run with the El Centro record (Magnitude 6.7) with those from a calculation with the record (Magnitude 5.0). At the request of the NRC Staff calculations were also made with the Temblor rock site record (tiagnitude 5.6) treated as an outcrop of the Sagninaw Formation bedrock. The calculations demonstrated that changes from one type of outcrop to another had a larger effect on the calculated response spectrum than did changing the accelergram. From the results of the SHAKE calculations, I calculated ratios of response spectra. These ratios are the ratio of the response spectral amplitude at the top of the fill to that spectral amplitude at the original ground surface at each frequency and are a measure of the effect of adding plant fill to the site. That is, it indicates that the amplitude of the site-specific response spectrum for the top of the fill in a linear relation to that for the orfginal ground surface. To do this, I also had to make SHAKE calculations for the site as it was before the fill was added. The Applicant developed a curve of the ratio of the response spectrum for the top of the fill to that at original ground surface versus frequency, by ratioing the deteministic site-specific response
-spectrum he developed for the top of the fill with another he developed for the original ground surface. This latter spectrum for top of till is reported in a study entitled " Site Specific Response Spectra, liidland Plant--Units 1 and 2, Part I, Response Spectra--Safe Shutdown Earthquake.
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s Original Ground Surface," (Reference 6). The ratio versus frequency curve developed by the Applicant is the upper solid curve in Attachment 5. Attachment 4 shows my results for all cases where the combination of site properties, outcropping layer and accelerogram were physically consistent with one another and with conditions at the site. The results obtained by the Applicant are shown in Attachment 5. As shown by these figures, with minor exceptions, the ratio for the site-specific response spectra (SSRS) developed by the Applicant exceeds those ratios calculated by the analytical method. This means that if one accepts the Applicant's site-specific response spectrum for the original ground surface as valid, then the Applicant's site-specific response spectrum for the top of fill is more conservative than the one which would be obtained by application of one-dimensional stress wave calculation methods using an original ground surface outcrop and a suite of accelerograms whose response, spectrum did not exceed the SSRS for the original ground surface. The _ SHAKE code calculations indicate that the amplification of the SSRS for the top of fill over that at the top of till developed by the Applicant through the analysis of empirical data is more conservative than one developed by application of theoretically calculated amplification , factors. Q6. What is the purpose of your testimony on the stability of the cooling pond dike slopes under earthquake loading. A6. In Reference 2 I raised a question concerning the need for a state-of-the-art analysis of the seismic safety of the slopes of the Emergency Cooling Water Reservoir (ECWR) because of possible effects of l
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movement of these slopes on service water return lines buried in these slopes. This was relayed to the Applicant by the NRC Staff in 10 C.F.R. I 50.54g, Question 45f. The testimony of Dr. A. J. Hendron, Jr., before this Board on August 12, 1981, addressed this question, and I have I thoroughly reviewed Dr. Hendron's testimony and spot checked the analysis presented as part of that testimony. The purpose of this testimony is to infom the Board of my conclusions on dike stability under seismic loading which was identified in the August session as an unresolved safety concern. Q7. What are your conclusions on the stability of the cooling pond dike under seismic loading? A7. In my professional opinion these analyses were conducted by the Applicant using an appropriate state-of-the-art method. In my opinion the analyses are based on ample shear strength data, and conservative assumptions were made throughout. The yield accelerations at which the slopes will' begin to defom inelastically (that is, start moving) as computed by Dr. Hendron are 2.8 to 10.3 times larger than a maximum acceleration of 0.19g, an acceleration which is well beyond the Safe Shutdown Earthquake. I have independently confirmed the lower end of this range and concur that Dr. Hendron's 2.8 value is the lower bound of seismic resistance which is actually available in these slopes. Since the yield acceleration must be exceeded before the service water discharge lines could be damaged, and since in my judgment the yield accelerations will not be exceeded during an earthquake even if its peak acceleration were as high as 0.5g I can
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- 9-definitely conclude that slope movement will not endanger the safety of the cooling pond water discharge lines.
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. ATTACHMENT la STATEMENT OF PROFESSIONAL QUALIFICATIONS OF PAUL F. HADALA i
I earned the Bachelor of Civil Engineering Degree from Union College l in Schenectady, New York in 1959. I then enrolled in the Graduate School of Arts and Sciences Harvard University, and studied Soil Mechanics ; under Professor A. Casagrande. I received the Master of Science Degree l from Harvard in 1960. After a short period of time with the Bureau of j Soil Mechanics, New York State Department of Transportation. I began what has thus far been a 21-year professional career of research and i engineering studies in the field of Soil Dynamics at the U.S. Amy Engineer Waterways Experiment Station (WES), Vicksburg, Mississippi.
- While employed at WES, I did additional graduate work in Theoretical and Applied Mechanics at Mississippi State University. I attended the Graduate College of the University of Illinois at Champaign-Urbana from September 1967 to August 1968 and took additional graduate work in Soil i
Mechanics, Foundation Engineering, Numerical Methods and Theoretical and Applied Mechanics. I did my Ph.D thesis research in absentia at WES and received the degree of Doctor of Philosophy in Civil Engineering from the University of Illinois in 1973. My thesis supervisor was Professor A. J. Hendron, Jr. My major was Soil Mechanics. I have a full minor in Theoretical and Applied Mechanics and a half minor in Numerical Methods of Structural Analysis.
'With the exception of the period of time in residence at the University of Illinois, I worked in the Soil Dynamics Division of the Geotechnical Laboratory of the U.S. Amy Engineer, WES, from 1960 until October 1977 as a Civil Engineer and later as a Research Civil Engineer.
