ML20077K451
| ML20077K451 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/25/1991 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | Fitzpatrick J KANSAS, STATE OF |
| References | |
| NUDOCS 9108080042 | |
| Download: ML20077K451 (486) | |
Text
{{#Wiki_filter:. h* KANSAS QAS AND ELECTRIC COAIPANY
#c7Ui.U.?I$I5n July 25, 1991 Mr. Jim Fitzpatrick Internal Revenue Service 412 S. Main Wichita, Kansas 67202
Dear Mr. Fitzpatrick:
We received a copy of your June 21, 1991 letter to Mr. Robert Martin, Region IV Administrator, of the Nuclear Regulatory Commission (NRC) on July 17, 1991. We are very disappointed in the letter and disagree factually with it as well. A. We are reasonably certain that the public document rooms in Topeka and Emporia contain all the documents that are required to be located at each location. I am certain if we had been asked to accompany your agents to locate the documents, we would have been able to locate the necessary documents at either of the locations. B. We have always provided all documents requested by you and your agency. In fact, the FSAR, was first requested on March 1, 1990 and ! provided March 2, 1990. It was requested several times during the agents field work and was provided each time. In fact, the FSAR was available on a daily basis to your agents. Items 1, 2 and 3 were also provided to the KCP&L audit team in response to data request 144, requested by Jim Jakes, dated June 26, 1991. Although the request did not specifically ask for items 1, 2 and 3 we felt they were responsive to the information requested. C. Items 4 and 5 were not requested until our meeting on July 12, 1991. I D. Please provide copies of the IDR's you state were given to us for items listed 1 through 7 in your letter and the agent and date requested so that we can verify our records that those items were either unavailable or that we refused to give them to you or your agents. Please provide these by July 31, 1991. If no IDR's were given to us for items 1 through 7, but some other form of request was made, please provide the documentation for the request. To expedite the completion of your examination I am providing the following information:
- 1. A copy of Chapter 14 of the Updated Safety Analysis Report (USAR).
- 2. A copy of the initial Startup Report (not previcusly requested).
910000004p 9jo7a5 PDR ADOCK 05000402 d N ( v. j) 120 E. &st - McNto, Kansas - Md! Address P.O. Box 208 / McNta. Kancos 67201 - Tolophone: Area Code (316) 261-6390 s
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- 3. Chapter 14 of the USAR included as item 1 is Wolf Creek's res)onse to the various NRC-regulations regarding testing of Wolf Cree <
Generating Station. This was available to your agents numerous times during their field work.
- 4. I am enclosing three documents in response to item 4; NPF 42 is the NRC full power license, the affidavit by KG&E as to the time Wolf Creek Generating Station met the criteria established by the Kansas CorporationCommission(KCC)forcommercialoperation'oftheplant and the KCC order setting the four "in service" criteria.
- 5. Enclosed is a copy of NRC Facility Operating License NPF 32, the low power license which restricts reactor power levels to less than 5 percent of full power.
- 6. The milestone schedule, not previously requested, is in Chapter 13 of the USAR. 13.1.1.1 is attached for your ir. formation.
- 7. There are numerous NRC regulations which address training that is required to be performed by >ower plant licensees. We have run a computer search of our data >ase of the 10 CFR (Code of Federal Regulations) keying-on the word training. The attached schedule, not previously requested, reflects the results of that search. We assume you have 10 CFR available if you intend to reference each item on the listing. If-you do not have 10 CFR available to you, we will make one available for your use.
We are very disturbed by the accusations included in your letter. We are certain that we have provided all of the documents that have been requested. However, we are willing to research any documentation you can provide which_would indicate we have not complied with requests for this information. Sincerj!1y, r
)
I 7 . /- , JTC/hf Attachments cc: Mr.RobertMartin(NRC) Mr. Doug Pickett (NRC) Ms. Carol Ann Reed (NRC) John Bailey (WCNOC) w/o attachment Kendall Coyne KCP&L) Dave Jacobsen Reid & Priest) w/o attachment Mr. Jim Jakes IRS) w/o attachment
WOLF CREEK ( )+ CHAPTER 14.0 l TABLE OF CONTENTS s INITIAL TEST PROGRAM l Section ,T.Lt_l e Pace 14.1 SPECIFIC INFORMATION TO BE INCLUDED 14.1-1 IN THE PSAR 14.2 INITIAL TEST PROGRAM 14.2-1 14.2.1
SUMMARY
OF TEST PROGRAM AND OBJECTIVES 14.2-1 14.2.1.1 Preoperational Test Program 14.2-1 14.2.1.2 Initial Startup Test Program 14.2-2 14.2.2 ORGANIZATION AND STAFFING 14.2-3 14.2.2.1 General Descrijtion 14.2-3 14.2.2.2 Startup Organization 14.2-4 14.2.2.3 Operating Staff 14.2-6 14.2.2.4 Major Participating Organizations 14.2-6 14.2.2.5 Quality Assurance 14.2-8 14.2.2.6 Qualifications of Key Personnel 14.2-8 I cs 14.2.3 TEST PROCEDURES (,) 14.2.3.1 Startup Test Procedures 14.2-8 14.2-8 14.2.3.2 Procedure Review and Approval 14.2-10 14.2.4 CONDUCT OF TEST PROGRAM 14.2-12 14.2.4.1 Administrative Procedures 14.2-12 14.2.4.2 Turnover From Construction to KG&E 14.2-13 Startup 14.2.4.3 Component and Prerequisite Testing 14,2-14 14.2.4.4 Preoperational Testing 14.2-14 14.2.4.5 Initial Startup Testing 14.2-14 14.2.4.6 Test Prerequisites 14.2-15 14.2.4.7 Test Evaluation 14.2-15 14.2.4.8 Design Modifications 14.2-15 14.2.5 REVIEW, EVALUATION, AND APPROVAL OF 14.2-15 TEST RESULTS 14.2.6 TEST RECORDS 14.2-16 14.2.7 CONFORMANCE OF TEST PROGRAMS WITH 14.2-16 REGULATORY GUIDES 14.2.8 UTILIZATION OF REACTOR OPERATING AND 14.2-16 TESTING EXPERIENCE IN DEVELOPMENT ,3 OF TEST PROGRAMS + > 14.2.9 TRIAL USE OF PLANT OPERATING AND EMERGENCY 14.2-18 PROCEDURES 14.0-1 Rev. 1 l
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} , ; , ,) TABLE OF CON'.'ENTS (CONTINUED)
Section Title Page 14.2.10 INITIAL FUEL LOADING CRITICALITY AND POWER 14.2-18 ASCENSION 14.2.10.1 Fuel Loading 14.2-19 14.2.10.2 Initial Criticality 14.2-22 14.2.10.3 Low Power Testing 14.2-23 14.2.10.4 Power Level Ascension - 14.2-23 14.2.11 TEST PROGRAM SCHEDULE 14.2-24 14.2.12 INDIVIDUAL TEST DESCRIPTIONS 14.2-25 14.2.12.1 Safety-Related Preoperational Test 14.2-25 Procedures 14.2.12.2 Nonsafety-related Preoperational Test 14.2-120 Procedures 14.2.12.3 Startup Test Procedures 14.2-156 9 4 14.0-11 Rev. O
WOLT CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 14.2-1 Safety-Related Preoperational Test Procedures 14.2-2 Non-Safety-Related Preoperational Test Procedures 14.2-3 Initial Startup Test
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i t l l l r% Q* , 14.0-111 Rev. O
WOLF CREEK TABLE OF CONTENTS (CONTINUED) 9 LIST OF FIGURES Figure No. Title 14.2-1 Deleted O 1 i l l l i l l l l l l l 14.0-iv Rev. O
WOLF CREEK e . ,~3 ' ls_._) CHAPTER 14.O I I' CI INITIAL TEST PROGRAM 14.1 SPECIFIC INFORMATION TO BE INCLUDED IN PRELIMINARY SAFETY ANALYSIS REPORTS This section is not applicable to an USAR. (m. () 14.1-1 Rev. 1 l
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'() f *H APTER 14.0 INITIAL TEST PROGRAM 14.2 INITIAL TEST PROGRAM 14.2.1
SUMMARY
OF TEST PROGRAM AND OBJECTIVES The Initial Test Program encompassed the scope of events following completion of construction and construction-related inspections and tests and terminating with Power Ascension Testing. The Initial Test Program was conducted in two separate and sequential subprograms: the Preoperational Test Program and the Initial Startup Test Program. At the conclusion of these subprograms, the plant was ready for normal power operation. Testing during the Initial Test Program was accomplished in four sequential phases: Preoperational Test Program Phase I - Preoperational Testing Initial Startup Test Program Phase II - Initial Fuel Loading and Zero Power Testing _ Phase III - Low Power Physics Testing Phase IV - Power Ascension Testing Prior to preoperational testing of a particular system, certain prerequisite and construction tests were conducted in order to verify the integrity, proper installation, cleanliness, and func-tional operability of the system components. 14.2.1.1 Preoperational Test Program The Preoperational Test Program is defined as that part of the Initial Test Program that commences with the completion of con-struction and construction-related inspections and tests and terminates with commencement of nuclear fuel loading. The Preoperational Test Program included botn safety-related and nonsafety-related preoperational tests. The Preoperational Test Program used a graded approach to determine the extent of testing to be performed. The safety-related preoperational tests (Table 14.2-1) demonstrated the capability of safety-related structures, systems, and components to meet performance requirements and to satisfy design criteria. The nontafety-related preoperational tests (Table 14. 2-2) were conducted on nonsafety-related systems and components to satisfy reliability and availability. Preopera-tional tests were conducted on those systems that: p
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- gh Aka) rell'ed"upon for safe shutdown and cooldown reactor under normal plant conditions and for maintain-of the h ing the reactor in a safe c ond i '. i on for an extended shutdown period;
- b. Are relied upon reactor under transient for safe and shutdown postulatedandaccident cooldown of the condi-tions and for maintaining the reactor in a safe condi-tion for an extended shutdown period following conditions; such c.
Are relied upon for establishing conformance with safety limits or limiting conditions for operations that are included in the technical specifications;
- d. Are classified as engineered safety features actuation systems or are relied upon to support or ensure opera-tion of engineered safety within design limits; features actuation systems
- e. Are assumed to function during an accident or for which credit is taken in the accident analysis;
- f. Are used to process, store, control, or limit the release of radioactive materials.
The objectives of the Preoperational Test Program were to:
- a. Verify that plant components lll and systems, including alarms and indications, are constructed and fulfill their design intent;
- b. Demonstrate, to the extent practicable, proper system / component response to postulated accidents; c.
Familiarize plant staff operating, technical, and main-tenance personnel with plant operation. The completion of preoperational testing constituted the comple-tion of Phase I of the Initial Test Program. 14.2.1.2 Initial Startup Test Program The Initial Startup Test Program is defined as that part of the Initial Test loading and Program that commences with the start of nuclear fuel terminates with the completion of power ascension testing. The initial startup tests (Table 14. 2-3) ensured that fuel loading was accomplished in a safe manner, confirmed the 14.2-2 Rev. O O
l WOLF CREEK ( design and ates-basis, demonstrated, where practical, that the_ plant oper-responds properly to anticipated postulated accidents, transients and and ensured that the plant can be safely brought to rated capacity and sustained power operation. ' r The objectives of the Initial Startup Test Program were tot
- a. Accomplish a controlled, orderly, and safe initial core loading;
- b. Accomplish a controlled, orderly, and safe initial criticality; i
- c. Conduct low power testing sufficient to ensure that design-parameters are satisfied and safety analysis assumptions are conservative;
- d. Perform a controlled, orderly, and safe power ascension with testing terminating at plant rated conditions;
- e. Provide sufficient testing of transient and accident conditions to verify safe operation during transient accident conditions. or The completion of initial startup testing constituted th'e comple-tion of Phases II, III, and IV of the Initial Test Program.
(~} 14.2.2 ORGANIZATION AND STAFFING 14.2.2.1 General Description-The Operating Agent, as defined in Section 1.4, was responsible for the overall startup administration program. In and technical direction of the WCGS recognition of this responsibility, the Director of Nuclear Operations, under the direction of the Vice President - Nuclear, established a startup organization c7 ordinate and to
-implementation direct the comprehensive planning, development,
( and performance of the cest program. The Startup Organization was headed by the Startup Manager who reported to the L Plant Manager both admin;stratively and technically. During the preoperational startup program, the Startup Manager acted to coordinate activities between the Startup the construction organization, staff, and tne operating staff. Prior to commencing L preoperational testing activities, a Joint Test Group (JTC) as described in Section 14.2.3.2.2 was formed review and recommend to dures, for approval startup administrative preoperational test procedures, and preoperational test proce-14.2-3 Rev. 0 l
UOLF CREEK results. A Plant Safety Review Conmittee (PCRC) as described in g Section 14.2.3.2.3 was organized with the Plant Manager acting as W chairman and it reviewed and recommended for approval initial startup test procedures and results. 14.2.2.2 Startup Organization The Startup Organization was directly responsible for the conduct of the WCGS preoperational test program. The duties and responsibilities of the startup organization also included:
- a. Familiarization of support personnel with specific tests,
- b. Direction to support personnel and others during per-formance of tests including approprie.te interface with station operators.
- c. Authority to disallow or terminate testing due to condi-tions which could endanger personnel or equipment.
- d. Identification of deficiencies that could adversely affect test performance,
- e. Assembly of test data and preparation of ' test reports for evaluation of test results by others.
The Startup Organization was composed of system startup engineers, technicians, planners, craft labor, and other support personnel. lll The Operating Agent provided these personnel and used contractors to supply manpower for those positions that it could not staff. The staffing level for the Startup Organization increased as the test program progressed and construction activities decreased. Typical schedules for the test program are given in Section 14.2.11. Staffing and training of perscnnel involved in testing at WCGS were planned to provide sufficient manpower to support the testing schedule. l The Startup Organization reported administratively and technically to the Startup Manager; the duties performed by key individuals within the Startup Organization are summarized below. l 14.2.2.2.1 Startup Manager l The Startup Manager had the authority and responsibility, as l delegated by the Plant Manager, for the overall direction and i administration of the functions and activities required to conduct l the Startup Program. The responsibilities and duties of the i Startup Manager also included: 1 O 14.2-4 Rev. O
WOLF CREEK f~ . a. (T/ Development of plans and schedules regarding the status Of the startup program.
- b. Review and approval of administrative and technical test procedures and results,
- c. Continuing analysis of construction and equipment in-stallation schedules for compatibility with testing schedules and recommendations for corrective actions to minimize conflict.
- d. Review and submittal of design related problems requiring engineering resolution,- encountered by the Startup organization in accordance with the appropriate Startup Administrative procedures,
- e. Maintaining liaison with all organizations supporting Startup and coordinating their activities.
14.2.2.2.2 Startup Section The Startup Section was comprised primarily of the system Test Group, the Electrical Test Group and the Instrumentation and Control ' Group which had primary responsibility within the Startup Organization'to perform testing. This section also reviewed and recommended the acceptance of system or subsystem turnover docu-mentation from-Construction and coordinated system turnover and () any subsequent system rework. It was responsible for. preparing the test procedures, conducting the tests, and reporting the test results. For preoperational testing, this section documented the test results and presented them before the Joint Test Group for its review and recommendation for approval. 14.2.2.2.3 operations Technical Support Section The Operations Technical Support Section was responsible for providing technical support to the Startup Section-during testing. Thefareas-in which this support was given were instrumentation and control, chemistry, computer, health physics and reactor engineering. This- section is also a permanent part of the WCGS operating staff and so is also involved in training, procedure preparation, and general preparation for support of plant operations. 14.2.2.2.4 Startup Scheduling Section The Startup Scheduling Section prepared and updated the Startup s Schedule, utilizing input from cognizant system startup engineers and the construction organizations. gg l \/ 14.2-5 Rev. 0
WOLF CREEK 14.2.2.2.5 Quality Control Section The Quality Quality ControlControl Section formulated acd i;.si mc ateu the Startup Program. This program monitorod the conduct of the Startup istrative and Organization's technical test testing activitier by reviewing admin-procedures, oy witnessing ma]or ovolutions and selected flushes, hydros, and preoperational tests and by reviewing turnover packages. The Quality Control Section was under the direction of the Director - Quality. They provided support to the Startup Manager. 14.2.2.2.6 Startup Technical Support Section The Startup Technical Support Section was responsible far pro-viding technical support conduct of the Startup Program. to the Startup organization during the Their responsibilities included test procedure and test results review and approval, planning of major milestone activities, technical startup organization training and startup program compliance to FSAR commitments. 14.2.2.3 operating Staff The WCGS operating staff was involved in the startup program in several capacities throughout preoperational and initial startup testing. This involvement included review of test procedures and results
- ng and the direct participation in test activities.
staff Operat-personnel were utilized by the startup organization as required for performance of testing under the direction of system startup engineers. lll Station operators engineers in performing tests and in the assisted system startup routine operations of systems. The operating staff directed the fuel loading and was responsible for plant operation during initial startup testing. The operating staff was divided into sections headed Superintendent of Operations, by the Superintendent of Maintenance, Superintendent of Plant Support, Superintendent of Technical Support, Nuclear Training Manager and Superintendent R:gulatory Quality and Administrative Services. These section superinten-dents reported administratively and technically to the Plant Manager. The duties and responsibilities of the operating staff during plant operations are described in Chapter 13.0. 14.2.2.4 Major Participating Organirations 14.2.2.4.1 Bechtel Bechtel provided engineering input into the Bechtel was contacted to provide personnel experienced startup program. plant startup to augment in nuclear the startup organization for WCGS. Bechtel employees were assigned consistent program schedules. with the startup 14.2-6 Rev. 0 O l l
WOLP CREEK (')T 14.2.2.4.2 Daniel International Corporation (DIC) DIC, as contractor for WCGS, was responsible for the construction completion, and orderly release of components and turnover of systems to KG&E consistent with the startup program This responsibility included: schedules.
- a. Certification that documentation for components, systems and structures, as required by purchase and installation specifications, is -complete and available; and the maintenance of these certification files which provide the documentary evidence, and
- b. Provision of dedicated craft manpower support as re-quired for performance of the startup program.
14.2.2.4.3 Westinghouse Electric Corporation Westinghouse, as the Nuclear Steam-Supply System (NSSS) supplier, was responsible for providing technical assistance to KG&E during preoperational and initial startup testing performed on the NSSS equipment and systems. nical guidance, Technical assistance is defined as tech-advice and installation, and testing practices.counsel based on current engineering, assigned consistent with the Westinghouse employees were Startup Program schedules. This responsibility _ included: ()
- a. Assignment of personnel to provide advice and assistance to KG&E for test and operation of all equipment and systems in the Westinghouse area of responsibility.
- b. Supportive engineering services, including special
; assistance during the initial fuel loading.
- c. Providing test procedure outlines and technical assist-ance for tests of Westinghouse furnished components and systems.
14.2.2.4.4 General Electric (GE) GE is the supplier and installer of the turbine generatot. GE supplied technical support for the startup and testing of the turbine generator. Some of the prerequisite testing (i.e., tur-bine oil flush) was performed by the GE personnel, GE has supplied recommended procedures for starting, operating, and shutting down equipment in their technical manuals for the turbine generator. 14.2-7 Rev. O
WOLF CREEK 14.2.2.5 Quality Assurance khk The KGLE Quality Branch was responsible for assuring of the quality construction, plant testing, and operations activities in cecordance with the WCGS Quality Program which is described in Chapter 17.2. 14.2.2.6 Qualifications of Key Personnel The qualifications for key plant operating personnel are described in Chapter 13.0. The qualification requirements for startup personnel involved in the WCGS startup program conformed to capability levels per ANSI N45.2.6 and Regulatory Guide 1.8 recommendations. All test personnel were indoctrinated An the startup administra-tive procedures, methods and controls. 14.2.3 TEST PROCEDURES The Initial Test Program was conducted in accordance with detailed preoperational and initial startup test procedures. KG&E main-tained the overall responsibility for test procedure preparation, review, and approval during the preparational stages. responsible KG&E was for final procedure revision, review, and approval. These activities were completed in a timely fashion to ensure that the ggg approved procedures for satisfying FSAR testing equipment commitments were available for review approximately 60 days- prior to scheduled implementation or fuel load for preoperational and initial startup tests, respectively. Preoperational and initial start-up testing commitments not available for review approxi-mately 60 days prior to scheduled implementation or fuel load, respectively, were handled on a case-by-case basis. The following sections describe the general methods employed to control procedure development and review, and they also describe the responsibilities of the various organirations which partici-pated in this process. The detailed controls and methods were described in the startup administrative procedures. 14.2.3.1 Procedure Preparation Tast procedures for the powerblock systems and components were developed by Westinghouse and Bechtel. Bechtel also prepared test procedures for the site safety-related systems and components. Test procedures for the site nonsafety-related systems and com-ponents were developed by various entities as acordinated by KG&E. 14.2-8 Rev. O O
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- (-) The format and content of the test procedures coveloped for the standard plant and safety-related site systems and components reflected the guidance provided in Regulatory Guide 1.68. The procedures contained as a minimum the following sections:
1.0 objectives The objectives section identified the general results to be accomplished by the test. 2.0 Acceptance criteria The acceptance criteria section clearly defined quan-titative and/or qualitative criteria against which the success or failure of the test procedure is judged. 3.0 References The references section identified those FSAR sections, vender manuals, drawings, etc. that were pertinent to the performance and/or development of the test-procedure. 4.0 Test Equipment The test equipment section identified temporary equip-ment required to conduct the test procedure and/cr (). ( collect data. 5.0 Notes and Precautions The notes and precautions sections listed limitations and precautions necessary to ensure personnel and equip-ment safety. Additional instructions needed to clarify the test procedure were also listed-in this section. 6.0 Prerequisites The prerequisites section identified those prerequisite tests and initial conditions that had to be completed and/or satisfied prior to the performance of the test procedure. 7.0 Test procedure The test procedure section provided a detailed step-by-step test method and instructions for data collection. All nonstandard arrangements required by the test pro-cedure section were restored either in the test proce-dure section or the system restoration section. O 14.2-9 Rev. O
-. - - . . - - - -- . . -_ ~ . . _ _ . ~ . - _- -
WOLF CREEK ._ 8.0- Test 10ata= Sheets I h: - Thei test data: sheet section provided specific forms _for
-dataicollection.=-Additional instructions, if!necessary, were alsoLidentified'for each data sheet' .
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- 9.0- System Restoration-The system-restoration section returned the system to a-safe operating or-standby condition. . Instructions for i the - removal and/or return;of system-temporary modifica-tions required-by the-prerequisite and/or test procedure sections.were clearly defined.
The- procedural. sections included,_ 'as applicable, appropriate-requirements for initials and/or signatures to control the per-formance_and' sequencing of the test.
. Thel _ test- _ procedures >were1 prepared using the latest design infor-mation-available and functional- requirements provided by the idesign' engineers.
This information was utilized in developing the detailed components test methods which verified the ability of_ systems. and to
. procedure preparation efforts function:within_their design _ specifications. The were started -more_ than- 2' years before the- first- procedure to be performed. This early start
- allowscfor an orderly development of-the ' test procedure program and of the--test procedures.
(The ll[i
- test procedures were reviewed by the cognizant design organi-zation to ensure.that the test procedure objectives and acceptance
-criteria are consistent- with current design document require-ments.'
Subsequent ance l criteria changes _to test procedure objectives or accept-
- during the preparational stage were based approved changesito design documents on with the design organiza-tion's concurrence.
14.2.3.2 ' Procedure' Review and Approval
- Followingiinitial the- procedure preparation, and prior to submittal to
-JTG ifor -review and approval recommendation, the test procedures were reviewed by the SNUPPS_ utilities (KG&E and Union Electric).' . ' Review comments were resolved between the SNUPPS
- utilities 1and the writing organization.
A-final _ revision was made by the writing organization, ating; all applicable design changes, incorpor-utilities for their review:and-approval. and was submitted to the 14.2-10 Rev. 0 O
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s. WOLF CREEK E(') N- Each utility-had various organizations, such as a startup organization, groupsi and committees, " safety review committee r compriwed of individuals initial test group, and a plant priate having appro-technical backgrounds and experience. these organizations, groups, and committees wereIndividuals responsiblewithin fort a. Reviewing procedures for accuracy and technical content;
- b. Verifying that the procedure has been revised to incor-potate known design changes;
- c. . Verifying procedure compatibility with field installa-tion of equipment;
- d. Verifying procedure conformance with FSAR requirements and plant operating technical specifications;
- e. Reviewing procedures against reactor operating and testing experiences of similar power plants.
14.2.3.2.2 Joint Test Group (JTG) A subcommitee of the PSRC, the JTG was organized by the Operating Agent to review test results. preoperational test procedures and preoperational .=-- The primary JTG functions were tot
- a. Review preoperational test procedures and recommend their approval by the Startup Manager.
- b. Evaluate and authorize changes to preoperational test procedures as detailed in the Startup Manual. Administrative
- c. Evaluate preoperational test procedure results and recommend their approval to the Startup Manager and Plant Manager,
- d. Review safety-related aspects of the startup adminis-trative procedures.
Membership in the JTG included the following personnel or designated representatives: their
- a. Superintendent of operations - Chairman
- b. Superintendent of Plant Support O 14.2-11 Rev. 0
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- c. Superintendent Services of Regulatory, Quality and Administrative
- d. Startup Technical Support Supervisor
- e. Assistant Startup Manager
- f. Operations Quality Assurance (non-voting member) 9 Bechtel Power Corporation-Engineering member) (non-voting h.
Westinghouse-Engineering (non-voting memoer) Others were requested to provide technical support tc the JTG. This support was based on the procedure being reviewed, required technical expertise or other the JTG meeting was with the concurrence of applicable factors. Participation in the JTG limited to technical input only. and was 14.2.3.2.3 Plant Safety Review Committee (PSRC) The PSRC was organited by the Oper4 ting Agent to ensure effective coordination of the engineering, construction, and operations activities affecting the startup program.
*he appropriate PSRC andmembers ensured sufficient review of initial 2ttitu? test procedures results. lll The pramary PSRC startup functions weret
- a. Review all initial startup test procedures and maka recommendations to the Plant Manager,
- b. Evaluation and authorization of changes to initial startup test prccedures.
c. Evaluation of initial startup test procedure results. Mombership in the PSRC is given in the Technical Soction 6.5.1.2. Specifications, 14.2.4 CONDUCT OF TEST PROGRAM 14.2.4.1 Adminietrative Procedures The conduct of theprocedures. preoperational startup program was controlled by administrative Thewas preparation, maintenance, implementation of these procedures and Startup Manager, the responsibility of the controls for startup The startup administrative activities such as: procedures prescribed 14.2-12 Rev. O O
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- a. Organization and interfaces;
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- b. Indoctrination and training'
- c. Preparation, review, approval, and' modification of test procedures;
- d. Format and content of test procedures;
- e. Tagging procedures;
- f. test scheduling and test conduct;
- g. Test deficiencies and resolution;
- h. Startup qualitv' control; and
- 1. Startup docusent control.
14.2.4.2 Turnover from construction to KG&E Startup construction completion was scheduled in accordanco with engineered system or subsystem boundaries. As systems or sub-systems were completed to support Startup testing, a turnover of the system or subsystem to KG&E Startup was processed. Turnover was conducted in accordance with established administrative prceedures. O As part of the turncver process, each' safety-related system er subsystem received physical walkdewns to provide assurance of readiness for Startup testing and verification that installation requirements had been met. Walkdowns were performed jointly by KG&E Startup and KG&E Construction personnel under the direction of the KG&E Conntruction Manager. Discrepancias identified during the walkdowns were tracked and resolved in accordance with established administrative and quality procedures. The system or subsystem Turnover Package prepared by the construc-tor was reviewed by KC&E Construction and KG&E Startup personnel for accuracy, completeness and acceptability for Startup testing. In conjunction with the Turnover Package review, Startup per-sonnel verified that the system or subsystem procurement and installation documentation review had been performed by construction, and that discrepancies had been addressed. Accept- ' ance of the Turnover Package by Startup followed satisfactory completion of the Turnover Package review. The Startup Manager was responsible for the approval and acceptance of the system or subsystem-and the associated Turnover Package. O 14.2-13 Rev. O
WOLF CREEK Individual camponents could be released to Startup for calibra-tion, testing or temporary operation prior to turnover. All components released in tnis manner were incorporated into the scope of a subsequent system or subsystem turnover. 14.2.4.3 Component and Prerequisite Testing Upon Startup acceptance of a turned-over system, released component, subsystem, or prerequisite-type testing was performed to demonstrate of, proper operability and functional ability and prior to, in support the performance of preoperational testing. Local containment leak rate testing, as described in Section 14.2.12.2.13, was performed at WCGS as part of the prerequisite test program. Administrative procedures were established to prerequisites were met before testing was ensure that all completion of all prerequisite initiated. Upon subsystem, tests applicable to a syste.n oc to a documented review was conducted verify that appropriate documentation was able and by Startup personnel that required prerequisite tests had been satisfactorily completed. All deficiencies which would prevent performance of preoperational tests or generate negative test. results were identified and dispositioned prior to implementdtion of the preoperational tests. 14.2.4.4 Preoperational Testing llh Technical direction and administration, including test execution and data sibility of recording, of the preoperational testing were the respon-the startup organization. The system startup engi-neera were responsible for the performance of tests and providing appropriate interface with station operators. The Startup Manager was responsible for the administration and surveillance testing activities during the preoperational test program. of all 14.2.4.5 Initial startup Testing During the initial startup testing phase, the Plant Manager had overall authority and responsibility for the startup program. The Startup Organization provided support to the plant operating staff which had responsibility for performing equipment operations and maintenance ing in accordance with the provisions of the plant operat-license. The WCGS operating staff was also responsible for onsoring that the conduct of testing did not place the plant in an unsafe condition at any time. 1 The_ shift supervisors had the authority to terminate or disallow tosting at any time. l 14.2-14 O Rev. 0 1
i= WOLF CREEK () 14.2.4.6 Test Prerequisites Each test procedure contained a set of prerequisites and initial conditions as prescribed by the procedures. startup administrative The system fied prerequisites were met prior to performing the test.spect-startup engineer ensured that all The format for test procedures is-described in Section 14.2.3.1. 14.2.4.7 Test Evaluation Upon completion of system preoperational testing, the test results were submitted to the JTG for its review and subsequent recommon-dation for approval to the Startup Manager and Plant Manager. Between test each major phase of-the initial startup test program, results for- all tests that were performed were reviewed the the PSRC. by This review ensured that all required systems were tested-' satisfactorily and that test results were approved before proceeding to the next stage of testing. These reviews are described in Section 14.2.5. 14.2.4.8 Design Modifications Modifications to the design of the equipment during the test program could be initiated in order to discovered as a result of testing. Any such correct deficiencies ( -either developed by the original design organization or other modifications were designated organizations. Modifications made to components or systems after completion of preoperational testing were reviewed for retesting requirements or initial startup on affected portions of the system. 14.2.5' REVIEW, EVALUATION, AND-APPROVAL OF TEST RESULTS The responsibility for-review, evaluation, and recommendation for approval of test results from all preoperational tests rested with the JTG. In the case of all initial start-up tests,. it rested with the PSRC. Following completion of a preoperational test, the responsible i system startup engineer. assembled the test data package for submittal to the members of the JTG for evaluation. Each test data package was reviewed to ensure that the test has been performed in accordance with the approved procedure and that all required data, checks, and signatures were properly recorded and that system performance met the approved acceptance criteria. O. 14.2-15 Rev. 0
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WOLF CREEK Members of tne JTG reviewed the evaluation findings and recem-mended corrective action deficiencies. to be taken to resolve any outstanding If the deficiencies were not resolved to the satis- lll faction of the JTG, then appropriate rotesting was required. If the evaluation indicated that deficiencies in the test, method were responsible for unsatisfactory test results, the test procedure was revised accordingly before retesting was initiated. The review and approval process for procedure revisions was carried out in the manner described in Section 14.2.3. Whenever an evalu-ation of test results indicated deficiencies in system perform-ance, the JTG reft. red the problem to the responsible engineering organization for evaluation. If the test documentation and system performance were acceptable, the JTG reconnended approval of the test by the Startup Manager and the Plant Manager. Following each major phase of the initial startup test program, the PSRC verified that all required tests were performed and that the test results were approved. This verification ensured that all required systems were operating properly and that testing for the next major phase was conducted in a safe and efficient manner. This type of review was performed to the extent required before major initial startup test phases such as fuel load, initial criticality, and power ascension. During the power ascension phase, review- and approval of initial startup test procedure results was completed as described in KMLNRC-84-235. 14.2.6 TEST RECORDS Test procedures and test data relating to preoperational and initial startup testing are retained in accordance with the measures described in Section 17.2.17. 14.2.7 CONFORMANCE OF TEST PROGRAMS WITH REGULATORY GUIDES The regulatory guidos applicable to the test program are listed, with positions, in Appendix 3A, Conformance to NRC Regulatory Guides. 14.2.8 UTILIZATION OF REACTOR OPERATING AND TESTING EXPERIENCE IN DEVELOPMENT OF TEST PROGRAMS Available information on reactor operating experiences was uti-lized in the development of the Initial Test Program, es ~~' lows: Rev, Ih ! 14.2-16 l
i WOLF CREEK () a. Bechtel reviewed and distributed pertinent Licensee Event Reports for use in the development of preopera- ; tional test procedures as follows: ' e
- 1. The Licensee Event Summary Reports and other per-tinent information were reviewed on a periodic !
basis, ' and those reports deemed to be useful for updating test procedures and items of a generic , nature were cataloged. A summary of these reports , was distributed within Bechtel. i
- 2. Copies of the specific reports vara then made and >
distributed for use in the preparation of proco- l dures. . In addition, these reports were coded and i filed in a computer retrieval system. i
- b. The operating experience assessment for Wolf Creek Generating Station Unit No. 1 (WCGS) was conducted by '
the nuclear divisions and_ plant staff who possess the appropriate experience in the area of concern. The sources of operating experience information included the use of the NETWORK and the INP0/NSAC SEEIN system. An j administrative system which controlled the- flow of ; information from NETWORK, INP0/NSAC SEEIN, etc., to the cognizant organizations including the Independent safety Engineering Group (ISEG) was developed and functioning prior to fuel load. The Licensing Section was responsible for coordinating , the review of the NRC Information and Enforcement (II) 1 Bulletins, circulars, and Information Notices. The Startup Group reviewed information provided by the other KG&E Nuclear Divisions and information provided by ' Bechtel and Westinghouse to determine its effect on the > Wolf Creek Initial Test Program, making revisions te test and administrative procedures as required. , An instrumented auxiliary feedwater water-hammer test was : performed only at Wolf Creek. (This test was not required to be performed. It was being performed for the purpose of gathering . engineering data only.) Procedure S-03ALO4, Auxiliary Feedwater ' System Water Hammer Test, required a visual and audible water hammer test and wai completed prior to the issuance of an operating license. See new Section 14.2.12.1.10. Procedure S-070017, Loss of Heater Drain Pump Test, was performed on callaway only. This test was conducted to verify analytical assumptions. No additional loss of hanter drain pump tests are 4 () . 14.2-17 Rev. 0 1 l
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l WOLF CREEK required, since the data obtained from tne first unit test equally valid for subsequent units. See Section 14.2.12.3.41. is lll Procedure S-07SF09 RCCA or Bank Worth Measurement at Power, was performed at 50 percent power only at Callaway. Wolf Cree < and i Callaway have the same core and Nuclear instrumentation system l design and the test at Callaway is considered a prototypical test { for Wolf Creek. This position was accepted by the NRC in a July 3, 1985 letter to KG&E. 1 l A natural circulation test was performed at callaway only to demonstrate the length of time to stabilize natural circulation, core flow distribution, and the ability to establish and maintain natural circulation. Operators participating in the tests were able to recognize when natural circulation had stabilized and were able to control saturation margin, RCS pressure, and heat removal rate without exceeding specified operating limits. were These tests conducted insofar as possible to include all available licensed operators. Licensed operators were trained in these same areas on the simulator. simulating natural The simulator has full capsoility of circulation, using Westinghouse dats initially. plant, When the above tests were accomplished on the Callawat actual data was incorporated into the Wolf Creek simulator program. See Chapter 18, item 1.G.1, and Section 14.2.12.3.43. 14.2.9 TRIAL USE OF PLANT OPERATING AND EMERGENCY PROCEDURES The plant operating procedures were utilized, where applicaole during tions, the test program, to support testing, maintain plant condi-and facilitate training. The trial use of opetsting pro-cedures served to familiarize operating personnal with systems snd plant operation during the testing phase and also served to ensure the adequacy of the procedures under actual er simulated operating conditions before plant operation begins. The emergency proce-dures were verified during startup as plant conditions, testing, and training warrant. Surveillance tests were performed as condi-tions warrant during the testing program, to demonstrate their adequacy. Plant operating procedures were same time frame as the preparation ofdeveloped in approximmrely tne preoperational and initial startup tests. The operating procedures were revised as necessary to reflect experience gained during the testing program. 14.2.10 INITIAL FUEL LOADING, CRITICAL:TY, AND POWER ASCENSION Prior to the commencement of fuel loading, required preoperatiora1 test procedures were evaluated, and appropriate remedial action 14.2-13 Rev. 0
WOLF CREEK () was taken if the acceptance criteria was not satisfied. completion of fuel loading, At the the reactor upper internals and pres-sure vessel head were installed, and additional mechanical and electrical tests were performed to prepara the plant for nuclear operation. After final procritieni tests, nuclear operation of the reactor began. This phase of testing included initial criti-cality, low power testing, and power level ascension. The purpose of these tests was to establish the operational characteristics of the unit and core, to acquire data for the proper calibration of setpoints, and to ensure that operation is within license require-ments. Section 14.2.12.3 summarizes the tests which are performed from fuel load to rated power. The fuel loading and post loading tests are described below. 14.2.10.1 ruel Loading The Plant Manager or his designated representative with technical assistance provided by Westinghouse, was responsible for the coordination of initial core loading. The overall process of initial core loading was, in general, directed from the operating floor of the containment structure by a licensed senior reactor operator. The licensed senior reactor operator had no additional responsibilities other than core load operations. The core configuration was specified as part of the core design studies conducted well in advance of fuel loading. In the event mechanical damage was sustained during core loading operations to a fuel assembly of a t pe for which no spare was available ensita, an alternate cora-loading scheme could have been' determined. Any such changes would have been approved by the appropriate Westinghouse personnel. Core loading-procedures specified the condition of fluid systems to prevent inadvertent changes in boron concentration of the reactor coolant; the movement of fuel to preclude the possibility of mechanical damage; the conditions under which loading could proceedt and the responsibility and authority for continuous and complete fuel and core component accountability. The following conditions were met prior to core loading
- a. The resctor containment structure was complete and containment integrity had been demonstrated,
- b. Fuel handling tools and equipment were checked out and operators familiariaJd in the use and operation of equipment. Inspections of fuel assemblies, rod cluster control assemblies, and reactor vessel were l satisfactorily completed.
l
)
14.2-19 Rev. 0 l
WOLP CREEK
- c. The reactor vessel and associated components were in a state of readiness to receive fuel. The water level was llh naintained above the bottom of the nozzles and recircu-lation maintained l
to ensure the required boron concen-traticn could be increased via the recirculation path or directly to the open vessel. Criteria for safe 1 immediatel.v if any ofloading required the following that loading conditions operations stop occur, l a. An unanticipated factor increane in the neutron count rates by a of two occurs on all responding nuclear channels during any single loading step after the initial nucleus of eight fuel assemblies is loaded,
- b. An unanticipated increase in the count rate by a factor of five on any individual responding during any single loading step after thenaclest channel initial nucleus of eight fuel assemblies is loaded.
- c. An unanticipated decrease in boron concentention greater than 20 ppm is determined from two successive sampics of the reactor coolant.
Loading operations could not be restarted until tne sitaattoq was ovaluated. An alarm in the containment and mein control room was coupled to the source range channels with a setpoint equal less than five times the current count rate. This to Sr alarm g automatically rate, alerts the loading operation personnel _of hign count and an immediate stop of all operatione suld be required until the situation was evaluated. alarm was actuated during core loading In the even the evacuation determined that no hazards to personnel and after it has been exist, personnel would be permitted to reenter the c;ntainment preselected to evalu-ate the cause and determine future action. The core was assembled in the reactor vessel and submerged in the roactor grade water containing sufficient dissolved boric acid to maintain or lower.a calculated core effective multiplication factor of 0.95 core loading. The refueling Core moderator, pool could be wet or dry during initial chemistry conditions (particularly boron concentration) were prescribed in the core loading procedure document and verified by chemical analysis of moderator samples token prior to and during core loading operations. At least two artificial neutron sources were introduced into the core ensure a detectorpoints at specified in the core during the loading program to response of at least 2 counts per second attri-butable to neutrons. 14.2-20 Rev. O
WOLF CREEK , () core loading instrumentation consisted installed source range type) of two permanently nuclear channels and two temporary incore source ran(pulse ge channels. A third temporary channel could also be used as a spare. The permanent channels, when responding, were monitored in the main control room, and the installed and monitored in the contain-temporary channels were > ment. At least one permanent channel was equipped with an audible count rate indicttor. Both plant channels have the capability of displaying the neutron flux level on a strip chart recorder. The temporary channels indicated on scalars, and a minimum of one channel was recorded on a strip chart recorder. Normally minimum count rates of two counts per second attributable to core neutrons were required on at least two of the four (i.e. two temporary and two permanent source range detectors) available nuclear source channels at all times following installation of the initial nucleus of eight fuel assemblies. A response check of nuclear instruments to a neutron source was performed within 8 hours prior to loading of the core, or upon resumption of loading if delay was for more than 8 hours. Tual assemblies, together with inserted components (control red assemblies, burnable, poison assemblies, source spider, or thimble plugging devices) were placed in the reactor vessel one at a time, according to a previously established and approved sequence developed to provide reliable core monitoring with minimum possi-bility of core nochanical damage. The core loading procedure
/~s documents prescribed the successive movements of each fuel (J assembly and its specified inserts from its initial position in the storage racks to its final position in the core. Fuel casembly status boards were maintained throughout the core loading operation.
An initial nucleus of eight fuel assemblies, one containing a neutron source, is the minimum source-fuel nucleus which permitted subsequent meaningful inverse count rate monitoring. This initial nucleus was determined by cciculation to be markedly suberitical (X,gf go.95) under the required conditions of loading. Each subsequent fuel addition was accompanied by detailed neutron count rate monitoring to determine that the just-loaded fuel assembly did not excessively increase the count rate and that the extrapolated inverse count rate ratio was behaving as expected. These-results for each loading step were evaluated before the next fuel assembly was loaded. The final, as loaded, core configura-tion was subcritical (K,ff go.95) under the required loading conditions. () 14.2-21 Rev. O y v- -
WOLT CRIEK 14.2.10.2 Initial Criticality
- ggg Prior to initial criticality, the following tests were performed and the results evaluated.
- a. At the completion of core loading, the reactor upper internals and pressure vessel head wore installed. A pressure test was conducted after filling, and venting was completed to check.the leaktightness of the vessel head installation.
- b. Hechanical and electrical tests were performed on the control red drive mechanisms. These tests included a complete operational checkout of the mechanisms and calibration of the individual red position indicators.
- c. Tests were performed on the reactor trip circuits to test manual trip operation, and actual control rod assembly drop times were measured for each control rod assembly. At all times that the control red drive mechanisms were being tested, the boron concentration in the coolant was maintained so that the shutdown margin requirements specified in the Technical Specifications were met. During individual RCCA or RCC bank notion, source range instrumentation was monitored for unexpected changes in core reactivity.
- d. The reactor control and reactor protection systems were checked with simulated inputs to produce trip signals lll for various trip conditions.
- e. A functional electrical and mechanical check was made of the incere nuclear flux mapping system near normal operating temperature and pressure.
Initial criticality was achieved by a combination of shutdown and control bank withdrawal and reactor coolant system boron concen-tration dilution. The plant conditions, precautions, and specific instructions for the approach to criticality were specified by approved procedures. Initially, the shutdown and control banks of control rods were withdrawn incrementally in the normal withdrawal sequence, leaving the last withdrawn control bank partially inserted in the core to provide effective control when criticality was achieved. The boron concentration in the reactor coolant system was reduced and criticality achieved by boron dilution or by subsequent rod with-drawal following boron dilution. Throughout this period, samples of the primary coolant were obtained and analyzed for boron con-centration. O 14.2-22 Rev. O
l l WOLF CREEK Inverse count rate ratio monitoring using data from the normal plant source range instrumentation was used as an indication of ! the proximity-and rate of approach to criticality. Inverse count rate ratio data was plotted as a function- of rod bank position daring rod motion and as a function of reactor makeup water addi- i tion during reactor coolant system boron concentration reduction. i 14.2.10.3 Low Power Testino Following initial criticality, a program of reactor physics measurements was undertaken to verify that the basic static and ,
+
kinetic characteristics of the core were as. expected and that values of the kinetic coefficients the analysis were conservative. assumed in the safeguards Procedures specified the sequence of tests and measurements to be conducted and the conditions under which each was performed in order to_ ensure both safety of operation and the validity and consistency of the results obtained. significantly from design predictions, If test results deviated if unacceptable behavior had been revealed, or if unexplained anomalies had developed, the plant would have been brought to a safe stable condition and the situation reviewed to determine operation. . the course of subsequent plant inese measurements were made at low power and primarily at or near () normal operating temperature and pressure. Measurements were made in order to verify the calculated values of control rod reactivity -worths, bank the isothermal temperature coefficient under various core conditions, differential boron concentration reacti-vity worth, and critical control rod configuration. boron concentrations as functions of In addition, measurements of the relative power distributions were made, and concurrent tests were conducted on the instrumentation, including source and intermed!- ate range nuclear channels. Gamma and neutron radiation surveys were performed at selected points throughout the station. Periodic sampling was performed to verify chemical and radio-chemical analysis of the reactor coolant. 14.2.10.4 Power-Level Ascension After the operating characteristics of the reactor were verified by low power testing, a program of power level ascension brought the unit to its full rated power level in successive stages. At each successive stage, hold points were provided to evaluate approve and test results prior to minimum test requirements for each successive stage of power proceeding to the next stage. The ascension were specified in the initial startup test procedures. O 14.2-23 Rev. 0
WOLT CREEK Measurements were made to determine the relative power distribu-tion bank in tne core as functions of power level and control assembly position. Secondary system heat balance measurements ensured that the indi-cations of power level were consistent and provide bases for calibtation of the power range nuclear channels. The ability of the reactor coolant system to respond effectively to signals from primary and secondary instrumentation under a variety of condi-tions encountered in normal operations was verified. At prescribed power leve?,s, the dynamic response characteristics of the primary and secondary systems were evaluated. response characteristics were measured for design step System load changes, rapid load reduction, and plant trips. Adequacy of radiation shielding was verified by gamma and neutron radiation surveys at selected points throughout the station at various rower levels. Periodic sampling was performed to verify the chemieul and radio-chemical analysis of the reactor coolant. 14.2.11 TEST PROGRAM SOHEDULE Detailed schedules for tasting were prepared, reviewed, and revised on a continuing basis as plant construction progressed. Preoperational tests which were schedule were not reviewed on a case-by-case basis. performed according to 4h Administrative procedures were established to ensure that all prerequisites were met before testing was initiated. Upon completion of all prerequisi:e tests applicable to a system or subsystem, a documented review was conducted by Start-up personnel to verify that appropriate documentation was available and that required prerequisite tests were satisfactorily completed. All deficiencies which would have prevented performance of preoperational tests or generated negative test results were identified and dispositioned prior to implementation of the pre-operational tests. Preoperational testing was scheduled to commence approximately 13 months prior to fuel loading. The preoperational tests were performed and sequenced during this period as a function of system turnover, system interrelationships, and acceptance for testing. Initial startup testing was scheduled to be conducted over a period of approximately 3 to 5 months, commencing with fuel load-ing. The sequential schedule for initial startup tests ensured, insofar as practicable, that test requirements were completed O 14.2-24 Rev. O
i i WOLF CREEK ;
) prior to exceeding 25-percent power for all plant structures, ,
systems, and components that are relied upon to prevent, limit, or mitigate the consequences of postulated accidents. The development of the test procedures was an ongoing process consisting of preparation, review, and revision. Preoperational test procedures were available for NRC review approximately 60 days prior to the performance of an individual test. If an indi-vidual the test procedure was not available 60 days prior to the test, NRC was notified of the test date and the date the test pro-cedure was available. Initial startup test procedures were avail- - aole for NRC review at least 60 days prior to fuel loading. 14.2.12 . INDIVIDUAL TEST DESCRIPTIONS Test abstracts were provided for both safety-related and selected nonsafety-related preoperational tests. The abstracts included test prerequisites acceptance criteria. and summaries of test methods, objectives, and 14.2.12.1 Safety-Related Preoperational Test Procedures The following sections contain test abstracts used for safety-related preoperational tests. Table 14.2-1 provides an index of these tests. () t The preoperational test procedures were designated SO3 (Safety-Related/ Common to WCGS and Callaway), SU3 (Safety-Related/WCGS Specific), : SO4 thru SO9 (Honsafety-Related/ Common to WCGS and ! Callaway) and SU4 thru SU9 (Honsafety-Related/WCGS Specific) as appropriate. , 14.2.12.1.1 Steam Dump System Preoperational Test (S-0 3AB01) 14.2.12.1.1.1 Objectives
- a. To demonstrate the operability of the steam dump control system control circuits in both the average temperature ,
and steam pressure modes of operation.
- b. To demonstrate the operation of the main steam dump '
valves and main steam cooldown valves, including valve response to safety signals,
- c. To verify the operation of the main steam line drain valves' control circuits, including valve response to a turbine trip signal.
O 14.2-25 Rev. 0
- _ . - - - - - _ - - - - - - . - _ . - ~ , . . . - . . . -
WOLF CREEK
- d. To verify the operation of the main steam to turbine- O driven feedwater pump supply valves' control logics, including valve response to an auxiliary feedwater actuation signal (AFAS),
- e. To verify the operation of the main steam power-operated relief valves' control circuits.
14.2.12.1.1.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are completed,
- b. Required electrical power supplies and control circuits are operational.
14.2.12.1.1.3 Test Method
- a. Operability of the steam dump control system control circuits is veri fied in both the average temperature and steam pressure modes,
- b. Operability of the main steam dump valves' and main steam cooldown valves' control circuits is verified, including valve response to turbine impulpa low pressure, low-1tw average temperature, and condenser shell high pressure signals. ll)
- c. Operability of the main steam line drain valves' control circuits is verified, including valve response to a turbine trip signal.
- d. Operability of the main steam to turbine-driven auxil-iary feedwater pump supply valvee' control logics is verified, including valve response to an AFAS.
- e. Operability of the main steam power-operated relief valves' control circuits is verified.
14.2.12.1.1.4 Acceptance Criteria
- a. The response of the main steam dump valves and the main steam cooldown valves to the .ssociated turbine impulse low pressure, low-low average temperature, and condenser shell high pressure signals is in accordance with system design.
- b. The main steam line drain valves open on receipt of a turbine trip signal.
O 14.2-26 Rev. O
WOLF CREEK () c. The main steam pump supply valves open on receipt of an AFAS. to turbine-driven auxiliary feedwater
- d. The reuponse of the main steam power-operated relief valves to pressure signals is in accordance with system design.
14.2.12.1.2 Main Steam Safety Valve Test (SU3-AB02) 14.2.12.1.2.1 Objectives To verify the pressure relief ontpoints of the main steam safety valves. NOTE: This objective may be accomplished either by bench testing or with a pneumatic test device. 14.2.12.1.2.2 Praraquisites The following prerequisites apply when a pneumatic test device is used,
- n. Required instrument calibration is complete,
- b. Hot Functional Testing is in progress.
r's
- V c. A Source of compressed air is available to provide air to the air set pressure device installed on the valve under test.
The following prerequisites apply when bench testing is performed.
- a. Bench testing facility is available,
- b. An apprcved WCC3 procedure is available to accomplish bench testing.
- c. A source of compressed gus is available to provida pressure to uhe valve under test.
14.2.12.1.2.3 Test Method The -following test. method applied when a pneumatic test device is used. Main steam pressure is adjusted within the required range, and ' air is admitted to the air set pressure device on the safety valve under test. Actual lift pressure is calculated, using the steam pressure and converted air pressure at the time of lift. O 14.2-27 Rev. 0
WOLF CREEK The following test applies when bench testing is performed. lll With the main steam safety valve mounted on the bench test facility, the spring assembly is preheated and the safety valve is pressurized with compressed gas. Actual set pres-sure is determined at the time of lift. 14.2.12.1.2.4 Acceptance Criteria Each main steam safety valve lifts within its respective setpoint tolerance. 14.2.12.1.3 !!ain Steam Line Isolation Valve Test (S-03AB03) 14.2.12.1.3.1 Objectives
- a. To verify the response of the main steam bypass, drain',
and auxiliary feedwater turbine warmup valves to steam line isolation signals.
- b. To demonstrate the operability of the main steam isola-tion valve control circuits, including control circuit response to a steam line isolation signal (SLIS).
14.2.12.1.3.2 Prerequisites
- a. Required component testing, instrument calibration, system flushing / cleaning are complete.
and a W
- b. Required electrical power supplies and control circuits are operational.
- c. The main steam line isolation valve accumulators are charged, and the associated hydraulic systems are operational.
14.2.12.1.3.3 Test Method An SLIS is initiated, and the response of the main steam bypass, main steam drain, and auxiliary feodwater turbine warmup valves is verified. 14.2.12.1.3.4 Acceptance Criteria
- a. The main steam bypass, drain, and auxiliary feedwater turbine warmup valves close on receipt of an SLIS.
14.2.12.1.4 Main Steam System Preoperational Test (S-03AB04) O 14.2-28 Rev. 0
I WOLF CREEK () 14.2.12.1.4.1 objectives
- a. To determine, during hot functional testing, the operat-ing times of the main steam isolation valves, main steam bypass valves, main steam dump valves, main steam cool ~
down valves, and the main steam power-operated relief valves.
- b. To verify the response of the main steam isolation valves to steam line isolation signals. .
14.2.12.1.4.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. Hot functional testing is-in progress,
- d. The condenser is available to receive steam from the main steam system. '
14.2.12.1.4.3 Test Method
- a. The main steam O
valves, main isolation valves, steam dump valves, main steam main steam cooldown bypass valves, and the main steam power-operated relief valves are operated, and operating times are recorded,
- b. An SLIS is. initiated, and the response of the main steam isolation valves is verified.
14.2.12.1.4.4 Acceptance Criter!a
- a. The operating times of the main steam isolation valves, main steam dump valves, main steam bypass valves, main steam cooldown valves, and the main steam power-operated relief valves are.within design spescifications.
- b. The main steam isolation valves close on receipt of a steam line isolation signal.
14.2.12.1.5 Main Feedwater System Preoperational Test (S-03AE01) 14.2.12.1.5.1 objectives
- a. To demonstrate the operation of the feedwater system valves and to verify the response of the feedwater system valves to a feedwater isolation signal (FIS).
14.2-29 Rev. O
WOLF CREEK
- b. To perform the initial operation of the steam generator feedwater pumps (SGPP) .
14.2.12.1.5.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The closed cooling water system is available to provide cooling water to the SGFP lobe oil coolers.
d. The compressed air system is available to provide air to system air-operated valves. e. The steam seal system is available to provide seal s tia n and packing exhaust for the SGFPs.
- f. The main turbine is available for operation. turning gear g.
The condensate system is available to supply suction for the SGFPs. h. The main condenser is available to receive SGPP exhaust. turbine k i. The auxiliary steam system is available to provide steam flow to the SGFP turbines. 14.2.12.1.5.3 Test Method
- a. Feedwater system valves are operated, and the proper response of required system valves verified, to an FIS is
- b. The turbine-driven SGFPs are operated as limited by steam, and operating data are recorded,
- c. The motor-driven SGFP is operated, and operating data are recorded.
14.2.12.1.5.4 Acceptance Criteria
- a. The feedwater control valves, steam generator feedwater isolation valves, feedwater chemical injection isolation valves, and feedwater bypass control valves close on receipt of an FIS.
O 14.2-30 Rev. O
WOLF CREEK !
- b. The closing time of the feedwater t O-within design specifications, isolation valves is
- c. The performance of the motor-driven SGFP is within ;
design specifications. 14.2.12.1.6 Steam Generator Level Control Test ( S-0 3 AE0 2) 14.2.12.1.6.1 objectives
- a. To . demonstrate the operability of the feedwater control valves (FWCVs).
- b. To demonstrate the operabi2(ty the valves.
of FWCV bypass
- c. To demonstrate the response of the FWCVs and bypass o valves to signals generated by the steam generator level control system.
14.2.12.1.6.2 Prerequisites
- a. Required component testing are complete. and instrument calibration
- b. Required electrical power supplies and control '
are operational. circuits 14.2.12.1.6.3 Test Method
- a. The FWCVs .are operated from their respective controllers, and the FWCVs response to feedwater flow, steamline flow, and steam generator lerei is verified.
- b. The FWCV bypass valves are operated from their respec-tive controllers, .and their response to steam generator level and neutron flux signal is verified.
14.2.12.1.6.4 Acceptance Criteria
- a. The response of the FWCVs to feedwater flow, flow, steamline and steam generator level is in accordance with system design.
- b. The response. of the FWCV bypass valves to steam gener-ator level and neutron flux signal is-in accordance with system design.
14.2-31 Rev. 0
WOLF CREEK 14.2.12.1.7 Auxiliary recdwater Motor-Oriven Pump and Valve Preoperational Test (S-03ALO1) 14.2.12.1.7.1 Objectives To demonstrate the operability of the motor-driven feedwater pumps, auxilisry determine by flow test their ability to supply water signals. to the steam generators, and verify their response to safety The operation of system motor-operated valves, including their response to safety signals, is also verified. 14.2.12.1.7.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
c. The condensate storage tank contains an adequate supply of domineralized water for the performance of this test,
- d. The steam generators are available to receive water from the auxiliary feedwater system.
14.2.12.1.7.3 Test Method
- a. Performance characteristics of the motor-driven auxil- O iary feedwater pumps are verified while discharging to the steam generators,
- b. System component control circuits are including the operation verified, of the motor-driven auxiliary feedwater pumps and system valves on receipt signals. of safety 14.2.12.1.7.4 Acceptance Criteria
- a. Motor-driven auxiliary feedwater pump per formance cha r-acteristics must be within design specifications.
- b. Motor-driven auxiliary reedwater pumps automat'.ca'ly start on receipt of an engineered safety features aw.u-ation signal (ESPAS) in the absence of an SIS signal and a Class IE 4.16 kV bus undervoltage signal.
- c. Auxiliary feedwater suction valves from essential ser-vice water system open, and suction v31ves from condon-sate storage tank close, on condensate otorage tank low-suction-pressure feedwater pump ESFAS.
signals, coincident with an auxiliary 14.2-32 Rev. 0 1
WOLF CREEK 14.2.12.1.8 Auxiliary Teodwater Turbine-Driven Pump and Valve Preoperational Test (SU3-ALO2) 14.2.12.1.8.1 Objectived I
- a. To verify the auxiliary feedwater pump turbine j mechanical trip and throttle valve automatic operation on an auxiliary feedwater actuation signal (AFAS) .
- b. To perform the initial coupled operation of the turbine-driven auxiliary feedwater pump. Full flow character-1stics of the turbine-driven pump will be demonstrated during hot functional testing.
- c. To perform five consecutive, successful, cold starts of the turbine-driven auxiliary feedwater pumps.
14.2.12.1.8.2 Prerequisites
- a. Required component testing,- instrument calibration, and system flushing / cleaning are complete.
b Required electrical power supplies and control circuits are operational.
- c. The steam generators are-available to receive water frem the auxiliary feedvater pumps.
- d. The steam generator blowdown system is available to maintain the normal operating levels in the steam gener-aters during auxiliary feedwater pump operation.
- e. The auxiliary steam system is available to supply stean to the auxiliary feedwater pump turbine,
- f. For the performance characteristic test of this pump, hot functional testing (HTT) is in progress.
14.2.12.1.8.3 Test Method
- a. An AFAS is simulated, and opening of the mechanical trip and throttle valve-is verified.
! The turbine-driven
- b. auxiliary feedwater pump is operated during HTT, and performance characteristics are recorded.
l
- c. The ability of the turbine-driven auxiliary feedwater pur.ps to start successfully five consecutive times frem cold conditions is verified.
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WOLF CREEK 14.2.12.1.8.4 Acceptance criteria ggg
- a. The auxiliary foodwater pump mechanical trip and throttle valve opens automatically on an AFAS.
- b. Operating characteristics of the turbine-driven auxil-inry feedwater pu=p are in accordanco with design.
- c. The turbino driven auxiliary feedwater pu=p starts successfully five consecutive times from a cold start.
14.2.12.1.9 Auxiliary Feedvater Motor-Driven Pump Endurance Tent (SU3-ALO3) 14.2.12.1.9.1 Objectives
- a. To demonstrate that the motor-driven auxiliary feedwater pumps can operate for 48 continuous hours without exceeding any of their limiting design specifications,
- b. To demonstrate that the motor-driven auxiliary fendwater pumps can operate for 1 hour after a cooldown from the 48-hour test.
- c. To demonstrate that the room environmental conditions are not exceeded during the 48-hour test.
14.2.12.1.9.2 Prerequisites lll
- a. Required component testing, instrument calibration and system flushing / cleaning are complete,
- b. Required electrical power supplion and control circuits are operational.
- c. The appropriate auxiliary foodwater pump room coolers are operational.
- d. The condensate storage tank is available as a water source and to receive recirculation flow.
14.2.12.1.9.3 Test Method Each motor-driven pump is started and operated for 48 hours after reaching rated speed and rated discharge pressure and flow, or a greater pressure and less flow. During the ondurance run, pump-operating data and the pump room environmental conditions are recorded. At the completion of each endurance test, the pump is cooled for 8 hours and until pump data returns to within 20 F of the original protest data. The pump is then started and operated for 1 hour. O 14.2-34 Rev. O
WOLF CREEK (8 l l 14.2.12.1.9.4 Acceptance Criteria
- a. The operating parameters (vibration. bearing tempera-etc.) of each motor-driven auxiliary feedwater tures, pump do not exceed the design specifications. ;
- b. The environmental conditions of each motor-driven aux 11-iary feedwater pump room do not exceed the design -
specifications. 14.2.12.1.10 Auxiliary Feedwater System Water Hammer Test (S- 3 03ALO4) 14.2.12.1.10.1 Objectivea To demonstrate that the injection of auxiliary feedwater at rated flow into a steam generator at or near normal operating tempera-tures will not causa damaging water hammer to the steam generators and/or feedwater system. 14.2.12.1.10.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits
() are operational.
- c. The steam generators are at or near normal operating :
temperature,
- d. The condensate storage tank is available as a water source.
14.2.12.1.10.3 Test Method , Auxiliary feedwater is injected into each steam generator. The feedwater system piping and the steam generators are monitored visually and audibly to verify that no damaging water hammer occurs. 14.2.12.1.10.4 Acceptance criteria No damaging water hammer occurs. 14.2.12.1.11 Auxiliary Feedwater Turbine-Driven Pump Endurance Test (SU3-ALOS) 14.2.12.1.11.1 Objectives O. 14.2-35 Rev. 0
WOLF CREEK
- a. To demonstrate that the turbine-driven aurt11ary feed-water pump can operate for 48 continuous nours without exceeding any of its limiting design specifications,
- b. To demonstrate that the turbine-driven auxiliary feed-water pump can operate for 1 hour after a cool down from the 48-hour test.
- c. To demonstrate that the room environmental conditions are not exceeded during the 48-hour test.
14.2.12.1.11.2 prerequisites
- a. Required component testing, instrument calibration and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The appropriate auxiliary feedwater pump roem coolers are operational.
- d. The condensate storage tank is available as a water source and to receive recirculation flow.
- e. A steam source is available.
14.2.12.1.11.3 Test Method The pump is started and operated for 48 hours after reaching rated speed and rated discharge pressure and flow, or a greater pressure and less flow. The turbine-driven auxiliary feedwater pump operating steam is as close to normal operating temperature as possible and is at least 400 F. During the endurance run, pump-operating data and the pump room environmental conditiens are recorded. At the completion of the endurance test, the pump is cooled for 8 hours and until pump data returns to within 20 F of the original protest data. The pump is then started and operated for 1 hour. 14.2.12.1.11.4 Acceptance Criteria
- a. The operating parameters (vibration, bearing tempera-tures, etc.) do not exceed the design specifications.
- b. The environmental conditions of the turbine-driven auxiliary feedwater pump room do not exceed the design specifications.
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i
+
I WOLF CREEK ( 14.2.12.1.12 Reactor Coolant Pump Initial Operation (S-03BB01) i 14.2.12.1.12.1 objectives To demonstrate the operating characteristics of the reactor cool- I ant pumps and verify the operation of their associated oil lift pumps. 14.2.12.1.12.2 Prerequisites ,
- a. Required componsnt testing, instrument calibration, and system flushing /ct ennina are complete. j
- b. Required electrical mower supplies and control circuits are operational,
- c. The chemical and volume control system is available to .
provide seal water to the reactor coolant pump seals. ; _ d. The component cooling water system is available to supply cooling vater to the reactor coolant pumps. 14.2.12.1.12.3 Test Method
- The reactor coolant pumps and associated oil lift pumps are cper-ated, and pump operating data are recorded.
O 14.2.12.1.12.4 Acceptance Criteria t Reactor coolant pump and oil lift pump operating characteristics are within desigr. specifications. 14.2.12.1.13_ Pressurizer Relief Tank Cold Preoperational Test (SO 3-BB0 2) 14.2.12.1.13.1 objectives - To demonstrate that the reactor makeup water system can supply design pressurizer relief tank (PRT) spray flow against design backpressure. The' operation of the PRT nitrogen isolation valves, including their response to a containment isolation signal, is also verified. 14.2.12.1.13.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational. ,
O 14.2-37 Rev. 0
- . _ _ . ~ _ _ . _ _ . . . _ _ _ _..____._.._.a. . , _ . ~ . _ .._,,.___a ,._,-..-_,_.,.._.2,.._.__
WOLF CREEK
- c. The reactor makeup water system is available to supply water to the PRT.
- d. The service gas system is available to pressurize the PRT.
14.2.12.1.13.3 Test Method
- a. With a design backpressure in the DRT, a reactor makeup water pump is operated to obtain the spray flow to the PTT.
- b. The response of the PRT nitrogen isolation valves to a containment isolation signal is verified.
14.2.12.1.13.4 Acceptance Criteria
- a. The reactor makeup water system supplies the design spray flow to the PRT with design backpressure in the PRT.
- b. PRT nitrogen isolation v.sives close on receipt of a containment isolation signal. Valve closure times see within design specifications.
14.2.12.1.14 RTD Bypass Flow Measurement (S03-8803) ggg At WCGS, test S-07BB01 (USAP Sectic n 14.2.12.3.3) was used to satisfy the requirement for verification of design specifications. 14.2.12.1.15 Pressurizer Pressure Control Tost (S-03BB04) 14.2.12.1.15.1 Objectives To demonstrato the stability and response of the pressuriser pressure control system, including the verification of pressurizer pressure alarm and control functions. 14.2.12.1.15.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The plant is at normal operating temperature and pres-sure with all reactor coolant pumps running, and hot functional testing is in progress.
O 14.2-38 Rev. O
4 WOLF CREEK (} 14.2.12.1.15.3 a. Test Method Pressurizer predsure is varied, and the ability of the pressurizer pressure control system to automatically control and stabilize pressurizer pressure is verified.
- b. Pressurizer pressure is varied, and pressurizer pressure control system alarm and control setpoints are verified.
14.2.12.1.15.4 Acceptance criteria ,
- a. The pressurizer pressure control system responds, in accordance with system design, to an increase and decrease in system pressure.
- b. Pressurizer pressure control system alarm and control setpoints are within design specifications.
14.2.12.1.16 Reactor Coolant System Hot Preoperational Test (S-03BB05) 14.2.12.1.16.1 Objectives
- a. To operate the reactor coolant systam at full flow con-ditions for a minimum of 240 hours to provide the neces-sary vibration cycles on the vessel's internal
. components prior to their inspection at core loading.
- b. To provide coordination and initial conditions necessary for the conduct of those preoperational tests to be performed during heatup, normal operating temperature and pressure, and cooldown of the reactor coolant system.
14.2.12.1.16.2 Prerequisites
- a. The reactor coolant system cold hydrostatic test is complete.
- b. The reactor vessel internals and head are installed, and the vessel is available to support this test,
- c. All systems and components required to support heatup, operations at normal temperature and pressure, and cool-down of the reactor coolant system are available,
- d. Required instrument calibration is complete.
.O 14.2-39 Rev. O
Y e WOLF CREEK l e.- The_ examination of the reactor internals in accordance-Ewi th -_ Sect ion _ 3. 9 (N) . 2. 4, is_ complete 14.2.12.1.16.3 Test-Method
- a. The- reactor coolant system is . operated ~at full flow l c*dditionssfor-a4 minimum of-240 hours.
;b. Tucsv preoperational tests . required to be performed-during heatap, normal operating- temperature and ;
l pressure, and cooldown of the reactor coolant system.are l completed,=as coordinated by this test. ' 14.2.12.1.16.4 Acceptance Criteria The reactor coolant system hasioperated at full flow conditions lor a minimum of 240 hours.
'Notest 1.. The acceptance criteria for individual systems are a-part of the individual test procedures sequenced by !
this procedure.
'2. A post-hot functional ~ examiration of the i .: n a, internals is- performed as described in e: a
-3. 9 (N) . 2. 4.
'14.2.12.1.17 ThermalLExpansion (S-0 3BB06)- llk 14.2.'.2.1.17.1 Objectives To verity that during heatup_and cooldown of the reactor coolant sjstem the associated' components, piping _, support, and restraint deflections are-unobstructed and within dusign specifications, j 14.2.12.1.17.2 Prerequisites i
a.- This test is conducted in conjunction 1 with hot- func--
- tional testing.
- b. Supports, restraints, and hangers are installed and reference points and predicted movements established,
- c. Required instrument calibration is complete.
14.2.12.1.17.3 Test Method-During~the reactor coolant system-heatup and cooldown, deflection data-are-recorded. ! O 14.2-40 Rev. 0
WOLF CREEK t' (m)/ 14.2.12.1.17.4 Acceptance Criteria
- a. Unrestricted expansion and movements are verified to be within design specifications.
- b. Components, piping, supports, and restraints return to their baseline cold position in accordance with system design.
14.2.12.1.18 Pressurizar Level Control Test (S-03BB07) 14.2.12.1.18.1 Objectives To demonstrate the stability and response of the pressurizer level control system, including the verification of pressurizer level alarm and control functions. 14.2.12.1.18.2 Prerequisites
- a. Required component testing and instrument calibration are comploto,
- b. Required electrical power supplies and control circuits are operational, c .- The letdown and charging portions of the chemical and
(~' - volume control system are available to vary pressurizar s leve:.
- d. The plant is at norma, operating temperature and pres-sure, and hot functional testing is in progress, 14.2.12.1.18.3 Tect Method
- a. Pressurizer level is varied and the ability of the pressurizer level control system to automatically con-trol and stabilize pressurizer level is verified.
- b. pressurizer level is varied, and pressurizer level control system alarm and control setpoints are verified.
14.2.12.1.18.4 Acceptance Criteria L a. The response and stability of the pressurizer level l contrcl system are within design specifications.
- b. The pressurizer level control system alarm and control functions are within design specifications.
i p) 14.2-41 Rev. O {
WOLF CREEK 14.2.12.1.19 Pressurizer Heater and Spray Capability Test (503-B808) 14.2.12.1.19.1 Objectives To determine the electrical capacity of the pressurizer heaters, and the rate of pressure increase from the operation of all pres-curizer heate rs. 14.2.12.1.19 2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational,
- c. The plant is at ncrmal operating temperature and pres-sure with all reactor coolant pumps running, and hot functional testing is in progress.
14.2.12.1.19.3 Test Method
- a. Pressurizer heaters are energized, and heater capacity is calculated,
- b. With the pressurizer spray valves closed, all pressur-izer heaters are energized, and the time to reach a lll 2,300 psig system pressure is measured and recorded.
14.2.12.1.19.4 Acceptance Criteria
- a. The capacity of the pressurizer heaters is within design limits.
- b. The pressurizer pressure response to the actuation of all pressurizer heaters is within design limits.
14.2.12.1.20 Reactor Coolant System Flow Meacurement Test (SU3-BB09) At WCGS, Test S-07BB03 (USAR Section 14.2.12.3.5) is used to ratisfy the requirements for verification of design
-specifications.
14.2.12.1.21 Reactor Coolant System Flow Coastdown Test (SU3-BB10) llk 14.2-42 Rev. 0 l
WOLF CREEK O At WCGS, Test S-07BB04 (USAR Section- 14.2.12.3.6) _is used to satisfy the requirements for verification of design , specifications.- . 14.2.12.1.22 Reactor Coolant System Hydrostatic Test (S-03EBil) 14.2.12.1.22.1 Objectives To verify the integrity and leaktightness of the reactor coolant system and the high-pressure portions of associated systems. 14.2.12.1.22.2 Prerequisites
- a. Required system flushing / cleaning are complete.
- b. The reactor coolant pumps are available to support this test.
- c. The reactor vessel's lower internals, upper internals, filter assembly, and the closure head are installed.
The studs are tensioned to design value for the associ-ated hydrostatic test pressure ,' d. Temporary + temperature' instrumentation is_ installed for measuring the temperature of-the steam generator tube f"s sheets, the bottom of the pressurizer, and the closure flange of the reactor vessel,
- e. A charging pump-or test pump is available to pressurice the system.
- f. Required instrument calibration is complete.
