ML19310A852
| ML19310A852 | |
| Person / Time | |
|---|---|
| Site: | Yellow Creek  |
| Issue date: | 06/27/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19310A853 | List: |
| References | |
| NUREG-0347, NUREG-0347-S01, NUREG-347, NUREG-347-S1, NUDOCS 8006300595 | |
| Download: ML19310A852 (56) | |
Text
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Evaluation Itcpori
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'rciated to construction of gericgNg7,r Yellow Creek Nuclear Plant Units 1 & 2 Docket Nos. S N 0-
!Tcnnossee Valley Authority June 1s7s
- Supplement No.1 l
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l Available from i
National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $6.00 ; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.
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NUREG-0347 Supplement No. 1 June 27, 1978 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF TENNESSEE VALLEY AUTHORITY YELLOW CREEK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. STN 53-566 AND STN 50-567
TABLE OF CONTENTS P, age
1.0 INTRODUCTION
AND GENERAL DISCUSSION.......
1-1 1-1 1.1 Introduction.........
1.2 General Plant Description......................
1-2 1.2.3 Other Major Structures.
1-2 1.9 Outstanding Issues.......
1-2 1.11 Exceptions to CESSAR Requirements..
1-5 2-1 2.0 SITE CHARACTERISTICS.......
2-1 2.1 Geography and Demography........
2.2 Nearby Industrial Transportation and Military Facilities...........
2-1 2.4 Hydrologic Engineering..................................
2-2 2.5 Geology and Seismology......................
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3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS.,
3-1 3.4 Water Level (Flood) Design Criteria....................
3-1 3.5 Missile Protection.................
3-1 3.5.1 Missile Selection and Description.......
3-1 3.6 Criteria.for Protection Against Dynamic and Environmental Effects Associated with the Postulated Rupture of Piping............
33 3.7 Seismic Design..................
3-5 3.7.2 Seismic System and Subsystem Analysis...........................
3-5 3.8 Design of Seismic Category I Structures.............................
3-5 4.0 REACTOR...........
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i TABLE OF CONTENTS (Continued)
Pag 5.0 REACTOR COOLANT SYSTEM.....
5-1 5.2 Integrity of the Reactor Coolant Pressure Boundary................
5-1 5.2.7 Leakage Detection System..................
5-1 6.0 ENGINEERED SAFETY FEATURES..
6-1 6.2 Containment Systems..............
6-1 6.2.1 Containment Functional Design.
6-1 6.2.3 Secondary Containment Functional Design............
6-4 6.2.4 Containme..t Isolation System..........
6-4 6.2.6 Containment Leak Testing Program..........
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6.3 Emergency Core Cooling System.........
6-5 6.3.1 Passive Failure Protection..................................
6-5 6.5 Engineered Safety Feature Atmospheric Cleanup Systems...
6-5 6.5.1 System Description and Evaluation of Auxiliary Area Emergency Gas Treatment System.
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- 7. 0 INSTRUMENTATION AND CONT ROL SYSTEMS......................................
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7.3 Engineered Safety Features Actuation System........................
7-1 7.4 Systems Required for Safe Shutdown........
7-1 7.6 All Other Instrumentation and Requirements for Safety...............
7-2 7.6.1-Environmental Qualification....................................
7-2 7.7 Control Systems Not Required for Safety....................
7-2 9.0 AUMILIARY SYSTEMS.............
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l 9.2 Water Systems................................
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TABLE OF CONTENTS (Continued)
PAGE 9.2.2 Component Cooling System........
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9-1 9.2.3 Ultimate Heat Sink....................
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9.5 Other Auxiliary Systems.........
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9.5.1 Fire Protection System........
9-1 15-1 15.0 ACCIDENT ANALYSES..
15.2 Postulated Accidents.
15-1 15.2.2 Radiological Consequences of Containment Purge During a 4
15-1 Postulated Loss-of-Coolant Accident..
1, 18.0 REVIEW BY THE ADVISORY COPMITTEE ON REACTOR SAFEGUARDS..
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20-1 20.0 FINANCIAL QUALIFICATIONS.
20.1 Introduction..........
20-1 20.2 Construction Cost Estimates.............
20-1 20.3 Sources of Funds..........................................
20-2 20.4 Conclusion............
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21.0 CONCLUSION
S.......
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APPENDICES PAGE APPENDIX A REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.
A-1 APPENDIX B CONTINUATION OF THE CHRONOLOGY OF REGULATORY REVIEW.
B-1 APPENDIX C ERRATA TO THE SAFETY EVALUATION REPORT.
C-1 APPENDIX D ADDITION TO THE BIBLIOGRAPHY.
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LIST OF TABLES PAGE TABLE 3.1 REVISED TORNADO MISSILE SPECTRUM - REGION I.
3-2 TABLE 3.2 CONCRETE WALL AND ROOF THICKNESS REQUIREMENTS TO RESIST THE EFFECTS OF TORNADO MISSILE IMPACT - REGION I..
3-4 TABLE 20.1 PRO FORMA STATEMENT OF SOURCES AND FUNDS FOR SYSTEM WIDE CONSTRUCTION EXPENDITURES DURING CONSTRUCTION OF YELLOW CREEK NUCLEAR PLANT, UNITS 1 AND 2.
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1.0 INTRODUCTION
AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Commission's Safety Evaluation Report in the matter of the Tennessee Valley Authority's appifcation to construct and operate the proposed Yellow Creek Nuclear Plant, finits 1 and 2, was issued in December 1977.
In that report, we concluded that upon favorable resolution of the outstanding matters set forth in Section 1.9 of the report, we would be able to reach the con-clusions required in accordance with the provisions of 10 CFR Part 50.35(a).
Section 1.9 identified 19 outstanding items requiring additional staff evaluation.
Since the Safety Evaluation Report was issued:
(1) The Advisory Committee on Reactor Safeguards completed its review of the Yellow Creek application on January 5,1978. The Committee's report is in-cluded in this supplement as Appendix A.
Our response to the Committees report is presented in Section 18.0 of this supplement.
(2) We identified two additional review matters. These are described as item numbers 20 and 21 in Section 1.9 of this supplement.
(3) The applicant has submitted Amendments number 10, 11, and 12 to the Yellow Creek Preliminary Safety Analysis Report, responding to the items identified in Section 1.9 of the Safety Evaluation Report and this supplement. These Amendments also describe recent station design modifications as described in l
Section 1.2 of this supplement.
(4) The applicant submitted information concerning its financial qualifications to construct and operate the Yellow Creek facility. We have completed our review of the information and our summary evaluation is included in Section 20.0 of this supplement.
(5) The applicant has submitted new information concerning a potential industrial hazard to the Yellow Creek nuclear plant. The issue was the location of an oil storage facility approximately 1.8 miles north and west of the proposed plant site. We evaluated the applicant's submittals and our summary evalua-tion and conclusion is contained in Section 2.2 of this supplement.
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The purpose of this supplement is to report our evaluation of the additional informa-tion submitted by the applicant since our issuance of the Safety Evaluation Report.
Our conclusions regarding the individual outstanding issues are found in the appro-priate sections in this supplement, and as stated in Section 21.0 of this supplement we are able to reaffirm our ccnclusions stated in Section 21.0 of the Safety Evaluation Report.
Appendix 8 to this supplement is a continuation of the chronology of significant events of the licensing review of the Yellow Creek application. Appendix C is a ifsting of errata to the Safety Evaluation Report. Appendix D incorporates addi-tions to the bibliography appended to the Safety Evaluation Report.
Each of the following sections in this supplement is numbered the same as the corresponding section of the Safety Evaluation Report. This supplement is an addition to and is not in lieu of the discussion in the Safety Evaluation Report.
1.2 General Plant Description 1.2.3 Other Major Structures In Amendment 10 to the Preliminary Safety Analysis Report, the applicant reported that two natural draft cooling towers will be used to reject the waste heat from the turbine generator's main condensers in place of the four mechanical draft towers. These natural draft towers will be in approximately the same location as the mechanical draft towers. Makeup and blowdown water service to the natural draft towers will remain as described in our Safety Evaluation Report.
The tower design and function are not safety related. The design change is re-ported here for accuracy and completeness. This design change is addressed in Section 10.4 of the Preliminary Safety Analysis Report and a complete description of applicant's design revision is found in the Yellow Creek Nuclear Plant Environmental Report, Revision 3, dated December 1, 1977.
1.9 Outstanding Issues The Safety Evaluation Report identified 19 outstanding issues which required further review in order to confirm that the proposed design would meet regulatory require-ments. We have completed our review fon each of those issues, and each has been acceptably resolved for the construction permit stage of re,few. The resolved outstanding issues. identified in the Safety Evaluation Report are summarized below and discussed in this supplement under the designated sections:
(1) We required that the applicant show by analysissthat the proposed ultimate i
heat sini. is capable of cooling the Yellow Creek plant Units 1 and 2 within the proposed design limitations (Sections 2.4, 9.2.3).
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(2) We required that the applicant show by analysis that the proposed acceleration levels for the operating basis earthquake are justified (Section 2.5).
(3) We required additional justification by the applicant that certain safety-related walls and roofs are of adequate thickness to resist penetration by tornado missiles. This relates to the applicant's planned utilization of concrete thicknesses which are based on a claimed concrete strength of 5000 pounds per square inch compressive strength one year after pouring (Section 3.5.1).
(4) We required thet the applicant provide a method for detecting intersystem leakage from high pressure systems to low pressure systems (Section 5.2.7).
(5) We required the applicant to provide a system to detect leaks outside contain-ment caused by passive failure of the emergeMy core cooling system during long-term cooling following a loss-of-coolant accident (Sections 5.2.7).
(6) We required the applicant to explicitly describe the design basis main steam line break and to report the containment pressure and temperature resulting from the break, and to report the assumptions and analytical methods used to determine the design basis results (Section 6.2.1).
(7) We required additional information to demonstrate that containment subcompart-ment pressure analyses are acceptable (Section 6.2.1).
(8) We required additional information to demonstrate that the minimum containment pressure analyses, for emergency core cooling calculations, is acceptable (Section 6.2.1).
(9) We required additional information to conclude that the applicant's pressure.
analysis for the auxiliary area is acceptable (Section 6.2.3).