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_2 During this time I conducted research on military application of soil dynamics. As part of this work I conducted experiments in stress wave propagation in soils, wrote the first stress wave propagation computer code at WES, perfomed numerous analyses and one- and two-dimensional - calculations of stress wave propagation and ground motion in earth media and wrote numerous reports on the subject. Their titles are among those given at the end of this attachment. Practical applications of this work included the development of ground motion design criteria for ballistic missile defense systems. i In October 1977, I was appointed as a Supervisory Research Civil Engineer and Chief of the WES Geotechnical Laboratory's Earthquake and Geophysics Division and switched from military to civilian applications of Soil Dynanics. In that capacity, I supervised and reviewed the research of others in the fields of earthquake induced stress wave propagation, dynamic properties of soils under earthquake excitation, stability of slopes during earthquakes, and geophysical field investigation methods. My organization also conducted, for other Corps of Engineers offices, earthquake analyses of Barkley and Richard B. Russell Dams during this period. I supervised these efforts and technically reviewed their results. Since January 1980, I have served as Assistant Chief of the Geotechnical Laboratory at WES. In addition to my duties as a manager, I continue to be active in soil dynamics research. I am a licensed Professional Engineering in the State of Mississippi, a member of. the American Society of Civil Engineers, and the author or co-author of six published papers and 53 WES technical reports of which l l 1 i
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all are in the field of Soil Dynamics. A resume containing more details follows: TECHNICAL QUALIFICATIONS AND SCIENTIFIC CONTRIBUTIONS OF DR. PAUL F. HADALA
- 1. Educational
Background:
- a. College (s) attended:
(1) Union College, Schenectady, New York, BSCE, 1959. (2) Harvard University, Cambridge, Massachusetts, MS,1960, received Gordon McKay Scholarship Prize awarded for excellence of undergraduate record, received Augston B. Mason Prize awarded by the Harvard' soil mechanics staff for excellence in soil mechanics graduate program. (3) University of Illinois, Urbana, Illinois, September 1967 - August 1968; canpleted course requirements for PhD in Civil Engineering degree under Department of the Army Program of Long-Term Training and Education of Civilian Employees. Thesis completed in absentia; PhD awarded in August 1973. (4) Mississippi State University - Vickisburg Mississippi Graduate Center, 6 semester hours Applied Elasticity (1965) and Theory of Plasticity 1966.
- 2. Degrees (kind and when awarded):
- a. BSCE, June 1959, Union Collge, Schednectady, New York,
- b. MS, Soil Mechanics, June 1960 Harvard University, Cambridge, Mass.
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- c. PhD, Soil Mechanics and Foundation Engineering, August 1973 University of Illinois at Champaign-Urbana.
- 3. Dr. Hadala has been a registered professional engineer with the State of Mississippi since 1964. He is currently a member of the American Society of Civil Engineers (ASCE) and is a reviewer for the ASCE Geotechnical Journal. He has received the Waterways Experiment Station Research and Development award (1973) the Meritorious Civilian Service Award (1976) recognition of his research contributions to the Corps of Engineers.
- 4. Professional experience:
- a. 07/60 - 11/60: Junior Engineer at the New York State Public Works Department, Albany, New York. Duties included supervision of laboratory testing, analysis, and design in connection with embankment stability and settlement problems, field inspection, report and specification writing plus limited supervision of drilling and piezometer installation as the Project Engineering at the Bureau of Soil Mechanics.
- b. 01/60 - 11/63: Civil Engineer Assistant with the US Army Engineering Waterways Experiment Station, Vicksburg, Mississippi. Duties included planning and execution of static and dynamic plate bearing tests, supervision of soils testing, analysis of data, development of scaling laws, and data reduction while working as the Assistant Project Engineer on Dynamic Bearing Capacity of Soils project.
- c. 12/63 - 10/64: Civil Engineering with the US Army Engineer Waterways Experiment Station, Vicksburg, Mississippi. Duties included development of design criteria for the foundation of buried protective structures, dynamic plate bearing test, supervision of technicians and
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enlisted civil engineer assistants, soils laboratory test supervision, analysis, report writing, and computer programing while working as the l Project Engineer on Dynamics Bearing Capacity of Soils project.
- d. 10/64 - 10/66: Research Civil Engineer with the US Amy Engineer Watereays Experiment Station, Vicksburg, Mississippi. Duties included planning and conducting of' soil dynamics research and monitoring contract research in the above areas. Work included supervision of other project personnel, cranduct, analysis, and reporting of test results, the planning of data reduction systems, and computer programing while working as the Project Engineer on foundation response studies and Blast Load Generator free-field test.
- e. 10/66 - 12/72: Research Civil Engineer with the US Amy Engineer Waterways Experiment Station, Vicksburg, Mississippi. Duties included acting as the Research Team Leader responsible for fomulating and conducting systematic research attacks on nuclear explosion ground shock and soil structure interaction. Involved soil mechanics, engineering mechanics, dynamics, and related disciplines. Also involved application of research results to SAFEGUARD System design and hardness assurance. Perfonned supervisory and administrative duties for Section Chief for six months.
- f. 12/71 - 10/77: Research Civil Engineer with the US Army Engineering Waterwyas Experiment Station, Vicksburg, Mississippi. Worked as an independent seniors research engineer and research team leader of 5-7 professionals with responsibility for fonnulating and supervising research on dynamic constitutive properties, ground shock, and projectile penetration and for applying results to military systems problems. Work
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included development and use of large finite differences and finite element computer programs. Functioned as Acting Chief during absences of Chief.
- g. 10/77 - 01/80: Supervisory Research Civil Engineer with the US Army Engineer Waterways Experiment Station, Vicksburg, Mississippi.
Worked as the Chief. Earthquake Engineering and Geophysics Division (12 professionals, 5 sub-professionals) engaged in field, laboratory, and analytical Research and Development on liquefaction and stability of earth dams in earthquakes, geophysical methods, and foundation vibrations. Managed Corps of Er gineerings Strong-Motion Instrument Program.
- h. 01/80 - Present: Supervisoy Civil Engineer with the US Army Engineer Waterways Experiement Station, Vicksburg, Mississippi. Works as the Assistant Chief, Geotechnical Laboratory (220 people, $15 M) engaged in laboratory personnel and resource management, review of research plans, and reports to sponsors in the fields of soil mechanics, engineering geology, rock mechanics, earthquake engineering, geophysic:.
vehicle mobility, pavements, theater of operations construction, and expedient surfacing. Consultation in soil dynamics and earthquake engineering to other Government agencies.