14.2.12.1.22.3 Test Method The minimum temperature for pressurizing the system is_ established. The reactor coolant pumps are operated as required to establish the required temperature. The system is then pres-surized to test pressure,-and system welds, flanges, piping, and components are monitored for leakage. 14.2.12.1.22.4 Acceptance Criteria The reactor coolant system and associated high-pressure systems are verified leaktight in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, " Nuclear Components," through the Winter 1975 Addenda. 14.2.12.1.23 Pressurizer Continuous Spray Flow Verification Test ( SU 3-BB12 ) O 14.2-43 Rev. 0 4
WOLF CREEK O At WCGS, Test S-07BB05 (USAR Section 14.2.12.3.7) was used to satisfy the requirements for verification of design specifications. 14.2.12.1.24 Pressurizer Relief Valve and PRT Hot Preoperational Test ( S-0 3BB 13) 14.2.12.1.24.1 Objectives To demonstrate that the operating times of the pressurizer power-operated relief valves are within design specifications. The ability of the reactor coolant drain tank portion of the liquid radwaste system to cool down the pressurizer relief tank (PRT) at , the design rate is also verified. 14.2.12.1.24.2 Prerequ'isites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The PRT'is at a normal operating level and is aligned for normal operation.
- d. The liquid radwaste system is available to cool down the O
PRT via the reactor coolant draia tank heat exchanger,
- e. The plant is at normal operating temperature and pres-sure, and hot functional testing is in progress.
14.2.12.1.24.3 Test Method
- a. Pressurizer power-operated relief valves are operated, and opening times recorded.
. Following the operation of the pressurizer power-operated relief valves, the PRT is cooled down via the reactor coolant drain tank heat exchanger, and the cooldown rate is calculated and recorded.
14.2.12.1.24.4 Acceptance Criteria
- a. Power-operated relief valve operating times are within design specifications.
O 14.2-44 Rev. O
WOLF CREEK b.
.The reactor coolant drain tank portion of the liquid ,
radwuste. system cools down the -PRT.at a rate within k design specifications. 14.2.12.1.25 ~ Reactor Coolant Loop Vibration Surveillance Test (S-03BB14)_ > 14.2.12.1.25.1 Objectives To verify that the dynamic effects experienced during reactor coolant _ loop steady flow and reactor coolant loop pump transients as measured during _ hot functional testing (HFT) do not exceed acceptance criteria for the primary loop piping and components. 14.2.12.1.25.2 Prerequisites
- a. Hot functional-testing is in progress,
- b. Reference points for vibrational- measurement of the reactor coolant piping and components are established,
- c. . All subject systems are available for the specified dynamic operation,
- d. Required instrument calibration is complete.-
() 14.2.12.1.25.3 Test Method
-a. The systems are aligned for the specified dynamic operation.
- b. _The specified dynamic event is initiated and the reactor coolant piping and component responses are monitored.
14.2.12.1.25.4 Acceptance Criteria The measured deflections for each of the test measurement points are within a specified percent of the calculated reference deflections.- 14.2.12.1.26 Leak Detection System Preoperational Test (SU3-BB15A) 14.2.12.1.26.1 Objectives
- a. To determine, during hot functional testing, the amount of identified and unidentified leakage from the reactor coolant system and verify that the leakage is within design limits.
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l- 14.2-45 Rev. 0 l
I WOLF CREEK
- b. To demonstrate the ability to detect an increase in reactor coolant system leakage.
14.2.12.1.26.2 Prerequisites
- a. Required instrument calibration is complet
- b. Hot functional testing is in progress, and the reactor ecolant system is at normal operating temperature and pressure.
- c. The volume control tank contains an adequate supply of water to support this test.
- d. The reactor coolant drain tank and associated pumps are available to support this test.
14.2.12.1.26.3 Test Method
- a. The reactor coolant system identified and unidentified leakage rates are determined by monitoring the reactor coolant system water inventory,
- b. A known leakage rate is initiated, and the ability to detect an increase in lenkage is verified.
14.2.12.1.26.4 Acceptance Criteria
- a. Reactor coolant system identified and unidentified leakage is within design limits,
- b. The ability to detect an increase in reactor coolant system leakage is verified.
14.2.12.1.27 Leak Detection System Preoperational Test (SU3-BB15B) 14.2.12.1.27.1 Objectives
- a. To demonstrate the operation of the leak detection system and to verify the ability of the system to detect leakage within the required time limit as specified by design.
- b. The operation of the containment particulate and radio-active gas ~ monitoring portions of the Leak Detection System are verified in SU4-SP01, Process Radiation Monitoring System Preoperational Test.
O 14.2-46 Rev. O
WOLP CREEK 14.2.12.1.27.2 Prerequisites. '
- a. Required component testing and instrument calibrLtion:
are complete.
- b. The containment normal-sumps, instrument . tunnel sump, floor' drain tank, auxiliary building sump and associated pumps are available to support this test.
14.2.12.1.27.3 Test Method
- a. A known simulated leakage is initiated, and the ability of. the system -to detect the leakage within the design time is verified.
14.2.12.1.27.4- Acceptance. Criteria
- a. The ability of the leak detection system to detect a-leak within the design time is verified.
14.2.12.1.28 RTD/TC Cross Calibration (S-03BB16) 14.2.12.1.28.1 Objectives To provide a functional checkout of the reactor coolant system resistance temperature detectors-(RTDs) and incore thermocouples
.( ): and to- generate. isothermal cross-calibration data for subsequent
. correction factors to indicated temperatures.
14.2.12.1.28.2 Prerequisiter-a.- Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. Initial plant heatup, during hot functional testing, is in progress, and all reactor coolant . pumps are operating.
14.2.12.1.20.3 Test Method At various temperature plateaus, RTD and incore-thermocouple data are recorded. Isothermal cross-calibration correction factors for individual. thermocouples and the installation corrections for individual RTDs are determined. O 14.2-47 Rev. O
WOLF CREEK 14.2.12.1.28.4 Acceptance Criteria 4
- a. Individual RTD readings are within the design specifications.
- b. The installation corrections of the RTDs are within design specifications.
14.2.12.1.29 Chemical and Volume Control System Major Component Test (S-03BG01) 14.2.12.1.29.1 Objectives To demonstrate the operation of the centrifugal charging pumps and essociated minimum flow valves, including their response'to safety signals. 14.2.12.1.29.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The refueling water storage tank supply of demineralized water contains an adequate for the performance of lll this test,
- d. The component cooling water system is available to provide cooling water to the centrifugal charging pump oil coolers.
14.2.12.1.29.3 Test Method
- a. Centrifugal charging pumps are operated, and performance characteristics are verified.
- b. Centrifugal charging pump and minimum flow valve control logics are verified, including their response to safety signals.
14.2.12.1.29.4 Acceptance Criteria
- a. Centrifugal charging pump performance characteristics are within design specifications,
- b. Each centrifugal charging pump receives a start signal from the load sequencer.
O 14.2-48 Rev. O
.. .- - - _ . - ~ . -
WOLF CREEK --O- e. -If a safety injection signal-is present, a centrifugal charging puinp minimum flow valve will open if the asso-- ' ciated. pump flow is low and will close if the associated pump flow is above the minimum flow requirement of the
- pump.
14.2.12.1.30 seal Injection Preoperational Test (SU 3-BG0 2) 14.2.12.11.30.1- Objective To demonstrate the ability of the chemical and volume control system to supply adequate seal water injection flow to the reactor coolant ~ pumps and verify the operation.of the seal water return containment isolation valves, including their response to a CIS. 14.2.12.1.30.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The volume control tank contains an adequate supply of demineralized water for the performance of this test.
([ d. Cooling water is available to the charging pumps. 14.2.12.1.30.3 Test Method
- a. With a charging pump in operation, seal water throttle valves are adjusted to maintain the required flow- to each reactor coolant pump.
- b. Seal water return containment isolation valves control logics are verified, including their response to a CIS.
14.2.12.1.30.4 Acceptance Criteria
- a. Seal water injection flow to each reactor coolant pump is within design specifications.
- b. Seal water return containment isolation valves close on receipt of a CIS. Valve closure times are within design specifications.
O 14.2-49 Rev. O
WOLF CREEK k 14.2.12.1.31 Charging System Preoperational Test (SU 3-BG0 3) 14.2.12.1.31.1 Objective ! To demonstrate positive displacement charging pu.np operating l characteristics and to verify the operation of the regenerative l heat exchanger inlet isolation valves and the letdown isolation I valves, including their response to a safety injection signal (SIS). 14.2.12.1.31.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
l l b. Required electrical power supplies and control circuits are operational.
- c. The volume control tank contains an adequate supply of l demineralized water for the performance of this test.
l
- d. Cooling water is available to the positive displacement charging pump.
I
- e. The reactor coolant system is available to receive 1
charging system flow. lll l 14.2.12.1.31.3 Test Method 1
- a. The positive displacement charging pump is operated, and pump operating data are recorded.
! b. Regenerative heat exchanger inlet isolation valve and letdown system isolation valve control circuits are verified, including valve response to safety injection signals. 14.2.12.1.31.4 Acceptance Criteria
- a. Positive displacement charging pump operating charac-teristics are within design specifications.
l b. Charging pump to regenerative heat exchanger inlet isolation valves close on receipt of an SIS. Valve closure times are within design specifications,
- c. The letdown line containment isolation valves close on receipt of a containment isolation signal. Valve closure times are within design specifications.
O l 14.2-50 Rev. 0 l l l
l WOLF CREEK 'O 14.2.12.1 32 -Boron -Thermal Regeneration System Preoperational Test (SU3-BG04)'
-i 14.2.12.1.32.1- Objective
)
To verify the operation of-the boron thermal regeneration system, ) and associated control circuits. Performance characteristics of the chemical-and volume contt>l system chiller pumps are also verified. 14.2.12.1.32.2 Prerequisites
- a. Required component testing, instrument calibration and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The volume control tank-contains an adequate supply of domineralized water for the performance of this test,
- d. . The chemical and volume control system chiller surge tank contains an adequate supply of demineralized water for'the performance of this test.
14.2.12.1.32.3 Test Method
- a. The chemical and volume control system chiller pumps are operated and performance characteristics are verified.
- b. Boron thermal regeneration system component control circuits are verified.
14.2.12.1.32.4 Acceptance Criteria
- a. The chemical and volume control system chiller pumps' operating characteristics are within design specifica-tions.
- b. The chemical and volume.-control system chiller pumps start automatically when the boron thermal regeneration system is placed in the borate or dilute mode of opera-tion.
14.2.12.1.33 Boric Acid Blending System Preoperational Test (SU3-BG05) 14.2.12.1.33.1 Objectives
- a. To demonstrate the operating characteristics of boron injection makeup and boric acid transfer pumps and 14.2-51 Rev. O
WOLF CREEK verify the ability of the boric acid blending system to O make up at design flow ra':es to the chemical and volume control system (CVCS).
- b. To verify the operation of system component control circuits in all modes of operation.
- c. To demonstrate by flow test the ability of the reactor makeup water system to supply water to the boric acid blender.
- d. To demonstrate by flow test the ability of the boric acid system to supply an emergency boration flow to the charging pump suction,
- e. To verify the operation of volume control tank valves and associated control circuits, including valve response to safety signals.
14.2.12.1.33.2 Prerequisites
- a. Required component testing, instrument calibration, and system. flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
lll
- c. The reactor makeup water system is available to supply water to the boric acid blender and boric acid batching tank.
- d. A charging pump is available to receive and discharge flow from the boric acid transfer pumps.
- e. The volume control tank (VCT) contains an adequate supply of demineralized water for the performance of this test.
14.2.12.1.33.3 Test Method
- a. The boron injection makeup and boric acid pumps are operated, performance data recorded, and the ability of the system to make up to the CVCS at design flow rates is verified.
- b. System component control circuits are verified in all modes of operation.
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WOLF CREEK'
- c. .With _a reactor makeup water pump in operation, flow is "erified to'the boric acid blender and boric acid batch-ing tank.
- d. With both boric acid transfer pumps in operation and a charging pump taking a suction from the VCT and dis-charging to the reactor coolant loops, the emergency boration flow rate from the transfer pumps to the charg-ing pump suction is recorded.
- e. The emergency boration flow rate via gravity feed from the boric acid tanks to the charging pump suction is recorded.-
- f. Proper operation of the reactor makeup water system is verified when the reactor makeup control system (RMCS) is operated in the manual, dilute, alternate dilute, and automatic modes,
- g. The operation cf the VCT outlet valves control circuits is verified, including their response to a safety injec-tion signal.
14.2.12.1.33.4 Acceptance' Criteria r a. The boron injection makeup and boric acid transfer pump (_]/ operating characteristics are within design specifications.
- b. The flow rate to the boric acid blender from the reactor makeup water _ system is within design specifications,
- c. The emergency _ boration flow rates to the charging pump
. suction-are within design specifications,
- d. The: boric acid transfer-pumps and -the reactor makeup water pumps start automatically on a low level in the volume control tank when the RMCS is in the automatic mode,
- e. VCT outlet valves close on receipt of a safety injection signal when the- associated-charging pump _ supply valve from the refueling water storage tank is open.
- f. Refueling water storage tank'to charging pump suction valves open on receipt of a safety injection signal.
- g. The boric acid transfer pumps stop on receipt of.a load shed signal.
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WOLF CREEK
- h. The botic acid filter to charging pump valve supply breaker trips open on receipt of a load shed signal.
14.2.12.1.34 Chemical and Volume Control System Hot Preopera-tional Test (S-03BG06) 14.2.12.1.34.1 Objectives
- a. To determine by flow test that all letdown and cleanup flow rates are within design specifications,
- b. To determine, by comparison of boron concentrations, that boric acid addition to the reactor coolant system has occurred, using the normal and emergency flow paths,
- c. To determine by flow test the ability of the chemical and volume control system (CVCS) to make up at design flow rates and boron concentrations to the reactor coolant system in all modes of operation,
- d. To determine by operational test that the letdown con-tainment isolation valve closure times are within design specifications.
- e. To demonstrate the ability of the pump room coolers to maintain room temperatures within design limits.
14.2.12.1.34.2 Prerequisites O
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. ' Required electrical power supplies and control circuits are operational.
- c. The plant is at normal operating te.i.perature and pres-sure, and hot functional testing is in progress,
- d. The CVCS pump rooms are closed, and their associated pump room coolers are operational.
14.2.12.1.34.3 Test Method
- a. The letdown throttle valves are adjusted to establish letdown flow within design specifications,
- b. Boric acid addition to the reactor coolant system is verified, using the normal and emergency flow paths, by comparing the change in boron concentrations.
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WOLF, CREEK
/~T Xl c. With a_ charging pump in operation, the ability ~ of the
'CVCS, in all modes of operation, to make up at design
' flow rates and boron concentrations to the-reactor cool -
ant system is verified,
- d. With letdown flow established, the letdown containment isolation valves are operated, and operating times are recorded.
- e. During CVCS pump operation, pump room temperature data are recorded.
14.2.12.1.34.4 Acceptance criteria
- a. All letdown and cleanup- flow rates are within design specifications
- b. The boric acid addition system is capable of . adding boron- to the reactor coolant system via both the normal
-and_ emergency flow paths.
- c. The C7CS-makeup flow rates and boron- additions to the reactor coolant system are within design specifications in all modes of operation.
- d. The letdown containment isolation valves' closure
() are within design specifications. times
- e. The CVCS pump room coolers maintain the room temperature within design limits.
- f. The boron ther' mal regeneration system (BTRS) can vary the reactor coolant boron concentration au required for daily load cycle at 85 percent core life.
14.2.12.1.35- Fuel Pool Cooling and Cleanup System Preoperational Test (SU3-EC01) 14.2.12.1.35.1 Objectives
- a. To demonstrate the operating characteristics of the fuel pool cooling, fuel pool cleanup, and pool skimmer pumps and to verify that the associated instrumentation and controls are functioning properly,
- b. To verify that the fuel pool cleanup pump refueling water storage tank (RWST) suction isolation valves close on receipt of a safety injection signal (SIS).
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l l WOLP CREEK
- c. To verify that each fuel pool cooling pump room cooler starts when the associated fuel pool cooling pump starts.
14.2.12.1.45.2 Prerequisites
- a. Required component testing, instrument calibration, and e
system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. Cooling water is available to the fuel pool cooling and cleanup system heat exchangers.
- d. The liquid radwaste system is available to drain the refueling pool to the RWST.
- e. The essential service water system is availaole to provice cooling water to the spent fuel pool pump room Cool?rs.
- f. The spent fuel pool and fuel transfer canals are filled to their normal operating levels, 14.2.12.1.35.3 Test Method gg
- a. The fuel pool cooling, fuel pool cleanup, and pool skimmer pumps are operated in their various modes, and pump operating data are recorded.
- b. System component control circuits are verified, includ-ing the operation of system pumps and valves on receipt of safety signals.
- c. The ability of each fuel pool cooling pump room cooler to start when the associated fuel pool cooling pump starts is verified.
14.2.12.1.35.4 Acceptance Criteria
- a. The operating characteristics of the fuel pool cooling, fuel pool cleanup, and pool skimmer pumps are within design specifications,
- b. The fuel pool cleanup pumps RWST suction isolation valves close on receipt of an SIS.
O 14.2-56 Rev. O
4 WOLF CREEK
' ): c. -Each fuel pool cooling pump trips-on a pool-level signal, low spent ' fuel
- d. Each- fuel pool cooling pump trips on receipt of a load shed signal,
- e. Each fuel. pool cooling pump room cooler starts.when the associated fuel pool cooling-pump starts.
14.2.12.1.36 Spent Fuel Pool Leak Test (S-0 3EC0 2) 14.2.12.1.36.1 -objectives-
- a. To demonstrate the-integrity of the spent fuel pool, cask loading pit, and fuel transfer canal,
- b. To demonstrate the leaktightness of-the cask loading pit gate and the fuel transfer canal gate.
14.2.12.1.36.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The spent fuel pool is filled to the normal operating level.
- d. The cask loading pit level is below the level of the fuel pool gate.
- e. The fuel transfer canal level is below the level of the fuel pool gate,
- f. The reactor makeup water system is available to provide demineralized water to the spent fuel pool.
.g. A' source of compressed air is available to pressurice the system standpipes..
L 14.2.12.1.36.3 Test Method The. cask loading pit gate and fuel transfer canal gate are visually inspected for leakage. A leak test is performed on the spent fuel pool,-cask loading pit, and fuel transfer canal, using
.the associated leak chase standpipes.
l ( 14.2-57 Rev. O
WOLF CREEK 14.2.12.1.36.4 Acceptance No leakage is observed from the spent fuel pool, cask loading pit, fuel transfer canal, cask loading pit gate, and fuel transfer canal gate. 14.2.12.1.37 Essential Service Water System Preoperational Test (SU3-EF01) Test SU3-EF02 combined with Test SU3-EF01, Essential Service Water System Preoperational Test. 14.2.12.1.37.1 Objectives
- a. To demonstrate the capability of the essential service water system to provide cooling water flow during the LOCA mode of operation. The operation and response of system valves to align the system in the LOCA flow mode on safety injection signals, load sequence signals, and low suction pressure signals are also verified,
- b. To demonstrate the operating characteristics of the essential cervice water (ESW) pumps and verify their response to safety signals,
- c. To demonstrate the operability of the backpressure g control valves, including their response to safety w signals.
14.2.12.1.37.2 Prerequisites l a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
i l c. The compressed air system is available to the system air-operated valves. 14.2.12.1.37.3 Test Method
- a. System operating characteristics are verified in the LOCA mode of operation,
- b. Safety signals are simulated, and the responses of the system valves and the ESW pumps are verified.
- c. The ESW pumps are operated and pump operating data are recorded.
O 14.2-58 Rev. 0
WOLF CREEK
- d. The operability of the backpressure control valves, including their response to safety signals is verified.-
14.2.12.1.37.4 Acceptance Criteria
- a. Components supplied by the essential service water system receive- flows that are within design specifica-tions in the LOCA mode of system operation,
- b. System valve operation in response to safety signals is within design requirements,
- c. System- valve operating times- are within design specifications. -
- d. The ESW pumps' operating. characteristics are within design specifications. ,
- e. Each ESW pump responds properly to load sequence and load shed signals.
- f. The time required . fr r each ESW pump to reach rated flow is within design specifications, g.- System backpressure valves close upon receipt of a LOCA
. sequencer or safety injection signal.
- h. An auxiliary feedwater pump low suction pressure signal will close the ESW- pump breakers if a zero sequencer signal is not present.
14.2.12.1.38 . Component Cooling Water System Preoperational Test (S-03EG01) 14.2.12.1.38.1 Objectives
- a. To demonstrate the capability of the component cooling water system to provide cooli.g water during the normal, shutdown, and post-LOCA modes of operation.
b.- To demonstr9te the operating characteristics of the component cooling water pumps and to verify that the associated-instrumentation and controls are functioning _ properly, including system response to safety signals. 14.2.12.1.38.2 Prerequisites
- a. Required-component testing, instrument calibration, and system flushing / cleaning are complete.
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WOLF CREEK
- b. Required electrical power supplies and control are operaticnal.
circuits lh 11 2.12.1.38.3 Test Method
- a. System operating characteristics are verified in the normal, shutdown, and post-LOCA modes of operation.
- b. Safety signals are simulated, and the response of system pumps and valves is verified.
14.2.12.1.38.4 Acceptance Criteria
- a. The performance characteristics of each component cooling water pump are within design specifications,
- b. Components supplied by the component cooling water system receive flows that are within design specifications with the system operating in the normal, shutdown, and post-LOCA modes,
- c. Component cooling water pump and valve responses to load sequence, containment isolation, and safety injection signals are within design specifications,
- d. . Closuro times for the component cooling water sup.11y and return valves to the reactor coolant system are within ggg design specifications,
- e. Component cooling water pump response to centrifugal charging pump start signals is in accordance with system design.
14,2.12.1.39 Residual Heat Removal System Cold Preoperational Test ( S U 3-E3 01) 14.2.12.1.39.1 Ob]ective To demonstrate the operability of the Residual Heat Removal (RHR) pumps, demonstrate by flow test their ability to supply water at rated pressure and flow, and verify their response to safety signals. The operation of system motor-operated valves, including their response to safety signals, are also verified. The RWST control and alarm circuits are also verified. 14.2.12.1.39.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
14.2-60 Rev. 0 O l l
WOLF CREEK <s; (,); . b. Required electrical power supplies.and control circuits are operational,
- c. The reactor vessel head is removed and the water level is above the nozzles,
- d. The refueling water storage tank contains an adequate supply of demineralized water for the performance of-this test,
- v. Coclin; water is available to the RHR pumps and heat exchangers.
- f. The. instrument air system is available to nupply air to system air-operated valves.
14.2.12.1.39.3 Test'hothod
- a. Performance characteristics of-the RHR pumps are veri-fied during discharge to the reactor c/clant hot and cold loops and test recirculation,
- b. RWST and RHR system component control circuits .are verified, including the operation of the RHR pumps and system valees on receitu of safety signals.
(} 14.2.12.1.39.4 Acce p'.anco Criteria
- a. RHR pump performance characteristics are within design specifications.
- b. RHR system components -align or actuate in accordance with system design to safety injection, containment isolation, load sequencing, load shed, and tank level signals.
- c. .The time required for each RHR pump to-reach rated speed is within~ design' specifications.
- d. RHR system motor-operated valve closure times are within design specifications.
14.2.12.1.40 Residual Heat Pemoval System Hot Preoperational Te3t ( S-0 3-EJ0 2) 14.2.12.1.40.1 Objectives
- a. To demonstrate the ability of the residual heat removal (RHR) system to cool down the reactor coolant system (RCS) at its design rate.
D V 14.2-61 Rev. 0
WOLF CRIEK
- b. To demonstrate the ability of the RHR pump room coolers to maintain room temperature within design limits. lll 14.2.12.1.40.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The component cooling water system is supplying water to each RER heat exchanger,
- d. The RCS is being cooled down during hot functional testing.
- e. The RHR pump rooms are closed, and their associated pump room coolers are operational.
14.2.12.1.40.3 Test Method
- a. While the RCS is being cooled down with the RHR system, the heat transfer is obtained by performing a heat balance acrcss each RER heat exchanger,
- b. When RER pu=p room temperatures have stabilized, recc a temperature data is recorded. T 14.2.12.1.40.4 Acceptance Criteria
- a. The RER system is capable of cooling down the reacter coolant system at its design rate,
- b. The RER pump room coolers can maintain room temperature within design limits.
14.2.12.1.41 Safety Injection System Cold preoperational Test (SU3-EM01) 14.2.12,1.41.1 Objectives To demonstrate the response of the safety injection pumps and associated valves to safety signals. 14.2.12.1.41.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete 9
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~. - - . _ _ . ~ _ - - _ _ - , . _- - .- .. .
WOLF CREEK
- b. Required electrical power supplies and control circuits are operational. :
14.2.12.1.41.3 Test Method The response of the safety injection pumps and associated valves to safety signals is verified 14.2.12.1.41.4 Acceptance Criteria
- a. The safety injection pumps and associated valves align or actuate in accordance with system design to contain-ment isolation signals, load ehedding signals, and load sequencing signals.
14.2.12.1.42 . Safety Injection Flow Verification Test ( SU 3- EM0 2 ) 14.2.12.1.42.1 Objectives
- a. To demonstrate the operating characteristics of the safety injection pumps and the centrifugal charging pumps,
- b. . To demonstrdte the capability of the sefety injection pumps to provide balanced flow to the reactor coolant system and prevent runout ~ flow in the cold leg r.nd hot
-() leg injection modes,
- c. To demonstrate the capability of the charging pumps to provide balanced flow to the reactor coolant system and prevent runout flow in the boron injection mode.
- d. To demonstrate the capability of -the residual heat removal pumps to provice required-net positive suction head to- the safety injection pumps and the centrifugal charging-pumps.-
- e. To-demonstrate that the safety injection and centrifugal
. charging pump room coolers maintain room temperature within design limits.
- f. To demonstrate that associated system. valve operating times are within specified limits.
14.2.12.1.42.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
O 14.2-63 Rev. 0
r WOLF CREEK
- b. Required electrical power supplies and control circuits are operational,
- c. The CVCS is available to supply rated flow to the reactor coolant system via the boroa injection path, while simultaneously supplying other required loads.
- d. The residual heat removal system is available to supply adequate suction her.d to the safety injection and centrifugal charging pumps durie required injection modes.
- e. The borated refueling water storage tank contains an adequcte supply of demineralized water for this test,
- f. The reactor vessel is available to receive water, and the temporary reactor vessel pumpdown system is opera-tional (if required).
- g. The auxiliary building HVAC system is available to cool the pump rooms and verify associated pump interlocks.
- h. The accumulator Jafety injection system piping from the safety injection syotem to the reactor coolant system is available, and en accumulator tank is capable of receiv.ing water.
- 1. Cooling water is available to required pumps and heat O
exchangers..
- j. The c:moressed a;c system is avsilable to supply air to associated system valves.
- k. The residual hewt rnmoval system hot leg and cold leg flow orifices have been sized for required flow.
14.2.12.1.42.3 Test Method
- a. The safety injection pumpa are operated in the cold leg flow mode to verify pump performance characteristics and to identify the weaker pump.
- b. The safety injection cold leg branch lines are balanced using the weaker safety injection pump and the balance checked with the stronger pump. The balance is per-formed so that injection flow is maximized while pre-venting pump runout.
9 14.2-64 Rev. O
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I
WOLF CREEK
- c. The safety injection hot leg branch lines are balanced, using-their respective safety injection pump. The balance is performed so that injection flow is maximized while preventing pump runout.
- d. The centrifugal charging pumps are operated in the boron injection mode to determine pump performance character-istics and to identify the weaker pump.
The boron injection branch lines are balanced, using the weaker centrifugal charging pump and the balance checked with the stronger pump. The balance is performed such that. injection flow is maximized while preventing pump runout.
- f. Each residual heat removal pump is operated in series with the centrifugal charging pumps and safety injection pumps to verify that the residual heat removal pumps can supply adequate suction head.
- g. With'each centrifugal charging pump and safety-injection pump operating, pump room temperatures are allowed to stabilize, and room temperature data are recorded.
14.2.12.1.42.4 Acceptance Criteria
- a. The safety injection and centrifugal chstging pump respons.e times and valve operating times are within design specifications.
- b. The safety injection pump room-coolers start with their respective pump.
- c. The NPSH provided by the residual heat removal pumps to the centrifugal charging pumps and safety injection pumps is within system design specifications,
- d. Safety injection cold leg, hot leg, and safety injection pump flows are within design specifications,
- e. Boron injection and centrifugal charging pump flows are within design specifications,
- f. The safety injection and centrifugal charging pump room coolers can maintain room temperature within design limits.
O 14.2-65 Rev. 0
WOLF CREEK 14.2.12.1.43 Safety Injection Check Valve Test (SU 3-EM0 3) 14.2.12.1.43.1 Objectives To demonstrate the integrity of accumulator outlet line and loop safety injection line check valves and bac';up check valves by performing backleakage tests. The operability of the various safety injection line check valves under their design pressure conditions is also verified. 14.2.12.1.43.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The reactor coolant system is at normal operating pressure.
14.2.12.1.43.3 Test Method
- a. Check valve leak testing is performed with the reactor coolant system at normal operating pressure,
- b. Check valve operability is performed by verifying flow through the check valves at reduced reactor coolant lll pressure.
14.2.12.1.43.4 Acceptance Criteria
- a. Check valve leakage rates are within limits established by Technical Specifications Section 3.4.6.2f.
- b. Injection line check valve operability is demonstrated by verification of flow through the check valves in each of the safety injection lines to the reactor coolant system.
14.2.12.1.44 Boron Injection Tank and Recirculation Pump Test (SU3-EM04) This test has been deleted at Wolf Creek since the boron injection requirements have been eliminated due to the decrease in required boron concentration. ) 14.2-66 Rev. 0
WOLF' CREEK 14.2.12.1.45 Containment Spray System Nozzle Air Test ( S-0 3EN01) 14.2.12.1.45.1 Objectives To demonstrate that the spray nozzles in the containment spray header are clear of obstructions. 14.2.12.1.45.2 Prerequisites A source of compressed air is available to pressurize the spray headers. 14.2.12.1.45.3 Test-Method Air flow is initiated through the contair4 ment spray headers, and unobstructed flow is verified through each nozzle. 14.2.12.1.45.4 Acceptance Criteria All containment spray nozzles are clear and unobstructed, as evidenced by air passing through each nozzle. 14.2.12.1.46 Containment Spray System Preoperational Test (SU3-EN02) l 14.2.12.1.46.1 Objectives
. Ng-)g
- a. To demonstrate the operation of system components, including their response to safety signals, and verify ,
that the associated instrumentation and controls are i functioning properly. System flow characteristics in the test and simulated accident modes are also verified.
- b. To demonstrate the ability of the pump room coolers to maintain room temperatures within design limits.
14.2.12.1.46.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The refueling water storage tank contains an adequate supply of demineralized water- for the performance of this-test,
- d. The auxiliary building HVAC system is available to cool the pump rooms and verify associated pump interlocks.
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WOLF CREEK
- e. The containment spray pump rooms are closed. llI 14.2.12.1.46.3 Test Method
- a. Performance characteristics of the containment spray pumps are verified in the test mode, recirculating te the refueling water storage tank, and in the simulated accident mode.
- b. System component control circuits are verified, including the operation of system pumps and valves on receipt of load sequence / shedder and CSAS/CIS signals, respectively.
- c. During system operatiens, spray additive eductor opera-ting characteristics are verified,
- d. During containment spray pump operation, pump room temperature data are recorded.
14.2.12.1.46.4 Acceptance Criteria
- a. Col.tainment spray pump performance characteristics are eithin design specifications for the tested modes of operation,
- b. Containment spray pump and response sequence / shedder and CSAS/CIS valve is verified, to Iced and the ll) associated response times -a within design specifications.
- c. Spray additive eductor operating characteristics are within design specifications,
- d. The containment spray pump room es niers maintain tne room temperature within design limits.
14.2.12.1.47 Accumulator Testing (S-03EP01) 14.2.12.1.47.1 Objectives To determine the operability of each safety injection accumulator and obtain, by flow test, each accumulator's discharge line resis-tance to flow. The ability of the accumulator discharge line isolation valves to open under maximum differential pressure conditions is verified, as is the response of accumulator system valves to safety signals. h 14.2-68 Rev. O l l
WOLP CREEK () 14.2.12.1.47.2 Prerequisitos
- a. Required component testing, instrument calibration, ano system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The reactor vessel head and reactor internals are not installed, and the vessel is available to receive water.
- d. A source of compressed air and nitrogen is available,
- e. The refueling water storage tank contains an adequate supply of demineralized water for the performance of this test.
14.2.12.1.47.3 Test Method
- a. Each accumulator is filled and partially pressurized with the discharge valves closed. The discharge valves are opened, discharging the accumulators to the reactor vessel, and performance data are recorded.
- b. Each accumulator discharge line isolation valve is operated under maximum differential pressure conditions of normal accun:ulator precharge pressure and zero reactor O coolant pressure, and the valve operating *imes
. are recorded.
- c. Accumulator system valve control circuits are verified, including their response to safety injection and contain-ment isolation signals.
1/.2.12.1.47.4 Acceptance Criteria
- a. Each accumulator's discharge line resistance to flow (L/D) is in accordance with design specifications,
- b. Each accumulator's discharge line isolation valve openin, time under maximum differential pressure conditions is within design specifications.
- c. The accumulator system nitrogen supply containment isolation valve closes on receipt of a containment isc.lation signal. Valve closure time is within design specifications,
- d. Each accumulator discharge isolation valve opens on receipt of a safety injection signal.
O 14.2-69 Rev. O I
WOLF CREEK 14.2.12.1.48 Auxiliary Feedwater Pump Turbine Preoperational lI Test ( S U 3- FC 01) 14.2.12.1.48.1 Objectives
- a. To demonstrate the operation of the auxiliary feedsater pump (AFWP) turbine and its support equipment, while uncoupled from the pump.
- b. To demonstrate control of the APWP turbine from the control room as well as the auxiliary shutdown pcnol.
14.2.12.1.48.2 Prerequisites
- a. Required component testing, instrument calibration and system flushing / cleaning are complete.
- b. Steam is available to the AFWP turbine.
14.2.12.1.48.3 Test Method
- a. AFWP turbine system valves are operated and required response to various signala is verified.
- b. The turbine is operated and proper control is verified from the control room as well as the auxiliiry shutdown panel, and operating data are recorded.
ll)
- c. The turbine is brought to high speed at which time the mechanical and electronic overspeed trips are verified.
14.2.12.1.48.4 Acceptance Criteria
- a. The AFWP turbine can be controlled from the control room panel and the auxiliary shutdown panel.
- b. The mechanical and electronic overspeed trips actuate to shut down the turbine in accordance with the design.
14.2.12.1.49 Essential Service Wcter Pumphouse HVAC Preopera-tional Test (SU3-GD01) 14.2.12.1.49.1 Objectives
- a. To demonstrate the capacity of the essential service water (ESW) pumproom supply fans.
- b. To demonstrate ESW pumproom unit heater response to a load shed signal.
O 14.2-70 Rev. 0
i WOLF CREEK () 14.2.12.1.49.2 Prerequisites
- a. Required component testing and instrument calibration are completed.
- b. Required _ electrical power supplies and control circuits ..!
are operational. l
- c. The ESW pumphouse HVAC system is air balanced.
14.2.12.1.49.3 Test Methods l
- a. The ESW pumphouse supply fans are operated and flow data- I are recorded. l
- b. Response of the ESW pumproom unit heaters to load shed signal is verified.
14.2.12.1.49.4 Acceptance Criteria I
- a. The ESW pumphouse supply fan capacities are within design specification.
- b. A load shed signal will trip the ESW pumproom unit !
heaters' circuit breaker. 14.2.12.1.50 Miscellaneous Building HVAC System Preoperational O Tests (SU3-GF01, SU3-GF0 2, SU3-GF0 3) 14.2.12.1.50.1 -objectives To demonstrate the capacity of; 1) the auxiliary feedwater pump f room cooler fans, 2) the main steam enclosure _ building supply and
- exhaust f ans and 3) the tendon access gallery transfer fans and to i verify that the associated' instrumentation and controls are func- !
tioning properly.-The responses of the main steam enclosure build-ing-dampers and tendon access gallery dampers to safety. signals ' are also verified. (At Wolf Creek Generating Station, this test'was performed in three independent parts. In addition, the auxiliary boiler room fan was treatsd as part of preoperational test SU4-GF01.)
- 14 . 2.12.~ 1. 5 0. 2 Prerequisites J
- a. Required component testing and instrument calibration.
are complete.-
- b. Required electrical power supplies and control circuits i are operational.
10 14.2-71 Rev. 0
WOLF CRCEX
- c. The miscellaneous building HVAC system is air balanced.
14.2.12.1.50.3 Test Method
- a. Flow data are recorded while the fans are operating,
- b. The response of system dampers to a safety injection signal (SIS) is verified.
14.2.12.1.50.4 Acceptance Criteria
- e. System fan capacities are within design specifications.
- b. The main steam enclosure building and tendon access gallery dampers close on receipt of a SIS.
14.2.12.1.51 Fuel Building HVAC System Preoperational Test (S-03GG01) 14.2.12.1.51.1 objectives To demonstrate that the emergency exhauwt fans are capable of maintaining a negative pressure in the fuel building or the aux 11-iary building during accident conditions with the buildings iso-lated. To demonstrate the capacities of the fuel building supply unit fans, emergency exhaust fans, and the spent fuel pool pump room cooler fans. The operability of system instrumentation and controls, including the components' response to safety signals, is lll also verified. 14.2.12.1.51.2 Prerequisites
- a. Requited component testing, instrument calibration, and system air balancing are complete.
- b. Required electrical power supplies and control circuits i are operational.
- c. The compressed air system is available to supply the air-operated dampers in the fuel building.
- d. Required portions of the auxiliary building HVAC syste.n have been air balanced and are available to support this test.
16.2.12,1.51.3 Test Method
- a. With the fuel building closed, the system is operated in its normal configuration, and the fuel building supply O
14.2-72 Rev. O
WOLF CREEK () unit fan and spent fuel pool pump room cooler fan capac-ities are verified. l
' b. With a fuel building isolation signal (FBIS) present, the emergency exhaust fan capacities and negative fuel building pressures are verified.
- c. With a safety injection signal (SIS) present and the auxiliary . building isolated, the-emergency exhaust fan capacities and negative auxiliary building pressures are verified.
14.2.12.1.51.4 Acceptance criteria
- a. The auxiliary building and fuel building pressures maintained by the emergency exhaust fans are within
-der,ign specifications. !
- b. The fuel building supply fans, emergency exhaust fans, and spent fuel pool. pump room cooler fans' capacities !
are within design specifications. !
- c. The fuel building ventilation system fans and dampers
, properly respond to FBIS and SIS, in accordance with system design.
O- '14.2.12.1.52 Control Building HVAC' System Preoperational Test (S03-GK01) . -! 14.2.12.1.52.1 Objectivas To demonstrate the capacities of the control building supply air unit, control building e.1haust fans, access control exhaust fans, - control room pressurization fans, control room filtration fans, control room air conditioning units, access control fan coil units, counting room fan coil unit, and class IE electrical equip-ment ac units. To demonstrate that the control room pressuri:a-tion fans are capable of maintaining a positive pressure in the -! control room following a. control room ventilation isolation signal (CRVIS). The system instrumentation and-controls, including the '
-components' responses to safety signals, are also verified. To demonstrate that the ventilation to battery rooms 1 through 4 is in accordance with system design.
14.2.12.1.52.2 Prerequisites
- a. Required component testing, instrument calibration, and l system ait balancing are complete.
14.2-73 Rev. 0 . . . a x .. - . . - . . - . - . - . _ .. - .. - . .
)
WoLP CREEK j
- b. Regtired electrical power supplies and control circuits llh are eperational. )
I
- c. The :ompressed air system is available to supply air to system air-operated dampers. j 14.2.12.1.52.', Test Method l
- a. The control building system fans are operated, and fan capacities are verified.
- b. Proper response of system components to control room ventilation isolation signals (CRVIS) and safety injec-tion signals (SIS) is verified,
- c. With a CRVIS present, the ability of each control room pressurization tan to maintain the control room at a positive pressure is verified.
- d. The air flow to battery rooms 1 through 4 is verified.
14.2.12.1.52.4 Acceptance Criteria
- a. The. control building HVAC system fan capacities tre within design specifications,
- b. The control building HVAC system fans and dampers pro- ggg perly respond to CRVIS and SIS in accordance with system design.
- c. The control room pressure maintained by the control room pressurization fans is within design specification,
- d. The air flow to battery rooms 1 through 4 is in accor-dance with system design.
14.2.12.1.53 Auxiliary Building HVAC System Preoperational Test (SU3-GLOl) 14.2.12.1.53.1 objectives To demonstrate the capacities of the auxiliary building supply unit fans, auxiliary / fuel building norma) exhaust tans, the auxil-iary building fan coil units, pump room coolers, penetration room coolers, decon tank exhaust scrubber fans, access tunnel transfer fan, and penetration cooling fan. The system instrumentation and controls, including components' response to safety and fire sig-nals, are also verified. O 14.2-74 Rev. O
l l WOLY CREEK l (]) 14.2.12.1.53.2 Prerequisites
- a. Required component testing, instrument calibration, and system air balancing are complete.
- b. Required electrical power supplies and control circuits are operational,
- c. The compressed air system is available to supply the air-operated dampers in the auxiliary building.
- d. The fuel building HVAC system has been air balanced, and is available to support this test.
14.2.12.1.53.3 Test Method
- a. The system is operated in its normal configuration, and the system fan capacities are verified.
- b. Proper responses of system components to safety injec-tion and fire signals are verified.
14.2.12.1.53.4 Acceptance Criteria
- a. The auxiliary building fan capacities are within design specifications.
() b. The auxiliary building fans and dampers properly respond to safety injection and fire signals, in accordar.ce with system design. 14.2.12.1.54 Diesel Generator Building HVAC Preoperational Test (S-03GM01) 14.2.12.1.54.1 Objectives To demonstrate the capacities of the diesel generator room supply fans and to verify that the system instrumentation and controls function properly, including the response of fans and associated dampers to a diesel generator run signal and room temperature signals. 14.2.12.1.54.2 Prerequisites
- a. Required component testing and instrument calibration are completed,
- b. Required electrical power supplies and control circuits are operational.
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UOLF CREEK
- c. The diesel generator building HVAC system is air cal- g anced. W
- d. The respective diesel generator is not operating while the room is under test.
14.2.12.1.54.3 Test Method
- s. Flow data are recorded, while the diesel generator room supply fans are operating.
- b. The responses of the diesel generator room supply fans and exhaust dampers to a diesel generator run signal and to room temperature signals are verified.
14.2.12.1.54.4 Acceptance Criteria
- a. The capacities of the diesel generator room supply fans are within design specifications.
- b. The diesel generator room exhaust dampers open on re-ceipt of a diesel generator run signal,
- c. The diesel generator room supply fans start on a high room temperature signal and stop on a low room tempera-ture signal.
14.2.12.1.55 Containment Cooling System preoperational Test lll (SU3-GN01) 14.2.12.1.55.1 objectives To demonstr' ate the capacities of the hydrogen mixing, co n t a i n.ne n t cooling, and pressurizer cooling fans and verify their associated instrumentation and controls function properly, including fan rosponse to safety signals. 14.2.12.1.55.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The essential service water system is available to supply water to the containment coolers,
- d. The containment cooling system has been air balanced.
O 14.2-76 Rev. O
_ _ _ .~,._. _ _ _ ._ _ ._. _ ___ .._...__ _.._ _ _ _ . _ - - . _ _ _ _ _ _ _ _ I WOLP CREEK (]) 14.2.12.1.55.3 Test Method l l
- a. The hydrogen mixing, containmerit coolina, and press- '
urizer cooling fans are operatad, flow data recorded, and fan capacities calculated.
- b. The response of the hydrogen mixing and containment cooling fans to safety signals is verified.
14.2.12.1.55.4 Acceptance criteria i i
- a. 'The capacities of the hydrogen mixing, containment cooling, and pressurizer cooling fann are within design i specifications.
- b. The hydrogen mixing and containment cooling fans align i or actuate in response to safety injection, shutdown sequencer, and LOCA sequencer signels, in accordance with system design. :
14.2.12.1.56 CRDM Cooling Preoperational Test (5-0 3GN0 2) 14.2.12.1.56.1 Objectives To demonstrate. the- operating characteristics of the cavity cooling, . control rod drive mechanisn- (CRDM), and the elevator
; machiae roc 4n exhaust fans and verify their associated instrumen- . tation and controls, including their response to safety signals.
14.2.12.1.56.2 Prerequisites
- a. Required component testing and instrument calibration.
are complete.
.b. Required electrical power supplies and control circuits are operational.
- c. The CRDM and cavity cooling portions of the containment t cooling system are air balanced.
14.2.12.1.56.3 Test Method
- a. The cavity cooling, elevator machine room exhaust, and CRDM _ fans are operated, flow data recorded, and fan capacities calculated.
- b. .The response of the CRDM fans to a safety injection signal is verified.
l (- 14.2-77 Rev. 0
WOLF CREEK 14.2.12.1.56.4 Acceptance Criteria ll)
- a. The capacities of the cavity cooling, elevator machine i room exhaust, and CRDM fans are within design specifica-tions,
- b. The appropriate CRDM fans supply breakers open on re- l ceipt of a safety injection signal. l 14.2.12.1.57 Integrated containment Leak Rate Test (SU3-GP01) !
14.2.12.1.57.1 Objective To demonstrate that the total leakage from the containment does not exceed the maximum allowable leakage rate at the calculated peak containment internal pressure. The operability of the con-tainment cooling fans at design accident pressure is also veri-fled. 14.2.12.1.57.2 Prerequisites
- a. The containment penetration leakage rate tests (type B tests) and containment isolation valve leakage tests (type C tests) are complete and the containment has been pressurized to 115 percent of the design pressure,
- b. All containment isolation valves are closed by normal actuation methods,
- c. Containment penetrations, including equipment hatches and personnel airlocks, are closed,
- d. Portions of fluid systems that are part of the contain-ment boundary, that may be opened directly to the con-tainment or outside atmosphere under post-accident conditions, are opened or vented to the appropriate atmosphere to place the containment in as close to post-accident conditions as possible,
- e. Required instrument calibration is complete.
14.2.12.1.57.3 Test Method
- a. The integrated containment leak rate test (type A test) is conducted, using the absolute method, described in the ANSI /ANS 56.8-1981 Containment System Leakage Testing Requirements. Measurements of containment atmosphere dry-bulb temperature, dew point and pressure are taken to calculate the leakage rate. A standard O
14.2-78 Rev. O
l WOLP CREEK statistical analysis of data conducted, using a O- linear least squares fit regression analysis to calcu-is late the leakage rate.
- b. On completion of the leak rate test, a verification test '
is conducted to confirm the capability of the data ' acquisition and reduction system to satisfactorily determine the calculated integrated leakage rate. The verification test is accomplished by imposing a known leakage rate on the containment, or by pumping back a known quantity of air into the containment through a calibrated flow measurement device. ,
- c. While at the design accident pressure, data is recorded for the containmen,t cooling fans. '
14.2.12.1.57.4 Acceptance Criteria The containment integrated leakage does not exceed the maximum allowable leakage rate at a calculated peak containment internal pressure, as defined in 10 CFR 50, Appendix J.- The containment cooling fan operation at design accident pressure , is in accordance with design.
-14.2.12.1.58. Reactor Containment Structural Integrity Acceptance Test (SU3-GP02) .O 14.2.12.1.58.1 Objectives To demonstrate the structural integrity of the reactor containment building.
14.2.12.1.58.2 Prerequisites a.- Containment penetrations are installed, and penetration leak tests are completed.
- b. Containment penetrations, including equipment hatches and personnel airlocks, are closed.
- c. Required' instrument calibration is complete.
14.2.12.1.58.3 Test Method The containment is pressurized at 115 percent of the design pres-sure, and deflection measurements and concrete crack inspections are made to determine that the actual structural response is within the limits predicted by the design analyses. . O 14.2-79 Rev. 0
WOLF CREEX 14.2.12.1.58.4 Acceptance Criteria ggg The containment structural response is within the limits predicted by design analyses. 14.2.12.1.59 Post-Accident Hydrogen Removal System Preopera-tional Test (S-03GS01) 14.2.12.1.59.1 Objectives
- a. To demonstrate that the hydrogen recombiner performance characteristics are within design specifications,
- b. To determine the operation of system dampers and valves, including the response of hydrogen purge and hydrogen monitoring containment isolation valves to a CIS.
- c. To demonstrate the operability of the hydrogen analyzers and their ability to sample the containment atmosphere.
14.2.12.1.59.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
14.2.12.1.59.3 Test Method
- a. Performance characteristics are recorded, while the hydrogen recombiners are operating.
- b. System valve and damper control circuits are verified, including the response of hydrogen purge and hydrogen monitoring containment isolation valves to a CIS.
- c. The hydrogen analyzers are operated, and performance data recorded.
16.2.12.1.59.4 Acceptance Criteria
- a. Hydrogen recombiner performance characteristics are within design specifications,
- b. Hydrogen purge and hydrogen monitoring containment isolation valves close on receipt of a CIS. Valve closure times are within design specifications.
O 14.2-80 Rev. O
l l i WOLF CREEK () 14.2.12.1.60 containment Purga System HVAC (S-03GT01) Preoperational Test 14.2.12.1.60.1 Objectives , To demonstrate the capacities of the containment minipurge supply ! and exhaust, shutdown purge supply and exhaust, and containment atmospheric control fans. The operation of system instrumentation and controls, including the response of system fans and dampers to t safety signals, is also verified. 14.2.12.1.60.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The containment purge HVAC system has been air balanced,
- d. The compressed air system is available to supply air te system valves and dampers.
~14.2.12.1.60.3 Test Method
- a. The containment minipurge supply and exhaust, shutdown
(~)g (_ purge rupply and exhaust, and containment atmospherie control fans are operated, flow data recorded, and fan capacities calculated,
- b. The response of system fans and dampers to safety sig-nals is verified.
14.2.12.1.60.4 Acceptance Criteria
- a. The capacities of the containmcat minipurge supply and exhaust, shutdown purge supply and exhaust, and contain-ment atmospheric control fans are within design specifi- ;
cations,
- b. System fans and dampers align or actuate in response to containment purge isolation and safety injection sig-nals, in accordance with system design. Damper closure times are within design specifications.
(:) l 14.2-81 Rev. 0
WOLF CREEK 14.2.12.1.61 Guseous (S-03HA01) Radwaste System Preoperational Test lll 14.2.12.1.61.1 Objectives
- a. To demonstrate the performance characteristics of the gas decay tank drain pump, vaste gas compressors, and catalytic hydrogen recombiners, including their response to safety signals.
- b. To verify the operability of system valves, including the response of the waste gas discharge val'/a to a high-radiation signal.
- c. To verify- that system instrumentation and controls function properly.
14.2.12.1.61.2 Prerequisites
- a. Required-component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The: component cooling water system is available to supply cooling water to the waste gas compressors and a catalytic hydrogen recombiners. W
- d. The service gas system is available to provide nitrogen, hydrogen, and oxygen to the catalytic hydrogen recom-biners, e .- The reactor' makeup water system is available to provide water to the waste gas compressors, catalytic hydrogen recombiners, and the waste-gas decay tank drain header.
14.2.12.1.61.3 Test Method
- a. Performance characteristics of the gas decay tank drain pump, and waste-gas compressors are verified,
- b. Hydrogen in introduced to the system and' the catalytic hydrogen recombiners performance are verified.
- c. System . component control- circuits are verified, including component response to-safety signals.
14.2.12.1.61.4 Acceptance Criteria
- a. Performance characteristics of the gas decay tank- drain L
pump, vaste gas compressors, and catalytic hydrogen recombiners are within design specifications. lll , 14.2-82 Rev. 0 l I
WOLF CREEK () b.- The waste as discharge high-radiat on signal. valve automatically closes on a c., The waste gas compressors trip on a high-high or low-low moisture separator level, high or low moisture separator - pressure, low compressor suction pressure, or low com- , ponent cooling water flow.
- d. The hydrogen recombiner oxygen feed valve closes on high-high hydrogen concentration in the recombiner feed, high-high _ oxygen concentration in the recombiner dis-charge, high cooler-condenser discharge temperature, high-high recombiner discharge temperature, low-low
.recombiner flow, and high-high recombiner reactor inlet temperature. i
- e. The hydrogen recombiner oxygen feed valve signal is blocked on high oxygen concentration in the recombiner feed and high catalyst bed temperature,
- f. The volume control tank vent valve closes on a hydrogen recombiner trip, low volume control tank and low waste gas compressor suction pressure. pressure, .
14.2.12.1."62 Emergency Fuel Oil System Preoperational Test 14.2.12.1.62.1 ec vos ' To demonstrate the capability of the system to provide an adequate fuel supply to the emergency diesel generator fuel oil day tanks and verify that the associated instrumentation and controls arn ' functioning properly. le 2.12.1.62.2 Prerequisites
- a. Required component te. ting, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are-operational. ^.
14.2.12.1.62.3 Test Method
- a. Fuel oil is transferred from the fuel oil storage tank to the fuel oil day tanks by means of the transfer pumps. Flow and pressure characteristics are recorded,
- b. Fuel oil day tank levels are varied to verify the trans-fer pump' automatic operations.
O . 14.2-83 Rev. O _ , _ _ _ . _ . _ _ . , _ _ -_ , _ . . _ _ ~ , - . _ _ _ . _ , _ _ _ . . _ . _ . _ - _ . _ _ . _ . - . .
WOLF CREEK
- c. Response to fire and emergency diesel generator start signals are verified.
14 . 2~.12 .1. 6 2 . 4 Acceptance criteria
- a. The transfer pump flow capacity is verified for later comparison to the fuel consumption rate (S-03NF02).
- b. Control circuit automatic operation from fuel oil day cank levels, fire signals, and diesel generator start signals is within design specifications.
14.2.12.1.63 Spent Fuel pool Crane preoperational Test (SU3-KE01) 14.2.12.1.63.1 Objectives
- a. To demonstrate proper operation of the spent fuel pool bridge crane control circuits and associated interlocks.
- b. To document the data obtained during testing of the spent fuel pool bridge crane at 125 percent of rated load.
- c. To verify the ability of the spent fuel pool bridge crane and associated fuel handling tools to transfer a dummy fuel assembly.
e T 14.2.12.1.63.2 prerequisites
- a. Required component testing and ins.trument calibration are completed.
- b. Required electrical power supplies and control circuits are operational,
- c. A dummy fuel assembly is available.
14.2.12.1.63.3 Test Method
- a. Operability of the spent fuel pool bridge crane control circuits and associated interlocks is verified.
- b. Ability of the spent fuel pool bridge crane and associa-ted fuel handling tools to transfer a dummy fuel assem-bly is verified.
O 14.2-84 Rev. 0 , l
WULF CREEK 14.2.12.1.63.4 Acceptance Criteria
- a. The spent fuel pool bridge crane electric and manual hoists support 125 percent of their rated load,
- b. The spent fuel pool bridge crane monorail center span deflection at rated load is within design specifica-T. ions.
- c. The spent fuel pool crane bridge, trolley and hoirt speeds at rated loads are within design specifications,
- d. All :ontrol circuits and interlocks associated with the spent fuel pool bridge crane operate in accordance with system design.
- e. While transferring a dummy fuel assembly, the spent fuel pool bridge crane and associated fuel handling tools operate in accordance with oystem design.
14.2.12.1.64 New Puel Elevator Preoporational Test (SU3-KE02) 14.2.12.1.64.1 Objectives
- a. To demonstrate proper operatf or. of the new fuel elevator control circuits and associated interlocks.
O b. To verify the ability of the new fuel elevator to and lower a dummy fuel assembly. raise 14.2.12.1.64.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. A dummy fuel assembly is available.
14.2.12.1.64.3 Test Method operability of the new fuel elevator including control circuita and associated interlocks is verified. 14.2.12.1.64.4 Acceptance Criteria
- a. All control circuits and interlocks associated with the new fuel elevator operate in accordance with system design.
O 14.2-85 Rev. O
WOLF CRdEK
- b. While raising and lowering a dummy fuel assembly, the g new fuel elevator operates in accordance with system W design.
14.2.12.1.65 Fuel Handling and Storage Prooperational Test ( SU 3-KE0 3) 14.2.12.1.65.1 Objectives
- a. To verify the ability of the spent fuel cask handling crane, and associated fuel handling tools to transfer a dummy fuel assembly,
- b. To demonstrate proper operation of the spent fuel cask handling crane control circuits and associated inter-locks.
- c. To document the data obtained during testing of the spent fuel essk handling crane at 125 percent of rated load.
14.2.12.1.65.2 Prerequisites
- s. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. A dummy fuel assembly is availsole.
14.2.12.1.65.3 Test Methed
- a. During the transfer of a dummy fuel assembly, the opera-bility of the spent fuel cask handling crane and assoc-iated fuel handling tools is verified.
- b. Operability of the spent fuel cask handling crane con-trol circuits and associated interlocks is verified.
14.2.12.1.65.4 Acceptance Criteria
- a. While transferring a dummy fuel assembly, the spent fuel cask handling crane and associated fuel handling tools operate in accordance with system design,
- b. All control circuits and interlocks associated with the spent fuel cask handling crane operate in accordance with system design.
O 14.2-86 Rev. O
WOLF CREEK The spent fuel cask hand!ing crane O c. percent of rated load. hoist supports 125
- d. The spent fuel cask handling crane bridge conter span deflection at rated load is within design specifica-tions, h
- e. The spent fuel cask handling crane bridge, trolley and ;
hoist speeds at rated loads are within design specifica- : 3 tions. d; ; i; , 14.2.12.1.66 Fuel Transfer System l'r e ope r a t i o n a l Test 3 (SU3-KE04) 14.2.12.1.66.1 Objectives
- a. To demonstrate proper operation of the fuel transfer system control circuits and associatted interlocks.
- b. To verify the ability of the fuel transfer system and associated handling tools to transfer a dummy fuel assembly.
14.2.12.1.66.2 Prerequisites l
- a. Required component testing and instrument calibration are complete.
O b. Required electrical power supplies and control circuits are operational,
- c. A dummy fuel assembly in available.
14.2.12.1.66.3 Test Method __-
- a. Operability of the fuel transfer system control circuits and associated interlocks is verified. Es
- b. During the transfer of a dummy fuel assembly, the opera-bility of the fuel transfer system and associated hand-ling tools is verified.
14.2.12.1.66.4 Acceptance Criteria
- a. All control circuits and interlocks associated with the fuel transfer system operate in accordance with system design.
O 14.2-87 Rev. 0 __-___m__.__.---- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' " ' - ^ - - ' - - - - - " - ' - - ' ' ' ' - - - - " - - - - - ^ ^ ~ - - - - - - - ' - - - - - --
WOLF CREEK While transferring a dummy fuel assembly, b. transfer system and associated handling tools operate in the fuel ll) accordance with system design. i 14.2.12.1.67 Refueling Machine and RCC Change Fixture Proopera-tional Test (SU3-KE05) 14.2.12.1.67.1 Objectives h
- a. To demonstrate proper operation of the refueling ma-I chine, rod cluster control change fixture and contain-ment building polar crane control circuits and associa-ted interlocks.
- b. To document the data obtained during testing of the containment building polar crane at 125 p4Ecent of rated load.
- c. Jo verify the ability of the refueling machine to trans-fer a dummy fuel assembly, 14.2.12.1.67.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits a are operational. W
- c. A dummy fuel assembly is available.
- d. A dummy control rod assembly is available.
1s.2.12.1.67.3 Test Method
- a. Operability of the refueling machine and rod cluster control change fixture control circuits and associated bridge, trolley, hoist and gripper interlocks is veri-fled,
- b. Opertbility of the containment building polar crane control circuits and associated interlocks is verified.
14.2.12.1.67.4 Acceptance Criteria
- a. All control circuits and interlocks associated with the refueling mtchine and rod cluster control change fixture operate in accordanie with system design.
O 14.2-88 Rev. 0 m _ _ _ _ . _ _ - --
WOLF CREEK f^)
- b. The control ;ircuits and interlocks associated with the containment building polar crane operate in accordance with system dosign.
- c. The containment polar crane main and auxiliary hoists support 125 percent of their rated load,
- d. The containment polar crane bridge center span deflec-tion at rated load is within design specifications.
- e. The containment polar crane bridge, trolley, and hoist speeds at rated loads are within design specifications,
- f. While transferring a dummy fuel assembly, the refueling machine operates in accordance with system design.
14.2.12.1.68 Refueling Machine Indexing Test (S-03KE06) 14.2.12.1.68.1 Objectives
- a. To verify the indexing of the refueling machine and establish bridge rail reference points for future opera-tions.
- b. To demonstrite the ability to transfer the dummy fuel assembly to the reactor vessel, fx
(_) 14.2.12.1.68.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational,
- c. A dummy fuel assembly is available.
14.2.12.1.68.3 Test Method
- a. While transferring a dummy fuel assembly with the re-fueling machine, the bridge rail is marked at key trano-fer points.
14.2.12.1.68.4 Acceptance Criteria
- a. The refueling machine can load a dummy fuel assembly in each of the reactor vessel fuel loading locaticns.
14.2.12.1.69 Puel Handling System Integrated Preoperational Test (SU3-KE07)
!v) 14.2-89 Rev. O
UOLF CREEK 14.2.12.1.69.1 Objective (g) To verify the ability of the ref ueling machine, new fuel ele rator, fuel transfer system, spent fuel b-!dge crano, spent fuel cask handling crane and associated fue? handling tools to transfer a dummy fuel assembly. 14.2.12.1.69.2 Prerequisites 1
- a. Required component testing and instrument calibration I are complete.
- b. Required electrical power supplies and control circuits j are operational. -
- c. The reactor vessel, refueling pool, refueling canal and spent fuel pool are filled with demineralized water,
- d. A dummy fuel assembly is available.
14.2.12.1.69.3 Test Method During the transfer of a dummy fuel assembly, the operability of the refueling machine, new fuel elevator, fuel transfer system, spent fuel bridge crane, spent fuel cask handling crane ano asso-ciated fuel handling tools is verified. 14.2.12.1.69.4 Acceptance Criteria I While transferring a dummy fuel assembly, the refueling machine, now fuel elevator, fuel transfer system, spent fuel bridge crane, spent fuel cask handling crane and associated fuel handling tools operate in accordance with system design. 14.2.12.1.70 Diesel Generator Mechanical Preoperational Test (S-03KJ01) 14.2.12.1.70.1 Objectives
- a. To demonstrate the performance characteristics of the diesel generators and associated auxiliaries, and verify that each diesel reaches rated speed within the required time.
- b. To verify the operability of all control citeuits asscc-lated with the diesel generator and diesel auxiliarios, including the control circuits response to safety sig-nals.
O 14.2-90 Rev. O
'10LP CREEK ;
i
- c. To demonstrate the capability of each air storage tank to g-g ) provide five diesel cranking cycles without being re- !
charged. l'4.2.12.1.70.2- Prerequisites '
- a. Required component testing, instrument calibration, and [
system flushing / cleaning are complete.
~
- b. Requirod electrical power supplies and control circuits '
are operational.
- c. The essential service water system is available to i provide cooling water to the diesel engine intercooler r heat exchanger.
- d. The emergency fJe1 oil system is available to provide fuel oil to the dlesel generators.
- e. The fire protection system is available to support this !
test. 14.2.12.1.70.3 Test Method-
- a. The diesel generators are started, and the time required to reach rated speed is recorded.
O L. With the diesel generators and associated- auxiliaries operating, performance characteristics are verified,
- c. The operability of all control circuits associated with the diosol generator and diesel auxiliaries, including the control circuits' response to safety signals, is
-verified.
- d. The ability of each air storage tank to provide _.five diesel cranking cycles, without being recharged, is verified.
14.2.12.1.70.4 Acceptance Criteria
- a. The time required for each diesel generator'to reach-rated speed is within design specifications. '
- b. The performance characteristics of the diesel generators and associated auxiliaries are-within design specifica-tions.
- c. Each diesel generator starts automatically on receipt of
- a safety injection signal or a bus under-voltage signal. ,
O 14.2-91 Rev. 0 ,,--y- . , ,g -- . . , - - . - w --e.y-.v.. ---..-,,,---m,.w._ .w,,,,,,,.:,, ,,,w.w_,.we,, ,,n-m+,,.e-,,-aw-m *-=-e+mn -v-*es--
WOLF CREEK
- d. Each diesel generator trips automatically on receipt of each of the following signals lll Lube oil pressure low Jacket coolant temperature high Crankcase pressure high Start failure Engine overspeed Diesel generator ground overcurrent Diesel generator differential current
- e. The diesel generator neutral ground overcurrent trip signal is bypassed when the diesel generator is opera-ting in the emergency mode,
- f. Each air storage tank is capable of providing five diesel cranking cycles, without being recharged.
- g. Each starting air compressor has the ability to charge its respective air tank from minimum to normal pressure within the required time.
14.2.12.1.71 4160-V (Class IE) system Preoperational Test (S-0 3NB01) 14.2.12.1.71.1 objectives
- a. To demonstrate that the 4,160-V Class IE busses can enetgized from their normal and alternate sources, be g
- b. To verify that a 4,160-V Class IE bus digital undervol-tage cignal trips the associated incoming feeder breakers.
- c. To verify that a degraded bus voltage condition will trip the associated incoming feeder breakers,
- d. To verify proper operation of system instrumentation and alarms.
14.2.12.1.71.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
14.2-92 Rev. O O
i WOLF CREEK () 14.2.12.1.71.3 Test Method
- a. The 4,160-V Class it busses are energized from their !
normal source, and bus voltages are recorded. :
- b. The 4,160-V Class !E busses are energized from their alternate source, and bus voltages are recorded. i t
- c. 4,160-V Class IE bus undervoltage signals are simulated, :
and proper operation of the 4,160-V Class It feeder . breakers is verified. ! 14.2.12.1.71.4 Acceptance criteria
- a. The voltage of each 4,160-V Class It bus, when supplied
, from its normal source, is within design specifications.
- b. -The voltage of each 4,160-V Class IE bus, when supplied from its_ alternate source, i s within design specifica-tions. i
- c. A 4,160-V Class IE bus digital undervoltage signal will trip the appropriate bus incoming feeder breakers.
- d. A degraded voltage condition on either 4,160-V Class - IE
.. bus will cause an alarm and, if it continues, trip the appropriate bus incoming feeder breakers.
[]}- , e.- A degraded voltage condition on either 4,160-V Class It bus coincident with a safety injection actuation signal will immediately trip the bus incoming feeder breakers. l 14.2.12.1.72 Diesel Generator Electric Preoperational Test (S-0 3tG01) 14.2.12.1.72.1 objectives a. To demonstrate that each diesel generator is capable of 35 consecutive valid starts with no failure.
- b. To demonstrate the ability of each diesel generator to carry the design ~ load for the time required to reach equilibrium temperature plus 1 ho'st, without exceeding design limits.
.c. To demonstrate the ability of each diosel generator to attain and stabilize frequency and voltage within the design limits and time.
O 14.2-93 Rev. 0
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l WOLF CREEK
- d. To demonstrate the capability of each diesel generstor 3 to withstand a full-load rejection without exceeding W speeds or voltages that cause tripping or damage.
- e. To demonstrate the operability of each diesel generator feeder breaker and associated interlocks.
- f. To demonstrate the ability of the diesel cooling water system to maintain the diesel temperature within design specifications, while the diesel generators are opera-ting at full load.
- g. to demonstrate the ability of each diesel generator to start and shed the largest single motor while supplying all other sequenced loads, maintaining voltage and frequency within design limits.
14.2.12.1.72.2 Prerequisites
- a. Required component testing and instrument ct.libration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The essential service water system is available to provide cooling water to the diesel generator inter-cooler heat exchanger. lll
- d. The emergency fuel oil system is available to provide fuel oil to the diesel generators.
- e. The fire protection system is available to support this test.
- f. The 4.16-kV busses are available for loading to support this test.
14.2.12.1.72.3 Test Method
- a. The ability of each diesel generator to undergo 35 consecutive starts with no failure is verified.
- b. The ability of each diesel generator to carry the design load for the time required to reach equilibrium tempera-ture, plus 1 hour, without exceeding design limits, is verified.
O 14.2-94 Rev. 0
F.-- WOLF CREER
- c. The ability of each diesel generator O to stabilire frequency and voltage within the design limits and time-is verified.
attain and
- d. The ability of each diesel generator to withstand a full-load rejection, without exceeding speeds or vol-tages that cause tripping, is verified.
- e. The operability of each diesel generator feeder breaker and associated interlocks is verified.
- f. While operating each diesel generator at full-load conditions, the ability of the diesel cooling water system to maintain diesel temperatures within design specifications is' verified.
- g. The ability of each diesel generator to start and shed !
the largest- fully loaded single motor while supplying
- all other sequenced loads and maintain voltage and ,
frequency within design limits is verified. ' 14.2.12.1.72.4 Acceptance criteria '
- a. Each diesel generator is capable of carrying the design load for the time required to reach equilibrium tempera-ture, plus 1 hour, without exceeding design limits.
O-
-v
- b. Each diesel' generator can attain and stabilize frequency and voltage within design limits and time,
- c. ~ Each diesel generator is capable of withstanding a full- '
load rejection without exceeding speeds or voltages that cause tripping. e
- d. When a diesel generator is operating in the nor. emergency (test) mode, the associated diesel generator feeder breaker trips on receipt of any of the following sig-
- nals:
Generator overcurrent : Reverse power Loss of field Underfrequency
- e. The diesel generator stops and the associated diesel '
generator feeder breaker trips on receipt of any of the following signals: l Generator differential current-Neutral ground overcurrent l I 14.2-95 Rev. 0
WOLF CREEK
- f. When a diesel generator is operating in the emergency mode, the following trip signals are bypassed: g Neutral ground overcurrent Generator overcurrent Reverse power Loss of field Underfrequency
- g. Each diesel generator cooling water systam, with the diesel generators operating at full-load, maintains the diesel temperatures within design specificat: ions,
- h. Each diesel generator has the capability of starting and
<5edding the largest fully loaded single motor while supplying all other sequenced loads, maintaining voltage and frequency within design limits.
- 1. Diesel generators are capable of 35 consecutive valid starts with no failure.
14.2.12.1.73 Integrated Control Logic Test (SU3-NF01) 14.2.12.1.73.1 Objectives
- a. To demonstrate that the actuation of the LOCA sequencer, shutdown sequencer, safety-related load shed, and non-safety-related load shed circuits on receipt of the appropriate undervoltage, safety injection, containment spray actuation, diesel generator breaker position, and normal and alternate 4,160-V feeder breaker position signals is in accordance with system design.
- b. To demonstrate that the LOCA sequencer, shutdown se-quencer, safety-related Icad shed, and nonsafety-related load shed circuits shed and sequence loads -n accordance with system design.
14.2.12.1.73.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operatinnal.
14.2.12.1.73.3 Test Method
- a. Undervoltage, safety injection, containment spray actua-tion, diesel generator breaker position, and normal and O
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WOLF CREEK () altarnate 4,160-V feeder breaker position signals are iro.t. Tted , and the actuation of the LOCA sequencer, s:m* A: vn sequencer, safety-related load shed, -and non-safeu -related load shed circuits is verified,
- b. Signals are initiated to actuate the 14CA sequencer, shutdown sequencer, safety-related load shed, and non-safety-related load shed circuits, and proper load shed and load sequencing are verified.
14,2.12.1.73.4 Acceptance criteria
- a. Actuation of the LOCA sequencer, shutdown sequencer, safety-related load shed, and nonsafety-related load shed circuits on receipt of under-voltage, safety injec-tion, containment spray actuation, diesel generator breaker position, and normal and alternate 4,160-V feeder breaker position signals is in accordance with system design.
- b. The LOCA sequencer, shutdown sequencer, safety-related load shed,.and nonsafety-related load shed circuits shed and sequence loads in accordance with system design.
14.2.12.1.74 LOCA Sequencer Preoperational Test (S-03NF02) 14.2.12.1.74.1 Objectives
}
- a. To demonstrate that initiation of a safety injection signal (SIS) will shed the nonsafety-related loads, start the diesel generator, and sequence the associated equipment. The ability of each 4,160-V Class IE load group to supply the sequenced loads while maintaining voltage within design specifications is also verffied,
- b. To demonstrate that a loss of offsite power concurrent with SIS will shed the safety-related loads, start the diesel generator, close the diesel generator feeder breaker, and sequence the associated equipment. The ability of each diesel generator to supply the sequenced loads while maintaining voltage and frequency within design specifications is also verified.
- c. To demonstrate the ability of each diesel generator to carry the short-time rating load for 2 hours and the continuous rated load for 22 hours, without exceeding design limits.
- d. To demonstrate that each diesel generator, following operation for 2 hours at the short-time rated load and I~T U
14.2-97 Rev. 0
WOLF CREEX 22 hours at the continuous. rated load, will start auto-matically on a loss of ac voltage concurrent with an g SIS, attain voltage and frequency within design limits and time, and accept the LOCA sequenced loads, while maintaining voltage and frequency within design limits.
- c. To demonstrate the ability of the diesel cooling water system to maintain the diesel temperature within design specifications, while the diesel generators are opera-ting for 2 hours at the short-time rating load and 22 hours at the continuous rating load.
- f. To determine the fuel oil consumption of each diesel, while operating for 22 hours at the continuous rating load.
9 To demonstrate the' ability of the 125 V de system to perform its design functions while at minimum voltage,
- h. To demonstrate the independence between the redundant on ac and de power sources.
14.2.12.1.74.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Each diesel generator and its associated auxiliaries are llh available,
- c. All components actuated by the LOCA sequencer and safety-related and nonsafety-related load shed circuits are available.
14.2.12.1.74.3 Test Method
- a. A train A SIS is initiated, and the following are veri-fled:
- 1. Group 1 nonsafety-related loads are shed.
- 2. Group 1 diesel generator starts.
- 3. Group 1 LOCA sequencer is actuated, and associated components ara sequenced. The times for sequenced pumps to reach full flow are verified.
- 4. With bus NB01 supplying the sequenced loads from its normal source, bus voltage is recorded.
O 14.2-98 Rev. O
WOLP CREEK (~' b.- With' group 2 de load group isolated from its power (-)E source and group 1 de load group; voltage set to minimum, a loss of offsite power is initiated-concurrent with a train A SIS, and the following are verified:
- 1. Safety-related group 1 loads are shed.
