ML19256A292
| ML19256A292 | |
| Person / Time | |
|---|---|
| Site: | Black Fox |
| Issue date: | 11/17/1978 |
| From: | Fate M PUBLIC SERVICE CO. OF OKLAHOMA |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811280250 | |
| Download: ML19256A292 (5) | |
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6212 DIN 8-013-523 PUBLIC SERVICE COMPANY OF OKLAHOMA A CENTRAL AND SOUTH WEST COMPANY P O BOX 201/ TULSA. OKLAHOMA 74102 / (918) 583-3611
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November 17, 1978 Executie Vice Presk/ent File: 6212.125.3500.32 L Public Service Company of Oklahoma Black Fox Station PSAR Amendment 13 g Docket STN 50-556 and STN 50-557 M Office of Nuclear Reactor Regulation Division of Project Management Light Water Reactors Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attn: Nteven A. Varga, Chief Gentlemen:
Public Service Company of Oklahoma hereby submits sixty (60) copies plus three signed originals of Amendment 13 to the Black Fox Station, Units One and Two, Preliminary Safety Analysis Report.
This Amendment is being submitted to update various areas of the docket as com-mitted to provide information to you. This information was previously submitted to you by letters dated November 7, 1978 (our 6212 DIN 8-013-291 and 6212 DIN 8-013-292) and November 10,1978 (our 6212 DIN 8-013-391) . Subjects covered are The Ability of Ultimate Heat Sink to Withstand Tornado Missiles, LPCI Diversion Effects On ECCS Performance and revised PSAR Figure 14.16a showing the containment elevator pit framing.
In addrion, Amendment 13 contains information on the educational and experience requirements for key personnel participating in the initial test program. This information is provided to update our response to staff question Q413.3A and reflects our present status in this area.
Finally, a copy of correspondence between General Electric Company and PSO regarding Calculated Peak Cladding Temperature From ECCS Analysis Following a Design Basis N Accident For Black Fox Station is included as a part of this Amendment for your information.
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Attacgmentsttached Service List w/o Attachments
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- BLACK FOX STATION SERVICE LIST XC: L. Dow Davis, Esquire Joseph R. Farris, Esquire William D. Paton, Esquire John R. Woodard, III, Esquire Colleen Woodhear, Esquire Green, Feldman, Hall & Woodard Counsel for NRC Staff 816 Enterprise Building U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74103 Washington, D. C. 20555 Andrew T. Dalton, Esquire Mr. Cecil Thomas 1437 South Main Street, Suite 302 U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74119 Phillips Building 7920 Norfolk Avenue Mrs. Ilene H. Younghein Bethesda, Maryland 20014 3900 Cashion Place Oklahoma City, Oklahoma 73112 Mr. Jan A. Norris Environmental Projects Branch 3 Mr. Lawrence Burrell U.S. Nuclear Regulatory Commission Route 1, Box 197 Phillips Building Fairview, Oklahoma 73737 7920 Norfolk Avenue Bethesda, Maryland 20014 Mrs. Carrie Dickerson Citizens Action for Safe Energy, Inc.
Mr. William G. Hubacek P. O. Box 924 U.S. Nuclear Regulatory Commission Claremore, Oklahoma 74107 Office of Inspection and Enforcement Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76012 Mr. Gerald F. Diddle General Manager Associated Electric Cooperative, Inc.
P. O. Box 754 Springfield, Missouri 65801 Mr. Maynard Human General Manager Western Farmers Electric Cooperative P. O. Box 429 Anadarko, Oklahoma 73005 Michael I. Miller, Esq.
Isham, Lincoln & Beale One 1st National Plaza Suite 4200 Chicago, Illinois 60603 Mr. Joseph Gallo Isham, Lincoln & Beale 1050 17th Street N.W.
Washington, D. C. 20036
+
RLACK & VEATCH TEL. (913) 967 2000 CONSULTING ENGINEERS TELEX 42-6263 1500 M E A DOW LAKE PAR KW AY M AILING ADDRESS: P O. BOX No. 8405 KANSAS CITY. M ISSOU RI 64114 In the Matter of the Application ) Docket Nos. STE 50-556 of PUBLIC SERVICE COMPANY OF ) STN 50-557 OKLAHOMA Black Fox Station ) 6212.125.2000.21 Units 1 and 2 )
CERTIFICATE OF TRANSMITTAL I, Edwin L. Cox, Project Licensing Engineer for Black Fox Station, Black
& Veatch Consulting Engineers, certify that a copy of Amendment 13 to Public Service Company of Oklahoma's Black Fox Station (STN 50-556 and STN 50-557) Preliminary Safety Analysis Report has been transmitted to the following by United States Mail, postage prepaid, this 21st day of November, 1978.
The Honorable Tommy Dyer Mr. Dale McHard Mayor of Inola Oklahoma State Department of Health City Hall N. E. 10th St. & Stonewall Inola, Oklahoma 74036 P. O. Box 53551 Oklahoma City, Oklahoma 73105 Director State Grand-in-Aid Clearinghouse Mr. Gerald Diddle 4901 N. Lincoln Blvd. Associated Electric Coop, Inc.
Oklahoma City, Oklahoma 73105 P. O. Box 754 Springfield, Missouri 65801 Environmental Impact Coordinator Encironmental Protection Agency Mr. Paul M. Murphy First International Building Isham, Lincoln, and Beale 1201 Elm Street One First National Plaza Dallas, Texas 75270 42nd Floor Chicago, Illinois 60690 Mr. Maynard Human Western Farmers Electric Coop Mrs. Ilene Younghein P. O. Drawer 429 3900 Cashion Place Anada.-ko, Oklahoma 73005 Oklahoma City, Oklahoma 73112 Robert A. Franden, Esq. Chief, Docketing and Service Section Green, Feldman, Hall & Woodard Off.'.ce of the Secretary 816 Enterprise Building U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74103 Washington, D. C. 20555
BLACK & VEATCH Mr. Harold R. Denton Mr. Lawrence Burrell Office of Nuclear Reactor Regulation Route #1 U. S. Nuclear Regulatory Commission Box 197 Washington, D. C. 20555 Fairview, Oklahoma 73737 Andrew T. Dalton, Esq. Joseph Farris, Esq.