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.- ATTACHMENT lb j
- LIST OF PUBLICATIONS i
(1) Jackson, J. G., Jr., and Hadala, P. F., " Dynamic Bearing Capacity , of Soils, Report 3, The Application of Similitude to Small Scale Footing Tests," Technical Report 3-599, Report 3, December 1964, US Army Engineer Waterways Experiment, Station, CE, Vicksburg, Mississippi. (2) Hadala, P. F. , " Dynamic Bearing Capacity of Soils, Report 4. Investi-gation of a Dimensionless Load-Displacement Relation for Footings on Clay " . i Technical Report 3-599, Report 4. June 1965, US Army Engineer Waterways Experi- ! , ment Station, CE, Vicksburg, Mississippi. l l (3) Hadala, P. F., and Jackson, J. G., Jr., "A Model Study of the Small l Boy Footing Behavior," Technical Report 3-793, August 1967, US Army Engineer i ! Waterways Experiment Station, CE, Vicksburg, Mississippi. (4) Hadala, P. F. , "The Ef fect of Placement Method on the Response of Soil Stress Cages," Technical Report 3-803, November 1967, US Army Engineer
- Waterways Experiment Station, CE, Vicksburg, Mississippi. (Also published l 4
in Proceedings of International Symposium on Wave Propagation and Dynamic Properties of Earth Materials, 1967, University of New Mexico Press, Albuquer-que, New Mexico.) (5) .Hadala, P. F., " Site Peculiar Airblast-Induced Ground Motion and ! Soil-Structure Interaction for Tactical Sentinel MSR Sites," April 1968, US j Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (6) Casagrande, D. R. , and Hadala, P. F., " Foundation and Materials Report, Ground Motion and Soil-Structure Interaction for Sentinel Defense i System MSR Site at Chicago, Illinois (U)," April 1968, US Army Engineer l Waterways Experiment Station, CE, Vicksburg, Mississippi, SECRET. (7) Hadala, P. F., " Sidewall Friction Reduction in' Static and Dynamic Small Blast Load Generator Tests," Technical Report S-68-4, August 1968, US j Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. , , I (8) Hadala, P. F., and Perry, E. B., " Preliminary Computations of Air- . blast-Induced Free-Field Ground Motions for Sentinel System MSR Site at Camp f ' Curtis Guild, Boston, Massachusetts (U)," October 1968, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi, SECRET.
. t (9) Hadala, P. F., and Perry, E. B., " Preliminary Computations of '
i , Airblast-Induced Free-Field Ground Motions for Sentinel System MSR Site at Boston, Massachusetts (U)," October 1968, US Army Engineer Waterways Experiment ' Station, CE, Vicksburg, Mississippi, SECRET. (10) Hadala, P. F. , and Taylor, H. M. , Jr., " Preliminary Computation of [ Airblast-Induced Ground Motion for Sentinel System PAR Site at Boston, Massachu- ; setts, (U)," October 1968, US Army Engineer Waterways Experiment Station, CE,
- Vicksburg, Mississippi, SECRET. l I -- (11) Hadala, P. F., and Perry, E. B., " Foundation and Materials Report.
4 Ground Motion and Soil-structure Interaction for Sentinel Defense System MS,R ' { Site at Boston, Massachusetts (U)," December 1968, US Army Engineer Waterways ;
- Experiment Station, CE, Vicksburg, Mississippi, SECRET. i i
l ! l
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t (23) Hadala, P. F., " Lower Bound Estimates to Allowable Dynamic Loads on Mortar Bas'e Plates," Hiscellaneous Paper S-69-14, March 1969, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (24) Day, J. D., and Hadala, P. F. , " Destruction of Earthen Tunnels," Miscellangous Paper I-930, October 1967 US Army Engineer Waterways Experiment l 1 Station, CE, Vicksburg, Mississippi. Also published in Proceedings of Second Counter-Interagency Research and Development Symposium, Houston, Texas, June 1967. i F (25) Taylor, H. M. , Jr. . .and Madala, P. F. , "Ef fect of Grid Size on Cutoff Frequency on'the Numerical Solution of a One-Dimensional Wave Propagation Problem," Technical Report S-71-2, February 1972, US Army Engineer Waterways l Experiment Station, CE, Vicksburg, Mississippi. t j (26) Hadala, P. F., " Evaluation of a Transmitting Boundary for a Two-
- Dimensional Wave Propagation Code," Technical Report S-71-16, December 1971 I
US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (27) Cunny, R. W. , and Hadala, P. F. , " Field Vane Shear Investigation, Willow Point, Louisiana," Miscellaneous Paper 3-667, July 1964, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. i (28) Cunny, R. W., Strohm, W. E., and Hadala, P. F., "Effect of Large-Size Particles on the Shear Strength of a Saturated Gravel," Miscellaneous 4 Paper 3-664, July 1964, US Army Engineer Waterways Experiment Station, CE, ! Vick.ourg, Mississippi.
' (29) Fowler, J. , and Hadala, P. F. , " Vibro-Seismic Survey, High-Stability ,
AEC Structure, Oak Ridge, Tennessee," Miscellaneous Paper S-71-26 October 1971, ' 3 US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (30) Hadala, P. F., "Effect of Constitutive Properties of Earth Media on Outrunning Cround Shock From Large Explosions," Technical P.eport S-73-6, j August 1973 US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (Also, Phd Thesis, University of Illinois at Champaign-Urbana, ' Illinois, May 1973.) i (31) Baladi, G. Y. , and Hadala, P. F., "Cround Motion Calculations for the Malmstrom PAR Site (U)," Miscellaneous Paper S-74-2, January 1974, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi, l SECRET. , (32) Hadala, P. F., et al, "An Assessment of the State of the Art for l Vulnerability and Hardness Analysis of Ballistic Missile Defense Facilities. ' Chapter 2, Blast and Shock (U)," Technical Report N-72-12, September 1973, US Army Engineer WaterwaysoExperiment Station, CE, Vicksburg, Mississippi, , 4 SECRET. , (33) Hadala, P. F., " Final Report on the SAFEGUARD Ground Motion Study
- (U)," Technical Report S-75-4. April 1975 US Army Engineer Waterways Experi-i ment Station, CE, Vicksburg, Mississippi. -
l ' 3
-, - . , . . . . . - , . , , , . - - - . , - _,.--,,--..__-,,..m,n,,,,.,n-, . , , , , , _ . , _ _ , _ , . , , , . _ , . . , , . , , , . _ -,..cm,-,,,-y,,_.m,,,.,.,-,,., ,w,.,,,,.,_w-,,,r,,,nem.