- 2. Group 1 diesel generator starts, and its feeder breaker closes.
- 3. Group 1 LOCA sequencer is actuated, and associated components are sequenced. The times for sequenced pumps to reach full flow are-verified.
- 4. With the group 1 diesel. generator supplying the sequenced loads, bus voltage and' frequency are recorded.
- 5. The group 2 ac and de busses are monitored to verify the absence of voltage on these busses and loads, indienting no interconnection at load groups.
- c. The ability of the group 1 diesel generator to carry the short-time rating. load for 2 hours without exceeding design limits is verified.
- d. The ability of the group 1 diesel generator to carry the
*(]) - continuous rated load for 22 hours without exceeding design limits is verified. Group 1 diesel fuel oil consumption is also determined.
- e. Following group 1 diesel generator operation for 2 hours at the short-time rated load and 22 hours at the contin-uous rated load, the group 1 diesel generator is shut-down, a loss of group 1 ac voltage is-initiated concur-rent with a train A SIS, and the ability of the group 1 diesel generator to start, attain voltage and frequency within design limits and time, and accept the loads resulting from the design accident loading sequence while maintaining voltage and frequency within design limits is verified. If this test is not satisfacturily completed, it is not necesuary to repeat the tests of items e and d prior to rerunning this test. Instead, prior to rerunning this test, the diesel generator may be operated at the continuous rated load for 1 hour or until operating temperature has stabilized.
O 14.2-99 Rev. 0
WOLF CREEK
- f. A train B SIS is initiated, fied:
and the following are veri- g
- 1. Gresup 2 nonsafety-related loads are shed.
- 2. Group 2 diesel generator starts.
- 3. Group 2 LOCA sequencer is actuated, and associated ccmponents are sequenced. The times for sequenced pumps to reach full flow are verified.
- 4. dith Bus NB02 supplying the sequenced loads from its normal source, bus voltage is recorded.
- g. With @toup 1 de load group isolated from its power source and group 2 de load group voltage set to .ninimum, a loss of offsite power is initiated concurrent with a train B SIS, and the following are verified:
- 1. Safety-related group 2 loads are shed.
- 2. Group 2 diesel generator starts, and its feeder breaker closes.
- 3. Group 2 LOCA sequencer is actuated, and associated components are sequenced. The times for sequenced pumps to reach full flow are verified.
- 4. With the group 2 diesel generator supplying the sequenced loads, bus voltage and frequency are recorded.
- 5. The group 1 ac and de busses are monitored to verify the absence of voltage on these busses and loads, indicating no interconnection of load groups,
- h. The ability of the group 2 diesel generator to carry the short-time rating load for 2 hours without exceeding design limits is verified.
- i. The ability of the group 2 diesel generator to carry the continuous rated load for 22 hours without exceeding design limits is verified. Group 2 diesel fuel oil consumpt'.on is also determined.
O 14.2-100 Rev. 0
l WOLF CREEK f~) j. Following group 2 diesel generator operation for 2 hours at the short-time rated load and 22 hours at the contin-uous rated load, the group 2 diesel generator is shut-down, a loss of group 2 ac voltage is initiated concur-rent with a train B SIS, and the ability of the group 2 diesel generator to start, attain voltage and frequency within design limits and time, and accept the LOCA se-quenced loads, while maintaining voltage and frequency within design limits, is verifled. If this test is not satisfactorily completed, it is not necessary to repeat the tests of items h and i prior to rerunning this test. Instead, prior to rerunning this test, the diesel gener-ator may be operated at the continuous rated load for 1 hour or until operating temperature has stabilized.
- k. The ability of the diesel cooling water system to main-tain the diesel temperature within design specifications, while the diesel generators are operating for 2 hours at the short-time rating load and 22 hours at the continuous rating load, is verified.
14.2.12.1.74.4 Acceptance criteria
- a. A train A SIS initiates the following, in accordance with system design:
- 1. Group 1 nonsafety-related loads are shed.
'2. Group 1 diesel generator starte.
- 3. Group 1 LOCA sequencer actuates, and the associated components are sequenced. Sequenced pumps reach full flow within the required times,
- b. Bus N301, while powered from its normal source, supplies tne sequenced loads while maintaining voltage within design specifications.
- c. With the group 2 de load group isolated from its power source and the group 1 de load group voltage at minimum, a loss of offsite power concurrent with a train A SIS initiates the following, in accordance with system design:
- 1. Safety-related group 1 loads are shed.
- 2. Group 1 diesel generator starts, and its feeder breaker closes.
14.2-101 Rev. 0 l
WOLF CREEK
- 3. Group 1 LOCA sequencer actuates, and the associated components are sequenced. Sequenced pumps reach lll full flow within design times,
- d. Group 1 diesel generator supplies the sequenced loads, while maintaining voltage and frequency within design specifications.
- e. With load group 1 supplying loads following a loss of offsite power concurrent with a train A SIS, the group 2 ac and de busses are verified de-energized, indicating no interconnection of load groups.
- f. Following group 1 diesel ger.erator operation for 2 hours at the short-time rated load and 22 hours at the contin-uous rated load, the group 1 diesel generator starts, attains voltage and frequency within design limits and time, and accepts the LOCA sequenced loads while main-taining voltage and frequency within design limits, on loss of group 1 ac voltage concurrent with a train A SIS.
- g. A train B SIS initiates the following, in accordance with the system design:
- 1. Group 2 nonsafety-related loads are shed.
- 2. Group 2 diesel generator starts.
- 3. Group 2 LOCA cequencer actuates, and the associated components are sequenced. Sequenced pumps reach full flow within design times.
- h. Bus NB02, while powered from its normal source, supplies the required loads while maintaining the voltage within design specifications.
- i. With the group 1 de load group isolated from its power source and the group 2 de load group voltage at minimum, a loss of offsite power concurrent with a train a SIS initiates the following, in accordance with system design:
- 1. Safety-related group 2 loads are shed.
- 2. Group 2 diesel generator starts, and its feeder breaker closes.
O 14.2-102 Rev. O
WOLF CRECK 1 l
- 3. Group 2 LOCA sequencer actuates, and the associated
-(]) components are sequenced. Sequenced pumps reach full flow within design times.
- j. Group 2 diesel generator supplies the required loads, while maintaining voltage and frequency within design specifications.
- k. With load group 2 supplying loads following a loss of offsite power concurrent with a train B SIS, the group 1 ac and de busses are verified de-energized, indicating no interconnection of load groups.
- 1. Following group 2 diesel generator operation for 2 hours at the short-time rated load and 22 hours at continuous rated load, group 2 diesel generator starts, attains voltage and frequency within design limits and time, and accepts the LOCA sequenced loads while maintaining voltage and frequency within design limits, on loss of group 2 ac voltage concurrent with a train B SIS.
- m. Each diesel generator is capable of carrying the short-time rating load for 2 hours and the continuous rated load for 22 hours, without exceeding design limits,
- n. Fuel oil consumption of each diesel, while operating at the continuous rated load, is within design specifica-O tions.
- o. Each diesel generator cooling water system, with the diesel generators operating for 2 hours at the short-time rating load and 22 hours at the continuous rating load, maintains the diesel temperatures within design specifications.
- p. The controls required for the loss of offsite power concurrent with a SIS (shedding, aequencing, etc.)
function with minimum de voltage available. 14.2.12.1.75 Shutdown Sequencer Preoperational Test (S-0 3NFF 3) 14.2.12.1.75.1 Objectives
- a. To demonstrate that de-energi:ation of either 4,160-V Class IE load group will start the associated diesel generator, close the diesel generator feeder breaker, actuate the associated group load shed, and actuate the shutdown sequencer. All sequenced components are veri-fied to start within required design times.
O 14.2-103 Rev. 0
i WOLF CREEK
- b. To demonstrate that each diesel generator will maintain voltage and frequency within design specifications while lll supplying the design shutdown loads,
- c. To demonstrate the ability of the emergency 4.16-kV loads to start at maximum and minimum design voltages.
14.2.12.1.75.2 Prerequisites
- a. Required component testing and instrument calibratf 1 are complete.
- b. Required electrical power supplies and control circuits are operational,
- c. Each diesel generator and its associated auxiliaries are available.
- d. All components actuated by the shutdown sequencer are available.
14.2.12.1.75.3 Test Method
- a. Class IE 4,160-V load group 1 is de-energized and the following are verified
- 1. Group 1 load shedder actuates.
ggg
- 2. Group 1 diesel generator starts, and its feeder breaker closes.
- 3. Group 1 shutdown sequencer is actuated, and asso-ciated components are sequenced. Components are verified to actuate within the required design times.
- b. Class IE 4,160-V load group 2 is de-energized and the following are verified:
- 1. Group 2 load shedder actuates.
- 2. Group 2 diesel generator starts, and its feeder breaker closes.
- 3. Group 2 shutdown sequencer is actuated, and asso-ciated components are sequenced, components are verified to actuate within the required design times.
O 14.2-104 Rev. 0
WOLE CREEK n I) c. Emergency 4.16-kV loads are started while their respec-tive diesel generators are supplying:
- 1. Minimum rated voltage
- 2. Maximum rnted voltage
- d. The ability of each diesel generator to maintain voltage and frequency within the design specifications while supplying the design shutdown loads is verified.
14.2.12.1.75.4 Acceptance Criteria
- a. De-energi:ation of Class IE 4,160-V load group 1 initin-tes the following, in accordance with system design:
- 1. Group 1 diesel generator starts, and its feeder breaker closes.
- 2. Group 1 shutdown pequencer actuaten, and associated components are sequenced. Components actuate within required design times.
- 3. Group 1 load shedder actuatas,
- b. De-energi:ation of Class IE 4,160-V load group 2 initia-tes the following, in accordance with system design:
U<~s
- l. Group 2 diesel generator starts, and its feeder breaker closes.
- 2. Group 2 shutdown sequencer actuates, and associated components are sequenced. Components actuate within required design times.
- 3. Group 2 load shedder actuates.
- c. The emergency 4.16-kV loads start and reach rated-speed within design times, with minimum and maximum design voltage.
- d. Each diesel generator maintains voltage and frequency within design specifications, while supplying the design shutdown loads.
14.2.12.1.76 480-V (Class IE) System Preoperational Test (S-03NG01) 14.2.12.1.76.1 Objectives To demonstrate that the 480-V Class IE load centers can be ener-cized from their normal and alternate sources and verify the () 14.2-105 Rev. O
WOLF CREEK operability of system breaker pr( tective interlocks. operation of system instrumentation and controls is also verified. Proper g 14.2.12.1.76.2 Prerequisites
- a. Required component testiny and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
14.2.12.1.76.3 Test Method
- a. The 480-V Class IE load centera are energized from their normal source, and voltages are recorded.
- b. The 480-V Class IE load centers are energized from their alternate source, and voltages are recorded.
- c. System breakers are operated, and breaker interlocks verified.
14.2.12.1.76.4 Acceptance Criteria
- a. The voltage for each 480-V Class Il load center, when supplied from its normal source, is within design spaci-fications,
- b. The voltage for each 480-V Class IE ioad center, when nupplied f rora its alternate source, is within design specifications,
- c. Syscem breaker interlocks operate in accordance with the system design.
14.2.12.le77 480-V Class IE System (ESW) Preoperational Test (SU3-!;G0 2) . 14.2.12.1.77.1 OL -j ecti ve s To demonstrate that the nonpower block 480-V Class IE MCC can be energized from their normal source and to verify theit bus voltage phase saquence. Proper operation of system instrumentation and controls is also verified. 14.2.12.1.77.2 Prerequisites
- a. Required component testing and instrument calibration are completed.
O 14.2-106 Rev. 0 l
WOLF CREEK
.I'E \I b.- Required electrical power supplies'and control circuits are operational.
14.2.12.1.77.3 Test Method The nonpower block 480-V Class'IE MCC are energired, voltages are recorded, and phase sequence is-verified. 14.2.12.1.77.4 Acceptance Criteria
- e. The voltage for each nonpower block 480-V Class IE MCC is within design specification.
- b. The bus voltage phase sequence of the nonpower block 480-V Class IE MCC is-in accordance with design.
14.2.12.1.78- 125-V (Class IE) DC System Preoperational Test (S-03NK01) 14.2.12.1.78.1 Objectives To demonstrate the ability of the batteries and chargers to pro-vide power during normal operations and the battery to provide power during abnormal conditions. The battery chargers' ability to recharge their respective battery is also demonstrated. Proper operation of- the system instrumentation und controls is also verified. 14.2.12.1.78.2 Prer'equisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. Ventilation for the battery rooms is available.
14.2.12.1.78.3 Test Method
- a. Each' battery is discharged, using a test load at the design duty cycle discharge rate.
- b. Each battery is fully discharged to determine its capac-ity factor,
- c. Each battery charger will charge its respective battery to normal conditions, after the battery has undergone a design duty cycle, while simultaneously supplying power-at a rate equivalent to the design emergency loading.
O 14.2-107 Rev. 0
WOLP CREEK 14.2.12.1.78.4 Acceptance Criteria ll)
- a. Each battery is capable of maintaining output voltage above the design minimum, during a design duty cycle.
- b. Each battery has a capacity factor greater than or equal to design,
- c. The battery chargers are able to recharge the batteries to normal conditions, after the battery has undergone a design duty cycle, while simultaneously supplying power at a rate equivalent to the design emergency loading.
14.2.12.1.79 Instrument AC System (Class IE) Preoperational Test (S-03NN01) 14.2.12.1.79.1 Objectives To demonstrate that the 120-V Class IE ac distribution panel-boards can be fed from their normal source inverters and from their backup source transformers by manual transfer. The opera-bility of system instrumentation and controls, including breaker protective interlocks, is also verified. 14.2.12.1.79.2 Prerequisites
- a. Required component testing and instrument calibration 3 are complete. W
- b. Required electrical power supplies and control circuits are operational.
14.2.12.1.79.3 Test Method
- a. The 120-V Class IE ac distribution panelboards are energized from their normal source inverters, and panel-board voltages are recorded.
- b. The 120-V Class IE ac distribution panelboards are energized from their backup source transformers by manual transfer, and panolboard voltages are recorded,
- c. The system breakers are operated, and breaker interlocks are verified.
14.2.12.1.79.4 Acceptance Criteria
- a. Each 120-V Class IE ac distribution panelboard voltage, when supplied from the normal source inverters of the panelboards, is within design specifications.
O 14.2-108 Rev. O
. . . . --.- - .- - -. - . - . .. - ~ - .. -. -
k # b EWOLF CREEK i
'(][, b. Each-120-VfClass IE.ac distribution panelboard voltage, >
when . supplied- from the backup source transformers,; is-within design-specifications.- c.4 System breaker interlocks 'operateL'in accordance with system design. 14.2.12.1.40 Engineered Safeguards (NSSS) Preoperational Test (SU3-SA01) 14.2.12.1.80.1 Objectives
- a. To demonstrate the. ability of the NSSS to Jinitiate-safety injection, containment isolation, containment a spray actuation, Emain-feedwater isolation, and steam line isolation signals on receipt of the associated input signals.
- b. To verify NSSS ESFAS loop response times.
- c. To demonstrate the ability of each solid-state .protec-tion system test panel to adequately test the associated NSSS;ESFAS and-reactor protection logic trains,
- d. To demonstrate the coincidence and redundancy of the.
NSSS ESTAS. l 4( ,/ e. To verify the operability of ESFAS block and permissive-interlocks. 14.2.12.1.80.2 Prerequisites
- a. Requiradicomponent testing and instrument calibration are complete.
- b. Required' electrical power' supplies are operational, 14.2.12.1.80.3 . Test Method
- a. The. ability of the'NSSS ESFAS to actuate safety injec-F tion, containment isolation, containment spray actua-tion,' main feedwater isolation, and steam line isolation signals on receipt of the. required coincidence of the following input signals for each redundant channel is-
- verified:
'High steam line pressure rate Low steam line pressure Low pressurizar, pressure O
14.2-109 Rev. 0
l WOLF CREEK lll High containment pressure (Hi-1, Hi-2, and Hi-3) High-high steam generator level Low Tavg Low-low steam generator water level
- b. Input signals are initiated, and loop response times are verified,
- c. The ability of each solid-state protection system test panel to test the NSSS ESFAS logic trains is verified.
- d. ESFAS block and permissive interlocks are verified.
14.2.12.1.80.4 Acceptance Criteria
- a. The NSSS ESFAS actuates safety injection, containment isolation, containment spray actuation, main feedwater isolation, and steam line isolation signals when their associated input signals are received from the following signals for each applicable channel:
High steam line pressure rate Low steam line pressure Low pressurizer pressure High containment pressure (Hi-1, Hi-2, and Hi-3) High-high steam generator level Low Tavg ' Low-low steam generator water level
- b. NSSS ESFAS loop response times are within design speci-fications.
- c. ESFAS block and permissive interlocks operate in accor-dance with system design.
14.2.12.1.81 Engineered Safeguards (BOP) Preoperational Test ( SU 3-S A0 2) 14.2.12.1.81.1 Objectives a.- To demonstrate the operability of the BOP ESFAS to initiate containment purge isolation, control room ventilation isolation, fuel building ventilation isola-ion, auxiliary feedwater pump actuation, auxiliary feedwater suction valve switchover to essential service water (ESW), and steam generator blowdown and sample isolation signals on receipt of the associated input signals. 14.2-110 Rev. O h 1
f'N WOLF CREEK b
- b. To verify BOP ESFAS loop response times.
- c. To demonstrate the ability of the BOP ESTAS test panel to adequately test the associated BOP ESFAS logic trains.
- d. To demonstrate the coincidence and redundancy of the BOP ESFAS.
14.2.12.1.81.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical p ver supplies are operational.
14.2.12.1.81.3 Test Hethod
- a. The ability of the BOP ESFAS to actuate containment purge isolation, control room ventilation isolation, fuel building ventilation isolation, auxiliary feedwacer pump actuation, auxiliary feedwater suction valve switchover to ESW, and steam generator blowdown and sample isolation signals on receipt of the required 7s coincidence of the following input signals for each
(_) redundant channel is verified, o Containment isolation (phase A) o High atmospheric radiation o High chlorine concentration o Loss of main feedwater flow o Low-low steam generator level o Loss of offsite power o Low feedwater pump suction pressure o Safety injection
- b. Input signals are initiated, and loop response times are verified,
- c. The ability of the BOP ESFAS test panel to test the BOP ESFAS logic trains is verified.
i 14.2.12.1.81.4 Acceptance Criteria
- a. The BOP ESPAS actuates containment purge isolation, control room ventilation isolation, fuel building venti-lation isolation, auxiliary feedwater pump actuation, auxiliary feedwater suction valve switchover to ESW, and 14.2-111 Rev. O l
WOLF CREEK I steam generator blowdown and. sample isolation signals when their' associated input signals are received from the following signals for each applicable channels o' containment isolation-(phase A) o High atmospheric-radiation o High chlorine concentration o Loss-of main feedwater flow o Low-low steam generator-level o Loss of offsite power o Low feedwater pump suction pressure o Safety injection
- b. BOP ESTAS loop response times are within design specifi-cations..
'14.2.12.1.82- Engineered Safeguards Verification Test (SU3-SA03) 14.2.12.1.82.1 Objectives To demonstrate the proper response of actuated components resul-ting from the following safety signals Safety-injection, con-tainment spray actuation,- main feedwater isolation,- steam line isolation, containment isolation, contro11 room ventilation isolation, _ containment purge fuel building isolation, ventilation isolation,. auxiliary feedwater pump actuation, auxiliary feedwater suction. valve switch over to ESW, and steam generator blowdown and sample isolation.
14.2.12.1.82.'2 Prerequisites
- a. Required component testing and instrument calibration are complete. '
b.- Required electrical power sources.and control circuits are operational.
- c. Components actuated by_the NSSS and BOP ESTAS are avail-
-able.
14.2.12.1.82.3 -Test Method NSSS and BOP ESFAS signals are initiated manually and the proper-response. and response times of the actuated components are veri-fled. O 14.2-112 Rev. O
(~ WOLF CREEK 14.2.12.1.82.4 Acceptance Criteria components required to actuate on receipt of safety signals re-spond properly in accordance with design specifications and within the times specified by design requirements. 14.2.12.1.83 Reactor Protection System Logic Test (S-03SB01) 14.2.12.1.83.1 Objectives
- a. To demonstrate the ability of the reactor protection nystem to initiate a reactor trip on input of the asso-ciated input signals.
- b. To verify reactor protection loop response times.
- c. To verify the operability of the reactor protection system block and permissive interlocks.
- d. To demonstrate the coincidence, redundancy, and fail safe (power leas) design of the reactor protection system.
14.2.12.1.83.2 Prerequisites () a. Required component are complete, testing and instrument calibration
- b. Required electrical power supplies and control circuits are operational.
14.2.12.1.83.3 Test Method
- a. The ability of the reactor protection system to initiate a reactor trip on receipt of the proper coincidence of the following trip signals for each redundant channel is verified: -
o Source range high neutron flux c, Intermediate range high neutron flux o Power range high neutron flux (low setpoint and high setpoint) o Power range high positive neutron flux rate o Power range high negative neutron flux rate o Overtemperature AT o Overpower AT o Low primary coolant flow o Reactor coolant pump bus undervoltage o Reactor coolant pump bus underfrequency O - 14.2-113 Rev. 0
WOLF CREEK o o High pressurizer pressure llh Low pressurizer pressure o High pressurizer level o Safety injection signal o Turbine trip signal
- b. Loop response times are measured for the above listed trip signals,
- c. Reactor protection system block and permissive inter-locks are verified,
- d. Power is isolated from the system, and the safo failure of the system is verified.
14.2.12.1.83.4 Acceptance Criteria
- a. The reactor protection systen initiates a reactor trip on receipt of the proper coincidence of the following signals for each epplicabla channels o Source range high neutron flux o Intermediate range high neutron flux o Power range high neutron flux (low setpoint and high setpoint) o Power range high positive neutron flux rate o Power range high negative neutron flux rate o overtemperature 4T o overpower AT o Low primary coolant flow o Reactor coolant pump bus undervoltage o Reactor coolant pump bus underfrequency o High pressurizer pressure o Low pressurizer pressure o High pressurizer level o Safety injection signal o Turbine trip signal
- b. Loop response times for the following trip signals are within design limits, o Power range high neutron flux (low setpoint and high setpoint) o Power range high negative neutron flux rate o Overtemperature aT o Overpower AT o Low primary coolant flow o Reactor coolant pur.p bus undervoltage o Reactor coolant pump bus underfrequency 14.2-114 Rev. O 9
l-
I I f- WOLF CREEK (>] o High pressurizer pressure o Low pressurizer pressure
- c. Reactor protection system block and permissive inter-locks operate in accordance with system design.
- d. The reactor protection system functions in accordance with system design on a loss of power.
14.2.12.1.84 Primary Sampling System Preoperational Test (C-03SJ01) 14.2.12.1.84.1 Objectives
- a. To set sample panels' flow rates and to verity the operability of the sample system containment isolation valves. Proper operation of system instrumentation and controls is also verified.
- b. To verify that the post-accident sampling system (PASS) containment isolation valves operate properly.
14.2.12.1.84.2 Prerequisites
- a. Required component testing instrument calibration, and
() system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operable.
- c. Plant conditions are established, and systems are avail-able, as necessary, to facilitate drawing samples frcm the sample points.
- d. The component cooling water system is available to provide cooling water to the auxiliary building sample station.
- e. The chemical and volume control system is available to receive discharge from the nuclear sampling station.
- f. The chemical and detergent waste system is available te receive discharge from the nuclear sampling station.
14.2.12.1.84.3 Test Method
- a. Sample panel flows are adjusted, and flow data are recorded.
~
14.2-115 Rev. O
l l WOLF CREEK
- b. Operability of the sample containment isolation valves is verified, including their response to an isolation signal. Valve operating times are recorded.
14.2.12.1.84.4 Acceptance Criteria
- a. The sample containment isolation valves close on receipt of an isolation signal.
- b. The sample containment isolation valves' closure times are within design specifications.
14.2.12.1.85 Process Radiation Monitoring System Preoperational Test (S-03SP01) 14.2.12.1.85.1 objectives To demonstrate the operation of the process radiation menitors and to verify the ability of the process radiation monitoring system to provide alarm and isolation signals, as applicable, apon re-caipt of high radiation signals, operability of the radicactivity monitoring control room microprocessor is also verified. 14.2.12.1.85.2 Prerequisites
- a. Required component testing and inhtrument calibration are complete.
ggg
- b. Required electrical power supplies and control circuits are operable.
14.2.12.1.85.3 Test Method
- a. The check source for each monitor is remotely positioned, and the actuation of each monitor and the operability of its associated alarms and isolation signals are verified.
- b. Operability of the radioactivity monitoring control room microprocessor is verified.
14.2.12.1.85.4 Acceptance Criteria The process radiation monitoring system provides alarm and isola-tion signals, in accordance with system design specifications. 14.2-116 Rev. O O
{} WOLF CREEK 14.2.12.1.86 Power Conversion and ECCS Thermal Expansion Test (SU3 a?04) 14.2.12.1.86.1 Objective To damonstrate snubber operability on all safety-related systems whose operating temperature exceeds 250'F. 14.2.12,1.86.2 Prerequisites
- a. Preservice examinations as specified in the Tedesco letter to KG&E dated 2/10/81 have been completed on the systems being checked within the last 6 months.
- b. Other required component testing and instrument caldbra-tion are completed,
- c. Required electrical power supplies and control circuits are operstional,
- d. Preoperational testing is in progress.
14.2.12.1.86.3 Test Method j-) a. During initial system heatup and cooldown, at specified (_/ temperature intervals, verify the expected snubber movement for any systam which attains operating tempera-ture,
- b. For those systems which do not attain operating tempera-ture, verify by observation and/or calculation that the snubber will accommodate the projected thermal movement.
- c. Observe snubber swing clearances at specified heat-up and cooldown intervals.
14.2.12.1.86.4 Acceptance Critoria
- a. The expected snubber movement for any system that at-tains operating temperature is within design specifica-tions.
- b. The expected snubber movement determined by observation and/or calculation for any system that does not attain operating temperature is within design specifications,
- c. Snubber swing clearance observed at specified heatup and cooldown intervals is within design specifications.
] -
14.2-117 Rev. 0 l
WOLF CREEK 14.2.12.1.87 Power Conversion and ECCS Systems Dynamic Test (S-03 0L '5) 14.2.12.1.07.1 Object 1/es To demonstrate during specified transients that the systems' moni-tored points respond in acc3rdance with design. 14.2.12.1.87.2 Prerequisites
- a. Reference points for measurement of the systems are established,
- b. Hot functional testing is in progress.
- c. All subject systems are available for the specified dynamic operations.
- d. Required instrument calibration is comp.iete.
14.2.12.1.87.3 Test Method
- a. The systems are aligned for the specified dynamic opera-tion.
- b. The specified dynamic event of pump operation, valve c;2 ration, etc., is initiated, and the systsm is meni- ggg tored for response.
14.2.12.1.87.4 Acceptance Criteria
- a. The total stress shall not exceed applicable code limits.
14.2.12.1.88 HEPA Filter Test (SU3-0006). 14.2.12.1.88.1 Objectives To demonstrate the leaktightness and particulate removal effi-ciency of all HEPA filters and to verify the leaktightness of their associated charcoal adsorbsrs. 14.2.12.1.88.2 Prerequisites
- a. The ventilation systems containing HEPA filters and charcoal adsorbers have been air balanced and are opera-tional and available to support this test.
O 14.2-118 Rev. 0 l
l
,~y WOLF CREEK v
- b. Required electrical power supplies and control circuits I are operational. -
- c. Required instrument calibration is complete.
14.2.12.1.88.3 Test Method
- a. HEPA filters are inplace tested with cold poly-dispersed DOP, in accordance with the procedures set forth in ANSI N510.
- b. Charcoal adsorbers are inplace tested with a suitable refrigerant, in accordance with the procedures set forth in ANSI N510.
14.2.12.1.88.4 Acceptance Criteria
- a. The airflow of each filter adsorber unit is equal to the design flow,
- b. Air flow distribution downstream of each HEPA filter is within 20 percent of the average velocity through the unit.
- c. HEPA DOP penetration is less than one percent at the
{j design air flow.
- d. Charcoal adsorber bypass leakage is less than .05 per-cent at the design air flow.
14.2.12.1.89 Cooldown from Hot Standby External to the Control Room (S-030008) 14.2.12.1.89.1- Objectives To demonstrate, using a plant procedure, the potential capability to cooldown the plant from the hot standby to the cold shutdown condition, using instrumentation and controls external to the control room verifying that:
- a. The reactor coolant temperature and pressure can be lowered to permit the operation of the residual heat removal (RER) system.
- b. The RER system can be operated and controlled.
- c. The reactor coolant temperature can be reduced 350' F, using the RER system, without exceeding technical speci-fication limits, f)3 14.2-119 Rev. O i
WOLF CREEK 14.2.12.1.89.3 Prerequisites lll
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The plant is in a hot standby condition.
- d. The authority and responsibility of the control room observers has been established and is specified in this procedure.
14.2.12.1.89.3 Test Method
- a. The plant is cooled from hot standby, RER is initiated, and a h50' F cooldown is performed with the RHR system transferring heat to the ultimate heat sink, using in-strumentation and controls external to the control room,
- b. All actions performed by the control room observers are documented within this procedure for use in evaluating their impact on the test results.
14.2.12.1.89.4 Acceptance Criteria The following actionu'are capable of being performed, the control room: external to g
- a. The reactor coolant temperature and pressure can be lowered to permit the operation of the RHR system.
- b. The reactor coolant temperature can be reduced g50' F, using the RHR system, without exceeding technical speci-fication limits.
14.2.12.1.90 Compressed Gas Accumulator Testing (S-030009) 14.2.12.1.90.1 Objeclives To demonstrate the ability of the auxiliary feedwater control valve /mainsteam atmospheric relief valve and main feedwater con-trol valve accumulators to provide the design backup supply of compressed gas for continued design valve operation following a loss of the normal motive source. 14.2-120 ' lll
I g_ WOIS CREEK \-) 14.2.12.1.90.2 Prerequisites Required component testing, instrument calibration and system flushing / cleaning are complete. 14.2.12.1.90.3. Test Method The accumulators are isolated from the compressed gas supply header and the associated valves are operated to demonstrate the ability of the accumulators to provida design motive force for the required valve cycles. 14.2.12.1.90.4 Acceptance Criteria The auxiliary feedwater control valve /mainsteam atmospheric relief valve, and main feedwater control valve accumulators provide the design backup supply of compressed gas to their associated valves. 14.2.12.2 Nonsafety-Related Preoperational Test Procedures The following sections are the test abstract for each nonsafety-related preoperational test. Table 14.2-2 provides an index of these tests. 7- 14.2.12.2.1 Turbine Trip Test (S-04ACO2) t
'~'
14.2.12.2.1.1 Objectives
- a. To demonstrate the ability of the turbine trip and monitoring system to initiate a turbine trip on input of the associated input signals.
- b. To demonstrate the response of the moisture separator reheater drain valves, feedwater heater extraction check valves, turbina main stop valves, turbine main stop
, valve above seat drain valves, turbine contro' valves, turbine control valve above seat drain valves, intermed-late stop valves, main steamline drain valvo0. startup drain valves, and intercept valves to a turnine trip signal,
- c. To demonstrate that a turbine trip signal initiates a reactor trip signal.
- d. To demonstrate that the turbine main stop valves opera-ting times are within design specifications.
(O~) 14.2-121 Rev. O
WOLF CREEK 14.2.12.2.1.2 Preroquisites
- a. Required component testing and instrument calibration is complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The main turbino control oil and lube oil syscems are available to provide oil to the turbine auxiliaries,
- d. The compressed air system is available to provide air to system air-operated valves.
14.2.12.2.1.3 Test Method
- a. The ability of the turbine trip and monitoring system to initiate a turbine trip signal on receipt of each of the following input signals is verified:
o Manual trip pushbutton depressed o Manual trip handle pulled o Generator trip (EHC vital trip) o Generator trip (unit trip) o Reactor trip o Loss of stator coolant o Low lube oil pressure & o Loss of EHC 125 V de power with turbine speed W below 75 percent o High turbine vibration o High exhaust hood temperature o Low hydraulic fluid pressure o Moisture separator high level o Low bearing oil pressure o Low condenser vacuum o Excessive thrust bearing wear o Backup overspeed (Electrical) o Loss of EHC 24-volt de power
- b. A turbins trip signal is initiated, and the response of the following valves is verified:
o Moisture separator reheater drain valves o Feedwater heater extraction check valves o Turbine main stop valves-o Turbine control valves o Intermediate stop valves o Turbine intercept valves o Startup drain valves o Main steam line drain valves 14.2-122 Rev. O 9
(~N WOLF CREEK x,_) . o Turbine main stop valve above sent drain valvos o Turbina control valve above seat drkin valves
- c. A turbine trip signal is initiated, and a reactor trip input signal is verified.
14.2.12.2.1.4 Acceptance Criteria
- a. The turbine trip and- monitoring system initiates a turbine trip on receipt of each of the following sig-nalst o Manual ~crip pushbutton depressed o Manual trip handle pulled o Generator trip (EHC vital trip) o Generator trip (unit trip) o Reactor trip-o Loss of stator coolant o Low lube oil pressure o Loss of EHC 125 V de power with turbine speed below 75 percent o High turbine vibration o High exhaust hood temperature o Low hydraulic fluid pressure
,o Moisture separator high level
<~) (,/ o Low bearing oil pressure o Low condenser vacuum o Excessive thrust bearing wear o Backup overspeed (electrical) o Loss of EHC 24-volt de power
- b. The following valves open on receipt of a turbine trip signal:
o Turbine main stop valve above seat drain valves o Turbine control valve above seat drain valves o Main steam line drain valves o Moisture separator reheater drain valves o startup drain valvon
- c. The following valves close on receipt of a turbine trip signal:
o Low pressure heater extraction check valves o Main stop valves o Turbine control valves o Intercept valves o Intermediate stop valves (O V 14.2-123 Rev. O
~ ___
WOLF CRIIx
- d. A turbine trip signal initiates a reactor trip signal. llh
- e. The turbine main stop valves operating times are within design sp6cifications.
14.2.12.7.2 Turbine System Cold Test (S-04AC03) 14.2.12.2.2.1 Objectives
- a. To demonstrate the operability of the turning gear and associated control circuits.
- b. To demonstrate the operability of the electro-hydraulic
(- control system. 14.2.12.2.2.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required e3ectrical power supplies and centrol circuits are operational.
- c. The main turbine control oil and lube oil systems are available to supply the turbine auxiliaries.
14.2.12.2.2.3 Test Method
- a. The operability of the turning gear and associated 9
control circuits is verified.
- b. A turbine simulator is utilized to verify the ability of the electro-hydraulic control system to perform its control functions.
14.2.12.2.2.4 Acceptance Criteria
- a. The turning gear motor trips on loss of bearing oil pressure, loss of all bearing lift pumps, or closure of the main transformer switchyard breaker.
- b. The turbine control and intercept valves close on a power load unbalance signal,
- c. The turbine load set is run back on a reactor overtemp-erature AT signal when in the manual mode.
- d. The turbine load set is run back on a reactor over-power AT signal when in the manual mode.
~
14.2-124 Rev. O
- ~ - - - - - - - - _ _ _ _ _
1 WOLF CRrrK
)
- e. The turbine load is not back on a loss of circulating water pump signal.
- f. Turbine loading is inhibited on a C-16 control interlock signal.
14.2.12.2.3 Condensate System preoperational Test (S-04AD01) , 14.2.12.2.3.1 Objectives To demonstrate the condann.te cumps' operating characteristics and verify the operation of syst6m valves and associated control circuits. The operability of the condensate storage and transfer system and associated components is also verified. 14.2.12.2.3.2 Prerequisites
- a. Required ccuponent testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplien and control circuits are operational.
- c. The feedvater system is available to receive flow frem the condensate pump discharge header.
O' d. The dominera11ted water system is availabic to provide water to the condensate pump seals and a source of makeup to the condensate stcrage tank.
- a. The condensate otorage tank is available to provide makeup to the conde.iser hotwell.
- f. The closed cooling water system is available to provide cooling water to the condensate pump motor bearing oil coolers.
14.2.12.2.3.3 Test Method
- a. Condensate pumps are operated, and performance charac-teristics are verified.
- b. The response of each condensate pump to a condenser low-low level trip signal in verified,
- c. The operability of the condensate pump recirculation valves is verified.
O 14.2-125 Rev. O
WOLF CRIEK 14.2.12.2.3.4 Acceptance Criteria h
- a. The operating characteristics of the condensate pumps are within design specifications.
- b. Each condensate pump will receive a trip signal on a 2/3 condanser low-low level signal,
- c. Each condensate pump recirculation valve operatos in accordance with design specifications.
14.2.12.2.4 Secondary Vent and Drain System Preoperational Test (S 04AF01) 14.2.12.2.4.1 Objectives
- a. To demonstrato the operating characteristics of the heater drain pumps.
- b. To demonstrate the operability of system valve and pump control circuits.
14.2.12.2.4.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits llk are operational.
- c. The compressed air system is available to the system air-operated valves.
- d. The closed cooling water system is available to supply cooling vatar to the heater drain pumps.
14.2.12.2.4.3 Test Method
- a. The heuter drain pumps are operated, and perforhance characteristics are verified,
- b. The operability of system valve and pump control cir-cuits is verified.
14.2.12.2.4.4 Acceptance Criteria The operating characteristics of the heater drain pu=ps are within design specificatiens. 14.2-126 Rev. O 0 I
WOLF CREEK 14.2.12.2.5 Condensate and Feedwater Chemical Feed System Pre-operational Test (S-04AQOl} 14.2.12.2.5.1 Objectives
- a. To demonstrate the operating characteristics of the condensate hydrazine addition pymps, condensate ammonia addition pumps, condensate hydrazine circulating pumps, condensate ammonia circulating pumps, feedwater hydra-zine ammonia addition pumps, and feedvater hydrazine ammonia circulating pump and verify the operation of the associated control circuits.
- b. To. demonstrate the operability of the drum dispensing pumps.
14.2.12.2.5.2 prerequisites
- a. Required component testing, instrument calibration, and systen flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
r~S c. The demineralized water storage and transfer system is (_) available to provide a source of demineralized water to the hydrazine and ammonia supply and mixing tanks.
- d. The compressed air system is available to provide air to the drum dispensing pumps.
- e. The service gas system is available to provide a source of nitrogon to the hydrazine and ammonia supply, measur-ing, and mixing tanks.
14.2.12.2.5.3 Test Method
- a. System pumps are operated, and performance characteris-tien are verified,
- b. The response of the condensate hydrazine circulating pumps, condensate ammonia circulating pumps, and the feedwater hydrazine ammonia feed pumps to a low level in their associated tank is verified.
O 14.2-127 Rev. O
WOLT CREEK 14.2.12.2.5.4 Acceptance Criteria
- a. The operating characteristics of the condensate hydra-zine addition pumps, condensate ammonia addition purps, condensate hydrazine circulating pumps, condensate ammonia circulating pumps, feedwater hydrazine ammonia addition pumps, feedwater hydrazine ammonia circulating pump, and the drum dispensing pumps are within design specifications.
- b. The condensate hydrazine circulating pumps, condensate ammonia circulating pumps, feedwater hydrazine ammonia feed pumps, and the feedwater hydrazine ammonia circula-ting pump trip on a low level signal frem their associa-tad tanks.
14.2.12.2.6 Reactor Makeup Water System Preoperational Test (S-04BLO1) 14.2.12.2.6.1 Objectives
- a. To demonstrate the operating characteristics of the reactor makeup water transfer pumps and verify that the associated control circuits are functioning properly.
- b. To demonstrate the operation of the system automatic valves, includin the response of the reactor makeup wat6r system conta nment supply valve to a CIS.
{lg 14.2.12.2.6.2 Prerequisites
- a. Required component testing, instrumant calibration, and system flushing / cleaning are complete.
- b. Required electrical power supp.*.ies and control ci:cuits are operational.
- c. The domineralized water storage and transfer system is available to provide a sobree of water to the reactor makeup water storage tank.
14.2.12.2.6.3 Test Method
- a. The reactor makeup water transfer pumps are operated, and pump operating data are recorded,
- b. Reactor makeup water transfer pumps and system automatic valves control logics are verified, including their response to safoty signals.
! 14.2-128 Rev. O i
l WoLP CREEK (
- c. The reactor makeup water containment supply valve is operated under flow conditions and operating timos recorded. l l
14.2.12.7.6.4 Acceptance criteria l
- a. The opeyating characteristics of the reactor makeup i vater transfer pumps are within design specifications. t
- b. Each reactor makeup water transfer pump trips on receipt of a reactor makeup water storage tank low level signal,
- c. Each ranctor makeup water transfer pump starts, after a time delay, with the other pump running and the receipt of a low header pressure signal.
- d. The reactor makeup water containment supply valve clos-ure time is within design specifications.
- o. The reactor makeup containment supply valve closes en receipt of a CIS.
14.2.12.2.7 Condenser Air Removal System preoperational Test (S-04CG01)
,-).
( 14.2.12.2.7.1 Objectives
- a. To demonstrate the operation of_the condenser air re-moval portion of the turbine building HVAC system motor-operated dampers, including automatic operation on a safety injection signal.
- b. To demonstrate the capacities of the condenser air removal filtration fans and verify the operhtion of their associated control circuits.
- c. To demonstrate the operability of the condenser air removal system vacuum pumps, control valves, and their associated control circuits.
14.2.12.2.7.2 Prerequisitos
- a. Required component testing, instrument calibration, and system flushing / cleaning'are complete.
- b. Required electrical power supplies and control circuits arra operational.
O 14.2-129 Rev. 0
WOLF CREEK
- c. The condenser air removal filtration system portion of the turbine building HVAC system is available to support this test.
- d. The condensate storage tank is available to provide a source of water to the vacuum pump seal water reser-voirs.
- a. The service water system is available to provide cooling water to the mechanical vacuum pump seal water coolers.
14.2.12.2.7.3 Test Method
- a. The condenser air removal filtration fans are operated, and fan capacities are verified.
- b. Operation of the condenser air removal filtration dam-pers is verified, including their response to a safety injection signal.
- c. The ability of the mechanical vacuum pumps to reduce condenser pressure during startup operation is verified.
- d. Operability of the mechanical vacuum pumps and their associated control valves' control circuits is verified, including their response to a low condenser vacuum signal.
ggg 14.2.12.2.7.4 Acceptance Criteria
- a. The condanner air removal filtration fans' capacities are within design specifications.
- b. The condenser air removal filtration dampers close on receipt of a safety injection signal.
- c. The rate at which the mechanical vacuum pumps reduce condenser pressure is within design specifications.
- d. The mechanical vacuum pumps start automatically on receipt of a low condenser vacuum signal.
14.2.12.2.8 Circulating Water System Preoperational Test (SU4-
, DA01) 14.2.12.2.8.1 Objective
- a. To demonstrate the operating characteristics of the circulating water pumps, vatar box venting pumps, and
- lh 14.2-130 Rev. O
( WOLF CREEK the condenser drain pump and verify the operation of their annociated control circuits,
- b. To demonstrate by operational test that the circulating water pump discharge valves operating times are within design specifications.
- c. To demonstrate that the gland water system flow to the circulating water pumps is within design specifications.
14.2.12.2.8.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are completed.
- b. Required electrical power supplies and control circuits are operational.
- c. The circulatng water system and condenser are available to receive flow from the circulating water pumps.
14.2.12.2.8.3 Test Method
- a. The circulating water pumps, water box venting pumps, and the condenser drain pump are operated and pump
( operating data is recorded.
- b. The response of the circulating water pumps and the condenser drain pump to control signals is verified.
- c. Circulating water pump discharge valve operating times are recorded.
14.2.12.2.8.4 Acceptance Criteria
- a. The circalating water pumps operating characteristics are within design specifications.
- b. The water box venting pumps operating characteristics are within design specifications.
- c. The condenser drain pump operating characteristics are
, within design specifications
- d. The condenser drain pump stops on receipt of a standpipe low-level signal.
- a. Each circulating water pump trips on receipt of a two out of three condenser pit high level signal.
O 14.2-131 Rev. O
WOLF CREEK
- f. Low gland seal water pressure or low gland seal flow will prevent start of the circulating water pumps,
- g. The gland seal water flow to each circulating water punp is within design specifications.
- h. The operating times of the circulating water pump dis-charge valves are within design specifications.
14.2.12.2.9 Service Water System Preoperational Test (S-04EA01). 14.2.12.2.9.1 Objectives
- a. To demonstrate the capability of the service water system and essential service water system to provide rated cooling water flow during the normal and normal-shutdown modes of operation to their respective loadn.
- b. To demonstrate the operating characteristics of the Service Water (SW) Pumps.
- c. To verify proper operation of site service water system controls and instrumentation.
14.2.12.2.9.2 Prerequisites
- a. Required component testing, instrument calibration, and '&
system flushing / cleaning are complete. W
- b. Required electrical power supplies and control circuits are operational.
- c. The essential service water system has been flow bal-anced in the LOCA mode.
- d. Site systen controls and instruments are calibrated.
- a. The SW system is available to receive flow from the SW pumps.
14.2.12.2.9.3 Test Method
- a. Service water and essential service water system flows are verified in the normal and normal-shutdown medes.
(The service water pumps provide the motive force.)
- b. The SW pumps are operated and pump operating data is recorded.
O 14.2-132 Rev. O
WOLF CREEK
)
14.2.12.2.9.4 Acceptance Criteria
- a. Components supplied by the service water system and essential service water system receive flows that are within design specifications with the system operating in the normal and normal-shutdown modes.
- b. The SW pumps operating characteristics are within design specifications.
14.2.12.2.10 Closed Cooling Water System Preoperational Test (S-04EB01) 14.2.12.2.10.1 objectives
- a. To demonstrate the capability of the closed cooling water system to provide cooling water flow to its asso-
-ciated components,
- b. To demonstrate the operating characteristics of the closed cooling water pumps and to verify that the asso-ciated instrumentation and controls are functioning properly.
14.2.12.2.10.2 Prerequisites
.(]
- a. Required component testin'g, instrument calibration, and system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
14.2.12.2.10.3 Test Method Performance characteristics of the closed cooling water pumps and flow data to supplied components are verified. 14.2.12.2.10.'4 Acceptance criteria
- a. The performance characteristics of each closed cooling water pump are within design specifications.
- b. Flow to all components supplied by the closed cooling water system is verified.
i 1 l 14.2-133 Rev. O i i
WOLF CREEK 14.2.12.2.11 Fire Protection System Preoperational Test (SU4-FP03) 14.2.12.2.11.1 Objectives
- a. To damonstrane the operating characteristics of the Fire Protection (FP) system jockey pump, motor-driven fire pump and th.s diesel-driven fire pump and verify the operation of their associated control circuits.
- b. To demonstrate the operability of the diesel oil system, including systen instrumentation and controls.
14.2.12.2.11.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are completed.
- b. Required electrical power supplies and control circuits are operational.
14.2.12.2.11.3 Test Method
- a. The jockey pump, the motor-driven fire pump and the diesel-driven fire pump are operated and operating data are recorded.
- b. The responsa of the motor-driven fire pump and diesel-driven firc pump to automatic start signals are veri-Ih fied.
- c. With the discol-driven fire pump operating at rated capacity, the capacity of the diesel oil day tank is verified.
14.2.12.2.11.4 Acceptance Criteria
- a. The FP pumps operating characteristics are within design specifications.
- b. The motor-driven fire pump and the diesel-driven fire pump automatically start upon receipt of their associa-ted decreasing fire protection system pressure signal.
- c. With the diesel fire pump operating at rated capacity, the capacity of the diesel oil day tank is within design specifications.
14.2-134 Rev. O O
i (^ t WOLP CRIIK
- d. With the diesel fire pump operating at rated capacity and upon receipt of a diesel oil day tank low level alarm, the remaining capacity of the diesel oil day tank is within design specifications.
14.2.12.2.12 Radwaste Building RVAC System preoperational Test (S-04GH01) l 14.2.12.2.12.1 Objectives
- a. To verify the radwaste building supply and exhaust fans' control circuits, including automatic transfer between exhaust fans.
- b. To' demonstrate the fan capacities of the radwaste build-ing supply and exhaust fans, recycle evaporator room fan coil unit, wasta evaporator room fan coil unit, control room (solidification) fan coil unit, sample laboratory fan coil unit, ground floor fan coil unit, basement floor fan coil unit, SLWS evaporator fan coil unit, and control room fan coil unit, and to verify that the associated instrumentation and controls function pro-perly.
14.2.12.2.12.2 Prerequisitos
- a. Required component testing, instrument calibration, and system air balancing are complete,
- b. Required electrical power supplies and control circuits are operational.
14.2.12.2.12.3 Test Method
- a. The radwaste building system fans are operated, and fan capacities are verified.
l b. Operability of the radwaste building supply and exhnust t fans' control circuits is verified. 14.2.12.2.12.4 Acceptance criteria
- a. The radwaste building system fan capacities are within design specifications.
- b. The radwaste building supply air unit will not opernte unless either radvaste exhaust fan is operating.
() 14.2-135 Rev. 0
..n. - -~ ,. c, ,
WOLF CREEK
- c. A low flow on the operating radwaste building 2xhaust fan will cause the operating fan to stop and the standby GI l fan to start.
14.2.12.2.13 Local containment Leak Rate Test (SU8-GP01) 14.2.12.2.13.1 Objectives l
-To determine the leakage rate of the containment penetrations and ;
the leakage rate of the containment isolation valves. 14.2.12.2.13.2 Prerequisites l
- a. All containment isolation valves are closed by normal actuation methods.
- b. Associated piping is drained, and v'ent paths for leakage are established.
- c. Required instrument calibration is complete. *
- 14 . 2 . .',2 . 2 .13 . 3 Test Method l i
The containment penetrations and containment isole. tion valves are leak tested by performing type ~B and type C tests, in accordance
-with 10.CTR 50, Appendix J. *-
14.2.12.2.13.4 Acceptance criteria The combined leakage from containment penetrations and containment , isolation valves is within design limits. 14.2.12.2.14 Liquid Radwaste System- Preoperational Test (S-04H301). 14.2.12.2.14.1 Objectives
- a. To demonstrate the cperating characteristics of the liquid radwaste system pumps and to verify the operation of their associated concrol circuits.
l b.- To demonstrate the operation of the liquid radwaste L system containment isolation valves, including their-response to a CIS.
- c. To determine by operational test that the liquid rad-
.wasta= system containment isolation valves' closure times
! are within design specifi:ations. 14.2-136 Rev. 0 O l l
+ -ww--g wv:e -y +w-y--e %y q,eg g-
q WOLT CREEK V 14.2.12.2.14.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The component cooling water system is available to provide cooling water to the reactor coolant drain tank heat exchanger.
14.2.12.2.14.3 Test Method
- a. The liquid radwaste system pumpe e.ro operated, and performance characteristics are recorded.
- b. The operability of the system pump and valve control circuits is verified.
- c. The liquid radwaste system containment isolation valves are operated under flow cenditions, and operating times are recorded.
14.2.12.2.14.4 Acceptance criteria
- a. The performance characteristics of the liquid radwaste system pumps are within deuign specifications.
- b. Each pump trips on receipt of a low-level signal from its respective tank.
- c. The liquid radwaste system containment isolation valves close on receipt of a CIS.
- d. The liquid radwaste system containment isolation valves' closure times are within design specifications,
- e. The liquid radwnste effluent discharge valve closes on a high process radiation signal.
14.2.12.2.15 Waste Evaporator Preoperational Test (SU4-HB02) 14.2.12.2.15.1 Objectives To demonstrate the operability of the waste evaporator and its associated pumps, valves, and control circuits. ( 14.2-137 Rev. O
WOLT CREEK 14.2.12.2.15.2 Prerequisites (g)
- a. Required couponent testing, instrument calibration, and system flushing / cleaning are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. Cooling water is available to the vaste evaporator.
J
- d. The auxiliary steam system is available to supply steam to the wasta evaporator.
- e. The vasta evaporator condensate tank and the primary evaporator bottoms tank are available to receive waste evaporator affluent.
14.2.12.2.15.3 Test Method
- a. The vasta evaporator is operated, and performance data is recorded.
- b. With the wasta evaporator in operation, a low feed inlet pressure signal is initiated, and the evaporator is verified to shift to the recycle mode.
- c. The vaste evaporator distillate pump is verified to trip on a low evaporator condenser level, ggg 14.2.12.2.15.4 Acceptanca Criteria
- a. The waste evaporator process flow is within design specifications.
- b. The waste evaporator goes into the recycle mod.e on low feed inlet pressure.
- c. The wasta evaporator cistillate pump trips on a low evaporator condenser level.
14.2.12.2.16 Solid Wasta System Preoperational Test (S-04EC01) 14.2.12.2.16.1 Objectives
- a. To demonstrate the operating characteristics of the solid waste system pumps and to verify the operation of their associated control circuits.
l l 14.2-138 Rev. O
l WOLF CRIEK O
- b. To demonstrate the ability of the decant station, drum-sing station, cement tilling station, and the solid redwaste bridge crane to process, solidify, and handle waste and to verify the operation of their associated control circuits.
- c. To demonstrate the ability of the dry waste compactors to process compressible wastes and to verify the opera-tion cf their associated control circuits.
14.2.12.2.16.2 prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operable.
- c. Reactor makeup water is available to provide a source of water to the decanting station.
14.2.12.2.16.3 Test Method
- a. The solid waste system pumps are operated, and the pump operating data are recorded.
P) \/ b. The system component control circuits are verified, and the ability of the solid radwaste system to process, solidity, and handle waste is verified. 14.2.12.2.16.4. . Acceptance Criteria '
- a. The operating characteristics of the evaporator bettens tank pumps (primary and secondary) are within design specifications.
- b. There are no free liquids present in the packaged waste. .
- c. The evaporator bottoms tank pumps (primary and secon-dary) trip on their respective tank lov level signal.
14.2.12.2.17 Solid- Waste Filter Handling System Preoperational Test (S-04HCO2) 14.2.12.2.17.1 Objectives To demonstrate the ability of the solid radwaste filter handling system to remove, transfer, and install a spent resin sluice filter assembly. ( ' 14.2-139 Rev. 0
WOLF CRIEK . 14.2.12.2.17.2 Prerequisites lll
- a. Requirnd conponent testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational. l l
14.2.12.2.17.3 Test Method I
- a. Operability of the solid radwasta monorail hoist and handling cask winch and associated control circuits is verified.
- b. The ability of the solid radwaste filter handling system to remove, transfer, and install a spent resin sluice filter assembly is verified.
14.2.12.2.17.4 Acceptance Criteria The filter handling system functions in accordance with design specifications. 14.2.12.2.18 Resin Transfer Preoperational Test (SU4-HCO3) 14.2.12.2.18.1 Objectives
- a. To demonstrate the ability to charge resins and acti-vated charcoal to those systems containing potentially lll contaminated dumineralizers or adsorbers. The ability of the spent resin sluice pumps to transfer resins and charcoal from demineralizers and avaorbers is also verified.
- b. To demonstrate tne operating characteristics of the spent resin sluice pumps, chemical addition metering pumps, and chemical drain tank pumps.
- c. To demonstrate the operability of system valve and pu=p control circuits.
14.2.12.2.18.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operational.
14.2-140 Rev. O
WOLF CRIEX (~))
- c. Those systems containing potentially contaminated domin-eralizers and adsorbers are available to cupport this test.
- d. The reactor makeup water system is available to provide a source of water for resin charging.
- e. A means of bulk disposal 10 available to receive vaste at the bulk disposal station.
14.2.12.2.18.3 Test Method
- a. Resins and charcoal are charged and transferred from selected potentially contaminated domineralizers and adsorbers.
- b. The spent resin sluice pumps, chemical addition metering pumps, and chemical drain tank pumps are operated, and performance characteristics are obtained.
- c. The response of the spent resin sluice pumps, chemical '
addition metering pumps, and the chemical drain tank pumps to a low-level trip signal from their respective tanks is verified. () 14.2.12.2.18.4 Acceptance criteria
- a. The operating characteristics of the spent resin sluice pumps, chemical addition metering pumps, and the chem-ical drain tank pump are within design specifications.
- b. The spent resin sluice pumps, chemical addition metering pumps, and the chemical drain tank pump trip on receipt of a low-level trip signal from their respective tanks.
14.2.12.2.19 Fire Protection System (Water) Preoperational Tact (SU4-KC01A, SU4-KColB) 14.2.12.2.19.1 Objectives
- a. To demonstrate the operability of the proaction sprink-1er system, wet-pipe sprinkler system, and the automatic water spray system, including system instrumentation, alarms, and interlocks.
- b. To demonstrate the operability of system valves, in-l ciuding their response to safety signals, i
l
~
14.2-141 Rev. 0
l l WOLF CREEK I
-c. To verify spray to the applicable electrical system 4 transformers.
14.2.12.2.19.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required electrical power supplies and control circuits are operable.
- c. The fire water pumps are available to provide a source of water to the fire protection system headers.
14.2.12.2.19.3 Test Method
- a. Response of the proaction sprinkler system, wet-pipe sprinkler system, and automatic water spray system to fare detection signals is verified, including the opera-bility of associated alarms, instrumentation, and inter-locks.
- b. The fire protection system containmont isolation valves are operated under flow conditions and operating times recorded,
- c. Response of the fire protection system containment a isolation valves to a CIS is verified. W
- d. Spray to the applicable electrical transformers is verified.
14.2.12.2.19.4 Acceptance criteria
- a. The proaction sprinkler system, wet-pipe sprinkler system, automatic water spray system and associated alarms, and instrumcntation and interlocks operate in accordance with system design specifications,
- b. The fire protection system containment isolation valves' closure time is within design specifications,
- c. The fire protection system containment isolation valves close on receipt of a CIs.
- d. The spray to applicable electrical transformers is within design specifications.
l 14.2-142 Rev. 0 l
WOLF CREEK O 14.2.12.2.20 Pire Protection Syr. tam (Halon) Preoperational Test (0-04KC02) 14.2 12.2.20.1 Cbjectives To demonstrate the operability of the halon cystem, including the associated instrumantation, control circuits, and alarms. 14.2.12.2.20.2 Preroquisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operable.
14.2.12.2.20.3 Test Method The operability of the halen system, including the associated instrumentation and alarms, is verified. System response to fire detection signals is also verified. 14.2.12.2.20.4 Acceptance Criteria The halon fire protection s operates in accordance with (~) system design specifications.ystem . v 14.2.12.2.21 Fire Protection System Detection and Alerm Preopera-tional Test (S-04KC03) 14.2.12.2.21.1 Objectives To demonstrate the operability of the fire protection system detectors and alarms not verified during the performanco of the halon and water system preoperational tests. 14.2.12.2.21.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operable.
14.2.12.2.21.3 Test Method Actuation of system alarms upon receipt of fire detection signals is verified. () 14.2-143 - Rev. 0 4 l ______.__.~..__..___.________.__..__._m _ _ . _ _ . - _ _ _ _ - - - - - _ -
l l 1 l WOLF CREEK l 1 l 14.2.12.2.21.4 Acceptance Criteria 1 Fire protection system detectors and alarms operate in accordance with system design specifications. l j 14.2.12.2.22 Oily Waste System precperational Test (S-04LE01) i 14.2.12.2.22.1 Objectives To demonstrate the sump pumps and miscellaneous condensate drain tank pumps' operating characteristics and response to sump / tank, level signals. The operation of system valves and associated control circuits and sump / tank level alarms are also verified. 14.2.12.2.22.2 Prerequisites
- a. Required corponent testing, instrument calibration, and system flushing / cleaning era complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The compressed air system is available to supply air to system valves and pumps.
- d. A water source (fire system) and a collection receptacle (oil / water separat.r, main condenser) are available for lll the testing of eac;. sump / tank.
14.2.12.2.22.3 Test Method
- a. The sump pumps and miscellaneous condensate drain tank pumps are operated, and performance characteristics are verified.
- b. The response of each pump and associated alarms to sump /
tank high and low level signals is verified.
- c. The operability of system air-operated valves is veri-fled, including the response to a precess radiation signal.
14.2.12.2.22.4 Acceptance Criteria
- a. The performance characteristics of the system pumps are within design specifications.
- b. The turbine building oily waste header discharge valve closes on a high-radiation signal.
l 14.2-144 Rev. 0 (l> l l -
WOLF CRIEK 14.2.12.2.23 Floor and Equipment Drain System Preoperational Test ! (SU4-LF01) ' 14.2.12.2.23.1 objectives To demcnstrate the sump pumps and hot machine shop oil interceptor pump's capacities and response to sump / tank level signals. The operation of system valves, their response to safety signals, and ' sump / tank level alarms are also verified. 14.2.12.2.23.2 Prerequisites '
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
t
- b. -Required electrical power supplies and control circuits are operational.
- c. The compressed air system is available to supply air to system valves and pumps.
- d. A water source (fire system or ESW) and a collection receptacle'(holdup tank, radwaste system, etc.) are available for the testing of each sump / tank.
O- 14.2.12.2.23.3 Test' Method a.- The sump pumps and hot machine shop oil inturceptor pumps are operated, and their capacities are verified.
- b. The response of each system pump, system-indication, and alarms, to sump / tank high and low level signals is veri-fled.
- c. The operability of system air- and motor-operated valves is verified, including their response to safety signals.
14.2.12.2.23.4- Acceptance Criteria
- a. The capacities of the floor and. equipment drain system pumps are within design specifications.
- b. System valves properly respond to safety injection sig-nals and containment isolation signals.
- c. The valve response times are within design specifica-tions.
l-14.2-145 Rev. 0 w.r-er ,,eww* -. A rm -ever ,em, .--v--w--,-+,.w,1,,-.,-m<v+--,-%--.~.<,--v*w-+,e-vey,,.wyvyve-y-,y-v-,wwt, +,-w-- ,+y-e w ,--we-e m - E+
WOLF CRIEK 14.2.12.2.24 13.8-kV System-Preoperational Test (S-04PA01) O 14.2.12.2.24.1 Objectives
- a. To demonstrate that the 13.8-kV busses can be energi:ed r from the startup transformer.
- b. To demonstrate that automatic fast transfer of the busses from the unit auxiliary source to the startup source is within design specifications.
- c. To demonstrate that the unit auxiliary source or startup trurce feeder breakers will trip on a stuck breaker condition.
- d. To demonstrate proper operation of system instrumentation and controla.
-14.2. . 2.24.2 Prerequisites -
- a. . Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. The 13.8-kV system has been energized. k 14.2.12.2.24.3 Test Method
- a. The 13.8-kV-busses are energized.frca the startup trans-former, and bus voltages are recorded.
- b. Automatic fast transfer from the unit auxiliary source to the startup source is verified.
- c. Stuck breaker conditions are simulated, and proper opera-tion of the 13.8-kV auxiliary source and startup source feeder breakers is verified.
14.2.12.2.24.4 Acceptance criteria
- a. The 13.8-kV bus voltages are within design specifica-tions, when energized from the startup transformer, b.- Automatic fast transfer of the -busses- from the unit auxiliary source to the startup source is within design specifications.
14.2-146 Rev. 0 O i
~
l ('h
<> h3LF CREEK
- c. The 13.8-kV auxiliary source and startup source feeder breakers trip on receipt of a stuck breaker signal.
14.2.12.2.25 4,160-V (Non-class IE) System Preoperational Test (S-04PB01) 14.2.12.2.23.1 Objectives A. To demonstrate that the 4,160-V busses can be energized from their normal and alternate sources, and to verify the optrability of supply breaker and bus tie breaker protectitre intarlocks.
- b. To demonstrate that automatic transfer is achieved through ths tie breaker from the normel source to the alternate source in the event of an elect tal fault.
- c. T- ,
anstrate proper operation of system instrumentation and controls. 14.2.12.2.25.2 Prerequisitek
- a. Required component testing and instrument calibration are complete.
() b. Required electrical power supplies and are operational. control circuits
- c. The 4,160-V (non-class IE) system has baen energized.
14.2.12.2.25.3 Test Method
- a. The 4,160-V non-Class IE busses are energi3ed from their l normal and alternate source, and bus voltages are re-I corded.
- b. System supply breakers and bus tie breakers are operated, and breaker interlocks are verified.
- c. System electrical fault signals are simulated, and auto-l estic transfer is verified through the tie breaker from the normal source to the alternate source for each 4,160-V bus.
14.2.12.2.25.4 Acceptance Criteria
- a. The voltage of each 4,160-V non-class IE bus, when sup-plied from its normal source and alternate source, is within design specifications.
14.2-147 Rev. O I
WOLF CREEK
- b. System supply breaker and bus tie breakar interlocks e
operate in accordance with the system design.
- c. Automatic ransfer is achieved through the tie breaker from the nc rmal source to the alternate source, for each 4,160-V but, upon receipt of an eluctrical fault signal.
14.2.12.2.26 480=Vo.'.t (Hon-Class IE) Systsm preoperational Test (S-04 P001) 14.2.12.2.26.1 objectives
- a. To demonstrate that the 480-V non-Class IE load centers can be energized from their normal sources and alternate sources, as applicable, and verify the operability of feeder breaker and bus tie breaker protective interlocks.
- b. To demonstrate that the 480-V busses supplied by 4160-V (Class IE) source breakers are shed on receipt of a load shed signal.
- c. To demonstrate proper operation of system instrumentation and controls.
14.2.12.2.26.2 Prerequisites
- a. Required component testing and instrument calibration are llh complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The 480-V (non-Class IE and Class IE) systems have been energized.
14.2.12.2.26.3 Test Method
- a. The 480-V non-Class IE load centers are energized from their normal source and alternate source, as applicable and voltages are recorded,
- b. System feeder breakers and bus tie breakers are operated, and breaker interlocks verified.
- c. A load shed signal is simulated, and the 480-V busses supplied by the 4,160-V (class IE) source breakers are V6rified to shed.
14.2-148 Rev. 0 O
WOLF CREEK 14.2.12.2.26.4 Acceptance criteria
- a. The voltage for each 480-V non-class IE load center, when supplied from its normal source and alternate source, as applicable, is within design specifications.
- b. system feeder breaker and bus tie breaker interlocks operate in accordance with the system design.
- c. The 480-V busses supplied by the 4160-V (Class IE) source breakers shed on receipt of a load shed signal.
14.2.12.2.27 250-V DC System Preoperational Test (S-04PJ01) 14.2.12.2.27.1 Objectives To demonstrate the ability of the battery and battery chargers to provida power to the busses. The battery chargers' ability to recharge their respective battery is also demonstrated. Proper operation of system instrumentation and controls is also verified. 14.2.12.2.27.2 Prerequisites
- a. Required component testing and instrument calibration are
() complete, I l
- b. Required electrical power supplies and control circuits 1
l are operational. '
- c. Ventilation for the battery room is available. l
- d. The 250-V de system has been energized.
14.2.12.2.27.3 Test Method
- a. The batter'y is discharged, using a test load at the design duty cycle discharge rate.
- b. The battery is fully discharged to determine its capacity i factor.
l l c. The ability of each battery charger to charge the battery to normal conditions, after the battery has undergone a design duty cycis, while simultaneously supplying power at a rate equivalent to the largest motor current load is verified. 1
- d. A load shed signal is initiated, and the battery charger PJ31 ac supply breaker is verified to trip.
O - 14.2-149 Rev. O I i y-
I l WOLP CREEK 14.2.12.2.27.4 Acceptance criteria O
- a. The battery is capable of maintaining output voltage above the design minimum, during a design duty cycle,
- b. The battery capacity factor is in accordance with design requirements.
- c. The battery chargers are able to recharge tho battery to normal conditions, after the battery has undergone a design duty cycle, while simultaneously supplying power at a rate equivalent to the largest motor current load.
- d. Battery charger PJ31 ac supply breaker trips on receipt of u load shed signal.
14.2.12.2.28 135-V (Non-class II) DC System Preoperational Test (S-04PX01, S-04PX02) 14.2.12.2.28.1 Objectives To demonstrate the ability of the batteries and chargers to pro-vide power to the busses. The battery chargers' e.bility to re-charge their respective battery is also demonstrated. Proper operation of system instrumentation and controls is also verified. 14.2.12.2.28.2 Prerequisites llI
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
- c. Ventilation for the battery room is available.
14.2.12.2.28.3 Test Method
- a. Each battery is discharged, using a test load at the design duty cycle discharge rate.
- b. Each battery is fully discharged to determine its capac~
ity factor,
- c. The ability of each battery charger to charge its respec-tive battery to normal conditions, after the tattery has i undergone a design duty cycle, while simultaneously I
supplying power at a rate equivalent te the design in-strumentation loading. 14.2-150 Rev. 0
O wo'r carrx l
- d. A safety injection load shed signal is initiated, and the batten charger PK21, PK22, PK23, and PK24 supply breaker is verified to trip.
14.2.12.2.20.4 Acceptance criteria
- a. Each battery is capable of maintaining output voltage above the design minimum, during a design duty cycle.
- b. Each battery capacity factor is in accordance with design requirements.
- c. The battery chargers are able to recharge the batteries to normal conditions, after the battery has undergone a design duty cycle, while simultaneously supplying power at a rate equivalent to the design load.
- d. Battery charger PK21, PK22, PK23, and PK24 supply brsever trips on receipt of a safety injection load shed signal.
14.2.12.2.29 Instrument AC (Ncn-Class IE) System Preoperational Test (S-04PN01) _ 14.2.12.2.29.1 Objectives k- To demonstrate that the 120-V non-Class IE ac distribution panels can be fed from their associated supply transformers. Proper operation of system instrumentation and controls is also verified. 14.2.12.2.29.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits j are operational.
14.2.12.2.29.3 Test Method The 120-V non-Clas: IE ac distribution panels are energi:ed from their associated supply transformers, and the panel voltages are recorded. 14.2.12.2.29.4 Acceptance Criteria Each 120-V non-Class IE ac distribution panel voltage in within design specifications. - 0 14.2-151 Rev. O
NOLF CREEK 14.2.12.2.30 Emergency Lighting System Preoperational Test (S-04QD01) 14.2.12.2.30.1 Objectives To demonstrate the capability of the emergency lighting system to provide adequate lighting. Proper operation of system instrumen-tation and controls is also verified. 14.2.12.2.30.2 Preraquisites Required electrical power supplies and control circuits are oper-able. 14.2.12.2.30.3 Test Method The ability of the emergency lighting system to provide adequate Jighting is verified. The operability of associated instrumen-tation and control circuits is also verified. 14.2.12.2.30.4 Acceptance Criteria The emergency lighting system operates in accordance with system design specifications. 14.2.12.2.31 Public Address System Preoperational Test (S-04QF01) 14.2.12.2.31.1 Objectives O To demonstrate the capability of the public address system to provide adequate intraplant communications and to verify the operability of the evacuation alarm system. 14.2.12.2.31.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operable.
14.2.12.2.31.3 Test Method
- a. The public address system is operated from all locations, and adequate communications verified.
- b. operability of the evacuation alarm system is verified.
14.2-152 Rev. O
(-) WOLF CREEK 14.2.12.2.31.4 Acceptance Criteria
- a. The evacuation alarm system operates in accordance with system design specifications.
14.2.12.2.32 Heat Tracing Freeza Protection System Preoperational Test (S-04QJ01) 14.2.12.2.32.1 Objectives To demonstrate the ability of the freeze protection system to automatically control the associated heat tracing circuits in accordance with system design. The operation of system instrumen-tation and controls is also verified. 14.2.12.2.32.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational.
14.2.12.2.32.3 Test Method
' Temperature signals are varied and the energization/ deenergiza-tion of the associated heat tracing circuits is verified.
14.2.12.2.32.4 Acceptance Criteria The freeze protection system automatically controls the associated heat tracing circuits, in accordance with system design. 14.2.12.2.33 Secondary Sampling System Preoperational Tesr (S-04RMC1) 14.2.12.2.33.1 Objectives
- a. To demonstrate the operating characteristics of the steam generator blowdown sample drain tank pump, sample chiller pump, and the condenser sample pumps, and verify the operability of their associated control circuits,
- b. To demonstrate that the system sample flows are within design specifications.
14.2.12.2.33.2 Prerequisites
- e. Required component testing, instrument calibration, and system flushing / cleaning are complete.
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i WOLF CREEK
- b. Required electrical power supplies and control circuits are operable,
- c. Plant conditions are established, and systems are avail-able, as necessary, to facilitate drawing samples from the sample points.
- d. The steam generator blowdown system is available to receive effluent from tha steam generator blowdown sample drain tank.
- e. The closed cooling water system is available to provide cooling water to the system sample coolers and chiller package.
14.2.12.2.33.3 Test Method
- a. The steam generator blowdown sample drain tank pump, sample chiller pump, and the condenser sample pumps are operated, and pump performance data recorded. Operabil-ity of their associated control circuits is also vari-fied.
- b. System samples are obtained, and flows are recorded.
14.2.12.2.33.4 Acceptance Criteria
- a. The steam generator blowdown sample drain tank pump, sample chiller pump, and coadenser sample pump perform-ance characteristics are within design specifications,
- b. Sample system flows are within design specifications.
14.2.12.2.34 Area Radiation Monitoring Preoperational Test (S-04SD01) 14.2.12.2.34.1 Objectives To demonstrate the operation of the area radiation monitors and to verif*f that a high radiation signal at each monitor will initiate an alarm. 14.2.12.2.34.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operatientl.
14.2-154 Rev. 0
rm -() WOL7 CREEK 4 14.2.12.2.34.3 Test Method A calibration source is utilized to actuate the area radiation monitors, and their operability and associated alarms are veri-fled. 14.2.12.2.34.4 Acceptance criteria Each area radiation monitor actuates the associated alarms, en receipt of a high radiation signal. 14.2.12.2.35 Seismic Monitoring Instrumentation Systen Preopera-tional Test (S-04SG01) 14.2.12.2.35.1 Objectives To demonstrate the operability of the seismic triggers and switches and otrong motion accelerometers, including their asse-ciated alarms Lnd recording and playback systems. 14.2.12.2.35.2 Prerequisites
- a. Required component testing and instrument calibration are 7-q complete.
- b. Required electrical power supplies anc. control circuits are operational.