(10) We required additional information on offsite radiological dose consequences incurred when a loss-of-coolant accident is postulated concurrent with con-tainment purging (Sections 6.2.4).
(11) We required additional information on the applicant's preliminary design plans for compliance with Appendix J to 10 CFR 50 (Section 6.2.6).
(12) We required additional information on the design of the ventilation system in the auxiliary area (Section 6.5.1).
(13) We required the applicant to provide safety grade instrumentation to notify the plant operator of the loss of component cooling water to the reactor coolant pumps (Section 7.4, 9.2.2).
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(14) We required the applicant to demonstrate that either (a) inadvertent actuation of containment sprays would have acceptable consequences, or (b) that current or proposed design features to preclude inadvertent actuation are or will be adequate (Section 6.2.1).
(15) We required additional information to assure that the control building and waste management building are protected against flotation due to the ground-water level at these locations (Section 3.4).
(16) We required the applicant's commitment to an augmented inservice inspection program for high energy fluid system piping between containment isolation valves (Section 3.6).
(17) We required additional information from the applicant to document that its fire protection program meets our requirements as specified in Branch Technical Position APCSB 9.5.1 " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976" (Section 9.5.1).
(18) We required additional information from the applicant to demonstrate the acceptability of emergency feedwater which is colder than the minimum specified by CESSAR (Section 1.11, Item 4).
(19) We required additional information on the Plant Monitoring System in order to complete our evaluation of the potentially safety-related aspects of this system (Section 7.7).
Two additional outstanding issues were identified at the meeting of the Advisory Committee on Reactor Safeguards on January 5, 1978. These items have also been resolved and are as listed and described below:
In Amendment 9 to the Yellow Creek Preifainary Safety Analysis Report, 1
Section 3.7.2.16, the applicant proposed a previously unreviewed method of computing the factor of safety against overturning of structures when the hydrostatic upif ft is taken into account. Additional information was required to justify the new method. Resolution of this issue is addressed in Section 3.7.2 of this supplement.
The applicant proposed to use the American Concrete Institute Standard 349-76 " Code Requirements for Nuclear Safety-Related Concrete Structures." We determined that, while application of that code was generally acceptable, certain features of that code are unacceptable and applicant commitments to acceptable alternate provisions were required. The applicant provided acceptable commitments, and we consider this issue resolved. This matter is discussed in Section 3.8 of this supplement.
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1.11 Exceptions to CESSAR Requirements In our Safety Evaluation Report thirteen exceptions to CESSAR interface require-ments were described. For each of these except items number (4) and (7), we con-cluded in the Safety Evaluation Report that our requirements were met. Items (4) and (7) concerned te.hnical matters for which we required additional information or commitments from the applicant, and as such these matters were IIsted as outstanding issues in Section 1.9 of the safety Evauation Report.
Item (4) concerned the minimum emergency feedwater temperature (outstanding issue number 18) and item (7) concerned the fire extinguishing facilitics to be provided inside containment (outstanding issue number 17).
Both of these items have been resolved.
With respect to Item 4, the applicant, in a letter dated February 14, 1978, sta+ed that the minimum emergency feedwater temperature will be 40 degrees Fahrenheit in compliance with CESSAR interface requirements and, therefore, it is now acceptable for this plant. We consider exception 4 (outstanding issue number 18) to be resolved.
Item 7 is within the scope of our overall fire hazards evaluation for the Yellow Creek facility. Our review of the applicant's overall fire hazards analysis has continued since issuance of the Safety Evaluation Report. The applicant has supplied additional information, and based on our review of this information we have concluded that the fire protection criteria and preliminary designs as presented to date are acceptable as discussed in Secticn 9.5.1 of this supplement.
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2.0 SITE CHARACTERISTICS 2.1 Geography and Demography We reported in the Safety Evaluation Report that the applicant had chosen the nearest population center (as defined in 10 CFR Part 100) to be the Florence-Muscle Shoals - Sheffield - Tuscumbia, Alabama complex, located approximately 35 miles east of the proposed plant site. The applicant has more recently reported, in a letter to the Atomic Safety and Licensing Board dated May 3, 1978, that population distribution data now show that the city of Corinth, Mississippi will probably reach a population of at least 25,000 persons during the lifetime of the Yellow Creek facility. Corinth is located about 15 miles west of the site; 10 CFR Part 100 guidelines indicate that the distance from the plant to the nearest population center should be at least one and one-third times the low population zone radius of the plant.
The low population zone radius for the Yellow Creek facility is three miles. Thus the projected relocation of the nearest population center from an area 35 miles east to a city 15 miles west of the site does not alter our previous conclusion that the population center distance meets the siting guidelines of 10 CFR Part 100 and is acceptable.
2.2 Nearby Industrial Transportation and Military Facilities In our Safety Evaluation Report we concluded that the plant will be adequately protected and can be operated with an acceptable degree of safety with regard to potential accidents which may occur as a result of nearby industrial, military and transportation facilities.
On January 25, 1978 the applicant moved the Atomic Safety and Board to reopen the hearing record in this case to permit receipt of the applicant's evidence analyzing the effect on site suitability of a recently completed oil storage facility located at the Yellow Creek Port, approximately 1.8 miles north and west of the proposed site' for the Yelinw Creek facility.
We subsequently reviewed the applicant's analysis and on February 6, 1978, moved the Board to receive staff evidence in which we oescribed our evaluation and con-cluded that "the oil tank facility as described [by the applicant] will pose no threat to the proposed Yellow Creek Nuclear Plant, and our prior conclusions as to the suitability of the proposed Yellow Creek Nuclear plant site remain unchanged."
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The Atomic Safety and Licensing Board has also considered the applicant's testimony, and in Paragraph 73 of the Partial Initial Decision issued on February 3, 1978, found that the proposed site is suitable and that no special design considerations are required for the proposed plant with regard to potential accidents which may occur as a result of nearby industrial, military, or transportation facilities. We consider this matter resolved.
2.4 Hydrologic Engineering In our Safety Evaluation Report we identified the applicant's proposed ultimate heat sink design as an outstanding issue, issue number 1.
We subsequently com-pleted our evaluation of the applicant's submittals of September 15 and October 7, 1977, involving the ultimate heat sink and in addition we completed additional independent analyses prior to reporting our position to the applicant in a letter from the staff dated January 24, 1978.
Our concerns regarding the applicant's proposed design were the following:
(1) The pond utilizes an unproven system of oriented sprays; (2) The analytical model used to predict performance has not been described in sufficient detail for us to independently review its capabilities; (3) The model has not been verified with high quality data from a cooling pond of the size and type proposed by the Tennessee Valley Authority; and (4) Review of the limited information on the predictive analytical model and other information at hand suggests that the model predicts a significantly more efficient cooling system than is indicated from similar test systems elsewhere.
The applicant has described the proposed design of the ultimate heat sink, and has identified the major features incorporated in the design to provide emergency cooling for the reactor plant. Following discussion with the applicant and issuance of the Safety Evaluation Report, we continued our analyses of the proposed design, and investigated the effects of variations of several parameters on the potential maximum pond temperature.
As a result of our analyses, we conclude that there is now reasonable assurance that the proposed ultimate heat sink will be capable of maintaining return tempera-tures below the design maximum of 100 degrees Fahrenheit. However, this conclusion is based upon results obtained from the staff's model, which, like the applicant's, has not been calibrated for the size and type of cooling pond proposed for the Yellow Creek facility. We have used our model conservatively and we have made the preliminary determination that the applicant's proposed system may perform more 2-2
efficiently than the prototype upon which our model was calibrated. However, because of the uncertainties involved in our analyses, we requested the applicant to make the following commitments to future work:
(1) To provide, no later than with submittal of the Final Safety Analysis Report, a detailed description of its analytical model to enable as to independently evaluate the model's capabilities.
(2) To obtain pre-operational data from the as-built heat sink ponds prior to a decision on issuance of an operating license, and to use the data to confirm that the pond model can be used to conservatively predict pond performance.
The pre-operational test plans should be submitted for our review prior to perfurmance of the tests, in order to assure that the type and quantity of data resulting from the tests will be sufficient to support our evaluation of the test results.
The applicant has included, in Section 9.2A.3.5 of the Yellow Creek Preliminary Safety Analysis Report, Amendment 11, commitments adequate to meet our requirements (1) and (2) above. The applicant stated, however, that "if the as-built spray pond is a commercially available oriented spray system, the need for validation and pre-operational testing.. will be considered.. and may not be necessary." To preclude future misunderstanding, the applicant is required to confirm, using pre-operational data, that the spray pond system as built for the Yellow Creek facility can be expected to perform at least as well as predicted either by the then-validated analytical model or by the design curves of a commercially available system approved by the staff.
We conclude that our independent analyses show that the proposed preliminary ultimate heat sink design meets the regulatory position of Regulatory Guide 1.27,
" Ultimate Heat Sink for Nuclear Power Plants," and that the applicant's commitments for future verification of the as-built heat sink system are acceptable at the construction permit stage of review. In the event that pre-operational testing shows that the heat sink system will not be capable of maintaining return water temperatures below the design maximum, the applicant will be required to make those modifications necessary to meet the recommendations set forth in Regulatory Guide 1.27, or to justify an alternate approach. On the basis of the foregoing, we conclude that there is reasonable assurance under 10 CFR 50.35(a) that (1) any safety questions resulting from our final design review can be satisfactorily resolved prior to issuance of an operating license for the proposed facilty, (2) the ultimate heat sink design and performance will be satisfactorily resolved prior to completion of construction of the Yellow Creek plant, and (3) taking into consideration the siting criteria of 10 CFR Part 100, the resolutten of the matter will, as regards this matter alone, allow construction and operation of the facility at tne proposed location without undue risk to the health and safety of the public. We consider this matter resolved.
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l 2.5-Geoloay and Seismology In our Safety Evaluation Report, we identified the matter of operating basis earth-quake accelerations as an outstanding issue, issue number 2.