,.-.~,- _ . . ._ _._ _ .. . . _ _
., , \ ,
(34) Hadala, P. F., " Comparison of One-Dimensional Ground Shock Calcu-lations for Site Defense Criteria Sites 1-4 with Results from Other BHD-Related i Studies (U)," August 1973, US Army Engineer Waterways Experiment Station, CE, y j Vi'cksburg, Mississippi SECRET. (35) Taylor, H. M. , Jr. , and Madala, P. F. , " Site Defense Ground Motion' ! Criteria S'tudies, One-Dimensional Work Propagation Calculations for Site 1 (U)," ! March 1973, US Army Engineer Waterways Experiment Station, CE, Vicksburg. ' ' Mississippi, SECRET. i (36) Taylor, H. M., Jr., and Hadala, P. F., " Site Defense Ground Motion j Criteria Studies One-Dimensional Work Propagation Calculations for Site 2 (U),"
- March 1973, US Army Engineer Waterways Experiment Station, CE, Vicksburg,
- Mississippi, SECRET.
! (37) Taylor, H. M. , Jr. , and Hadata , P. F. , " Site Defense Ground Motion 4 Criteria Studies, One-Dimensional Work Propagation Calculations for Site 3 (U)," April 1973, US Army Engineer Waterways Experiment. Station, CE, Vicksburg, i
! Mississippi, SECRET.
(38) Taylor, H. M. , Jr. , and Hadala, P. F. , " Site Defense Ground Motion Criteria Studies, One-Dimensional Work Propagation Calculations for Site 4 (U)," i May 1973, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Missis- ! sippi, SECRET. ' '
- (39) Rohani, B. , and Hadala, P. F. , " Penetration Analysis of a 12,000-Pound GP Bomb into a Concrete Target," January 1976, US Army Engineer Waterways Exper-j iment Station, CE, Vicksburg, Mississippi.
(40) Hadala, P. F., " Laboratory Support Studies for and Pretest Predictions j for Rock Penetration Tests in TTR Welded Tuff," October 1975, presented at the l DNA Penetration Review Meeting, 30 October.1975 at Harry Dismond Laboratories, j US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. (41) Hadala, P. F., " Phase I Study of an Earth Penetrator to Carry a Shallow Burst Munition (U) " September 1976, Final Summary Report, US Army , Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi, SECRET 4 RESTRICTED DATA, CNWDI. j j (42) Hadala, P. F., Letter to Director, Defense Nuclear Agency, dated l j 27 July 1976, subject: Hard Copy of SAGE Presentation on SBM Penetrator Design. j } US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi, j SECRET RESTRICTED DATA, CNWDI. ! (43) Butters, S. W.; Swolfs, H. S.; Johnson, J. N.; Butler, D. K.; and Hadala, P. F., " Field Laboratory and Modeling Studies on Mount Helen Welded ! Tuff for Earth Penetrator Test Evaluations," Technical Report 75-9, August 1976,' i Terra Tek, Salt Lake City, UT (Appendices A and B prepared by WES). ! (44) Hadala, P. F., " Earth Penetration, FY 74 Results and FY 75 Activities " l February 1975, presentation at the DNA Strategic Structures Review Meeting, 19-20 ( February 1975 at Stanford Research . Institute, Menlo Park, California (300 pre-
- prints handed out), US Army Engineer Waterways Experiment Station, CE, Vicksburg, l Mississippi.
4
i (45) lladala, P. F. , " Evaluation of Empirical and Analytical Procedures i Used for Predicting the Rigid Body Motion of an Ear h Penetrator," Miscellaneous Paper S-75-15, June 1975, US Army Engineer Waterways Experiment Station, CE, . Vicksburg, Mississippi. j (46) Peterson, R. W. , and Hadala, P. F. , " Shear Strength Properties for ! the Antelope Lake Target. Tonopah Test Range, Nevada " Miscellaneous Paper S-76-7, I May 1976, US Army Engineer Waterways Experiment Station, CE, Vicksburg, Mississippi. i (47) Hadala, P. F., " Earth Penetrating Nuclear Weapons," a presentation I j at the DNA SPSS Biennial Review Conference, February 1977, published by Defense Nuclear Agency, Washington, DC. 3 (48) The Data Analysis Working Group "A Review of High Explosive Testing to Investigate Ground Motions Pertinent to the MX Multiple Aimpoint Systes," , 13 May 1977, DAEG TR-1, R&D Associates, Marina del Rey, California.
; (49) Hadala, P. F., Spangler, D. R., and Stong, T. D., "The Technology i of Earth Penetrating Weapons (U)," pp 37-64, Journal of Defense Research, i Vol 10 No. 1. Spring 1978, Battelle of Columbus Labs for DARPA, SECRET-FRD l (Paper is C-FRD).
1 4 (50) Hadala, P. F. , Rohani, B. , Spangler, D. R. , and Stong, T. D. , " Earth Penetrating Weapons Technology Program (U)," Technical Report SL-79-5, August i 1979, CONFIDENTIAL,140 pp. (51) Hadala, P. F. , and Cooper, S. S. , "Cround Motions Due to Explosive ; Structural Demolition Over a Point Bar Depcsit," September 1978, unpublished
- report to New Orleans District, CE, 30 pp.
(52) Curro, J. R. , Hadala, P. F. , and Landers, G. B. , " Seismic Attenuation , j Tests at the Portsmouth Ohio, Caseous Diffusion Add-On Site," MP-S-78-4, April 1978. ' (53) Marcuson, W. F. , III, Hadala, P. F. , and Franklin, A.* C. , " Current i Methods of Dynamic Analysis for Seismic Stability of Earth Dams," In: S..L. Koh (Editor), Mechanics of Landslides and Slope Stability, Engineering Ceology, j No. 16, pp 19-28. s .
- (54) Earthquake Engineering and Geophysics Division and Engineering Geology l j and Rock Mechanics Division staff members, " Engineering and Design - Geophysical '
- Exploration," EM 1110-1-1802, 28 February 1979.