14.2.12.2.35.3 Test Method A test signal is initiated, and the operability of the seismic triggers and switches and strong motion accelerometers, including their associated alarms and recording and playback systems, is verified. 14.2.12.2.35.4 Acceptance Criteria The seismic triggers and switches and strong motion accelerome-ters, including their associated alarms and recording and playback syste=s, operate in accordance with system design specifications. 14.2.12.2.36 Loose Parts Monitoring System Test (SU4-SQO2). 14.2.12.2.36.1 Objective To demonstrate the operability of the accelerometers, signal conditioning devices and diagnostic equipment, including associa-ted alarms and recording and playback systems. D 14.2-155 Rev. O
WOLF CREEK. O 14.2.12.2.36.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control ciroudts are operational.
- c. Reactor coolant system is filled with water,
- d. Reactor coolant system is at normal operating temperature and pressure with all reactor coolant pumps running, and hot functional testing is in progress (for those portions of the testing to be performed during hot functional testing).
- e. Reactor coolant system is at normal operating temperature and pressure with all reactor coolant pumps running after fuel loading during startup testing (for those portions of the testing to be performed during startup testing).
14.2.12.2.36.3 Test Method
- a. Test signals are initiated and the operability of the accelerometers, signal conditioners, and diagnostic circuitry, including alarms and recording and pl~yback systems, is verified.
a g
- b. Channel audio outputs are also recorded during het fune-tional testing and aftnr fuel leading during startup testing to obtain a record of the reactor coolant system noise " signature."
14.2.12.2.36.4 Acceptance Criteria The accelerometers, signal conditioners, and diagnostic circuitry, including alarms and recording and playback systems operate to detect loose parts as specified in USAR Section 4.4.6.4. 14.2.12.2.37 Plant Performance Test (SU8-0007) 14.2.12.2.37.1 Objectives
- a. To monitor the balance-of-plant and electrical systems under loaded conditions during het functional and power ascension testing. The ability cf the ventilation sys-tems to maintain ambient temperatures within design limits is also verified. To monitor the concrete temp-eratures surrounding hot penetrations and to verify evacuation alar = audibility in high noise areas.
O 14.2-156 Rev. O
(g) WOLF CREEK 14.2.12.2.37.2 Prerequisites
- a. Required component testing, instrument calibration, and system flushing / cleaning are complete.
- b. Required HVAC systems have been balanced.
- c. Maguired electrical power supplies and control circuits are operational.
14.2.12.2.37.3 Test Method This procedure does not provide a test method. It provides a monitoring and data collection function only, with the resultant datum evaluated against provided design values, as applicable. 14.2.12.2.37.4 Acceptance Criteria
- a. Evacuation alarm audibility in high noise areas is veri-fled.
- b. The containment coolers maintain containment temperature within design.
N Note Each monitored point is evaluated throughout the test to ('J N- verify that the applicable system or component is functioning per design. 14.2.12.2.38 Electrical Distribution System Voltage Verification Test'(S-090023) 14.2.12.2.38.1 Objectives To record actual loaded electrical distribution parameters during various steady-state and transient conditions. 14.2.12.2.38.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Requi::ed electrical power supplies and control circuits are operational.
14.2.12.2.38.3 Test Method The bus voltages and loadings of the electrical distribution system (down to the class lE 120/208 V ac system) are recorded for O 14.2-157 Rev. 0
WOLF CREEK various steady state configurations. Data is also recorded during the starting of the largest class lE and non-class 1E motors. All monitored busses are loaded to at least 30 percent. 144 2.12.2.38.4 Acceptance Criteria Not applicable. Note: The data obtained from this test procedure ara used to verify electrical system voltage analysis. 14.2.12.3 Startup Test Procedures The following sections are the test abstracts for each ctartup test. Table 14.2-3 provides an index of these tosts. 14.2.12.3.1 Automatic Steam Generator Level Control (S-07AB01) 14.2.12.3.1.1 Objectives
- a. To verify the stability of the automatic steam generator level control following simulated transients at low power conditions and the proper operation of the variable speed feature of the feedwater pumps,
- b. To demonstrate the performance characteristics of the steam generator feedwater pumps.
(l) 14.2.12.3.1.2 Prerequisites
- e. The steam generator level control system has been checked and calibrated,
- b. Steam generator level instruments and set points have been set and calibrated.
- c. Main feedwater is operational.
14.2.12.3.1.3 Test Method
- a. Induce simulated steam generator level transients to verify proper steam generator level control rssponse,
- b. Verify the variable speed features of the steam generator feedwater pumps by manipulation of controllers and test input signals, and verify the perfor=ance characteristics of the steam generator feedwater pumps.
l 14.2-15a Rev. O
1 l p) s
- WOLE CREEK 14.2.12.3.1.4 Acceptance Criteria
- a. Automatic steam generator level control system response must be in accordance with the vendor's technical manual,
- b. The steam generator feedwater pump's performance charac-teristics are within design specifications.
14.2.12.3.2. Dynamic Automatic Steam Dump Control (SU7-AB02) 14.2.12.3.2.1 Objectives To verify automatic operation of the T average steam dump control system, dsmonstrate controller setpoint adequacy, and obtain final settings for steam, pressure control of the condenser dump valves. 14.2.12.3.2.2 Prerequisites
- a. The reactor coolant system is at normal operating pres-sure and temperature.
- b. The reactor is critical,
- c. The steam dump system has been checked and calibrated.
- d. Main feedwater and the condenser are operational.
14.2.12.3.2.3 Test Method
- a. Reactor power is increased by rod withdrawal and steam dump to condenser to demonstrate setpoint adequacy.
- b. Pressure controller setpoint is increased prior to switching to T average control, which will rapidly modulate open condenser dump valves,
- c. Simulate turbine cperating conditions with reactor at power, then simulate turbine trip, resulting in the rapid opening of the steam dump valves.
14.2.12.3.2.4 Acceptance Criteria The steam dump system controllers must maintain stable eactor coolant system T average at the controllers set point with no divergent oscillations. 14.2.12.3.3 RTD Bypass Flow Measurement (S-07BB01) 14.2.12.3.3.1 objectives t ( To determine the , flow rate necessary to achieve the design reactor coolant transport time in each resistance temperature detector 14.2-159 Rev. 0 l
WOLF CREEK (RTD) bypass loop and to measure the flow rate in each RTD bypass loop to ensure that the transport times are acceptable. 14.2.12.3.3.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The reactor core is installed, and the plant is at normal operating temperature and pressure with all reactor coolant pumps running.
14.2.12.3.3.3 Test Method The flow rate necessary to achieve the design reactor coolant transport time for each hot and cold leg bypaus loop is calcu-lated, utilizing the hot and cold leg RTD bypass loop piping lengths. Hot and cold RTD bypass loop flow data are recorded. 14.2.12.3.3.4 Acceptance Criteria The flow rate in each het and cold leg RTD bypass loop, required to achieve the design reactor coolant transport time, is within design specifications, h 14.2.12.3.4 Pressurizar Heater and Spray capability Test (S-07BB02) 14.2.12.3.4.1 Objectives To determine the rate of pressure reduction caused by fully open-ing the pressurizer spray valves and the rate of pressure increase from the operation of all pressuri er heaters. 14.2.12.3.*.2 Prerequisites
- a. Required component testing and instrument calibration are complete.
- b. Required electrical power supplies and control circuits are operational,
- c. The reactor core is installed with the plant in the hot shutdown condition at normal operating temperature and pressure with all reactor coolant pumps running.
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- d. The final setting of the continuous upray flow valves is complete.
- e. The reactor coolant system is berated to the value re-quired for fuel loading.
- f. .This test is performed ~ prior to initial criticality.
14.2.12.3.4.3 Test Method
- a. With the pressurizar spray valves closed, all pressurizer heaters are energized, and the time to reach a 2,300 psig system pressure is measured and recorded.
- b. With the pressurizar heaters deanergized, both spray valves are fully opened, and the time to reach a 2,000 psig system pressure is measured and recorded.
14.2.12.3.4.4 Acceptance Criteria The pressurizer pressure response to the opening of the pressur-iter spray valves and to the actuation of all pressurizer heaters is within design limits. () 14.2.12.3.5 Reactor Coolant System Flow Measurement (S-073B03) 14.2.12.3.5.1 ' Objectives
- a. To confirm, after core installation but before initial critical operation, that reactor coolant system (RCS) flow rate as measured by loop elbow differential pressure readings is greater than or equal to 90 percent of the thermal design flow rate,
- b. To confirm during initial power operation that RCS flow rate as computed from calorimetric data is greater than or equal to the thermal design flow rate.
14.2.12.3.5.2 Prerequisites
- a. Required component testing and instri. ment calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
- c. The reactor core is installed, and the plant is at normal operating temperature and pressure.
A V 14.2-151 Rev. 0
. . . .- . . .- .- -- - . - -~ - - - - - . . - - -
=
-WOLF CREEK-
. 14.2'.12.3.5.3- Test Method- -
- a. -Before critical operationi loop elbow differential pres-sure readings =are taken with all greactor coolant- pumps running, and RCS flow rate is calculated.
- b. During initial power ~ operation, calorimetric data are taken=from Procedure 5-075C03, " Thermal Power Measurement and Statopoint Data ' Collection,"- and RCS flow rate is calculated.
14.2.12.3.5.4 Acceptance criteria-- RCS flow rate by loop elbow differential pressure' measurement is greater than or- equal-to.90 percent'of the thermal design value Land;by calculationffrom calorimetric data is greater than_.or equal tosthe thermal-design-value. 14.2.12.3.6 ' Reactor Coolant System Flow coastdown' Test (SU7-BB04) 14.2.12.3.6.1( Objectives-
- a. To- measure: the rate at which reactor coolant' flow changes,.' subsequent to simultaneously tripping all reac-tor coolant pumps.-
- b. To -datermine~ that the reactor coolant system-low-flow &
delay time is less than or equal to- the total low-flow W-
- delay time -assumed in the safety analysis'for loss of flow.
-14.2.12.3.6.2 Prerequisites a.--. Required component testing and' instrument calibration are complete.-
- b. _ Required electrical power supplies and control circuits are operational.-
- c. The reactor core is installed, and the. plant is at normal
- operating.Ltemperature and pressure with all reactor c . coolant pumps. running.
'14.2.12.3.6.3- Test Method Flow. coastdown' stabilization and loss cf coolant delay-time data
.are. recorded while tripping reactor-coolant pumps.
14.2-162 Rev, 0
f) v WOLF CREEK 14.2.12.3.6.4 Acceptance Criteria
- a. The rate of change of reactor coolant flow is- within design specifications.
(-
- b. The reactor coolant system low-flow delay tina is less than or equal to the total low-flow delay time assumed in the safety analysis for loss of flow.
14.2.12.3.7 Pressurizar Continuous Spray Flow Verification (S-07BB05) 14.2.12.3.7.1 Objectives To establish a setting for the pressurizer continuous spray flow valves to obtain an optimum continuous spray flow. 14.2.12.3.7.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits are operational.
e) (~ c. The reactor core is in~ stalled with the plant in the het shutdown condition at normal operating temperature and pressure with all reactor coolant pumps running.
- d. The reactor coolant system is borated to the value re-quired for fuel loading.
- e. This test shall be performed prior to initial critical-ity.
- f. The preliminary setting of the continuous spray flow valves has been completed during het functional testing.
14.2.12.3.7.3 Test Method Continuous spray flow valves are adjusted to establish the optimum continuous spray flow, and the valve throttle positions are re-corded. 14.2.12.3.7.4 Acceptance Criteria The continuous spray flow valves are throttled to establish the optimum continuous spray flow to keep the spray line warm and minimize normal steady-state pressurizer heater loads, f~h u.) 14.2-163 Pev. O
WOLF CREEK 14.2.12.3.8 RTD/TC Cross Calibration (S-07BB06) O 14.2.12.3.8.1 Objectives
- a. To provide a functional checcout of tha reactor coolant system resistance temperatura detectors (RTDs) and Incore thermocouples and to generate isothermal cross-calibration data for subt.equent correction factors to indicated temperatures.
NOTE 1 This portion of the test ne ed be performed only if the data collected in S-03Bhl6, RTD/TC Cross Calibration, l during hot functional testing, so warrants. I
- b. To provide a functional checkout of the core subcooling l monitor system including the detecting thermocouples. l 14.2.12.3.8.2 Prerequisites
- a. Required component testing and instrument calibration are i complete.
1
- b. Required e19ctrical power supplies and control circuits are operational.
- c. Plant heatup, following core loading, is in progress, and O all reactor coolant pumps are operating.
14.2.12.3.8.3 Test Method
- a. At various temperature plateaus RTD and incere thermo- l couple data are recorded. Isother:al cross-calibration correction factors for individual thermocouples and the insta11ction corrections for individual RTDs are deter-mined.
- b. At normal operating temperature, the thermocouple core i subcooling monitors' operational and programmable func- '
tiens are verified, including associated alarms, dis-plays, and printouts. 14.2.12.3.8.4 Acceptance Criteria
- a. Individual RTD readings are within the design specifica-tions.
- b. The installation corrections of the RTDs are within design specifications.
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- c. The thermocouple core subcooling monitor alarms, dis-plays, and printouts function in accordance with design specifications.
14.2.12.3.9 Core Loading Instrumentation and Neutron Source Requirements (S-07SC01) 14.2.12.3.9.1 Objectives To verify proper alignment, calibration, and neutron response of the temporary core loading instrumentation prior to start of fuel-loading; to check the neutron response of the nuclear instrumenta-tion system (NIS) source range channels prior to start of fuel-loading; and to check the neutron response of the temporary and NIS source range instrumentation prior to resumption of fuel-loading following any delay of B hours or more. To verify the signal to noise ratio is greater than 2. 14.2.12.3.9.2 Prerequisites .j a. Hot functional testing is completed,
- b. The nuclear instrumentation system is installed and calibrated.
'3 r 'w d 14.2.12.3.9;3 Test Method
- a. A portable neutron source (1-5 curie), plus preshipment equipment checkout data, is used to verify proper align-ment, calibration, and neutron response of the temporary core-loading instrumentation,
- b. A portable neutron source (1-5 curia) is used to check the neutron response of the NIS source range detectors.
- c. A portable neutron source,(1-5 curie) or movement of a sourcs-bearing fuel element to produce the desired change in neutron level to verify the neutron response of the temporar and NIS source range instrumentation prior to resumpt_.. of fuel-loading following any delay of 8 hours or more.
- d. Perform a statistical evaluation of 10 observations for each channel, to verify operability of the equipment.
14.2.12.3.9.4 Acceptance Criteria Neutron instrumentation is operational, calibrated, and indicates a positive / negative change in count rate as the neutron level is increased and/or decreased. The signal to noise ratio is greater than 2. 14.2-165 Rev. O i l
WOLF CREEK O 14.2.12.3.10 Thermal Power Measurement and Statepoint 'cata Col-lection (S-07SCO3) 14.2.12.3.10.1 Objectives To measare core therme,1 power and obtair, data for instrumentation calibration. 14.2.13.3.10.2 Prereqaisites
- a. Calorimetric instrumentation is installed.
- b. This test is performed at 30-percent, 50-percent, 75-percent, 90-percent, and 100-percent power.
14.2.12.3.10.3 Test Hethod Collect data and calculate thermal power, obtain statapoint data, compute the averaga fcr each parameter measured, convert to the appropriate units, end summarize the dsta for each RCS loop. 14.2.12.3.10.4 Acceptance Criteria This test is for the collection of data. 14.2.12.3.11 Nuclear Instrumentation System Test (SU7-SE01) g 14.2.12.3.11.1 Objectives The purpose of this test is to verify that the nuclear instrumen-tation system performs the required indications and control func-tions through the source, intermediate, and power ranges of opera-tion prior to core loading. 14.2.12.3.11.2 Prerequisites
- a. The nuclear instrumentation system is installed, calibra-ted, aligned, and operational for a period of at least 4 hours.
- b. The plant is at ambient temperature and pressure.
14.2.12.3.11.3 Test Method
- a. The source and intermediate range channels are subjected to various test signals to verify that the appropriate indicators alarm, illuminate, or actuate, and the source range local and remote speakers function.
14.2-166 Rev. O O
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- . .. . -. - .- .~ - - . . . - - - .~.. .-
5 e f! WOLF CREEK-b.. The? power? range channels are subjected to-various test signals to observe proper meter reading'and function of the comparator and rate circuitry.
- c. =The-high' voltage circuitry of the source und intermediate range channels.is tested.
~
14.2.12.3.11.4 -Acceptance-Criteria The1 control and indication functions and the reactor trip set points of the nuclear-instrumentation system' source, .inte rmediate , and power range channels have been verified. 114.2.12.3.12;- operational : Alignment of Nuclear Instrumentation (S-075E02) 14.2.12.3.12.1 - objectiver To. establish ~and determine voltage settings, trip settings, opera-
-tional. settings, alarm settings, and overlap of channels en source range,: intermediate range, and power. range instrumentation from prior to initial criticality to:at or near full reactor power.
; .14.2.12.3.12.2 Prerequisites a.- 13un nuclear instrumentation system has been aligned.
I : b.- This testiis conducted prior to criticality, during power escalation, and at or near full power. $ -14.2.12.3.12.3. Test Method
- a. All= functions.'are -
calibrated,' tested, and verified, utilizing permanent 1y' installed controls and- adjustment mechanisms. b.- Operational modes of the source range, intermediate range, .and power range channels are set for their-proper functions, as'per:the test instructions. 14.2.12.3.12.4 Acceptance Criteria-The' overlap.between th's source, intermediate, and power -range channels must be at' least 1-1/2 decades, -and the power range-channels are-capable of being adjusted:to. agree with the results of plant calorimetric calculations. L 14.2-167 Rev. 0 J ., - . l. a . - . . . - , - - . , . , - . . . - - ,_... .. . - o #
WOLF CREEK 14.2.12.3.13 Axial Flux Difference Instrumentation Calibration (S-07SE03) 14.2.12.3.13.1 Objectives To derive calibration factors for overpower, overtemperature, and T setpoints, based on incore flux data, calorimetric data, and excore nuclear instrumentation detector currents. 14.2.12.3.13.2 prerequisites
- a. The axial flux difference instrumentation system has been aligned.
- b. Data has been obtained from flux maps taken at 30-percent and 50-percent power.
14.2.12.3.13.3 Test Method Collect data, as required by test instruction, at 50-percent and 75-percsnt power, perform FaI calculations to calibration factors, and extrapolate results for use at the 100-percent power plateau. 14.2.12.3.13.4 Acceptance Criteria Calibration factors agree with Technical Specifications. 14.2.12.3.14 Control Rod Drive Mechanism operational Test (S- O 07SF01) 14.2.12.3.14.1 Objectives To demonstrate the proper operation of the red drive mechanisms under both cold and hot plant conditions and to provide verifica-tion of proper slave cycler timing.
~
14.2.12.3.14.2 prerequisites
- a. The control rod drive mechanisms are installed.
- b. The rod drive motor-generator sets are installed and power is available,
- c. For the control rod drive mechanism timing test, the core is installed, rod position indication is installed, and the control rod driving mechanism cooling fans are opera-tional,
- d. Nuclear instrumentation channels operable and operating.
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- e. A fast speed oscillograph (Visicorder or equivalent) to monitor test parameters is available.
14.2.12.3.14.3 Test Method
- a. With the reactor core installed and reactor in the cold shutdown condition, confirm that the slava cycler devices supply operating signals to the proper CRDM stepping magnet coils,
- b. Verify proper operation of all CRDMs under both cold and hot shutdown conditions. CRDM magnet coil currents and audio noise signals are recorded.
14.2.12.3.14.4 Acceptance Criteria The control rod drive mechanisms conform to the requirements for proper mechanism operation and timina as described in the magnetic control rod drive mechanism instruction manual. 14.2.12.3.15 Rod Control System (S-07SF02) 14.2.12.3.15.1 Objectives ('T
's,) To demonstrate and document that the rod control system performs the required control and indication functions just prior to init-ial criticality. To demonstrate operation of the rod inhibit functions.
14.2.12.3.15.2 Prerequisites
- a. The reactor coolant system is at normal operating pres-sure and temperature.
- b. The rod control system is installed and aligned.
- c. The source range nuclear instruments are operable.
- d. The rods are capable of withdrawal.
- e. The rod position indication system is operable.
14.2.12.3.15.3 Test Method
- a. With the reactor at no load opsrating temperature and pressure, and just prior to initial criticality, the control is checked for each applicabin position of the bank selector switch for proper operation.
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WOLF CREEK
- b. Status lights, alarms, and indicators are verified.
14.2.12.3.15.4 Acceptance Criteria The control and indication functions in accordance with the red position indication system and rod control system manuals. Rod motion is inhibited upon application of an inhibit function. 14.2.12.3.16 Rod Drop Time Measurement (SU7-SF03) 14.2.12.3.16.1 Objectives To determine the red drop time of each rod cluster control assem-bly under no-flow and full-flow conditions, with the reactor in the cold shutdown condition and at normal operating tenperature and pressure. 14.2.12.3.16.2 Prerequisites
- a. Initial core loading is completed.
- b. Rod control system is installed and tested.
- c. Individual red position indication is installed and checked.
14.2.12.3.16.3 Test Method ggg Withdrav each rod cluster control assembly, interrupt the elec-trical power to the associated rod drive mechanism, and measure and record the red drop time. This test is performed with the reactor at cold and hot conditions and at no-flow and full-flow. 14.2.12.3.16.4 Acceptance Criteria Tne rod drop tiras are acceptable in accordance c..th plant tech-nical specifications. 14.2.12.3.17 Rod Position Indication System (SU7-SF04) 14.2.12.3.17.1 Objectives To verify that the rod position indication system satisfactorily performs required indication functions for each individual rod and that each rod operates satisfactorily over its entire range of travel. 14.2-170 Rev. O
V- WOLF CREEK 14.2.12.3.17.2 Prerequisitas
- a. Plant system conditions are established as follows:
- 1. Test performed at Tavg s200' F, nominal RCS pressure for Tavg noted
- 2. Test results verified at Tavg nominally 557' F, RCS pressure nominally 2235 psig and at least one reactor coolant pump in service.
14.2.12.3.17.3 Test Method
- a. All shutdown rod banks are fully withdrawn by bank stop-ping at 18,210 and 228 steps to record the rod position, the Digital Rod Position Indication display (DRPI), and the group step position indication.
- b. All control rod banks are fully withdrawn by bank in 24 step increments while recording rod position as indicated by the plant control room DRPI readout, and the group step position indication,
- c. In addition, the pulse-to-analog converter chassis bank g-)
(_ position digital readout is recorded for all control rod banks. 14.2.12.3.17.4 Acceptance Criteria The rod position indication system performs the required indica-tion functions, and each rod operates over its entire range of travel within the limits of the rod position indication instruc-tion manual and the plant precautions, limitations,setpoints manual, and WCGS Technical Specifications. 14.2.12.3.18 Automatic Reactor Control System (S-07SF05) 14.2.12.3.18.1 Objectives To demonstrate the capability of the reactor control system to respond properly to input signals and to transmit proper control signals to other plant control systems and components. 14.2.12.3.18.2 Prerequisites
- a. The reactor is at approximately 30-percent power.
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WOLF CREEK-
.b. Pressurizer level and pressure, steam dump, steam genera-tor --- level, and main feed pump. speed control systems are in automatic.
14.2.12.3.18.3 Test Method T average will be successively varied from the- Tref att point to verify the transient recovery capabilities of the auto reactor control system. 14.2.12.3.18.4 Acceptance-Criteria \.
- a. No manual intervention should be required to bring the plant conditions to equilibrium values following initia-tion of a 6 F temperature transient.
i b.- Tavq should return to within +1.5 F of Tref f0110ViU9 l_ initiation of a 6 F temperature t?ansient.
- c. Rod motion is inhibited'by application of the appropriate inhibit inputs.
14.2.1243.19 Incore Flux Mapping-(S-07SR01, S-07SR02) 14.2.12.3.19.1 Objectives i: To obtain' core power and tamparature profiles for evaluating L performance. core lll 14.2.12.3.19.2 Prerequisitna
- a. The incore monitoring system has been functionally i l tested. '
- b. This test is performed at low power, 30 , 50 , 75 , 90 ,
i and 100-percent power.
- c. The reactor is stabilized prior to taking a map.
14.2.12.3.19.3 Test Method i The movable detectors are inserted into the core, data is obtained, and thermocouples are monitored while at a stable power. The obtained data is retained for evaluation. 14.2.12.3.19.4- Acceptance Criteria Flux and temperature data is-obtained at the various power levels. 14.2-172 Rev. O O
() O WOLP CREEK 14.2.12.3.20 Incore Instrumentation Test (S-07SR03, S-07SR04) 14.2.12.3.20.1 Objectives To set up and demonstrate operation of the incore instrumentation system. 14.2.12.3.20.2 Prerequisites
- a. The incere instrumentation system is installed.
- b. Proper rotation and limit switch operation has been verified.
- c. Testing is performed at cold shutdown and hot standby.
14.2.12.3.20.3 Test Method At cold shutdown a dummy cable is inserted into each thimble, and proper rotation and limit switch operation is verified. At het standby the detectors are inserted into the thimbles to demon-strate performance in all operational modes. 14.2.12.3.20.4 Acceptance Criteria \ The incere instrumentation system is capable of taking a flux map. 14.2.12.3.21 Operational Alignment of Procesc Temperature Instru-mentation ($-07SF06) 14.2.12.3.21.1 objectives To. align AT and Tava process instrumentation under isothermal conditions, prior to criticality and at power. 14.2.12.3.21.2 Prerequisites
- a. This alignment is performed prior to initial criticality and again at 75-percent power. Alignment is checked at 100-percent power.
- b. All reactor coolant pumps shall be operating.
14.2.12.3.21.3 Test Method
- a. Align- A T and Tavg per test instructions under isothermal conditions prior to criticality and at approximately 75 O)
\_ 14.2-173 Rev. O
WOLF CREEK percent power. Extrapolate the 75-percent. power data to O determine-oT: and Tavg-values for the.100-percent power plateau,
- b. ~At or_near full power, check the alignment of the AT and Tave channels for agreement with the results-of thermal pow 4r measurement. Realign any channels, as necessary, to meet test specifications.
14.2.12.3.21.4 Acceptance _ Criteria # The 100 percent power indications for AT and Tave channels must be within the maximum design values as specified in vendor design documents. 14.2.12.3.22 Startup Adjustments of Reactor Control- System (S-07SF07)- 14.2.12.3.22.1 Objectives To obtain the optimum plant efficiency. 14.2.12.3.22.2 Prerequisites
- a. The reactor coolant system is at normal operating pres-sure and temperature.
- b. Plant instrumentation shall have been aligned according lll to operational Alignment of Process Temperature Instru-mentation. -
- c. The turbine control system shall_have been aligned.
14.2.12.3.22.3 Test Method
- a. obtain system 1 temperature and steam pressure data at steady-state conditions for zero-power _and at hold points during power escalations.
- b. Evaluation of these data will provide the-basis for adjustments to the reactor control system.
14.2.12.3.22.4 Acceptance Criteria The-Tavg controller must be capable of maintaining full load steam pressure within turbine pressure limitations specified in the vendor's technical manual. $ 14.2-174 Rev. O O
l WOLF CREEK 14.2.12.3.23 RCCA or Bank Worth Measurement at Zero Power (S-07SF08) 14.2.12.3.23.1 Objectives To determine the differential and integral reactivity worth of a red cluster control bank (RCC) or an individual rod cluster con-trol assembly (RCCA). 14.2.12.3.23.2 Prerequisites
- a. The reactor is critical with the neutron flux level within the range established for zero power physics testing.
- b. The reactor coolant system is at normal operating pres-sure and temperature.
14.2.12.3.23.3 Test Method RCC and RCCA worth are validated by constant addition and/or dilution of boron in the reactor coolant system, causing rod movement to compensate for the boren addition and/or dilution. This rod movement will cause step changes in reactivity which are used to compute the worths. 14.2.12.3.23.4 Acceptance Criteria The integral reactivity worth of the RCC or RCCA over its entire range of travel agrees with acceptance criteria given in the Nuclear Design Report within tolerance values specified in vender design documents. 14.2.12.3.24 RCCA or Bank Worth Measurement at Power (SU7-SF09) 14.2.12.3.24.1 Objectives
- a. To measure RCCA worth for a rod ejected from the HFP rod insertion limit position,
- b. To determine in-core response resulting from a dropped rod with all other control rods near fully withdrawn.
14.2-175 Rev. O i 1 J
L WOLF CRIIK I 14.2.12.3.24.2 Prerequisites (l> Testing will be performed at 30-percent power with the reactor stable.
- 14.2.12.3.24.3 Test Method
- a. Ejected rod - Compute the change in reactivity associated with the change in RCCA position.
- b. Dropped rod - Determine the quadrant power tilt ratio and hot channel factors by use of the in-core flux mapping system.
14.2.12.3.24.4 Acceptance Criteria
- a. Ejected rod - The rod worth of the ejected rod is within tolerance values specified in vendor design documents.
- b. Dropped rod - The peaking factors are within the limits specified in vender design documents.
14.2.12.S.25 Reactor Systems Sampling for Core Load (S-07SJ01) 14.2.12.3.25.1 Objectives To verify uniform boron concentr'ation, prior to core load, reactor coolant system and directly connected auxiliary systems. in the lll 14.2.12.3.25.2 Prerequisites
- a. Beric acid tanks, pumps, and transfer lines are all filled with 4 percent beric acid solution,
- b. Reactor coolant system is filled with reactor grade water which has been borated to a concentration as specified in the technical specifications.
14.2.12.3.25.3 Test Method
- a. Filling and circulating the reactor coolart system with borated water should be accomplished, utilizing normal flow paths as much as possible.
- This test was performed at 50 percent power at Callaway.
Callaway has the same core and Nuclear Instrumentation System as Wolf Creek. Wolf Creek Core parameters measured prior to the pseudo rod drop test were compared with the corresponding results for Callaway to verify that the plant response was the same. This exemption was approved in a July 3, 1985 letter from the NRC. g 14.2-176 Rev. 0
1 () WOLF CREEK
- b. Collect and analyze four samples taken at equidistant depths in the reactor vessel simultaneously with one sample from the operating residual heat removal loop to check uniform boron concentration.
14.2.12.3.25.4 Acceptance criteria Boron concentration of the samples obtained from the designated sample points must be within a 30-ppm range of values. 14.2.12.3.26 Initial Core Loading (SU7-0001) 14.2.12.3.26.1 Objectives
- a. To load fuel in a controlled manner.
- b. To measure boron concentration.
14.2.12.3.26.2 Prerequisites
- a. Sufficient preoperational testing has been completed to ensure the necessary equipment and attendant instruman*
tation is functional. (~') b. Required technical specification surveillance is com-x- plated and the necessary systems are operable. 14.2.12.3.26.3 Test Method Instruction includes a core-loading sequence which specifies the loading-in a step-by-step fashion-with the appropriate data col-lection records. 14.2.12.3.26.4 Acceptance Criteria A permanent record of the final as-loaded core configuration has been made, and the configuration is consistent with the fuel assembly core loading plan. Boron concentration is as specified in the Technical Specifications. 14.2.12.3.27 Inverse Count Rate Ratio Monitoring For Core Leadin7 (S-070002) 14.2.12.3.27.1 Objectives
- a. To obtain nuclear monitoring data during initial core leading.
- b. To prevent criticality during core loading.
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WOLF CREEK O 14.2.12.3.27.2 Prerequisites
- a. Temporary and plant source range nuclear instrtmentation has been operational for a minimum of 4 hours ss achieve stable operstion,
- b. Plant is prepared for initial core loading.
14 2.12.3.27.3 Test Method Data from the nuclear monitoring channels is used to assess the safety with which core loading operations may be conducted. Inverse count rate ratio is plotted and evaluated to prevent any unexpected deviation from suberiticality. The core la monitored and maintained in a suberitical configuration throughout the core loading. 14.2.12.3.27.4 Acceptance Criteria The core is load.ed without achieving criticality. J*.;.12.3.28 Inverse Count Rate Ratio Monitoring for Approach to Initial Criticality (S-070003) 14.2.12.3.28.1 Objectives
- a. To obtain nuclear monitcring data during initial critic-ality,
- b. To anticipate and determine criticality.
14.2.12.3.28.2 Prerequisites
- a. Both source range and intermediate range nuclear channels alarm, trip functions, and indicating devices have been checked out and calibrated.
- b. Both source range and intermediate range nuclear channels have been energized a mininna of 4 hours to insure stable operation.
14.2.12.3.28.3 Test Method -
- a. Obtain base line count rates prior to red withdrawal and boron dilution. After each increment of rod withdrawal, and periodies11y during boren dilution, count rates are obtained, and inverse count rate ratio is evaluated,
- b. Core ' reactivity is monitored during the approach to criticality.
14.2-178 O Rev. O
/^T V WOLF CRIEK 14.2.12.3.28.4 Acceptance Criteria To determine criticality.
14.2.12.3.29 Initial Criticality (S-070004) 14.2.12.3.29.1 Objectivce To achieve initial criticality in a controlled manner. 14.2.12.3.29.2 Prerequisites
- a. Initial core loading is completed,
- b. Required technical specification surveillance is com-plated and the necessary systems operable.
- c. Sufficient post-core loading precritical testing has been completed to ensure the necessary equipment and attendant instrumentation is functional.
14.2.12.3.29.3 ' Test Method
- a. At preselected points during rod withdrawal and/or boren
- /~)
?s/
dilution, data is taken and inverse count rate plots made to enable extrapolation to be carried out to the expected critical point.
.b. Initial criticality la achieved by boron dilution or, if desired, by withdrawing control rods.
14.2.12.3.29.4 Acceptance Criteria The reactor is.gritical with the flux level established at approx-imately 1 x 10 amps on the intermediate range nuclear channels. 14.2.12.3.30 Determinktion of Ocee POWur Nahge for Physics Testing (S-070005) 14.2.12.3.30.1 Objectives To determine the reactor power level at which effects from fuel heating is detectable and to establish the range of neutron flux in which zero power reactivity measurements are to be performed. 14.2.12.3.30.2 Prerequisites
- a. The reactor is critical and stable in the intermediate range.
O sJ 14.2-179 Rev. O
WOLF CREEK
- b. Control rods are sufficiently doop in the core to allow positivo roactivity insortion by rod withdrawal. (g)
- c. Reactor coolant temperature is establish 6d at a value thet minimizes the moderator temperature coef.ficient reactivity feedback.
14.2.12.3.30.3 Test Method
- a. Withdraw control rod bank and allow the nontron flux level to increaso until nuclear heating effects are indicated by the reactivity computer.
- b. Record the reactivity computer picoammotor flux level and, if possible, the corresponding IR channel currents at which nuclear heating occurs, to obtain zero power testing range.
14,2.12.3.30.4 Acceptanco Critoria The power level at which zoro power testing is conducted is dotor-mined. 14.2.12.3.31 Boron Endpoint Datormination (S-070006) 14.2.12.3.31.1 Objectives To determine the critical reactor coolant system boron concentra-tion appropriate to an endpoint configuration (RCC configuration). a W 14.2.12.3.31.2 Prerequisi?,os
- a. The reactor is critical within the range for zero power testing and stable,
- b. The reactor coolant is at normal operating pressure and temperature.
- c. Rods are at the approximate and point configuration.
14.2.12.3.31.3 Test Method Boron endpoints are measured by deterwining the boron concentra-tion of the reactor coolant system with the rods close to or at the desired configuration. If not, the rods are then quickly moved to the desired configuration with no coron adjustment. The
- change in reactivity is measured, and this reactivity is converted I to an equal amount of boron to yield the endpoint at that particu-lar rod configuration.
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l L- ._
WOLF CREEK 14.2.12.3.31.4 Acceptance Criteria .O-v The results of the boron endpoint calculations meet the require-ments of the Nuclear Design Report within tolerance values specified in vendor design documents. 14.2.12.3.32 Isothermal Temperature Coefficient Measurement (S-070007) 14.2.12.3.32.1 Objectives To determine isothermal temperature coefficient, then derive the moderator temperature coefficient from the isothermal data. 14.2.12.3.32.2 Prerequisites
- a. The reactor is critical within the range for zero power testing and stable.
- b. The reactor coolant is at normal operating pressure and temperature,
- c. Control rods are at the approximate end point configura-tien.
14.2.12.3.32.3 Test Method The isothermal temperature coefficient is determined by O heating / cooling the reactor coolant system at a constant rate temperature versus reactivity. The moderator temperature and plottinfentmaybederivedfromisothermaldata, if desired. coeffic 14.2.12.3.32.4 Acceptance criteria The average of the measured values of the isothermal and, if desirod, the derived' moderator temperature coefficient agrees with acceptance criteria given in the Nuclear Design Report within tolerance values specified in vendor design documents. 14,2.12.3.33 Power Coefficient Determination (S-070008) 14.2.12.3.33.1 Objectives To verify the power coefficient of reactivity. 5 14.2.12.3.33.2 Prerequisites
- a. Reactor power level, reactor coolant tomperature and pressures, and RCCA and RCC bank configuration are as follows:
O 14.2-1?1 Rev. 1 l
. - _.-.. . . . _ . , . _ . _ . - - - _ . - _ . _ _ _ _ . _ - _ , . _ . ~ . . ~ . . . _ _ . _ . . -
J WOLF CREEK
- 1. RCS pressure - norinal 2235 psig ggg
- 2. RCCA, RCC bank configuration - nominally all rods out, D at bite position
- 3. Reactor power level - nominally 30, 50, 75, and 90 percent RTP
- 4. T - consistent with the nominal value corres-p8X81ng to the power levels, T,yg program at the identified nominal
- b. All subsystems which affect overall plant transient response should be in automatic mode of operation with the exception of the rod control system and automatic makeup. The CVCS domineralizer shall be bypassed.
14.2.12.3.33.3 Test Method
- a. As generator electrical load is changed, the primary side is permitted to freely respond without any control rod motion.
- b. The power coefficient verification factor is calculated by measuring the change in RCS temperature and the cor-responding change in core power.
14.2.12.3.33.4 Acceptance Criteria lll The average value of the measured verification factor agrees with that obtained from design predictions of the isothermal tempera-ture coefficient and doppler power coefficient. This agreement is within limits given in the test instructions. 14.2.12.3.34 Loaa Swing Tests (S-070009) 14.2.12.3.34.1 Objectives To verify proper nuclear plant transient response, including cutomatic control system performance, when load changes are intro-duced at the turbine generator. 14.2.12.3.34.2 Prerequisites Step load changes are initiated from steady state conditions at approximately 30 , 75 , and 100-percent power. O l 14.2-182 Rev. 1
WOLF CREEK
~) 14.2.12.3.34.3 Test Method V
- a. Manually reduce the turbine generator output as rapidly as possible to achieve an approximate 10-percent load decrease / increase,
- b. Plant variables are recorded, along with values observed on the normal clant instrumentation, during the load transient for those parameters required.
14.2.12.3.34.4 Acceptance Criteria The following acceptance criteria are to be used to determine successful test completion. Failure to meet these criteria does not constitute a need for stopping the test program, but corree-tion of any deficiences saould be accomplished, as required, consistent with the current plant schedule,
- a. Reactor and turbinc must not trip.
- b. Safety injection is not initiated.
- c. Neither steam generator relief valves ncr safety valves shall lift,
- d. Neither pressurizer relief valves nor safety valves shall lift.
- e. No manual intervention shall be required to
(]) conditions to steady state. bring plant
- f. Nuclear power overshoot (undershoot) must be less than 3 percent for load increase (decrease).
14.2.12.3.35 Large Load Reduction Test (S-070010) 14.2.12.3.35.1 Objectives To demonstrate satisfactory plant transient response to various specified load changes, to monitor the reactor control systems during these transients, and, if necessary, optimize the reactor control system setpoints. 14.2.12.3.35.2 Prerequisites Step load reduction changes of 50 percent are initiated frem steady state conditions at approximately 75- and 100-percent power. O 14.2-183 Rev. 1 l
WOLY CREEK l 14.2.12.3.35.3 Test Method
- a. Manually reduce the turbine generator output to achieve an approximate 50-percent load reduction.
- b. Monitor plant response during the transient and record plant variables, as required.
- c. If necessary, adjust the reactor control system setpoints until optimal response is obtained.
14.2.12.3.35.4 Acceptance Criteria The following acceptance criteria are to be useo to determino successful test completion. Failure to meet these criteria does not constitute a need for stopping the test program, but correc-tion of any deficiencies should be accompilched, as required, consistent with the current plant schedule.
- a. Reactor and turbine must not trip.
- b. Safety injection is not initiated.
- c. Steam generator safety valves shall not lift.
- d. Pressurizer safety valves shall not lift.
- e. No manual intervention shall be required to bring plant h conditions to steady state.
14.2.12.3.36 Plant Trip From 100 Percent Power (S-070011) 14.2.12.3.36.1 Objectives To verify the ability of the plant automatic control systems to sustain a trip from 100 percent and to bring the plant to stable conditions following the transient, to determine the overall response time of the hot leg resistance temperature detectors, and to evaluate the data resulting from the trip to determine if changes in the control systen setpoints are warranted to improve transient response based on actual plant operations. 14.2.12.3.36.2 Prerequisites
- a. The rod control system, steam generator level, pressur-1:or pressure and level, and the stoam dump system are in the automatic control mode,
- b. The plant is operating t normal steady state full power.
- c. Diesel generators in %:.,ndby idling condition.
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] 14.2-184 Rev. 1
WOLF CREEK 14.2.12.3.36.3 Test Method
-{ }
- a. Initiate a plant trip by opening the main generator output breaXer, monitor plant response, and record plant variables, as required.
- b. If necessary, adjust the control system setpoints to obtain optimal response.
14.2.12.3.36.4 Acceptance Criteria The system parameters must stay within the limitations specified in the vendor's design transient analysis document. 14.2.12.3.37 Rods Drop and Plant Trip (S-070012) 14.2.12.3.37.1 Objectives , To demonstrate that the negative rate trip circuit will trip the reactor and to monitor plant response. 14.2.12.3.37.2 ' . - trequisites
- a. The rod control system, steam generator level, pressur-izar pressure and level, and the feedwater pump speed control are in the automatic control mode. Steam dump control system is in the T,yg mode.
() b. The plant is operating at a steady state power of 30 to 50 percent.
- c. The rod group and the selected rods to be dropped have been identified.
14.2.12.3.37.3 Test Method
- a. Drop two RCCAs from a common group whic'., t because of their worth and location, are the most difficult to detect by the nuclear instrumentation system (NIS).
- b. Monitor systems behavior and plant response to trip from an intermediate power level prior to the plant trip test from full power.
14.2.12.3.37.4 Acceptance Criteria l The following acceptance criteria are to be used to determine successful t.est completions i l' 14.2-185 Rev. 1 l l
WOLF CREEK
- a. The reactor shall have tripped as a result of the nega-tive rata trip.
ggg
- b. All RCCAs shall release and bottom on receipt of a trip signal.
- c. The pressurizar safety valves shall not lift,
- d. Steam generator safety valves shall not lift,
- e. Safety injection is not initiated.
24.2.12.3.38 Shutdown and Maintenance of Hot Standby External to the Control Room (S-070014) 14.2.12.3.38.1 Objectives To demonstrate, ucing a plant procedtre, that the plant can be taken from $10 percent power to hot standby conditions, and verify that the plant can be maintained in hot standby for at least 30 minutes with a minimum shift crew, using controls and instrumenta-tion external to the control room. 14.2.12.3.33.2 Prerequisites
- a. Required component testing and instrument calibration are complete,
- b. Required electrical power supplies and control circuits lh are operational,
- c. The plant is at normal operating conditions at g10 percent power.
- d. The authority and responsibility of the control room observers has been established and is specified in this procedure.
- 14.2.12.3.38.3 Test Method
- a. The plant is taken from g10 percent power to hot standby conditions, using a plant procedure, minimum shift crew, and controls and instrumentation external to the control room.
- b. Hot standby conditions are maintained for at least 30 minutes.
O l 14.2-186 Rev. 1
WOLF CREEK l () c. All actions performed by the control room observers is documented within this procedure for use in evaluating i their impact on the test results. 14.2.12.3.38.4 Acceptance Criteria The plant can be taken from g 10 percent power- to hot standby conditions which are maintained for g30 minutes, using a plant procedure, minimum shift crew, and controls and instrumentation external to the control room. 14.2.12.3.39 Power Ascension Thermal Expansion and Dynamic Test (S-070015) 14.2.12.3.39.1 Objectives
- a. To demonstrate during specified power ascension transi-ents that the_ systems' monitored points respond in accor-dance with design,
- b. To demonstrate during the heatup to full power tempera-ture that the systems' piping can expand without obstruc-tion and-that the expansion is in accordance with de- '
sign. Also, during the subsequent cooldown to ambient temperature, the piping returns to its cold position in () accordance with system design. 14.2.12.3.39.2 Prerequisites
- a. Reference points for measurement of the systems are established.
- b. Power ascension testing is in progress,
- c. All subject systems are available for the specified dynamic operations,
- d. Required instrument calibration is complete,
- e. A preservice inspection of the associated piping snubbers has boon completed within 6 months.
14.2.12.3.39.3 Test Method
- a. Record cold baseline data,
- b. Obtain measurement data at various specified temperature plateaus.
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WOLF CREEK -
- c. The systems are aligned for the specified dynamic opera-tien.
- d. The specified dynamic event of pump operation, valve operation, etc., is initiated, and the system is moni-tored for response.
- e. On completion of cooldown to ambient temperature, obtain measurement data.
14.2.12.3.39.4 Acceptance Criteria
- a. There shall be no evidence of blocking of the thermal expansion of nny piping or components, other than by design,
- b. The total stresses shall not exceed applicable code limits.
- c. Spring hanger movement must remain within the hot and cold set points, snubber swing clearance remains satis-factory., and snubbers must not become fully retracted er expanded.
- d. Piping and components must return to their baseline position on cooldown in accordance with system design. ,
- e. The measured thermal movement shall be within 25 percent of the analytical value or 10.25 inch, whichever is greater.
14.2.12.3.40 Biological Shield Testing (S-070016) 14.2.12.3.40.1 Objectives
- a. To measure and record the neutron and gamma ray radiation levels in accessible areas of the plant where radiation levels above background are anticipated,
- b. To determine locations if any, where shielding is defic-ient.
- c. To ensure that plant personnel are not subjected to overexposure from radiation as a result of inadequate shielding.
14.2.12.3.40.2 Prerequisites
- a. Required instrument calibration is complete.
14.2-188 Rev. 1 9
WOLF CREEK () b. Appropriate reactor power levels are attained.
- 14.2.12.3.40.3 Test Method Neutron and gamma ray surveys are conducted in each of the follow-ing reactor power-level ranges.
Iegi % Reactor Power Rance Preoperational Shield Tests g0 Low Power Tests 0-5 Intermediate Power Tests 5-50 High Power Tests- 50-100 14.2.12.3.40.4 Acceptance Criteria Neutron and gamma ray radiation surveys in all accessible areas of the plant where radiation levels above background tre anticipated reveal no shielding deficiencies; or identify and implement appro-priate administrative controls in accordance with 10 CFR 20 for the areas determined to be radiation arer*. 14.2.12.3.41 Loss of Heater Drain Pump Test (S-070017)*
- 14.2.12.3.41.1 Objectives
() To verify proper pump accident. nuclear plant responce to a loac of heater drain 14.2.12.3.41.2 Prerequisites The plant is operating at steady state conditions at 90-porcent power. 14.2.12.3.41.3 Test Method The heater drain pumps are tripped and plant variables are re-corded, along with values observed on the normal plant instrumen-tation, during the transient for those parameters required. i 1 L This test was performed at Callaway only, with the consent of l-l the NRC, as Callaway and Wolf Creek have identical Heater Drain-Systems. I L i O 14.2-189 Rev. 1 l
*b-- -
WOLF CREEK 14.2.12.3.41.4 Acceptance Criteria ggg The following acceptance criteria are to be used to determine successful test completion. Failure to meet these criteria does not constitute a need for stopping the test program, but correc-tien of any deficiencies should be accomplished as required, consistent with the current plant schedule,
- a. Reactor and turbine must not trip.
- b. Safety injection is not initiated.
- c. Neither steam generator relief valves nor safety valves shall lift.
- d. Neither pressurizer relief valves nor safety valves shall lift.
- o. No manual intervention shall be required to bring plant conditions to steady state.
14.2.12.3.42 Calibration of Steam and Feedwater Flow Instrumenta-tion at Power Test (S-070018) 14.2.12.3.42.1 Objectives
- a. To calibrate the steam flow transmitters against feed-water flow, gg
- b. To perform a cross-check verification of all signals indicating feedwater and steam flow.
14.2.12.3.42.2 Prerequisites
- a. Test equipment, including transmitters, has been cali-brated for expected ranges of plant conditions,
- b. The plant shall be at steady state conditions for each power level at which testing is performed.
14.2.12.3.42.3 Test Method At 30 and 50 percent power, perform Step a if the steam flow / feedwater flow mismatch alarm actuates. At 75 and 100 percent power, perform Steps a and b.
- a. Verify calibration of the steam flow by comparing steam flow signal to referenced feedwater flow, d
O l 14.2-190 Rev. 1
WOLF CREEK
- b. Compare, using plots, the steam and feedwater flow values
- O- to determine if recalibration is necessary prior to the next power escalation.
14.2.12.3.42.4 Acceptance Criteria
- a. Steam flow /feedwater flow mismatch alarm does not actuate at 30, 50, 75, and 100 percent power,
- b. Steam flow indication should be within 14 percent of feedwater flow panel indicator at 75 and 100 percent power.
- c. The test feedwater flow instrument versus plant feed-water flow instrument and plant steam flow instrument curves should be within 12.5 percent and 13.0 percent of their respective ideal curves at 75 and 100 percent power. ,
14.2.12.3.43 Natural Circulation Test (S-090024)* 14.2.12.3.43.1 Objectives To demonstrate- the length of tire required to stabilize natural circulation; to demonstrate core flow distribution during natural circulation using incore thermocouples. 14.2.12.3.43.2 prerequisites (}
- a. Required low power physics testing has been completed.
- b. Required instrumentation in installed and calibration complete,
- c. .The plant is operating at steady state conditions at 3 percent power.
14.2.12.3.43.3 Test Method All reactor coolant pumps are simultaneously tripped while at 3 percent rated power. The transients are monitored and establish-ment of natural circulation verified. l Due to similar plant design for Callaway and Wolf Creek, the NRC allowed WCGS to use Callaway Natural Circulation test data and pertinent results. l O 14.2-191 Rev. 1 l l l
..-, ,, - _ . . , , . . . . . ,, . , ,, ..n~ - - . . , . . , . - , , , . . . . . . . . . , - , , , . . .
, WOLF CREEK 14.2.12.3.43.4 Acceptance Criteria Natural circulation has been demonstrated. The measured core aT e
as a function of core power under natural circulation condition: is no greater than the limiting reactor coolant system AT based on dcaf.gn requirements. O O } 14.2-192 Rev. 1
WOIJ CREEK 14.2.12.3.43.4 Acceptance Criteria Natural circulation has been demonstrated. The measured core oT as a function of core power under natural circulation conditions is no greater than the limiting reactor coolant system AT based on
~dosign requirements.
O O : 14.2-193 Rev. 0
WLF CREEK (~N LJ TABLE 14.2-1 SATLTl-RI2ATED raruATICt&L TEST FIOCE:L"AES Test Abstract Test h bar Title USAR Section 5-03AB01 Steam Dap System Preoperaticnal Test 14.2.12.1.1 SU3-AB02 Main Steam Safety Valve Test 14.2.12.1.2 S-03AB03 Main Staam Line Isolation Valve Test 14.2.12.1.3 S-03AB04 Main Steam System Prscpuational Test 14.2.12.1.4 S-03AE01 Main Feedwater System Precperational Test 14.2.12.1.5 S-03AE02 Steam Generator Laval Control Test 14.2.12.1.6 S-03ALO1 N'v414a4 Feedwater Mctor-Driven Ptr:p and Valve Preoperational Test 14.2.12.1.7 SU3-AI42 P'v414a7 Feedwater Turbine-Driven Pung ard Valva Preoperational Test 14.2.12.1.8 SG -AI43 Aw414a7 Feedwatar Metcr-Driven Ptr:p Erdurance Test 14.2.12.1.9 S-03ALO4 Atociliary Feedwater System Watar Hananar Test - 14.2.12.1.10 SU3-ALCS Al'xiliazy Feedwater Turbine-Drivan Pump ( Endurance Test 14.2.12.1.11 ( S-03BB01 Reactor Coolant Pump Initial Cperation 14.2.12.1.12 SU3-BB02 PRT Cold Prooperational Test 14.2.12.1.13 SU3-BB03 RID Bypass Flow Measurement 14.2.12.1.14 S-03BB04 Prussurizar Pressure Control Test 14.2.12.1.15 S-03BB05 Reactor Ceolant System Hot Preeparational Test 14.2.12.1.16 S-03BB06 Thermal Expansion 14.2.12.1.17 S-03BB07 Pressurizer Level Centrol Test 14.2.12.1.18 SU3-BB08 Pressuriter Heater ard Spray Capability Test 14.2.12.1.19 SU3-BB09 Reactor Coclant System Flow Measurament Test 14.2.12.1.20 SU3-BB10 Reacter Coolant System Flow coastdcun Test 14.2.12.1.21 S-03BB11 Raactor Coolant Systen Hydrostatic Test 14.2.12.1.22 SU3-BB12 Pressurizar Contirrxus Spray Flow Varification Test 14.2.12.1.23 5-03BB13 Pressurizar Ralief Valve ard PRT Hot Preoperational Test 14.2.12.1.24 S-03BS14 Reactor Coolant Io:p Vibration Su:veillance Test 14.2.12.1.25 SU3-BB15A Leak Detection System Precperational Test 14.2.12.1.26 SU3-BB15B Leak Detection System Prec;xtrational Test- 14.2.12.1.27 S-03BB16 RID /TC Cross Calibration 14.2.12.1.28 ( Rev. 0
i WOLF CREEK h TABLE 14.2-1 (Sheer. 2)
- Test Numkgg Title 222R Section S-03BG01 Chemical and Volume Control System Major Component Test 14.2.12.1.29 SU3-BG02 Seal Injection Preoperational Tert 14.2.12.1.30 SU3-BG03 Charging System Preoperational Test 14.2.12.1.31 SU3-BG04 Boron Thermal Regeneration System Prooperational Test 14.2.12.1.32 SU3-BG05 Boric Acid Blending System Preopera-tional Test 14.2.12.1.33 S-03BG06 Chemical and Volume Control System Hot Prooperational Test 14.2.12.1.34 SU3-EC01 Fuel Pool Cooling and Cleanup System Preoperational Test 14.2.12.1.35 S-03EC02 Spent Fuel Pool Leak Test 14.2.12.1.36 SU3-EF01 Essential Service Water System Pre-SU3-EF02
. operational Test 14.2.12.1.37
- Essential Service Water Pump Preopera-tional Test 14.2.12.1.37 S-03EG01 Component Cooling Water System Pre-operational Test 14.2.12.1.38 SU3-EJ01 Residual Heat Removal System Cold Pre-operational Test 14.2.12.1.39 SU3-EJ02 Residual' Heat Removal System Hot Preoperational Test 14.2.12.1.40 SU3-EM01 Safety Injection System Cold Pre-operatienal Test 14.2.12.1.41 l SU3-EM02 Safety Injection Flow Verification Test 14.2.12.1.42 SU3-EM03 Safety Injection Check Valve Test 14.2.12.1.43 SU3-EM04 Boron Injection Tank and Recirculation Pump Test 14.2.12.1.44 S-03EN01 Containment Spray System Nozzle Air Test 14.2.12.1.45 SU3-EN02 Containment Spray System Preoperational Test 14.2.12.1.46 S-03EP01 . Accumulator Testing 14.2.12.1.47 SU3-FC01 Auxiliary Feedwater Pump Turbine Preoperational Test 14.2.12.1.48 SU3-GD01 Essential Service Water Pumphouse HVAC-Preoperational Test 14.2.12.1.49 SU3-GF01 Miscellaneous Building HVAC System SU3-GF02 Preoperational Tests SU3-GF03 14.2.12.1.50 S-03GG01 Fuel Building HVAC System Preoperational Test 14.2.12.1.51 SU3-GK01 Control Building HVAC System Proopera-tional Test 14.2.12.1.52 SU3-GLO1 Auxiliary Building HVAC System Pre-operational Test 14.2.12.1.53 S-03GM01 Diesel Generator Building HVAC Pre-operational Test 14.2.12.1.54 \g v
l Rev. 3
g ICIF CREEK TAEE 14.2-1 (Sheet 3) Test !?a:tber Ti t1_e_ USAR 6ection SU3-G101 Contairment Cooling Sptam Prooperatioral Test 14.2.12.1.55 S-03GIO2 CRI24 Cooling Proopuntional hat 14.2.12.1.56 SU3-CP01 Integrated Containmint Isak Rate hst 14.2.12.1.57 SU3-GP02 Reactor Contairment Structaral Intag-rity Acceptan::e Test 14.2.12.1.58 S-03GS01 Post-Accident Hydrxxgen Removal System Preoperational hat 14.2.12.1.59 S-03GT01 Containment Nrge System INAC Pre-operational Test 14.2.12.1.60 S-03HA01 Gaseous Radwasta System Preoperational Test 14.2.12.1.61 S-03JE01 E:nerg.rcy Fual oil System Prooperatiend Test 14.2.12.1.62 SU3-E01 Spent Fual Pool Crane Prooperational hst 14.2.12.1.63 SU3-E02 liew Fual Elevator Prooperational hst 14.2.12.1.64 SU3-E03 Nal liandling ard Storage Prooperational Test 14.2.12.1.65 - SU2-E04 Fuel Transfer System Preoperational Test 14.2.12.1.66 SU3-IT05 Refueling 14achina and RCX: Change Fixture Preoperational Test 14.2.12.1.67 S-03E06 Rafaeling linchine Irdaxing hst 14.2.12.1.68 SU3-E07 Nel Handlig Systau In+cagrated Preoperational Test 14.2.12.1.69 S-031U01 Ljesel Generator Metanical Procpera-
*ional hst 14.2.12.1.70 S-03NB01 4160-V (Class IE) Systen Preoperatio.ul Msv 14.2.12.1.71 S-03NE01 Diesel Generator Electrical Preopera-tiend i%st 14.2.12.1.72 SU3-NF01 Integrated hel Irgic Test 14.2.12.1.73 S-03NF02 IDCA Segaencer Preoperr tional hst 14.2.12.1.74 S-03NF03 Shutdown Sa m 'r Preoperational Test 14.2.12.1.75 S-03NG01 480-V (Class IE) b/' stem Prec p ational Test 14.2.12.1.76' SU3-!CO2 460-V Class IE System (ESW) Prec%ra-tional Test 14.2.12,1.77 S-03NK01 125-V (Class IE) DC Systa: Preopera-
'tional Test 14.2.12.1.78 S-03NN01 Inst:mant AC System (Class 13) Pre-operational hst 14.2.12.1.79 SU3-SA01 Ergineered Safoguards (NSSS) Procpera-tional Test 14.2.12.1.80 0
Rev. 0
ICII CREEK TAE!.E 14.2-1 (Sheet 4) h 1 Test Ihr*1xu' Title t.'SAR Section SL*3-SA02 Dyincerod Safcqmrds (BCP) Pre-coaraticnal Test 14.2.12.1.81 l SL'3-SA03 ?hgireared Safoguards Verification Test 14.2.12.1.82 1 S-03SB01 P.uctor Prctection System Iagic Test 14.2.12.1.83 S-03S701 Primary Sa::pling System Procperational hst 14.2.12.1.84 S-03SP01 Process Radiaticn Manitoring S}ttem Procperaticnal Test 14.3.12.1.85 SL'3-0004 Pcver cc:nversic:n and EC Syste.~s hermal Expansion Test 14.2.12.1.86 S-030005 Perser Cer: version and ECCS Systars Dynarde hst 14.2.12.1.87 SU3-0006 HEPA Filtar Test 14.2.12.1.88 S-030008 Cooldcwn from Hot Standby DC.cInal to the Control Poc:m 14.2.12.1.89 S-030009 Cc:: pre"rA Gas Ammiator Testirs 14.2.12.1.90 O O Rev. O
O == TABIE 14.2-2 NCt4SATETV-REIKIED PRIDPEPATICIRL TESTS Test Ntrnbar T_itle USAR Section I S-04ACO2 Turbine Trip Test 14.2.12.2.1 S-04ACO3 Turbine Systen Cold Test 14.2.12.2.2 S-04AD01 Condensata Systen PIsoperational 'Nst 14.2.12.2.3 S-04AF01 Saoandary Vent and Drain System Pre-operational Test . 14.2.12.2.4 S-04AQ01 Condansata and Feedwater h4M Feed System Pzwoperational Test 14.2.12.2.5 S-04BI41 Ranctor Makeup Water Systan Pzwopkra-tional Test 14.2.12.2.6 S-04CG01 Condanner Air Rameval Systen Pre-operational Test 14.2. 12.2.7 SU4-EA01 Circulating Water Systen Proeperational
'Nst 14.2.12.2.8 S-04EA01 Service Water System Preoperational Test 14.2.12.2.9 S-04EB01 Ciceed Cooling Water System Proopera-tional 'Nst 14.2.12.2.10 SU4-FP03 Fire Protection System Prooperational Test 14.2.12.2.11 S44GiO1 Radwasta M41d1% HBC System Pre-cparatimal 'Nst 14.2.12.2.12 SUS-GP01 local Ccritainment laak Rata Test 14.2.12.2.13 S-04HB01 Liquid Radwasta System Preoparational Test 14.2.12.2.14 SU4-HB02 Wasta Evaporator Prooperational Test 14.2.12.2.15 S-04HC01 Solid Wasta Systen Prooperaticmal Test 14.2.12.2.16 S-04HC02 Solid Wasta Filter Handling Systen Prooperational 'Nst 14.2.12.2.17 SU4-HCO3 Rasin Transfer Prooperational Test 14.2.12.2.18 SU4-KC01A Fire Protection Systan (Water). Pre- '
SU4-EC01B operational 'Not 14.2.12.2.19 S-04KCO2 Fire Protection Systen (Halan) Pre-operational 'Nst 14.2.12.2.20 S-04EC03 Fire Protection System Detection and Alarm Preoperational 'Nst 14.2.12.2.21 S-041E01 Oily Wasta System P g ational Test 14.2.12.2.22 SU4-LF01 Floor arri Equipnant D ain System Pre-operational Test . 14.2.12.2.23 S-04PA01 13.8-kV System Prooperational Test 14.2.12.2.24 S-04PB01 4,160-V (Non-Class IE) Systen Pre-operational Test 14.2.12.2.25 S-04PG01 480-Volt (Non-Class IE) Systen Pre-cperational Test 14.2.12.2.26 O Rev. 0
MLT CSECK R ELE 14.2-2 (Sheet 2) O Test !? rte.r Title USAR Sertien S-04R701 250-V DC Syste:n Precperaticr.a1 Test 14.2.12.2.27 S-04FK01 125-V (Nan-class II) DC Syra:n Pre-S-04PK02 cperaticnal Test 14.2.12.2.28 S-04PN01 InstrLN AC (Non-Class IE) Syste:n Pre-operational Test 14.2.12.2.79 S-04QD01 Emergetry LightinJ Systc:n Prtx:pera - tictvtl Test 14.2.12.2.30 5-04CT01 Public Address Syste:n Precperational Test 14.2.12.2.31 S-04CC01 Heat Tracirq Prueze Protection Sys+w:n Preoperational Test 14.2.12.2.32 S-04FM01 Seccrdary SamplirrJ Syste:n Procperational Test 14.2.12.2.33 S-04SD01 Area Paiiation Monitcrirq Preptional Test 14.2.12.2.34 S-045G01 Seimic Manitoring Instru:non*ation Syste:n Proc;xtraticnal Test 14.2.12.2.35 SU4-SCO2 Iccco Parts Mcnitcrirq Systc:n Test 14.2.12.2.36 SU8-0007 Plant Perfor: nance Test 14.2.12.2.37 S-090023 Electrical Distributic.) Syste:n Voltage Verification Test 14.2.12.2.38 O l l l PIrv. O O e
p J 1017 CREEK
'IABlZ 14.2-3 Di1TIAL STARIUP TCt Test thunber Title USAit Section S-07AB01 Autantic Steam Generator Imval Cental 14.2.12.3.1 l SU7-AD02 Dynamic Autmatic Steam Dump Control 14.2.12.3.2 S-07BB01 RrD Bypass F1w Haasww rit 14.2.12.3.3 S-07BD02 Pressuriter Heater ard Spray capability Test 14.2.12.3.4 S-07BB03 Reactor Coolant System Flow }haswmr_ii; 14.2.12.3.5 SU7-BB04 Reactor Coolant Systaa Flow Coastdcwn Test 14.2. 12.3.6 S-07BB05 Pasmtrizer contiuous Spray Flow Verification 14.2.12.3.7 S-07BB06 RID /IC Cross Calibration 14.2.12.3.8 S-075C01 Core frMhg Instrumantation and Neutron Scurre Requiruments 14.2.12.3.9 S-07SC03 'Ihermal Tower Measurement ard Statepoint Data Collection 14.2.12.3.10 577-SE01 liuclear Instrunantation Systen Test 14.2.12.3.11 S-07sE02 operational Alignment of liuclear Instru-(] S-075E03 mentation Axial Flux Difference Instrumentation 14.2.12.3.12 Calibration 14.2.12.3.13 S-07SF01 Control Rod Driva }%chanism Operational Test 14.2.12.3.14 S-075F02 Rod Contzel System 14.2.12.3.15 SU7-SF03 Red Drop Time Measumingit 14.2.12.3.16 SU7-SF04 Rod Position Iniication System 14.2.12.3.17 S-07SF05 Automatic Reactor Control System 14.2.12.3.18 S-075R02/ Incore Flux Mappiry S-075R02 14.2.12.3.19 S-075R03/ In c re Ins
- N tion Test S-075R04 14.2.12.3.20 S-07SF06 Operational Alignment of ITocess Tenparaturu Instrumantation 14.2.12.3.21 S-07SF07 Startup Adjustments of Reactor control Syntan 14.2.12.3.22 S-075F08 RCCA or Bank Worth Measurement at Zero Power 14.2.12.3.23 SU7-SF09 RCCA or Bank Worth Measumierit at Power 14.2.12.3.24 S-07S701 Reactor Systems Sampling for Core load 14.2.12.3.25 SU7-0001 Initial Core leadirg 14.2.12.3.26 5-070002 Inverse Count Rate Ratio Moni*arirg for Core leading 14.2.12.3.27 Rev. O
WU' CFIEK TAEII 14.2-3 (Sheet 2) h Test theter Title U"JJ1 Section S-070003 Inverse Count Rata Ratio Monitcrirq for Approach to initial Criticality 14.2.12.3.28 S-070004 Initial Criticality 14.2.12.3.29 S-070005 Datarmination of Core Power Pargo for Hrfnics Testing 14.2.12.3.30 S-070006 Bcran Dx! point Determination 14.2.12.3.31 S-070007 Isothermal Temperature Coefficient Measurement . 14.2.12.3.32 5-070008 Power Coefficient Determination 14.2.12.3.33 S-070009 Icad Swing Tests 14.2.12.3.34 S-070010 large load Patrtion Test 14.2.12.3.35 S-070011 Plant Trip frten 100 Percent Rm.r 14.2.12.3.36 S-070012 Rods Drop ani Plant Trip 14.2.12.3.37 S-070014 Sim'n and Maintenance of Hot StarxD:rf D:ternal to the Centrol Pocm 14.2.12.3.38 S-070015 Power Ascensicn 'Ihermal Dqansien ard Dinamic Test 14.2.12.3.39 S-070016 Biolcgical Shield Testirq 14.2.12.3.40 S-070017 loss of Heater Crain Pep Test 14.2.12.3.41 S-070018 Calibration of Steam and Feedwater Flcv Inst:t'.~.entation at Power Test 14.2.12.3.42 5-090024 Natural Ciln11ation Test 14.2. 12.3.43 O O Pav. 0 , 1 1
~s--- as
..~,~na-n.
~-- ~~a.-+----we-,+m-su-as.~a-.w- - - - -
=a----u. - -a m m.r anax pr r
- 6
)
I, t f .,.-$ WOLF CREEK j 1 F
+
[ h v v i s h f I p I k
+
FIGURE 14.2-1 , (Deleted) g , c k l l
.l l
Rev. 0 l l
-:,---~- -------------~~-~~~~~~~~~r~ ~ ~ " ~ ~
n eosa strr v. i. r INTEROFFICE CORRESPONDENCE
'IO: Merlin Williams FRQ48 William B. Norton DATE: Novenber 22, 1985 SUIkmCT Startup Report moLKKO 85-779 Please fini attached four (4) copies of the Wolf Creek Generating Station Startup Report to be transmitted to Licensing. W.is report fulfills the requirements of Technical Specification 6.9.1.3, ADM til-039, and ADM 01-070.