1437 South Main St., Suite 302 Green, Feldman, Hall & Woodard Tulsa, Oklahoma 74103 816 Enterprise Building Tulsa, Oklahoma 74103 John R. Woodard, Esq.
Green, Feldman, Hall & Woodard Atomic Safety and Licensing Appeal 816 Enterprise Building Board Panel Tulsa, Oklahoma 74103 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Ms. Carric Dickerson P. O. Box 924 Claremore, Oklahoma 74107 Copies of PSAR Amendment 13 were transmitted to PSO Legal Counsel, Isham, Lincoln and Beale, Washington, D. C., for service to:
Sheldon J. Wolfe, Esq.
Mr. Frederick J. Shon Dr. Paul M. Purdom Edwin L. Cox
STATE OF OKLAHOMA COUNTY OF TULSA Martin E. Fate, Jr., being first duly sworn, deposes and states; That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the Applicant herein; that he has read the following Amendment 13 to the Black Fox Station Units One and Two Preliminary Safety Analysis Report and knows the contents thercof; that the same is true as he verily believes.
DATED: This /66 day of 'l') mm./uA) ,1978.
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, m f Mart [E.'TMJr.
Executive Vice President h)
Subscribed and sworn to before me this /// 6 day of %c<am/A/ ,1978.
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. Notary Public in and for the# dounty of Tulsa, State of Oklahoma My Commission expires b(etum f /9f L j /
STATE OF OKLAH0MA COUNTY OF TULSA Martin E. Fate, Jr., being first duly sworn, deposes and states; That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the Applicant herein; that he has read the following Amendment 13 to the Black Fox Station Units One and Two Preliminary Safety Analysis Report and knows the contents thereof; that the same is true as he verily believes.
DATED: This /[o #3 day of '7) m _ A A ) ,1978.
Sign O b Mar,t47I~E M , Jr. (
Executive Vice Presidenf Subscribed and sworn to before me this /(, M day of 'h u v d o) ,1978.
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Notary Public in and for the County of Tulsa, State of Oklahoma My Lomission expires h>M uuo f / 91 L '
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BFS PSAR AMENDMENT 13 DOCKET STN 50-556 and STN 50-557
STATE OF OKLAHOMA COUNTY OF TULSA Martin E. Fate, Jr. , being first duly sworn, deposes and states; That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the Applicant herein; that he has read the following Amendment 13 to the Black Fox Station Units One and Two Preliminary Safety Analysis Report and knows'the contents thereof; that the same is true as he verily believes.
DATED: This 16th day of November , 1978.
Signed s/ Martin E. Fate, Jr.
Martin E. Fate, Jr.
Executive Vice President Subscribed and sworn to before me this 16th day of _, November , 1978.
Roberta L. Parkey Notary Public in and for the County of Tulsa, State of Oklahoma My Commission expires February 5, 1982
BFS ERRATA AND ADDENDA SHEET AMENDMENT 13, NOVEMBER 17, 1978 Remove Page Insert Page Chapter One 1.9-22 1.9.22 1.9-23 1.9-23 Chapter Three 3C.14-17a 3C.14-17a Chapter Six 6.3-1 through 6.4-1 6.3-1 through 6.4-1 6.3-la through 6.4-la 6.4-la Chapter Nine 9.2-11 9.2-11 Chapter Thirteen 13.1-9 13.1-9 13.2-6 13.2-6 13.4-la 13. 4-la Chapter Fourteen 14.1-1 14.1-1 14.1-2 14.1-2 14.1-3a 14.1-3a 14.1-3b 14.1-3b 14.1-6 Amendment 7 413-9 4 13 - 9 Place the Amendment 13 page tab, notarization letter, and Amendment 13 instructions behind the Amendment 12 Questions i
13-111778
BFS ERRATA AND ADDENDA SHEET (Continued)
Remove Page Insert Page and Responses. The remainder of Amendment 13, consisting of the LPCI Diversion Ef fects Analysis and GE correspondence, is to be placed behind the Amendment 13 instructions.
11 13-111778
BFS TABLE 1.9-1 (Continued)
Regulatory Guide Title and Applicant's Position 1.111 Methods for Estimating AtmospLaric Transport and Dispersion of Gr aeous Ef fluents in Routine Releases from Light-Water-Cooled Reactors (Rev. O, 3/76 and Errata, 1/77)
Iodine dose pathway models follow this guide. A less con-servative analysis (General Electric's model) was also used.
See the response to Regulatory Guide 1.109 for additional CoEments.
1.112 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Fower Reactors (Rev. O, 4/76)
Calculations utilized iodine and liquid radwaste source terms similar to those in this guide. In addition, calcu-lations were also performed using source terms from General Electric's report, NED0-21159, for the above parameters.
See PSAR Sections 11.2 and 11.3 for additional details.
1.113 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I (Rev. O, 5/76)
The aquatic dispersion model from WASH-1258 was used in the preparation of the PSAR. That model is similar to that pre-sented in this guide.
1.114 Guidance on Being Operator at the Controls of a Nuclear Power Plant (Rev. O, 2/76)
PS0 intends to meet the provisions of this guide.
1.115 Protection Against Low-Trajectory Turbine Missiles (Rev. O, 3/76)
PS0 will meet the provisions of this guide in the design of BFS.
1.116 Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems (Rev. O, 6/76)
PSO intends to meet the provisions of this guide.