] (55) Franklin, A. G., Marcuson, W. F., III, and Hadala, P. F., " Soil
- Dynamics - State of the Art - 1980," Proceedings of the Sixth Southeast Asian t
Conference on Soil Engineering Vol 2, pp 19-23, May 1980. 3 (56) US Army Corps of Engineers, " Theater of Operations, Construction in
. the Desert, A Handbook of Lessons Learned in the Middle East," January 1981, l Joint WES-CERL report by Kao, Anthony and Hadala, Paul F.
- (57) Nadala, Paul F., and Ballard, Robert'F., Jr., Discussion of ASCE
- Geotechnical Journal paper by Allen, N. Foster, Richard, F. E., Jr. , and Woods, R. D., " Fluid Wave Propagation in Saturated and Nearly Saturated Sands, pp 698-
) 701, vol 107, No. GT 5, May 1981. - 4 5 5 i 5 i
[ ATTACHMENT 2 i i i r REFERENCES
- 1. . " Site Specific Response Spectra, Midland Plant - Units 1 and 2 Part II, Response Spectra Applicable for the Top of Fill Material at the Plant Site," April 1981. Weston Geophysical Ccoperation, Westboro, Massachusetts.
- 2. Hadala, P. F., Memorandum for Record, " Visit to Midland Nuclear Power Plant on 27-28 February 1980, A Review of the Midland Plant Units 1 and 2 FSAR (Including Revisions 1-27) 30 May 1980," US Amy Engineering Waterways Experiment Station, Vicksburg, Mississippi.
- 3. Hadala, P. F. Letter dated 16 June 1981 to District Engineer, USAE District, Detroit (ATTN: Mr. Neil Gehring)
Subject:
Review of Amendment 85 - Midland Nuclear Power Plant, USAE Waterways Experiment Station, Vicksburg, Mississippi.
- 4. Hadala, P. F., Memorandum for Record, "Effect of Plant Fill on Seismic Ground Motion Environment at the Midland Michigan Nuclear Powr Plant," 3 August 1981 USAE Waterways Experiment Station Vicksburg, Mississippi.
- 5. Hadala, P. F., Letter dated 24 August 1981 to District Engineer, USAE District Detroit ( ATTN: Mr. Neil Gehring)
Subject:
Deismic Safety of Baffle Dike, Perimeter Dike and Emergency Cooling Pond Slopes at Midland Nuclear Power Plant.
- 6. . " Site Specific Response Spectra, Midland Plant -
Units 1 and 2, Part I, Response Spectra - Safe Shutdown Earthquake, Original Ground Surface," February 1981 (also Addendum to Part I dated June 1981), Weston Geophysical Corporation, Westboro, Massachusetts.
- 7. Schnabel, P. B., Lysmer, J., and Seed, H. B., " SHAKE, A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites,"
EERC Report 72-12, December 1972 University of California at Berkeley, Berkeley, California.
- 8. Lysmer, J., " Analytical Procedures in Soil Dynamics," pp 1267-1317, Vol III, Proceedings of the ASCE Geotechnical Engineering Division Speciality Conference on Earthquake Engineering and Soil Dynamics, June 1978. Pasadena, California.
4 e i 4
- - _ -- - - -- .-- - - - , . . , - - . - ~ - _ . -
. k' NTAC i.Ek'T $
. DEPARTMENT OF THE ARMY WATERWAYS EXPERIMENT srATION CORPS OF ENGINEERS P. o. sox est vicxsmuRG. Mississippi asteo
.. .a nv e nn a '* WESCA 3 August 1981 I
MEMORANDUM FOR RECORD
SUBJECT:
Effect of Plant Fill on Seismic Ground Motion Environment at the Midland Michigan Nuclear Power Plant m>
- 1. INTRODUCTION. Under IA0 Number CE-IA-80-047 from the Detroit District CE, (who are in turn supporting the Site Analysis Branch of the Nuclear Regulatory Co= mission), the undersigned has participated in a continuing review of the plant fill at Midland Nuclear Power Plant. My participation has been ILaited to seismic considerations. References 1 and 2 addressed a number of questions but were primarily concerned with evaluation of liquefaction potential. This memorandum, which consists of two parts, addresses the effect of the plant fill on the earthquake induced ground motion environment. Part I is a review of Appendix B of Reference 3 requested by NRC. Part II consists of a series of SHAKE Code (Reference 4) one-dimensional wave propagation calculations performed by WES to study the effects of changing some of the parameters used in similar calculations in Reference 3.
- 2. PART I - REVIEU OF WESTON'S REPORT. In the main body of Reference 3, the only portion this writer is competent to evaluate is Section 2.2. The P- and S-wave velocity profile given in Figure 1 of Reference 3 and plotted
.in Figure 4 of that reference are considered reasonable. Inclosure 6 of Reference 1 shows the S-wave velocity (V) data for the plant fill to be consis-tent with the 440-1060 ft/sec range adopted by Weston'. A closer look at Inci 6 of Reference 1 indicates the vast majority of the data lies between 575 and 900 ft/see and that there is a slight trend of increase in V with depth. In the upper part of the fill, 700 ft/see is an upper bound ,to nearly all the data. The P- and S-wave velocity profiles used for that portion of the profile below original ground surface come from the Weston preconstruction geophysical tests (see FSAR section 2.5.4.7.2) and the effect of the addition of the fill should be only a very slight increase in V,. The amount of the increase is judged to be so small that it could not be resolved because it is below the sensitivity of the seismic testmethods (see Reference 7). This '-
reviewer is satisfied that the seismic profile used in the selection of records for use in the development of a site specific response spectra for the plant fill is physically reasonable for the site and consistent with the available data.
- 3. Appendix B of Reference 3 contains the results of a series of SHAKE Code one-dimensional wave propagation analyses performed to study "possible local amplification effects on earthquake ground motion at the Midland Plant Site."
/
--, -.. -- -. c . . - - , - - , - , , -
1 UESCA 3 August 1981
SUBJECT:
Effect of Plant Fill on Seismic Ground Motion Environment at the . Midland Michigan Nuclear Power Plant In Section 2.0 of Appendix B and Figures B-1, B-3, B-4, and B-% soil profiles and properties used in the analyses are presented. This reviewer has no disagreement with the range of layered systems investigated, the densities, damping, and shear-wave velocities used. However, by specifying modulus factors, the authors of Appendix B effectively negated their choice of shear- I wave velocities and substituted much lower values for the plant fill as shown in Figure 1. This point was first pointed out to the authors and the applicant's representatives at a meeting at NRC in Bethesda, MD on 30 June 1981.