If you have any questions please contact me at X2275. William B. Norton Reactor Engineering Supervisor WBN/bss ces G. Koester, w/a
'[] J. Bailey, w/a
'" F. Rhodes, w/a' G. Boyer, w/a RMS I
(,., e e
1 l KANSAS CAS h ELICTRIC COMPANY (^)3 (, WOIE CREEK GENERATING STATION STARTUP REPORI' Docket No. STN 50-482 Licenso No. NPF-42 Prepared By W.B. Norton Reactor Engineering Supervisor [a^'s L.D. Arnold E.E. Lehmann Power Ascension Engineer Reactor Engineer W.G. Ryder J.D. Stamn Reactor Engineer Ibwer Ancension Engineer R.L. Sirrdi W.C. Brandt Power Ascension Engineer Engineering Technician D.S. Stephens Typist rs (_)
fABf2 OF CONTENTS PAC tD. O Table of Con'.ents i List of Tab es iv List of Figures vi Introduction 1 1.0 Ini*.ial Core Loadityy 1.0-1 2.0 f:ost Core Loading Procritical Testing 2.0-1 2.1 control Rod Testing 2.0-3 2.1.1 Cold and Hot Ro3 Control System Testing 2.0-5 2.1.2 Roi Control System Test 2.0-10 2.2 Incore Hovable Detector Systan 2.0-11 2.3 Pressurizer Continuous Spray F1w Setting 2.0-14 ard Pressurizer Heater and Spray Capability Tests
. 2.4 Reactor Coolant Syston Flow Measuronent 2.0-18 2.5 Reactor Coolant Systna Flow Coastdown '!bst 2.0-20 2.6 RTD Dypass Flow Measuranent 2.0-24 2.7 Preliminary Data Collection for Instnment 2.0-26 Calibration 2.8 Nuclear Instrurentation Systan 2.0-27
- 2.9 frID/TC Cross Calibration Tests 2.0-28 2.10 '1hermocouple Core Subcooling m nitor 2.0-32 Syntan Test 2.11 Special Test Procedure For the Pressurizer 2.0-33 Relief Valves 2.12 Imse Parts Monitoring Systan 2.0-35 3.0 Initial Criticality And Low Power Physics 3.0-1 Test Sequence 3.1 Initial LYiticality 3.0-3 3.2 Control Rod Bank Wrth thasuratents 3.0-13 O 3.3 inets - 1 = m om Ce m ciene 3.0 2s 1
l TABLE OF CONTENTS
"^"" " '
O 3.4 Boron Erdpoint ard Doron Worth Moanutment 3.0-28 3.5 Power Distribution h3anururent 3.0-31 4.0 Power Ascension Testing 4.0-1 4.1 At Power thynien Tnatirvj 4.0-2 4.1.1 Incore Movablo Detector Arti Thermocouple 4.0-3 Mapping at Power 4.1.2 Axial Flux Difference Instrumentation 4.0-5 Calibration 4.1.3 Power Coefficient Determination 4.0-11 4.1.4 Pseudo Red Ejection Test 4.0-13 4.2 Control Synton Dynamic Desponso 4.0-16 4.2.1 Dynamic Autonatic Steam Dump control 4.0-17 4.2.2 Automatic Reactor Control 4.0-19 4.2.3 Automatic Steam Generator Level Contre,1 4 . 0 -2 '! 4.3 Transient and Trip Teats 4.0-22 4.3.1 Load Swirv) Tests 4.0-23 4.3.2 Large Load Pc3uction Tests 4.0-31 4.3.3 Shutdown And Maintenance' of Hot Standby External 4.0-34 To The Control Roam 4.3.4 Rods Drop And Plant Trip 4.0-35 4.3.5 Plant Trip From 100 Percent Power 4.0-37 4.4 Instrumentation Calibration arri Aligment 4.0-40 4.4.1 Thermal Power Meanutement arri Statepoint 4.0-41 Data Collection 4,4.2 Calibration Of 5 team and Fc-adwater Flow 4.0-43 Instrumentation 4.4.3 Operational Alignment of Nuclear Instrumentation 4.0-44 4.4.4 Operational Alignment of Process Temperature
'0 Instrumentation 4.0-53 11
TABII 0F Cot #E!JTS l PAGE to. 4.4.5 Startup Adjustments of The Reactor Control 4.0-56 i System 4.5 Steam Generator Ptilsture Carryover itwourunent 4.0-58 4.6 NSSS Acceptarne '!bst 4.0-60 4.7 IVmr Anconsinn ihnrmal Afrj Pftr.ie "tst 4.0-62 4.8 131ological Shield Testing 4.0-64 4.9 Plant Ibefornutn> Test 4.0-65 4.10 Turbino Cenerator 1bsta 4.0-66 4.11 Special 'Ibsts 4.0-68 4.11.1 Moisturo separator Reheater Test 4.0-69 4.11.2 Peactor Vosnel Level Irr31 cation System (RVLIS) 4.0-70 Appendix As Chronology of The Post Puol Load Startup Program A-1 Apper>3ix B: Power Ascension Testing Synopsia 3-1 Appendix C Unplannoi Reactor Trips During Post Fuel Load C-1 Test Program l l O k) lii
LIST OF TABLES O TABLE TITLE PAGE 10. 2.1.1-1 Rod Drop Time Sumnary 2.0-7 2.4-1 -ICS Loop Flow Determination Prior To 2.0-19 Initial Criticality 2,5-1 Plow Coastdown Rate Calculations 2.0-21 2.5-2 Iow-Flow Reactor Trip Time Delay 2.0-23 Calculations i 2.6-1 RTD Bypass Flow Measurement Results 2.0-25 2.9-1 Results of Initial RTD/'It Cross 2.0-29 Calibration Test 2.9-2 Results of Second RTD/'IC Cross 2.0-31 Calibration Test 2.11-1 Results of PORV Opening / Closing 2.0-34 Test 3.2-1 Control Rod Bank Worth Sumnary 3.0-14 h 3.3-1 Isothermal Tenperature Coefficient Results Sumnary 3.0-26 3.4-1 Borcn Endpoint Sumnsry 3.0-29 3.1-2 Differential Boron Worth Sumnary 3.0-30 3.5-1 Power Distribution Sumnary 3.0-32 4.1.1-1 Incc:.e Flux Map Sumnary During Power 4.0-4 Ascension 4.1.2-i Incore/Excore Correction Factor 4.0-6 4.1.2-2 100% NIS Current Values 4.0-7 4.1.2-3 Gain Valu;s for Delta I Function 4.0-9 Generator 4.1.2-4 Delta q Values At Spacified Power 4.0-10 Plateaus 4.1.3-1 Doppler Coefficient Verification Factors 4.0-12 0 1.,
LIST OF TABLES O TAtu.E TITLE PAGE NO. 4.1.4-1 D-12 Rcd Worth From HFP RIL 4.0-14 4.1.4-2 t' lux Map Results Frm Rod D-12 Ejection 4.0-15 4.3.1-1 Load Swing From 30% to 20% Power 4.0-24
' 3.1-2 Load Swing Frem 20% to 30% Power 4.0-25 4.3.1-3 Load Swing Frcm 75% to 65% Power 4.0-26 4.3.1-4 Load Swing From 65% to 75% Power 4.0-27 4.3.1-5 Load Swing From 100% to 90% Power 4.0-28 4.3.1-6 Load Swing From 90% to 100% Power 4.0-29
- 4. 'J . 2-1 Large Load Reduction Test From 75% 4.0-32 Power 4.3.2-2 Large Load Reduction Test From 100% 4.0-33 Power 4.3.4-1 Rods Drop Arrl Pl?nt Trip Test Data 4.0-36 O Sumury 4.3.5-1 Plant Trip From 100 Percent Power 4.0-39 J Test Sumary 4.4.1-1 BCS Flow From Calorimetric Measurement 4.0-42 4.4.3-1 Nuclear Instrumentation Overlap Data - 4.0-45 Source Range and Intermediate Range 4.4.3-2 Nuclear Instrumentation Overlap Data - 4.0-46 Intermediate Range and Power Range 4.4.4-1 Tenperature Alignment Data at 100% Power 4.0-55 4.5-1 Steam Generator Moisture Carryover Test 4.0-59 Results O V
LIST OF FICURES O FIGURE PACE No. TITLE, 1.0-1 Core Loading Sequence - Legend 1.0-4 1.0-2 Core Loading Sequence Steps 30 to 7 g 1.0-5 1.0-3 Core Loading Sequence StepsC7 to 7 D 1.0-6 1.0-4 Core Loading Sequence Steps 6 to 34 g 1.0-7 1.0-5 Core Loading Sequence Steps 35 to SSC 1* ~ 1.0-6 Core Loading Sequence Steps 55 to $6 3 1.0-9 0 1.0-7 Core Loading Sequence Steps 57 to 86 g 1.0-10 1.0-8 Core toading Sequence Steps 87 to 118 3 1.0-11 1.0-9 Core Loading Sequence Steps 119 to 158 g 1.0-12 1.0-10 Core Loading Sequence Steps 159 to 193 1.0-13 1.0-11 ICRR Plot For Core Loading Source Range N31 1.0-14 N 1.0-12 ICRR Plot For Core Icading Source Range N32 1.0-15 1.0-13 ICRR Plot For Core Loading Tenporary 1.0-16 Chan.el A 1.0-14 ICRR Plot For Core Loading Temporary 1.0-17 Channel B 1.0-15 ICRR Plot For Core Loading Temporary 1.0-13 Channel C 1.0-16 Wolf Creek Generating Station 1.0-19 Cycle 1 Final Core Loadirg Map 2.1-1 Control Rod Locations 2.0-4 2.2.1 Movable Detector Locations 2.0-12 2.3-1 Nominal Pressure Response to Opening 1.0-16 of Both Pressurizer Spray Valves 2.3-2 Pressure Response to Actuation of 2.0-17 All Pressurizer Heaters
?.1-1 ICRR During Rod Bank Withdrawal Channel 3.0-4 O
vi
LIST OF FIGURES O FIGURE TITLE PACE No. 3.1-2 1CRR Durin3 Rod Bank Withdrawal Channel 3.0-5 N32 3.1-3 ICRR vs. ICS noron Concentration 3.0-6 Channel N31 3.1-4 ICRR vs. ICS noron Concentration 3.0-7 Channel N32 3.1-5 ICRR vs. Time of ICS Dilution Channel 3.0-8 N31 3.1-6 ICRR vs. Time of TCS Dilution Channel 3.0-9 N32 3.1-7 ICRR vs. Reactor nakeup Water Addition 3.0-10 Channel N31 3.1-8 ICRR vs. Peactor Makeup Water Addition 3.0-11 Channel N32 3.2-1 Differential and Integral Bank Worth 3.0-15 Plot (CBD) 3.2-2 Dif ferential and Integral Bank Worth 3.0-16 Plot (CIC) 3.2-3 Dif ferential and Integral Bank Worth 3.0-17 Plot (CBB) 3.2-4 Differential and Integral nank Worth 3.0-18 Plot (Cin) 3.2-5 Differential and Integral Bank NortP 3.0-19 Plot (SDE) 3.2-6 Dif forential and Integral Bank Worth 3.0-20 Plot (SDD) 3.2-7 Dif ferential and Integral Bank Worth 3.0-21 Plot (SDC) 3.2-8 Differential and Integral Bank Worth 3.0-22 Plot (Ejected Rod D-12) 3.2-9 Differential and Integral Bank Worth 3.0-23 Plot (SDB, F-10, Stuck Rod) O
., a
.... . . - ~ . -.
LIST OF FIGURES _C ' FIGURE TITLE PAGE NO. 3.2-10 Differential and Integral Bank Worth 3.0-24 Plot (Overlap) 3.3-1. Rod Withdrawal Limits 3.0-27 4.4.3-1 Channel Current Vs. Reactor Power - 4.0-48 Channel N41 4.4.3-2 Channel Current vs. Reactor Power - 4.0-49 Channel N42 4.4.3-3 Channel Current Vs. Reactor Power - 4.0-50 Channel N43 4.4.3-4 Channel Current Vs. Reactor Power - 4.0-51 Channel N44 (~h .a O vili
INTRODUCTION O This report presents the results of initial startup testing at the Wolf Creck Generating Station from initial fuel load until the completion of the Power Ascension Test Program. The plant performed exceptionally well during the startup phase allowing the entire past fuel load test program to be completed in 169 days. Wolf Creek is a Standardized Nuclear Unit Power Plant System (SNUPPS) unit located in Coffey County, Kansas. The Nuclear Steam Supply Systan (NSSS) is a four loop, Westinghause pressurized water reactor (PAR) rated at 3411 megawatts thermal (%T) (3425 ST including reactor coolant pump (PCP) heat). Gmeral Electric provided the turbine generator for Wolf Creek. Bechtel Power Corporation was the architect for the entire power block. Sargent & Lundy acted as architect-engineer for the site-relate 3 portions of the project that were not part of the SNUPPS design. Daniel International Corporation was the site constructor for Wolf Creek. License No. NPF-32 was issued by the Nuclear Regulatory Comission (NPC) on March 11,1985, which authorized Kansas Gas & Electric to proceed with initial fuel loading and low power testing, including initial criticality and low power physics tests, at prawer levels not in excess of 31 rated thermal power (RTP) . The first fuel assembly was inserta into the core ca March 12, 1985, and fuel loading was completed on March 17, 1985. After the installation of the upper internals and the reactor vessel head, the RCS was filled and vented. Cold plant testing was authorized on March 27, 1995, and was gompleted on April 17, 1985. The RCS was at hot standby conditions (557 F, 2235 psig) on April 30, 1985. Post core loading precritical testin7 was completed and approved on May 19, 1985. The reactor was tsken critical on May 22, 1985, and low power physics testing was comenced. Iow power physics testing was completed on June 3, 1985. On June 4,1985, the NRC lif ted the 5% power restriction and issued license number NPF-42 for full power operation. Wolf Creek entered Made 1 (>5% power) for the first time on June 6,1985. The turbine generator was synchronized to the grid on June 13, 1985. The 100% plant trip test, performed on August 28, 1985, was the final test in the Power Ascension Test Program. Following a brief maintenance outage the unit was declared commercial on September 3,1985, at 0114. t 0 1
l 1.0 INITIAL CORP. LOADIfC O The initial core loading test sequence consisted of activities requirai prior to and daring the actual loading of fuel, Many of these tests were required to be performe3 within certain time periods prior to the fuel load. The Nuclear Instrumentation Systen (NIS) was tested well in advance of core load. Proper functioning of the systan was verified including all alarm and trip mc<;hanisms. Numerous deficiencies were encountered during the performance of this procedure, however, all were resolved prior to closing the procedure. Most of the deficiencies dealt with camputer points not giving expected responses and were successfully resolved by having the computer group rebuild the [nints. Two cases of equipnent failure were discovered during the test, one of whicil was a high voltage supply and the other a control board meter. Both pieces of equipnent were replaced and successfully tested. Following the functional check of the NIS, the source range preamplifier and pulse amplifier gains were adjuste3 for optimum settings. This was accomplished using a portable neutron source with a strength of approximatell' one curie. The neutron source was placed near the source range detector housing and the high voltage bias adjusted to obtain data which was then plotted. By determining the point at which the curve deviated from a best ilt straight line, optimtrn bias settings were chosen. These settings were then verified by disabling the high voltage power supply bias supply and checking that the count rate settled out below 5 counts per O second. Actual testing was performed using surveillance (STS) proceiures which satisfied the same requirenents as the startup procuiures. No problans were experienced with those portions of the STS procedures that were related to the startup testing. The renainder of the requiranents of the core loading test sequence were restricted by time limits. Some of the limits were imposed by Technical Specifications, others by Westinghouse or plant administrative recommendations. Tre first such' limit was that within 7 days prior to core load, tanoorary nucl ear monitoring instrumentatien should be setup and checkel and that valve 'ineups be performed in preparation for chenistry sampling. The tanporary instrumentation package, which consisted of 3 neutron detectors and all associated equipmmt, was supplied by Westinghouse for use during core load. The equipment was setup on the refueling deck inside the contain.nent building and settings adjusted in accordance with a Westinghouse pre-shipment calibration. A neutron source was placed near the detectors and the high voltage variel to obtain data for voltage plateaus. The results of the plateaus provoi to be consistant with the pre-shipment calibration indicating no changes in settings were necessary. The high voltage operating setpoint to be usai was determino3 to be 2100 volts on all three detectors. Due to the time requir?nents, this test was repeat +3 three times because the operating license was not received as anticipated. 1.0-1
As required by Plant Te:hnical Specifications, containment ventilation, containment penetrations, and the refueling machine were proven operable j O within 100 hours of core load. Again, these procedures were repeated as the license was delayed. Verification of valve lineups and boron concentration sampling commence 3 72 hours prior to the anticipated coce load time and was repeatc3 several times due to delays in license receipt. Sample results indicated that the boron injection tank was out of specification high and the boric acid storage tank was out of specification low. By recirculating the tanks, both were brought within specification. A sanple from the excess letdown heat exchanger line could not be obtained since the reactor coolant system fill elevation was below that of the sanple point. This line was then isolated and valves tagged shut to prevent any possible mixing of the bao systens. After being placci in the reactor vessel and within 8 hours of the start of core load, the temporary detectors were responso checked. This verified that the count rate would increase when exp9sai to a neatron source and was accomplished by lowering a source into the vessel and placing it next to the detectors. The initial position of the detectors within the reactor vessel is shown on Figure 1.0-2. Alco during this 8 hour period, an analog channel operational test was performed on the NIS source range channels. On March 11, 1985, operating license NPF-32 was issued to Wolf Creek. Prerequisites to the core loading procedure were completed and a briefing was held with all involved personnel to confirm responsibilities. Background count rates were determined for all detectors, tenporary and permanent. ICRR monitoring was performed concurrently throughout the entire O core load. As expected, very low background count rates were obtained which were on the order of 0.02 to 0 05 counts per second. As reference counts were taken, tanporary detector A did not respond reliably. Since temporary detector a was not considered to be a responding detector until late in the core loading sequence, it was switched with detector A thus making detector 3 the inoperable detector. Detector 3 was never declared operable but there were at least two available responding detectors at all times. After verifying that all requiraments for core load were completed and that the refueling equipment was operating properly, the Plant Manager's permission to begin core load was obtained. At 0747 on March 12, 1985, the first fuel assembly, assembly C04, was removed f rom the spent fuel storage pool and placed into the upender. The first problen was encountered when the transfer cart would not traverse all the way to the containment building. Assembly C04 was returned to the spent fuel pool in order to troubleshoot the transfer cart. It was founi that the energency pullout cable was tangled in the cart device mechanism. A scuba diver was sent down to remove the cable. Instead of replacing the cable, it was decide 3 to use the cart without the cable since this was a new core and with no hic' radiation present, a diver could, if necessary, go down to attach a n u cable at a later time. At 1216, af ter a 4.5 hour delay, assenbly C04 was again renove3 f rom storage and this time successfully loaded into care position L-15. O 1.0-2
As assembly C04 was moved over the reactor vessel area, the high flux at shutdown alarm sounded. !bs alarm was then blocked as this was not O unexpected. Assembly C04, as well as the second assembly loaded, assably C30, contained the primary sources. The sources within these assmblies were Californium - 252 (Cf-252) which were previously installed on January 12, 1985. Since this was a new core, the water was only at the level of the nozzles within the reactor vessel and just above the transfer tube within the transfer canal. Therefore, there was no moderator between the fuel assembly and the source range detector hausing which allowed an increased number of neutrons to reach the detector. The high flux at shutdown alarm remained blocke3 until both assemblies were installed in the core. Core loading operations were suspended after the first fuel assembly was unlatched in the core to investigate a problem with the spent fuel bridge crane. The hoist of the crane was " chattering" as the fuel assemblies were being lifted. Since core loading was performed in a smi-dry condition, the lack of bouyancy presented an increased weight to be lif ted by the hoist. The original setpoints for the baist were for wet conditions and were thus adjusted to compensate for the present dry condition. Af ter doing so, the crane was functionally checked and upon successful completion, core loading resumed. Following this 3 hour delay, the second assably was terreved from storage at 1602. Since the first two assemblies loaded were source bearing assemblies, the count rate significantly increased as expected. The count rate from source range channel N31 increased to 6.24 counts per second and N32 to 3.75 counts per second. Before proceeding with core loading, the high flux at shutdoan setpoints were adjusted to five times these values. Core loading then Q continued as illustrated in Figures 1.0-1 through 1.0-10. As each fuel assembly was being inserted into the reactor vessel, outputs of responding tsporary detectors were monitored on a strip chart recorder at the core loading station inside the containment building. Permanent plant instrumentation was manitored in the control room. Af ter verifying the core was not approaching criticality, the fuel assablies were unlatched from the manipulator crane. Count rate data from four operTole detectors (2 temporary, 2 permanent) were obtained by averaging the results of threa counting periods. This data was then used to plot an inverse count rate ratio (ICRR) curve. The source bearing assemblies were loaded into the core first, however, effective reactivity nunitoring could not be achieved until the sources were
- toved to their final position and a cluster of assablies built around them. Therefore, ICRR monitoring was not initiated until af ter steo 13 of the loadina sequence. Data was then obtained througtrut the loading sequence and ICRR plots updated following each steo of the sequence. These plots are included in Figures 1.0-11 through 1,3-15.
At no time was core loading interrupted due to high count rates or unexpected changes in the ICRR. However, there was some concern following step 87. After this assably (assembly A04) was loaded, only one of the resconding detectors was indicating greater than 2 counts oer second. The Final Safety Analysis Report requires at least two detectors indicate this O count rate. This did not present an icmediate concern since another detector that was not identified as a resnoniing detector did indicate this 1.3-3
m FIGURE 1.0-1 [u 4 CORE LOADING SEQUENCE LEGEND Legend for Core Loading Figures
-- Assembly loaded in pernanent position in previous step.
Assembly loaded in temporary position in previous step. Assenbly loaded into position during loading step Number N.
- Location of Tetporary Detector A (B and C) .
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,0 't) Not as yet loaded. Note: Arrows indicate detector or fuel movement. q LJ 1.0-4
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value'thus satisfying the FSAR_ requirement. Nowever, the concern was that toward the end of the loading sequence, the temporary detectors have to be removed in order to place the final fuel assemblies in the core. This would only leavelthe two NIS detectors which were not indicating the required 2 counts per second. The situation was evaluated and the requirment of 2. counts per second was changed to 0.5 counts per second following a 10CFR50.59 evaluation. Boron samples of the reactor coolant system taken throughout the core load ~ resultedinanaveragevalueof2104ppmwitharangeof2039ppmto2123 ppg. Residgal Heat Removal Systs temoeratures averaged 95 F, ranging from 83 F to 101 F. With the exception of the transfer cart and the spent fuel bridge crane hoist, no significant equipment problems occurred during the core loading operation. A few instances of minor difficulties were encountered with the tm porary monitoring equipment all of which were resolved in short periods of time. Core loading was successfully completed at 0600 on March 17, 1985, as the last assembly was unlatched in core position L-15. The total elapsed time for loading the complete core of 193 fuel assemblies.was 118.25 hours. Including delay time the average ass ably loading rate was 1,63 assemblies per hour. Core loading operations continued 24-hours a day using two 12 hour shifts. Imediatelv following completion of the core load, a video map of the core O V was taken to verify proper fuel assembly positioning and that the proper insert was contained in each fuel assembly. An underwater television camera and video cassette recorder were used for this verification. The video tape was than reviewed- for double verification. The results of the map are illustrated in Figure 1.0-16. Upon completion of the map review, the initial core loading operation was complete and accomplished in an exemplary manner. l l L l O 1.0-19
- _ _ _ _ - _ _ _ _ _ . . . . _ _ _ _ _ _ . _ _ _ _ _ , __ ~
FIGURE 1.0 WOLF CREEK GENERATION STATION - CYCLE 1 FINAL CORE LOADItG t%P
- 8 R P N M L K J H G F E D C R A C27 C43 C60 C12 C07 C50 KZo L l l l 11605 810P. 430S B10P. 405 B10P.12805li en
; 1 w 110 i Q~"
C36 C32 C64 A38 Col A22 C04 A16 C28 C)2 , 6 19PS1 RS1 12PSD R34 A9P1D B9P60 R50 12P2D R17 20P90 R43 056 C51 B46 A30 B03 A40 B45 A42 B19 A33 l B47 C15 C17 A9P40 09P50 92DS 20P25:428 16P12D R14 l6P17D R47 16P3C R13 20P24C 77DS 3 C47 B28 B24 B48 A43 B34 A52 B43 AD1 B49 059 B39 C05 4 R26 20P2D R22 20P11C 0551 16P28C RS2 16P13C 3SDS 70P330 R35 20P80 R04 C06 C42 A49 B16 A51 B20 A32 B61 A23 B09 A14 B22 A28 C25 C45 5 4005 12P40 R12 20P22C 23DS 16P2D 25D5 16P18: 4605 16P310 75DS 20P50 R02 12PtC 7905 C3B A06 B37 A05 B62 A41 B10 A26 B31 A10 B63 A46 B14 A62 C3* 6 A10P4h R05 l6P32D 530S 16P16i R29 20P19; R20 20P3C R31 !6P23D 6605 16P8D R48 A10P2L Cat C03 A13 B12 A53 BIS AIS B07 A03 B13 A63 806 A61 C55 C40 7 8305 20P27 R16 16P4D 3305 :0P32: 5D5 20P200 SDS 20P70 97DS 16P11f R19 20P15: 2405 C22 A37 B18 A36 B38 AE0 Bil A29 B53 A35 B32 A21 B57 A21 C13 0 A10P11 RS3 16P300 R40 16P260 RDB ?0P260 R39 20P170 R41 16P10D ROI ?6P20: R1r L10P3D C20 C61 A53 B50 Ab4 B33 A47 817 A44 B02 A56 B64 A48 l C1f C24 9 540S 20P21C R10 16P220 340S 20P60 4005 20P40 45DS 20P23: 47DS 16P70 R36 20P)Et 39D5 C09 A02 856 A17 B35 A13 B23 Att 001 A55 B26 A04 005 A39 C39 10 410P6C R25 16P25; 8805 :6P27C R10 20P34D R15 'OP13C R23 l6P210 56D5 16P29: R37 L10P50 C08 A25 804 A50 B42 A27 B58 A20 BS2 A59 B40 A65 C64 C49 C53 ' 125DS 12P80 R09 20P31011?DS l6P190 1705 16P50 4405 16P15 134Di?0P140 E49 12P6D 1405 C19 854 836 B51 A31 830 At2 B60 A07 B21 'B27 855 CIS p' R18 20P100 R54 00P28C 7305 16P240 R07 16P60 0552 PDP12; R33 20P16: R32 C35 C59 329 A08 B$1 A57 B44 A45 B25 A54 BOS C33 C46 A9P2: 55D5 20P30: R55 16P90 16 Pit R27 l6P140 R06 10P29D t27DSld9PSD I' C16 C23 C21 A09 EJ1 A34 C30 A19 C62 C57 CAB ,# ' A9P3: R44 12P7D R38 20P10 R46 19PS2 R03 12P30 R42 .19P 70 C44 C14 CS2 $p, C37 C58 C29 35 9805 310P7D 3005 ,y 9005 110PSD 65DS O Top - Fuel Asserbly ID Bottcm - Insert ID XXDS - Thimble plug RXX - Control Rod Assenbly YYPXXD - Burnable poison with YY poison rods (symnetrical) XYYPXD - Barnable poison with YY poison rods (non-symnetrical) XXPSX - Primary Source Assably XSSX - Secondary Source Ass mbly 9 1.0-19
2.0 POST CORE LOADItC PREERITICAL TESTI?G O After completion of initial core loading, preparations were begun to perform the post core loading precritical testing phase of the Startup Test Program. This test phase performed the final testing and alignment of various plant systans prior to initial criticality. This ginvolved testing at cold shutdown conditions (RCS average temperature <200 F), testing during plant heatup and pressurizatign and testing at hot no! oad 1 conditions (RCS average tanperature 557 +0,-5 F, TCS pressure 2235 + 25 psig). The upper internals were installed in the reactor vessel on March 17, 1985, and the reactor vessel head was set on March 18, 1985. With the completion of the tensioning of the reactor vessel head studs on March 21, 1985, the plant entered Mode 5 (Cold Shutdown) . The fill and vent of the Reactor Coolant Systen was completed on March 29, 1985. During the Cold Shutdown period the following testing was performed:
- 1) CRDM polarity checks,
- 2) Cold, no flow tod control system,
- 3) Cold, full flow rod ccMrcl systan,
- 4) Initial incore novable detector tests,
- 5) Thermal and dynamic testing of the main steam and feedwater systens O ciattiet deta coltectioa),
- 6) Initial testing of the reactor vessel level instrumentation systcm (RVLIS).
Cold shutdown testing was completed on April 14, 1985, and the plant entered Mode 4 (Hot Shutdown) on April 17, 1985. During the heatup phase, the following testing took place:
- 1) RCS RTD/'IC cross calibration,
- 2) Continuation of the thermal and dynamic testing of the main steam and feedwater systens,
- 3) Special test of the pressurizer power operated relief valves (PORV) at operating RCS pressure (2235 psig),
- 4) RVLIS data collection.
The plant entered Mode 3 (Hot Standby) on April 26, 1985. Tecting at 450 CF RCS average tenperature was completed on April 28, 1985. Loose parts noise was noted on Channel 2 of the Loose Parts Monitor. Investigatian of the noise continued during the boatup and subsequent testing at 557"F. The noise was determined to be caused by a vibrating thimble (tube 42) in the O tocore moott><1ea syste ,- 18ts mes eveteeted es e :rimor Pr 8te,8,a dia , e impact later testing. 2.3-1
0 The RCS was at 557 F, 2235 psig on April 30, 1985. Testing performed at this plateau included:
- 1) Setting of continuous spray flow valves,
- 2) Verification of pressurizer spray and heater effectiveness,
- 3) Verification of RCS flow rate,
- 4) Determination of transport time in the RCS RTD by? ass loops,
- 5) Hot, full flow control rod system,
- 6) Hot, no flow control rod systen,
- 7) Checkout of the incore movable detector system, d
- 8) Measurement of ~ the reactor coolant loop flow coastdown time,
- 9) Precritical alignment of the Tavg and delta T instrumentation,
- 10) Precritical alignment of the Nuclear Instrumentation System (NIS),
- 11) Initial data collection at hot no-load tenperature and pr~ -" e or startup adjustments of the reactor control system.,
~
- 12) Checkout of the loose parts monitoring system,
- 13) Checkout of the thermocouple core monitor systen,
- 14) Background data collection for the biological shield testing, The final test of the rod control system prior to initial criticality was completed on May 18, 1985. All precritical testing was campleted and the ~
Plant Safety Review Committee approved the test packages in May 19, 1985. The following pages of this section contain detailed diset ssions or the post core loading precritical test program. Those ongoing procedures that were completed later in the Startup Test Program are discussed in Section 4 of this Startup Report:
- 1) Power Ascension Thermal and Dynamic Test - Section 4.7.
- 2) Biological Shielding Testing - Section 4.8,
- 3) RVLIS - Section 4.11.2.
O 2.0-2 1
2.1 C0!(mOC. TKX) TESTI!C O A major partion of the testing performed during the post core loading phase t of the test program was involved with the vatious canponents of the rM control systs. The full length control rod cyst s consists of 53 ro3 cluster control assemblics (RCCA) with each assembly consisting of individual rods constructed of hafnium encapsulate 3 in cold worked stainless steel tubing. The assemblies are divided into 5 shutdown banks (SBA, SBB, SIC, SBD and SBE) and 4 control banks (CBA, CDU, CBC, CBD) . Each bank, except shutdown banks SBC, SBD snd SBE, is divided into two groups. Shutdown banks SBC, SBD and SBE have only one grcup. Each group consists of 4 rods each except control banks CBA (2 rods in each group) an$ CBD (2 to3s in group 1 and 3 rods in gcoup 2) . The rod banks are located in the core as sho N ' Figure 2.1-1. Ne . length rods are mved by a Westinghouse magnetic jack type dri"e mechanism. Each control rod drive mechanism (CRDM) contains three magnetic induction ails which energize in a cyclic sequence to mye the rods. Loss of power to tacco coils causes he rods to drop to the fully inserted position. The rod control system is designed to allow individual novment of the various banks or moveent of the control banks in the overlap mMe. In the overlap nodo, one control bank is withdrawn until it reaches a predetermine] setpoint where the next control bank in the sequence bmins to cove in synchronization with the first bank. During control bank withdrawal the sequence is CBA - CBB - CBC - CBD. Control bank insertica in overlap reverses the withdrawal sequence: CBD - CBC - CBD - CBA. During the stepping of a bank, each rod group alternates motien to provide a more uniform reactivity change. In addition to rod novment by bank, an individual rod can be moved by use of the lif t disconnect panel. This allows an out of position rod to b<: testored to the bank position. O 2.0-3
i FIGURE , l '~ 1 CONTROL RCD l.CCA TIONS { R P N ML K J H G F E 0- C B" A s.a-1 . 2 4 @ @ @ $ .. a @ 4 4 4 ! 4 4 @ $ 19 4 e # % . e @ @ @ @ " 7 4 @ e -
@j $ @ ~
@ @ $ @ no- .
e @ @ g to @ @ @ @ @ 11 4 _ 4 12 4 @ $ @ $ - is @ @ 4 4 14 4 @ @ @ $
$5 o'
t BANK NO. CLUSTEPS A- 8 OSHUTDOWNBANK h D 4 E 4 A 4 O CONTROL BANK B 8 C 8 D 5 2.0-4
r 2.1.1 CotD AND 110' ROD COtt!TOL SYSTIM TESTI!C O During the pont core loading test scquence, major tests of the rod control syston were performed under four different sets of PCS conditions:
- 1) Cold, no flow (<200 0F, >350 psig) 3
- 2) cold, full flow (<200 F, 2 0 peld 0
- 3) Hot, full flow (>551 F, 2275 +25 psig) 0
?) Hot, no flow Q500 F, 2235 +25 poig)
The cold, no ficu and h?t, full flow tests performe checks of CRDM operation aM of rod position inoication, veritial rod movement over the entire range of travel, verifiai proper slave cycler timing (cold, nb flow only) aM determined the rod drop time for each individwil rod. The effectiveness of the thimble dashpot region for decelerating the control rod ' was verifici during each rod drop. The cold, full flow and hot, ao flow tests measural rod drop tiines and verified thirrble dashpot effectiveneus only. The initial testitig petformed was the slave cycler timing. One rod in a power cabinet was withdrawn to 54 steps and coil current traces (lift coil, movable gripper coil and stationary gripper coll) were taken during a six (N step withdrawal and a six step insertion nequence. The ru3 was then N inserted to rod bottom and the process was repeated until one rod in each power cabinet (five) had been tested. Each slave cycler operated satisfactorily. The remaining testing was done by rod bank. For the ccid, full flow and hot, no flow testing, only those steps necessary to measure the rod drop times were performed. Each rod bank was withdrawn to 54 steps while taking rod position indicition data at intermediate points. At 54 steps, each rod was individually withdrawn 6 steps and insertai 6 steps while recording coil currents. An additional 4 step withdrawal, 4 step insertion trace was taken for Westinghouse evaluation. These traces were used to determine that CRDM operation was satisfactory. When coil current traces had been collected for each rod in the test bank, the bank was withdrawn to 228 steps while taking rod position data at intermediate poin h. Test equipmmt was then setup to monitor stationary gripper coil current aM digital rod position indication (DHPT) coil output. Each rod was dropped individually by pulling the novable gripper coil fuse aM then pulling a stationary gripper coil fuse while takin] a visicorder trace. These trsees were used to determine the drop time for each rod aM the effe tiveness ot the thimble dashpot region to decelerate O 2.0-5
i 9 l the dropped rod. O arter 11 roa ia the te t de"x a a deem arovvea, tae deex 228 steps aM then inserted until the rod bottom LED's energized to witaar , te . demonstrate satisfactory operation over the critire range of travel (withdrawal and insertion). This procedure was then repeated for each Rod Bank (control and shutdown) until all banks had been tested under the applicable test conditions. After all the rods had been dropped, the standard deviation (sigma) was determinM > and any rod having a drop time outside a + two-sigma limit was dropped six additional times to ensure performance roTiability during subsequent operation. The following test results were obtained from this series of tests:
- 1) The slave cycler for each rod control syste power cabinet functioned 7 satisfactorily during control rod withdrawal and insertion operations, t
- 2) CRDM operability was deonstratai at cold, no flow and hot, full flow conditions. All CRDM's operated satisfactorily,
- 3) The perfomance of the DRPI syste das dmonstrated at cold,- no flow and hot, full flow conditions. The DRPI syst s and related rd position indications satisfied all acceptance criteria,
- 4) All rods operated satisfactorily when withdrawn and inserted over their entire range of travel at both cold, no flow and hot, full flow O, cenditio#s, l 5) The rod drop times for each individual full-length shutdown and control *
! -rod were less than 2.2 seconds under all test conditions: cold, no l flow; cold, full flow; bot, full flow; hot, no flow. The rod drop l times are summarizo$ in Table-2.1.1-1, l. 6)'During cold, full flow rod drop testing, rod D-2 (SBA) was outside the upper two-sigma limit (1.47 seconds) with a drop time of 1.48 seconds. i Six additional rod drops were performed with a minimum time of 1.48 seconds and a maximum time of 1.51 seconds. This range of 0.030 seconds was greater than the 0.020 seconds allowable. After engineering evaluation, the drop times for D-2 were determined to be acceptable. The D-2 drop time was within the two-sigma limit for the r eaining two tests at hot plant conditions. During hot, full flow rod drop testing rod B-10 (CBB) was outside the lower two sigma limit (1.33 seconds) with a drop time of 1,32 seconds. l: Rod B-10 was dropped six additional times with drop times of 1.33 to - L 1.34 seconds,
- 7) The rod drop traces were consistent with no signs of binding or other abnormalities under all test conditions. The rod deceleration through j' the thimble dashpot region was similict for all rods at each set of test conditions. The thimble dashpot region was effective for O. de=etereeias the ccetrot roa aerias e ca roa aron-2.0-6
TABLE 2.1.1-1 ROD DROP TIME-
SUMMARY
Rod Drop Time to Dashpot Entry . R00 CORE Cold, No Flow Cold, Full Flow Hot, Full Flow Hot, No Flow BANK Coord. H-6 1.22 1.44 1.36 1.18 j CBA H-10 1.22 1.45 1.37 1.19 F8 1.20 1.43 1.35 1.18 K-8 1.21 1.46 1.33 1.19 F-2 1.21 1.44 1.36 1.20 B-10 1.22 1.40 1.32* 1.20 K-14 1.21 1.44 1.35 1.20 CBB P-6 1.21' 1.43 1.36 1.21 B-6 1.19 1.42 1.36 1.21 F-14 1.20 1.41 1.35 1.20 P-10 1.20 1.41 1.36 1.21 K-2 1.22 , 1.44 1.37 1.22 11 - 2 1.21 1.43 1.37 1.21 B-R 1.22 1.44 1.35 1.20 H-14 1.20 1.45 1.34 1.20 CDC P-8 1.21 1.41 1.34 1.21 F-6 1.21 1.42 1.'7 1.21 F-10 1.21 1,44 1.37 1.20 K-10 1.22 1.43 1.36 1.19 K-6 1.22- 1.44 1.38 1.22 D-1 1.22 1.44 1.15 1.21 M-12 1.20 1.43 1.35 1.21 - CBD- D-12 1.21 1.41 1.34 1.20 M-4 1.21 1.43 1.35 1.23 H-8 1.20 1.45 1.37 1.19 D-2 1.19 1.48* . 36 1.21 SBA B-12 1.18- 1.40 :. 35 1.19 M-14 1.20 1.44 1.36 1.20 P-4 1.21 1.41 1,39 1.21 B-4 1.19 1.44 1.33 1.20 D-14 1.19 1.45 1.38 1.21 P-12 1.18 1.41 1.36 1.22
'M-2 1.20 1.46 .1.39 1.22:
O-2.0-7
TABLE 2.1.1-1 (Cont) RCD DROP TIME
SUMMARY
Rod Drop Time to Dashpot Entry RCD CORE Cold, tb Flow Cold, Full Flow Hot, Full Flow Hot, tb Flow BANK Coord. G-3 1.21 1.44 1.36 1.22 C-9 1.22 1.43 1.34 1.20 J-13 1.20 1.44 1.36 1.20 SBB N-7 1.20 1,43 1.36 1.22 C-7 1.20 1.40 1.34 1.19 G-13 1.20 1.43 1.36 1.19 ,
'N-9 1.19 1.41 1.36 1.20 J-3 1.20 1.44 1.38 ,
1.23 E-3 1.19 1.43 1.34 1.22-SBC c 11 1,20 1.41 1,33 1.20 L-13 1.19 1.43 1.34 1.21 N-5 1.20 1.43 1.34 1.23 C-5 , 1.20 1.42 1.36 -1.22 SBD E-13 1.19 1.42 1,36 1.20 N-11 1.19 1.40 1.34 1.21 L-3 1.21 1.44 1,39 1.23 H-4 1.21 1.42 1.35 1.21 SBE D-8 1.19 1.40 1.36 1.22 i - H-12. 1.21 1.46 - 1.36 1.20 L M-8 1_.22 1.42 1.35 1.23
- Aver-
! ace 1.20 1.43 1.36 1.21 l Redroppe3 isix times l l l l l iI
'()k:
2.0 8 i V+WW* - -P9's=-W'T "W *f 79 'f--f- -m4ipyi
- ewW M f -PewP%ie-'-*4-r'y6 1=ier yW4r4+4wLMmew '3m- p., ws_,vmgm-ie3w.yer4-9-.,is.
. . . - - . - _ . . - -- . . _. - - . - _ - _ . _ ~ - - . . . . -
i The following problems were encountered during the performance of these test procedures !
- 1) During withdrawal of CBD, DRPI indication for Rod K14 was lost.
Detector encoder card A406 in CRPI Data Cabinet A was fourd to be defective and was replaced (cold, no flow test),
- 2) During withdrawal of CBD, computer indication of rod position reained at zero. A power supply in cabinet RJ048 was found rowered down for investigation of an unrelated problem causing the zero indication on the computer. The computer data was coliccted af ter the powm supply was energized (cold, no flow test),
- 3) During withdrawal of SBC, demand position as displayed by the computer indicated zero when all other indications were 18 steps withdrawn.
During investigation, tha group step counter for SBC would not inerment/decreent durity rod r:ovment. The following itms were correctcd during the investigations , a) The Kl9 relay in the Logic Cabinet was found to be sticking and - was replaced, b) The slave cycler card for shutdown banks C, D and E (Power Cabinet SCDE) had failed and was replaced, c) The A t. B DRPI coil cables for rod Cll were interchanged at the Reactor vessel head connections. These cables were reconnectai in their correct locations, d) A cable was found terminated in RJ049 instead of RJ048 Since this termination was correct per the scheme drawings, a Tmporary W)dification was used to correct the termination. The T e pcrary Modification was later made permanent. The required rod position indication was then collected when SBC was withdrawn for rod drop testing (cold, no flow test). L l l O-2.0-9
2.1.2 noo CORrROL SYST m TEST Prior to initial criticality, the full length rod control systmi was teste3 to verify satisfactory performance of the require 3 control and indication functions and to verify that a manual rod block prevents manual withdrawal of the full length control rcds. All the acceptance criteria of the text were met. This final operational check was performed by rod bank. Each rod bank was withdrawn to either 18 steps (shutdown banks) or 48 steps (control banks) while annitoring proper operation of the group step counters, individual rod position indication (DRPI), rod speed indication and the console indicating lights. A set of rod position indication data (DRPI) was then taken. Each group in the withdrawn bank was individually put on the DC hold bus and the fusible disconnects for the appropriate power cabinet were opened. After verifying that rod position did not change, the fusible disconnects were closed and the CC hold switch for the rod group under test was returned to the off position. The bank was tnen inserted to the zero step condition (rod bottom). This process was repeated until all rod groups had been checkoi on the EC hold bus. The DC hold bus perfor.rsi satisfactorily for all rod groups. No problens were noted with Rod Position Indication or any other control or indicetion function. Bank overlap was checked using overlap setpoints that allowed the control banks (CBA, CBB, CDU, CBD) to be withdrawn in MANUAL to approximately 30 O steps. After the overlap and rod position data had been collected, the control rods were inserted in MANUAL to ver,ify that the bank overlap function worked satisfactority. The bank overlap systen functioned to start ard stop the coatrol banks (CBA, CBB, CBC, CBD) at the correct setpoints. A manual rod withdrawal block was simulated by lif ting the lead of cable 5SFS10AH at TP-PP-16 in RPO40. An attempt was then made to withdraw Control Bank A in MANUAL. Control Bank A did not withdraw. The lead was then relanded ard the ability to withdraw Control Bank A was danonstrated. l l l O 2.0-10
_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ ~ _ _ _ _ _ _ 2.2 ItLDRE ffNABLE LIMX'IVR SYSTEN O The Incore Movable Detector Systan or Flux Mapping Systan is designed to provide a three-dimensional reactor core power profile through the use of movable fission chambers (neutron detectors) being moved axially in radial positions throughout the core. Dy placing the detectors into selected positions within the core, detector currents are provided and stored. This data is then processed yielding necessary information for reactor core surveillance. The Flux Mapping Systen consists of two major itenst a flux mapping console and a detector drive systan. The detector drive syster consists of four (4) trains cach providing a me:hanical means of routing a datector into any ont, of 58 guide thimbles in the reactor core. The guide thimble positions are shown in Figure 2.2-1. Each detector is routed through a 6-path transfer device, a 15-path transfer device, and the seal table before reaching the guide thimble. Via the 6-path transfer device, each detector is capable of accessing another train's 15-path transfer device. The Flux Mapping Console provides a means of remotely controlling the dete: tor drive systen including all tho drive units and transfer devices. Two identical sections within the flux mapping console provides backup capability should one of the sections fail. The detector currents, which are a measure of neutron intensity, and therefore core power, is reported to a CRT, printer, floppy diskp and the Westinghouse P2500 plant process computer. After core loading, the guide thinbles were inserted in the core and final O installation of the detector drive systen was complete 9. The operation of the system was verified in two parts:
- 1) The operability of the detector drive systan was danonstrated.
- 2) The operability of the integral Flux Mapping Systen using the renotely statione$ flux mapping console was demonstrated.
OPERABILITY OF DETFrTOR DRIVE SYSTUi The checkout of the drives and transfer mechanisms was done using a manual controller operated locally, and four dumny detectors (detector shaped hut without co-axial output cable and fission chamber). Each one of the four detectors was run into all thimble combinations for that drive and the safety mechanisms were checked with regard to transfer rotation with a detector inserted and winding of retracted detectors onto the takeup reel. The path length of each thimble combination was measured to give an initial path limit value for subsequent checkout of the integral systen. Problans were encountered with operation of the portable controller due to printed circuit card failures, but repairs were made to the necessary components to mable its use. Spring tension on the detector takeup reels requir33 adjustment for proper detector withdrawal operation and a flexible braided storage path thimble required replacement. With these problems corcecte3, checkout of the detector drive systan was successfully completed. 2.0-11
FIGURE 2.2-1 MOVABLE DETECTOR LOCA TIONS R P N M L K J H G F E D C B A 100* 1 o o _ 2 o o o V 3 o o o' o 4 o o o 5 o o o o 6 o o o o o 7 o o o o 8 DO* o o o o o o o o 270' 9 o o o o 10 o o o 11 o o o o o s 12 o o o 13 o o o o 14 o o o o 15 o o O* O Movable Detector (58) O 2.0-12
t OPERABILITY OF '!HE FLUX MAPPItC SYSTEM O- The chockout of the integral systen was done from the flux mapping consoles inside the control room. The menu driven consoles consist of the input computer terminal, printer, floppy disc unit, and data link to NSSS cmputer. The two re3cndant consoles were tested for all modes of operation. Initial checkout of the system was accomplished using the four dumy detectors to verify the thimble path lengths for all detector path combinations. Following path length verifichtion, the dumy detectors were replaced with the operating fission cha:nbers. During the checkout, numerous computer card failures were encountered which were corrected by card replacanent or card / chassis reseating. Several problans with detector drive notors not being de-energized when the detector position escoder indicated no detector motion were corrected by adjustmcot of controlling relays. Problens with data link failure to the Westinghouse P2500 process computer , were corrected by giving flux mapping a higher priority level on the computer systen. Due to systen failures, it was not possible to verify all four detectors could individually obtain a full flux map from all 58 thimbles. However, it was shown that the systen could be used to insert any one of the four detectors in any of the 58 thimbles. Subsequent work on the , flux mapping systew verified the integral flux mapping systan functione3 as - designed. Curing flux mapping in the power ascension program, the top of fuel and path length limits were revised based on grid strap data depressions to give proper detector position / core height correlation. Initial flux maps were run using only two detectors for a full flux map due to detector /canputer O c ra rettore 'ot the aeces==rv aete cor etux =>e eaetvet= *>e deeiaea-Card contacts and failures continued to be a source of proolens during subsequent flux maps but reseating and/or replacanent of the questionable cards enabled use of all four detectors during the test program. O 2.0-13
2.3 PRESSURIZER CORTINUOUS SPRAY PfiW SETTI!G AND PRESSURIZER HEATER AND SPRAY CAPABILITY TPSPS The reactor coolant systs pressurizer establishes primary plant pressure by maintaining a saturated liquid aM vapor environment in the pressurizer at the desired pressure. Activation of tha two spray valves in the opray lines from two RCS cold legs to the pressurizer aM imersion heaters within the pressurizer acts to control saturation pressure, thereby controlling FCS pressure. The continuous spray bypass valves are in parallel with the pawer operated spray valves. These valves provide a small continuous spray flow to warm the pressurizer spray lines and rozzle in order to limit thermai stresses when the spray valves actuate and to assure that the boron concentration in the pressurizer is not dissimilar from that in the reactor coolant loops. PRESSURIZER CONTINUOUS SPRAY FtLW SETTItG This test was performed to establish a setting for the pressurizer continuous spray throttle valves to obtain an optimum continuous spray flow and to establish the setpoint for the pressurizer spray line low tmperature alarms. Each continuous spray throttle valve was opened in discrete increents while tronitoring pressurizer spray line tmperature. Spray line tmperature was allowed to reach equilibrium prior to opening the valve further. Equilibrium spray line temperature was plotted against valve turns opm. The continuous spray throttle valves were set at the break point on the curve where further valve opening had a minor effect on equilibrium I] spray line temperature. Valve Setting Equilibrium Spray Line Taperature BB-V082 51/4 turns opm 535 F 0 BB-V383 5 turns open 522 F The spray line low tmperature slarm bistables were set at 10 + 5 0F below the equilibrium spray 1Mc temperatures. Sistable Sr.tpoint 0 TB-451 521 F TB-452 521 F A review of the FC3 chemistry data dmonstrated that the new settings for the continuous .* pray throttle valves assured that the boron concentration in the pressurize:. was not dissimilar from that in the RCS. The completion of this test satisfied an outstanding test discrepancy from the Preoperational Hot Functional Test Program. PRESSURIZER HEATER AND SPRAY CAPABILITY TEST The purpose of this test was to determine the rate of pressure reduction caused by the opening of both pressurizer spray valves and the rate of pressure increase caused by the operation of all the pressurizer heaters. 2.0-14
i While at hot, no-load conditions (5570F, 2235 psig), the pressurizer spr sy O valves were opened fully. Pressurizer pressure, pressurizer level, pressurizer water temperature and spray line temperatures were recorded on a strip chart recorder. The transient was stoppa$ when pressurizer pressure had fallen to approximately 2000 psig gy shutting the pressurizer spray valves and the RCS was returnal to 557 F, 2235 psig. The results of this transient are shown as Figure 2.3-1. , While monitoring the same parameters as in the spraydown test, the pressurizer heaters were tested by energizing both banks of backup heaters and inje: ting a full danand signal to the control heaters. When pressurizet
~
pressure reached 2330 psig, the pressurizer heaters were secure 3 and the TCS i was returned to 557 F, 2235 psig. The results of this transient are shown f as Figure 2.3-2. i The pressurizer response to the opening of the pressurizer spray valves ICV 455B and PCV455C was within the allowable range as shown on Figure 2.3-
' 1. However, an engineering evaluation of the data determined that both spray valves should open within 5 seconds.
An additional test section was written and performed to determine the stroke tig fran full closed to full open for the pressurizer spray valves at 557 F, 2235 psig. The opening times weret Valve Ooening Time PCV 4558 3.9 seconds O ecv asse 6 97 eeeoa= An additional engineering evaluation, determined that these opening times in conjuction with the previously determined pressure decrease rate were acceptable. The pressurizer response to actuation of all pressurizer heaters was within the allowable bands as shawn on Figure 2.3-2 indicating an acceptable response. l l O 2.0-15
-__i_..._._.._-_.__._-_.-.__-__-_____.____-_----_~-----------------
O O O FIGURE 2.3-1 NOMINAL PRESSURE RESPONSE TO CPENING OF BOTH PRESSURIZER SPRAY VALVES (W1 th Allowable Band) 2250 ,
\
? \x
& 2200 \ _. \' N S \ g rv y \ \
b t \s \ i D 2150 -
\
m \ N
$ 3 N
w \ e \ . N E N T 2200 ' - D \ \ \ N N
" \ \ - Honinsi Response T
- a \ N h 2050 \ \ --Allowabid Band w
T s x - x
\ . Test Dats 4 \ < \
. \
2000 l
\ 1 \
0 20 40 60 80 100 120 140 TIME (s e conds)
(- J \v G FIGURE 2.'3-2 PRESSURIZER RESPONSE TO ACTUA TION OF ALL PRESSURIZER HEATERS (Wi th Allowable ' Band) 2350 /
/
/
/
% 2330 '
/ /
~ E S
? '
; nn /
b 2310 / , h / '/ *
,/
k / *
/ *
/ - Nominal k eponse
$ gggo / ,/
/ r '
G ,
/ ~~ Allowable Band f /
,/
m /
/ / */
M /
- Test Data
/
N 2270 - -
/
- 4. '
x /l,/
/
2250 0 40 80 120 1go yz, 24g TIME (seconds)
2.4 RFJCIOR COOT ## SYSTIN FLOW MF'LSURIMf2R J O Reactor Coolant Systs (TCS) flow indication is obtained from measurement of the differential pressure across the RCS coolant loop piping elbow which connects the steam generator piping and the reactor coolant pump suction (cross-over leg). Each RCS loop has three differential pressure flow transmitters which provide visual flow indication in the control room, flow information to the plant emputer, and voltago signals to the prote: tion system for the loss of flow0 reactor trip. This test was perfo6mai at hot, no-load conditions (557 + 2 F, 2235 + 15 psig) to determine that RCS flow, as indicated by the loop ~ elbow diffe7ential pressures, was equal to or greator than 90 percent of the thermal design value. With four reactor coolant pumps (BCP's) operating and the plant at the required steady state conditions, data was collected from the plant computer and the main control board indicators. This data was then averaged and convertoi to gpm. The total flow as determined from the plant emputer was 435,221 gpm as compared to the minimum required flow of 344,520 gpm. The results are shown in Table 2.4-1. As a result of this test, the twelve differential pressure flow transmitters were adjusted to indicate 100% flow at hot no-load conditions and four PCP's operating. Verification that RCS flow rate is greater than or cqual to the thermal design flow rate using calorimetiic data was done during the power ascension phase of the Startup Test Program and is discussed in Section 4.4.1 of this report. O 2.3-13 l 1
--.,, _,,_,.,_._,.m_ _ . . - - , , - - . 2 --------. y.--m.-_,__,,_y,,,4,y,,, ,.
.~. - ... - .- -..-. _ _ . - . - . . - . .-
t TABLE 2.4-1
'O: RCS LOOP FID4 DETERMINATION j v' PRIOR M INIT!AL' CRITICALITY -
l t t Plant Ccrnouter Main Control Board Irdicator , Percent GH Percent GH RCS 110.3 108,999 111.0 109,668 Loop 1 i ICS _ 110.1 108,749 109.8 100,515. Loop 2 RCS . 109.1 108,594 110.0 109,680 Loop 3 1
- ICS 110.2 108,878 110.3 109,009 Loop'4 Total - 435,221 - 435,873 Flow-Acceptance Criteria, Total Flow > 344,520 gpn i
l l 1 i.
.i EO 2.0-19 p
s
. . . . . . - ..m,-
......_,_.....~...._,-.,.,_..-,,......_...~...,,_,_._...~_,.,___,,,__,....-..,..w,- e,..-_ _ _ . , ,,J.__m--m-
i l 2.5 ItP>c!VII C00fMR SYSTIM lTIM COASTDOWN TEST O This test was performed to measure the rate at which reactor coolant flow changes subsequent to a simultaneous trip of all four reactor coolant pumps and to determine the reactor coolant systen low-flow reactor trip time delay. While operating at hot, no-load conditions, the permissive P-8 was simulated by lif ting leads in the nuclear instrumentation cabinets. This ensuro3 that low flow (190%) in one reactor coolant loop would open the reactor trip breakers. -Reactor coolant systen flow, reactor coolant pump breaker status and reactor trip breaker status was recorded on a high speed chart recorder . The four reactor coolant pumps were then simultaneously tripped by simulating an underfrequency condition with a test circuit installo3 in the safeguards test cabinet (SB-030) . ' The chart recorder trace was analyzed to determine normalized loop flos fractions at discrete points in time following the ICP trips. At each time, the loop flow fractions for all four loops were averaged to datermine the core flow fraction. The relationship: F'(t) = ( l 1 0 895 , 1 F(t) where F(t) = core flow fraction was then used to determine F'(t) (the inverse flow fraction) for each second O- during the first tm seconds following the reactor coolant pemp trip. Using the method of least squares, F'(t) was- fit as a function of time for the period of 2 seconds through 10 seconds: P' ( t) = At + B where A = slope B = Y-intercept The Westinghouse acceptance criterion for the flow coastdown was: TAU g = 1 = Flow Coastdown Parameen A TAUg > 11.70 seconde The test dm.a TAUq was 12.56 seconds which was satisfactory. Table 2.5-1 5"m urizes thesec'alculations. The chart recorder data was also used to determine the low flow reactor trip time delay. The loop inverse flow fractions were least squares fit for a period from three to ten seconds following ' the reactor coolant pump trip. . The least squares fitting parameters were thm used to calculate the sensor delay time for each loop: O f' = 1/f where f = loop flow fraction f ' (t) = At + B 2.3-20
i t i i TABLE 2.5-1 FfD4 COASTD0hh RATE CALCULATIONS O ! l Timo Core Average Flow F'(t) = ( 1 _)0' 9$ -1 (soc) Fraction (F(t)) F(t) { 0 1.0 0 i 1 0.945 0.052 2 0.873 0.129 . 3 0.807 0.212 i 4 0.749 0.295 5 0.699 0.378 i 6 0.660 0.450 + 7 0.619 0.536 8 0.588 0.608 9 0.566 0.691 10 0.529 - 0.768 -i Method of'Least Squares used to fit data F' (t) = At + B
. t A = alope B = Y - intercept
- . A = .0796 B = .0258.
O ,Av, = no n sed - e,r -eeer
=1 A
= 12.36 seconds Westingbuse acceptance criteric.n TAUg > 11.70 s e nds NOTE: Data fit for period from 2 through 10 seconds, t
l I i i L. _ _ O ' 2.0-21
.a....--......_..----.--_.__-...- . - . . . - . - _ _ . _ - - . - . . . . , . . . . - . , . . . . - - . . . . . . . . , . . - - . , . .
f i i Scosor delay time = TD " II ~ '
^
O The longest sensor delay time was 0.532 seconds in loop 4 Combining the ; sensor delay time (Tp ) with the time from whcn the first loop flow decrease 3 - to tha low flow trip setpoint (90%) until the second reactor trip breaker { had oponod.(T4 = 0.109 seconds) and the gripper release time (Tg = 0.15 ; seconds) yieldal a low-flow reactor trip time delay of 0.792 seconds. The : Westingho'ase acceptance criterion was less than or equal to 1.0 seconds, therefore the results were natisfactory. Table 2.5-2 sumarizes these calculations. j The results of this test confirmed, based on the Westinghouse acceptance ; criteria, that the flow coastdown following a trip of all four reactor i coolant pumps is adequate to prevent the departure from nucleate boiling ratio (DtBR) from docreasing below the limiting value of 1.3. t
)
O' L a 0 2.0-22 . ~ , - - . _ . . . _ . < . . _ _ . . . . , . _ , _ - . . . , - . , _ . - , _ , , _ , _ _ , . , , , , , , , , . . . _ . . . , , - _ . _ , _ . , , _ , , , , _ _ . , , _ , _ , , , _ _ . , , , , _ _ _ , , , . , _ _ .,i
l l TABI,E 2.5-2 l IO4-FID4 RrJCTOR TRIP TIME DEIAY O cat 4DIATIONS f,oop Inverso Flows (1) Time Loop 1 toop 2 Loop 3 Loop 4 ! (soc) l 3 0 1.0 1.0 1.0 1.0 1 1.058 1.056 1.065 1.055 2 1.139 1,136
~
1.158 1.148 3 1.241 1.236 1.243 1.234 4 1. 345 1.327 1.335 1.332 5 1.443 1.412 1.438 1.426 6 1.539 1.502 1.508 1.518 7 < 1.659 1.574 1.623 1.616 8 1.709 1.679 1.711 1.706 , 9 1.833 1.742 1.816 1.809 10 1.914 1.880 1,873 1.901 Ah 0.096 0.961 0.089 0.967 0.092 0.095 B 0.969 0.949 Td
'4
(', 0.406 0.371 0.339 0.533 TOTES:
- 1. . Inverse loop flow = f' = 1/f where f = loop flow fraction 2 A and B are least squares fitting parameters for f' as a function of -
tI f' (t) = At + D. Data is fit from 3 through 10 seconds. TD= Sensor Time Delay = (1 - B)/A 3. Time from first loop flow at low flow trip setpoint until second reactor . trip breaker trips: Ty = 0.109 sm. Maximrn sensor delay time TD= 0.533 sec ; Grippar release timo (from rod drop data): T =g 0.15 s m l l- Im-flow reactor trip time delay (T +T y D+T g ): T IE .M sec
=
*riestinghouse acceptance criterion: Tg, < 1.0 seconds l
t i O 2.0-23
i 2.6 IrrD HYPASS FI/M MFASURINNP O Each loop in the TCS contains a loop bypass mnifold in which resistance tmperature detectors (RTD's) are placed to monitor hot and cold 103 tm peratures. Signals from the tonperature detootors are us&3 to compute the reactor cool. ant delta T (tenperature of the hot log, Tg ,,, minus the tmporature of the cold leg, T andanaveragereactor9,6clant tmperature (Ta ,) . This infoN) ion is used by the reactor protection nystan as well as the process instrumentation systen. 4 The manifolds for the cold leg tunperatures draw a sample from the RCS flow at the reactor coolant pump discharge. The manifolds for the hot leg tanperatures draw samples frcxn RTD scoops in the reactor vessel outlet piping. The bypass lines join downstream of each set of tanporature detectors (hot and cold) an3 discharge into a cannon line. The combine 1 bypass flow passes through a flow elanent before dischartiing into the suction side of the reactor coolant pump. , The purpose to this test was to measure the Pm bypass flow rate in each loop to ensure transport times of less than or equal to one secon3 through the loops. The one second transpcet time was a design parameter for the overall RTD time response. The following combinations of flows were measure 3:
- 1) The combined hot and cold leg bypass flow rate was measure 3, O 2) The co1d 1e3 byvese f1ew rete wee meeeures with the het 1e> iso 1eted.
- 3) The hot leg bypass flow rate was measure 3 with the cold leg isolated.
New low flow alarm setpoints were calculated at 90 percent of the total bypass flow for each loop. Since the sum of the individual hot and cold leg bypass flows with the other leg isolated was greater than the measured total bypass flow, a correction was app 11e3 to obtain the actual flows based on the ratio of the measured hot and cold leg bypass flows. Table 2.6-1 su:nnarizes the results and shows that the calculated bypass flows are greater than that required for a one second transport time through the bypass loops. In addition, with total bypass flow at the low flow alarm value of 90% and cold leg flow at its normal value, the hot leg flow stil1 has a transport time less than i second. The nes low flow alarm setpoints were incorporated into the Wolf Cre@, l Generating Station Total Plant Setpoint Document (TPSD). The completion of l this test satisfied an outstanding test discrepancy fran the Prooperational Hot Functional Test Program. I l O 2.0-24 _ _- . _ _ _ . . _ _ . . _ . , _ . . . ~ . -
O O O TABIE . 2.6-1 RTD BYPASS ETJDW MFAStiL J2Tr RESULTS (1) (2) Total Calculate 3 Minimiza (3) Loop Bypass Iow Flow Leg . Recp2 ire 3 : Measured Calculatal Calculate 3 Flow Hot No. Loop Flow- Alarm Setpoint Flow. Flow Flow Leg Bypass at Low Flow (opm) (om) (apm) Alarm Setpoint 1 265 238.5 Hot 103.43 155 146.7 120.2 Cold 67.62 125 118.3 - m 2 265 238.5 Hot 103.43 155 149.4 122.9 o 4; Cold 67.62 120 115.6 - un 3 295 265.5 liot 103.43 185 . 176.0 146.5 Cold 67.62 125 119.0 - 4 282 253.8 liot 113.88 178 16S.4 143.2 Cold 67.62 120 113.6 - Notes: (1) 90 percent of total bypass flow (2) Basel on actual pipe volu~es arx1 ene secon3 transport time (3) Total IPipass Flow at 931, cold leg flow constant
4 .l 2.7 PRILIMINAlW DATA CotLICTION IMR INSTPUMIMP cat 41 BRAT 10ti
.O While operating at hot, zero power conditions prior to icitial criticality several preliminary data colle: tion proce3ures were perlotmo3 for precoiures ,
that were to be performed during the Power Asecosion portice of the test program. The calibration procsiures are discussed in more detail in Sost.$on 4 of this Startup Test Report. An alignment was performe3 of the degts T and Tavg process instr:rentation at hot isothermal conditions (557 +2 F). Using test cards, known resistances were input into ICS lo?p T and T instrument loops snd the loop outputs were verified as well as Eta T. Ten at hot isothermal conditions actual-instrumult loop outputs were recorde1. ICS spare T and T outputswerecomparedtothenormalinstrumentswithsatisfactoY rh2b2ts. Overtemperature. delta T and overpressure delta T setpoints were also verified. Data was co11ecte3 for those components which are used to develop the rod speed control signals. This proca operating and the ICS at 557 + 0, gre was2235 5 P and performed +25 psig. with Parameters four ICP's monitored for each loop were T T Tav Generator Pressure, Turbine Imhs,e M2s,ure,g, and readwater in addition, Flow,auctioneered Steam NIS power, auctioneered Tavg an3 Tpy were also monitored. Test instrumentation was installed to nonitor fee 3 water flow and steam generator pressure during later tests. In addition, :e zero of tha steam flow and fee 3 water flow transmitters was verified at the process esbinets while the plant-was in hot standby conditions. With the stan generator isolated, the transmitter voltages were measure 3 at the process cabinets to ensure no zero shift, One transmitter, AB-FT-542, had to be replace 3 because of static pressure shift. i i i t L l O 2.0-26
-. - - - - . . - . - . _ . - . - . - - . - . . - . - - _ . - . _ - . _ _ _ - ~ . - _ _ . - .
7.8 tRCIAR INSTitlNDITATION SYST1H O The nuclest instrtrnentation systan (NIS) was testo3 to verify that all voltage settings, trip settirrjs, and slant settinJs in the NIS were within expceted tolerances and that all ranges were functioning properly. This was accmplished by empleting a series of inst rument and control channel calibration surveillance procedures written specifically to msure the NIS is operable in accordance with Technical Specifications. A total of twelve ceparate proccdures worst performed to complete this verification. One test was perfomed on each individual source range, intermediate range and power range drawer. The scaler timer and audio ecune rato drawer, ccanparator and rate drawer and the flux deviation and miscollaneous controla drawers were also checke3 out. No major problems or deficienclos were found during any of this testing. A more detailed sumnary of all trating on the nuclear instrumentation systen is givm in section 4.4.3. O O 2.0-27 __.__.1.-__-__.._. -
.- ._-_ - -----..-. =.- _-- - -.- - - _ --
2.9 frID/IC CHOSS CALIBRATION TESTS O The purpose of this test procedure was to provide a functional check out of the reactor coolant systen resistance tenperature detectors (RTD's) and the incore thermocouples (T's) and to generate isothermal cross calibration data for subsequent determinati' > of individual RTD installation correction factors. 0 Thet<sstw3sperformg3atfourdu o ' ant tunperature plateaus 250 p, 345 T', 450 F and 557 F (within +57 At each tmperature plateau, RTD resistances were read directly at t.he field wires using test boxes with
- multitrasition switches ord a digiul vo1Lutoc. Concurrently, T readouts wwe dotsined trom the plant computer, and steam generator satur dion pressures were obtaino3 from tmporary test gauges installed on the mijn steam lines. The RTD resistances were then converted to tenperatures using vendor curves; the saturation pressures were converte3 to tm peratures using the ASME Steam Tables, and the T tenperatures were averaged. All of these tmperatures were comparsi and isothermal corre: tion factors wero calculated.
The initial performance of this test commenced on April 16, 1985 and was ensentially conplete on April 30, 1985. The results are sumnsrized in Tablo 2.9-1. The acceptance criteria for this test were ,
- 1) All converta3 average RTD tanperatures were within +20 F of the calculated iCS average tanperature for the given taiperature plateau.
- 2) All average stgam saturation torperatures for each tanperature plateau were within 12 F of the calculat(d FCS average temperature.
- 3) All thermgcouple average tmperatura for each tenperature plateau are within 12 F of the calculated TCS temperature.
All acceptgnce criteria were met except fo.t the steam saturstion tanperature at the 345 F plateau. Based on the accuraeles of the Heise guages used to determine the stean pressure (11.5 psi) and the consistency of the RTD average, this was determined to be acceptable. Although acceptance criteria were met, as di scusse3 above, during the i- initial performance of the test, further investigation of the data by ,- Westinghouse resultai in a decision to repea the cross-calibration test. l The main concern was whether the TCS was in a truly steady state or isothermal condition whan the initial data wos collectol. In the second ! performance of this test, additional precautions were taken to achieve tne i highest degree of stea3y state and- isothermal conditions in the ICS. In l aldition, data was collected with the RTD-man. fold-return valves in both the I closa$ and the open position. The normal contiguration is with the RTD-manifold-return valves open but Weatinghouse d eterminal that closing the ( valves would result in exposing both hot-leg a1d cold-ley narrow range RTD's l to water at a single tmperature thereby giving better test data. 2.0-28
t 0; O 4 nBL 2.9-1. . . RNTS OF INITIAL RID /IC CROSS CALIBRATION TEST f a t I RTD No. Calculate 3 Installation Correction Fac-tors, OF > 0 0 U 0 250 F 345 F 450 F 557 7 f (
- TE-410A -0.5 0.0 -0.1 -4 ' +
, TE-41GB -0.7 -0.4 -0.1 -0.3 1 ! TE-411 A -0.5 -0.1 -0.2 -0.1 I TE-411B -0.7 -0.3 -0.1
-0.3 TE-413A -0.1 0.4 0.3 0.3 TE-423 A 0.5 [
0.6 0.6 - 0.7 i [ TE-433B 0.3 0.2 0.3 0.3 i !' TE-443B -0.3 0.0 ' O.2 -0.3 l l' TE-430A 0.4 0.4 0.5 0.0 ! I. TE-430B 0.2 -0.5 .-0.3 0.1 l !< TE-431A 0.5 -0.5 -0.5 0.1 ! TE-431B 0.3 [
-1.4 -1.1 0.1
! F TE-4 20 A 3.2 -0.2 -0.5 . -0.1 i $, - eu
? ~TE-4 20B 0.0 -0.3 -0.2 0.1 j c TE-421A !
0.2 -0.1 -0.3 -0.1 ; l TE-4218 0.1 -0.3 -0.1 0.1 ; TE-4138 -0.4 0.1 0.3 0.1
'TE-433A [
0.6 0.4 0.5 0.3 l TE-44UA 0.0 -0.4 -0.8 -0.2 ! j TE-4408 -0.2 0.1 0.3 -0.4 ; i TE-441A 0.0 0.7 0.7 0.0 ! !. TE-441B -0.2 0.9 0.9 -0.8 i { TE-423B -0.1 -0.3 0.0 0.3 i TE-4 4 3 A 0.3 0.4 0.4 l ' 0.' t 4 Avg pt Tor.p. t r l F 248.3 340.7 447.3 556.8 j Steag .at Temp , , F 248.6 339.4 444.8 556.0 4 1 Tm T SAT
.3 1.3 i G 2.5 0.8 1
F Avg E Temp I* 249.8 T U 342.3 447.6 556.7
! F -
-T i lT=ToAW T -1.5 -1.6 -0.3 0.1
'I 3 g l 3 - - --
d 6
i I After reducing power from 30%, tgin secgnd test wg3 performed on June 30, 1985. Data was collected at 375 F, 450 F and 557 F with the RTD-manifold- ' return valves both open and shut. The results of the second test are
] sumarized in Table 2.9-2. The acceptance e riteria runaincd the same for this test with the exception of: ;
- 1) All converted average RTD tanperatures were within +1.70 F of the ,
calcugated RCS average tmperature for the giva tauperature plateau 162.0 F in the initial te: t) . In gli cases, thgs trore stringent acceptance criterion was met (TE-413B was 1.7 F at the 0375 F plateau with the RTD bypans return valves open). For 0 both the 450 F plateau cases, the average 'IC tatpergture was more0 than 2.0 F less titan the average RTD tanperature (closed - 2.4 F, open - 2.8 F) . This was determined to go acceptable sgnce the instrumait accuracy of the thermocouple is +5gF. At the 375 P plateau with RTD bypass return valves open, T was 272 F. A review of the data showed that this may have been
- caused thTa-slight non-isothermal condition in the RCS. .Since T was in specification for the remaining six cases, this was determined n$kTto be a problem.
The value of obtaining data with the RTD manifgld return valves closed was highly questionable. Note for example the 375 F data for the loop 1 RTD's. Theo two hot Igg RTD's (-410A and ~411A) were cooler than the average by 0.4 F and 0.6 F, respectively, while the two cold leg RTD's (-410B and - 411B) were higher in temperature than the RCS average temperature. For the open-manifold case, this trend was reversed. In general, the quality of the data for the open-manifold case was better than that of the closed-manifold case. The scatter in data (indicated by large installation correction factors) appears to decrease at the higher tanperature plateaus. This could have been due to several factors, two of the nest likely being:
- 1) A higher degree of isothermal conditions were obtained at the higher tenperatures.
- 2) The RTD calculations are trore accurate at higner temperatures.
In general, the test was satisfactorily ecmpleted. The most important l accgptance criterion (individual RTD tunperatures not different by more than ! 1.7 F from the average) was satisfied for all RTD's at all tcmperat"re l plateaus for both open and closed manifold cases. Those acceptance criteria not satisfied at all points were determined to be acceptable. The test results were supplied to Westinghouse for possible determination of an l improved calibration curve for the RTD's. Canpletion of this test satisfied l -. an outstanding test discrepancy from the Preoperational Hot Functional Test L Program. i 2.0-30
. _ _ _ _ __ .-. _ _ . _ ~ . _ _ . . - - _ _ _ . _ _ - _ _ . _ _ _ _ _
O ,m O -2 O RESULTS OF SECOt0 I?fD/1C CROSS CALIBRATION TL'7P RTD No. m Calculatal Iastal14_t y p rtion Factors, F m 450'C l 557E 375F _, Valves CL _Valg _.? ' valves CL valves OP valves OP Valves CL Valves OP 0.7 -0.1 0.8 0.1 0.7 TE-410 A - 0.4 -0.2
-d.4 -0.1 I 0.0 -0.3 -0.2 -0.6 410B
-0.2 0.0 J.2 0.7 -0.1 0.8 (
-411A O.6 -0.3 0.3 0.1 -0.3 -0.1
-411B -0.1 0.1 -0.1 0.5
~
0.3 0.0 0.3
-0.2 1.1 0.0 C-413A _
9.4 0.2 0.5 0.1 0.5 0.5. l -423A 0.2 0.2 0.3
-?.3 0.0' -0.5
-433n -0.5 -0.2
-0.3 0.0
~
0.0
-443B 0.1 0.4 0.P G . .'~
U.1 -0.1 0.0
-4 30 A -0.7 -0.6 -0.2 -~-0.3 0.5 -0.8 -0.1 -0.1
-430n -0.9 -0.7 -0.7
-0.3 0.1 -9.1 -0.1
-4 31 A -0.5 -0.4 -0.1
~
-0.5 -0.8 -0.2 -0 2
-131B -1.1 -0.8 -0.8
.N -0.4 0.9 -0.1 -0.2 0.1 -0.7 0.4
? ~420A
-0.3
~~~'
O.0 -0.2
-420B MI~ -421A i -0.1 0.3
-0 .1
-9.5
-0.2 0.4 0.0
-0.3 0.8
^~
-0.1 -0.2 j 0.1 -0.2 0.0 -0.3
-421B 0,0 0.1 -0.1 _,
0.9 -0.1 0.2 0.3
-413B 1.4 1.7 0.5 0.3 0.3
' 0.1 -0.1 0.2 -0.?
-433A -0.4 -
0, / 0.0 -0.1
-440A 0.4 0.3 0.4 -0.2
-0.3
-0.2 0.1 4.3 -0.2
-440B -0.1 0.2 ~~
- 0. f'
~
0.6 0.0 0.8 0.1 I -441A U.8 0.2
-0.5 -D.6 0.4 -0.3 0.0 -
! -441B j 0.2
-0.6 -0.1 -d.3 0.6 -0.4 -0.1
-423B i 0.3 0.1 O.3 0.3
-443A 0.0 0.4
~~
0.1 0.4 Avg 170D 449.7 556.9 556.2 556.6 remp 371.4 372.1 449.4
' Steam Eat. 556.6 555.7 555.4 369.9 369.9 448.t 448.0 Trap ,
8 T=T - 1.7 0.3 0.S 1.2 TsaP 2.0 2.2 A 1.3 Ave 1C 555.4 554.4 554.2 371.4 370.7 447.0 446. Tunp T=T ~ 2.8 1.5 1.5 2.4 T, lcn# 3.5 1.4 2.4 LA ja
2,10 THERMoCoGP[E CORE SUDCOOLUG MONIER SYSTEM TEST G U The Thermocouple Core Subcoolit., Monitor Systen (TCP) consists of two trains which monitor fifty incore thermocouples, primary systen pressure and selected T ,and TC RTD's. The ECM therefore nornally monitors primary systen pres 3, tire and 2LDanperature. More importantly, the microprocessor controlled system will calculate saturation tauperature and pressure during an accident condition and will alarm when the margin to saturation is reduced 6.o a preset level and again if the core ever reaches the saturation level. The purpose of this test was to perform a preoperational type functional checkout of the TCCM. The checkout include $ verifying all thermocouples were functioning properly and the ECM IfD displays, alarms, calculations, outputs and printers were working correctly. With the plant at normal operating tanperature and pressure, normal TCM displays were verified to be functioning properly. The), in order to be able to change input parameters, normal field inputs were disconnected and a signal injection test box was connected to the TCCM. Only one train at a time was taken out of service for testing so that the other train was always operable. Using the test box, test engineers were able to inject a wide range of normal and abnormal temperature and pressures. This method tested that CCD displays gave proper outputs, calculatiot.s performed by the TCM were correct, alarms activated at the expected setpoints and verifiad that . 7 poper outputs were being sent to the pir.nt canputer, analog indicators and (d thi. TCM printers. A final check was perforned after the signal test box was urnoved to ensure all the normal field inputs were reading properly. During tlO test, Train B gave some unexpected displays when certain test signals we w injected into the systen. After troubleshooting, it was Ciscovered part of the A/D converter circuitry was out of calibration. A recalibration of Train B was pe 'vmed and all the discrepancies were clearad. , tr. ring the final verification of normal field inputs, it was discovered some of the thermocouples were not reading properly. Further investigation determined that threa 2hermocouples were defective. Since the plant must be in Mode 6 in order to repair or replace the incore thermocouples, this problem will probably not be corrected until the first refualing. However, since Techr.ical Specifications only require 4 thermocouples per core quadrant to be openble, the systen still far exceeds the minimum
- requiratents.