As discussed in the Safety Evaluation Report in describing the operating basis earthquake, page 2-36, the applicant originally proposed to use acceleration values of 0.15g for soil supported structures and 0.125g for rock supported structures to scale the modified response spectra. These valtes correspond to one-half the accelerations used for the safe shutdown earthquake for soil-supported structures and rock-supported structures, respectively. This would nominally satisfy the criteria in Appendix A to 10 CFR Part 100 that "the maximum vibratory ground acceleraticn of the operating basis earthquake shall be at least one-half the maximum vibratory ground acceleration of the safe shutdown earthquake."
Subsequently, the applicant proposed to use acceleration values for the operating basis eart.iquake which meet the definition pf the operating basis earthquake given in Appendix A but which are less than one-half the acceleration of the safe shut-down earthquake. Such a departure is permitted within the guidelines of Appendix A which states: "If an applicant believes that the particular seismology and geology of a site indicate that some of these criteria, or portio'is thereof, need not be satisfied, the specific sections of these criteria should be identified in the license application, and supporting data to justify clearly such departures should be presented." The definition in Appendix A which the applicant proposed to meet is: "The ' operating basis earthquake' is the earthquake which, considering the regional and local geology and seismology and specific characteristics of local subsurface material, could reasonably be expected to affect the plant during the operating life of the plant."
l In support of such an approach, the applicant provided a probabilistic analysis of the seismicity in the region. This analysis assumed random occurrence of earth-quakes within the defined source zones and tectonic pravinces; then, using the historical record of seismicity in the region, the return period (or probability of occurrence in a given time interval) for a specified level of ground motion at the site was calculated. Using this approach the applicant determined that the return period for the site to experience a Modified Mercalli intensity intermediate between VI and VII was 500 to 1450 years. The applicant used standard intensity /
acceleration relationships to obtain its predicted acceleration values at the Yellow Creek site.
In reviewing the applicant's approach, we cc isidered the significance of alternate source zones on the determination of the probability. We found that limiting the source zone of the earthquake activity around New Madrid to an area south of Cairo, Illinois resulted in a decrease in the return period of intensity in the range of VI to VII to about 100 years. Historical data show that most of the i
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earthquake activity near New Madrid has occurred south of Cairo, Illinois. However, an additional factor which should be considered is that the source zone of the New Madrid earthquakes is more than 100 miles from the Yellow Creek site.
Recent studies by Computer Sciences Corporation (Murphy and O'Brien,1977) indicat-ed that the acceleration levels corresponding to a particular intensity level at a site were considerably lower for distant earthquake sources than for those occurr-ing nearby. If this factor is considered the return period for an acceleration level of 0.10g at the site would be greater than 300 years, approaching the return period computed by the applicant. The probability of an event with a 300 year return period occuring during the 40 year operating life of the plant would be on the order of 13 percent. And finally, even this latter finding appears conserva-tive considering a recent report by Algermissen and Perkins (1976) which indicated that an acceleration level of about 0.04g has a return period of 475 years or only about a 10 percent probability of being exceeded at sites in the vicinity of the Yellow Creek site during a 50 year time interval.
The applicant proposed to use modified Regulatory Guide 1.60 spectra, similar in shape to those used for the safe shutdown earthquake, to describe the operating basis earthquake. These spectra were scaled to 0.10g for soil suppcrted structures and to 0.08g for rock-supported structures. If one considers the low probability that the site will experience such acceleration levels, we find the ground motion for the operating basis earthquake defined by the applicant to be a conservative estimate for an earthquake which "could reasonably be exp(cted to affect the plant site during the operating life of the plant." We consider this issue resolved.
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3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.4 Water Level (Flood) Desian Criteria in our Safety Evaluation Report, we identified the matter of potential control 4
building and waste management building flotation as outstanding issue number 15.
In Amendment No. 9 to the Preliminary Safety Analysis Report the applicant committed j
to install rock anchors in the foundation of the control building and the waste management building. The anchors will consist of American Society of Testing Materials A615 Grade 60 reinforcing bars grouted into holes drilled into the rock l
foundati n We find that these anchors will provide a sufficient factor of safety against tiotation. We have completed our evaluation of the applicant's commitments and consider this issue resolved.
- 3. 5 Missile Pr4ty tior}
o 3.5.1 Missile Selection and Description The Safety Evaluation Report identified tne matter of the design of safety-related walls and roofs for protection against tornado missiles as outstanding issue number 3.
Subsequent to issuance of our Safety Evaluation Report, we discussed this matter with the applicant and clarified our p3sition in a meeting held on December 21, 1977. The applicant responded in a letter dated February 14, 1978. We notified the appilcant in a letter dated April 5, 1978 that its response was unacceptable.
The applicant provided additional information in letters da' ' April 13 and May 26, 1978.
Th applicant's commitments now include the specification of a revised missile spectrum to be considered in the desigi of the facility to nrovide protection against missiles generated by the design basis tornado.
The revised tornado ge 4: sted missile spectrum for Region I (as defined in Regula-tory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants"), the region of the country in which the proposed plant will be located, is shown in Table 3.1 of this supplement. This spectrum is in.iccordance with our requirements specified in Section 3.5 of the Standard Review Plan. On this basis, we find the revised missise spectrum information to be acceptable.
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i TABLE 3.1 REVISED TORNADO MISSILE SPECTRLM REGION I a
Mass Velocity Missiles Dimensions (meters)
(kilograms)
(meters per second)
Wood Plank 0.092 x 0.289 x 3.66 52 83 6-inch Schedule 40 Pipe 0.168 Ofaseter x 4.58 130 52 1-inch Steel Rod 0.0254 Ofaaeter x 0.915 4
51 Utility Pole 0.343 Ofaaeter x 10.68 510 55 t
12 inch Schedule 40 Pipe 0.32 Diameter x 4.58 340 47 Automobile 5 x 2 x 1.3 1810 59 aVelocities are horizontal velocities. For vertical velocities, 70 percent of the horizontal velocities are acceptable.
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We have also evaluated the thicknesses and strengths of building wall panels and roof slabs which would afford protection against the impact of tornado generated missfies. Based on the revised tornado missile spectrum and the results of Bechtel-Calspan tests on the impact effects of various missiles ("Ra wits of Missile Impact Tetts On Reinforced Concrete Panels," by J. V. Rotz, Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, Louisiana, December 1975), we have estabitshed minimum acceptable wall panel and roof slab requirements for tornado missile protection for the three regions of the country. Table 3.2 of this supplement lists the minimum concrete thickness and strengths that would be required to resist the impact effects of the tornado missiles listed in Table 3.1 for Region I.
The applicant's letters of April 13 and May 26, 1978, proposed that tornado missile protection for safety-related 2tructures would be provided by enclosing them in concrete structures having exterior walls and roofs of thickness and strength as described in Table 3.2.
Further, the concrete thickness requirements will be based on the 28 day strength of the concrete conifrmed by testing concrete cylinders at a frequency which conforms with Amarican National Standards Institute Standard N 45./.5.
We conclude that the proposed design criteria for protection of the facility against the impact of tornado generated missiles are acceptable. We consider this matter resolved.
- 3. ti Criteria for Protection Against Dynamic and Environmental Effects Associated With the Postulated Rupture of Piping In our Safety Evaluation Report, we identified as outstanding issue number 16 the need for an augmented inservice inspection program for high energy fluid system piping between containment isolation valves. Our position on this matter was further clarified and reiterated in a letter to the applicant dated May 24, 1978.
The applicant responded to our requirements in a letter dated May 26, 1978 stating that; (1) If " break exclusion" is claimed, then an augmented inservice inspection pro-gram will be implemented.
(2) Where breaks are not assumed to be excluded, breaks will be postulated in accordance with the criteria cf Branch Technical Position MEB 3-1 " Postulated Break and Leakage Locations In Fluid System Piping Outside Containment" of Standard Review Plan Section 3.6.2, and the plant will be designed to with-stand the consequences of tNse breaks, including the full dynamic effects of the breaks.
(3) Where guard pipes are used on process pipes to mitigate the consequences of a rupture of the process pipt the guard pipes will be designed for the full dynamic effects of a longitudual or circumferential break of the process pipe.
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TABLE 3.2 CONCRFTE WALL AND ROOF THICKNESS REQUIREMENTS TO
-~TE515T THE EFFECTS (F TORNADO MISSILE IMPACT k'iGION I Concrete Strength Wall Thickness Roof Thickness (pounds per square inch)
(inches)
(fnches) 3000 23 18 4000 20 16 5000 18 14 3-4
We find the applicant's commitments regarding postulated breaks in high energy piping to be in accordance with our requirements as specified in our letter of
.May 24, 1978, to the applicant and therefore, are acceptable. We consider this issue resolved.
3.7 Seismic Desian 3.7.2 Seismic System and Subsystem Analysis As described in Section 1.9 of this supplement, we identified an outstanding issue concerning an unreviewed method oi computing the factor of safety against overturning of structures when the hydrostatic uplift is taken into account. This issue was identified at the meeting of the Advisory Committee on Reactor Safeguards on January 5, 1978.
In Amendment 9 to the Yellow Creek Preliminary Safety Analysis Report, the appli-cant revised the method of computing the structural overturning moment. This method had not previously been proposed or approved in any license application. In a letter to the staff dated February 24, 1978, the applicant committed to computing the overturn
- g moments for seismic Category I structures by the method outlined in the Bechtel Power Corporation Topical Report BC-TOP-4-A, " Seismic Analysis of.
Structures and Equipment for Nuclear Power Plants," Revision 3.
We have reviewed this report and found it acceptable. We consider this issue resolved.
3.8 Desfon of Seismic Cateaory 1 Structures In Amendments 11 and 12 to the Preliminary 4fety Analysis Report the applicant proposed the use of the American Concrete Institute Standard 349-76 " Code Requirements for Nuclear Safety Related Concrete Structures" as the controlling concrete code instead of American Concrete Institute Standard 318-71.
- This matter was discussed at the meeting o the Advisory Committee on Reactor Safeguards on January 5, 1978.
At the tima of issuance of the Safety Evaluation Report, the staff had not accepted the American Concrete Institute Standard 349 Code. In the interim, Regulatory Guide 1.142, " Safety-Related Concrete Structures for Nuclear Power Plants (other
]
than Reactor Vessels and Containments)," April 1978, has been issued for comment by the staff, endorsing the American Concrete Institute Standard 349 Code with some exceptions. The applicant committed to accepting our position on the use of this i
Code.