- 4. SHAKE, when given modulus factors, uses them in preference to V to compute j the initial shear modulus G o. Since the shear modulus-shear strain
- curve for
- each layer which is used in the code is normalizedto G,, shear moduli for all strain levels are controlled by its specification. Figure 1 also shows that the upper 50 f t of the till was represented as being five times stiffer than it actually was. This reviewer verified that the solid curve in Figure 1 was what was actually used for Case A by performing a duplicate of one of the authors' calculations and reproducing the results in Figure B-8 of Reference 3.
- 5. The normalized shear modulus and damping versus strain curves used in the Appendix B calculations were not given in the report. However, they were supplied to the reviewer by letter (Reference 6) and were determined to be those developed in Reference 8 The curves are shown in Figures 2 and 3.
The use of these curves is considered an acceptable state-of-the-art practice.
- 6. In Reference 6, the authors provided the reviewers with a set of new results from SHAKE Code calculations for the Case A profile in which the modulus factors were adjusted to produce initial shear moduli consistent with the shear-wave velocities. This revised set of properties is referred'to as Case A, variation 1.
Figure 4 shows the effect of changing modulus factors on the amplification factor versus frequency curve generated by the SHAKE code calculation for Case A soil profile. The El Centro 5/18/1940 acceleration record for the Imperial Valley station scaled to 0.12 g was used as an outcrop of the Saginow bedrock (which underlies the profile at a depth of 371 ft) in this' calculation. As shown in Figure 4, there are substantial percentage increases in amplification factors at frequencies between 1/2 and 10 hz when more realistic modulus factors are used.
- 7. In References 3 and 6, a substantial number of compater code parameter studies were-performed. A wide variation in soil profiles was considered along with four different earthquake records. In all cases the records were scaled to 0.12 g and input as outcrop motions for an outcrop of the Saginaw Formation. In the - -
case of the El Centro record, which was recorded on deep alluvium, a question i arises as to the appropriateness of using this record as a bedrock outcrop. The other three records used are all from the Lytle Creek Earthquake (Richter magnitude 5.4). Two of the three records have been classified as being recorded at " intermediate" sites and one, the Cedar Springs, Allen Ranch record is from a rock site. The use of these records as rock outcrop motions is more realistic j than'is the use of the El Centro record. The effect of the choice of the layer
- selected for the outcrop will be examined in Part 11.
li i
- - . .- -~~ .,, -,.,,,---.__,---,n ,_ -. . - , - - . . , , . . - , - . . , , . , , --
WESCA 3 August 1981 SU; JECT: Effect of Plant Fill on Seismic Ground Motion Environment at the Midland Michigan Nuclear Power Plant
- 8. Also, the El Centro record, which was used in most of the Reference 3 SHAKE calculations is from the ocar field of an earthquake of magnitude 6.7 and is rich in low frequencies. The site. specific response spectra developed in the report was' based on a 5.3 m earthquake adjacent to the Midland site. Why the record ,
b for a substantially larger event' was used as the baseline or pivot point of .. the parameter study is unknown. The other three records are from an earthquake of the appropriate magnitude. .
- 9. In the final analysis, all of the work in Appendix B of Reference 3 is summarized in a plot of the ratio of response spectra for the top fill to that for the original ground surface. See Figure 5, which is a copy of Figure 9
, of Reference 3. The fact that the-analytically developed curve (a) lies belov the curve for the comparable ratio of empirically developed site specific response spectra (SSRS) and, (b) has a similar shape been used as an argument that the empirically based SSRS are physically reasonable and are more conservative than
- the top of fill spectrum that would have been obtained via a SHAKE code calcula-tion. The additional ratio of r'sponse. e spectra calculations furnished in Refer-ence 6 are superimposed on Figure 5 and'show that the effect of revising the site properties was to increase the calculated ratios of response spectra (RRS) in the 1/2 to 10He range. However, the analytically developed RRS still (with minor exception) lie below the RRS for the empirically developed site specific l response spectra.
- 10. Figure 6 shows the 84th peretntile, 5 percent damped site specific response spectra developed via empirical methods in References 3 and 5 for the top of plant fill and the original ground surface, respectively. Figure 6 indicates i that the empirical approach produces substantial amplification of response at periods greater than 0.25 see (f requencies less than 4 ,hz), which is supported by the additional calculations reported in Referece 6 and displayed in Figure 5.
- 11. PART II - SHAKE CODE PARAMETER STUDY. One of the objectives of the parameter study wa. to determine the dif ferences between the ground motion environment at the top of plant fill and the original ground surface as calcu-i lated under the assumption of one-dimensional vertical shear wave propogation and all of the other assumptions implicit in the use of the SHAKE Computer code (see Reference 4 ). Another was to study effects of variations in the soil properties assumed for the plant fill. A third objective was to study
- the ef fect of varying the input accelerogram on ground motion. The second
- and third objectives have already been addressed in Reference 3 and the work l - reported herein merely extends the range of'that parameter study. The fourth objective was to examine the effect of choico'of outcropping layer. In Reference 3 all. calculations were made assuming that the'accelerogram represented the motion at a bedrock (Saginaw Formation) outcrop near the site. An equally (or perhaps more) reasonable assumption for some of the accelerograms is that they represent original ground surface (top of till) motions. Table 1 is a list of the thirteen j SHAKE code calculations performed that shows the variation of parameters performed.