2.3-22
2.11 SPIrIAL 7EST PROCEDURE NR THE PRESSURIZER RELIEP VALVES The pressurizer power operated relief valves (PORV's) are solenoid actuated valves which respord to a signal from a pressure sensing system or to manual control. Motor-operated block valves are provided to isolate each power operated relief valve if excessive leakage ~ develops or if the PORV fails to close. The power-operated relief valves provide the safety-related means for_ reactor coolant syste depressurization to &:hieve cold shutdown. During the preoperational hot functional test program, the following discrepancies were noted against- the _ PORV's: BB-PCV-455A - did not reliably and consistently close when operated from an initially ambient valve body teperature BB-PCV-456A - did not reliably and consistently close when operated from an initially ambient valve body tenperature and leaked through the seat with the result that a water seal could not form in t.he inlet piping. The purpose of this test was to test the PORV's after rework and demonstrate satisfactory operation. With RCS pressure at 2235 + 15 psig and a water ~ Teal formed at the valve inlets, each PORV was testcd individually oy opening the valve in manual and allowing pressurizer pressure to decrease by 200 psi. The pressurizer heaters were deenergized during the tesc. The PORV sas then closed. RCS pressure was then restored to 2235 psig and the
~
other PORV was tested. A high speed chart recorder was used to nunitor l valve opertng ard closing times. Both BB-PCV-455A and BB-PCV-456A operated satisfactorily. The opening and closirn times were within specification ard thero was no problem with valve closure. There were also no indications of valve seat. leakage. The results are shown in Table 2.11-1. This test closed an outstanding test discrepancy from the Preoperational Hot Functional Test Program. l t i lO 2.0-33 L
- - - . . ~ . - , , - , , . ,
TABIE 2.11-1 RESULTS OF PORV OPEtiltM/CLOSItG TEST Valve Opening Acceptance Satisfactory Time Criteria C1csure After 200 psig Pressure Drop
- BB-PCV-455A 0.28 seconda 1 2 ceconds Yes BB-ICV-456A 0.4 seconds 1 2 seconds Yes
*Both BB-ICV-455A and BB-PCV-456A closed in less than 1 second. ,
l-l ll l L O 2.0-34
- 1. - - - - - . .. ,- - . ..
2.12 LOOSE PARTS MONMVRitC SYSTFM The purpose of this procedure was to obtain baseline data from the loose parts monitoring systs af ter the reactor core had been loadgd. With plant conditions at normal operating temperature and pressure (557 F, 2235 psig) ard four ICP's in operation, tape recordings of the noise from each of the twelve accelerometers (channels) was obtained. A reference signal was introduced using the installed simulator while recording channels 1 through
- 4. Decibel levels were also obtained using the installed meter for each of the twelve accelerometers.
Ch:.anel 2, which is one of two accelerometers nounted on incore thimble guide tubes at the bottom of the reactor vessel, indicated a high vibration. An alarm was present for vibration and loose parts at the loose parts nonitoring panel and the main control panel for channel 2. After a review of the data by Westinghouse, it was determined that the noise was not characteristic of a loose part. For the following reasons, the noise was suspected to be normal thinble tube vibrations:
- 1) The noise only occurs on one (1) accelerometer as opposed to both accelerometers at the bottom of the vessel.
- 2) The noise stops whm :he ICP's are turned off (lack of full flow) .
p 3) Thimble tube vibration is consistent with qualitative experience at V other Westinghouse plants with similar signals. Based on Wmtinghouse analysis of the test data and additional nunitoring of channel 2 with four ICP's in operation, the alert level alarm for channel 2 was increased from 1.813 volts to 3.0 volts. i l l I 1' l O 2.0-35
3.0 INITIAL CRITICALITY AND IIM POfER TEST SEQUENCE rm The initial criticality and low power test segment of the startup program enempassai a number of activities ranging frm bringing the reactor critical for the first time to verifying design parameters of the core. An integrated test procedure was used to accmplish these activities. It started by bringing the reactor critical inmadiately followed by determining the power range to be used for further testing. Nuclear Instrumentation Systs checkouts were also incorporated into this portion of the test. Control rod bank worths, isothermal taperature coefficients, and boron endpoints were measured for various rod configurations. The worths of the most reactive rod and a pseudo ejected rod were then determined followed by restoration of the reactor to a normal configuration. An additional rod swap test was performed to gather information to be used by the KG&E Nuclear Fuels group. This dealt with swapping shutdown bank B, whose worth was known, with the reaining control rod banks to detemine their worth 1.e., insert shutdown bank A, withdraw shutdown bank B or vise versa. D ta was gathered on site, however, the analysis was done by the fuels group as this test was conducted for information only. A combined effort of KG&E and Westinghouse personnel was utilized to cmplete this extensive sequonce of testing in a timely manner. On May 22, 1985 at 0745, the plant was brought critical. Imediately following, the zero power physics resting began. All testing within this segment was completed below a power level of 5 percent raced thermal power as allowed by the ep- ating license. Acceptance criteria used for test results were based on the core design report which was provided by Westinghouse. A su;miary of the results of the tests performed during this segment of +.he startup program follows:
- 1) Isothermal tmperature coefficients at CBC and CED inserted, CBD inserted and thg all-rods-out configurations wert, measured to be within 1.5 pc:t/'F of t criteria of 13.0 pcn/geF.expected values A positive thus meeting nederator the acceptance tmperature
- coefficient was calculatai for the all rods out configuration.
l Consequently, rod withdrawal limits were developed for use during the ! first cycle, l- 2) Control bank worths for control banks A through D were measured to be l within 4.0% cf the predicted values, well within the acceptance criterion of 110%,
- 3) Shutdown Bank worths for shutdown banks C through E were measured to be within 2.6% of the predicted values, well within the acceptance criterion of 110%,
- 4) Total rod worth was measured to be within 5.4% of the predicted value meeting the 110% acceptance criterion,
- 5) Critical boron concentrations were measured for six different control rod configurations. With the exception of the all-rods-out configuration (ARO), all concentrations met the acceptance criterion 3.3-1
. . - . . - . .- . ..~ _ ...
. .. _ . . . _ . . . - . - .- .- .- .. ~.
- t; of +10% of the predicted values. An evaluation of the all-rods-out case n' yieToed no-impact on the safety analysis from the critical boron A); concentration being 3 ppm out of tolerance high,
- 6) The differential boron worth was measured to be within 0.19 pen / ppm of the predicted value well within the +10% acceptance criterion, 7)~ core power distributions determined using flux maps were acceptable for the configurations of all-rods-out, CBD-in,-hot-zero power insertion limit,. arx1 pseudo ejected rod.
L 3.0-2
~ - . - - - - -
3.1 INITIAL CRITICEITY 0:
-In preparation for bringing the reactor critical, the Nuclear Instrumentation Systs source range channels N31 and N32 were verified operational and, within twelve hours of criticality, analog channel operational test surveillance procedures were performed on each of the intermediate and power range channels. Reference counts were obtained for the source range channels using a 132 second interval and ten seperate counting periods for use in ICRR nonitoring. The reference counts were 1181 counts for channel N31 anr31300 counts for channel N32.
The initial approach to criticality began at 1032 on May 21, 1985 at which time the reactor coolant systs (RCS) boron concentration was 2041 ppm and - all control and shutdown banks were fully inserted. Beginning with Bank A, the shutdown banks were withdrawn in 50 step increments stopping to obtain counts for use in plotting inver*a count rate ratios (ICRR's) . These ICRR plots for rod withdrawal verifiai the core would not be critical with the next 50 step withdrawal and are illustrated in Figures 3.1-1 and 3.1-2. Rod withdrawal continued with the control banks until control bank D was at 160 steps. Dilution to criticality began at 2040 on May 21, 1985 with a dilution rate of 60 gallons per minute. Boron conce.1tration in the RCS was determined every 20 minutes. Plots of inverse count rate ratio versus time of
-dilution, RCS boron concentration, and makeup water addition were made during the approach to criticality in order to predict criticality. These A() plots are shown in Figures 3.1-3 through 3.1-8. The dilution rate was changed to 30 gallons per minute at 0650 on May 22 and criticality was I- achieved at 0745 with boron concentration of 1343. Control bank D was then
- _ .used to maintain the reactor just critical.
l l Just after the reactor went critical, readings were taken to determine overlap between the NIS source and intermediate range channels, hhen l intp* mediate range channels showed a positive indication, i.e, greater than
~
10 amps,-both source and intermediate range indications were recorded. Theindicagonswereagainrecordedwhentheintermediaterangeindication showed 10- amps. Itwasnotpossibletogetoverlapregdingsatany higherlevelsincethesourcerangereactortripisag10 counts oer second even though the source range scale indicates up to 10 counts per second. The overlap data is shown in Section 4.4 of the report and shows overlap l ' between the source and intermediate range is greater than the required 1 1/2 1 decades. With the reat ,e stabilized and just critical, the range of core power for physics testing was determined. This was accomplished'using a reactivity computer, supplied by Westinghouse, with an input signal coming from NIS power range channel N-42. Control bank D was withdrawn thereby increasing l the flux level until the effects of nuclear heating were observed (i.e. incre E 5.2 x 10 9se_ in RCS
- amps average on the reactivity tenperature) comeuter. picoameter, This occurred at a and 8 power x 10 -71evel amps on both NIS intermed! ate range channels N35 and N36. The te declared to be 1/10 to 1/100 of these values or 5.2 x 10'gting to range wag 5.2 x 10~
amps on the reactivity computer. 3.0-3
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After the testing range was determined, a reactivity computer checkout'was (U].- .mrformed. Positive. and negative reactivity insertions were-introduced by
' control rod movement with calculations of the change in reactivity made using the neutron flux doubling time. Comparing these calculated values to theoretical values developed from the In-Hour equation, resulted in an average difference of 0.7%. Reactivity changes of approximately 25, 50 and 75 pcm were used for the calculations.
- Only one significant problen was encountered during the approach to criticality. Water had been pumped from the spent' fuel pool into the refueling water storage tank (RWST) . Rcd withdrawal was suspended until the boron concentration of the RWST could be verified to be within Technical Specification requirements. Af ter a 45 minute delay, verification was made -
and rod withdrawal resumed. O 3.C-12
3.2 CONTROL IKX) HANK WORTil MFASURIMNPS 01 Control rod bank reactivity worths were measured by monitoring reactivity changes associated with RCS boron and control rod bank exchanges. Differential reactivity worths, being the ratio of the change in reactivity to the corresponding change in bank position, and integral reactivity worths, being the total reactivity change due to the travel of the entire rod bank height, were obtainoi via these exchanges. After establishing a constant ICS boron dilution or boration rate, the control rod banks were pariodically inserto3 or withdrawn to compensate for the changing boron concentration. The changes in reactivity dae to control rod bank movement were indicated on a strip chart recorder connecte3 to the reactivity computer. From an all-rods-out starting condition, boron endpoints and individual control rod bank reactivity worths for control ' banks A through D and shutdown banks C through E were obtained by RCS dilution. The worth of the most reactive rod, rod F-10, was then determined by withdrawing it and compennating with the insertion of shutdown banks A and B. When rod F-10 was-fut1 out, insertien of shutdown banks A and B continued by RCS dilution until-the banks were fully inserted. The reactor, which had previously been tripped to realign rod F-10, was then brought critical with shutdown banks withdrawn arf 3 the Control banks inserted. Initiating RCS boration,-the control banks were withdrawn in G order to measure their worth in the overlap mode. Upon conpletion, the rods U were repositioned at the hot zero power insertion limit. By FCS boration, rod D-12 was thm withdrawn to simulate an ejected rod and its reactivity worth measured. Rod D-12 was then realigned and the reactor manually trippe3 as this was the end of rod worth measurenents. A sumnary of the results of the rod worth measurenents is presented in Table 3.2-1. The differential and integral reactivity worths of all cases have been plotted in Figures 3.2-1 through 3.2-10. As indicatedt in Table 3.2-1, all measured rod worths were within the acceptance criteria. One significant problem experienced during rod worth measuranents was that at one point, individual rod 9toups within a rod bank became misaligned when switching back and forth from bank to bank on the selector switch. The largest misailgnment encountered was four steps. Upon completion of rod worth testing, the reactor was trippe3 and the step counters reset in order to realign the rods. The only other problen encountered was during the worth measurements of the most reactive rod, F-
- 10. Technical Specification 4.10.1.2 requires that prior to this event, the rods must be tripped from the 50% withdrawn posicion to show insertion capability. When withdrawing control bank C to demonstrate this, a counts
- doubling occurred causing automatic boration of the RCS. After terminating the baration, rod withdrawal continued slowly in anticipation of possible criticality. The reactor did go critical prior to reaching the 50% pmsition but was quickly brought nubcritical by rod insertion. Following an A evaluation, which determined that control Bank C was worth nore than U predicted in this configuration (all other rods inserted), the RCS was borated conservatively and testing proceeded without any further difficulties.
3.0-13
.. ~ . .. .- ~~ - , . , . - . . - - . . . _ . ~ . . . . . . . .-
TABLE 3.2-1 i - WOLF CREEK GENERATIMG STATION CYCLE 1 BOL-PHYSICS TEST.
. CONTROL'_ ROD LWNK-WORTH SUttiARY Bank / Rod ' Measurai Worth Acceptance Criteria
. Configuration * -(pcm) (pcm)
CBD 650.4 650 + 65 CDC (CBD 0 0) 1194.3 1240 + 124 CUB (CBD,CDC @ 0) 1910 970 + 97 CBA (CBD,CBC,CBB@0) 658 680 + 68 SDE _ (CB @ 0) 846.9 870 + 87 SDD (CD,f-SDE @ 0)- -758 740 + 74 SDC - (CB,SDO SDD@0) . 954.7 960 + 96 ARI - I 6322.5 6680 + 668 Ejecto3 Hod D-12 548.5 < 860
*CB = Control banks SD u Shutdown banks ARI = AI rods in l'
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i I 3.3 IS0WFRMM. T!MPERAWRE COEFFICIERI' O The isothermal temperatore coef ficient measurments we5e *PtiSh*d "ei"7 a constant heatup or cooldown rate'of approximately 10 F ger hour. Witg - reactor power within the physics testigg range, 5.2 x 10' to 5.2 x 10 amperes, a cooldawn of approximately 5 F was made by adjustment of the steam dump system. Reactivity changes during the cooldown were recorded by the reactivity empgter. This process was then repeated for a heatup of approximately 5 F. Pressurizer level was maintained steady or slightly increasing during the process in order to eliminate boron reactivity etfects due to outflow from the pressurizer. Isothermal temperature coefficient measurements were performed at three different control rod configurations; all-rods-out; control bank 0-in; and control banks D and C-in. A heatup and cooldown was performed for each configuration. The isothermal tmperaturo coefficient was taken to be the average of the values of the slopes of the heatup and cooldown plots from the reactivity emputer. The results of the isothermal taperature coefficient measurements were all within the acceptance criteria as presented in Table 3.3-1. Using the value of the isothermal temperature coefficient, the mxlerator temnerature coefficient - (M'IC) was calculated for the- all-rods-out configuration. This is simply the isothermal temperature coefficient minus the Doppler (fuel) tmperature coefficient. The M'IC is required to be negative (i.e. as O temreretere tacre see "es tive reectiviev i= tatt eocea) nowever the results of the calculation was a positive 1.03 pc Y F.* This required rod withdrawal limits to be instated which would preclude operation of the plant with a positive moderator tmperature coef ficient. These limits are illustrated _ in Figure 3.3-1. No major probles were encountered during the isothermal temperature coef ficient measurments and with rod withdrawal limits in place, all test - results were satisfactory. l l-i I i 1 I O-- 3.0-25 l l l
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SUMMARY
Rod / Bank MeasuredOValue Acceptance Criteria Configuration (pca/ F) (pcm F): ARO -0.92 -2.03 + 3.0 CBD.9 0 -2.05 -3.36 + 3.6 CBD, CBC 0 C ,
-5.47 -6.66 + 3.0 -
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3.4 BORON ENDPOUR AND BORON WOR'!H MEASURD4DCS Boron endpoints were measured in an integrated test procedure wttich also measured the control bank worths. With the reactor at hot rero power, RCS boron concentrations were measured at several rod bank configurations for cmparison to design predictions. The RCS conditions were stabilized with the controlling rod bank at the desired endpoint position. Critical boron concentrations were then measured for the endpoints. Results of the mdpoint measuranents are presented in Table 3.4-1. With the exception of the all-rods-out configuration endpoint, all measured endpoints were within 6.2 percent of the design predictions. In the all-rods-out case, the measured concentration was 3 ppm outside the allowable tolerance. T'le situation was evaluated by Westinghouse with the result that no impact on the safety analysis was presented, therefore the measured endpoint was acceptable. A differential boron worth was calculated using values of measured boron concentratiors and rod worths. Dividing the total worth of the control banks (in overlap) by the difference of the critical boron concentrations fran the all-rods-cut to the control banks in configuration resulted in a differential boron worth of -10.08 pen / ppm, well within the + 10% of design prediction tolerance as indicated in Table -3.4-2. h O
.3.0-28
_ ._ . - _ - _. _ . . . _ _._m _ . TABLE 3.4-1 DORON ENDPOINT SUM %RY Control Rod Configuration Measurei Value (ppm) Acceptance Criteria (ppn)
-l ARO 1352.2 1299 + 50
._.- CBD @ 0 (ARO-CBD) 65.3* 63 1 6 CBD, CT @ 0 126.9* 123 + 12~
- (CBD-CBC) -
. CBD, CBC, ~ CBB @ 0 87.9* 93 1 9 (CBC-CBB) CBD, CBC, CBB, CBA @ 0 69.3* 65 + 7
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- Values are the dif.ference be W m successive endpoint measurenents.
O 3.0-29
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. DIFFERFRfIAL POROti WORTl! SUFNARY r Differential Boron Worth = Measured Value Acceptance Criteria (ocm/ cpm) (ren/pom) Control Bank Worth (overlap) -10.08 -10.27 + 1.03 - Cg Wo)-g pA@0)_ D N / 3.0-30
3.5 POWER DISTRIBUTION MFASUREMENPS p-d The core power distributions were measured using the incore flux mapping systen. Data from this systen was then analyzed using the Westinghouse INCORE computer program. Four flux mLps were obtsined during the low power physics testing segmmt of the startup program. The maps were takm at different control rod bank configurations which were all-rods-out, control bank D at 0 steps, hot zero power insertion limits, and a pseudo-ejected rod. Incore power tilts, reaction rates, and hot channel factors were reviewed for the flux maps all of which resulted in acceptable values. Results of the maps along with acceptable values are given in Table 3.5-1. Analysis of the first flux maps indicated abnormalities between thimbles R-08 and N-08. After a detailed examination of plant documentation, it was determine 3 that the two thimbles had been interchanged at the seal table. The problen will be remedied during the first refueling by appropriate repositioning of the thimbles at the transfer device. .The flux map data is currently being corrected for each map by manually interchanging the data for thimbles R-08 and N-08. Several equipment problems hampered the running of the flux maps. In most cases, not all of the detector drives were available for operation. However, constant maintenance and the redundant design of the systan allowed the flux maps to be obtained although in a less than timely manner. l l l t u l l I l l 3.0-31
O O .O TABLE 3.5-1 l IOdER DISTRIIRTI* ION
SUMMARY
l riap Control Rod Incore Tilt Reaction Rate Enthalpy Rise flot Axial Offset Nunher Configuration Error Channel Factor Fg(Z) F A liN Accm table Measural vreptable Measured Acceptable Measurel Acceptable Measural Measured Acuptablo l 1.0157 <1.04 -6.0% < +10% 1.4305 1.3910.14 2.3467 N/A -1.788 N/A CIM002 ARO t CIM001 CBD @ 0 1.0125 <1.04 8.8% < +101 1.5792 1.5710.16 2.7702 N/A -16.577 N/A 1.0174 <1.04 -9.5% < +10% 1.5775 N/A 2.7654 N/A -36.928 N/A CIM003 . lot zero power ~ -~ insertion limit 1.9513 N/A +50.8 N/A 3.8293 N/A 6.6099 <7.03 -33.363 N/A ClM004 Pseudo rod Eiection u O b
4.0 POWER ASCENSION TESTIlG O , 1 Following the completion of low power physics testing, the power ascension ! phase of the startup test program was begun. The testing in this phase of ' the test program included various at-power physics tests, control syst e dynamic response tests, overall transient and trip testing of the plant, and calibration and alignment of plant instrumentation and control syst as. Additional testing included a steam generator moisture carryover test, the NSSS acceptance test, thermal and dynamic testing of the main steam and main feedwater systas, biological shield surveys and turbine generator testing. The tac lif ted the 5% power restriction on June 4,1985. The low power physics test sequence was approved by the PSRC and the Plant Manager authorized the start of power ascension testing on June 5, 1985. The turbine generator was synchronized to the grid on June 13, 1985. The initial phase of power ascension testing was approved by the PSRC and, after Plant Manager authorization, power was increased to 30% on June 19, 1985. The plant was shutdown from outside the control room on June 29, 1985. Dower was increased to 50% on July 6,1985. The rods drop and plant trip test was performed on July 16, 1985, and testing at 50% power was completed on July 18, 1985. After PSPC re/lew and approval of the test procedures, the Plant Manager authorized the start of 75% power testing on July 19, 1985. Testing at 75% power was started on July 20, 1985 and completed on July 29, O 1985- rae esrc reviewea es eer vea ' 75' e wet tese et cea ree "a the Plant Manager authorized the start of tne 90% power test sequence on July 30, 1985. The plant was at 901 power on August 4,1985, testing was completed on August 6,1985 and, af ter PSRC approval of the test procedures, the Plant Manager authorized the start of 100% power testing on August 9, 1985. , The plant was initially brought to 100% power at 1607 on August 8,1985. A 100 hour continuous run at 100% power was completed at 2007 on August 12, 1985. Power was then decreased to 55% power while vibration probleer with the "B" main feedwater pump were investigata'. The plant was returned to 100% power on August 21, 1985. The 100% power plant trip test was successfully performed on August 23, 1985, and after a checkout of the NIS system to verify that the systs had not degraded during 100% power operation, the 100% power test sequence was completed. The reaining test packages were approved by the PSRC on August 30, 1985. The unit was declared commercial at 0114, September 3, 1985. O 4.0-1
4.1 AT tuER PHYSICS TESTUC O The At-Power Physics Tests are a series of tests to verify accuracy of the physics models use3 in core design and accident analyses, to verify that hot channel factors and control rod worths in a rod ejection accident are conservative, and to obtain calibration data for the Nuclear Instrumentation System (NIS) . A sumary of the physics tests performed during pwer ascension follow:
- 1) Incore Movable Detector and Thermocouple Mapping at Power - Flux maps were taken at power levels of 30%, SM , 75%, 9M , and 100% full power. The hot channel factors were measured to be within WCCS Technical Specifications. All measurement parameters met their design and accident analysis acceptance criteria after clarification by Westinghouse,
- 2) Axial Flux Difference Instrumentation - Flux map results were used to calibrate the Nuclear Instrumentatior: Systen to axial power distribution in each section of the core. Resetting of the NIS circuitry was accomplished to duplicate certain key flux map parameter results,
- 3. Power Coefficient Determination - A comparison of the Doppler only Power Verification Factor showed the measured value was well within the design value tolerance of +5%,
O 4. Pseudo Rod Ejection Test - With control rod D-12 pulled to 223 steps, core power distributions associated with a flux map taken at that time indicated hot channel factors were within the limits of WCGS Technical Specifications and the partial worth of the ejected rod was within the value for design tolerances. O 4.0-2 l
'4.1.1 INCORE MOVABLE DETFL'rOR MAPPI?G AT POW'ER O
The core power distributions were measured during the power ascension portion of the startup testing program using the incore movable detector flux mapping system. The results of the flux maps are give in Table 4.1.1-
- 1. The flux map taken with an ejected rod will be discussed in Section 4.1.4.
The flux maps taken and listed in Table 4.1.1--I were taken over a-range of 30% to 100% inwer at various control rod configurations. The flux maps were taken to provide results to verify the accuracy of the physics tredels used in the core design, to verify operation of the reactor within iCGS Technical Specification power distribution requirements during typical operation, to obtain calibration data for the Nuclear Instrumentation Systan (NIS), and to obtain baseline data for the target axial flux distribution surveillance. As shown in Table 4.1.1-1, normally 58 thimbles are used for a full flux map. Several flux maps taken during the incore/excore detector calibration used only 16 thimbles because it was only desired to monitor the axial offset. The measured power distribution parameters are comcared with tCGS Technical Specification limits in Table 4.1.1-1. The power distribution for all flux maps met their design, accident, and ICGS Technical Specification limits. The predicted vs. measured axial powers were within the 10% value as required by design when clarification was obtained from Westinghouse (~) indicating that this was only applicable for the larger unrodded core region V during normal operation. All power distribution measurenent results were acceptable when c~npared with the design, accident analysis, and Technical Specification ilmits. The core exhibits a small natural tilt to quadrant II, NIS channel N43. The small natural tilt does not exceed the tCGS Technical Specification limits. l i O 4.0-3
O O 4 1 1-1 O ' VCGS ItCORE FLUX t%P SUf1MRY DURItC POWER ASCENSION Map No. Date IP Core Rod Fn F 3y F y., No. Thimbles B/O Pos. "
~~' ~
Used W D/k m Max. Loc. Limit Max. Loc. Limit Max. AX. Pr. Limit ClM005* 58 6/27/85 30 134 !!FPRIL 2.2070 D3 4.640 1.4216 .D7 1.6985 1.5662 9 1.950 CIM006* 58 6/27/85 30 134 IIFPRIL 2.2384 E14 4.640 1.5304 E14 1.6986 1.8390 9 1.9494 D12@22fi ! ClM007 58 7/9/85 50 229 D@202 2.0976 D9 4.640 1.42211 D9 1.639 1.4515 39 1.705 ClM008 58 7/9/85 50 280 00200 2.1117 D9 4.640 1.4246 D9 1.630 1.6186 9 1.705 ClM009 58 7/21/85 68 370 D@202 2.0086 D12 3.412 1. 346fl D9 L.5854 1.4752 39 1.649 ClM010 58 7/26/85 75 523 D@210 1.9891 D9 3.093 1.353H D9 1.5645 1.3665 39 1.6275 CIM012# 16 7/26/85 75 533 D@l52 2.2292 lill 3.093 1.3950 till 1.5645 1.5242 9 1.7955 CIM013 58 7/27/85 75 533 D@l58 2.2523 D9 3.093 1. 379il D9 1.5645 1.5307 9 1.7955 ,, ClM019# 16 7/27/85 75 533 D@210 1.9932 1111 3.093 1.3491 till 1.5645 1.3585 36 1.6275 ClM020 58 7/27/85 75 537 D@210 2.0373 D4 3.093 1.3635 D4 1.5645 1.3721 17 1.6275 ClM321 58 8/6/85 92 737 D@214 2.0360 D9 2.522 1.3634 D9 1.5138 1.3670 23 1.5743 CIM023 58 8/12/85 100 932 Da215 2.0313 11 5 2.320 1.3621 D9 1.49 1.3663 16 1.55 CIM024 58 8/22/85 100 1192 D@218 ,2.025U IIS 2.320 1.3632 115 1.49 1. 370 J 16 1.55
'?
Radial Tilt Incore AO ('t) Map AO No. N41 N42 N43 N44 N41 N42_ N43 N44 CIM005* -12.214 0.9903 1.0137 1.0134_ 0.9826 -11.956 -12.501 -12.106 -12.240 CIM006* -6.014 0.9461 1.0935 0.9945 0.9659 -8.484 -0.468 -7.571 -7.535 CIM007 -4.269 0.9896 1.0172 1.0128 0.9805 -4.382 -3.907 -4.618 -4.171 ClM008 -5.410 0.9894 1.0159 1.0130 0.9817 -4.497 -6.001 -5.299 -5.844 '3 CIM009 -7.762 0.9890 1.0105 1.0100 0.9905 -7.601 -8.084 -7. SUS -7.458 CIM010 ~6.468 0.9915 1.0084 1.0106 0.9895 -6.379 -6.629 -6.577 -6.287 CIM012# -21.813 1.000 1.000 1.000 1.000 -21.813 -21.813 -21.813 -21.813 ClM013 -25.133 0.9915 1.0073 1.0120 0.9892 -25.013 -25.406 -25.165 -24.919 4 ClM019# -1.171 1.000 1.000 1.000 1.000 -1.171 -1.171 -1.171 -1.171 ClM020 2. 389 0.9907 1.0084 1.0111 0.9898 2.493 2.381 2.349 2.334 ClM021 -10.035 0.9937 1.0082 1.0106 0.9875 -9.863 -10.049 -10.244 -9.985 CIM323 -10.962 U.9946 1.0065 1.0109 0.9881 -10.767 -10.996 -11.098 -10.986 CIM024 -9.926 0.9945 1.0060 1.0100 0.9895 -9.719 -9.924 -10.102 -9.958
- Flux Maps for ejecto3 rod verification.
#0uarter core flux maps.
tJOTE: tbp's l's CIM0ll, CIM014-CIM018 were quarter core maps taken during the incore-excore calibration arx1 not usel. ClM022 is non-existent. m
4.1.2 AXIAL FLUX DIFFERENCE INSTRUMErfrATION CALIBRATION O This test consisted of three sections:
- 1) A preliminary Incore-Excore calibration after the trip at 50% Rated Thermal Power an:1 prior to escalation above 50% power. It was done using the incore flux maps taken at 30% and 50% power,
- 2) The Incore-Excore calibration at 75% power using a seriea of S11 core and quarter core flux maps taken at the 75% power level,
- 3) A " fine tuning" of the Incore-Excore calibration at 100% power.
The Axial Flux Difference Instrumentation Calibration is based on a number of flux maps taken at various axial offsets (AO) to determine a correlatica between Incore and Excore detectors (A0 is defined as difference of % power in top half of core an bottm half of core divided by the sum of % power in top and bottom of core). Plots of Incore A0 vs. Excore A0 were generated for each excore channel. The slope of a least squares straight line drawn through the Ao points yields the correction factor to correlate the NIS excore detectors to the measured ircore power distribution. The correction factor .is equal to the least squares straight line slope and is shown in Table 4.1.2-1 for each channel. The correction factor is entered into the plant computer at the computer addresses shown in Table 4.1.2-1. (~ Another plot of excore (NIS channel) top and bottcm detector current vs.
\
incore A0 is used to predict 100% power current values from each detector. 100% power currents are used because at that power level, delta q (defined as % power in top half of core minus % power in bottom half of core) is equal to AO and subsequent calculations can be minimized. The resulting . least squares fit straight line is used to derive expected current values which are input into the NIS circuitry to check control board ard emputer values for consistency at specific AO values. Figure 4.1.2-2 shows 100% power current values input into the NIS circuitry to calibrate the execre system to the incore measurement.
~
'Ibe preliminary calibrat'en at 50% power indicated that delta q values from the plant emputer did not favorably car. pare with predicted values. A closer analysis showed the computer was using an average power from all NIS channels instead of only the channel being tested at this point in time, as
,. assumed in calculating the predicted value. The predicted values were l corrected to actual NIS indications and close agreement between predicted
- and' actual computer values resulted. Since this was only a preliminary l calibration, it was decided power escalation to 75% could be safely achieved, and a trore accurate Axial Flux Calibration done at that power l
level. A series of flux maps were taken at various A0 values for the Axial Flux Calibration at 75% power by inducing an axial xenon oscillation in the reactor core. The flux maps used wre maps ClM010, CLM012, CLM013, ClM019, , ard ClM020 shown in Table 4.1.1-1. These maps allowed for use of five l points to-obtain the least squares straight- line values to perform an
- l. accurate Axial Flux Calibration. Inputting the 100% power current values into the MIS circuitry resulted in a discrepancy between predicted and 4.0-5 ,
,m , , . , . , , r-, .
. _ . _ . . . . . . . - - ~
. - - ~ . . _
. - . . . . . . . - . . - . . . . . . . . . . . _. . . . . . . . . - ~ . . - - .
~
> TABLE 4.1.2-1 Ny J-INCORE/EXCORE CORRDMION FACTOR M^H LEVIL Correction Factor K0554 K0552 KJ551 K0553 (N41) (N42) (N43) (N44)
' 50% 1.350 2.192 1.602 1.525 75% 1.8188 1,8418 1.7949 1.9342-100% 1.8188 1.8418 1,7949 'l.9342 -
P OL 4 I l-i l O 4.0-6 i 1 _ . . , - , , _ _ , - , , . __-..,_.---m._.__ -- -._- _ _ . - . . - . . - . . . . - ,
t TABLE 4.1.2-2
~
100% NIS CURRENT VALUES Power- A.O. NIS Channel Currents (Ha) Plateau (%) (%) 41 42 ' 43 44 Top Bottom Top Bottan Top Bottom Top Bottom 50 -30 213.0 329.5 231.3 327.3 207.2 329.3 229.9 341.1
-75 135.8 430.5 164.2 375.4 117.2 395.4 168.7 430.2
-35 204.4 340.9 223.8 332.6 197.2 336.6 223.1 351.0 0 264.5 262.2 276.0 295.2 267.2 285.2 270.7 281.7
+7 - 276.5 246.7 286.4 287.7 - 281.2 274.9 280.2 267.8
+75- 393.2 93.9 387.7 214.9 417.2 174.9 372.7 133.2
+30 316.0 194.9 320.7 263.1 327.2 241.1 311.5 222.3 75- -30 226.3 316.1 239.7 341.7 . 22f f 339.2 236.1 328.4 154.3 378.1 167.6 411.3 22 .2 411.2 163.8 389.1
-35 218.3 323.0 231.7 349.5 218.8 347.2- 228.3 335.1 0- 274.4 .274.7 287.9 295.4 272.8 291.2 283.1 287.9
+7 285.6 265.0 299.1 284.6 283.6 280.0 294.0 278.4
+75 394.4 171.2 408.1 179.5 388.5 171.1 400.3 186.6
+30 322.4 233.3 336.0 249.0 319.1 243.1 G30.0 ~
-247.4 100% 0 274.7 269.6 284,8 287.4 272.5 285.6 281.4 282.4
- O O
4.0-7
l measured values from the delta I penalty generator and certain process canputer va)ues for delta q. A new gain for th? delta I penalty generator O was esiculated using voltage signals to prevent otherwise required gain changes each time an Axial Flux Calibration was done. The newly calculated gains produced the expected function generator output thereby . satisfying the acceptance criteria. The delta I penalty function generator gain values for each NIS channel are shown in Table 4.1.2-3. The process computer flagge3 all values with A0 +75% as unreliable. This was understandable since the computer was scaled ~for voltages of +10 to -10 volts and +75% A0 values would correspoM to voltage magnitudes greater than 10 voTts. With the preceding discrepancies explained and/or corrected, the Axial Flux Calibration was satisfactorily ccrnpleted at 75% power and power escalation to 100% was allowable for Axial Flux Calibration verification curposes. At 100% powe.r, a flux map was taken (map ClM023 as shown on Table 4.1.1-1) to verify the Axial Flux Calibration values input at 75% power were 1) correct, or 2) needed " fine tuning". It was determined a " fine tuning" of the NIS was needed to give the required accuracy between the control board delta I meters, process canputer delta q, and flux map (Incore) A0 measurenent. A one point correction for detector currents was done to give the required identical values, within acceptable tolerances, for process computer delta q, and control room delta I meter values (ar 100% cower percent delta I = percent delta q = AO) . With the " fine tuning" completed another flux map was taken at 100% power to verify that the " fine tuning" gave successful correlation between control room delta I indication, process computer delta q output, and flux map (Incore) A0 results. The results shown in Table 4.1.2-4 indicate the Axial O Flux Calibration at 100% power was within the required accuracy. The Axial Flux Difference Instrumentation Calibration showed improved accuracy each time it was checked from 50% power to 100% power. Eacn adjustment to the NIS circuitry brought the incore/ccmputer/ control room meters into closer agreenent until the final adjustment at 100% power verified incore delta q vs. control room meter delta q and cceputer delta q was within the design acceptance criteria of 1.5% and tne control room meter delta q vs. canputer delta q was within the design acceptance criteria of
+0.5%.
O 4.0-8
.. . - - -. . .-. -. . . ~ , - . - - ~ . . .-
' d j.
_. TABLE'4.1.2-3 GAIN VAIIUES FOR DELTA I FUNCTION GENERATOR - l l
%P N41 N42 N43 N44 50 1.661 2.480 1.978 2.073 75 1.0 1.0 1.0 1.O !
100 1.832 1.845 1.800 1.965
\
O L l 4.0-9
O = o 0 .2-4 DELTA. g VAUJES AT SPk_/IED POWER PIATFAUS . O N41 N42 N43-Power Meter / Comp Meter / Comp Mete:/ Comp Plateau Meas. Meter Comp. Error' Meas. Meter Comp. Error Meas. Meter Comp Error
+30 +30 +18.4 11.6- +30 +30 +15.3 14.7 +30 +30 +17.4 12.6 i 50% +7 67 +4.4 2.6 +7 +7 +3.8 3.2 +7 +7 +4.0 3.0 0 0 * -
0 0 0.1 -0.1- 0 -0 * -
-30 -30 -19.3 -10.7 -30 -30 -16.4 -13.6 -30 -30 -12.1 -17.9
+30 +30 +22.3 7.7 +30 +30 +22.2 7.8 +30 + 30 +21.7 8.3-t7 751 +7 +5.3 1.7 +7 +7 +5.2 1.8 +7 +7 +5.1 1.9 0 0 0 0 0 0 0 0 0 0 0 0
- i
-30 -30 -22.4 -7.6 -30 -30 -21.2 -8.8 -30 -30 -22.2 -7.8 100% -10.8 -9.0 -9.0 -0 -11.0 -9.0 -9.4 -0.4 -11.1 -9.0 -10.3 -1.3
-9.7 -9.8 -10.0 -0.2 -9.9 -10.1 -10.0 0.1 -10.1 -9.8 -10.1 -0.3- !
Acreptance Error 0.6 - 0.6 - 0.6 y !
?
~
c$ N44 Max. Power Meter / Comp Meas./ Comp error Plateau Meas. Meter Conp. Error Ch. flo.
+30 +30 +14.9 15.1 15.1 , N44
- Computer flatyJel as unreliatle 50% -47 +7 +3.4 3.6 3.6 N44 0 0 0 0 - -
-30 -30 -14.0 -16.0 -17.9 N43
+30 430 +22.2 7.8 8.3 N43 75% +7 +7 45.3 1.7 1.9 N43 0 0 -11.9 11.9 -11.9 N44
-30 -30 -22.6 -7.4 -8.8 N42 -
I 100% -11.0 -11.0 -9.0 -2.0 -2.0 - N44
-10.0 -11.1 -10.7 0.4 0.7 N44 3 Acceptable Error 0.6 1.5 -
l B b
. - - . .- - - ._ = _ - . -- - .- - - . - . - . - . .
4.1.3 00NER COEFFICIENT DETEIC11 NATION O Power coefficient measurement was done during the power ascension progr:m to verify values used in the nuclear design and accident analysis prediction for the Doppler Only Power Coefficient. The measurement values were achieved at powc plateaus of 30, 50, 75, aM 90% power by causing a series of small load swings at the turbine-generator. These series of load swings resulted in corresponding temperature swings in the Reactor Coolant System, but of opposite sign to the turbine generator load swing. The Reactor Coolant Systen tcmperature change for each turbine-generator load change was recorded and used in a subsequent calculation to determine a Doppler Coefficient Verifiqation Factor. The measured Doppler Coefficient Verification Factor (C") is the ratio of the change in core average tenperature and the change in power due to the doppler effect. This ratio is equivalent to the predicted ratio of the Doppler(C coefficient ong) . Power Coefficient Since for and theofisothermal small changes temperature, temperature the void coefficient portion of the power coefficient is insignificant, the ratio is also proportional to the power coefficient. The predicted value of the verification factor was calculated using the design values for the Doppler only and isothermal temperature coefficients. The measured value of the Doppler Coefficient Verification Factor was compared with the predicted value. The results of the comparison are shown in Table 4.1.3-1. The results show the values used for the nuclear desian .O end eccident eaetvsie eredictica ere very c1ose to these meeeere3 end thee the measured values are well within the required acceptance criteria. O 4.3-11
.. .. . . . . . . . - - ~ . , - . . . - . - . . - .. . . . . - . . . . - .
5" TABLE 4.1.3-1 L- i
- DOPPLER COEFFICIEW VERIFICATION FACTORS
~
Power CM C P C'M+CP -Acceptable Level . (%)- C" + CP t (UF/%) ( F/%) 30 -2.35 2.34 0.01 < 0.5
- 50 -1.90 1.89 0.01 < 0.5 75 -1. ~.6 1.21 0.05 < 0.5 90 -1.087 -1.1P 0.093 < 0. 5 -
i.
~
O . r \ i O. 4.0-12 _ _,,_,___-.-..__....a._..-..__..._ . . . -- __._____.-._..._.-_.-.._;,_ .:._,,._.--_~
-- ..~ - -- - . - _ - . - . - .. -. . - - - - - -- -
4.1.4 PSEUDO ROD F.71CTION TEST
.. O This test is used to verify that the values in the design and accident analysis for_ ejection of the most reactive rod from the Hot Full Power Rod Insertion Limit (Hf P RIL) are conservative.
Two main parameters are measured by this test for comparison with design values: 1) the positive reactivity worth added from the HFP RIL to the full out position for control rod D-12 (designated as most reactive rod) and 2) the core power peaking factors resulting from ejection of rod D-12 from the core with all other control rods at-the HFP RIL. The_ worth of control rod D-12 was measured by pulling only rod D-12 from the. HFP RIL to the full out position and measuring the resulting change in l moderator / coolant temperature, Tavg. The mathematical product of the isothermal tmperature coefficient and tenperature change results in the reactivity worth of rod D-12 from the HFP RIL to 228 steps withdrawn. The
-results and acceptance criteria for D-12 rod worth are shown in Table 4.1.4-1.
Following D-12 rod worth determination, rod D-12 was reinserted to the HFP RIL and the calculated worth was equated to boron worth. Rod D-12 was
. subsequently borated to the full out position with all other rods reaining at the HFP RIL and a full core flux map was obtained. Pertinent core 3 peaking-factor results from analysis of a flux map before and after rod D-12 ;
O eseotico, ere enown ta redte 4 1 4 !! 4 O 4.3-13 _~_ ~ . _ _ _ _ . _ . _ _ _
JAJ: + s.- --a,,u.. e 4-A_.a,A-4,y._=.4_4ehah.d.4.4 4 SJ-4..,. a 5 ts%.4 %_ ,, m 4 TABLE 4.1.4-1 D-12 ROD WORTH-FROM IEP RIL
%P- 30 delta red oosition (steos) 67 reactivity change (ocm) 17.04 1.13 times reactivity change (ren) 18.74 acceotable reactivity chance < 230 L
1 4 C 4.0-14
4.. TABLE 4.1.4 - LO~ FLUX MAP RESULTS PROM ROD D-12 FJD: TION Prior to D-12 ejection D-12 ejected Core Parameters Measured. Limit Measured Limit
'F-delta H Nuclear 1.4216 1.6986 N/A N/A Fxy 1.5662 1.95 N/A N/A
~
FQ (2) 2.2070 4.64 2.2415 - 2;26'~ QPTR 1.0137 1.02 N/A N/A-2- O 4.0-15 4- _ .r - --- .,-.,.r ,mw-, .-se.r,- ,,-e-rw,. -.--...ww.-e.n.-_r ~,re.,-.-- , - - . . . . , . .
l 4.2 ComT10L SYSTEM DYNAMIC RESN)NS'd
. I Prior to the completion of the 30% power plateau testing, the proper response of the re?ctor control system, the steam dump control syst s and j the steam generator level control was verified. The purpose of these tests '
was to verify that the controller settings resulted in- relative stable operation. The steam generator level control system section also discusses testing done at higher power levels. The overall plant response to
- transient and trip testing is discussed in section 4.3.
ps *
- 20 O
4.0-16
4.2.1 DY!Wi!C AU10% TIC !rTEAM IXMP CORTH0!. O The steam dump (turbine bypass) syston consists of cwelve valves which are designo3 to handle 40% of full turbine stexn flow at full load precure. Scven valves dincharge into the low pressure condensor, four vslves discharge into the intermediate condenser, aM a sinolo valve discharges into the high pressure adenser. The systen is designed tot
- 1) Allow a 50 oereent step electrical loM reduction without reactor trip (Section 4.3.7) . Tha syst m will allow a turbine and resctor trip from full power witnnut lif ting the main stexn safety valves,
- 2) Control steam generator pressure at no-load conditions,
- 3) Dypass steam to the main condenser during plant startup and permit a l normal mnual cooldown of the ICS from a hot standby condition to a !
point consistrat with the initiation of residual heat rmoval systan operation. ' The steam dump sycts, during normal operating transienta for which the plant is designa), is automatically regulatoj by the RCS tanparature control syntan to maintain the programed coolant taperature (Tavg mode) . The programed coolant tmperature (Tref) is derival fran the high pressure turbine first stage impulso pressure, which is a load reference signal. The dif feronce between Tref and measured Tavg is uso3 to activate the steam dump systan under automatic control. The syst m operates the valves in two O fundxnental modes. In one mode, twc groups of six valves each trip open sequential'.y in approximately 3 seconds. This operatf oral mode is rtivate3 during a large reactor-to-turbine power mismatch (load reje tion contrailer). In the socond node, four groups of three valves coch modulate opor1 sequentially in approximately 10 so:onds (plant trip controller) . When the plant is at no load and thera is no turbine load reference, the systan is operatoi in the pressure control mode. The measured main strern , header pressure is c. mpared to the pressure set by the operator in the control rocru. The pressure control mode is also used for plant cooldwn. Both the Tavg node and the stexn pressure node of control were individually tested. In the steam pressure mode, reactor power was raisoi from approximately 2 percent to approximately 10 percent to verify that the steam dump synts would adjust valve position to maintain the stexn header pressure at the set value. In the Tavg node, two individual tests were performe3; one to verify croper plant trip controller response, and one to verify load rejection controller response. The plant trip contrM irr response test was performed by simulating P-4 reactor trip signal aM then adjusting Tavg by changini reactor power to verify proper automatic steam dtrnp operation. The load reje: tion controller response test was, performed by simulating a sudden tons of load at approximtely 6 percent reactor p>wer and verifying proper automatic stexn dump operstion. DurinJ the testing, steam d e p systen O parameters were annitorol using strip chart recorders. 4.0-17
i The following acceptance criteria were satisfied during the testingt O 1) The plant trip gentroller tha-M-500D) responded properly to maintain Tavg at 562 + 2 F at approx mately 6% power. Af ter steady state power was achieved ~there were no divergent oscillations in tmperature,
- 2) The load rejo: tion contrglier (AB-E-500A) respondo3 properly to maintain Tavg at 561 + 2 P at approximately 6't power. After steady state power was achieved, there were no divergent oscillations in tmperature,
- 3) The stern header pressure controller (An-PK-507) responde3 properly to maintain pressure at the normal no-load pressure of 1092 + 25 psig.
No control tr/ stem s ' Ments were requirel as a result of thin test. The test was ettnplete3 on June 11, 1983, after several days delay while adjustments were made t0 the automatic steam generator level control systen. O~ O 4.0-13 l
l l l 4.2.2 AUWMATIC RFJCIVR C0ffrHOL O To control TCS average tenperature (Tavg) as power is increased er de:ressed a reference tenperature, Tref, corresponding to turbino load as indicated by ' first stage turbine imp:1se pressure, is provided to the resctor control systs. Tref is ths) canparei to the actual Tavg in the RCS loops. The auctioneerei high Tavg from the four loop Tavg's fs used as the basis for the comparison. If a difference of more than 1.5 F exists, the rod control systs, when in the automatic node, will insert or withdraw the control rods to align Tavg with Tref. The purpose of this test was to verify initial satisfactory operation of the systs in automatic. Additional testing to perform adjustments on the systen is diacussed in Section 4.4.5. To verify systen operation, Tavg was increasei by approximately 6 0F by withdrawing the control rods in MANUAL. Rod control was then placed in AUTO and various parameters were 'tonitored on strip chgrt recorders as the plant respondai. The process was then repeatoi for a 6 F decrease in Tavg by inserting the control rods in MWJAL and then placing the rod control systan in AUM. This test was performed with the plant at steady state 30% power conditions on June 27, 1985. For both the increase and decrease in Tavg, control rods innediately started gto nove when the rod control systen was places in AWo. Tavg was within 1.5 F of Tref in less than one ndnute in both cases. The following acceptance criteria were satisfied: No manual intervention was required to bring plant cont tions to 1) equilibrium values following the initiation of the transients, 0
- 2) Tavg returned to within -+ 1.5 F of Tref following the itiation of the transients,
- 3) Tenperature variations were less than 5 F0 peak-to-peak and tenperature oscillations had a period of greater than one minute following the initiation of the transient, Wheg the data was analy;:e3 slight renparature oscillations of less than 0.5 F with a period of approximately twenty seconds were noted both before and after the initiation of the transient. However, there was no taperature oscillation caused by the reactor control systen during the transient, therefore, the period was infinity and acceptance criterion 3) above was satisfied. Westinghouse concurred that the results were satisfactory.
O 4.0-19
l l 4.2.3 AtTI0MATIC STrhi C12iERA'10R LIN!1 COffrROL TPST O This series of tests was performe3 to verify the satisf actory operation of the various canponents of the automatic steam generator level control syntan at steady state conditions as well as under increasingly severe plant transients. The head capacity curves of both stern driven main feWwater p:mps were also verified. Initial testing was performo3 at approximately 10% reactor p]wer to verify the ability of the bypass feo3 water control valves to control stexn generator level at ::ero electrical load in automatic. Level offsets were introduce 3 in each steam generator and then the bypass feedwater control valves were placed in automatic and a110wa3 to restore level. The nuclear fee 3 forward signal in the bypass salve control circuitry was chceke3 by performing 3% power decreases and increases and monitoring valve perfarmance. The next test was performed in the 10% - 19% power range to verify the initial satisfactory operation of the mm a feo3sater control valves. The bypass valves were closed, the main fee 3 water control valves were in manual, the feo3 water pumps were in auto and the pump spee3 controls were in auto. A level of fset was introduced in each steam generator and the applicaole control valve was place 3 in automatic aM a110we3 to restors level. 31 and 11 power changes were also introduced with the main fee 3 water control vsives in automatic. Overall operation was satisfactory. Slight setting cnanges were made in the controls for the C steam cenerator to reduce coupling with Q the B stern generator. Two defective controller cards were replaced in the pump spee3 control syste. Testing at 30% power included verification of the operating characteristics of the steam driven main foo3 water pJmps. Both pumps exceede3 predicted performance. Main fee 3 pump spee3 control performed satisf actorily under these conditions. A ten percent power decrease was perforw3 aM then a ten percent power increase. Overall ooeration of the sten generator level . control systan was satisfactory although levels in the 3 an3 D stexn generators went below 40% briefly at the beginning of the transient. Pump performance was again verifici to be satisfactory at 5 n power. steam generator level control was monitoro3 during a 25% power decrease and then a 15is power increase. The power increase was terminate 3 at 40% due to a xenon tr ansient. As a result of these transients adjustments were ma3e to the main fee 3 water valve controllers aM the master pump spee3 controller. Also, the fee 3 water pump delta P/ speed controller did not achieve the desired setpoint at 251 power for the dif ferential between fee 3 water hea3er pressure and steam header pressure. The steam ficw indication loops were realigned. At 75% power, steam generator level control was "onitorei during the large lo M reauction test (Section 4.3.2) . The overall response of the control systes was satisf actory during the transient, although levels went natside the 40-631 baM and were outside the 45-551. band af ter 5 minutes. As noted previously at 25% pcwer, the feetaater pump delta P/ speed controller was still not operating satisf actorily at 251 power thus requiring further Mj ustment. 4.3-20
. . . _ _ _ _ . . . . _ ~- . - _ .
_ _ _ . . _ - _ _ . _ . _ . _ _ . _ _ _ _ . . . . . . , ~ . _ . - During-XV test (Section100% 4.3.1) power
. Controltesting, systondata was collected performance during tAe !;scy.
was satiritan TheM .wanj control O- syntans were then nonitored during the large load re3uction test (Section I.evel did go outside the 40% -60% band with a low level of 34.5% in
- 4. 3. 2) .
the "D" stexn generator. After review of the data, Westinghouse determined that response was satisfactory since no levels had been outside of initial level +15%. The feedwater pump delta P/ speed controller parformo3 satisfactorily during the transients. In sumnary, the steam generator water level controls were satisfactority adjusta3 to handle norms 1 operation as well as major oscillations. The main feo3 water pumps head and capacity curves were botter than expo:ted. O k O l- 4.0-21 e
4.3 TitN4SIINP Ato TilIP T13TS O Plant response to transients of varying magnitude was determine 3 during the power ascension phase of the startup progrxt. These test; were utili::a3 to analy::o the overall behavior of the major plant control s*j stans during a p)wer swing or an actual plant trip. The need for further control systen adjustments was determined by monitoring various plant parameters before, during and after each transient. The specific tests performo3 consistel of a series of 10% power load swinJs at 301, 75% and 100% power, a 501 loa 3 reduction at 751 and 1001 power, a trip to determine ability to shutdown and maintain hot standby external to the control room, a trip at 50% power to verify the abilit'/ of the nuclear instruments to detect droppe3 roda, and a unit trip at 100% power. O O 4.0-22
4.3.1 ftwD switC TICTS Tha purpose of the lott swing tests was to verify ti e prepar nuclear plant transient response, includinj autantic ' control systan perfomance, wnen 1M load changes, both decrease and increase, were introluceJ at the turbino generator. Tests were performai at the 30%, 7511 an] 100% power test plateaus. The tests were started with stable plant conditions at the desired test plateau an1 the followiry) systuns in automatics
- 1) Steam Generator Main ?celwater Control,
- 2) Pressurizer Pressure Control,
- 3) Pressurizer Heater Croups A anj B,
- 4) Pressurizer Heater Group C in CLOSE,
- 5) Pressurizer Level Control,
- 6) Steam Dmip Control in Tavg Control Mode,
- 7) Main Foolwater IV1p Tutbine Spwd Control.
Af ter initial plant data was collected to verify plant stability, the electro hydraulic controller (IrH) was usod to achieve a 1M load dc:reme as rapidly as possible. When the plant was in a atable condition additional data was collected. During the transient cert.ain parameters were mnitoral on multichannel strip chart recorders. 7 The DiC controller was then used to increase the plant output as rapid 1y as (d possible to achieve a 10% lor! increase artl attain a final plant level at approxirutely the original test plateau. Af ter the plant wu in a stable condition, a final set of data was collected. The acceptance criteria for these tests weret
- 1) No reactor trip was generated,
- 2) No turbine trip was generated,
- 3) Safety injection was not initiatel,
- 4) Neither the steam generator relief valves nor safety valves lif ted,
- 5) Neither the pressurizer relief valves nor sahety valves lif ted,
- 6) No manual intervention was required to bring the plant conditions to stea3y state,
- 7) Nuclear power overshoot was less than 3% for load increase,
- 3) Nuclear power undershoot was less than 31 for load decrease.
The data from the three test plateaus is sumnarizal in Tables 4.3.1-1 through 4.3.1-6. 4.0-23
TABLE 4.3.1-1 LOAD SWING FRCH 30% 'IO 20% NfdER Parameter Initial Durina Transient _ Final i Minimum Maximum t Plant Operating Level ; (No-Gross) 290 - - l90 ' Nuclear Power (%) , 32 20 32 72 Tavo - auctioneernd (XF) 566 561.5 566 M ; Tref ("F) 566 - - s63 Delta T - Loop 1 (%) 38 2$ 36 7 Overpower Delta T Setpoint(%) 108 - - 108 overtorporature Delta T Setpoint (%) 140 - - 147 Pressurizer Pressure (psig) 2230 2219 2283 2230 Pressurizer Level ()) 35 29 36 30 Steam HeLer Pressuro (psig) 1030 1024 (1) 1068 (1) 1030 Steam Flow (lbght x 10') Loop 1 1.1 - - 0.7 Loop 2 1.15 - - 0.8 Loop 3 1.2 - - 0.85 Loop 4 1.0 - - 0.55 _ Narrow Range Steam Generator Level (%) Loop 1 49 38 57 48 Loop 2 50 40 58 50 Loop 3 49 40 58 50 Loop 4, 49 39 56 49 Feedwater Tenperature ("F) Loop 1 347.7 - - 316.8 Loop 2 347.8 - - 317.3 Loop 3 347.8 - - 317.0 Loop 4 347.2 - - 316.1 Feedwater (lbpr x 10Flgw) Loop 1 1.0 - - 0.4 Loop 2 0.9 - - 0.4 l Loop 3 1.0 - - 0.45 l , Loop 4 1.0 - 0 0.5 Feodwater Pump Discharge (1) (1) Pressure, esig 1120 1064 1155 1110 l Feed Pumo A Speed (RPM) 3900 1 38,70 4200 3600 Feed Pumo B Speed (RPM) 1000 (1) - (1) - 1000 Control Bank D Position (steos) 194/195 - - 146/145 (1) Fran test recorders Time to reach eTuilibrium following load change: 4 minutes. O 4.0-24 _ _ _ _ . . ~ . _ _ _ , . . . _ , - . _ . . _ _ . . , _ . - , _ . -
l l l TABLE 4.3.1-2 LOAD SWING FROM 20% 10 30% PGiER Parameter Initial During Transient Final Minimum Maximum Plant Operating Level (Mie-Gross) 190 - - 290 Nuclear Power (%) m 22 22 32.5 32 Tavo - auctioneered (%F) 562 559 565.5 565.5 Tref ('F) 563 - - 565.5 Delta T - Loop 1 (%) 27 27 38.5 38 Overoower Delta T Setpoint (%) 108 - - 109 Overtenperature Delta T Setpoint (%) 147 - - 138 Pressurizer Pressure (psig) 2230 2215 2275 2225 Pressurizer Level (1) 30 26.5 36.5 34 Steam Header Pressure (psig) 1030 983 (1) 1030 (1) 1028 Steam Flow (1b g /hr x 10") Loop 1 0.7 - - 1.1 Loop 2 9.8 - - 1.15 Loop 3 0.85 - - 1.2 Loop 4 0.55 - - 1.0 Narrow Range Steam O- ceaer tor Levet <s) Loop 1 48 41 57.5 48 Loop 2 50 41.5 61 50 Loop 3 50 40.5 58 49 Loop 4 m 49 43 59.5 49 Feedwater Tatperature ('F) Loop 1 316.8 - - 347.9 Loop 2 317.3 - - 348.0 Loop 3 317.0 - - 348.4 Loop 4 316.1 - - 347.6 feedwater (lbg/hr x 10 Flgw) Loop 1 0.4 - - 0.95 Lcop 2 0.4 - - 0.90 Loop 3 0.45 - - 1.0 Loop 4 0.5 - - 1.0
- l. Feedwater Pump Discharge (1) (1) i Pressure, osig 1110 1021 1196 1120 Feed Pump A Speed (RM1) 3600 (1) 3720 (1) 3785 3900 Feed Pumo B Soeed (RPM) 1000 (1) 1000 (1) 1000 1000 Control Dank D Position l
(steps) 146/145 - - 194 l (1) From test recorders l Time to reach equilibrium following load change: 3 minutes. O 4.0-25
i 1 l TABLE 4. 3.1-3 LOAD SWI!G FROM 75% M 65% IWER ; O ! Parameter Initial During Transient Final l Minimum Maximum Plant Operating Level (tiMross) 775 - - 650 Nuclear Power (%) 75.5 75.5
~
m 60 t2.5 _Tavo - auctioneered (XF) 579 575 581 576 Trer _ ( F ) 581 - - 578 Delta T - Loop 1 (%) 75 64 76 67 Overpower Delta T Setpoint(%) 109 - - 109 Overtsperature Delta T Setpoint (%) 122 - - 128 Pressurizer Pressure (psig) 2250 2219 2288 2230 Pressurizer Level (%) 54 45 56 46 Steam Header Pressure (psig) 1000 1014 (1) 1366 (1) 1000 Steam Flow (lny' /hr x 10") Loop 1 2.7 - - 2.3 Loop 2 2.5 - - 2.2 Loop 3 2.7 - - 2.3 Loop 4 2.5 - - 2.3 Narrow Range Steam Generator Level (%) O Loop 1 48 42 53 48 Loop 2 49 42 53 50 Loop 3 49 42 55 50 Loop 4, 49 42 52 48 Fetrlwater Tcmperature ('F) Loop 1 416.6 - - 433.3 Loop 2 416.9 - - 403.6 Loop 3 416.8 - - 433.7 Loop 4 41C.1 - - 402.8 l FeedwaterFlgw) (lbg/hr x 13 Loop 1 2.7 - - 2.3 l Loop 2 2.7 - - 2.2 l Loop 3 2.8 - - 2.3 Loop 4 2.7 - - 2.3 Feedwater Pump Discharge (1) (1) Pressure, osig 1180 1188 1204 1143 Feed Pump A Speed (RPM) 4400 (1) 4200 (1) 4608 1 4100 Feed Pu:rn B Soeed (RPM) 4600 (1) 4392 (1) 4824 4300 Control Bank D Position (steos) 199/198 - - 146/145 9 (1) Frcm test recorders I-Time to reach equilibrium following load change: 5 1/2 minutes. l O 4.0-26
, , - L. . , .- . - . - - - _ . . .. - - , . . . - .- .- -
TABLE 4.3.1-4 LOAD SWI!G FROM 65% 'IO 75% M,iER Parameter Initial During Transient Final Minimum Maximum Plant Operating Level (Mio-Gross) 660 - - 760 Nuclear Posagt _( %) m 62.5 62 74.5 74.5 Tava - auctioneered ('4F) 576 575 580 580 Tref ("F) 578 - - 582 Delta T - Loop 1 (%) 66 66 76 75 Overpower Delta T Setpoint (%) 109 - - 109 overtmperature Delta T Setpoint (%) 117 .. - 121 Pressurizer Pressure (psic) ,__ 2238 2220 2275 2235 Pressurizer Level (%) 47 44 53 53 Steam Header Pressure (psig) 1010 975 (1) 1014 (1) 1010 Steam Flow (lbpr x 10 ') ' Loop 1 2.3 - - 2.85 Loop 2 2.25 - - 2.8 Loop 3 2.43 - - 2.85 Loop 4 2.35 - - 2.8 Narrow Range Steam O Generator Level (%) Loop 1 48 42 55 48 Loop 2 47 43 58 50 Loop 3 48 42 57 49 Loop 4, 48 43 55 49 Feedwater Teperature ("F) Loop 1 402.7 - - 415.1 Loop 2 403.0 - - 415.2 Loop 3 403.0 - - 415.5 Loop 4 402.1 - - 414.5 FeedwaterFigw) (1bpr x 10 Loop 1 2.4 - - 2.8 Loop 2 2.4 - - 2.8 Loop 3 2.45 - - 2.85 Loop 4 2.4 - - 2.8 Feedwater Pump Discharge (1) (1) Pressure, psig 1150 1077 1116 1M0 Feed Pumn A Soeed (RPM) 4100 (1) 4032 (1) 4392 4400 , Feed Pumo B Soeed (RPM 4300 (1) 4104 (1)-4536 4530 Control Bank 0 Position (steos) 148 - - 199 (1) From test recorders Time to reach equilibrium following load change: 5 minutes. O 4.0-27
TABLE 4. 3.1-5 Q LOAD SWIM FROM 100% to 90% FOER Parameter Initial During Transient Final I Minimum Maximum Plant operating Level (Mie4ross) 1050 - - 930 Nuclear Power (%) , 99.8 87.5 99.8 87.5 Tava - auctioneered (XF) 587.5- 585 587.5 585 Tref ('T ) 588.5 - - 585 Delta T - Loop 1 (%) 101 90.5 101 90.5 Overpower Delta T Sotpoint(t) 108 - - 108 overtenperature Delta T Setpoint (%) 112 - - 115 Pressurizer Pressure (psig) 2230 2210 2300 2220 Pressurizer Level (%) 61 57 65 57 1000 1000 (1) Steam Header Pressure (psig) 1040 (1) 1910 Steam Flow (1bpr ' x 10") Loop 1 2.99 - - 2.61 Loop 2 2.91 - - 2.53 Loop 3 2.95 - - 2.61 Loop 4 2.95 - - 2.57 , Narrow Range Steam O cemer tor 'evet (s) Loop 1 48 42 50 48 , Loop 2 50 44 53 50 Loop 3 48 45 53 49 Loop 4, 49 46 52 49 Fe(dwater Tanperature IF) Loop 1 439.1 - - 429.5 , Loop 2 439.3 - - 429.7-Loop 3 439.5 - - 429.6 Loop 4 438.6- - - 428.7 FeedwaterFlgw) (1bpr x 10 Loop 1 2.99 - - 2.65 Loop 2 2.87 - - 2.49' Loop 3 2.95 - - 2.57 Loop 4 2.95 - - 2.53 Fe<dwater Pump Disenarge (1) (1) Pressure, osig 1200 1188 1240 1190 Feed Pump A Speed (RPM) 5000 (1) 4714 (1) 5143 4700 Feed Pump B Soeed (RPM) 5200 (1) 5000 (1) 5429 4900 Control Bank D Position (steos) 222 - - 162 t (1) Fron test recorders Time to reach equilibrium following load change: 6 1/2 minutes. O 4.0-28 _ . _ _ - , _ . - . _ _ , _ _ _ _ . . _ . _ _ _ _ . _ . _ . _ _ _ _ . . _ . . . _ , _ . . . . _ . . _ . _ , _ _ . . ~ , , ,
TABLE 4. 3.1-6 LOAD SWING FRCli 90% to 100% PCNER Parameter Initial _D,oring Transient Final ! Minimum Maximum Plant Operating tavel (M4o-Gross ) 935 - - 1010 ~ Nuclear Power (%) m 89 99 93 93 Tavg - auctioneered (;F) 585.5 581.5 582 582 Trof ( F) 7 85 - - 587 Delta T - Loop 1 (%) 91 91 98 98 Overpower Delta T Setpoint (%) 109 - - 1 03 Overtmperature Delta T Setpoint (1) 116 - - 120 , t Pressurizer Pressure (psig) 2245 2220 2250 2250 Pressurizer Level (%) 59 52 59 54 Steam Header Pressure (psig) 1010 962 (1) 1014 (1) 962 Steam Flow (lbg /hr x 10) Loop 1 2.65 - - 2.91 Loop 2 2.57 - - 2.84 Loop 3 2.65 - - 2.84 Loop 4 2.61 - - 2.90 Narrow Range Steam n Generator Level (%) V Loop 1 48 46 54 48 Loop 2 50 47 58 50 Loop 3 50 45 56 50 Loop 4, 49 47 55 49 Feedwater Tmperature ("F) Loop 1 429.7 - - 436.4 Loop 2 429.9 - - 436.8 Loop 3 429.9 - - 436.5 , Loop 4 429.2 - - 436.0 Feedwater (1bghtx10Flgw) Loop 1 2.65 - - 2.91 Loop 2 2.33 - - 2.80 Loop 3 2.57 - - 2.84 Loop 4 2.57 - - 2.84 Feaiwater Pmp Discastge (1) (1) Pressure, psig 1200 1123 1188 1155 Feed Pump A Soeed (RPM) 4750 (1) 4571 (1) 4714 4800 Feed Pump B Speed (RPM) 4950 (1) 4857 (1) 5143 5000 Control Bank D Position (steps) 194 - - 223 (1) Nm test recorders Time to reach equilibrium following load change: 6 1/2 minutes. O 4.0-29
- - . - _ _ _ _ , _ _ _ _ _ , . __ _ _ . . _ . _ . _ _ . . . _ - _ _ _ - _ _ _.J. _ .. _ .-.. _ _ _
The test at 30% power was performed on June 29, 1985. All of the acceptance criteria as outlined above were satisfied. tere wre several prameters ' O that went outside their expected band:
- 1) Maximtrn pressurizer pressure was 53 psig above initial pressure (3,50 psig expected),
- 2) Steam generator levels were expected to be within + 10% of initial-levels but steam generator A dropped 11% on the load decrease ard steam generator B increased 11% on the load increase,
- 3) Tavg was not expected to overshoot (undershoot) itsffnalvalueon load increase (decrease) but on the load decrease 0.5 r undershoot was noted.
These slight deviations from expected valves wre reviewed by Westinghouse ard determined to be acceptable. The test at 75% power was perfomed on July 28, 1985. All of the acceptance , criteria as previously outlined were satisfied. Durin3 the initial 10% ' power decrease, the operators had to take manual control of feedwater pump , speed control to maintain steam generater feed flow. After adjustments to the gain on the feed flow / steam flow mismatch cards (FY510D - FYS400), the 10% pawer decrease ard the 10% power increase were performed without further difficulty. As for the 30% power test, several parameters, including steam pressure overshoot /undershoot, steam generator level swing and Tavg undershoot were outside the Westinghouse expected range by a slight amount p but this was determined not to impact the test results ard no additions 1 v control system adjustments were required. The test at 100% power was performed on August 22, 1985. All of the acceptance criteria as previously outlined were satisfied. The p wer increase was actually closer to 7% than 10%. Pods had to be withdrawn somewhat prior to the power increase to bring axial flux difference within Technical Specifications limits. This left insufficient rod worth to complete the 10% power increase. A review of the test data showed that all systcrns performed satisfactorily ard that there was no need for further testing. As noted in the previous load swing tests, some parameters were slightly outside the Westinghouse expected values including RCS pressure swing and steam pressure overshoot. Wese were reviewed by Westinghouse and determined not to impact the test results. No control system adjustments were required. l-l L O 4.0-30 l
4.3.2 IWG: IDAD Rl:DOCr!ON TF'3TS The large load reduction tests were performal during the 75 percent and 103 percent pw'I testing plateaus to verify the ability of the primary plant, secondary pl nt and the automatic reactor control systens to sustain a 50 percent stcy wl reduction. The data obtaino3 was usal to evaluate the interaction bet. % n the control nystems and to determine if any setpoint or gain adjustments e Te n eessary. During each load reduction test, selected plant paraaieters were trended on multi-channel strip chart records. Stable operation was verifie3 with rod control, steam generator MFW control, pressurizer pressure and spray control, pressurizer level control an3 feo3nter ptrnp turbine spee3 control systens in automatic. Steam dump control was in the Tavg ntxle. Using the standby lon3 set, generator load was reduced by 50 percent within approximately 1 minute with all test recorders in high spee3. No manual intervention with the control systens was allowed durinq and following the 50 percent load reduction. The plant was allowal to stabilize with centrol systuns in automatic. The large load reduction test at 75 percent power was performed on July 28, 1985. Without manual intervention, the automatic control systans did sustain the load reduction and allow the plant to be returne3 to stable conditions. All acceptance criteria were natisfied: 0 t) Twe re>ctor did not trin.
- 2) The turbine did not trip,
- 3) Safety injection did not initiate,
- 4) Steam generator safety valves did not lift,
- 5) Pressurizer safety valves did not litt,
- 6) Ho manual intervention was require 3 to reach equilibrium plant conditions.
The actual load reduction was 627 MWe (54.51) . As a result, the steam dump valves did not shut until after 9 minutes, 40 segonds after start of trgnsient (expecte3 8 minutes) and Tavg peakel 6'F cave its initial value (5 F expected). A Westinghouse evaluation determine 1 that both these values were consistent with a 54.5% power decrease. The test data is sumnariza3 in Table 4.3.2-1. No control syntan setpoints or pins were change 3 as a result of this test. The large load reduction test at 103 percent p>wer was performe3 on August 29, 1985. Without manual intervention, the plant systans reached equilibrium conditions. All the major acceptance criteria were met as discussa3 above for the 751 power test. No control systan setpoints or O aeios were cheooe4 se a reso1t t taie test-Table 4.3.2-2. The test det> i s '2,n > < i ma i , 4.3-31
TABis 4.3.2-1 : O LARGE LOAD REDUCTION TEST FROM 75% POWER L Parameter Initial During Transient Final '; Minimum ' Maximum Nuclear Power (%) , 75.5 20 75 26 Tava - auctioneered (;F) 580 562 586 ' 564 Tref ("F) 581 - - 565 , Delta T (%) 77 26 77 32 Overpower Delta Setpoint (%) 109 - - 109 over tenperature Delta T Setnoint (%) 121 _ 144 Pressurizer Pressure (psig) 2240 214J 2320 2240 Pressurizer Level (%) 52 2 59 31 Steam Header Pressure (psig) 1014 1014 1118 1040 Sten flow (lb g/hr x 10) Loop 1 2.75 - - 0.9 Loop 2 2.6 - - 1.0 Loop 3 2.8 - - 1.0 Loop 4 2.7 - - 1.0 Narrow Range Steam Generator Level (%) Loop 1 48 31 62 47 !