1.117 Tornado Design Classification (Rev. O, 6/76)
All systems, structures, and components required by this guide to be protected against the effects of tornados are protected by being housed in Category I structures. The exception to this protection is the Off-Gas System which 13 is located in the Turbine Building. For this system, PS0 has adopted the GE position that the release of radioactivity as a result of damage to the system by a tornado or tornado-generated missiles will not exceed 1.9-22 13-111778
BFS TABLE 1.9-1 (Continued)
Regulatory Guide Title and Applicant's Position 10 CFR 100 guidelines.
13 1.118 Periodic Testing of Electric Power and Protection Systems (Rev. O, 6/76)
The PS0 supplied scope will conform to IEEE 338-71 as stated in PSAR Section 8.1. GE supplied scope will also meet the criteria of IEEE 338-71 as stated in GESSAR Table 7.1-1.
1.119 Surveillance Program for New Fuel Assembly Designs (Rev. O, 6/76)
This guide is not applicable to BFS since a previously proven fuel design will be utilized.
1.120 Fire Protection Guidelines for Nuclear Power Plants (Rev. O, 6/76)
In lieu of following Regulatory Guide 1.120, PSO intends to comply to the extent practicable with Branch Technical Iosition APCSB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants," as modified by Appendix A to Branch Technical Position APCSB 9.5-1, entitled " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976."
In November, 1977 PS0 will submit a detailed fire hazard analysis for Black Fox Station.
1.121 Bases for Plugging Degraded PWR Steam Generator Tubes (Rev. O, 8/76)
Not applicable to BFS.
1.122 Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components (Rev. O, 9/76)
PSO will meet the requirements of this guide in the design of BFS with the exception of C.1, Boundary Peaks, which is discussed in Subsection 3.7.2.8.
1.9-23 13-111778
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PARTIAL PL AN EL 590'- 9 5/8' T O S CONTAINMENT EL 59T-7' . NOTES L STRUCTURAL STEEL FRAMING AT EL 589' 8 SH ALL BE EL 58T- MS/g* I i LIMITED TO SECTIONS FOR WHICH THE WIDTH DOES NOT EXCEED 20~.
- 2. SUPPRESSION POOL HWL CL 570'-8*.
- 3. STRUCTURAL STEEL AT OR A80VE EL 589*-s* WILL BE P LA T E- DESIGNED IN ACCORDANCE WITH SECTION 12.0 OF BFS PSAR APPENDIX 3C, SPECIFICALLY FIGURE 12 2 SECTION 2 TO PREVENT THE ELEVATOR OR ANY PARTS THEREOF FROM FALLING INTO THE SUPPRE$$ ION POOL.
6' It' 2' 0 5' 10' 3/16" = l'-0*
PUBLIC SERVICE COMPANY OF OKLAHOMA OI ACE FOR ST ATION - Unit t REACTOR BUILDING-PLATFORM FRAMING 14-16 o ELEVATOR PIT FRAM!NG REACTOR BUILDING-PLATFORM FRAMING ELEVATOR PIT FRAMING u
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BFS 6.3 EMERCENCY CORE COOLING SYSTEMS (CESSAR) 6.3.1.4 Capability to Meet Functional Requirements (CESSAR)
Addition: (CESSAR) page 6.3.2)
(8) Additional design conservatism for the pump suction lines passing through containment is provided as described in Subsection 3.9.3.5.
Justification: To identify design commitment described in Subsection 3.9.3.5.
6.3.2.2.1 High Pressure Core Spray System (HPCS). (GESSAR Confirmed) 6.3.2.2.7 ECCS Pump Suppression Pool Screens (GESSAR).
Addition: The capability of ECCS and RCIC equipment to function as needed in the presence of particles which can pass through the Supression Pool suction line inlet screens will be reviewed and the appropriate requirements will be included in specific equipment specifications, as required, to ensure the proper performance of the ECCS and RCIC systems.
6.3.3.10 Conformance with ECCS Acceptance Criteria of 10CFR50.46. (CESSAR)
Addition: Since the issuance of the 238 NSSS GESSAR changes have occurred which affect the calculated peak cladding temperature (PCT).
Correspondance received by PS0 from GE summarizing these changes is incorporated into this PSAR. A copy of the correspondance is located in Volume X of this PSAR along with the LPCI Diversion report behind the index tab for Amendment 13.
13 6.3.3.12 Use of Dual Function Components (GESSAR)
Addition: Analysis of EPCl diversion on ECCS performance for Black Fox Station Units 1 and 2 supplies specific information pertaining to ECCS performance during diversion of flow from two of the RHR pumps to containment spray. This analysis is included behind the question tab for Amendment 13.
6.3.3.14 Thermal Shock Considerations. GESSAR Confirmed) 6.3.4 Tests and Inspections (GESSAR C6nfirmed) 6.4 HABITABILITY SYSTEMS 6.3-1 through 6.4 13-111778
6.4.1 Habitability Systems Functional Design The habitability area design employs several systems and provisions to assure habitability for a suitable period following postulated design basis accidents. These habitability systems and provisions include the following.
(1) Shielding (2) Habitability Area HVAC System (which includes the Emergency Air Cleanup Filtration System)
(3) Food and water storage (4) Kitchen and sanitary facilities 6.4.1.1 Design Bases. The functional design of the habitability systems and provisions are established by the following design bases.
(1) Shielding (a) The whole bcdy gamma radiation dose due to direct shine from internal and external sources, will not exceed 5 rem for 30 days following a DBA.
(2) Habitability Area HVAC System (a) The thyroid dose from the inhalation of radioactive idione for the 30 day period following a LOCA will not exceed 30 rem to the thyroid.
(b) The beta skin dose from airborne activity within the habi-tability area for the 30 day period following a IOCA will not exceed 30 rem.
(c) The habitability area environment will be maintained at a minimum drybulb temperature of 70 F and a aaximum drybulb temperature of 80 F. Maximum relative huuidity will be 60 per cent. For other design bases relative to the Habitability Area HVAC System, refer to Subsection 9.4.1.