j Table 2 describes the earthquake accelerograms used as input. l l i 3 l
-. . ~ . -- , . ~ .. _ . - ..- . - ~
?_ . .. . .. _ .. _ . . . . ... ._ -~ . -. .. . . . .-.__~_i WESCA 3 August 1981
SUBJECT:
Effect of Plant Fill on Seismic Ground Motion Environment at the ' Midland Michigan Nuclear Power Plant 4'
- 12. The reference point for the parameter study reported herein is the Case A soil profile given in Reference 3 (see also Table 3), the El Centro earthquake accelerogram described in Table 2, which was scaled to 0.12 3 in this study as well as in Reference 3. Each SHAKE run used an earth pressure at rest coefficient of 0.45 and the following options (see Page 15 of Reference 4): 1,2,3,4,5, '
4 8, 9, and 15. The first calculation performed (Run 1) was performed to duplicate one of the calculations in Reference 3. It used the Case A soil profile (Table 3 and dotted line in Figure 1) even though it was considered much too sof t in the fill layers and,too stiff in the upper part of the glacial till. It was used because it was the one used in Reference 3. The modulus and damping versus strain relations used in Option 8 are those given in the sample data set on Page 40 of Reference 4 and are essentially those shown in Figures 2 and 3. The accelerogram (El Centro, S00E, 5/18/40) scaled to 0.12 g was used only because of its choice by the authors of Reference 3: As shown in Table 2, it is a long record and was recorded in the near field region of an earthquake substantially larger than the safe shutdown earthquake for the Midland site. The outcrop location in the rock of:the Saginaw Formation (which underlies the Case A profile at a depth of 371 f t) was chosen again because it was used in Reference 3. The actual accelerogram was recorded at the surface of a deep soft alluvial deposit ' and its frequency content is probably not appropriate for a rock outcrop. This calculation (the solid line curve in Figure 7) duplicated the amplification curve shown in Figure B-8 of Reference 3.
- 13. Figure 7 shows the effect of some variation in site profile on the ampli-
! fication factor curve for the motion at the top of fill with respect to outcrop motion. Figure 8 shows the effects of the same variations in site profile on 5 percent damped shock spectra at the ground surface. Profile F, whose ? roper-ties are given in Table 4 and whose low strain level shear modulus vs depth relationship is given by the solid line in Figure 1 represents this writer's judgment of the lower limit of fill stiffness and the best estimate of the stiffness of the natural ground. Profile C, whose properties are given in ' Table 5, represents this writer's judgment, on the upper limit of the fill stif f-
~ '
ness. Below the plant fill it is the same as F. Profile H (see Table 6) is the same site profile as F and G, but with the fill removed. The top of the profile in this case is the original ground surface. Case I is the same as A but with the fill removed. l
- 14. Figures 7 and 8 show that changing profiles make substantial differences in the results of the calculations. The effects are best shown on the shock
. spectra in Figure 8. When profiles A and I are compared (wrong site properties-with and without fill) it is seen that the effect is also deamplification at frequencies above 1 Hz but to a lesser degrey, 11gure 7 shows that in the lower
, frequency range (below 1 Hz) the effect c7 F o e esence of the fill (either Case A or F) is substantial amplification over C4;t I
- 15. Figure 7 also shows for Profile F what happens when the outcrop location is changed to the top of till (i.e. the original ground surf ace), a location
considered more realistic for a ground motion record recorded on the surface ef' t
- . 4 L
- ~- - . , _ . _ . . . . _ . . . . . _ = _ ~ . . - - -
. . ~ . . . ..-wv. - . . . . . . ~-- - - . .
..o. . - . . . ~ . . ... + . + .
. WESGA 3 August 1981
SUBJECT:
Effect of Plant Fill on Seismic Ground Motion Environment at the Midland Michigan Nuclear Power Plant deep alluvium. The nature of the amplification curve changes radically, the l amplification factor remains close to unity at low frequencies and drops toward zero at the higher frequencies. This is in accord with common sense, as the top of fill is only 30 ft above the outcrop's elevation in the profile and radically different low frequency motions at the two points are a physical p impossibility. f
- 16. While this writer considers the spectra in Figure 8 for the A and I pro-files inappropriate because of the bedrock outcrop location used in the calcu-lation and the size of earthquake in which the record was obtained, the spectra in Figure 8 can be compared with the 84th percentile empirically developed spectra shown in Figure 6 by overlaying the transparency of Figure 6 given at the end of this report on Figure 8. In fact, the same comparison can be made with any of the spectra which follow.
- 17. Figure 9 shows the spectra for the top of the F, G, and H profiles calcu-lated if the El Centro record is used as a till outcrop which is physically more realistic than using it:as a rock outcrop record. In the 1 to 5 Hz frequency range, the Profile G spectra is almost the same as that for the original ground surface while the spectra for Profile F shows modest amplification. While the El Centro record is from an earthquake of larger magnitude than the safe shut-down earthquake, it is still of interest to compare Figure 9 with Figure 6.
All three spectra exceed the empirical one for the top of fill at frequencies below 2 Hz.
- 18. The NRC staff requested a series of SHAKE calculations be performed with an accelerogram from the Fogoria-Cornino station and the 9/11/76 earthquake at Friuli, Italy. This record was among those used in the development of the empirical site specific response spectra. The card deck for this record was
- furnished to WES by NRC. Since the. record was recorded in the surface and the site profile bears some general similarity in layering and stiffness to the
. Midland site, the original ground surface is the appropriate outcrop layer and was used as such in Runs 4, 5, and 11,(Table 1). Results from these three calculations are shown in Figures 10 and 11. Because of the physical proximity of the outcropping layer to the top of the fill, only limited amplification can occur. As expected, Profile F, the softer profile shows more amplification than Profile G. The peaks as shown in Figure 10 are at frequencies which are bounded from above by those obtained from the formula: f=V, ., 4ii Where H = layer thickness = 31 ft f = frequency V = average small strain level shear-wave velocity of fill (670 ft/see for Profile F and 770 ft/sec for Profile G) Since Profile F is softer and the response of the fill is, on a relative scale, much further into the nonlinear regime, the greater overprediction by'the' formula for Case F appears reasonable. I 1
e
., w WE5GA 3 August 1981 SL"6 JECT: Effect of Plant Fill on Seismic Ground Motion Environment at the Midland Michigan Nuclear Power Plant .
- 19. Figure 11 shows the spectra for the same 3 cases. The top of fill spectra for Profile G is essentially the same as that for the original ground surface (Profile H) while that for. Profile F shows some amplification in the 2-5 He frequency range. Since the acceleration record, soil profiles and outcrop are all physically reasonable with respect to the, assumptions made in the develop-ment of the empirical site specific response spectra, a comparison with trans- I parency of Figure 6 is appropriate. Such a, comparison shows that except for two very slight excursions, the spectra in Figure 11 are all below the 84th percentile empirical spectra for,the top of fill.