'O 'oca 2 Loop 3 Sa 50 33 31 66 66.5 so 50 Loop A 50 34.5 64.5 49 Feedwater Tanperature(F)
Loop 1 416.4 - - 333.2 Loop 2 416.8 - - 333.4 Loop 3 416.6 - - 333.8
~
Loop 4 415.8 - - 332.9 Fealwater (1bg /hr x 10 Figw) Loop 1 2.75 -
- 0.8 Loop 2 2.75 - -
0.45 Loop 3 2.8 - - 0.65 Loop 4 2.7 - - 0.8 Feedwater Pump Discharge Pressure, psig 1162 1136 1240 1136 Control Bank D Position (steps) 181 - - 95/94 Time to reach equilibrium following load change: 36 minutes. 1 0 - 4.0-32
, - . _ _ . . _ _ . _ , _ _ _ . _ . , _ _ . _ . - . - . . . . . - _ _ _ . , - , ~ , _ . _ . _ . _ . - -
7 1 TABLE 4.3.2-2 l l LARGE LOAD REDUCTION Os TEST FROM 100% IVdER Parameter Initial During Transient Final Minimum Maximum Nuclear Power (U _ m 99.5 53 99.5 53 Tay1 - auctioneered (F ) 588 573 588 573 l Tref ("F) 588 - - 573 Delta T (U 100 62 100 62 3verpower Delta T Setpoint(%) 108 - - 108 Over tenperature Delta T Setpoint (t) 110 - - 131 Pressurizer Pressure (psig) 2235 2155 23461 2255 Pressurizer Level (%) 62 44 62 44 Steam Header Pressure (psig) 1020 1020 1118 1027 Stoan Flow (lbg /hr x 1P') Loop 1 3.8 - - 2.1 Loop 2 3.6 - - 2.0 Loop 3 3.7 - - 2.2 Teop 4 3.6 - - 2.1 Narrow Range Steam , Generator Level (%) Loop 1 48 31.5 56 48 Loop 2 50 33 57 50 Loop 3 49 33 58 50 Loop A 49 32.5 58 50 Feedwater Tenperature("F) Loop 1 438.2 - - 392.0 Loop 2 438.6 - - 392.3 Loop 3 438.5 - - 392.2 Loop 4 437.6 - - 391.3 Fes3 (1bprwater x 10Flgw) Loop 1 3.8 - - 2.1 Loop 2 3.6 - - 1.9 Loop 3 3.7 - - 2.0 Loop 4 3.7 - - 2.0 Feedwater Pump Discharge Pressure, psig 1188 1162 1331 1162 Control Bann D Position (steos) 217 - - 97 Time to reach equilibrium following load change: Approximately 12 minuten (based on steam dump denand trace) . O) L 4.3-33
l l 4.3.3 SilUTDOWN AND f4Altm2WCE OF llOf STANDUY EXTERNAL 'IU (]/ u 'IME CONTROL ROO1 The purpose of this test was to dmonstrate that, using plant operating procedures (OFN Control Roam Not Habitable), the plant can be taken from > 10 percent power to hot stanty conditions and then maintaine3 in hot I standby for at least 30 minutes with a minirntrn shif t cres using centrols and instrumentation external to the Control Room. The minimum shift crew was that specified in OFN-13. This test was performed on June 29, 1985. During the evacuation of the Control Room by the minimum shift crew, a standby cred reained in the Control Room to nonitor plant conditions. The standby crew was to take no actions during the transient. The reactor was tripped at 1111 by manually tripping the reactor trip breakers from the Control Room. The minimum shift crew thm evccuated the Control Room to take control of the plant at the Auxiliary Shutdown Panel (ASP) and other duty ::tations as outlined by OFN-13 and as dire:ted by the Shift Supervisor. Hot stan6y conditions were established at 1147 and maintained until 1220. During this period, pressuri::er pressure, pressurizer level, steam generator level and FCS temperature were maintained from outside the Control Rcom. The acceptance criteria of the proce3ure were met. O 4.0-34
4.3.4 H0DJ DROP AND MNTP TRI? O This test was performed at the md of the 50 percent power test plateau to demonstrate operation of the negative rate trip circuitry by dropping two ICCA's from a comnon rod group and to revies plant response and control systens behavior to a plant trip from an intermediate power level prior to the plant trip test from 100 percent power. During the performance of the test, a high speed chart recorder was used to nonitor the state of the reactor trip broskers, the rod-on-bottom alarms, nuclear instrumentation systen (NIS) power and negative rate trip histable output. While at steady state plant conditions (50 + 5% power), the two control rods (D-4 and M-12) in group 1 of control bank D (CDD) were transferred to the DC hold bus. The drop of the two rods was initiated by removing the stationary gripper coil fuses for D-4 and M-12 in power cabinet LAC and then deenergizing the DC hold bus. The trip test was perforned July 16, 1985 with satisfactory results. The two dropped rods did caune a reactor trip due to power range high negative rate and all rods dropped normally. The pressurizer safety valves did not lift, the steam generator safety valves did not lift, and safety injection was not initated, thus satisfying all the acceptance criteria for the test. In addition, the reactor trip generated a turbine trip. The steam dumps operated to reject heat to the condenser and feed flow, steam flow, and O steam generator narrow range levels all went to zero as indicated on the process insttumentation. The traces on the high speed chart recorder did not show a change of state for the NIS negative rate bistables because the recorder was set up with a 5 to 120 VAC range. Investigation determined the NIS negative rate bistables tripped at 9 VAC therefore the event recorder bistable in the recorder did not show a change of state. However, the negative rate bistable trip was shown by the indicator lights on the front panel of the NIS negative rate trip drawer and by the first-out indicators on the main control board. Beth of these had to be reset prior to subsequent startup.
- Table 4.3.4-1 su
- tmarizes the data from this test.
i O 4.0-35
l TAntI 4. 3. 4-1 RODS DROP AfD Pf/WT TRIP l TEST IM1A SU't%RY 4 i Plant Before During Transient After Para:wter Trip Minimum Maximum Trip Tavo Auctioneered, OP 568.5 555.0 568.5 554.5 i Tref , ~F 571.5 - - 558.0 _Dolta T - Loop 1 , % 55 2 55 2 Overpower Delta T Setpoint - f.oo3 1, % 134 - - >150 overtemperature Mlta T
'Sotpoint - Loop 1, % 109 - - 108 Pressurizer Pressure, psig 7225 2125 -2245 2245 Pressurizer !
t.evel, % 38 24 38 25 Narrow Range, -- ; Steam Generator , Level - Loop 1, % 49 0 53 14 k O ' i l s e t l-O 4.0-36 __. _ _ _...._ . - _ _ _ .~.. _,.- . , . . ~ . . _ . _ _ . . _ . - . _ _ . , _ . , , . , _ . . _ , _ . , , . _ _ , , . _ - . _ . , _ _ . _ . ,
4.3.5 14Mir TRIP FROM 100 PFICDfr twER O The purpose of this test was:
- 1) To verify the ability of the primary and secondary plant aM the plant automatic control systans to sustain a trip fran 100 percent power and to bring the plant to stable conditions following the transient,
- 2) To detot nine the overall response time of the reactor coolant hot icg resistance tunperature detectors aM,
- 3) To evaluate the data resulting from this test to determine if changes in the control systan setraints are wirranted to improve transient response based on actual plant operation.
For the performance of this test, the plant was operated at 100 percent power with the following control systems in aatonatiet
- 1) Reactor Ra$ Control,
- 2) Steam Generator Main Feedwater Coretrol,
- 3) Pressurizer Pressure aM Spray Control,
- 4) Pressurizer Heater Groups A and B,
- 5) Pressurizer Heater Group C in close position,
- 6) Pressurizer Level Control, O 7) 8)
steam o'ree coatrot ta tav9 coaerot moee, thin Feedwater Pump Turbine Controls. All shutdown aM control rod banks were fully withdrawn except control bank D which was positioned to maintain axial flux difference within the limits specified in WCCS Technical Specifications. Various plant parameters were input to strip chart recorders to monitor the plant performance during the trip. Af ter initial data was recorded ard the test recorders were operational, the plant trip was initiated by momentuity jumpering two terminals of the Turbine Soquential Trip relay (AR in Cabiret t% 104B) which opened both Main Generator Output Breakers. After the trip, the plant was restored to a stable condition using normal plant operating procedures.. The' plant was tripped from 100% power at 0512 on August 28, 1985. Using the opening of Genera:or Output Breaker #2 as time zero, the turbine tripped at 0.075 neconds and the reactor trip breakers opened at 0.155 seconds. The steam dtrnp valve deman$ signal went to 100% (opan) at 0.225 seconds and then mdulated to 0% (closed) at 46.895 seconds. The steam dump valves were placed in pressure control modo at 0517. Feedwater isolation signal was received at 3514
'!he following major acceptance criteria of the test were met during the O trie:
- 1) All control rods dropped, 4.0-37
- 2) Reactor Coolant System Pressure remained less than 2450 psig (maximum pressure of 2260 psig was reached at 16.5 minutes during plant recovery),
- 3) Steam Header Pressure reained less than 1150 psig (maximtrn pressure of 1090 psig was reached 'at 5.965 secon3s),
- 4) Satety Injection was not initiated,
- 5) Usirg tha strip chart recordings, the overall Hot Leg RTD resporea times were determined for each loop as Loop 1 6.000 seconds toop 2 5.900 seconds Loop 3 5.630 secords Loop 4 5.880 seconds All response times were less than the requirei 8.4 seconds. ,
The overall Hot Leg RTD response time includes rerponse of the RTD
-itself, and was defino3 for the purpose of this test as the interval of time measured between-the point where the neutron flux has decreased by 50 pareent of its initial value to the point where the hot leg temperature signal (as measured by the RTD output) has decreased by a value equivalent to 331/3 percent of the initial loop delta T value in degrees Fahrenheit.
The remaining major acceptance criterion required that neutron flux drop below 15 percent power within 2 seconds of the last generator output breaker opening. Since the pen for Generator output Breaker #1 stopped inking when the recorder was shif ted to high speed just prior to the trip, it was impossible to determine the exact time this breaker opened. However, a , review of the recorder traces 'and plant performance as discussed above indicate that the generator output breakers opened at essentially the srne
-time. _ Based on the opening time of generator output breaker #2. Nuclear Flux was below 15 percent at 1.325 seconds.
In addition, a review of the strip chart recording showed that the plant control systans responded satisfactorily to this transient and no changes to control system setpoints were required to improve plant response. Table 4.3.5-1 suntnarizes the data from this test. l O 4.0-38 _ _ _ _ _ _ . ._,_._._ _. _ 2 ._... - _ ,_ _ , . _ - . _ . -
. . - ~ . . . . _ . _ _ - . . . _ - _
. _ . ~
t TABLE 4.3.5-1 PINE TRIP FROM 103 PERCEtR IUdER
- O- TEST DATA SUmARY Plant Bafore During Transient After Parameter Trip Minimum Maximum Trip i Tava A4ctioneered, OF 588.2 555 538.2 555 Tref, 'F 588.5 - - 557 1 Delta T - Loop 1, % 99.5 4 99.5 4 !
- Overpower Delta T Setpoint - toon 1, 1 108 - - 108 Overtemperature Delta T Setpoint - Loop 1, % 110 - - 146 -
Pressurizer i Pressure, psig 2230 1990 2255 2255 Pressurizer Level, % 62 27.5 62 27.5 Narrow Range Steam Generator Level - Loop 1, % 48 0 48 16 O : O- ' 4.0-39
- . . , _ , . - . , , . , _ . . . . _ . . . _ . . . , _ , . . . , , , . . . . , _ . ..,-,...._,...y-,,,..,.,...r,,,,.-.,._-m,. . ~
. _ . . - . - . - - - ._ _ _ - _ - ~ - _ _ - . - . - - - _ . - . . . _ _ . _ _
4.4 INSTRUMDfrATION CALIBRATION AND ALIGNDff O During the power ascension test program, a serles of tests was performe3 to calibrate and align various plant control and instrumentation syster.s. The first test discussed determined thermal power as well as collecting statepoint data at various steady state power levels. The data collected was then used for the calibration of the sten and feedwater flow instrumentation and startup adjustments of the reactor control systen. In addition, testing was performed to operationally align the nuclear instrumentation systan arr3 the process tenperature instrumentation. O i O 4.0-40 l l
~ . _ . , . . . . . . - . . - . - , . . - . . - - . . - . . - - . - - - - - - . . . - - - - - - , - - - - - - -- - - - = ~~- ~ - - - -
4.4.1 '!1[ERMA!, l'OWER MEASURIM2R AND STATEICI!U (O V DATA CO[D CTION 1 This series of tests had three purposes:
- 1) To provide a method for determining reactor thermal power,
- 2) To collect control and protection instrumentation calibration data at '
steady state power levels (statepoints) during the power ascension test program,
- 3) To determine ES flowrate using calorimetric data (50%, 75%, 90%,
100%). After the initial test setup was verified during the post core loading pre:ritical testing (section 2.7), data was collected at 30%, 40%, 53%, 75%, 90%, 98% and 100% power. The control and protection instrummtation data cerved as an input to the calibration of the steam aM feedwater flow instrumentation (section 4.4.2) and to the startup adjustments of the ratetor control systen (section 4.4.5) . Tesc instrummtation was installed to nonitor feedwater flow to each steam geserator and steam pressure in each steam generator. To improve the po:ential accuracy of the test data, the range of the feedwater flow test itstruments was incressa5 as power was increased. The initial set of O ia4tromeate were o-5 reta waite the ria t et ee tastr==e"e= ~~e o-so reia-Us 'trJ the information fran the test instruments as well as feedwater tetperature from process' instrumentation and the ASME Steam Tables, the enthalpy rise across each steam generator was calculated. Steam quality was assumed to be 1.0 (see section 4.5 for actual noisture carryover) and for all but the 100% power test steam generator blowdown was isolated. After correcting the heat input to the steam generators for the actual EP heat input as determined during the preoperational hot functional test, the actual percent of design power was determined. RCS flow was detennined using calorimetric data at 50%, 75%, 90%, 98% and 100% power. The flow calculation used the thermal heat output for each TCS loop as determined by the calorimetric, T and T for each loop, the enthalpies of the water in the CS bot lehand coNegs, and the specific volume of the cold leg water to detennine the RCS loop flow. The total RCS flow was the sum of the loop flows. The data is sumnarized in Table 4.4.1-
- 1. RCS flow was greater than the minimum allowable flow at all test plateaus.
O 4.0-41
- , . _ _ . _ __ _ _ . _ _ . __ _ . _ - _ _ _. _ _ _ _ . _ _ ~ . _ , _ _ . _ _ _ _ _ _ . _ _ _ -
l t TABt2 4.4.1-1 - f RCS FIDd FROM CALORIMfi7TRIC MFASUREMEtTP
~
Nominal Power Laval Loop 50% 75% 90% 981 1 100% 1 106,795 gpn 105,648 apn 108,107 gpn 105,480 gpn 103,890 opn 2 103,069 gpn 106,374 opn 105,484 gpn 101,621 ann 99,210 gpn : 3 105,036 gpn 103,521 apn 103,813 gpn 103,401 gpn 100,460 apn 4 103,796 gpn 101,778 gpn 103,665 opn 102,939 apn 102,150 gpn Total 418,696 gpn 417,321 cpn 421,069 gpn 413,521 atrn 405,710 gtrn M ceptance Criteria >382,803 gun >388,542 gpn >388,542 gpn * >389,547 otxn
* 'I'1e test at 98% power was performed for information only, therefore there was no acceptance criteria in the procedure.
O : b I O 4.0-42 I m ~+ , _ . s.v., ...+.,_..,..#... . . , , ..r,._ .,,m.,. , _ _mm..,,,.,_.,_- _.m. , ., ,.-_._.,~.,_,__.-_.__..,..,._.,_.,y_ . , ~ - , - _ - , . _ , . . _ . -. . . , - ,
. _ ._--. - - _- - - - - - - - . . - ~ - - . . - . - . _ . . - _ - _ - - _ -
1 1 4.4.2 CALIBRATION OF STlW1 AND Fl'.lWATim P!N INSTlRMNTATION The purpose of this test was to collect data during the power ascension testiry; to allow the calibration of the steam flow transmitters against feedwater flow. Test instr'..mantation was use3 to measure the fetdwater flow clanent differential pressure ard the flow calculations were perfotmo3 as part of the thermal power measuronent (section 4.4.1) . In addition, the flow data collected was camparn3 to design values for stean flow ord fee 3 flow. i During the post core loading precritical testing, the static zero shift of the installn3 instrumentation was verifie3. One steam flow transmitter had to be replaced (Secti. n 2.7) . The ranainder of the testing used data collected at 30%, 50%, 75% and 104 power plateaus during the thermal tower measuremmt test (Section 4.4.1) . At 30% and 50% power, data was accumulcted and it was verifie$ that the steam flow /feedwater flow mismatch alarm did not actuate. After the 75% data was analyzal, the steam flow transmitters were respannoj and additional i data was co11 ecto 3 Also, the Tref program was adjuste3 prior to the second set of data resulting in higher steam pressure and lower flows for a given - power level (section 4.4.5) . The steam flow /feo3 water flow mismatch did not actuate. The 100% power data was collected with steam generator blowdown in service. Aftet correcting for the blowdown flow, it was detenine3 that the following acceptance criteria were satistied
- 1) The steam flow /feedwater flow mismatch alarm did not actuate,
- 2) Stoan flow indication on the main control board was within + 4.0% of feedwater flow indication on the main control board, ~
- 3) Plots of fea3 water flow fran the test instrumentation versus fee 3 water flow from installed plant instrumentation were within +2.5% of the ideal (design) curves,
- 4) Plots of feedwater flow from the test instrumentation versus steam flow from installed plant instrumentation were within +3.0% of the ideal (design) curves.
Although all acceptance criteria were satisfie3, the 100% power data was analyrai to determine optimum spans for the steam flow transmitters. The spans were adjusted after the completion of the test program and additional steam flow and fee 3 water transmitter output data was colle:tel at 100% power. O 4.0-43 t l . , . . -. -, - , - - - - - - - - - ~ - - -- -r- -- -~
4.4.3 OPLRATIONAL ALIGtECfr OF NUCLPAR INSTRlMNTATION O The Nuclear Instrumentation Systm (NIS), because of its importance in monito'ing reactor operation, was checkal and calibrato3 throughout the startep test program. Earlier sections of this report have detailed NDI testing which was performed before initial fuel load, during post-core load testing and during initial criticality and low power physics testing. This section sunmarizes all testing and calibration done on the NIS throughout l the test program including the additional testing done during power ascension testing. i PRIOH M FUEL IDAJ Test!ng before fuel load was done in three separate phases as described in Secti >n 1.0 First, an extensive preoperational-type functional test was performed on all tFc circuits, alarms, bistables and meters in all three ranges of the NIS. Second, the high voltage, discriminator voltage and preamplifier setcings were determined and the source ranges calibrate 3 accordingly. Finally, just hours prior to the movanent of the first fuel assembly, analog channel operational tests (ACOT) surveillance proce3ures were performed on each source range channel. PDST CORE IDAD TESTING As detailed in Section 2.8, additional functional testing was done on all of the NIS .. .lizing channel calibration surveillance procedures. INITIAL CRITICALITY Just prior to initial criticality, analog channel operational test (ACOT) surveillance procedures were performed on each intermediate and power range channel. After initial criticality, readings were taken to determine the amount of overlap betwecn the source and interme31ste range. This data is shown in Table 4.4.3-1. KMER ASCENSION TESTItX3 The power ascension testing had three main objectives:
- 1) To determine overlap between the intermediate and power ranges,
- 2) To plot power range detector currents versus reactor power,
- 3) To adjust power range indication to agree with secondary calorimetric calculations.
Overlap testing began at 3% power and continue 3 through 100%. Readings were taaen on both intermediate and power range channels to verify that there is at least 1 1/2 decades of overlap between the two ranges. The overlap data is shown in Table 4.4. 3-2. O Power range detector current and reactor oower readings were taken from 10% to 100% power to verify detector linearity. Figures 4.4.3-1 through 4.4.3-' depict curves plotting detector current against power and shows that all 4.0-44
. .. TABLE 4.4.3-1 A ' NUCLEAR INSTRUMENTATION OVETd.AP DATA k.)= SOURCE RANGE AND INTERMEDIATE RANGE t
A First Positive Intermaliate Indication On Rangeyndicates mediate Range (>10,gter- Amps) 10 Amps SOURCE RAtEE Channel N31 - - Main Control Board 8 x 10 2 1 x 10 4 NI Drawer 1.25 x 10 6.5 x 10 3 Channel N32 - - Main Control Board 8 x 10 2 1 x 10 4 NI Drawer 1.25 x 10 7 x 10 3 INTEPMEDIATE RANGE
-n '-
O Channel N35 - - Main Control Board 1 x 10~11 1 x 10~10 NI Drawer 1 x 10~11 1 x 10-10 Channel N36 - -
' Main Control Board 1 x-10'11 1 x 10'l NI Drawer j 1 x 10~l 1 x 10~
O ' 4.0-45
O O O TABLE 4.4.3-2 fiUCLEAR ItJSTRUMEtTTATIOt1 OVERIAP DATA Iri"MiciEDIATE RAtEE A!JD P3 DER RA!EE INTERMEDIATE RAfCE 01 10% 30% 40% Channel 1135 -4 -4 Main Control Board 7 x 10
-6 3.5 x 10 -5 1.7 x 10 3.0 x 10 tJI Drawer 6.5 x 10~ 5 x 10~ 1.5 x 10 2.8 x 10~
Clunael N36 -5 ~4 -4 tiain Control Board 7 x 10
-6 4.0 x 10 1.9 x 10 3.0 x 10 i
-0 -5 g,7 x 7g 4 2.9 x 10 tJI Drawer 6 x 10 5 r 10
?
o IOWER RAtJGE S Channel tJ41 10% 34% 43% t-tain Control Board 1% 1.5% 10.1% 34% 42.5% tJI Drawer . Channel tJ42 42% 10% 36% Main Control naard !.% 1.5% 11% 371 42.51 tJI Drawer Channel T143 44% 1 1 ", 36% riain Contro! naard 1% 1.5% 10.8% 35.25% 42.751 tJI Drawer Channel tJ44 35.5% 42% 01 10% tiain Control Board 1.si i11 36% 42.5% i tJI Drawer
O
~
O O
~ TABLE 4.4.3-2 (Cotfr)
NUCLEAR INSTRUf4ErlTATIOt3 OVERIAP DATA INTEIC4EDIATE RANGE Ar#) POWFJI RAICE l INTERMEDIATE RAtGI 50% 751 90% 100% Channel N35 _4 Main Control Board 3.5 x 10 4.5 x 10 6 x 10 4 6 x 10 4-NI Drawer 3 x 10 4.5 x 10 5.5 x 10 6 x 10 Channel N36 Main Control Board 4 4 4 4 3.5 x 10 5 x 10 6 x 10 7 x 10 r31 Drawer 3.2 x 10 5.1 x 10
-4 6 x 10
-4 -#
7 x 10 IUdER HAtJGE u
"a Channel 1141 Main Control Board 49% 75% 92% 100%
NI Drawer 481 75% 92% 100% Channel N42 Main Control Board 49% 75% 92% 100% NI Drawer 481 75% 92% 100% Clunnel IJ43 Main Control Board 491 _ 76% 93% 101% NI Drawer 481 75% 92% 100% Channel N44 Main Control Board 491 75% 92% 100% NI Drawer 48% 761 92% 100%
'tyrfE: Overlap realings at 0%, 10% and 30% were taken lefore Power Range Gain was aljustcsl to nutch sceorrlary calorianetric.
( FIGURE 4.4,3-
. ,r _ _
CHANNEL CURRENT VS. REACTOR POWER CHANNEL N41 800 - l 1 l . I i x
<1 i
E00 l
----,. , ' - l
! /i i t .
TOTAL I i r , g s i j . j ! 9 400 X
@ I-
/ l l
? / l l o / ,
,b /
/ ! !, i.
4
'/ i l
L I
/ b LOhER l l300 j/
X/
/
l l
!, 7 - jj t
?
.-! .4 j
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g A 1 ; y / ,q',/y , UPPER l
- 5 200 ' ' "
Q /l l / i /* f.
/ l
/ N/,s / ,s \,
X.' x/
.- C/.// \ 1 l f 00 ,
/ . 4/
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i l i , I I
/
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/ l !
/ /
/ / i
/' $ i I p ,
0 ./ l 0 10 20 30 Jo 50 so 70 80 90 100 t
}-
REACTOR POWER (percent) 4.0-48
. . . ._- - . ~ - - . . _ . , . .- _ .. . ..- . -
FIGURE 4.4.3-2
-0 CHANNEL CURRENT VS. REACTOR POWER .
CHANNEL ' N42 600 .-. I I i , t I i l , ! !k ! 500_l_ T i i l-2{ / l l I ,' i
. TOTAL i r
i 1 s x , i i
+ ,, t i l / i ,
i ! j j / I' 4 { ' -' 3 400 , ._., 1 C3. i EE to < i. I t 3
/
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n 300 ' u ' < i \LOWERt , X' I I Lo i i / t l ,
/fi -
,+
, 1 ,
, l ',
U fI I I l l,g a L3
- !i
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./f
+'
j
- UPPER y 200 _
I ! ,. ' ,,r"
/
I o i ,
, , . t
/ / '
l i I l-s ., x/ s c/ / ./ Ve 4X l I i I \ 1
-/ ./,4/
< 4 l
i
* .l l -100 ' / _ .1-- ._
l
/ o //, ,1, , l l ,
L
/
/ ,.p- /+/l ! t {
I t, . . i <
/ x,j /; i j !
l \ l
//1 i 0 E ! .L I
- 0 10 20 30 40 50 60 70 80 90 C0 1
REACTOR POWER (percent) 4.0-49
-c ~-
J 1 .w ey- -tv- %----mr.w+Wmy 7 ----'s----p-- ,g'Tm4-+- -- ** ----+W.g v gr--- +eee---ea m *54 +-m.+== a----g e Twy N ew we+wT9 +9'r
FIGURE J.4.3-3 0 CHANNEL CURRENT- VS. REACTCR POWER CHANNEL N43 600 . i x I .
/ / -,I t i i i 1 i.
l X 500 2 f/- i-t j i , i 1 < ' i i l TOTA. t l / Q 400 l l'<x/ r i m ,
'd -
t i
- ,/ i O < t c j Q /
/ 1 8
k /
/ ! 0
,/ LOWERi ,..-
l l \ y n l j 'O
, i i Q _f }_ ,a N /
jx i ,e i Y ! ,o' ,
/ yhpEp h \ /:' ,,
./ l 200 '
}O / ,,./ i '
l i i x/ y ,-l i i
/ i
/O't[; '
i I i ' 100 '
/ I l
,- f p4 ' l l'
+
\
/
of l l i l
/j "V I l l
,,- I i ; ! i
'Ix //
l er l j l I { ! !i l~ 0 - 1 __4 . 0 10 20 30 40 50 60 70 80 90 100 REACTOR POWER (percent) 4.0-50
.- . . . _ _ _ _ . _.- . ._ _ _ . _ m._._-.._. . ._
FIGURE 4.4 3-4' CHANNEL CURRENT VS. REACTOR P0HER : CHANNEL N44 600 I I i
}
/4
, /
I i 'sx i 500 1
' I l i l l* ! l-
/l ;
l . TOTAL i : ; 4 l ,
/ i I , i
+
3 400 I ! I i
'/x I o <
/ I l
t p .:
.o ,V '
\
& /
l
; j LOWER ,/}
i i 5 e 0' i
/
4 / j X l
,/
/ ./i,A 'I l UPPER l
/ c' 200 ,
/',- i <
y" l /
/ /
,/['-
[
\
l l l
.I i Nh,Y, l l
/ / '
/.'l.V l i goo i A I I I j
/ if '
l i x -
/' l i
/~ j i ,
O!/ l O- o 10 20 30 40 50 60 70 80 90 100 l REACTOR POWER (cercent) 4.0-51 j
.. . . _ _ _ . - _ _ . _ _ _ _ _ _ - _ _ _ _ . . _ _ _ . _ _ . _ . _ _ . . , _ _ ~ = detectors responded linearly.
V Beginning at-30% power, the power range channels were adjusted to' match
. reactor power as calculated using a secondary calorimetric. This adjustment
-is required by Technical Specifications to be done on a daily basis.
Therefore, normally the operations surveillance procedure was used to make tnis adjustmmt. A few times, however, the adjustments were made based on a calorimetric value from the Startup Test Program. Either method was acceptable and the test program was mar 3e trore flexible by allowing either type of calorimetric to be used to make the power range adjustments. When the plant reached 100% power, additional testing was performed to make final adjustments to the NIS. The high voltage settings on both the intermediate and power range detectors were verified acceptable by plotting a plateau curve. _ The intermediate range 25% reactor trip histables were also reset using the actual 100% power intermediate range current readings. Af ter the plant was trippe3 frcxn 100% power, a test was performe3 to set the intermediate range compensation voltage and verified the source range detectors had survived 100% power operations by completing a voltage plateau check on each source range detector. O l l g i l LO 4.0-52
- ,..,r,.~,, , ,,--+c, ,,--.,-,,,--v,a -
,n.,.e . .-~ .v.y--- - - --e- ,
4.4.4 OPERATIGIAL ALIGNFRP OF PROCESS .& TFMPERA'IURE INSTRUMFNPATION V The process instrumentation receives temperature data from the primary coolant systan RTD's. This data is use3 to generato delta T and Tavg values for use by the plant control and protection systens. Prior to initial criticality while at hot, isothermal conditions (557 +20F), an alignment was performed of the delta T *nd Tavg process instrumentation. See section 2.7 for a discussion of this test. During the power ' ascension test program, data was collected at a series of power plateaus to align the delta T and Tavg instrumentation. The naninal power levels were 30%, 50%, 75%, 90% and 100%. With the plant at steady state conditions, voltage readings were measured ard recorded for ea;;h protection channel T T and Tavg at the applicable protection cabinet . TherecordEv,oltShDes were then converted to tenperatures and the measured Tavg was compared with the Tavg calculated from the measured T and T y Also at all power levels except 100%, T and T spare' k,tD resisk2$e,s were, measured for each protection cabineh These bsured spare R7) resistances were then converted to tanperature and camparei to the applicable loop T and T The core exit E values were also recorded as additional inf b tion COW. At 30%, 50%, 75% and 90% power '.evels the following acceptance criteria were OLJ satisfied:
- 1) The0 1 p Tavg sumnator output converted to0 F shall agree within +
0.5 F of the Tevg value calculatai f rom the measurements of the loop operational Tg and T s, C
- 2) ThehoopoperationalTyRTD output converted to F shall agree within
+1.27 of the value canputo3 from thg measured output of the loop Installed spare T gRTD converted to F,
- 3) The foop operational T RTD output converted toO F shall agree within
+1.2 F of the value co:gftputed from thg measured output of the loop Tnstalled spare T RTD converted to F.
C At 75% power, using the temoerature and calorimetric data from the previous tests as well as the 75% power test, the 100% power loop delta T's and Tavg's were axtrapolated: Delta T, F Tavg, F Loop 1 56.28 583.3 Loop 2 55.56 582.4 Loop 3 56.97 583.6 O Loop 4 56.37 583.6 s 4.0-53
The loop delt3 T's compared reasonably well with the expected best estimate value of 56.8 F quoted in design documents. The extrapolata3 maximum loop _ _O'- Tavg for 103% power fell outside the original acceptance criteria but further test data at 75% pawer shawed that it was satisfactory (See Section 4.4.5). As a result of the extrapolated 100% power delta T's, new gains were calculated for the delta T sumator cards: Loop- New Delta T Sumator Card Gain 1 1.45 2 1.44 3 1.40 4 1.41 0 At lgs % power, no loop Tavg exceeded 588.5 F and no loop delta T exceeded 59.4 F. However, to bring the loop delta T sumator output converted _to % power within +1% of the calorimetric power (100.1%), it was necessary to increase the outputs of loops 1, 3, and 4 delta T sumators. This was accomplished by calculating new gains: Loop Original New Delta T Sumator Power Card Gain 1 97.5% 1.445 3 98.63% 1.420 4- 98.63% 1.451 In addition, although not -required by the acceptance criteria, the output of the loop 2 delta T sumator card was optimized by adjusting the gain to 1.459 (original power 99.15%) . Table 4.4.4-1 sumarizes- the temperature data at the 100% power plateau. I i i f3 l= V i 4.0-54
O 0: O TABIE '4.4.4-1 TU4PERATURE AI,IONENT DATA AT 100% . POWER Inop H/E Convertor Delta T Sumnator: Delta T Tavg Sumnator Tavg No.- Ogtput Outixit g Calglated Ogt[xit - Calgalated P Power (%) .F F F F t 1 Tg ,, 615.3 97.5 54.87 55.4 587.6 587.6 T COfD 559.9 , 2 T - 614.4 99.15 55.09 54.9 587.1 587.0 Y ,'D 559.5 COE 3 T, ,,,, 616.5 98.63 56.19 56.5 588.2 588.3 Ty{g 560.0 e g 4 T, m., 615.'4- 98.63 55.59 55.7 587.5 587.6-T]{D 559.7 i Calorimetric Power = 100.1% 4 b
_ . _ _ ._ _ _ ~ _ _ _ . _ 1 I 4.4.5 STARTUP AIlJUSmEMPS OF 'nlE REAC'IOR p COMPROL SYSTEN ()- i The reactor control system in the automatic mode positions control rods to maintain Tavg in the reactor coolant systs a5 ^ r*fer*"o* t" P*rature' Tref. If Tavg is different from Tref by +_1.5 F or mre, the reactor control system steps the control rods to restore Tavg to the desired temperature band. The reactor control syste uses the highest loop Tavg for comparison to Tref. Turbine power based on first stage turbine impulse pressure and nuclear power signals are used to form a power mismatch contribution to the rod control system. The combined temperature error signal (Tavg-Tref) and the power mismatch signal determine rod speed as well as direction of motion. The purpose of this series of tests was to determine the Tavg program resulting in the highest possible steam pressure and thus optimum plant efficiency without exceeding gressure limitations for the turbine, or the design full power Tavg (588.5 F) . As discussed in section 2.7, basgline data was collected in the postcore loading precritical tests at 557 E and 2235 psig. Additional data was collected at 30%, 50%, 75% and 100't power in conjunction with the thermal power measurements, section 4.4.1. Data collected for each loop included T Tavg, fee 3 water flow, and steam generator pressure. Firstshe,Ttbh2n,e impulse pressure was also recorded as well as Tref, auctioneered Tavg and auctioneered NIS power. After the 75% data was collected, first stage turbine impulse pressure, Tavg Oi and steam generator pressure were extrapolated to 100% power with the following results: First stage turbine impulse pressure ~ 685 098i# Tavg ~ 584 F Steam generator pressure ~ 935 psia Since Tavg and steam generator pressure were extrapolated to be low, the gain and bias for the control cards TY-505A and TY-505E were checked. The bias on TY-505A was lower than the required value of 7.921. Additional data was then taken and extrapolated to 100% power: First stage turbine impulse pressure ~ 691.8 gsia Tavg ~ 587.5 F Steam generator pressure ~ 990.5 psia At 100% power, the final se5 f data was collected with Tree cound to se slightly greater than 588.5 F. TgeoutputofcontrolcardTY-505Awas adjusted to correct Tref to 588.5 F at nominal 100% power. After this cberection, steam generator pressures were: Loop 1 1007 psia Loop 2 1010 psia Loop 3 1004 psia Loop 4 , 1010 psia 4.0-56
iAt the same time, first stage turbine impulso pressure was an average of 712.8 psia.- Various control cards were adjusted to match the actual turbine
-: impulse pressure vs. plant power. The steam generator pressures at 100%-
pawer were determined to be satisfactory.. As a result of this. test, the' reactor control system was adjusted to provide a sufficient supply of steam at rated pressure to support 100% power operation. O: 0 4.0-57
4.5 STEAM GENETWIOR MOISTURE CARRYOVER
-%J
() MEASUREMENT This test was performed to determiJe the average moisture carryover content in the steam from the steam generators at 100% power. The radioactive tracer method war used to determine the rmisture carryover. With the plant operating at 100% steady state conditions, a one curie liquid radioactive tracer (sodium-24 in the form of a sodiu:u nitrate solution) was mixed with approximately 20 gallons of danineralized water in a tanporary mixing tank and then injected into each steam generator feedwater line using the four feedwater hydrazine amnonia addition pumps. The feedwater lines transported the radioactive tracer into each steam generator. A large volume of demineralized water was then injected to flush out the chenical addition lines. Steam generator blowdown flow had previously been secured to prevent dilution (and loss) of the radioactive tracer. Also, the condensate polishing system and the condensate makeup reject line back to the condensate storage tank had previously been isolated to prevent dilution (and loss) of the radioactive tracer. After a 30 minute stabilization perkd to allow mixing of the radioactive tracer in the steam generators and arryover of the radioactive tracer with moisture into the condensate /feedwater systems, three sets of samples cere taken at 15 minute intervals from each of the four steam line probes, from each steam generator upper shell, and from the main feedwater systen. The samples were then analyzed for sodium -24 activity. Using the results of the sample analysis, a percent moisture carryover was calculated for each set of steam generator upper shell samples and steam line samples. Since the steam generator upper shell results were considered to be more accurate, the steamline samples were used as a backup. This test was satisfactorily performed on August 10, 1985 with the plant at 99.74% power as determined by the plant calorimetric procedure. The moisture carryover results are sumnarized in Table 4.5-1. The average l moisture carryover was 0.015% as ecmpared to an acceptance criterion of I 0.25%. 1 l O 4.0-53
l TABLE 4.5-1 STEAM GENERATOR MOISTURE CARRYOVER TEST RESULTS Sample Point Sample Set Percent Carryover Steam Generator 1 .0166% Upper Shell 2 .0168% 3 .0122% 4 Average .015% Main Steam Probe
- 1 .0167%
2 .0168% 3 .0122% Average .015% O 4.0-59
4.6 NSSC ACCEPTA!CE TEST O v The purpose of the NSSS Acceptance Test was 1) to danonstrate the availability and reliability of the Nuclear Steam Supply Systs and 2) to measure the NSSS power output. The test did not verify any safety criteria, but rather verified the NSSS vendor had supplied an acceptable systan for contractual and warranty purposes. The reliability of the NSSS is dmonstrated by maintaining the plant at ratal output for 250 hours without incurring a load reduction or plat. *rio due to a NSSS malfunction. Ideally, the 250 hours should be continuous uninterruptai operation. However, it is acceptable to have only 100 hours of continuous 100% power operation with the remaining 150 hours being accumulated time rather than continuous. Power may be reduced during the 150 hour portion, but the accumulation of time is stopped until power is returned to rated output. The measursent of the NSSS output is acl.ieved by calculating the enthalpy rise across the steam generators. This enthalpy rise is determined by measuring the inlet feedwater flow, tapperature and pressure and its outlet steam pressure. The steam flow is considered to be equal to the fee 3 water flow since steam generator blowdown is isolated during the test. The reliability portion of the test comenced at 1607 on August 3,1935 aM 100 hours of continuous operation was achieved at 200 on August -12,1935. ^ At no time during that period was power reduced beloa its acceptable value of 3425 & 0,-5% M M. I:miediately following the completion of 103 hours of operation, the plant was reduced to approximately 60% power to investigate vibration problans with one of the main feedwater pumps. The pump vibration problami was corrected and at 0630 on August 21, 1985,
. power was returned to the acceptable range D254-3425) for testing and the additional accumulation of 150 hours began. On August 22, 1985 power was reduced below the test band for 3 hours and 14 minutes during the performance of another test procedure.
The measurement of the NSSS output portion of this test was conducted on August 23, 1985. Four sets of feedwater and steam data were collected over a four hour period. The results proved to be very consistent and well within the acceptance criteria of 3425 +0, -2% 3R with the calet. lated values being 3417.5 Mc, 3410 MC, 3422.4 Mc, and 3416.7 MC. These values also easily met the additional requirements that each calcul:.ted value be within 11% of the average of the 4 calculated calorimetric values. The variation frcm the average was less than 0.2%. The total accumulation of 250 hours was officially signed complete at 1634 on August 27, 1985. Canputer trends and operaticas calorimetric procedures were used to verify that the power output remained inside the acceptable range during the test. The 250 hour run was completed without any NSSS malfunction and the only problan encountered during the test was due to secondary plant. O 4.9-60
. ., .- . . . .- . - - -- .. _ . ~. - . _ -
L The test was considered successful in that it proved the NSSS is capable of pl : . sustained operation at rated output and is capable of producing an output at s
. the warranted rating- of : 3425 BWP. Though not a strict test acceptance criterion, the - test was - also ustd to document - that the secondary side of the -
. plant was capable of operating.at 95% of its rated electrical output.
O 4.0-61
-4.7 IWER ASCENSION 'nG2 MAL AND DYNN11C TEST O
v The purpose of this test was to monitor those systems or portions of systens that could not be monitored during the preoperational hot functional test thermal expansion prc> gram because systems did not reach normal operating tenperature and also to remonitor those points that did not meet the acceptance criteria for the preoperational hot functional test. The testing included:
- 1) Demnstrating that the following systems (or portions of systems) which were not monitored during hot functional testing are free to expand thermally as designed:
a) Main Steam System, from the main steam headers to the condenser via the condenser dump lines, b) Main Steam System, from the main steam header to the the steam generator feedwater pump turbines, c) Main Feedwater Svstem, from the steam generator feedwater pumps to the steam generators,
- 2) Monitoring the dynamic response cf the main steam system to a plant trip from 100 percent power, p 3) Remonitoring the snubbers and spring hangers whose measured movementa V during hot functional testing were outside of acceptance criteria or associated piping did not reach the normal operating temperature,
- 4) Visually monitoring (measurement, if required) stex1y-state vibration of pressurizer surge piping with reacgor coolgnt syste.n primary loop at nonnal operating mode (RCS at 557 F + 10 F, four FCP 's running).
For the thermal expansion portion of the test, 81 lanyard transducers and 37 resistance tenperature detectors (RWs) were installed at selected locations to measure piping movements and corresponding temoeratures in the main steam and main feedwater systens. These instruments were connected to a data acquisition sysram (DAS) provided by Westinghouse where data was collected ard printed out at the selected tenperature plateaus. At these temparature plateaus, walkdown inspections were performed to verify that no piping was being restrained, other than by design, from thermal growth, swing clearances were checked on all required snubbers and snubber and I spring hanger settings were recorded, t cor the dynamic response to the 100% power plant trip, 41 lanyard transducers and 4 pressure transducers were installed at selected locations in the main steam system. The instruments were connected to the Westinghouse DAS where the dynamic response for each channel was recorded on a FM tape recorder and a digital data acquisition system. O 4.0-62 l l
...- . ~ . . - - -.
This testing was begun with the start of the post core loading precritical test program and continued through the power ascension test program until O. the 100% power plant trip was performed on August 28, 1985. During this time, all monitored systes were heated up to normal operating tmperature and cooled down to ambient. Thg one exception was the main feedwater system whichwasheatedupto340__+10F,cooleddowntoambient, heated"Pt o 440
+ 10 F (normal operating temperature) and cooled down to 340 + 10 F. Pre-test ambient data, intermediate heat up plateau data (when plant conditions allowed), normal operating temperature (hot) data and post-test ambient data were collected. All data was reviewed an3 approved onsite by a Bechtel stress engineering team. All acceptance criteria pertaining to thermal expansion were satisfied.
A visual inspection was performed of the pressurizer surge piping and displacment and velocity data was taken at various locations using a vibration monitor. The ICS was at normal operating temperature and pressure with 4 ICP's operating. The highest peak to peak displacement was 3 mils with a velocity of 0.1 in/see giving a frequency = .1 in/sec = 33.3 Hz.
.M3 in The data obtained were within allowable limits.
For the 100% power plant trip, data was collected on the N tape recorder 2 minutes prior to the trip and 3 minutes following the trip. Data was collected on the digital data acquisition system approximately 30 seconds prior to the trip and 60 seconds after the trip. The tima history plots for each channel were obtained for the period of the monitored transient. These Plots were evaluated by Bechtel Stress Engineering. This evaluation .O c "ctoaea thee waite ett vietms move emte were witaim the ettow m=e e applicable codes, some additional pipe supports should be added to the steam dump piping. This will be accomplished as permitted by plant conditions. O 4.0-63
4.8 BIO [DGICAL SHIELD TESTING O The purpose of this test was to measure and record the neutron and gama-ray radiation levels in accessible areas of the plant where radiation levels above background were anticipated and to determine locations, if any, where shielding was deficient thereby. ensuring that plant personnel would not be subjected to overexposure from radiation as a result of inadequate shielding. The test procedure established the Health Physics requirments for the biological shield survey point selection and survey techniques as well as methods of documentation. To meet the requirenents of this test, a series of four biological shield surveys were performed during the period frm May 17, 1985 to August 9, 1985. The first survey was performed on May 17, 1985 prior to initial criticality. This Preoperational Survey was intended to provide baseline data and demonstrate that no sources of radiation were present that would effect subsequent surveys. The survey was successfully performed with no abnormal findings. The Low Power Survey was performed on May 24, 1985, with reactor power at 3%. No unexpected radiation readings were noted. General readings taken inside steam generator labyrinths were in excess of 100 mrm per hour. The containment hatches were posted as High Radiation Areas in accordance with O tocra2o "a 9tocea" ret re,"treme"te- The re at"a= e ""a were witai" the expected ranges due to N-16 shine from primary piping. The Intermediate Power Survey was originally started on July 6, 1985, in containment. This portion of the survey was terminatc4 due to extensive flux mapping interference with gama readings. The survey outside containment was performed on July 8,1985 with reactor power at 49.4%. One abnormal neutron reading near two electrical penetrations was noted. The area was resurveyed with another instrument and no neutron readings were detected. It is believed that m interference affected meter deflection. The containment portion of the Intermediate Power Survey was completed on July 15, 1985, with reactor power at 48%. No readings beyond expected extrapolated values were noted. The Containment portion of the High Power Survey was performed on August 3, 1985, with reactor power at 100%. No unusual radiation roadings were noted. The balance of the survey was performed on August 9, 1985 and no unusual readings outside containment were noted. O 4.0-64
4.9 PLAl# PERFORMAfCE TEST A ~NJ This test was used to monitor performance of various plant systems durina the power ascension test program. Specific test objectives were:
- 1) To monitor the balance of plant and electrical systems under loaded corditions,
- 2) To obtain data to verify the ability of ventilation syst es to maintain ambient tmperature within design limits,
- 3) To verify evacuation alarm audibility in high noise areas,
- 4) To monitor concrete taperatures surrounding hot penetrations.
During the powr ascension test program, baseline data was collected at steady state conditions. Plant systems monitored during this test included: Main Steam Feedwater Feedwater Heater Extraction; Drains and Vents Station Service and Essential Service Water Transformer Electrical Load Centers AC Inverters 125 V AC and DC Containment Cooling ,]L Auxiliary Building ventilation Control Building Ventilation Steam Tunnel Ventilation Auxiliary Feedwater Pump Room Ventilation Auxiliary Boiler Room Ventilation Ra3 waste Building Ventilation Turbine Building Ventilation Cmponent Coolirg Water Steam Generator Blowdown Condensate Demineralizer Water The acceptance criteria of the test were satisfied:
- 1) The audibility of the evacuation alarms was verified throughout the plant,
- 2) The containmegt air coolers maintained containment air temperature less than 120 F throughout the power ascension test program.
- n addition, baseline data was collected for many systems in the plant. One plant parameter was fourd to be significantly outside its predicted range.
Condensate pump C suction pressure was 7.5 psia whereas 3.3 to 5.0 psia was expected. This condition is being investigated but does not impact system operation. O 4.0-65
. . - . . . . . . ~ .
9 4.10 TURBINE GENERATOR TESTS
- O The General Electric turbine generator was checked out with a series of tests which monitored the turbine, generator and associated support systes from the time the turbine was first brought on line until the plant was at 100% of rated power. Testing was performed with the generator off line and with the plant at 20%, 40%, 60%, 75%, 90% and 100% of rated power.
Tha turbine was first brought to rated speed using nuclear power at 0308 on June 12,1985. Turbine bearing vibration was monitored during acceleration to check for critical speeds. While at rated speed, several tests were performed; General Electric performed a checkout of the exciter, the turbine lube oil systen was checked, turbine steady state vibration was recorded, EHC control circuits were tested, the generator core monitor was started up and adjusted for proper flow, and a number of other parameters were monitored and recorded using permanent plant and test instrumentation and the plant canputer. After all exciter checks had been completed, the generator was first synchronized to the grid at 1857 on June 12, 1985. The unit tripped almost imnediately due to a minor problem with the turbine control circuits. The secon3 tinn the generator was synched it again tripped due to a slightly different control circuit problem. At approximately 0230 on June 13,-1985 the generator was successfully synched ani Wolf Creek began supplying electrical power for the first time using nuclear fuel. However, the Q D turbine only ran for a few hours before it was shut down due to high vibration. It was determined that the vibration was due to the fact the turbine was operated nearly- 24 hours without a load. When the generator was finally put on line and the turbine loa _ded down, it cooled off rapidly causing the vibration. After the turbine had been off line and cooled off, it was brought back on line and the generator synched without any vibration problens. l With the generator on line and the load at approximately 20%, additional
- - turbine generator monitorirg was performed. All parameters which had been l checked with the generator offline were again monitored and recorded. A l final checkout of the exciter was performed and the generator hydrogen seal l oil flow was measured. A check of the reverse power relay and turbine overspeed tests were also performed.
Turbine generator monitoring continued at various testing plateaus through 100% power. Parameters which had previously been checked were again monitored and recorded. In addition to those previously mentioned, these parameters included generator stator bar tenperatures, ETC control signal parameters, power load unbalance circuitry, thermal expansion movement, seal and lube oil pressures, turbine performance data, MSR parameters and a number of other miscellaneous parameters. As before, permanent plant and test instrumentation as well as the plant canputer were utilized for recording data. ' Sane additional checks were done at 100% power. The main field and alterrex field carbon brush vibration was measured and a full load check and recalibration of pressure tranducers to reflect actual versus design 4.0-66
. . _ _ . . . . - - . . _ _ _ . _ . . _ . . . . _ . - _ - _ . _ . _ . . . - - . . _ . _ _ _ . . _ _ . _ . ~ . _ _ _ _
-l i
settings was performed. -O
' Although a number of minor problems were found and subsequent 1v mr acted, the only problon of any major significance was m Un, da .ac noise testing. Several different types of measurements were performed to check control valve instability and dynamic noise. The problen was first encountered at 40% power and was seen in each succeed!Ng plateau in varying degrees with each type of measweent. The probl e ha: Seen partially 4 corrected by fixing a ground loop found in the Dic control cabinet. General Electric feels the proble will be fully resolved af ter a filtering circuit is installed in the control valve signal amplifier circuitry.
O O 4.0-67
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4.11 SPFCIAL TESTS Two special tests were performed during the post fuel load test program. The first monitored the performance of the troisture separator reheaters. (MSR's) .The second special test collected baseline data on the reactor vessel level instrumentation syst s (RVLIS).
~,
i 4.0-68
4.11.1 MOIS'IURE SEPARA'IOR RDIFATER TEST The General Electric turbine utilizes four moisture separator reheaters to reheat steam discharging frm the high pressure turbine before sending it to the three low pressure turbines, he steam is reheated by tapping high tanperature, high pressure steam off the main steam line. The~ main steam is supplied to the reheaters in two stages. 'Ihe first stage has no control circuitry. It is manually turned on aM off by operators. The second stage, however, has a more complex control circuit. From 5% to 65% turbine load the reheater outlet pressures are monitored to control main steam flow to the reheaters using a low load control valve. When the turbine load increases above 65% a high load control valve then opens up. (This is an on-off control valve, not proportional control). Since the operation of the reheater was not being checked in the turbine generator testing, a special test was written to monitor operation of the moisture separator reheaters (MSR's), especially the second stage reheat control circuitry while the turbine was being loaded. The test was performed while the turbine was being loaded during startup ad power ascension af ter the 50% plant trip. Both the first and second stage reheat supply lines were put into service usirg the normal plant operating , procedures. Then, while the turbine was being loaded, data was recorded on moisture separator reheater valve positions, inlet tenperatures, inlet O eteeeeree, rtow retes, ooetet temver>teree, ooetet vreeeores, e#a c " trotter outputs. The data was recorded every 5% to 75% of rated load. This data was then analyzed and compared to expected values. The moisture separator aM first stage ' aheater drain tank condenser dump valves closed as expected at approximately 10% load. The second stage reheater drai1 tank condenser dump valves closed at approximately 20% load as expected. The rehert supply line drains closed when the associated reheat supply lines were opened. The secoM stage reheat low load control circuitry controlled flow as expected as the turbine load was increased to 65%. And finally at approximately 65% load the high load valve opened up as it should have. The only problems encountered were some computer points which did not indicate properly. The test showed that the overall performance of the rgheaters maintained all the low pressure turbine inlet temperatures within 50 F as reqaired by General Electric. Though the reheaters met the minimun requirements, it is
- _ felt the overall systen performance could be improved by tuning the system.
This tuning and additional monitoring will be done as a part of normal plant l_ performance testing and monitoring. l O l 4.0-69 l
4.11.2 RPJCIVR VESSEL LEVIL INSTRUMDUATION SYSTEM (RVLIS) RVLIS is a redundant safety grade systen which provides reactor vessel water level indication. The systm utilizes two sets of two d/p cells. These cells measure the pressure differential between the bottom of the reactor vessel and the top of the vessel. Cells of differing ranges are utilized to cover different flow behavior with and without BCP operation:
- 1) Reactor Vessel Narrow Range (Delta PB)
This instrument provides an indication of reactor vessel level from the bottom of the reactor vessel to the top of the reactor during natural circulation conditions.
- 2) Reactor Vessel Wide Range (Delta Pg)
This instrument provides an indication of reactor core and internals pressure drop for any combination of operating BCP's. Comparison of the measured pressure drop with the normal, single-phase pressure drop will provide an approximate indication of~the relative void contant or density of the circulating fluid. The indication of coolant density is significant only when subcooling margin is near zero. This instrument nonitors coolant conditions on a continuing basis during force 3 flow conditions, d To provide the required accuracy for level measurement, temperature measurenents of the impulse lines are provided. These temperatures, together with existing reactor coolant temperature measurenents and wide-range RCS pressure, are enployed to compensate the d/p transmitter outputs for differences in systen density and reference leg density . This would be particularly important during the change in the environment inside the containment following an accident. The RVLIS test collected baseline data on systen operation under a variety of plant conditions. Specific test objectives were to:
- 1) Check the RvLIS wide range compensating function by recording dynamic head and full range level indicator readings with all FCP's running over full RCS power / pressure / temperature,
- 2) Obtain plant specific RVLIS dynamic head and full range level indications for 3,1, 2, 3 and 4 BCP's operation at both cold shutdown and hot standby conditions,
- 3) Record RVLIS RTD and other inputs at selected paints during heatup after fuel load, l
- 4) Observe hydraulic isolater operation during heatup and initial plant operation.
4.3-73
l Data collection started with the post core _ loading precritical test sequence and was completed with the plant at 100% power. (]"- The hydraulic isolators operated per design. With the exception of one data point, all-indications were within the expected ranges. The one exception was for full range indication LI-1321 at a hat zero flow condition where the expected range was 104 +6%._ The actual recorded reading was 111%. Mare difficulty was experienced with agreanent between the full range indicators and with agreenent between the dynamic range indicators. There are two indicators of each type and a two pen recorder with one pen assigned to each type. Required-agreenent for a given type of indication (full range or dynamic) was within 2% for all three indicators. In some cases the actual agreenent was 3% or greater. Although the system operates as designed, investigation is coatinuing to determine the cause for excessive difference between the indicators of a given type. This work will be completed and the systen fully operational by startup following the first refueling. O O 4.0-71
I Ath)CIX A
.p V' QiRONOI/XIY OF ' HIE POST FUEL START'JP PROGRAM March 11,1985 -
NRC issued Iow Power License authorizing initial fuel load and pre-critical testing, initial criticality and low power physics testing. March 13, 1985 - Comenced fuel loading - Entered Mode 6 March 17, 1985 - Completed fuel loading. March 21, 1985 - Reactor vessel studs tensioned - Entered Mode 5. March 26,1985 *
- Reactor coolant system filled.
March 27, 1985 - Initial fuel load procedures approved by PSRC. Received authorization from the Plant Manager to comence post core load precritical testing at 1300. March 31,1985 - Comenced cold control rod testing. April 10, 1985 - Completed cold control rod testing. 0 _ () April 17,1985 - Entered Mode 4 (>200 F) at 0750 0 April 26, 1985 - Enteral Mode 3 (>350 F) at 2300. April 30, 1995 0 RCS At 557 F, 2235 psig at 0500. May 4, 1985 - Comence:1 hot control rod testing. May 7, 1985 - Completed hat control rod testing. May 19, 1985 - Post core loading pre-critical testing approved by the PSRC. Received authorization from the Plant Manager to comence initial criticality and low power testing at 2330. May 21, 1985 - Comenced diluting for initial criticality at 3843 - Entered Mode 2. May 22, 1985 - Reactor critical at 074$. May 31, 1985 - Completed low power physics testing at 1130. June 4, 1985 - NRC lif ted 5% power restriction on license. June 5, 1985 - Initial criticality and low r.cwer physics testing l O verovea 'v the esac- aeceivea 8 ea =i=eti , er = the Plant Manager to comence initial synchronization and 20% pawer test sequence at 1555. A-1 l
i June 6, 1985 - Entered tbde 1 at 2222. I l f} 1- June 13, 1985 - Turbine-generator synchronized to the 9 tid at 0204. - Jene 18, 1985 - Initial symhronization and 20% p wer test coquence complete au 2200 : June 19, 1985 - Testing approved by the PSRC. Received authorization from the Plant Manager to comnence the power _ , ascension and 50% p wer test sequence at 0830. Jur.e 29,1985 - Performed plant shutdown external to the control t room. , t July 6, 1985 - Plant at 50% pwer. July 16, 1985 - Performed rods drop and plant trip test. July 18, 1985 - Power ascension and 50% power testing complete at 0200. , 1 July 19, 1985 - Testing approved by PSRC. Received authorizatier from the Plant Manager to comnence the 73% power test sequence at 1312. July 20,1985 - Plant at 75% power.
.,uly 29, 1985 - '5% power test sequence complete at 0645.
J011 30, 4985 - Testing approve 3 by the PSRC. Receive 3 authorization from the Plant Manager to coaner':n the 901 power test sequence. Augest 4, 1985 - Plant at 90% power. August 6, 1985 - 90% power test sequence complete at 1630. August 8, 1985 - Testing approve 3 by the PSRC. Received authorization from the Plant Manager to comnence the 100% power test sequence. Plant at 100% powe.r at 1607. August 12, 1985 - 100 hour continuous run conglete at 2007 Power reduced to ~ 55% be ause of main feed pump "B" vibration problens. August 21, 1985 - Plant at 100% power. August 28, 1985 - Performed 100% plant trip at 0513. 100% test sequence complete. Q August 30, 1985 - Testing approved by PSRC. September 3,1985 - Unit declared comnercial at 0114 A-2 l
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APPI21 DIX C UNP!ANNTD itEAC"10ft TitII"J DUllI!*3 IMT INIL IIMD Tl:ST PitOGitAM Date/ Time Power Level Cause Low-low steam generator level due to feelwater control ) June 6, 1985/2247 6% - 7% oscillations _j inadvertently opened reactor June 23, 1985/1333 30% trip breaker duriry performance of surveillance procedure Low-low steam generator level July 9, 1985/1115 50% due to test recorder corrections July 10, 1995/0820 ~45% Hi-hi steam generator level caused MMf! and low-low steam July 11, 1985/0230 12% - 15% generator level due to more fea! water control valve leakage. , Loose lug in INU7/lHU3 caused July 23, 1985/0815 75% loss of S/G feedwater pump Positive rate trip on NIS July 31, 1985/0030 75% channel N41 dua to spike with another channel in test TOTALI 7 unplar.ned reactor trips l l l l l w l O C-1
_ . . . . . ___m_ . _ . . _ _ . _ . _ _ . . _ _ _ _ _ , <_._m___. . _ _ _ ,
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UNITED STATES - h . Ef . [ q NUCLEAR REGULATORY COMMISSION g, . WASHINGT ON, D, C. 20656 f %,""" /
. am4 as Docket No. STN 50 482 Mr. Glenn t..;Koester Vice President - Nuclear Kansas Gas and Electric Company 201 North Market Street ,
Post Office Box 208 Wichita, Kansas 76201
Dear Mr..Koester:
Subject:
Issuance of facility Operating L'icense NPF Wolf Creek Generating Station, Unit 1 The U.S. Nuclear Regulatory Comission (NRC) has issued the enclosed Facility Operating License NPF-42, together with Technical Specifications and Environ-mental Protection Plan for Wolf Creek _ Generating Station, Unit 1. Based upon the findings of the Comission as reflected in the enclosed license and: the favorable vote by the Comission on full-power operation, License No. NPF-42. authorizes operation of the Wolf Creek Generating Station, Unit 1 at reactor l (' . core power levels not in excess 'of 3411 megawatts thermal (100% power) -and supersedes License No. NPF-32, issued on March 11, 1985. Enclosed is a copy of a related' notice, the original of which has been for-warded to the Office of the Federal' Register for publication. Fnur signed copies of Amendment No. 3 to Indemnity Agreement No. B-99 which , covers the activities authorized under License No. NPF-42 are also enclosed. Please sign all' copies and return one to this office. Sincerely, 11 > H gh . Thompson r., ector - D sion of Licensing . Office of Nuclear Reactor Regulation
Enclosures:
- 1. Facility Operating License NPF-42
- 2. Federal Register Notice
- 3. Amendment No. 3 to Indemnity Agreement No. B-99 cc w/ enclosures:
{- See next page
- . - - -. -S: +-;MLAj.: t.
-s WOLF CREEK l ',
g4 1935 (~ . Mr. Glenn L. Koester . Vice President - Nuclear Kansas Gas and Electric Company PO! North Market Street Pott Office Box 208 Wichita, Kansas 67201 cc: Mr. Nicholas A. Petrick H5.'Wanda Christy . Executive Director. SNUPPS 515 N. 1st Street 5 Choke Cherry Road Burlington, Kansas Rockville, Maryland 20850 C. Edward Peterson, Esq. Jay $11 berg, Esq. tegal Division Kansas Corporation Ccruission. Shaw,'Pittman, 1800 M Street, N. W.Potts & Trowbridge State Office Building, Fourth' Floor
-Washington. 0.-C. 20036 Topeka, Kansas 66612 Mr. Donald T. McPhee John M. Simpson, Esq.
Vice President - Production Attorney for Intervenors Kansas City Power & Light Company 4350 Johnson Drive, Suite 120
~1330 Baltimore Avenue Shawnee Mission, Kansas 66205 Kansas City, Missouri 64141 -
Regional Administrator Ms.-Mary Ellen Salava U. S. NRC, Region IV (. Route 1, Box 56 611 Ryan Plaza Burlington, Kansas 66839 Suite 1000
.. . Arlington, Texas 76011
- A. Scott Cauger-Assistant General Counsel
- Mr. Allan Mee-
'Public Service Comission ProjectCoordinator P. O. Box 360 Kansas Electric Power Cooperative, Inc.
Jefferson City, Missouri 65101 Post Office-Box 4877 Gage Center Station Mr. Howard Bundy Topeka, Kansas 66604 Resident inspector / Wolf Creek NPS c/o U.S.N.R.C Regional Administrator Post Office Box 311 U.S.N.R.C. - Region III-Burlington,. Kansas 66839 799 Roosevelt Road u Glen Ellyn, Illinois 60137 H Mr. Robert M. Fillmore State Corporation Commission . Brian P. Cassidy, Regional Counsel u State of Kansas Federal Emergency Manage.s. Agency L -Fourth Floor, State Office Bldg. Region I. Topeka, Kansas - 66612 J. W, McCormack POCH Boston', Mass;chusetts 02109
W0t.F CREEK M4 BB5 ( . cc: Terri Sculley, Director' ' Special Projects Division Kansas Corporation Commission State Office Building. Fourth Floor Topeka, Kansas 66612 Mr. Gerald Allen . Public Health Physicist Bureau of Air Quality & Radiation Control , Division of Environment Kansas Dept. of Health & Environment
^
Forbes Field Bldg. 321 Topeka, Kansas 666?0 Mr. Bruce Bartlett Resident inspector / Wolf, Creek hPS c/o U.S.N.R.C Post Office Box 311 Burlington, Kansas 66839 1 l ( l
O JUN 4 1985 WOLF CREEK - OTHER (.
. , l cc: Office of the Governor State of Kansas Topeka, Kansas 66612 AttorneyGeneral 1st floor - The Statehouse Topeka, Kansas 66612 ,
Chaiman, Coffey County Comission Coffey County Courthouse Rurlington, Kansas 66839 EIS Review Coordinator EPA Region VII 324 East-lith Street Kansas City, Missouri 64106 (
- . , . .,m- - .w,- yp q
UNITED sT ATEs [ ft NUCLEAR REGULATORY COMMISSION W A$ mNG T ON, D. C. 70%8
- r. ( ,
(' a..... KANSAS GAS AND ELECTRIC COMPANY ! KANSAS CITY POWER & LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE, INC. DOCKET'NO. STN 50-482 WOLF CREEK GENERATING STATION, UNIT NO.1 FACILITY OPERATING LICENSE License No. NPF-42 l' . The Nuclear Regulatory Connission (the Connission) has found that: A. The application for license filed by Kansas Gas and Electric Company.
-Kansas City Power & Light Company, and Kansas Electric Power Cooperative, Inc. (licensecs), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Connission's regulations set forth in 10 CFR Chapter I, and all required notifica- ,
p tions to other agencies or bodies have been duly made;
- B, Construction of the Wolf Creek Generating Station Unit No. 1 (the facility)hasbeensubstantiallycompletedinconformitywithCon-struction Permit No. CPPR-147 and the application, as amanded, the provisions of.the Act, and the regulations of the Connission; C. The facility will operate in conformity with the applicatinn, as amended,'the provisions of the Act, and the regulations of the Com-mission, (except as exempted from compliance in Section 2.D below); ;
D. .There is reasonable assurance: (i)thattheactivitiesauthorized by this operating license can be conducted without endangering the health and-safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth-in10CFRChapterI.(exceptasexemptedfromcompliancein Section 2D below); E. Kansas Gas and Electric Company
- is technically qualified to engage in the activities authorized by this license in accordance with the Comission's regulations set forth in 10 CFR Chapter I;
- Kansas Gas and Electric Company is authorized to act as agent for the Kansas p City Power & Light Company and the Kansas Electric Power Cooperative, Inc.,
and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
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T. The licensees have satisfied the applicable provisions of 10 CFR Part 140 " Financial Protection Requirements and Indemnity Agreements," of the Comission's regulations; G. The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public; H. After' weighing the environmental, economic, technical and other bene-fits of the facility against environmental and other costs and con-sidering available alternatives, the issuance of this facility Oper-ating License No. NPF 42, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied; and I. The receipt, possession, and use of source, byproduct and special nu-clear material as authorized by this license will be in accordance with the Comission's regulations in 10 CFR Parts 30, 40 and 70.
- 2. Pursuant to approval by the Nuclear Regulatory Comission at o meeting on June 3,1985, the License for Fuel Loading and Low Power Testing, License No. NPF-32, issued on March 11, 1985, is superseded by facility Operating License No NPF-42 hereby issued to Kanns Gas and Electric f
Company, Kansas City Power & Light Company, and Kansas Electric Power Cooperative, Inc. (the licensees) to read as follows: A. The license applies to the Wolf Creek Generating StWn, Unit No.1, a pressurized water nuclear reactor and associated t ipment(thefa-cility), owned by Kansas. Gas and Electric Company, Kensas City Power
& Light Company, and Kansas Electric Power Cooperative, Inc. The facility is located in Coffey County, Kansas, aproximately 28 miles eut-southeast of Emporia, Kansas, and is descriaed in the licensees'
" Final Safety Analysis Report", as supplemented and amended, and in the licensees' Environmental Report, as supplemented and amended.
B. Subject to the conditions and requirements incorporated herein mission hereby licenses Kansas Gas and Electric Company (KG&E), the Com
, Kansas City Power & Light Company (KCPL) and Kansas Electric Power Cooperative, Inc.(KEPCO).
(1) Pursuant to Section 103 of the Act and 10 CFR Part 50 " Domestic Licensing of Production and Utilization Facilities," KG&E, to possess, use and operate the facility at the designated location in Coffey County, Kansas, in accordance with the procedures and limitations set forth in this license; (2) KCPL and KEPC0 to possess the facility at the designated location in Coffey County, Kansus, in accordance with the procedures and limitations set forth in this license; (
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-3 (3) KG&E, pursuant to the Act and 10 CFR Part 70, to receive, possess and use'at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended, 9
(4) KG&E, pursuant to the Act and 10 CTR Parts 30, 40 and 70, to re- ! ceive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radia-tion monitoring equipment calibration, and as fission detectors in amounts as required; (5) KG&E, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) KG8E, pursuant to the Act and 10 CTR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This license shall be deemed to contain and is subject to the conditions specified in the Comission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Comission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) MaximumPowerLevel _ KGAE is authorized to operate the facility at reactor core levels not in excess of 3411 megawatts thennal (100% power) power in accordance with the conditions specified herein and in Attach-ment I to this license. The activities identified in Attachment 1 to this license shall be completed as s hereby incorporated into this license. pecified. Attachment 1 is (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental r'rotection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. KG&E shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (
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4 . f . (3) Antitrust conditions - Kansas Gas & Electric Company and Kansas City Power & Light Company shall comply with the antitrust conditions delineated in Appendix C to this license. (4) Environmental _ Qualification (Section 3.11. SSER #4. Section 3.11. SSER f5)* . r All electrical equipment within the scope of 10 CFR 50.49 shall be qualified by November 30, 1985. , F_ ire protection (Section 9.5.1
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(5) SER. Section 9.5.1.8. SSER #5) - (a) KG8E shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPS Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, and as approved in the SER through Supplement 5, subject to provisions b & c below. (b) KG&E may make no change to the approved fire protection program which significantly would decrease the level of fire ( protection in the plant without prior approval of the Commis-sion. To make such a change the licensee must submit an application for license amendment pursuant to 10 CFR 50.90. (c) KG&E may make changes to features of the approved fire pro-tection program which do not significantly decrease the level of fire protection without prior Commission approval, provided: (1) such changes do not otherwise involve a change in a Itcense condition or technical specification or result in an unreviewed safety question (see10CFR50.5.9). (ii) such changes do not result in failure to complete the fire protection program approved b Comission prior to license issuance. y the
*The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements g wherein the license condition is discussed.
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( . 4 j KG&E shall maintain, in an auditable fonn', a current record. of all such changes including an analysis = of the effects of the change on the fire protection program and shall make such records available to NRC ins)ectors upon request. All i changes to the approved program saall be reported to the ' Director of the Office of Nuclear Reactor Re ulation, along with the FSAR revisions required by 10 CFR 5 .71(e). (6) Qualification of Personnel (Section 13.1.2, SSER #5, Section 18, 55ER #1) KGAE shall have on each shift operators who' meet the requirements
- I described-in Attachment 2.
(7) _NUREG-0737 Supplement ! Conditions (Section 12. SER) J. KG&E shall complete the requirements described in Attachment 3 to the satisfaction of the NRC. These conditi'ns o reference the .
.' appropriate items in Section 22,LaTMI Action Plan Requirements for App 1tcants for Operating Licenses," in the Safety Evaluation .
Report and Supplements- 1, 2, 3, 4, and 5 of NUREG-0881. I. (E) 151T75) Post fuel-Loading Initial Test Program (Section 14, SER Section !
, Any changes in the Initial Test Program described in Section 14 of-the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
(9) Inservice Inspection Program (Sections'5.2.4 and 6.6, SER) By December 11.-1985, KG&E shall submit for staff review and approval, the inservice inspection program which confonns te the ASME Code in effect on March 11, 1984 (10) Emergency Planning . , In the event'that the NRC finds that the lack'of progress t
-in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350,-is an l
_ indication that a major substantive problem exists in L. . achieving preparedness, or the maintaining an10adequate provisions of state CFR Section of emergency (2) 50.54(s) [ will apply. . ( . L
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I' ' (11) Steam Generator __'u g ,Ruiture (Section 15.4.4, SSER #5) Prior to restart following the first refueling outage, KG&E shall ! submit for NRC review and approval an analysis which demonstrates ' that the steam generator single-tube rupture (SGTR) analysis pre- > sented in the FSAR is the most severe case with respect to the release of fission products and calculated doses. Consistent ' with the, analytical assumptions, the licensee shall propose.all necessary changes to Appendix A to this' license. (12) LOCA Reanalys t (SectiM 15.3.7, SSER d5) :
. L '. i Prior to restart following the first refueling outage, KG&E shall :
submit for NRC review and approval a reanalysis for the worst large ' break LOCA using an approved ECCS evaluation model. (1?) Generic Letter 83-28 KG&E shall-submit responses to and implement the requirements of Generic Letter 83-28 on a schedule which is consistent with that given in their February 29, 1984 and February 6, 1985 letters. (- (14) Surveillance of Hafnium Control Rods (Section 4.2.3.1(10),SER ~ pnd_55ER #2) - KG&E shall perfom a visual inspection of a sample of hafnium - control rods during one of the first five refueling outages. A sumary of the results of these inspections shall be submitted ' to the NRC. D. Exemptions from certain requirements of Appendix J to 10 CFR Part 50, and from a portion of the requirements of General Design Criterion 4 of Appendix A to 10 CFR Part 50, are described in the Safety Evaluation Report. These exemptions-are authorized by law and will not endanger life or property or the comon defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions the facility will operate, to the extent authorized herein, in conformity with the application, as amended, tne provisions of the Act, and the rules and regulations of the Comission. E.- KG&E shall fully implement and maintain in effect all provisions of
.the Comission approved Physical Security, Guard Training and Quali-fication, and Safeguards Contingency plans, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR l
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( , 50.54(p), which are part of the license. These plans, which contain Safeguards Information protected under 10 CFR 73.21 are entitled 1
" Wolf Creek Generating Station, Physical Security Plan; Safeguards Contingency Plan and the Security Training and Qualification Plan".