(d) Carbon dioxide and oxygen concentration in the habitability area atmosphere will be maintained within safe levels, both during complete airpatch isolation of the habitability area and when outside air intakes are open.
(c) The habitability area envelope will be designed for minimal leakage such that airborne radioactivity inleakage is within acceptable limits following a design basis accident.
(f) Habitability requirements are based upon the assumptions of 6.4-la 13-111778 13
BFS 9.2.5.3 Safety Evaluation. The Ultimate Heat Sink is capable of providing sufficient cooling for more than 30 days:
(1) To permit simultaneous safe shutdown of both nuclear reactor units and to maintain them in a safe shutdown condition and (2) In the event of an accident in one unit, to permit safe control of the accident and also permit simultaneous safe shutdown of the other unit and to maintain it in a safe shutdown condition.
The UHS, consisting of cooling towers, fans, basin, pump house, and makeup basin will be designed to withstand, without a loss of functional capability to meet the requirements of items (1) and (?.) above, the following natural phenomena; safe shutdown earthquake, probable maximum flood, and tornado wind forces and tornado borne missiles.
The UHS cooling towers, basin, and the pump house will be constructed of concrete walls and roofs.
The Ultimate Heat Sink is capable of withstanding the effects of other applicable site-related events, reasonably probable combinations of less severe phenomena, and any single creditable failure of any active component without loss of the sink capability to provide the heat rejection necessary to meet requirements of items 1 and 2 above. A single nonmechanistic failure of a man-made structural feature of the UHS is considered to be incredible since the UHS is designed to Seismic Category I requirements. Refer to Tables 9.2-2 and 9.2-3 for a failure analysis of the Standby Service Water System and the UHS.
The Ultimate Heat Sink in conjunction with the SSWS is designed to withstand, without loss of safety function, the disabling of any one cooling tower fan simultaneous with a single active failure in any one Unit 1 or Unit 2 division during a loss of preferred power.
13 9.2-11 13-111778
BFS 13.1.2 Operating Organization This section describes the organization structure, functions, and responsibi'ities of the PSO operating organization for Black Fox Station.
13.1.2.1 Plant Organization. The organizational structure of Black Fox Station is shown on Figure 13.1-4. This structure is based upon information presented in ANSI N18.1-1971, WASH 1130, and as a result of review of the organization of other nuclear power stations. The number of persons in each position is shown on Figure 13.1-4 as is the license requirement to comply with 10 CFR 55. These requirements also comply with the information presented in 10 CFR 50.54.
13.1.2.2 Personnel Functions, Responsibilities, and Authorities 13.1.2.2.1 Station Manager. The Station Manager has overall responsi-bility and authority for all phases of operation of the staff. He will report to the Vice President, Power Generation in the PS0 General Office.
He is directly responsible for the safe, orderly, and efficient operation of the station including administration of the BFS Emergency Plan and is responsible to see that the station is operated in accordance with the license and applicable regulations. He serves as the station's NRC liaison for all communications concerning station operation. He is responsible to maintain a qualified staff and assure that they are properly trained. By supervisiag the activities of the Training Coordinator, he shall assure that an effective training and retraining program is maintained.
The Station Manager shall also approve all administrative procedures and policies regarding station operation and maintenance. Through the Security and Office Supervisor, the Station Manager shall direct a security force and the office staff.
13.1.2.2.2 Station Superintendent. The Station Superintendent reports directly to the Station Manager and acts on behalf of the Station Manager in his absence. He will be responsible for coordinating and directing the activities of the Technical Supervisor, Radiation Safety and Chemistry (RS&C)
Supervisory, Operations Supervisor, and the Maintenance Supervisor. He will be chairman of the Operations Committee and the Test Working Group. 3 13.1.2.2.3 Security and Office Supervisor. The Security and Office Supervisor will be responsible for implementing and supervising the station security.
13.1-9 13-111778
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DESIGNA TES APPROXIMA TE ON-BOA RD DA TE.
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DESIGNATES TIME ALLOTTED FOR INITIAL TRAINING. (THENUMBER REFERS TO THE COURSENUMBER SHOW" DESIGNA TES TIME ALLOTTED FOR ON-SITE TRAINING.
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BFS The Station Superintendent will serve as the chairman of the Test Working Group.
Members will be appointed as required to ensure competent representation on the TWG from the BFS project organization, B&V site organization, GE (NSSS) site organization, and the project field management organization.
Other parties may be assigned by the chairman for advice and consultation as appropriate.
Education and experience for the members of the TWG are as follows:
(1) For the Chairman refer to the Station Superintendent under subsection 13.1.3.2.2.
(2) Black & Veatch representative shall have a BS in Engineering or the Physical Sciences and four years power plant design or operation experience. He shall have general familiarity with the design of BFS which is the responsibility of B&V.
(3) General Electric representative shall have a BS in Engineering 13 or the Physical Sciences and four years power plant design, construction, or operations experience at least two of which shall be nuclear. He shall be familiar with the GE designed portion of the plant.
(4) BFS Project Engineering representative shall have a BS in Engineering or the Physical Sciences and four years power plant design or operations experience. He shall have general familiarity with the design of BFS.
(5) Proj ect field management member shall have a high school education plus four years power plant construction experience.
He shall be familiar with BFS construction procedures.
13.4-la 13-111778
BFS 14.0 INITIAL TESTS AND OPERATION 14.1 TEST PROGRAM (CESSAR) 14.1.1 Administrative Procedures (Testing)
The preoperational testing and initial startup programs for all safety-related systems and components will be coordinated by the station Test Working Group (TWG) and executed under the direction of the Station Manager as described in Subsection 13.4.2. Overall responsibility will rest on the Station Manager for the performance of the tests by the BFS staf f augmented as necessary by other PSO personnel.
The plant operating and engineering personnel will have the responsibility for the preparation of the preoperational and initial startup test procedures with technical direction as needed from B&V and GE or other consultants.