- 20. The third acceleration record the NRC staff requested be used in SHAKE calculations was the Temblor record from the Parkfield earthquake (see Table 2).
While the record was obtained in an earthquake of a magnitude close to that of the safe shutdown earthquake, this record was recorded at a rock site in fairly close proximity to the fault rupture. At the 30 June meeting at NRC, there was much discussion over the appropriateness (and lack thereof) of the Parkfield records for the Midland site. There is additional discussion on the subject in Reference 9. This issue falls within the expertise of the seismologist and the writer is not a seismologist. However, what is clear is that if the Temblor record is to be used in SRAKE calculations, it should be used as a , rock outcrop record.
- 21. Figures 12 and 13 show the amplification factors and shock spectra computed using the Parkfield-Temblor record. The figure shows substantial amplification due to the fill at frequencies up to 4 Hz and the spectra showed in Figure 12 exceed those in Figure 6 substantially in the range below 5 He. It should be expected that,if the seismologists decide the Parkfield-Temblor record and others like it should make up a substantial part of the data set for an empir-ical analysis then the empirically developed spectra for the plant fill would exceed that in Reference 3. Reference 9 showed that such was also the case for the spectra for the original ground surface. .
- 22. Figure 14 shows the largest effect of any single variation made in this study. Changing from a rock outcrop (which is appropriate for the Temblor record) to a till outcrop substantially decreased the shock spectra amplitudes in the entire region of interest.
- 23. Figures 15 and 16 show ratios of response spectra calculated by the same equation used in Reference 3. These ratios are the ratio of the spectra for the top of fill to that for the outcrop motion. With only one exception, thosa cases where the writer judged the outcrop layer was reasonably matched to the actual conditions at the accelerograph secondary station are shown. The excep-tion is the case of Runs 1 and 13 on Figure 16. This is shown becruse it was one of the cases examined in Reference 3. While the new calculation is close to the results in Figure B-14 of Reference 3, it is not an exact match. This is probably because the calculations which resulted in Figures 15 and 16 were performed with an amplification factor data set which had less resolution in the frequency domain (af = 0.1 He) and no smoothing of the input. The key point is that except for the cases involving the Farkfield-Temblor record, all lie
. 6
)
' \
. . l
., l UESCA 3 August 1981 l
SUBJECT:
Effect of Plant Fill on Seismic Ground Motion Environment at the l Midland Michigan Nuclear Power Plant well below the curve for the ratios of site specific response spectra given in Figure 9 of Reference 3 (also Figure 5 of this memorandum). It is also of interest to note that the ratios are generally closer to unity over the entire frequency range for the till outcrop cases than for those with the Saginaw outcrop. 24
SUMMARY
. The analytical parameter study has shown that the effect of the addition of the fill on the ground motion environment can be different depending on the stiffness of the fill, the acceleration record chosen, or which layer is chosen for the outcrop. The largest variations observed in the top of fill ground shock environment occurred due to the change from Profiles A to F (or I to H). This represents a major change, not just in the fill but in the top 50 ft of original ground. The combined effects of making the site properties more realistic was to raise the shock environment (Figure 8). The second major variation was the result of changing outcrops. In general, changing from Saginaw to till as the outcropping layer decreased the shock environment substantially (compare Figures 8 and 9; see also Figure 14). The effect of stiffening the fill only was to decrease the shock environment at the top of the fill (Figure 11) by a modest amount. I
- 25. The 5 percent damped shock spectra calculated for the top of the fill for all cases involving the Fogaria record were below the empirically developed 84th percentile site specific response spectra (SSRS). The use of the El Centro record as a till outcrop and realistic site properties produced spectra which exceeded the SSRS to a modest degree in a limited frequency range. The spectra calculated using the Parkfield-Temblor record as a rock outcrop significantly exceeded the SSRS. Whether the El Centro and Parkfield records are reasonable ones to use in the first place has been questioned. This writer cannot answer that question; it should be posed to the seismologists.
- 26. All ratios of response spectra GUU) calculated except those for the Parkfield-Temblor record fell within the envelope of the RRS from the Reference 3 study.
b -/ 3 Incl: PAULF.HkDALA
- 1. Tables Assistant Chief
- 2. Figures Geotechnical Laboratory
- 3. References 7
sS I, i Table 1 ~. '
,t' 4 SHAKE COI)E PARAMETER STtIDY
- s s Run Number Soil Profile Record Outcrop Assumed Remarks
- g. 1 A El Centro Saginaw Dupilcate of Run From Reference 3, 2 .76/.93 F El Centro Till Hore Realistic Outcrop, Vary Soil Profile 3 C El Centro Till More Realistic Outcrop, Vary Soil Profile
.4 .76/.85 F Fogaria
- Till Vary Soil Profile 5 .76/.86 C Fogaria Till ,
I 6 F Parkfield-Temblor Saginaw Vary Soil Profile 7 C Parkfield-Temblor Saginaw 8 F El Centro Saginaw "' Compare with 1 (Soil Profile), Compare with 2 (Outcrop) 9 .76/.91 F Parkfield-Temblor Till Compare with 6 (Outcrop) 4 10 H El Centro Till Compare with 2 and 3 (Natural Cround vs j Top of Fill)' , i 11 H Fogaria Till Compare with 4 and 5 (Natural Ground vs Top of Fill) 12 H Parkfield-Temblor Saginaw Compare with 6 and 7 (Natural Ground vs ' Top of Fill) 13 I El Centro Saginaw Compare with 1 (Natural Ground vs Top I of Fill) , t i i i e 4
~
s Table 2 2 I I CROUND HOTION RECORDS USED j a Record Time " max i Source Of Earthquake Length Interval After Scaling H fore caling sec. g. Hecord
- Hag. Date Station Component g sec.
0.12 CIT Vol II El Centro Imperial S00E 0.35 53.8 0.02 6.7R5/18/40 Valley #117 Card Deck Friule After- Fogaria- EW 0.11 15.0 0.005 0.12 Provided by shock 5.0 b C '"I"" 9/11/76 CIT Vol 11 Parkfield Temblor 525E 0.35 30.4 0.25 0.12 l 5.6 6/27/66 , i
- All records were run without any further attempt at baseline connection.
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