]
To meet our requirements the applicant has stated that walls and roofs designed to provide protection from tornado missiles will have a design strength verified by tests at 28 days after pouring. However, walls and slabs which will not be utilized for protection from tornado missiles may be based on a design strength verified by test at 90 days after pouring.
3-5
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The design strength and wall thickness for tornado missile protection are described in Section 3.S.1 of this supplemen'..
We consider the applicant's use of the American Concrete Institute Standard 349-76, as modified above, to be in accordance with our requirements and, therefore, acceptable.
1 3-6
4.0 REACTOR Since issuance of the Safety Evaluation Report, we have identified a technical matter regarding the use of control element assemblies containing boron carbide encapsulated in relatively thin walled cladding. The Combustion Engineering con-trol element design referenced by the applicant uses such a design.
While elements of this type supplied by Combustion Engineering have been used successfully, we believe that such elements, comprised of tubing containing a water 4
soluble material and designed to remain in the reactor coolant for a lifetime measured in years, should be checked periodically to assure that the element integrity and reactivity worth are acceptably maintained. The applicant's plans for the conduct of such periodic surveillance during reactor operation are the type of technical information which can reasonably be left for later corsideration and will be evaluated during the operating licepse review. Consequently, we will require that the applicant submit, with the Final Safety Analysis Report, plans for routine surveillance designed to assurs that the reactivity worth designed into the control elements is not being lost through some failure mechanism. Acceptable programs could include control element reactivity checks during the first core cycle and worth measurements of all assembly banks at refueling outages.
On the basis of the foregoing, we conclude that there is reasonable assurance that this matter of control element surveillance will be satisfactorily resolved prior to the completion of construction of the proposed facility, and that, regarding this matter alone, and taking into consideration the siting criteria contained in 10 CFR Part 100, the proposed facility can be construtted and operated at the propose 'ocation without undue risk to the health and safety of the public.
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5.0 REACTOR COOLANT SYSTEM A
5.2 InteQrity of the Reactor Coolant Pressure Boundary i
5.2.7 Leakage Detection System The Safety Evaluation Report identified two areas concerning leakage detection i
provisions as outstanding issues:
(1) Intersystem leakaoe from high pressure systems to low pressure systems Olssue j
number (4)o.
(2) Leaks outside containment caused by passive failure of the emergency core cooling system during long-term cooling following a loss-of-coolant accident i
Olssue number (5)o.
The applicant submitted a letter dated January 26, 1978, in which was enclosed a copy of a letter to the applicant from Combustion Engineering. The letter from Combustion Er.gineering acknowledged that the intersystem leakage detection equip-ment is within the nuclear steam supply system scope of supply and will be addressed in the CESSAR (System 80) Final Design Approval application. We will evaluate the intersystem leakage detection system during that review. We find that the appli-cant's commitments, as discussed in its Preliminary Safety Analysis Report and the combustion Engineering letter of intent together give reasonable assurance that the final design of the intersystem leakage detection system can and will meet our requirements. We will verify this during the operating license review for the Yellow Creek facility. We consider this issue resolved.
l The applicant has also provided a system to detect leaks outside containment caused by passive failure of the emergency core cooling system during long-term cooling foilowing a loss-of-coolant accident. The system consists of drainage paths from the emergency core cooling system equipment to the aux'iliary area sumps. Sump level will be monitored and will alarm in the control room when an excessive level increase occurs. Assuming the maximum credible leak of a failed pump seal, the applicant has shown that there is sufficient time to isolate the faulted train before net positive suction head is lost or the redundant train becomes flooded.
Based on our review, we find this system acceptable since it provides assurance that the plant operator will be supplied with sufficient information and will have enough time to ensure that essential. emergency cooling is not' interrupted following i.
a loss-of-coolant accident. We consider this issue resolved.
1 5-1
6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Desian Containment Analysis We reported in the Safety Evaluation Report that the applicant had identified the worst case main steam line break as a 7.3-square foot break area at 105 percent of rated full power, assuming the failure of a main steam isolation valve to close.
The applicant, however, had not provided the results of this worst case main steam Ifne break accident to verify the adequacy of the containment functional design.
We identified this matter as outstanding issue number 6.
The applicant had provided the results of a main steam line break accident based on CESSAR mass and energy release data which did not consider the failure of a main steam isolation valve. This postulated main steam ifne break accident, without a main steam isolation valve failure, resulted in a peak calculated containment pressure of 37.4 pounds per square inch, gauge, which is 20 percent lower than the containment internal design pressure.
In Amendment 11 to the Yellow Creek Preliminary Safety Analysis Report, the appli-cant committed to provide for review a main steam line break accident analysis, including a single failure analysis, by September 1,1978. This analysis will be based on revised mass and energy release data. The new release data will be based on revised CESSAR system 80 steam generator inventories, revised main feedwater flow and feedwater isolatir.1 valve closure time. The main steam line. break mass and energy release will also be revised to account for integral flow restrictors in the main steam lines. The effect of incorporating the design changes and revised system parameters should be to reduce the mass and energy release to containment in the event of a postulated main steam line break.
The app 1tcant has stated that the net effect of the revised mass and energy releases along with consideration of the main steam isolation valve failure to close will result in a decrease in the calculated containment pressure from the value of 37.4 pounds per square inch, gauge, that is currently reported in the Preifminary Safety Analysis Report.
We have reviewed the information provided by the applicant regardirig this matter and have concluded that there is reasonable assurance that the revised main steam line break accident analysis will result in a calculated peak containment pressure.
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that is at least 10 percent less than the containment internal design pressure.
The 10 percent margin is in accordance with our standard requirement on calculated peak containment pressure at the construction permit stage of review. Therefore we expect the additional information to be submitted by the applicant to be
- confirmatory. However, if as a result of our review we find the analysis or results unacceptable, we will require the applicant to modify the design as necessary in order to meet our requirements. On this casts, we concluda that the containment internal design pressure is acceptable at the preliminary design and construction permit stage of review. We will perform our confirmatory evaluation of the applicant's September 1, 1278 submittal early in the post-construction permit period to provide time for any design changes if found necessary. We consider that outstanding issue number 6 is reso?ved.
In our Safety Evaluation Report we stated that we had not completed our review of design features provided by the applicant to preclude inadvertent actuation of the containment spray system. In a letter to the staff dated February 14, 1978, the applicant committed to design the plant to accept the consequences of an inadvertent spray actuation. In the same letter the applicant submitted information outlining alternative design approaches which individually or in combination would acceptably limit the consequences of an inadvertent spray actuation.
The applicant identified and discussed the following possible design measures to mitigate the consequences of inadvartent spray operation:
(1) Increase the minimum refueling water tank water temperature.
(2) Provide a redundant, qualified vacuum relief system on the primary containment.
(3) Increase the structural capacity of the primary containment to resist the effects of external pressure.
Based on our review, we conclude that there are several design approaches, employing state-of-the-art designs, which could be taken which would result in a maximum external differential pressure less than the design external pressure.
At the operating license review stage we will evaluate the applicant's detailed analysis of the specific design measures taken to design for an inadvertent spray actuation, demonstrating that the resultant containment external design pressures are acceptable. We will require the applicant to perform analyses for the most severe containment initial conditions.
4 We conclude that there is reasonable assurance that the matter of inadvertent spray actuation will be satisfactorily resolved prior to completion of construction of 6-2
the facility, and that, regarding this matter alone, and taking into consideration the siting criteria contained in 10 CFR Part 100, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. We consider outstanding issue number 14 to be resolved at the construction permit stage of review.
Containment Subcompartment Analysis We stated in the Safety Evaluation Report that t s 'rSlicant had not provided sufficient justification for the nodalization of the subcc'npartments for the struc-tural design evaluation or the component supports design evaluation. We, therefore, could not conclude in the Safety Evaluation Report that the results of the subcom-partment analysis were acceptable for use in the design of the subcompartment structures or the component supports.
The applicant submitted, in Amendment 9 to the Preliminary Safety Analysis Report, additional information in support of the subcompartment analysis. We have reviewed the applicant's analysis and performed a confirmatory analysis which demonstrated good agreement with the peak differential pressures calculated by the applicant.
Our confirmatory analysis also showed acceptably small differences between the pressure transients calculated by the applicant and the staff.
Further, the applicant has investigated the effect of modifying his analysis with respect to the nodalization of the reactor cavity and determined that there would be no significant changes in the design bases. We conclude that the applicant has developed an acceptable model and we find the subcompartment analysis acceptable for the construction permit review stage.
We will further review the subcompartment analysis at the operating license review stage and will require the applicant to meet any analytical or design requirements resulting from resolution of a staff generic study concerning asymmetric loads caused by a postulated loss-of-coolant accident. We consider outstanding issue number 7 to be resolved.
Containment Pressure Evaluation for Emergency Core Coolina Calculations We reported in the Safety Evaluation Report that we were reviewing the applicant's analysis of the containment backpressure to be used in the emergency core cooling system performance evaluation. Previously the applicant had revised several plant input parameters to the containment backpressure analysis and had included the effects of concurrent operation of a large volume purge system (four 48-inch lines). Following discussions with the staff, the applicant revised the Yellow Creek plant design to permit containment atmosphere purging with only a pair of eight-inch lines during the reactor operating modes when containment integrity is required. The effect of purging on the containment backpressure analysis is negligible as a result of reducing the purge line size and number. Thus, applicant 6-3
has referenced the CESSAR minimum containment back pressure analysis and has presented Yellow Creek containment design parameters to demonstrate that the minimum backpressure for the Yellow Creek containment would be no lower than that presented by Combustion Engineering in the CESSAR application.
Based on our review, including comparison of the pertinent plant parameters, we conclude that it is acceptable to reference the CESSAR minimum containment backpressure analysis for the Yellow Creek emergency core cooling system performance evaluation. We consider outstanding issue number 8 to be resolved.