Scoping documents, prepared by B&V and GE, which describe the test objectives and acceptance criteria for each test, will be used in the development of detailed test procedures. The procedures will identify all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service. The procedures will insure that the 13 testing is done within the acceptance-rejection criteria for performance listed in the appropriate design documents. The procedures will assure that all prerequisites for the given test have been met, adequate instrumentation and equipment are available, and that the tests are performed under suitable environmental conditions adhering to approved test methods.
The test procedures are reviewed by TWG and the plant operating and engineering personnel to assure incorporation of scoping documents, establishment of test prerequisites and compliance with applicable regulatory requirements.
The appropriate station staff supervisors shall approve each procedure.
Final authorization for use shall be the responsibility of the Station Manager. Selected preoperational test procedures for system important to safety will be reviewed by the Review and Audit Committee to ascertain whether the procedures will satisfy the test objectives, meet acceptance criteria established by the system designer, require documentation sufficient to verify results, and establish base line data for use during plant life.
Execution of the test procedures will be carried out by the operating staff of PSO with technical direction as needed from GE and Black & Veatch 14.1-1 13-111778
BFS or other consultants. The test results are prepared in final form by the operating staff or designates and i viewed by the station staff supervisors to determine that test objectives have been satisfied and that the established acceptance criteria have been not. The prropriate system design organization will review the test results to determine that the system design objectives have been met.
The procedure approval and the test execution, review, and approval will be documented according to established QA procedures. The approved test procedure, results report, and data are retained at the station. These documents will be retrievaole throughout the plant life or until they are no longer required in accordance with ANSI N45.2.9-1974. Personnel qualifications 13 are discussed in Chapter 13 and subsection 14.1.6.7. Organization of the review committees is discussed in Chapters 13 and 16.
14.1.2 Administrative Procedures (Modifications)
The appropriate system design organization shall review all test-results for safety-related systems and components. If test results, test data, or system performance are unsatisfactory, the system design organization shall recommend appropriate corrective action.
If the deficiency is a result of equipment performance due to improper installation and does net require a design change, appropriate project field management shall assure the deficiency is corrected by the appropriate contractor. The documented corrective action will be reviewed and verified by an appropriate member of project field management, and the testing resumed.
Modifications which require a design or hardware change or which alter the intent of the procedure must be reviewed and approved by B&V or GE engineering support personnel as appropriate.
Modifications which require technical specification changes or concern an unreviewed safety item must be reviewed and approved by tne Review and Audit Committee prior to submittal to the NRC as required by 10 CFR 50.59.
Minor modifications to test procedures which do not change the intent of the procedure, do not require design or equipment changes, and clearly do not sacrifice plant or personnel safety may be made by the appropriate station staff supervisor with postapproval by the Station Manager.
The station Test Working Group shall serve to interpret the type of deficiency for other than minor modifications and identify responsibility for correction. All modifications to systems or procedures must be soproved by the Station Manager.
14.1-2 13-111778
BFS with station procedures which are formulated using the guidance of ANSI N18. 7-1972, " Administrative Controls for Nuclear Power Plants," as described in Subsection 13.5.1.
14.1.6.4 Utilization of Plant Operating and Testing Experiences at Other Reactor Facilities. During the preparation of detailed preoperational testing procedures and operation procedures licensee event reports for similar reactor plants, and operating experiences will be reviewed to identify potential problens. Precautionary measures will be factored into procedural 13 steps to preclude or minimize their occurrence. The schedule for this evaluation activity will be coincident with the schedule for preparation.
14.1.6.5 Test Program Schedule. The test program will follow the schedule indicated on Figure 14-1 and will follow the typical NSSS test sequence presented on Figure 14.1-1 (GESSAR). The scheduled time period for development of detailed test procedures will commence approximately 6 months prior to performing the initial tests which is consistant with the schedule for hiring and training of plant operating and engineering personnel presented on Figure 13.2-1. Staffing and training for the Black Fox Station is discussed in Chapter 13. Plant operating and emergency procedures will be developed during the initial testing phase concurrent with the preparation of pre-operational test procedures and the performance of preoperational tests.
14.1.6.6 Trial Use of Plant Operating and Emergency Procedures. During the preoperational testing program, the operating and emergency procedures for the plant will be placed in trial use, to the extent practical, to verify their adequacy end appropriater.ess. The experiences gained during the trial use ,
period will be factored into the final operating and emergency procedures as appropriate.
14.1.6.7 Staff for Conduct of Test Progran. As described in Subsection 14.1.1, PSO's staff will be augmented as needed with technical direction provided by B&V and GE or other consultants for the preparation of preoperation.
initial startup test procedures. Execution of the test procedures will oe carried out by the operating staff of PS0 with technical direction as needed 13 from CE & B&V or other consultants. Such technical direction will consist of providing scoping documents containing test objectives, performance requirements, and acceptance criteria and providing consultation as needed. Augmentation of 14.1-3a 13-111778
BFS the normal BFS operating staf f during the performance of the preoperational tests will be provided as needed by members of the PSO Power Generation staff as mentioned in subsection 13.4.2., B&V, GE and other consultants. The interrelationships, interfaces and the general qualifications of the participants, PSO, B&V, and CE, in the test program are described in this chapter and chapter 13. The schedule for augmenting PSO's staff will be consistent with the schedule for preparation and execution of the test procedutes.
The person in responsible charge of preparing preoperational test procedures shall have a bachelor of science degree in engineering or the physical sciences and four years experience in design and/or operation of power plants at least one of which will be in the preoperational phase of a nuclear plant test program. An additional four years of experience shall 13 be required without the degree. This individual may also be in responsible charge of the conduct of preoperational tests, but if another is selected that person will meet the same requirements.
The person (s) assigned to be in responsible charge of developing post-fuel loading test procedures and of conducting the post-fuel loading tests shall meet the "American National Standard for Selection and Training of Nuclear Power Plant Personnel" ANS 3.1-1978 for the responsible person in charge of reactor engineering.
For both the preoperational testing and startup testing phases, personnel developing procedures and supervising the conduct of tests shall be qualified using a program which will comply with " Qualifications of Inspection, Examinations, and Testing Personnel for the Construction Phase of Nuclear Power Plants", ANSI N45.2.6-1973.
14-1-3b 13-111778
TEST PROGRAM SCHEDULE
- FIGURE 14-1 FUEL
-24 MOS. - 18 MO S. -12 MOS. -6 MOS. LOAD +6 NOS.
PREPARE ADMIN. AND PERFORM COMPONENT
, COMPONENT TEST PROCEDut,3 IESTS I
Ah0 US PERFORM PRE-0P TESTS FUEL LOAD 2 PRECRITICAL, INITIAL CRIT.
POWER ASCEN SION
< PRE-OP PHASE ; ; STARTUP TEST PHASE *
- AFPLIES TO UNIT ONE. THE DURATION OF PROCEDURE PREPARATION FOR UNIT TWO IS EXPECTED TO BE SHORTER.
BFS Question 413.3A .
Question: We requested in Question 413.3 additional clarifying information on the responsibilities of organizations participating in the development, conduct and review of test results for each phase of the initial test program. Additionally, we requested that minimum educational and experience requirements for key personnel participating in *he initial test program be described. To date, this information has not been provided. We will require that sufficient information be provided to demonstrate that adequate plans and minimum requirements have been established to assure that adequate testing will be conducted.
Response: The Applicant is fully committed to properly performing aaa staffing an initial test program - including the preoperational, fuel load, low power testing, and power ascension phases. The Applicant is also full; cognizant of the need for early planning for the initial test program to avoid staffing and procedural problems.
Preliminary plans for conducting the initial test program are described in Section 14.1 of the Black Fox Station PSAR.
Commitments to Regulatory Guide 1.68 and ANSI N18.7-1972 are described in GESSAR NSSS Docket STN 50-550 which is referenced by PSAR Section 14.1.
The Applicant further commits to provide the remainder of the information concerning personnel qualifications in the latter part of 1978.
The detailed description of the initial tes: program will 13 be included in the FSAR. PSO excepts submittal of the FSAR to be at least 24 months prior to the date scheduled for loading fuel in Unit 1.
413-9 13-111778
BFS ANALYSIS OF LPCI DIVERSION EFFECTS ON ECCS PERFORMANCE FOR BLACK FOX STATION UNITS 1 AND 2 NOVEMBER 2, 1978 DOCKETS STN 50-556 STN 50-557 13-111778
BFS This report was previously transmitted to the Nuclear Regulatory Commission by letter from Public Service Company of Oklahoma on November 10, 1978. Two telephone conversations, November 16 and November 17, have taken place between the NRC Staff and PSO relative to information in the report and information requested in the NRC lette- dated October 11, 1978. This report incorporated in the PSAR as part of t_endment 13 incorporates one correction on page two, paragraph number ', where the previous word " increases" has been changed to " decreases". Public Service Company of Oklahoma will respond fully to the balance of the NRC Staff concerns on this matter in PSAR Amendment 14.
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BFS Purpose This analysis was performed to investigate the effect on the ECCS analysis for the Black Fox Nuclear Power Station of diverting low pressure coolant injection (LPCI) pumps to the containment spray mode ten minutes after a loss-of-coolant accident (LOCA) initiation.
Automatic diversion of LPCI flow to containment spray has been provided in response to an NRC requirement to assure containment integrity for postulated high steam flow bypassing the suppression pool. Such flow diversion would occur only if a high containment pressure (>9 psig) signal is present after ten minutes. The assumption of sufficient bypassing to cause such a pressure has been shown by CE to be extremely conservative
_nd unrealistic.1 Conclusion The results show that the worst single failure / break type combination is the high pressure core spray (HPCS) line break (approximately .02 ft )
assuming the failure of the low pressure and core spray (LPCS) diesel generator (D/G) which powers one LPCS pump and one LPCI pump. This single failure / break type combinat'on yields the highest peak cladding temperature (approximately 1985 F) o' '1 the cases affected by LPCI diversiv.: at ten minutes. The peak c'. adding temperatures experienced by the cases affected by LPCI diversio- re below the limits established in 10 CFR 50.46 (2200 F).
This temperate s also below the peak clad temperature (PCT) calculated for the break of a recirculation line (2038 F) which is not adversely affected by LPCI diversion at ten minutes.
Assumptions (1) A maximum of two LPCI pumps (specifically LPCI "A" and LPCI "B")
can be fully diverted at ten minutes to the containment spray mode. (NOTE: LPCI "A" shares an emergency diesel generator with the LPCS: LPCI "B" and "C" share an emergency diesel generator.
The pump associated with LPCI "C" cannot be diverted to containment sprays.)
1 NEDO-10977 Drywell Integrity 'udy:
. Investigation of Potential Cracking in BWR/6 Mark III Containment.
1 13-111778
BFS (2) The standard FSAR assumption of one automatic depressurization system (ADS) valve failure combined with the worst additional single failure was retained because this assumption is built into the present model. This bounding assumption yields conservatively higher calculated peak cladding temperatures (PCTs) by approxi-mately 100 F. The PCT reported on Page 1 does not include this
. assumption.
(3) Approved Appendix K analysis models were used, except that some LPCI flow to the reactor vessel was stopped ten minutes after the accident.
General Observations from the Analyses only those accident cases which are not reflooded to the hot node before ten minutes are affected by the assumed LPCI diversion. Once the
. core has been reflooded, only one ECCS pump is necessary to keep the core covered. Thus, the breaks affected include small breaks less than approximately 0.2 ft (depending on the break location) and outside steam line breaks (OSLB). The effect of the assumed LPCI diversion on the OSLB is small and is discussed in a later section of this report.