6.2.3 Secondary Containment Functional Design We reported in Section 6.2.3 of the Safety Evaluation Report that the applicant had net provided sufficient information for us to conclude on the acceptability of the lower region of the secondary containment (which is called the auxiliary area) pressure analysis. The applicant provided in Amendment 11 to the Preliminary Safety Analysis Report, a description of the analytical methods and containment input parameters used in the analysis to determine the pressure response of the auxiliary area following a postulated loss-of-coolant accident. We have reviewed the applicant's analysis for determining the pressure response of the auxiliary area and find it acceptable.
The appitcant has also committed to perform functional tests to periodically demonstrate that the auxiliary area emergency gas treatment system will be capable of reducing the auxiliary area pressure and maintaining the prescribed negative pressure. We find this commitment acceptable and will review the details of the test program in the applicant's Final Safety Analysis Report. We consider outstanding issue number 9 to be resolved.
6.2.4 Containment Isolation System We reported in the Safety Eval,atson Report that the applicant had not provided an acceptable containment atmosphere release calculation for use in determining the radiological consequences of a loss-of-coolant accident while purging the contain-ment. Since the issuance of the Safety Evaluation Report the appitcant has modified the design of the containment purge system. Previously the app 1tcant had proposed the use of a large volume (fouf 48-inch lines) purge system during power operation.
The applicant has now committed to purge with only a pair of eight-inch lines during reactor operational modes when containment integrity is required, i.e.,
during power operation, startup, hot standby and hot shutdown. The applicant has committed to limit the use of the large volume purge system only to the reactor operational modes of cold shutdown and refueling.
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The applicant has also committed to insta11 redundant, Class IE radiogas monitors in the eight-inch purge ducts. These monitors will isolate the purge system on receipt of a high radiation signal and thereby provide diverse containment isolation signal parameters for a spectrum of postulated loss-of-coolant accidents.
The applicant submitted, in Amendment 9 to the Preliminary Safety Analysis Report, a containment atmosphere release calculation which we have found acceptable for use in the determination of the acceptability of the offsite radiological consequences.
The offsite radiological consequences of a loss-of-coolant event while purging are discussed in Section 15.2.2 of this supplement.
Based on the applicant's commitments and supporting analyses, we conclude that the containment purge system design and performance are acceptable. We consider outstanding issue number 10 to be resolved.
6.2.6 Containment Leak Testina Frocram We reported in the Safety Evaluation Report that we would need additional informa-tion regarding the design provisions to permit venting and draining of systems for the containment integrated leak rate test, and to permit local leak testing. 'The applicant has since provided a listing of the systems which will be capable of being vented and drained to the containment atmosphere so that the accident environment may be simulated across the containment isolation valve.
The applicant has also committed to local leak testing (Type C) in accordance with Appendix J to 10 CFR Part 50, " Preliminary Reactor Containment Leakage Testing for Water - Cooled Power Reactors." We have reviewed the applicant's containment leak testing program description and conclude it is acceptable for the construction permit stage of review.
We will review details of the containment leak testing program such as individual isolation valve test provisions, for conformance with Appendix J at the operating license review stage.
We conclude that outstanding issue number 11 is resolved.
6.3 Emeraency Core Coolina System 6.3.1 Passive Failure Protection The resolution of outstanding issue number 5, concerning leak detection for passive failures outside containment, is discussed in Section 5.2.7 of this supplement.
6.5 Encineered Safety Feature Atmosphere Cleanup System 6.5.1 System Description and Evaluation of Auxiliary Area Emergency Gas Treatment System We reported in our Safety Evaluation Report that we had requested that the appli-cant provide us with a detailed description of the location cf the supply and l
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return ventilation ducts for the auxiliary area emergency gas treatment system so 4 -
that we could evaluate the amount of credit to be given for mixing of air in the
.1 1
auxiliary area.
The applicant's response in Amendment 10 to the Preliminary Safety Analysis Report j
indicates that the individual supply and return vents within individual cubicles in the auxiliary area will be located to promote good mixing within the cubicles, and
. that a detailed description of the system performance will be given in the Final l
Safety Analysis Report.
Our own loss-of-coolant accident analysis indicates that a degree of mixing of even as low as 10 percent would not result in unacceptable dose consequences. Thus, we l
believe that the potential for unacceptable dose consequences due to possible "short circuiting" of ventilation air flow is extremely small, and therefore find the proposed auxiliary area emergency gas treatment system design acceptable. We consider outstanding issue number 12 to be resolved.
i 4
l
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P 6-6
7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.3 Engineered Safety Features Actuation System In Section 7.3.4 of our Safety Evaluation Report we noted the applicant's position 4
that the design features to be implemented in the solid state plant protection system and solid state component control system would preclude an inadvertent containment spray system actuation. This issue was identified as outstanding issue number 14. On this basis, the applicant stated that analysis of the effects of inadvertent spray actuation was not required for Yellow Creek.
We stated that our acceptance of the applicant's position would require an additional depth of design review and evaluation, over and above that normally performed at the construction permit stage of review. Such a review would require evaluation of Topical Report CENPD-172, " Plant Protection Syster," the solid state component control system and the plant monitoring system. We requested that the applicant provide additional design and analysis information to support their position. As discussed in Section 6.2.1 of this supple-e t, the applicant has agreed to perform an analysis of the effects of inadvertent spray actuation in lieu of providing detailed design and analysis information to demonstrate that the solid state plant protection system and solid state component control system design will preclude an inadvertent containment spray system actuation.
We conclude that this is an acceptable resolution of the instrumentation and control systems concerns regarding inadvertent spray actuation identified in Section 7.3.4 of our Safety Evaluation Report. Additional evaluation regarding the analyses of the effects of inadvertent spray actuation is presented in Section 6.2.1 of this supplement. As discussed in Section 6.2.1 of this supplement we consider outstanding issue number 14 to be resolved at the construction permit stage of review.
- 7. 4 Systems Required for Safe Shutdown in the Safety Evaluation Report we noted the applicant's exception to the require-ment that safety grade instrumentation be provided for monitoring the component cooling water flow to the reactor coolant pumps. This issue was identified as issue number 13. As stated in Section 9.2.2 of this supplement, the applicant, by letter dated January 26, 1978, has committed to provide safety grade instrumentation to monitor component cooling service to tre reactor coolant pumps.
As further discussed in Section 9.2.2 of this supplement, we find this commitment to be acceptable. We consider outstanding issue number 13 to be resolved.
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- 7. 6 All Other Instrumentation and Requirements for Safety 7.6.1 Environmental Qualification
~
In Section 1.8 of the Safety Evaluation Report, which concerns our requirements for future technical information, we stated our requirement "that the applicant commit to provide a technical report within one year after a decision on issuance of construction permits identifying (1) how each piece of balance-of plant Class IE equipment has been, or will be, qualified; (2) the acceptance criteria; and (3) test procedures and test results, if available, or a schedule for submittal of test procedures and results."
In response the applicant has stated in Amendment 10 of the Preliminary Safety Analysis Report that:
"To demonstrate that an acceptable qualification program, consistent with the objectives of IEEE-323-1974, is implemented for balance-o -plant Class IE equipment, a report will be prepared (withn one year of ',ne Yellow Creek CP issuance) that either:
a.
Identifies 1.
How each piece of Class IE equipment has been, or will be qualified.
a
- 11. The acceptance criterion, and 111. Test procedures and test results, if available, or a schedule for providing same.
i or:
b.
Reference to applicable topical reports for Class IE equipment supplied by CE which is outside the CESSAR scope."
We find this commitment by the applicant to be acceptable for concluding that there is reasonable assurance that an acceptable equipment qualification program, consis-tent with the objectives of the Institute of Electrical and Electronics Engineers Standard number IEEE-323-1974, will be implemented for the balance of plant Class IE equipment.
- 7. 7 Control Systems Not Required for Safety In the Safety Evaluation Report we identified a concern regarding the safety significance of the plant monitoring system with respect to its interaction with
)
the solid state component control system and with respect to its safety-related display function. This issue was identified as issue number 19. We were concerned j
l 7-2 l
that failures or misoperation of the plant monitoring system would adversely affect the operation of safety systems or would incorrectly represent the plant status to an operator, possibly causing the operator to err in the control or operation of a safety system. In response, the applicant stated that the plant monitoring system will provide only backup status monitoring and display information for safety systems and is not required to perfore any safety function during or after any of the transients or accidents identified in Chapter 15 of the Preliminary Safety Analysis Report. The applicant also stated that the plant monitoring system will be isolated from all safety system signals, power sources and structures with isolation devices which will be designed and qualified in accordance with Regulatory Guide 1.75, " Physical Independence of Electric Systems."
Although hardwired indication will be provided for display of important safety system information, the plant monitoring system cathode ray tube displays will be a more convenient and a direct source of information to the plant operator. Further-more, the plant monitoring system offers advantages over hardwired indication for identifying, comparing, diagnosing, and verifying information, and for alerting the operator. Therefore, we believe that the plant monitoring system will play an important role in the operability of the plant by identifying the values of key parameters upon which the safety analysis and safety system design are based.
Operation outside of these limits may invalidate the assumptions used in the safety analysis. Thus, a failure or misoperation of the plant monitoring system could lead an operator to make adjustments in plant operating parameters that would (without the direct knowledge of the operator) cause plant operation in violation of limiting conditions for operation. Should an anticipated transient occur while oporating in this condition, the effects could be more severe than those predicted by the accident analyses.
i Since the plant monitoring system will not be required to perform any safety functions during or after any of the transients or accidents identified in Chapter 15 of the Preliminary Safety Analysis Report, we agree with the applicant's position that the plant monitoring system, with the exception of isolation of its interfaces with the safety systems, is not required to meet the requirements for safety systems identified in Section 7.1 of the Safety Evaluation Report. However, as noted, we consider that the plant monitoring system would have an important role in providing information to the operator. Thus, we required that the applicant document for review the scope, objectives, design bases and design criteria for the plant monitoring system. The purpose of this information was to provide assurance that failure in the plant monitoring system would not directly affect the operation of any safety systems and that the plant monitoring system would accurately and reifably provide information to the operator regarding plant status and operating conditions during normal operations.