Af ter reviewing the effect of diversion on the rest of the small breaks, general statements can be made to describe the results in the area of interest:
- 1. The calculated PCTs (no LPCI diversion) in the small break regions affected by LPCI diversion generally decrease with decreasing break size. This follows from the fact that the core is uncovered for shorter periods and that the decay heat is lower at the time of uncovery as the break size decreases.
- 2. The maximum temperature for the assumed LPCI diversion case for any given break location occurs at approximately that break size where the LPCI system would normally inject flow into the vr el starting at 600 seconds (i.e. the assumed LPCI diversion time).
Bigger breaks get some reflooding benefit from the LPCI pumps before diversion. Smaller breaks have the same ECC systems available as this maximum break, but the smaller break area has 2 13-111778
BFS a lower calcularca PCT, as discussed previously. As an exa aple, this worst break is ind' ated on Figure 1. A longer LPCI diversion time muld have correspo.adingly smaller breaks where the maximum temperature would occur and hence lower calculated PCT.
- 3. Diverting LPCI from its ECCS flooding function does not always result in higher PCTs. When compared to no LPCI diversion, a reduction in PCT can be observed as a result of diverting LPCI if the LPCS is available. The reduction of subcooled LPCI water results in a reflooding mixture (due largely to LPCS flow) of steam and water which has higher voids. Thus, in the case where little LPCI flow is available for re looding, even though less ECCS flow is entering the vessel, the swollen level inside the lower plenum is higher and reflooding can occur sooner. In such cases the calculated PCTs are extremely low and changes in PCT in either direction are insignificant.
- 4. Because this investigation is primarily concerned with small breaks, the failure of the HPCS, for non-core spray line breaks, is the worst single failure for this study. If the HPCS were operable, the break sizes being analyzed would reflood earlier than ten minutes with the very small break sizes never uncovering.
he following break locations were considered: A) core spray line, B) recirculation line, C0 feedwater line, D) the steam line, and E) LPCI line. A brief summary of each analysis is provided below.
A. Core Spray Line Break (HPCS Line) - It is conservatively assumed that no flow enters the vessel through the broken line independent of the break size. For this case, the failure of the diesel generator associated with LPCS and LPCI "A" is the worst single failure since all credit for core spray cooling is eliminated.
The ECC systems remaining before diversions are 2 LPCI + ADS after diversion at ten minutes. Because in both cases the reflooding the vessel, there is a longer ~eflooding time associated with the diverted case with reduced ECCS flow. The results of this investigation are shown in Figure 1. Because the temperature 3 13-111778
BFS increase from the non-divered case is a result of a loss of reflooding flow from 1 LPCI pump, intermediate cases (loss of part of the flow) will experience intermediate (lower) temperature increases.
This particular failure / break type combination was the most adversely affected by the assumed LPCI diversion. However, the peak cladding temperatures are still below the limit of 2200 F.
B. .eirculation Line Break - For this break, the worst single failure is the HPCS failure, as described previously. The ECCS remaining before diversion are 3 LPCI + LPCS + ADS and, after diversion, 1 LPCI + LPCS + ADS. Since in the diverted case the remaining LPCI flow is not enough to significantly quench the voids in the lower plenum, the mixture in the lower plenum will reflood with a higher voided mixture. This higher void fraction for the diverted case more than offsets the reduction in ECCS flow entering the vessel due to this div .... of LPCI. Hence, there is a net reduction in PCT due to a shorter reflooding time and the recirculation line break without diversion which has already been reported is bounding relative to a line break with diversion.
A representative break (.01 ft ) was analyzed which confirmed these results. The results of this investigation are shown in Table 1. Intermediate cases (diversion of less than the full flow from two pumps) should result in smaller temperr.ure decreases.
C. & D. Feedwater and Steam Line Breaks - For these breaks, the worst single failure is the HPCS failure, as described previously.
The ECCS remaining before diversion are 3 LPCI + LPCS + ADS and after diversion 1 iPCI . LPCS + ADS. For the diverted case, there will be a reduct!.on in calculated PCT for the same reasons discussed for :he recirculation line break. A representa-tive break (i.e. 01 fc ) was again analyzed which confirmed the anticipated results. The results of this investigation are shown 4 13-111778
BFS in Table 1. For both cases, insignificant decreases in calculated PCT result from LPCI diversion. Tae outside (isolated) steam line break was also considered with similar results.
E. LPCI Line Break - As in the case of the core spray line break, it is conservatively assumed that no flow enters the vessel through the broken line independent of the size. For this break, the worst single failure is the HPCS failure, as described previously. The ECCS remaining before diversion are 2 LPCI +
LPCS + ADS ano, tfter diversion, LPCS + ADS (if the break is in line "C") or LPCS + LPCI + ADS (if the break is in lire "A"/ or "B") .
In either case there is insuffic'-nt LPCI flow to significantli quench the voids in the lower rienum. Therefore, the core will reflood with a voided mixture. This higher void fraction more than offsets the reduction in ECCS flow entering the vessel due to diversion of LPCI. Hence, there is a net reduction in PCT due to a shorter reflooding time.
As above, the .01 ft break was analyzed which confirmed the anticipated results. The results of both diverted cases are shown in Table 1.
Response to Question (1)
The system provided for diversion of LPCI flow is a safety grade system.
Consequently, it has a high reliability in performing its intended function.
Postulation of a failure of this system to perform its function in combination with another single failure is not required under CDC 35 or 10 CFR 50.46.
Response to Question on Operator Action The operation of the ECC systems including diversion of LPCI to contain-ment sprays requires no operator action for at least 10 minutes following accident initiation. Ten minutes is the present licensing basis for operator manual action time following automatic actuation of the ECC system.