In response to our request, the applicant, in transmittals dated February 24, 1978 and March 16, 1978, provided additional information regarding the plant monitoring system. The information included:
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(1) 'A general description of the plant monitoring system functions, (2) The design principles to be used to develop the cathode ray tube display and control room arrangements including:
(a) Considerations for presentation of information (i.e., density, color, format);
(b) Operator interface, response, and performance; (c) Human factors assessment, including mock up, modeling, and simulation; (3) Design bases and criteria for isolating the plant monitoring system from safety systems; (4) Other safety considerations such as evaluating the adequacy of displays of safety information and operator training and procedures for normal, abnormal, and accident conditions; (5) A summary of the testing program, including integrated system testing, to assure that the interfaces between the safety system and the plant monitoring system are functioning properly and that the plant monitoring system is functioning as designed.
We have reviewed the information provided by the applicant. The interfacing of the safety systems to the plant monitoring system will be designed in accordance with Section 4.7, Control and Protection System Interaction, of IEEE 279-1971 and as modified by Regulatory Guide 1.75.
Based on this information, we conclude that tiie electrical isolation between the plant monitoring system and the safety systems is acceptable. On this basis, we conclude that the plant monitoring system will be isolated from all safety systems and that failures or misoperation of the plant m?nitoring system will not effect the operation of any safety systems in response to the abnormal or accident conditions identified in Chapter 15 of the Preliminary Safety Analysis Report.
1 We conclude, based on the design principles and testing program described by the applicant, that the plant monitoring system will accurately and reliably provide information to the operator regarding plant operating status and conditions during normal operation. As the primary source of operator information during normal operation, the plant monitoring system will play an important role in the operabil-ity of the plant in assisting the operator to maintain the plant within the limits upon which the safety analyses are based. Thus, we will impose technical specifi-cations governing the normal operation of the plant with the plant monitoring system out of service. The technical specifications will be structured to ensure that the operator can determine, promptly and accurately, that plant operating 7
4 conditions are within the limiting conditions for operation when plant monitoring system equipment and fur:ctions which normally provide this information are unavall-able. We require that the applicant provide the allowed outage tires, equipment and functions, testing frequency, and operatfog procedures to be in these technical specifications anc analyses to support them in tAe Final Safety Analysis Report.
i We also require that additional information regarding the plant monitoring systrm final design and testing results to deronstrate that the design principles have been implemented shall be documented in the Final Safety Analysis Report.
On the basis of the information provided by the applicant, and the requirements to be specified in the technical specifications, we conclude that the plant monitoring system design basis and criteria are acceptable for the construction permit stage of the licensing review. We consfoer outstanding issue number 19 to be resolved.
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9.0 AUXILIARY SYSTEMS 9.2 Water Systems 9.2.2 Component Coolina System In Section 1.9 of the Safety Evaluation Report, we identified as outstanding issue number 13 a concern that inadvertent failure or closure of any one of the containment isolation valves could terminate coolant flow to two reactor coolant pumps, which potentially could result in fuel damage because of a locked rotor. In response to our concern, the applicant in a letter dated January 26, 1978 has committed to providing safety grade instrumentation (this matter is also addressed in Section 7.4 of this supplement) to notify the plant operator to take corrective action upon loss of component cooling the reactor coolant pumps. Based on this commitment, we conclude that the component cooling system will meet the single failure criterion and is, therefore, acceptable. We consider outstanding issue number 13 to be resolved.
9.2.3 Ultimate Heat Sink In our Safety Evaluation Report we sta.ted that the design of the ultimate heat sink did not meet the thermal interface requirements of CESSAR. This issue was identified in Section 1.9 of the Safety Evaluation Report as outstanding issue number 18.
The ultimate heat sink design has been reevaluated and is discussed in Section 2.4 of this supplement. Based on the information provided by the applicant, we now find the design criteria for the ultimate heat sink acceptable. We consider out-standing issue number 1 to be resolved at the construction permit stage of review.
9.5 Other Auxiliary Systems 9.5.1 Fire Protection System In Amendment 8 to the Preliminary Safety Analysis Report the applicant submitted a comparison of the fire protection design for Yellow Creek with Appendix A to our Branch Technical Position APCSB 9.5.1, " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976," including a fire hazards analysis of the plant. However, the fire hazards analysis was inadequate. In Section 1.9 of the Safety Evaluation Report we identified this matter as outstanding issue number
- 17. In response to our request for additional information, the applicant, in a 1etter dated January 26, 1978, submitted a revised fire hazards analysis, inci ding a set of drawings keyed to the defined fire areas. In the revised submittal, the applicant provided the following information that was not included in its original submittal:
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(1) Identification of fire areas on the drawings and keyed to the fire hazard analysis table.
(2) Identification on drawings of rooms and equipment protected by fixed suppres-sfon systems.
(3) The fire loading for the different fire zones provided in the fire hazards analysis table.
Based on our review of the applicant's fire protection program, we conclude that although the applicant has not yet submitted all information needed by the staff to complete its review of the Yellow Creek fire protection system design against Appendix A to the Branch Technical Position APCSB 9.5 1, the applicant has supplied the staff with sufficient information to permit it to make a finding pursuant to 10 CFR S 50.35(a) that the fire protection system design is adequate for the level of review necessary at the construction permit stage. The basis for this conclusion is as follows.
I 1)
The applicant has described the principal architectual and engineering design criteria with respect to plant fire protection for the existing seismic design conditions, and has identified the major features or components incorporated therein with respect to the plant fire protection systems for the protection of the health and safety of the public.
2)
The presently designed facility has sufficient design flexibility to allow implementation of any design changes that may be necessary to assure compItance of the Yellow Creek facility with Appendix A to the Branch Technical Position APCSB 9.5-1, and which may reasonably be left,for later consideration. The final design can be supplied in the Final Safety Analysis Report (although the staff has requested this information to be supplied far in advance of the submission to the staff of the Final Safety Analysis Report).
3)
There are no safety questions associated with the features of the fire protection system that require any research and development.
4)
On the basis of the above, there is reasonable assurance that:
(!) any safety questions regarding the fire protection system will be satisfactorily resolved at or before the latest date stated in the i
application for completion of construction of the Yellow Creek facility, and (11) with respect to the fire protection system, and taking into consideration
{
the site criteria contained in 10 CFR Part 100, the proposed facilities 1
l 9-2
can be constructed and operated at the proposed locetion without undue risk to the health and safety of the public.
We consider outstanding issue number 17 to be resolved at the construction permit stage of review.
1 i
l 9-3
15.0 ACCIDENT ANALYSES 15.2 Postulated Accidents 15.2.2 Radiological Consequences of Containment Purce Durina a Postulated Loss-of-Coolant Accident As discussed in Section 6.2.4 of this supplement, and identified as outstanding issue number 10 in Section 1.9 of the Safety Evaluation Report, the applicant has proposed purging the containment while the reactor is at ' power through two eight-inch diameter purge lines. The four 48-inch diameter purge lines will only be used when the reactor is at cold shutdown conditions. The applicant has committed to a design which will isolate the containment within five seconds in the event of a postulated loss-of-coolant accident while purging. During this period, 14,640 pounds of steam containing some iodine fission product activity from the primary coolant system was originally estimated by t'.e applicant to be released from the 48-inch-diameter purge lines prior to containment isolation (the original design included the use of the 48-inch lines at power). The use of the eight inch lines will reduce the quantity of steam released prior to isolation while operating at power. Nevertheless the applicant has conservatively assumed the same quantity of steam that would be released if four 48-inch diameter purge lines were used. In evaluating the acceptability of this event, the staff has used the criterion that the radiological consequences of the loss-of-coolant accident plus purge should be within the guideline value of 10 CFR Part 100. The radiological consequences of the loss-of-coolant accident have been evaluated (see Section 15.3 of the Safety Evaluation Report) and were shown there to result in a two-hour thyroid dose at the nearest exclusior boundary of 104 rem. We concluded that this value is within the values considered appropriate at the construction permit stage by Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a loss of Coolant Accident for Pressurized Water Reactors" as well as within the guidelines of 10 CFR Part 100.
Although the source term for the design basis loss-of-coolant accident, as given in Regulatory Guide 1.4, stipulates that 25 percent of the core iodine inventory and 100 percent of the noble gases are assumed to be immediately available for leakage from the containment, it has been our position that, if the containment could be isolated within five to ten seconds after the loss-of-coolant accident, the appropriate source term to be used in conjunction with release of primary coolant through the purge valvJs would be the Iodine activity contained in the primary coolant released.
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The normal iodine activity in the primary coolant is expected to be on the order of 0.1 microcuries of iodine per gram or less, I-131 equivalent. We estimate that a release of 14,64G pounds of primary steam at this level of contamination would result in' a dose of about 0.5 rem to the thyroid at the nearest exclusion area boundary under adverse meteorology. Clearly, the loss-of-coolant accident plus purge doses for normal fodine activity and assuming only eight inch purge Ifnes would be well within the guidelines of 10 CFR Part 100.
Our concern, however, has been with the possibility of iodine coolant activity concentrations elevated significantly above normal values. The possibility of higher than normal iodine coolant concentrations, referred to as iodine spiking, raises the possi,bility that the sum of the loss-of-coolant accident plus purge doses might be substantially higher than indicated above. Iodine spiking may raise iodine coolant concentrations by factors of 100 or more within relatively short times. Although the mechanism of fodine spiking is not well understood, it has been shown to be well correlated with significant changes in power level and/or pressure. $1nce the applicant has not proposed a restriction on the timing of containment purging, we consider it likely that purging might occur when the coolant activity is elevated due to a spiking condition. Hence, we consider that the probability of a loss-of-coolant accident simultaneous with a purge when the coolant activity is elevated due to an fodine spike to be essentially the probability of a loss-of-coolant accident simultaneous with the purge event !tself.
Although the maximum iodine concentration permitted by the Combustion Engineering Standard Technical Specifications is one microcurie of iodine per gram of I-131 equivalent during steady-state power, the technical specifications permit coolant concentrations as high as 60 microcuries of fodine per gram at 100 percent power no more than 10 percent of the time, and permit even higher concentrations at reduced power levels. If a loss-of-coolant accident were to occur during containment purge, and if the fodine concentration was elevated to a level of 60 microcuries of fodine per gram I-131 equivslent, we estimate that the consequences of the purging alone would be in excess of the guideline values of 10 CFR Part 100.
The applicant in Amendment No.12 has proposed a program of administrative controls together with measurements of the coolant activity. The intent of these controls is to permit purging when the coolant activity is low (2.95 microcuries of fodine per gram) and to prevent it when it is unduly high. These proposed controls are as follows:
Power Operation (1) The lodine-131 activity in the reactor coolant system will be determined by analysis immediately prior to a planned containment purge.
(2) If this analysis indicates an Iodine-131 activity less than 2.95 microcuries of fodine per gram, the primary containment may be purged for personnel access subject to the following restrictions:
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(a) If there has been a change in power level greater than 10 percent in the time period between the prepurge analysis and initiation of containment purge, additional reactor coolant system Iodine-131 analyses will be performed again. Analyses will be performed over a sufficient time period to demonstrate that the Iodine-131 concentration is no longer I
increasing as a result of a spike condition.
(b) If a change in power level greater than 10 percent occurs during a con-tainment purge, the purge system will be isolated within 15 minutes of power change. Purging may not resume unless conditions (1) and (2)(a) are met.
(3) Containment purge will not be initiated if the Iodine-131 activity, as deter-mined in (1) or (2)a, equals or exceeds 2.95 microcuries of iodine per gram.
We find such a program to be acceptable and an effective means of controlling the offsite doses that might result from a containment purge coincident with a loss-of-coolant accident. On this basis, we have concluded that the appifcant's proposed program is acceptable for the construction permit review stage. We consider outstanding issue number 10 to be resolved. We will evaluate the final procedures for monitoring the primary coolant at the operating license review stage.
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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS At its 213th meeting c, January 5-7, 1978, the Advisory Committee on Reactor Safe-guards completed its review of the application by the Tennessee Valley Authority for permits to construct the Yellow Creek Nuclee? Plant, Units I and 2.
A copy of the Committee's report on the Yellow Creek facility, dated January 13, 1978, which contains certain comments and recommendations, is included as Appendix A to this report. The action we have taken or plan to take in response to the Committee's comments and recommendations are described in the following paragraphs.
(1) The Committee noted that our Safety Evaluation Report identified a number cf outstanding safety items, and the Committee recommencled that these matters should be resolved 8n a manner satisfactory to the staff.
Since the Januarv 5, 1978 meeting with the Committee, we have reviewed addi-tional information and commitments submitted by the applicant concerning the outstanding issues identified by the staff. We have resolved all issues. Our evaluation and resolution of these issues are reported in appropriate sections of this supples.ent to our Safety Evaluation Report.
(2) The Committee identified thost generic issues relating to large water reactors which were discussed in the Committee's report to the Commission on November 15, 1977 (Report No. 6), and which the Committee considered relevant to the Yellow Creek facility. The Committee noted that the issues should be dealt with by the staff and applicant as solutions are found.
We hwe treassitted the Committee's recomendations to the applicant for its consideration in proceeding with the Yellow Creek design. Appendix C to'the Safety Evaluation Report discusses the disposition and status of the generic matters raised by the Advisory Committee on Reactor Safeguards.
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l 20.0 FINANCIAL QUALIFICATIONS 20.1 Introduction Section 50.33(f) and Appendix C of 10 CFR Part 50 are the Commission's regulations relating to financial data and information required to establish financial qualifica-tions for applicants for facility construction permits. In accordance with these regulations, the Tennessee Valley Autority (the applicant) submitted financial information with its application, as well as providing additional financial infor-mation in response to our request. The following analysis summarizes our review of the application and the additional information and addresses the qualifications of the appitcant to finance the costs of designing and constructing the Yellow Creek Nuclear Plant, Units 1 and 2.
The applicant is a corporation wholly owned by the United States. Its overall objective of resource development includes the generation, transmission and sale of electricity to Federal agencies, industry, and to municipal and cooperative utili-ties. The applicant serves approximately 2.5 million customers. in parts of Tennessee, Kentucky Alabama, Mississippi, North Carolina, Georgia and Virginia.
The applicant's total operating revenues from its power program were $1,881.5 mill'+n for the 12 months ended June 30, 1977, and net income amounted to $126.1 million for the same period. The power program capitalization amounted to $6,701.1 million at June 30, 1977, and consisted of 70 percent long-term debt and 30 percent proprietary capital (including appropriation investment and retained earnings reinvested !n the power prcgram). The applicant's power bonds are rated "Aaa" (highest quality) by both Hoody's Investors Service and Standard and Poor's.
20.2 Construction Cost Estimates The appilcant has submitted construction cost estimates for the facility as follows:
(dollars in millions)
Ye)1ow Creek Units 1 and 2 Unit 1 Unit 2 Nuclear production plant costs................ 1,060.0 1,034.0 Transmission, distribution and general plant costs............................
26.6 29.4 Nuclear fuel inventory cost for first core.................................
123.7 134.i_
TOTAL
$T W
$1,T!RI u 20-1 l
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We have compared the cost of the proposed nuclear production plant estimated by the applicant with the cost projected by the CONCEPT costing model developed by the Energy Research t.id Development Administration (now the Department of Energy).
This analytical model projected the costs to be $2,383 elllion, compared with the applicant's estimate cf $2,120.8 all11on. Since the CONCEPT model is used primarily as a rough check on the (. cst estimates made by the applicant and is not intended to be a substitute for detailed engineering cost estimates, we conclude that it is reasonable to use the appifcant's estimate in our financial analysis.
20.3 Sources Of Funds The applicant plans to finance the Yellow Creek Nuclear Plant as an integral part of its overall power facilities construction program. Financially, the power program is separate from other applicant programs. It is required to be self-supporting and self-liquidating. New power facilities are financed largely from borrowings made through the sale of power revenue bonds and notes and, in part, from internally generated funds including retained earnings and depreciation.
Section 15d of the TVA Act (Tennessee Valley Authority Act of 1933, as amended) authorizes the applicant to issue bonds, notes and other evidences of indebtedness in wpport of its construction program. Debt service on these obligations, payable solely from the applicant's net power proceeds, has precedence over payments to the U. S. Treasury, which include a return on the net appropriation investment in power facilities plus repayments of such investment.
Table 20.1 indicates the applicant's projected sources of funds for system-wide construction expenditures during the period of construction of Yellow Creek.
In the 12-month periods ended June 20, 1976, and June 30, 1975, the applicant issued $800 al111on and $750 million, respectively, of long-term power bonds, net of redemptions. In the most recent fiscal year (the 12 months ending June 30, 1977, the applicant has issued $900 million of long-term debt. At June 30, 1977 the appilcant had $4.725 billion in long-term debt outstanding, thus indicating a history of successful bond issues.
The security for the applicant's power revenue bonds and other obligations is the revenue from its operation. Subsection 15d(f) of the TVA Act gives the applicant's Board of Directors the power, independent of any regulatory authority, to set its own rates with the objective that they be as low as feasible but requiring that they be high enough to meet the total of its financial obligations, to fully pro-tect its bondholders, and to protect the, equity of the Federal Government. In addition to its basic rates, the applicant has in operation a monthly cost of fuel andpurchasedpoweradjustmentclause.
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TABLE 20.1
' PRO FORMA STATEMENT OF SOURCES AND FUN 05 F00 SYSTEM WIDE CONSTRUCTION EXPENDITURES DURING CONSTRUCTION OF YELLOW CREEK NUCLEAR PLANT, UNITS 1 AND 2 (dollars in millions)
Fiscal Year YEAR 1978 1979 1980 1981 1982 1983 1984 1%5 1986 Long tern UsSt Issued 1,575.0 1,975.0 1,775.0 1,150.0 1,050.0 800.0 490.0 630.0 760.0 Internally Generated Funds 291.7 406.1 458.7 737.0 674.5 883.0 941.9 1,007.9 1.105.6 Total Construction
' Expenditures '$1,866.7
$2,381.1 $2,333.7 $1,887,0 $1,724.5 $1,683.2 $1,431.9 $1,637.9 $1,865.6
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20.4 Conclusion Based on the preceding analysis, we have concluded that the applicant is financial-1 ly qualified to design as$ construct the Yellow Creek Nuclear Plant, Units 1 and 2.
Our conclusion is based on the determination that the applicant has reasonable assurance of citaining the funds necessary to complete the design and construction activities including related fuel cycle costs. This conclusion is significantly influenced by the following factors as discussed above: (1) the applicant's his-tory of successful issuance of billions of dollars of its highly-rated power revenue bonds, and (2) the applicant's independent rate-setting authority: Our analysis includes the basic assumption of a viable capital market, due to the lengthy future period involved and the expected heavy dependence on external financing, h
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21.0 CONCLUSION
S In Section 21.0 of the Safety Evaluation Report, we stated that we would be able to make certain conclusions upon favorable resolution of the outstanding matters set forth in Section 1.9 of the Safety Evaluation Report. We have discussed these matters in this supplement and indicated favorable resolution of each matter.
Accordingly, we reaffirm the conclusions as set forth in Section 21.0 of the Safety Evaluation Report.
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l APPENDIX A REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS f
UNITED STATES y
)e a,
NUCLEAR REGULATORY COMMISSION T. '
l OVISORY COMMITTEE ON REACTOR SAFEGUARDS WA$mGToN. O. C. 20655 January 13, 1978 Honorable Joseph M. Hendrie Chairman U. S. Nth: lear Regulatory Comission Washington, DC 20555
Subject:
REPORP Of YELIIN CREEK NUCLEAR PUNr, UNITS 1 AND 2 Dear Dr. Hendrie During its 213th meeting, January 5-7, 1978, the Advisory Comittee on Reactor Safeguards conpleted' its review of the application of the Tennessee Valley Authority (the Applicant) for a permit to construct the Yellow Creek Nuclear Plant, Units 1 and 2.
The application was reviewed at a Subcomittee meeting in Corinth, Mississippi on Decem-ber 16, 1977. A tour of the site was made by Subcommittee members on December 15, 1977. During its review, the Comittee had the benefit of discussions with representatives and consultants of the Applicant, Contustion Engineering Incorporated, and the Nuclear Regulatory Commis-sion (NBC) Staff. We Comittee also had the benefit of the ducuments listed.