There is no requirement either in 10CFR50.46 or GDC 35 for assuming no operator action 20 minutes after the initiation of the accident. Ten minutes continues to be the licensing basis used and supported by General Electric.
It is also the basis for the containment- performance evaluation as it has been for other BWR plants.
5 13-111778
TABLE 1 THE EFFECT ON THE PCT OF DIVERTING LPCI FLOW AT 10 MINUTES FOR VARIOUS .01 FT2 BREAK TYPES PCT PCT Break Type No Diversion With Diversion Recirculation Line 948 F 877 F Feedwater Line 917 F 836 F Inside Steam Line 920 F 831 F LPCI Line 834 F 804 F(
964 F( }
NOTE: (1) PCT if break occurs in LPCI line "A" or "B" (2) PCT if break occurs in LPCI line "C"
(
6 13-111778
2000 REFLOODING TIME <10 MINUTES 2000 -
/p \ '
/ \
mm.m. # \
E, un g 3500 -
4 m
E 1
w DI in h
g 3000 -
BREAK SIZE FOR WHICH LPCI INJECTS
$ FLOW INTO THE VESSEL AT 10 MINUTES U
k BASE CASE (NO DIVERSION)
E
- ==== ammme i LPCI REMAINING AFTER 1r DIVERSION AT to MINUTES Soo i I I I I I o
0.02 0.05 0.1 0.2 0.3 g o.002 0.005 0.0 t BREAK AREA (f t2)
O
$ Figure 1. Peak Cladding Tenperature versus Break Area for an HPCS Line Break Assuming an LPCS Diesel-Generator Failure. Systems Remaining: 2 LPCI + ADS
60 -
40 -
w J
TAF b
a M m 5 M YJ ~
BAF g
0 I I l 0 3m 600 1200 1500 C TIME isec)
L C
D Figure 4a. Water Level Inside the Shroud Following a HIGH PRESSURE Core Spray Line Break, LPCS DG Failure, Break Area = 0.02 ft2 (SBM).
1200 -
_i .00 _
ua 5
a w
E d
N w w e > ns t/2 400 l l l O 300 000 900 1200 1500 m
u TIME (sec) i Z
C Figure 4b. Reactor Vessel Pressure Following a IIIGli PRESSURE Core Spray 5 Line Break, LPCS DC Failure; Break Area = 0.02 ft2 (SBM).
0 E
N.o t
4 3
E e
9 i
8 U
O ,
2 a
i 10,000 25 I l l l 0 300 000 900 1200 1500 C
b TIME (sec)
C Z Figure 4c. Convective lleat Transfer Coefficient Following a llIGil PRESSURE Core Spray Li.ne Break, LPCS DG Failure, Break Area = 0.02 ft2 (SBM).
3000 -
E us 3 2000 -
4 5
t 9
0 a !a
=
e d
4 som -
t 8
l I 0
0 600 900 1200 1500 C TIME (setl
,L Z Figure 4d. Peak Cladding Temperature Following a llIGli PRESSURE Core Spray
2000 t
i e
D a
S
+ i000 -
5 .
~ *i N w *
- l w 0 t l l
$ 0 500 1000 1500 TIME (sec)
N Figure 4e. LPCI Flow Rate Following a llIGli PRESSURE Core Spray Line Break, LPCS DG Failure, Break Area = 0.02 ft2 (SBM).
BFS GEN ER AL @ ELECTRIC NUCLEAR ENERGY PROJECTS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MC 392, (408) 925-3217 November 14, 1978 DIN # 6212 DIN 8-013-528 FILE: 6212.312.5560.21L Mr. T. N. Ewing Project Manager Black Fox Station Nuclear Project Public Service Company of Oklahoma P. O. Box 201 Tulsa, OK 74102 Attention: Mr. Vaughn Conrad Gentlemen:
SUBJECT:
CALCULATED PEAK CLADDING TEMPERATURE (PCT) FROM ECCS ANALYSIS FOLLOWING A DESIGN BASIS ACCIDENT (DBA) FOR BLACK FOX PLANT '
It has become apparent recently that there is some confusion as to what the appropriate PCT following a DBA at Black Fox should be. This letter describes the changes which have occurred in the PCT from the issuance of 238 NSSS GESSAR which is referenced by Black Fox, to the present, along with the reason (s) for each change.
Table 1 illustrates the ghronology of changes to the PCT starting with the GESSAR value gf 2180 F and ending with the current value for Black Fox which is 2038 F. The reason (s) for the changes are also noted. It is of interest to note that the trend in the PCT since GESSAR due to these changes has been decreasing. It should also be noted that the recent ECCS analysis with low pressure coolant injection (LPCI) diversion does not change the current limiting calculated PCT of 2038 F for Black Fox.
Sincerely QA. J. Levine, Manager Project Licensing Unit 1 AJL:daj/873 cc: Dr. M. J. Robinson L. R. Cannard 13 '
13-111778
BFS CHANGES IN PEAK CLADDING TEMPERATURE (PCT)
SINCE GESSAR PCT (*F) Reason for Change 2180 GESSAR Table 6.3-3 2129 1. Change from REFLOOD 03 to REFLOOD 04 to incorpor-ate a correction to the calculation of vapor flow split between the fuel bundles and jet pumps.
2042 1. Correct the water level setpoint in SAFE which signals initiation of the high pressure core spray system.
- 2. Modify the core power in REFLOOD to be consistent with requirements of 10CFR50 Appendix K (102% licensed power level)
- 3. Correct initial core average quality to improve code accuracy in representing bubble rise velocity.
- 4. Correct recirculation line break area for the suction end of the recirculation suction line break in SAFE calculation. The limiting flow area is based on the vessel nozzle safe end inside diameter.
- 5. The guide tube thermal resistance was increased a factor of 10 due to deciaml point error.
2038 1. Implement additional 1 psi core pressure drop in REFLOOD as required by the NRC.
- 2. Use of more accurate pressure table, h 14 13-111778