The Yellow Creek Plant site is located in Tishomingo County, approxi-mately 15 miles east of Corinth, Mississippi. The minimum exclusion area distance is 695 meters; the low population zone radius is three miles. The nearest populaticn center is the Florence-Muscle Shoals-Sheffield-Tuscumbia, Alabama couplex (1970 population of about 62,900) which is located approximately 35 miles east of the site.
The application for the Yellow Creek Plant was subnitted in accordance with the Comission's standardization policy as described in Appendix 0 to Part 50, " Licensing of Production and Utilization Facilities," and Section 2.110 of Part 2,
- Rules of Practice," of Title 10 of the Code of Federal Regulations. For this application, the reference system is the Combustion Engineering Standardized Nuclear Steam Supply System known as Standard Reference System-80. This design has been reviewed by the ACRS and was discussed in its report of September 17, 1975, "Coubustion Engineering Standard Safety Analysis Report - CESSAR-80."
The reactor contairunent scheme for the Yellow Creek plant consists of a spherical steel vessel and a reinforced concrete shield building, generally similar to those for the Cherokee and Perkins plants.
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I Honorable Joseph M. Hendrie January 13, 1978 An acceleration of 0.25g for rock-supported structures and 0.30g for soil-supported structures has been specified for the safe shutdown earthquake selected for the Yellow Creek Plant. The applicant has used a probabilistic treatment to choose an operating basis earthquake with
' which are associated ground level accelerations of 0.08g for rock sup-ported structures and 0.10g for those supported on soil. The re=ni ttee considers these values acceptable.
The NRC Staff has identified a nLaber of safety items which will require resolution before issuance of a mnstruction permit. These matters should be resolved in a manner satisfactory to the NRC Staff.
With regard to generic problems cited in the Committee's report, " Status of Generic Items Relating to Light-Water Reactors: Report No.
6," dated November 15, 1977, items mnsidered relevant to the Yellow Creek Plant are:
II-1, 2, 3, 4, 5B, 6, 7, 9, 10; IIA-2, 3, 4; IIB-1, 2; IIC-1, 2, 3A, 3B, 4, 5, 6; IID-1, 2; IIE-1. These problems should be dealt with by the Staff and Appl!. cant as solutions are found.
The Advisory Conmittee on Reactor Safeguards believes that if due con-sideration is given to the foregoing, the Yellow Creek Nuclear Plant, Units 1 arxl 2 can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public.
Sincerely yours Stephen Lawroski Chairman References I
1.
Yellow Creek Nuclear Plant, Units 1 and 2 Preliminary Safety Analysis Report, Voltanes 1 through 6 and Amenchments 1 through 10.
2.
Safety Evaluation Report related to construction ot' Yellow Creek Nuclear Plant, Units 1 and 2, NUREG-0347, December 1, 1977, Docket Nos. STN 50-566 and STN 50-567.
3.
Letter from J. E. Gilleland, TVA to O. D. Parr, NRR, on safety grade instrumentation to detect the loss of cooling water to the reactor molant ptmps, dated September 14, 1977.
4.
Letter from J. E. Gilleland, TVA to D. B. Vassallo, NRR, on revised operating basis earthquake, dated October 14, 1977.
i S.
Intter from J. E. Gilleland, TVA, to O. D. Parr, NPR, on plant I
monitoring system, dated November 7,1977.
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Honorable Joseph M. Hendrie January 13, 1978 6.
Letter from J. E. Gilleland, TVA, to D. B. Vaanalin, NRR on out-standing issues, dated November 14, 1977.
1.
I4tter from J. E. Gilleland, TVA, to O. D. Parr, NRR, on out-standing issues, dated November 23, 1977.
8.
Letter from G. R. Lanning, Jr., to the Nuclear Regulatory t'nmunis-sion Advisory censmittee [ sic], on geology and emergency plans, dated December 16, 1977.
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APPENDIX B CONTINUATION OF THE CHRONOLOGY OF REGULATORY REVIEW December 1, 1977 Letter from applicant transmitting additional information on the fire hazards analysis.
December 5, 1977 Letter from appilcant transmitting Amendment No. 10 to the Yellow Creek Preliminary Safety Analysis Report, consisting of revised text.
Decuber 8,1977 Letter to applicant transmitting twenty copies of Safety Evaluation Report.
December 14, 1977-Letter from applicant concerning the plant monitoring system and the emergency feedwater temperature.
December 16, 1977 Letter from applicant transmitting additional financial qualifications information requested in staff letter of October 6,1977.
December 16, 1977 Advisory Committee on Reactor Safeguards Subcommittee meeting with staff and applicant, in Corinth, Mississippi.
December 21, 1977 Representatives from applicant and staff meet in Bethesda, Maryland to discuss several unresolved issues on the Yellow Creek project.
January 3, 1978 Letter to applicant concerning further definition of staff position regarding minimum containment pressure analyses.
January 5, 1978 Advisory Committee on Reactor Safeguards Full Committee meeting with staff and applicant, in Washington, DC.
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l January 13, 1978 Advisory Committee on Reactor Safeguards issues letter report to the Commission Chairman concerning the Committee's review of the Yellow Creek project.
January 18, 1978 Letter to appifcant transmitting a copy of the January 13, 1978 report by the Advisory Committee on Reactor Safeguards.
January 24, 1978 Letter to applicant regarding staff position on resolution of outstanding issue number 1 concerning the ultimate heat sink.
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January 24, 1978 Letter from applicant transmitting new information about an oil storage facility located approximately two miles from the proposed Yellow Creek site.
January 26, 1978 Letter from applicant transmitting results of an analysis entitled " Yellow Creek Post - Tank Form Hazards."
January 26, 1978 Letter from applicant transmitting additional information concerning issues 4, 6, 11, and 13 in staff's Safety Evauation Report.
February 9, 1978 Letter to applicant issuing a Limited Work Authorization to conduct a number of nonsafety-related activities and one safety-related activity.
February 14, 1978 Letter from applicant transmitting additional information concerning issues 3, 6,14,18, and 21 in staff's Safety Evaluation Report.
February 24, 1978 Letter from applicant transmitting additional information concerning issues 1, 19, and 20 in staff's Safety Evaluation Report.
March 2, 1978 Letter to applicant concerning status of the review of outstanding issues.
March 31, 1978 Letter from applicant transmitting Amendment No.11 to the Yellow Creek Preliminary Safety Analysis Report, consisting of responses to outstanding issues and miscellaneous design changes.
April 5, 1978 Letter to applicant concerning the resolution of outstanding issue number 3, design strength of fly ash concrete.
April 13, 1978 Letter from applicant transmitting a response to staff letter of Apris o, 1978 concerning resolution of outstanding issue number 3 on design strength of fly ash concrete.
May 5, 1978 Letter to applicant transmitting draft copy of NUREG-0219 and requesting
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comments.
May 24, 1978 Letter to applicant restating the staff position on outstanding issue number 16 concerning augmented in-service inspection of high energy
{
piping between containment isolation valves.
j Jine 7, 1978 Letter from applicant transmitting Amendment No. 12 to the Yellow Creek Preliminary Safety Analysis Report, consisting of responses to outstanding issues and miscellaneous design changes.
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APPENDIX C ERRATA TO THE EVALUATION REPORT 1.
Page 1-11, Item (2), "...(Section 2.5.2.3)." should read "..(Section 2.5.2, page 2-36)."
2.
Page 1-12, item (14), "..(Section 6.2.1, 7.3.5)." should read "..(Sections 6.2.1, 7.3.4)."
3.
Page 1-12, Item (10) "...(Sections 6.2.4, 15.2.2)." should read "..(Sections 6.2.4, 15.2.2 page 15-6)."
4.
Page 2-10, second paragraph, ".. (approximately 10.5 pounds per square foot)..." should read "...(approximately 9.5 pounds per square foot).
5.
Page 2-11. fourth paragraph, second line, "..onsite joint frequenry...", should be
".. representative joint frequency..".
6.
Page 2-11, Section 2.3.3, third line, "6000" should read "3600".
7.
Page 2-12, Table 2.2, last entry in lef t hand column, "F6" should read "less than 6 feet".
8.
Page 2-13, second paragraph, next to last line, "...using values presented by Gifford for.. atmospheric conditions." should read ".using diffusion estimates presented by Gifford."
9.
Page 2-14, first paragraph, ".. Routine Releases from Light-Water-Cooled Reactors,"
(Errata January 1977).. " should read ".. Routine Releases from Light-Water-Cooled-Reactors," Revision I, July 1977.. "
10.
Page 2-20, last paragraph in Section 2.4.6; "...Section 15.7.." should read
"...Section 15.4...".
- 11. Page 2-14, first paragraph, "...provided in Regulatory Guide 1.111 adjustments to..."
should read "...provided in Regulatory Guide 1.111, March 1976, adjustments to..."
- 12. Page 6-4, last line on page; delete "...was informed that the steel are'a proposed for use in the calculation was...".
- 13. Page 7-9, fourth paragraph, "...and in a letter to the staff dated October, 1977,.
should read "...and in a letter to the staff dated October 7, 1977,...".
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- 14. Page 15-8, second paragraph, "...the lowest average annual flow of record for the..."
should read "..the average annual flow for the.. "
- 15. Page'17-10, delete "..., subject to the appitcant's satisfactory documentation concerning
-i c revision in the Combustion Engineering quality assurance organization....".
- 16. Page B-5 (Appendix B), Entry: November 16, 1977 should be November 17, 1977.
- 17. Page B-5'(Appendix B), on entry of November 23, 1977:
[.
"...concerning seven outstanding fssuls." should read "..concerning 26 outstanding issues."
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APPENDIX 0 I
ADDITION TO THE BIBLIOGRAPHY 1.
Algermissen, S.. T. and D. M. Perkins, a Probabilistic Estimate of Maximum Acceleration I
in Rock in the Contiguous United States, U. S. Geological Survey Open-Flie Report 76-416.
2.
Murphy, J. R., and L. J. O'Brien,1977, The Correlation of Peak Ground Acceleration Amplitude with Seismic Intensity and Other Physical Parameters, Bulletin of the Seismological Society of America, Vol. 67, pp. 877-915.
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