F. Except'as otherwise provided in the Technical Specifications or En-vironmental Protection Plan, the licer.see shall report any violations of the requirements contained in fection 2.C of this license in the followlng manner: initial notification shall be made within 24 hours to the NRC Operations Center via the Emergency Notification System with written followup within thirty da descdbedin10CFR50.73(b),(c)ysinaccordancewiththeprocedures
.and(e).
G. The licensees shall have and maintain financial protection of such type and in such amounts as the Comission shall require in accordance with ' Section 170 of the Atomic Energy Act of.1954, as amended, to cover public liability claims. H. This license is effective as of the date of issuance and shall expire at Midnight on March 11, 2025. FOR THE NUCLEAR REGULATORY COMMISSION
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Harold R. Denton', Director Office of Nuclear Reactor Regulation Attachments / Appendices:
- 1. Attachment 1 - Tests and Other Items which must be Completed
- 2. Attachment 2 - Operating Staff Experience Requirements
- 3. Attachment 3 - NUREG-0737, Supplement 1, Require;nents
- 4. Appendix A - Technical Specifications (NUREG-1136)
- 5. Appendix 8 - Environmental Protection Plan
- 6. Appendix C - Antitrust Conditions Date of Issuance: JUN 4 IE
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.l ATTACHMENT 1 t
1 Prior to startup following the first refueling outage, the licensee shall complete, to the satisfaction of Region IV, the following activities' {
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- a. Replace the emergency diesel generator lobe oil keep warm pumps, with pumps satisfying ASME Section .Ill Class 3 requrements. -
(482/8455-01) i e h P G
(" ATTACHMENT 2 Opdrating $taff Experience Requirements _ KG&E shall have a licented senior operator on each shift who has had at least six months of hot operating experience on a same type plant, including at least six weeks at power levels greater than 20% of full power, and who has had start-up and shutdown ' experience, for those shifts where such an individual is not available on the plant staff, an advisor shall be provid?d who h'as had at least four years of power plant experience, including two years of nuclear plant expe-rience, and who has 1ad at least one year of experience on shift as a licensed
, senior operator at a similar type facility. Use of advisors who were licensed only at the R0 level will be evaluated on a case-by-case basis. Advisors shall be trained on plant procedures, technical specifications and plant systems, and shall be examined on these topics at a IcVel sufficient to assure far..f11arity with the plant. For each shift..the remainder of the shift crew shall be trained in the role of the advisors. These advisors, or fully trained and qualified replacements, shall be retained-until the experience levels identified in the first sentence above have been achieved. The names of any replacement advisors shall be certified.by XG&E prior to these individuals being placed on shift.
The NRC shall be notified at least 30 days prior to the date KG&E proposes to release the advisors from further service. ( . (
o j (' ATTACHMENT 3' l NUREG-0737, SUPPLEMENT 1. REQUIREMENTS (1) Tunctional and Task Analysis (I.C.1, SSER #5) Prior to startup following the first refueling outage, KG&E shall submit for f staff review and a? proval, a description of the process used to complete the functional and tas( analysis, including a description and justification for all information and control deviations from the Westinghouse 0<ners Group Emergency Response Guidelines. Revision 1. (2) Emergency __ Response Capabilities (Generic lette'r 82-33, Supplement 1 to NLREG 0737)
, i Prior to restart following the first refueling outage, KG&E shall have a fully functional Technical Support Center and Emergency Operations facility and a fully operable Emergency Response facilities Information System (ERFIS).
(3)RegulatoryGuide1.97(Section7.5.2.3,SSER#3) Prior to restart following the first-refueling outage, KG&E shall have installed and operable the following instrumentation. ( (a)-- Source range -instrumentation qualified to post-accident conditions (b) Reactor vessel water level instrumentation (c) Subcooling monitors ' (d) Radiation monitors for releases from steam generator safety / relief
' valves or atmospheric dump valves, and (e) Auxiliary feedwater~ pump turbine exhaust monitor i a
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(. ' APPENDIX B
'TO FACILITY OPERATING LICENSE NO NPF-42 KAN'SAS GAS AND ELECTRIC COMPANY KANSAS CITY POWER & LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE, INC.
WOLF CREEK GENERATING STATION UNIT 1 ( DOCKET NO. 50-482 ENVIRONMENTAL PROTECTION PLAN (NONRAD10 LOGICAL)
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( WOLF CREEK GENERATING-STATION l UNIT NO 1 i ENVIRONMENTAL PROTECTION PLAN l; (NON-RADIOLOGICAL) l July, 1984 . l t TABLE OF CONTENTS Section Paa ge, 1.0 Objectives of the Environmental Protection Plan . . . . . . 'l - 1 2.0 Environmental Protection Issues . . . . . . .-.-. . . . . . 2-1 2.1 Aquatic Issues -
. . 2-1 .
2.2 Terrestrial Issues . .. .. ..................
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3.0- Consistency Requirements ................. 3-1 - 3.1 : Plant Design and Operation ................ 3-2 3.2 Reporting Related to the NPDES Pennit and ! State Certification . . . . . . . . . . . . . . . . . . . . 3-3 (' 3.3 Changes' Required for Compliance with Other _-Environmental Regulations . . . . . . . . . . . . . . . . . 3-3 ; 4.0 Environmental Conditions ..-. . . . . . . . . . . . . . . . 4-1 4.1 Unusual'orl.Important Environmental Events . . . . . . 4-1 4.2 Environmental Monitoring and Management . . . . . . .. .. .. . 4-1 4.2.1 Fog Monitoring . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2.2 Waterfowl Impaction . . . . . . . . . . . . . . . . . . . . 4-1 4.2.3 Land Management . . . . . . . . . . . . . . . . . . . . . . 4-1 5.0 Administrative Procedures . . . . . . . . . . . . . . . . . 5-1 5.1 Review and Audit . . . . . . . . . . . .-. . . . . . . . . 5-1
- 5. 2 Retention of Program Documentation . . . . . . . . . . . . 5-1 5.3 Changes in Environmental Protection Plan . . . . . . . . . 5-1 5.4 Plan Reporting Requirements . . . , , . . . . . . . . . . . . 5-1 .
5.4.1 Routine-Reports .:. . ... , . . . . . . . . . . . . . . . . 5-1 5.4.2 Nonroutine Reports . . . . . . . . . . . . . . . . . . . . 5-2 t 9
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1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of nonradiological values during operation of Wolf Creek Generating Station. The principal objectives of the EPP are as follows: (a) Verify that the facility is operated in an environmentally acceptable the Final Environmental Statement Operating manner, License Stage asNUREG established 0878 by(FES-OLS), and other NRC environmental im assessments. (b) Coordinate NRC requirements, assure they are suitably "*>1 filled and maintain consistency with other Federal, State and local requirements for environmental protection. , (c) Keep NRC informed of the environmenal effects of facility operation and of actions taken to control those effects.
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' Environmental concerns identified in the FES-OLS which relate to water quality matters are regulated by the NPDES permit issued by the State of Kansas.
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f . 2.0 Environmental Protection Issues In the FES-OL dated January,1982, the staff considered the environmental im-j pacts associated with the operation of Wolf Creek Generating Station (WCGS). l Certain environmental issues were identified which required monitoring, study i or license conditions protection to resolve environmental concerns and to assure adequate of the environment. 2.1 Aquatic Issues * (a) The impacts of Wolf Creek Generating Station operation on the aquatic environment of the John Redmont Reservoir - Neosho River system will be negligible during periods of nomal and above-normal hydrologic conditions in the upstream watershed. However, should a severe and prolonged drought occur, the withdrawal of cooling-lake inateup water from the Redmont Dam tai.lwaters area would contribute to a marked draw-down of water in the reservoir and to reduced streamflow in the river, thus severly depleting available aquatic habitat and adversely affecting resident biota. (FES Section 5.5.2.1). (b) Some of the operat'ional effects on aquatic organisms in the cooling lake will be locally severe. For example, periodically high concentrations of total residual chlorine in the vicinity of the cooling water discharge ( outlet is expected to cause appreciable mortality among aquatic organisms, especially during periods when temperatures in the area are insufficient to cause fish and other motile species to avoid the area. (FES Section 5.5.2.2). . (c) Cold' shock effects on fish due to reactor shutdowns could cause significant mortality to aquatic species in the cooling lake. (FESSection5.5.2.2). (d) Impingement and/or entainment impacts on aquatic biota are expected to be significant since the appro'ich velocity of water flow to the facility are relatively high. (FES Section 5.5.2.2). (e) Discharge from the cooling lake to Wolf Creek is expected to influenq6 the composition of aquatic comur ities imediately downstream from toe discharge outlet, but aquatic biota of the Wolf Creek-Neosho River ur-fluence will not be adversely affected by the discharge. (FES Section 5.5.2.3). The NRC will rely on the State of Karsas for determination of the need for monitoring or pemit limitations related to these and othar aquatic issues. ( 2-1
I k 4 ( , 2.2 Terrestrial Issues (a) Thatthecompositionandstructureofvegetationinthe453ha(1120 l acre)exclusionzonewillbeselectivelycontrolledtobecompatible - l with the function and security of station facilities. (FES-OLS: Section5.5.1.1;StationSite) (b)" That' the vegetation within a buffer zone surrounding the cooling lake will be retained naturally occurring in or allowed biotic communities. to develop FES-OLS:(toward
.Section a 5.5.1.1; natural state, f.e.
StationSite) , i (c) That herbicides used for the maintenance of transmission line corridors will .be limittd to herbicides approved by the V. S. EPA and the State of Kansas at the times of such use. (FES-OLS: Section 5.5.1.2; Energy-TransmissionSystem) , (d) That in the event a serious disease problem involving waterfowl attribut-able to station operation occurs, the actions specified in the reference
' kill be initiated following technical evaluation if deemed necessary.
(FES-OLS: Section 5.5.1.1; Station Site) (e) The need for a wildlife monitoring program which includes a general survey I. program for waterfowl collision events be accomplished. (FES-OLS: Sec-tion 5.5.1.2; Energy-TransmissionSystem) (f) The need for a fog monitoring program to document any potential increase in fogging due to the operation of the cooling lake heat' dissipation system. (FES-OLS: Section 5.4.1; FogandIce) 2-2
3.0 Consistency Requirements 3.1 Plant ~ Design and Operation The licensee may make changes in station design or operatien or perfonn tests or experiments affecting the environment provided such activities do not in-volve an unreviewed environmental question and do not involve a change in the EPP*. Changes in station design, operatio'n, perfonnar.ce of tests or experi-ments which do not affect the environment are not subject to requirements of this EPP. Activities ments of this Section. governed by Section 3.3 are not subject ta the require-Before engaging'in additional construction or operational activities which may significantly affect the environment, the licensee shall prepare and record an environmental evaluation of such activity. Activities are excluded from this requirement if all measurable nonrediological environmental effects are con-fined to the on-site areas previously disturbed during site preparation and plant construction, khen the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activity and obtain prior NRC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be imple-mented only in accordt.nce with an appropriate license amendment as set forth in Section 5.3 of this EPP. A proposed change, test or experiment shall be deemed to irvolve an unreviewed environmental question if it concerns: (1) a matter which may result in a sig.nificant increase in any adverse environmental impact previously evaluated
'in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Boardt or (2) a significant change in effluents or power level (3) a matter not previously ' reviewed and evaluated in the doc-uments specified in (1) of this Subsection, which may have a significant adverse environmental impact.
- This provision does not relieve the licensee of the requirements of 10 CFR 50.59, 3-1
o . 9 ( The licensee shall maintain records of changes in facility design or o and of tests and experiments carried out pursuant to this Subsection. peration These records shall include written evaluations which provide bases for the deter-mination that the change, test, or experiment does not involve an unreviewed environmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1,0. The licensee shall in-clude as part of the Annual Environmental Operating Re 5.4.1)briefdescriptions, analyses, interpretations, port and(perSubsection evaluations of such changes, tests and experiments. 3.2 Reporting Related to the NPDES Permit and State Certification Changes to, or renewals of, the NPDES Pennit or the State certification shall be reported to the NRC within 30 days following the date the change or renewal is approved. If a pennit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted. The licensee shall notify the NRC of changes to the effective NPDES Pennit pro-posed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPDES Permit at the same time the application is submitted to the permitting agency. ( 3.3 Changes Required for Compliance with Other Environmental Regulations Changes in plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, and local environmental regulations are not subject to the requirements of Section 3.1. ( 3-2
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( I 4.0 Environmental Conditions t 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or coald result in significant environmental impact casually related to plant. operation shall
' be recorded and promptly reported to the NRC within 24 hours followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of
, 1973, fish kills, increase in nuisance organisms or conditions, and unantici-pated or emergency discharge of waste water or chemical substances.
No routine monitoring programs are required to implement this condition. 4.2 Ei.vironmental Monitoring and Management Environmental monitoring and management activities shall be undertaken as out-lined,in.Section 2 and as described in the following. 4.2.1 Fog Monitoring A fog monitoring program shall be accomplished to document the frequency (. of occurrence of natural fog and future cooling lake operation induced fog througn the first year of comercial operation of WCGS. A visiometer and continuous recorder shall be utilized in a conservative location throughout the program. ' 4 ?.2 Waterfowl Impaction A general survey program shall be accomplished to document significant water-fowl collision events and detennine if mitigation is warranted. 4.2.3 Land Management There shall be a land management program instituted at WCGS to provide for revegetation, maintenance, and restoration of the WCGS site. This program shall attempt to achieve a balance between production and conservation values on site property through the implementation of conservation and wildlife management techniques. There shall be no reporting requirements associated With this condition, 4-1 l l-l _ . ,~ _ _ _ _,
( - 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the EPP. The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. : A description of the organization structure utilized to achieve the independent review and audit function and available for results of the audit activ. ties shall Jae maintained and made inspection. 5.2 Retention of Program Documentation i Program documentation relative to the environmentai aspects of plant operation shall be made and retained in a manner convenient for review and inspection. Program documentation shall be made available to NRC on request. Documentation of modifications to plant strvetures, systems, and components detennined to potentia 11'y affect the continued protection of the environment shall be retained for the life of the plant. All other infonnation, data, and finalized reports relating to this EPP shall be retained for five years or, where appitcable, in accordance with the requirements of other agencies. 5.3 Changes in Environmentel Protection Plan (. Requests for changes in the EPP shall include an assessment of the environmental impact of the proposed change and a supporting justification. Implementation of such changes in' the EPP shall not comence prior to NRC approval of the pro-posed changes in the form of a license amendment incorporating the appropriate revision to the EPP. 5.4 Plan Reporting Requirements 5.4.1 Routine P.eports An Annual Environmental Operating Report describing implementation of this EPP for the previous calendar year shall be submitted to the NRC prior to May 1 of ' each year. The initial report shall be submitted prior to May 1 of the year following issuence of the operating license. The period of the first report shall begin with the date of issuance of the operating license. 5-1
i t . . The report shall include stunaries and analyses of the results of the environ- ! rnental protection activities required by Subsection 4.2 of this EPP for the t report period, including a compa ~rison with related preoperational studies, . operationalcontrols(asappropriate),andpreviousnon-radiologicalenviron- ! mental monitoring resorts, and an assessment of the observed impacts of the ' plant operation on tae environment. If hannful effects or evidence of trends ! toward irreversible damage to the environment are observed. the licensee shall provide a detailed anal alleviate the problem ysis of the data and a proposed cour se of action to i
-The Annual Environmental Operating Report shall also include:
(a) A list of EPP noncompliances and the corrective actions taken to . remedy them.-- - (b) A list of a11' changes in statien design or operation, tests, and experiments:made=in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental issue. (c) A list of no'nroutine reports submitted in accordance with Subsection 5.4.2. , in the event that some-results are not available by the report due date, the .(, report shall be submitted noting and explaining the missing results. The miss-ing results shall be submitted as soon as possible in a supplementary report.
. - 5.4.2 Nonroutir.e Reports
- A' written report shall be submitted to the NRC within 30 days of occurrence ofanunusualorimportantenvironmentalevent(seeSection4.1). The report shall-(a) describe, analyze,andevaluatetheevent,includingextentand magnitude-of the impact, and plant operating conditions. (b) describe the probable cause of tie event, (c) indicate the action taken to correct the '
reported event, (d) indicate the corrective action talen to preclude repett-tion of the event and to prevent similar occurrences involving similar components or systems preliminary responses., and (e) indicate the agencies notified and their Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this Subsection. The NRC shall be provided a copy of such report at the time it is submitted to
- the other agency..
A 5-2
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/JTDINX C WOLP CREEX, Ull1T 1 l
nit 1'EtVST CQiDITIQl.", FOR KAllSAS GAS AllD ELECTRIC COMPAll't
- 1. As used herein:
(a)
- Licensee" means Kansas Gas and Electric Core pany.
(b} " Licensee's Service Area" means those counties I located in whole or in part within the area certificated'to l Licensee by the appropriate state regulatory commission. (c) *Dulk Power" means the electric power, and any attendant energy, supplied or made available at transmis . sion or subtransmission voltage by one entity to another. (d)
- Emergency support" is capacity and energy .'
as available from one system, and as needed by another system to replace capacity and energy made unavailable-due to forced outages of generating equipment or trans-mission facilities. (e) 'Haintenance support" is capacity and energy planned by one system to be made available to another sys-tem to replace capacity and energy made unavailable due to maintenance of generating equipment or transmission facil-ities. . (f) " Entity" means a financially responsible pri-vate or public corporation, governmental agency or author-ity, municipality, cooperative, or lawful association of any of the foregoing, owning, contractually controlling, or operating, or in good faith proposing to own, contrac-tually control, or operate, facilities for the generation and trant, mission of electricity for bulk power supply which meets each of the following criteriat (1) its exist-ing or proposed f acilities are technically feasible of interconnection with those of Licensee; (2) with the ex-ception of municipalities, cooperatives, government agen-cies or authorities it is, or upon commencement of opera-( tions will be, a public utility subject to regulation with respect to rates and services under the laws of Kansas. (
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( (g) " Participating entity" is an " entity" partici-pating in the ownership of or power output from Wolf Creek Unit 1. . (h) *FEC" refers to Kansas Electric cooperatives, Inc. or Kanst8 Electric power Cooperative, Inc. insofar as
'i t shall becums. s successor in interest.
(i) "KEC members in Licensee's Service Area" refers to the following KIC member rural electric co-operatives: the Butler Rural Electric Cooperative As-sociation, Inc. ; the Caney Valley Electric Cooperative Association, Inc. ; Cof fey County Rural Electric Coopera-tive Association, Inc.; thefJnited Electric Cooperative, Inc.; the Radiant _ Electric Cooperative, Inc.; the Sedgwick County Electric Cooperative Association, Inc.; the Sekan . Elcetric Cooperative Association, Inc.; and Sumner-Cowley Electric Cooperative, Inc. (j) The " total demand requirements" or the " power requirements" of the KEC members in License s Service Area refers to the sum for all such KEC members of the KG&E delivery point contributions to the maximun, monthly sum of the intagrated 15 minute non-coincidental demands for each member. . 2, (a) Licensee shall offer an opportunity to parti-l cipate in Wolf Creek Nuclear Unit Number 1 and any other nuclear generating unit (s) which it may construct, own I and operate severally ot jointly, during the term of the instant license or an extension or renewal thereof, to any entity (ies) in Licensee's Service Area. Such par-l ticipation shall be in reasonable amounts, by an ownership l interest, or at the option of the entity, by a contractual l right to purchase a portion of the output of such units, or on any other mutually agreeable basis. The transmis-sion provisions herein relate to participation in Wolf l Creek Nuclear Unit No. 1 and not to any transmission which may be associated with participation in other nuclear gen-l erating unit (s) which Licensee may construct, own and l operate severally or jointly. 1 k l l
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-3 (b) Licensee shall make available _to 'XEC an u'n-divided-17 percent ownership participation in the Wolf Creek _Huclear Unit Number 1 upon mutually agreeable taras and conditions, which ownership participation KEC shall acquire as of the date of issesnee of the Nuclear Regula-tory _ Commission's construction permit for the Wolf-Creek
-Nuclear Unit Number-1 or as soon thereafter as KEC can recure_the necessary regulatory and financing _appro'vals pursuant to' the terms of the May 20, 1976 settlement agreement between Licensee and KEC.. The power which KEC obtains from the Wolf Creek Nuclear Unit-Number 1 shall be utilized first to satisfy the power requirements of the XEC members in Licensee's Service Area to the maximum extent' reasonable and efficient to do so and to the extent consistent with KEC's other power supply ob-ligations to XIC's members in Kansas. During the'calen-dar_ year in which the Wolf Creek Nuclear Unit Number 1 'com- .
mences commercial _ operation and,,in each succeeding year of
, operation of- the Wolf Creek Nuclear ' Unit Number 1 no. less than 42 percent of the_ total demand requirements of the KIC members in Licensee's Service Area sh4 11 be satisfied by KIC by use of its available power from Wolf Creek Nu-clear Unit Number 1. XIC's po'wer from Wolf Creek Unit-
' Number 1 shall be transmittedi by Licensee for KEC to such delivery and interconnection points on Licensee's system and in'such-amounts as specified by;KEC, pursuant to para-
- graph 5 (a) below.
3._ Licensee shall interconnect with any participating entity in Licensee's Service Area which requests such in-I terconnection-and provide for the followings L (a) maintaining and coordinating of reserves, in-l cluding, where applicable, the purchase and; sale'of re-serve capacity and energy, l. (b) emergency support, (c) maintenance-support, _g (d) delivery of " unit power" or other participa-L tion power out of Wolf Creek Unit 1 from the Licensee, L and
_ _ , _ . . _ . m -. , _ , .- .m ,. . v m ...; n.. ,, p . ( . . (c) transmission services for the above and as ' d; scribed subsequently.
- 4. (a) Licensee _and the participating entities in Licensee's Service Area having a rese'rve coordination crrangement prot ided for in Paragraph 3, above, shall from time to time jointly establish the minimum reserve roquirements to be installed and/or provided under con-tractual arrangements as necessary to maintain in total o reserve margin sufficient to provide adequate reliabil-ity of power supply to the interconnected systems of the parties. To have reserve coordination rights, other than roserves for Wolf Creek, with the Licensee, a participat-ing entity must own or have contractual rights to generat-ing capacity other than of Wolf Creek Nuclear Unit Number
- 1. Unless otherwise agreed upon, the minimum reserve re- '
quirement shall be calculated as a percentage of the projected annual peak load, adjusted for ' purchases and coles of firm power, including partial requiremen'ts firm
,pswer. The parties to such a reserve coordinating arrange-I ment shall provide such amounts of opernting (ready and spinning) reserve capacity as may be adequate to avoid the irposition of unreasonable demands on the others in meet-sing the normal contingencies of operating their systems.
H3 wever, in no circumstances shall any party's spinning er operating reserve requirement exceed the minimum re-carve requirement as provided above. (Horeover, if the p2rties to a reserve coordination agreement cannot agree upon a minimum reserve requirement, the participating antities' minimum reserve requirement shall be neither loss than nor greater than Licensee's minimum system re-carve requirement.) (b) Emergency and/or scheduled maintenance bulk power service shall be provided by each party to the ex-tent required by the system in need, and be furnished to the fullest extent available from the supplying system. Licensee and each party (ies) within Licensee's Service Area shall provide to the other emergency and/or sche-duled maintenance bulk power service if and when avail-cble from its own generation bnd from generation of k t l l l l
-5 . -
( others with whom Licensee is interconnected to the extent it can do so without impairing service to its customers. including other electric systems to whom it has firm com-mitme n t s .) .
- 5. (a) Licensee shall transmit the power from a parti-cipating entitys share of Wolf Creek ' Nuclear Unit Nunber 1 to the participating entity, or for the ace %unt of such participating entity, to delivery or interconnection points on Licensee's system and in amounts as specified by the participating entity. Such deliveries shall be reasonable as to the number.of points, system adequacy and frequency of schedule changes.
(b) Licensee shall transmit power from an entity (ies) outsi'de Licensee's Service Area to a participating entity-within Licensee's Service Area in an amount at least ecual
- to the share of Wolf Creek Nuclear Unit Number 1 of a par-ticipating entity within Licensee's Service Area when the output of this unit 'is reduced or unavailable because of maintenance or for other reasons. .
( (c) If capacity and energy from a participating ontity's (ies ' ) portion of Wolf Creek Nuclear Unit Number 1 are delivered to other entities, Licensee shall also provide transmission for a later scheduled return of such
'onergy within the same calendar year, in an equal amount of mwhrs, from these other entities to the delivery point of the participating entity within Licensee's Service Area, provided that such transmission arrangements can be reason-ably accommodated from a functional and technical stand-point. (For example, any Wolf Creek power transmitted (per-udssible within the terms of these conditions) out of Licensee's system shall create in participating entitylles) the right to call upon Licensee, and the corresponding obligation of Licensee, to transmit equal pcwer back into Licensee's system for account of said participating cntity(ies), all_within the same calendar year. At any point in time the transfer of power back in (for account of participating entity (ies)) could occur simultaneously with full delivery of that participating entity's(ies')
l power from Wolf Creek Nuclear Unit Number 1. l k
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- 6. -(a) -Licensee shall' sell _ power at its filed and ef fective rates (for total or partial requirements) to
, any entity in. Licensee's-Service l Area now engaging or pro ,
posing to engage in the wholesale or retail sale _ of elec - .
, tric power, o (b) _ Insof ar as the power requiremento of the XEC members in Licensee's . Service Area are satisfied by power' which is not Licensee's power and which has been transmit-
'ted- by- Licensee - for KEC pursuant to Paragraphs 2 (b). and 7 Eof these--license conditions, Licensee's sale of fuil or_ partial 1 requirements to KEC or to its members in
- Licensee's Service - Area pursuant to Paragraph 6 (a) above '
shallL be correspondingly - reduced. 7, . ' In addition to the transmission of fered by Licensee .
-in Paragraphs 2 (b) , 3, 5,. and 6 above, Licensee shall, consistent with Paragraph 8 below and with the terms of I< the May 20, 1976 settlement agreement between Licensee and KEC, transmit, for KEC the following power:
, (a) Commencing July 1,1980, and _ until the _ Wolf
, Creek Nuclear Unit Number 1 commences commercial operation or?is. finally abandoned, (i) 30 megawatts of preference
. customer power, to the extent available, which KEC obtains from Southwestern Power Administration, provided that such' preference power is delivered to Licensee's Neosho.substa-
~
tion near Parsons,? Kansas; (ii)1said- 30 megawatts may be increasti to 90 megawatts of _ Southwestern Power Adminis-tratiori p :eference customar power once the Harry S. L - Truman :- .un commences commercial. operation, but not before; (b) When Wolf Creek Nuclear Unit Number 1- commences commercial operation, and thereafter in each succeeding calendar year until the Project ceases operation or-until
- calendar year 2021, whichever is later, a total of 90 -mega-
. watts of preference customer power (inclusive of the power described in subparagraph 7(a) above) -which KEC obtains from the. Southwestern Power Administration. or from a source _
or- sources which as a matter of law are administrative 1y 4;
- foreclosed to Licensee by virtueo ' f a statutory or regula-
. tory preference. The power described in subparagraphs 7(a)
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c . ... , t _ , l . 4 and . (b) shall be transmitted by Licensee 'for KEC on a con-Ltractual buy-sell arrangement unless by entering into such
;crrangement Krc would lose its entitlement to such prefer-once power ,
'(c)_'.When Wolf Creek Nuclear Unit. Number 1 commences
-commercial operation and'thereafter so'long as this Unit continues operation, or until calendar year 2021, which-ovor is'later, (i) any additional quantities of; power which KEC generates from a source.other.than Wolf Creek: Nuclear Unit Number.1, or which KEC obtains -from any power source
- cr --sources which as- a matte.r _ of law are not administratively '
< foreclosed to' Licensee by virtue.of a statutory or regula-
; tory preference,-provided that such power is transmitted-by Licensee to KEC members'in' Licensee's Service Area for ths use of _ such membersit and . (ii) any _9ther quantities- of
- t power which KEC generates from a source other than Wolf Croek Unit Number 1, or which.KEC obtains from any power escurce or sources-which.ae a matter of law are not admin-istratively foreclosed"t'o' Licensee by virtue of a statu-tory or regulatory preference, to the.same extent that Licensee would' reasonably agree to transmit such power.for cny other electric utility. All' the power described in this_ subparagraph shall be transmitted by-Licensee for KEC upon reasonable and timelyLrequest for such transmission only on the basis of a contractual _ buy-sell arrangement similar.in duration.to Licensee'sLthen existing comparable i buy-sell l contractual: arrangements with. other electric util-
, ities.; Pursuant to'any such buy-sell contract, Licensee challLpurchase the designated power'as delivered by the es11er- or any other entity at delivery or interconnection points _'on Licensee's' system-and shall resell the same to KEC . at _ the . Licensee's purchase price plus an amount which constitutes Licensee's transmission cost including a rea-cenable return on=the investment allocable solely to the transmission of such power;
'(d) All of the power transmitted by Licensee for KEC pursuant _ to the provisions of subparagraph _7 (a) (i) above, shall be used to satisfy the power requirements of the KEC members in Licensee's Service Area. When the
{ Harry S. Truman ' Dam commences- commercial operation and in
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cach succeeding year 1of commercial operation of the Project, no>1ess than 40 megawatts of the power transmitted by L1/ cansee for' KEC pursuant to the provisions of subparagraph 7 (b): above,- if available, shall be used to satisfy the - power rgquirements of the KEC members in Licensee's Ser- ' vice Areas and (ef Insof ar as the power of KEC from Wolf Creek
. Nuclear Unit Number 1land the power transmitted by Licen- ,
coe _ for: KEC in the manner provided in subparagraphs 2 (b) and 7 (a) through (d).is not utilized in Licensee's Sarvice Area, as reasonably and f airly determined by KEC
-_in accordance with the foregoing provisions, Lic-nsee shall,_ upon reasonable and timely request for such service,
- transmit such excess power for KEC from and to such inter-connection points on Licensee's system and in such amounts na specified.by KEC on terms and conditions as provided .
Ein subparagraphs 2 (b) and 7 (a) through (c) above. . 8.- The transmission described in these license condi-
-tions shall be made available only upon terms-which
-. fully, compensate Licensee for its. costs, including _any
(!- trcnsmission power losses and a reasonable return on in-vostment allocable solely to such transmission and re-flocted.in Licensee's-schedules or tariffs filed with the Xcnsas Corporation Commission or the Federal Power Comnis- ,
- clon.- The transmission described in Paragraphs 2 (b) and 7 above_shall be available to KEC for the transmission of r: quested amounts of ~ power in the manner specified in Paragraphs 2 (b) and 7(a) through (e) above, provided ,
that KEC gives Licensee reasonable advance notice-of-tho transmission required and, provided further, that cuch transmission arrangements _ can be reasonably accorno-dated from a functional and technical standpoint and to tha~ extent that Licensee can do so without impairing ser-vice to its customers including other electric systems to which it has firm commitments. Nothing herein imposes a , rcquirement on Licensee to become a common carrier.
- 9. Licensee shall include in its planning and construc-tion of' additional transmission facilities sufficient trans-
- ; mission capacity'to accommodate the transmission described ,
i in Paragraphs 2 (b), 7 (b) , and 7 (c) (1) above, provided that ll - L( L i
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4 - 9, KE'C gives Licensee sufficient advance notice as may be nec'- ossary to acconnodate such requirements from a functional
- and technical standpoint. Licensee and'KEC shall consult with respect to the planning and construction of additional tr'ansmis sion : f acilitie s .
- 10. The foregoing conditions shall bel implemented in a conner not inconsistent with the provisions of, and as pro-vided under, the Federal Power Act and all other applicable raderal and State laws and all rates, charges and practice's in connection therewith are to be subje'et to the approval of regulatory agencies having jurisdiction over them.
9 C. . 9 4 6 8 4
,- ,- -m -z-e p, APPDiDIX C g
WOLP CREEX,-UNIT 1 ANTIWCT CONDITIONS FOR
~
. KANSAS CITY POWER & LIGHT COMPANY ,'
- 1. As used hereini (a) " Licensee" means Kansas City Power ti. Light
' Company. .
(b) " Licensee's Service Area" means those portions of the States of Missouri and Xansas which are certificated to Licensee by the respective state regulatory. commissions. An entity shall be deemed to be in the " Licensee's Service Arsa" if it has electric power generation, transmission or distribution facilitics located in whole or in part in or cdjacent to the above-described area or in counties served
.-in part at retail by Licensee.
(c). " Bulk Power" means the electric power, and any , ottendant energy, supplied or made available at transmission cr subtransmission voltage by one entity to another. (d) " Entity" means person, a private or public cor-F paration, a municipality, a cooperative, a joint stock assoc-
.iction, b~usiness trust or a lawful associatien of any of the foregoing constituting, a separate legal entity owning, oper-oting or proposing to own or operate equipment or facilities for .the generation, transmission, or distribution of elec- '
l_ tricity, provided that, except for municipalities and coop-l_ oratives, an " entity" is restricted to those-which are or L will .be' a public utility. under the laws. of the state in L which the entity transacts business or under the Federal Powar Act and are or will be providing electric service un- ' dor a contract or. rate schedule on file with and subject to the regulation of a state regulatory commission or the red-crol Power Commission. L (e) " Cost" means any and all operating, maintenance, l ganoral and administrative expenses, together with a.ny and L 011 ownership costs, which are reasonably allocable to the
- transaction consistent with industry practices, cost shall include.a reasonable return on Licensee's investment. The
,( . l
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.o9 , -. . , :~ . ; ..~ ,~ c.n.e . ; c .: m.t. . : .; , . ., p 1 ., 11 neln of a peartion of the capacity of a generating unit ,
chall be upon the basis of a rato that will recover to -, tha. seller the pro rata part of the fixed costs and oper-Oting, maintenance, general and administrative expenses of tho unit, provided that, in circumstances in which Licen-sco and one or more entities in Licensee's Service Area cach takesHan undivided interest in a unit in fee, con-ctruction costs and operation, maintenance, general and cdministrative expinses shall be paid pro rata. 2 (a) Licensee shall intercennect with and coordinate cparations (by means of reserve sharing and the sale and purchase of emergency ana/or scheduled maintenance and/o'r other classes of bulk power) with any entity (ies) in Licen-aco's Service Area engaging in or proposing to engage in electric bulk power supply on terms that will fully compen-cote Licensee for its. costs in connection therewith. Such coordination arrangements will allow the other party (ies) full access to the benefits of coordination. (b) Emezgency and/or scheduled maintenance bulk (1powerserviceshallbeprovidedbyeachpartytotheex-tcnt required by the system in need, and be furnished to tha fullest extent available from the supplying system. Licensee and each party (les) shall provide to the other emargency and/or' scheduled maintenance bulk power service if and when available from its own generation and from ganeration of others to the extent it can do so without it.< pairing service to its customers including other elec-tric systems to whom it has firm commitments and the re-coiving party shall fully compensate the other party for
'its costs in connection therewith.
(c) Licensee and the other party (ies) to an inter-connection and reserve sharing arrangement shall from time to time jointly establish the minimum reserves. to be in-ctclied and/or provided under contractual arrangements as nocessery to maintain in total a reserve margin sufficient to provide adequate reliability of power supply of the inter-connected systems of the parties. Unless otherwise agreed upen, minimum reserves shall be calculated as a percentage of estimated peak load responsibility. No party to the ar-rangement shall be required to maintain greater reserves (
. - . .s -
,, 64
.u- .
L +-than such minimum, provided that, i'f the reserve requirements -_cf c- party are increased over the amount _ such party would - be rsquired to maintain or have available without such inter . cannoction, then the other party (les) to such interconnection -chc11 be required to carry-or provide for, in addition to such ndnimum reserves, - die full amount in kilowatts of such "incroase. If in addition to sharing reserves, one party sells' ccpacity to another in order for that other to meet its re-sorvo responsibility, the seller shall'be appropriately com-panscted for such sale in accordance with applicable filed rotos. (d) -The parties to such a reserve sharing arrange-mont each shall provide such. amounts of operating (ready , cnd spinning) reserve capacity as may be adequate to avoid
~
the imposition of unreasonable demands'en the other in meet- .ing the normal contingencies of operating its system, How-ovor, _ in-no circumstances shall a party's operating reserve ! requirement exceed its minimum installed reserve require-ment' as determined in 2 (c) . I' (e) Interconnections will not be limited to low voltages when higher voltages are available from Licensee's installe?. facilities in the ' area where interconnection is cpprop-Jate, if and when the proposed: a'rrangement is found to b7 technically and economically feasible. Control and
'tcir.ntering facilities shall be provided as required for
. cafe, and. prudent operation of the interconnected systems. (f) -Interconnection and coordination agreements (c.1011 not embody any_ unreasonably restrictive provisions
- parteining to intersystem coordination. Good industry practice as developed in the area from time to_ time (if not unreasonably restrictive) will satisfy this provision.
- 3. Licensee will sell bulk power from new generating ccpocity planned or under construction at its. cost or pur-chtso bulk power- from any other entity (ies) in Licensee's 50rvice Area engaging in or proposing to engage in genera-tion of electric power when such transaction would serve to reduce the overall costs of new bulk power supply for itself or for the (other) party (les) to the transaction. This refers m
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N- . , T f . (spacificallyyto the opportunity.to coordina'.e-in the plan-- - ning of new. generation,' transmission and related1tacili--
- tios . : This provision shall not'be construed _to-require -
-_ Licensee to purchase or sell bulk power if itLfinds_'such fpurchase or sale -infeasible or its costs in connection with - ,
such purchase or sale would exceed its benefits therefrom.-
-de Licensee;and any successor in. title shall offer an opportunity:to participate in Wolf creek Nuclear-Unit 1 to ,
- cny entity (ies) in Licensee's Service Area which shall
-ladicate; its interest therein in writing delivered- to- Li-
.consee prior to' October, 31, 1974, and ? in any other nuclear -
ganorati'ng L unit (s)..which 'they or either of them, may. con-
. atruct, cnns and op.erate severally _ or^ jointly, during the ltarm of the instant license,;or any extension or renewal
- thsreof, by: either;aircasonable ownership interest in such Lunit(s),:or by aL. contractual right'to purchase a' reasonable portion of _ the output of such unit (s) - at the cost thereof
'if theD entity (ies): so elects. -Upon timely _ offer by Licen-ceo,-- notice of intention _ to participate Lin future . nuelear
_ , units must;be given to Licensee in writing ~priorito the
- (y lacement of. orders = for- major equipment. In connection--with i
- cuch access, -Licensee will also offer transmission service-ca.may:be required.for delivery of such power-to such entity
' (ios)1 on- a -basis that-. willi fully, compensate Licensee for ito: cost.
- L -
il. - (a) Licen'see sh'all' facilitate-the' exchange of bulk h .:pcwariby7t ransmission overtits transmission: facilities to,
~
k frem,-between or;amongcanyLentities in Licensee's Service - L ' Arco with which itEis at-any time interconnected, and be-itwoon any such interconnected: entity (ies) and. any other L . cntity (ies)' engaging in; bulk _ power supply- out' side Licen-coo's- Service Area between whose- f acilities Licensee's transmission ~11hes and the transmission lines of others twould form 'a continuous electrical path, ' provided that' (1)
- p. 4thsinaces'sary rights to utilize such(other)--' transmission
.linsa have been obtained, (2) the_ reliability of Licensee's bulk ' power < system is not- thereby: impaired, and (3) the ar-rangement,s reasonably 1can be accommodated from a. functional L .and : technical standpoint.- ;such transmissionishall be on
'torms that1fu11y compensate Licensee for its cost, including transmission losses associated therewith. Any entity (ies)
/
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requesting : such- transmission - arrangements L shall give rea- ' : conable Ladvance notice to Licensee of its (their) schedulo hnd requirements for bulkLpower to be scheduled by Lican- -
- soofoyer Licensee's transmission f acilities..
~
-(b) . Licensee shall include in its- planning and '
construction Lof ?.f acilitiesito be owned:by Li'censee suffi-s cient transmission capacity 'asimay be contractually re-lcorvad: for the-type of transactions referred to in subpara-grcph (a)' off this paragraph, provided that ;the entity (ies) Lin' Lice'nsee's Service Area give Licensee sufficient' advance
, notice as .mayl be necessary to accommodate its (their) re- .
Lguirsments from a ' functional 1 and1 technical ' standpoint and-Lprovided that such entity (ies) fully compensates Licensee ' cforf the contractual reservation by Licensee.of_ capacity in
/its transmission facilities.
6.- Licensee will sell power- for resale to any entity (les) , 'in Licensee's ServicefArea-now-engaging in or proposing to (fljage .ini retail _ distribution of electric power under con-
~
Ltracts? for ;its 1(their); full.or- partial requirements;at Li- + csnase?s applicable filed rates to the extent' Licensee can do: co without impairing service ,to its retail customers.- .
- 7. The-foregoing conditions shall be-implemented:in a-mannor .not-inconsistent with the provisions of,' and as _ pro- -
ildad under, . the Federal: Power.. Act and all other - applicable FQd3ral andLState laws and all rates, charges and practices 'iin connection therewith are to be subject to the approval ef E rogulatory_ agencies. having jurisdiction over them.- ~ O 1 o L .
1 c, : -
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7590-01 j(; ',
. I KANSAS GAS & ELECTRIC COMPANY VANSAS CITY POWER & LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE,-INC.
, DOCKET NO. STN 50-482 '
WOLF' CREEK GENERATING STATION, UNIT NO.1-NOTICE'0F ISSUANCE OF FACILITY OPERATING LICENSE Notice-is hereby given that the U. S. Nuclear Regulatory Comission (the Comission or NRC), has issued Facility Operating License No. NPF-42 to Kansas
; Gas & Electric Company,' Kansas City Power & Light Company and Kansas, Electric Power Cooperative, Inc. (the licensees) which authorizes operation of the Wolf Creek Generating Station, Unit No. I at reactor core power levels not in exces; of 3411 megcwatts thermal in accordance with the provisions'of the License, the Technical Specifications and .the Environmental Protection Plan.
l. The issuance of this license was approved by the Nuclear' Regulatory Comis- . sfon at a meeting on June 3,1985, and it superseded the License for Fuel j Loading and Low Power Testing, License No. NPF-32, issued on March 11, 1985. License No. NPF-42 incorporates changes to the technical specifications - that were made subsequent to the issuance of NPF-32 and supersedes NPF-32. Wolf Creek Generating Station, Unit No.1 is a pressurized water reactor located approximately 28 miles east-southeast of Emporta, in Coffey County,
.Kanst1. The application was submitted and accepted for review under the Com-mission's standardization policy statement of March 5,1973. Kansas Gas & 'k I'
1 9
^
9 7590-01 3 , f. 9 Electric Company was-one of five utilities who joined together under'the acro-nym SNUPPS_ (Standardized Nuclear Unit Power Plant System) to submit applications
~
(for. Construct 1on Pennits for a standard plant design for review under the Com-niission's standardization policy, using the duplicate plant option described in' Appendix N -to the Ccmission's regulations in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50), " Licensing of Production and r-- Utilization facilities." This option allows for a simultaneous review of the safety-related parameters of a libited number of duplicate plants which are to be constructed within a limited time span at a multiplicity of sites. The license is effective as of the date of issuance. The application for the license complies with the standards and requirements
.t
\ oftheAtomicEnergyActof1954,asamended(theAct),andtheConrission's regulations. The Comission has made appropriate findings as requ! red by the
- Act and the' Comission's regulations in 10 CFR Chapter I which are set forth in the License. Prior public notice of the overall action involv'ing the pro-posed issuance ~of an operating license was published in the Federal Register on December 8,1980(45FR83360).
- The Comission has detennined that the issuance of this license will not
- result in any environmental impacts other than those evaluated in the Final Environmental Statement since the activity authorized by the license is en-
. compassed by the overall action evaluated in the Final Environmental Statement.
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For further details:with respect to -this action, see (1) Facility =0perating License No. NPF-42, with Technical Specifications (NUREG-1136) and the' Environ- '
-_ mental Protection Plan; (2) the report of the Advisory Cocinittee on Reactor .
Safeguards, dated May 11,1982;(3) the Comisiton's Safety Evaluation Report, dated April 1982 (NUREG-0881), and Supplements 1- through 6; (4) the Final Safety Analysis Report-and Amendments thereto; (5) the Environmental Report and supplements-thereto;_and (6) the Final Environmental Statement, dated-June
-1982 -
These items' are available for inspection at the Commission's Public Docu-ment Room located at 1717 H Street, N. W.', Washington, D. C. 20555, and at the i Emporia State University, William Allen White Library,1200 Commercial Street. ( Emporia, Kansas 66801; and at the Washburn University School-of Law Library, Topeka, Kansas'.- A copy of. Facility Operating License NPF-42 may be obtained . upon request addressed to the U. S. Nuclear Regulatory Consnission,- Washington, D. C.~20555, At'tention: Director, Division of Licensing. Copies of the Safety Evaluation Report and Supplements 1 through 6-(NUREG-0881) and the Final Environ-mentalStatement(NUREG-0878)maybeorderedbycalling(202)275-2060or(202) - 275-2171 or by writing to the Superintendent of-Documents, U. S. Government Printing Office, Post Office Box 37082. Washington, D. C. 20013-7082. All orders should clearly-identify. the NRC publication number and the requester's GP0 deposit account, VISA or Mastercard number and expiration date. Anyone' {
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'7590-01 wishing to inquire about a subscription account or subscribe to a periodic NRC publication may do so by calling GPO at (202) 783-3238.
The NRC will continue to participate in. the National Technical Information Service Program and individuals or organiza'tions may continue to purchase NRC documents at current rates from the National Technical Information Service, Department of Comerce, 5285 Port Royal Road, Springfield, Virg .a 2216
~
Dated at Bethesda, Maryland this Y'Cday of
%, /9 f$
s j FOR THE NUCLEAR REGULATORY COMMISSION. 1 ff , 'bi -( 3. J Y ngblood Chief Lis .ai g Branc No. 1 Division of Licensing
herap , >
-8 y#j9,i UNIIED STATES
. . NUCLEAR REGULATORY COMMISSION j-
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WASHINGTON, D. C. 205$5. f:*3,3 *. Docket No. 50-482 AMENDMENT TO INDEMNITY AGREEMENT NO. B-99 _ AMENDMENT NO. 3 Effective June 4 1985
, Indemnity' Agreement No.-B-99 between Kansas. Gas
&. Electric.Compan,y, Kansas City Power & Light Company, and Kansas Electric Power Cooperative, Inc. and the Nuclear Regulatory Commission dated May.9 1984, as amended,. is hereby further amended as follows:
Item 3 of the Attachment to the indemnity. agreement is deleted in
, its entirety and the following substituted therefor:
'. Item 3 - License number or numbers-
, SNM-1929 (From 12:01 a.m. , May 9,1984 to 12 midnig t, March 10, 1985 inclusive .
.( NPF (from 12:01 a.m. , March 11, 1985 to Y
12 midnight, June 3, 1985 inclusive) NPF-42 (From 12:01 a.m. June 4, 1985
)
FOR-THE UNITED STATES NUCLEAR REGULATORY COMMISSTION b$$ $A JeromY Saltzman,' Assistant Diregt,or State and Licensee Relations- V Office of State Programs Accepted ,1985 Accepted , 1985 By: By KANSAS GAS & ELECTRIC COMPANY KANSAS CITY POWER & LIGHT COMPANY
.{ Accepted , 1985.
By KANSAS ELECTRIC POWER COOPERATIVE, INC.
, p.y KANSAS CAS AND ELECTMIC COMPANY u esenu: ces,a the O t r a n t es t es t September 3, 1985 e
Judith A. McConnell Executive Secretary State Corporation Commission 4th Floor, State Office Bldg. Topeka, Kansas 66612-1571 RE: Docket Nos. 120.924-U, 142,098-U, 142,099-U, 142,100-U Dear Ms. McConnell Enclosed please find the original and five copies of Affidavit of Glenn L. Koester declaring Wolf Creek Generating Station to be in commercial service as of 1:16 A.M. September 3,1985. I have enclosed an additional copy which I ask that you stamp " filed" and return to me.
'Ihank you for your assistance.
Sincerely, W b onathan L. Heller JLH/mbs Enclosures 9/6/85 xc: JMEvans, w/a BGoshorn, w/a Records Mgm, w/a - MS2-03 EDProthro/IDFile , w/a Exec. Af fidavit - 202 GO 201 N. Market - Wehrts Kansas - Mast Addrest Po. Dos 200 l W<hita. Kansas 67201 - Te4 phone Aree Code (316) 2614297 .
AFFIDAVIT OF GLENN L. K0 ESTER STATE OF KANSAS )
) SS:
SEDGWICK COUNTY- ) GLENN L. K0 ESTER, Vice Fresident - Nuclear for Kansas Gas and Electric Company, of lawful age, being first duly sworn, deposes and says:
- 1. All necessary preoperational and startup component testing at Wolf Creek Generacing Station has been demonstrably completed.
- 2. The 250-hour full warranted output performance test of the nuclear steam supply system has been satisfactorily completed by actual test demonstration as required by contract.
- 3. There exists demonstrably sufficient transmission capacity in place (either owned or otherwise obtained) to carry the total design electrical capacit.y of the respective Owners from the generating station to the distribution systems of the Owner utilities.
- 4. The unit is fueled by its predominant fuel source, slightly enriched uranium.
- 5. All information necessary to confirm the above four statements has been provided Staf f of the Kansas Corporation Commission so that it may independently verify the above.
- 6. Because Wolf Creek Generating Station has satisfied all four comme. -lal service criteria established by the Kansas Corporation Commission in its Order of October 9, 1984, in Docket Nos. 120,924-U, 142,098-U, 142,099-U and 142,100 3 and at 1:16 A.M. September 3, 1985 was generating electrical power that served all Owners' customers af ter satisfaction of all four criteria, the Owners hereby declare Wolf Creek Generating Station to be in commercial service as of 1 :16 A.M. September 3, 1985.
FURTHER, AFFIANT SAITH NOT. O&_.- TJ1enn $d Y L. Koester Subscpt##4,,,and sworn to before me this 3rd day of September, 1985. ' j/p'giDO'tB
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j i ~ '@ .D I j ml irks: lA 15; @ '/ MyC@%,h/)gg..y f... %
o 4 THE STATE CORP 0RAll0N COMMISS10!i 0F THE STATE OF KANSAS 2 - . C r ,,-,, mk i;, L.. g , ') C MICHAEL LENNEN, CHAIRMAN
.- .., .)
-7.:
,'/
BEFORE COMMISSIONER $l RICH ARD C. (PE T E) LOUX . . , KEITH R. Hr.NLEY ' N ',M, . 4 /
)
{N THE MATTER OF A GENERAL lNVESTIGATION BY)THE DOCKET NO. LOMMISSION OF THs PR9JECTED COSTS AND NELATED ) 120,924*U M AT T ERS OF THE WOLF CREEK NUCLE AR GENER Af l0N ) FACILITY AT BURLINGTON, KANSAS.
) DOCKET NO.
IN THE MATTER OF THE APPLICAfl0N OF KANSAS 6AS) 142,098-U AND ELECTRIC COMP ANY REQUESTING PROPOSED CHANGES )84-KG&E-197R IN ITS CHARGES FOR ELECTRIC SERVICE.
) DOCKET NO.
IN THE MATTER OF THE APPLICATION OF KANSAS CITY ) 142 099-U POWER AND llGHT COMPANY MEQUESTING PROPOSED CHANGES )3 IN ITS CHARGES FOR ELECTRIC SERVICE. DOCKET NO. IN THE MATTER OF THE APPLICATION OF KANSAS ELECTRIC 142 }00-0 ) POWER COOPERATIVES, lNC. REQUESTING ELECTRIC SERVICE. P90eosED CHANGES))84-KEh IN ITS CHARGES FOR PREHEAR110J1RDER NOW, UPON MOTION OF THE STAFF OF THE STATE CORPORATION (COMMIS$10N), THE ISSUE OF THE APPROPRIATE COMMISSION { "IN' SERVICE" OR COMMERCIAL OPERATION CRITERIA TO BE UTILIZED IN DETERMINING, FOR RATEMAKING PURPOSES, WHETHER THE WOLF CREEK 6ENERATING STATION IS IN COMMERCIAL OPERATION, COMES ON FOR CONSIDERATION AND DETERMINATION BY THE COMMIS$10N.
-APPEARANCES IHE MATTER WAS HEARD BY THE COMMIS$10N ON JULY 9, 1984, IN HE ARING ROOM B, ST ATE CORPOR AT ION COMMIS$10N, FOURTH FLOOR, STA1E OFFICE BUILDING, IOPEKA, KANSAS. IHE PARTIES WERE REPRESENTED BY COUNSEL AS FOLLOWSI WILLIAM D. WEBB, DVERLAND PARK, KANSASJ WARREN B. WOOD AND MARK ENGLISH, KANSAS CITY f MISSOURI, APPEARING ON BEHALF OF KANSAS Cl1Y POWER & L10HT COMPANYJ WICHITA KANS AS, JAMES HAlNES AND JONATHAN HELLER, ELECTRIC dOMPANy; APPEARING ON BEHALF OF KANSAS 6AS a CLlrFORD BERTHOLF, WICHITA, KANSAS, APPEARING ON BEHALF OF KANSAS ELECTRIC POWER COOPERATIVES, lNC.J JOHN SIMPSON, FAIRWAY, KANSAS, APPEARING ON DEHALF OF KANSAS N ATUR AL RESOURCE COUNCILJ MILO M. UNRUH AND d!LO M. UNRUH, JR., WICHITA, KANSAS, APPEARING ON BEHALF OF VULCAN MATERIALS COMPANYJ
y
' THOMAS M. VAN CLEAVE, JR., PRAIRIE VILLAGE, KANS AS, AND ROBERT C. JOHNSON, ST. LOUIS, MISIOURl, APPEARING ON BEHALF OF B0EING MIL 1TARY AIRPLANE COMPANYJ
( BRIAN MOLINE, GENERAL COUNSEL AND RosERT M. FILLMORE, ASSISTANT OENERAL COUNSEL, APPEARING ON BEHALF OF THE I COMMIS$10N STAFF AND PUBLIC GENERALLY. INTRODUCTION
- THIS ORDER CONCLUDES PROCEEDINGS WHICH BEGAN ON MAY 22, 1984, WHEN THE COMMISSION STAFF FILED WITH THE COMMIS$10N A MOTION FOR AN ORDER ESTABLISHING A PREHEARING CONFERENCE IN THE CAPfl0NED DOCKETS. THE STAFF MOTION INDICATED THE PURPOSC OF THE PREHEARING CONFERENCE WAS To DETERMINE (1) THE APPROPRIATE TEST YEAR To BE UTILIZED IN RATE APPLICATIONS RESULTING FROM COMMERCIAL OFIRATION OF THE WOLF CREEK GENERATING STATION, (2)
THE APPROPRIATE IN* SERVICE CRITERIA TO BE UTILIZF.3 IN DETERMINING, FOR RATEMAKING PURPOSES, WHETHER THE %%F CREEK GENERATING STATION IS IN COMMERCIAL OPERATION, AND (3) WHETHER A HEARING EXAMINER SHOULD BE APP 0lNTED FOR THE PURPOSE OF RECElVING EVIDENCE AND L:"lTING ISSdES PRIOR TO THE HEARING OF RATE APPLICATIONS R hu'. T I N G FROM COMMERCIAL OPERATION OF THE WO' F CREEK 6ENERATING STATlJ BY ORDER ISSUED MAY 31, 1984, Tus COMMISSION GRANTED THE STAFF'S MOTION. FOLLOWING PROPER NOTICE TO ALL INTERESTED PARTIES, A PREHEARING CONFERENCE WAS HELD ON JULY 9, 1984, IN HEARING ROOM B, STATE CORPORATION COMMISSION, FOURTH FLOOR, STATE OFFICE BUILDING, IOPEKA, KANSAS. UPON THE MOTION OF ST W , THE COMMISSIONERS RETIRED . T0 CHAMBERS AND AN OFF-THE-RECORD DISCUSSION WAS HELD AMONG THE FARTIES IN AN EFFORT To LIMIT ISSUES SUBJECT TO EVIDENTIARY PROCEEDINGS BEFORE THE COMIS$10N. SUBSEQUENTLY THE HEARING WAS , RECONVENED AND THE PARTIES ANNOUNCED THAT IT WAS THEIR INTENT TO CIRCULATE FOR SIGNATURES A Si: DULATION CONCERNING (1) THE APPROPRIATE TEST YEAR TO BE UTILIZED IN RATE APPLICATIONS RESULTING FROM COMMERCIAL OP ER ATI ON OF THE WOLF CREEK GENERATING e STATION AND (2) THE APPOINTMENT OF A HEARING EXAMINER FOR THE (. PURPOSE OF RECEIVING EVIDENCE AND LIMITING ISSUES PRICR TO THE 1 2
y HEARING OF RATE APPLICATIONS RESULTING FROM COMMLHCIAL OPERATION OF THE WOLF CREEx GENERATING STAi!ON. IHE PARTIES INDICATED THAT THERE EXISTED SUBSTANTIAL DISAGREEMENT CONCERNING THE APPROPRIATE IN-SERVICE CRITERIA TO BE UTILIZED IN DETE3 MINING, FOR RATEMAKING PURPOSES, WHETHER THE WOLF CREEK 6ENERATING STATION l$ IN COMMERCIAL OPERATION AND THAT EVIDENTIARY PROCEEDINGS BEFORE THE COMMISSION WOULD BE REQUIRED. FOLLOWING COMPLETION OF THE HEARING THAT DAY, THE RECORD WAS CLOSED, SUBJECT ONLY TO RECEIPT OF TIMELY-FlLED BRIEF$, AND THE MATTER WAS 1AKEN UNDER ADVISEMENT. A COMPLETE RECORD AS DEVELDPED BEFORE THE COMMIS$10N. INE EVIDENCE INCLUDES 195 PAGES OF TESTIMONY AND THREE EXHIBITS. IN ADDITION, THE COMMIS$!ON HAD AVAILABLE IN ITS FILES THE ANNUAL REPORTS AND OTHER RECORDS OF THE KAN$AS 6AJ & ELECTRIC COMPANY (KG&E), KANSAf CITY POWER f, llGHT COMPANY (KCPL), AND KANSA9 ELECTRIC POWER COOPERATIVE, INC. (KEPCO). lHE FOLLOWINCi WITNESSES APPEARED ON BEHAl,F OF THE COMMIS$10N STAFF: HOSSEIN A. NOVIN CONSTRUCTION INSPECTOR FOURTH FLOOR, STATE OFFICE BLDG. IOPEKA, KAWS AS 66612 dAYNE A. WEBER CHIEF ENGINEER FOURTH FLOOR, STATE OFFICE BLDG. IOPEKA, KANSAS 66612 IHE FOLLOWING WITNESS APPEARED ON BEHALF OF KG&E: FORREST I. RHODES PLANT MANAGER KANSAS 6AS & ELECTRIC CO. P.O. 80x 208 - WICHITA, KS b72Ol NEITHER KCPL, KEPCO, NOR THE INTERvENORS PRESENTED TESTIMONY. UPON EXAMINING T et E TESTIMONY, EXHIBITS, FILES AND RECORDS IN THESE DOCKETS AND BEING OTHERWISE DULY INF0dMSD AND ADVISED IN THE PREMISES, THE COMMISSION MAKES THE FOLLOWING FINDINGS AND CONCLUSIONS: L s
O I. IlMLSEHON QYEllllE fAllHES 1 KGtE IS A CORPORATION ORGANilED AND EXISTING UNDER THE LAWS OF THE STATE OF KANSAS WITH ITS PRINCIPAL PLACE OF BUSINESS LOCATED AT 201 NORTH MARKET STREET, WICHITA, KANSAS. K6&E IS AN ELECTRIC PUBLIC UTILITY PURSUANT TO THE PROVISIONS OF K.S.A. 66-101 n n2 AND 17 HOLDS APPROPRIATE CERTIFICATES OF CONVENIENCE AND AUTHORITY TO ENGAGE IN THE BUSINESS OF THE SALE OF ELECTRIC SERVICE AT RETAll FOR DOMESTIC, COMMERCIAL AND INDUSTRIAL USES IN THE STATE OF KANSAS. 2 KCPL IS A CORPORAfl0N ORGANf2ED AND EXISTING UNDER THE L i,W S OF THE STATE OF MISSOURI WITH ITS PRINCIPAL PLACE OF BUSINESS LOCATED AT 1330 BALTIMORE AVENUE, KANSAS CITY, MISSOURI. KCPL l$ AN ELECTRIC PUBLIC UTILITY PURSUANT TO THE PROVISIONS OF K.S.A. 66 101 n LES. AND IT HOLDS APPROPRIATE CERTIFICATES OF CONVENIENCE AND A9THORITY TO ENGAGE IN THE BUSINESS OF THE SALE OF ELECTRIC SERVICE AT RETAll FOR DOMESTIC, COMMERCIAL AND INDUSTRIAL USES IN THE STATE OF KANSAS. 3 KEPCO 15 A CORPORATION ORGANIZED AND EXISTING UNDER THE { LAWS OF THE STATE OF KANSAS WITH ITS PRINCIPAL PLACE OF BUS t riE S S LOCATED AT 5709 WEST 21ST STREET, IOPEKA, KANSAS. KEPCO 15 AN ELECTRIC PUBLIC UTILITY PURSUANT TO THE PROVISIONS OF K.S.A. 66-101 H 1 12 AND IT HOLDS APPROPRIATE CERTIFICATES OF CONVENIENCE AND AUTHORITY TO E !4 G A G E IN THE BUSINESS OF THE SALE OF ELECTRIC SERVICE FOR RESALE TO 25 RURAL ELECTRIC COOPERATIVE MEMBERS IN THE STATE OF KANSAS, 4 A PETITION FOR LEAVE TO INTERVENE IN 00CKET NO. 142,098-0 WAS FILED BY THE VULCAN MATERIALS COMPANY-CHEMICALS OlVISION (VULCAN) ON JUNE 28, 1984 THE COMMISSION FOUND THAT THE PETITIONER APPEARED TO HAVE AN INTEREST WHICH WAR 1 ANTED PARTICIPAT! Ort IN THE PROCEEDINGS ANP, ACCORD I NGL Y, ITS PETITION FOR LEAVE TO INTERVENE WAS GRANTED FY AN ORDER DATED JULY 26, 1984 u l l l
o. N 5 A PETITION FOR LEAVE TO INTERVENE IN 00CKE7 NO.'S 120,924-U, 142,098-U, 142,099-0, AND 142,100-0 WAS FILED BY THE KANSAS NATURAL RESOURCE COUNCIL (KNRC) ON JULY 5, 1984. THE COMMISSION FOUND THAT THE PETITIONER APPEARED TO HAVE AN INTEREST WHICH WARRANTED PARTICIPATION IN THE PROCEEDINGS AND, . ACCORDINGLY, ITS PETITION FOR LEAVE (0 INTERVENE WAS GRANTED BY AN ORDER DATED JULY 26, 1984 6 A PETITION FOR LEAVE M INTERVENE IN DOCKET NO. 142,098-U WAS FILED BY THE 8 0 E '. . . *ilL I T ARY AIRPLANE COMPANY (BOEING) ON JULY 9, 1984 IHE COMMISSION FOUND THAT THE PETIT 10NER APPEARED TO HAVE AN INTEREST WHICH WARRANTfD PARTICIPATibN lM THE PROCEEDINGS AND, ACCORDINGLY, ITS PETITION FOR LEAVE TO INTERVENE WAS GRANTED BY AN ORDER DATED JULY 26, 1984 7 NOTKE OF THE JutY 9, 1984 HE ARING IN THE CAPTIONED DOCKETS WAS % RVED BY KC&E, KCPL AND KEPC0 UPON THEIR CUSTOMERS BY PUBLICATigN l '4 NEWSPAPERS OF GENERAL CIRCULATION WITHIN THE 6ERVICE TERRl10RY OF EACH RESPECTIVE UTILITY. KG&E, .KCPL AND KEPC0 FILED 4F71 DAVITS OF PUBLICATION WHICH VERIFIED PUBLICATION OF NOTICE. /t001 T 10N AL L Y , KG&E, KCPL AND KEPC0 MAILED NOTICE To m 8NTERVENORS IN EACH UTilliY'S LAST PRIOR RATE PROCEEDING BEFORE THE COMMIS$10N. THE NOTICE ISSUED WAS REASONABLE AND PROPER. 8 IHE COMMi!.SiON HAS JURISDICTION OVER THE PARTIES IN THE CAPTIONE'J PROCfEDING. II. E!ULCT MAlf ER JURISDiCTI0t{ 9 THE COMMIS$10N STAFF HAS ARGUED IN ITS BHIEF THAT THE. COMMISSION'S AUTHORITY AND JURISDICTION TO DETERMINE APPROPRIATE COMMERCIAL OPERATION OR "!N* SERVICE" CRITERIA FOR PUBLIC UTILITY PROPERTY IS ESTABLISHED AND SET FORTH IN 1984 KAN. SESS. LAWS, CHAPT. 247, SEC. 1 (1984 H.B. 2927) AS FOLLOWS: L 5
K.S. A. 66-J38 l$ HEREB7 AMENDED 70 READ AS FOLLOW $l 66-128 IME STATE CORPORATION COMMISSION SHALL ( DETERMINE THE REASONABLE VALUE OF ALL OR WHATEVER TRACTION OR PERCENTAGE OF THE PROPERTY OF ANY COMMON CARRIER OR PUBLic UTILITY GOVERNED BY THE PROVISIONS OF i Yh!S ACT WHICH PROPERTY IS USED AND REQUIRED TO BE USED 11 ITS SERvlCES TO THE PUBLIC WITHIN THE STATE OF FANSA3, WHENEVER THE COMMISSION DEEMS THE ASCERTAINMENT
. OF SUCH VALUE NECESSARY IN ORDER TO ENABLE THE COMMISSION TO FIX FAIR AND REASONABLE RATES, JOINT RATES, TOLLS AND CHARGES. lN MAKING SUCH YALUATIONS THE COMMIS$10N MAY AV All ITSELF OF ANY REPORTS, RECORDS OR OTHER THINGS AVAILABLE TO THE COMMIS$10N IN THE OFFICE OF ANY NATIONAL, STATE OR MUNICIPAL OFFICER OR BOARD. FOR THE PURPOSES OF THIS ACT, fROPERTYJF A PLY, f.y_P L ! C U_T.l L I T Y WHl(H H6S NOT B(fN COMPLETED _AlfR D.EDlfAIED TO CQMMERCIAL SERVICE SHALL NOT BE DEEMED 4Q EL.1L}_LD AND REQUl&ED TO BE U$12, IN THE PUBLIC UTILITY $
SERylCE TO THE PUBLIC, EXCEPT THAT, ANY PROPERTY OF A PUBLIC UTILITY, THE CONSTRUCTION OF WHICH WILL BE COMMENCED AND COMPLETED IN ONE YEAR OR LESS, MAY BE DEEMED TO BE COMPLETED AND DEDICATED TO COMMERCIAL SERVICE. (la. AT 1255, EMPHASl3 ADDED). 10 IHE STAFF CONTENDS IN ITS BRtEF THAT K.S.A. 66-128, AS AMENDED, ESTABLISHES A TWO-FOLD TEST FOR RATE BASE INCLUSION OF PUBLIC UTILITY PROPERTY. THE FIRST TEST IS THAT PUBLIC UTILITY PROPERTY MUST BE COMPLETED AND DEDICATED TO COMMERCIAL SERYlCE PRIOR TO RATE BASE INCLUSION (WITH AN EXCEPTION NOT APPLICABLE HERE). THE SECOND TEST, WHICH l$ RtACHED DNLY IN THE EVENT THE FIRST TEST 15 MET, !$ THAT PUBLIC UTILITY PROPERTY MUST BE USED AND REQUIRED TO BE USED AT THE Tine OF RATE BASE INCLUSION TO PROVIDE SERYlCE TO THE PUBLIC WITHIN THE STATE OF KANSAS. THE STAFF ASSERTS THAT THE COMMISSION'S AUTHORITY TO ADOPT COMMERCIAL OPERATION OR "IN-SERVICE" CRITERIA IS DERIVED FROM THE STATUTORY PROHIBITION OF RATE BASE TREATMENT FOR CONSTRUCTION WORK IN PROGRESS AND THAT THE ADOPTION OF SUCH CRITERIA IS REQUIRED IN ORDER TO DETERMINE THAT THE FIRST TEST HAS BEEN MET. 11 KG&E ARGUED AT HEARING (IR P. 165-106) AND IN ITS BRIEF THAT SECTION 1 0F K.S.A. 66-128, AS AMENDED, DEFINES THE COMMISSION'S AUTHORITY TO ASCERTAIN APPROPRIATE CRITERIA TO DETERMINE WHEN AN ELECTRIC GENERATING FACILITY HAS BEEN COMPLETED AND DEDICATED To COMMERCIAL SERvlCE. KCPl INDICATED AT HEARING THAT IT DEFERS IN THIS MATTER TO THE POSITION OF K6&E (IR P. 156). NO PARiv TO THif DROCEEDING MAS ARGUED THAT THE COMMIS$10N L 6 l l l
)
4 is WITHOUT SUBJECT MATTER JURISDICTION TO DETERMINE APPROPRIATE "lN-SERVICE" OR COMMERCI AL OPERATION CRITERIA. 12 THE COMMISSION BELIEVES THAT K.S.A. 66-128 (MPOSES UPON
!T A STATUTORY RESPONSIBILITY TO INSURE THAT PUBLIC UTILITY
- PROPERTY IS COMPLETED AND DEDICATED TO COMMERCIAL SERVICE PRIOR To RATE BASE INCLU$ ION. FURTHER, THE COMMISSION BELIEVES THAT IT IS PARTICULARLY IMPORTANT IN THE CASE OF PLANT WHICH REPRESENTS VERY COMPLEX AND INNOVATlvE TECHNOLO3Y THAT SPECIFIC OBJECTIVE CRITER!A BE UTILIZED IN MAKING THE DETERMINATION THAT SUCH PLANT IS COMPLETED AND DEDICATED TO COMMERCI AL SERVICE.
13 THE CCMMISSION HAS JURISDICTION OVER THE SUBJECT MATTER IN THE CAPTIONED PROCEEDING. Ill. Eg0aOSED "IN-SERVICE" OR COMMERCIAL UPERATION CRITERIA A. CRITERIA PROPOSED BY THE COMMIS$10N STAFF 14 RELYING UPON SEVERAL AUlHORITIES, THE COMMIS$10N STAFF g
, CONTENDS THAT UMTll SUCH TIME AS THERE EXISTS NOTHING SUBSTANTIAL REMAINING TO BE DONE ACCORDING TO CONTRACT DUE TO THE FACT THE WOLF CREEK GENERATING STATION IS ABSOLUTELY FINISHED, ENTIRE, AND FREE FROM DEFICIENCY, THE PLANT CANNOT BE DEEMED " COMPLETED" FOR RATEMAKING PURPOSES IN A C C 0 F: D A N C E WITH K.S.A. 65-128, AS AMENDED (IR, P. 6).
16 IHE STAFF ASSERTS THAT THE CRITERIA SPONSORED BY 7 RIASONABLY ASSURE THAT THE WOLF CREEK GENERATING STATION IS
" ABSOLUTELY FINISHED" AND " FREE FROM DEFICIENCY" AT THE TIME OF COMMERCIAL OPERATION. IHE COMMIS$10N STAFF PROPOSES THE ADOPTION OF THE FOLLOWING FOUR CRITERIA: (1) ALL NECESSARY PREOPERATIONAL AND START UP COMPONENT TESTING MUST BE DEMONSTRABLY COMPLETED (IR, P. 36, 63-64; STAFF BRIEF, P. 6-21); (2) THE TWO*HUNDRED FIFTY HOUR FULL WARRANTED OUTPUT PERFORMANCE 1EST OF THE NUCLEAR STEAM SUPPLY SYSTEM SHOULD BE SATISFACTORILY COMPLETED BY ACTUAL 7
TEST DEMONSTRATION AS ~ REQUIRED BY CONTRACT (IR, P. 29 31, 6ti-66; STAFF BRIEF, P. 21-29); (3) THERE MUST EXIST DEMONSTRABLY SUFF1-CIENT TRANSMISSION CAPACITY IN PLACE (EITHER OWNED OR OTHERWISE OBTAINED) TO CARRY THE TOTAL DESIGN NET ELECTRICAL CAPACITY OF THE RESPECT.VE OWNERS FROM THE GENERATING STATION TO THE DISTRI-BUTION SYSTEM OF THE OWNER UTILITIES (TR, P. 36-37, 69; STArF BRIEF, P. 29)J AND (14 ) THE UNIT MUST BE FIRED BY ITS PREDOMINANT FUEL SOURCE (IR, P. 69) STAFF BRIEF, P. 29). 16 MR. NOVIN EXPLAINED THE NEED FOR THE FIRST STAFF CRITERION AS FOLLOWS, INTER ALIM PRE-OPERATIONAL TESTING SHOULD BE ACCOMPLISHED TO DEMONSTRATE THE PROPER PERFORMANCE OF STRUCTURES, SYSTEMS, COMPONENTS, AND DESIGN FEATURES IN THE ASSEMBLED PLANT. TO ENSURE VALID TEST RESULTS, THE PRE *0PERATIONAL TESTS SHOULD NOT PROCEED UNTIL THE CONSTRUCTION OF THE SYSTEM HAS BEEN ESSENTIALLY COMPLETED AND THE DESIGNATED CONSTRUCTION TESTS AND INSPECTIONS HAVE BEEN SATISFACTORlLY COMPLETED. THE INITIAL TEST PROGRAM SHOULD BE DESIGNED TO DEMONSTRATE THE PERFORMANCE OF STRUCTURES, SYSTEMS, COMPONENTS, AND DEslGN FEATURES THAT WILL BE USED DURING NORMAL OPERATIONS OF THE FACILITY AND ALSO DEMONSTRATE THE PERFORMANCE OF STANDBY SYSTEMS AND FEATURES THAT MUST FUNCTION TO MAINTAIN THE PLANT IN A SAFE CONDITION IN THE EVENT OF MALFUNCTION OR ACCIDENTS. (TR., NOVIN, C P. 30) FURTHER, MR. WEBER JUSTIFIED THE FIRST STAFF CRITERION BY HIS FOLLOVING TESTIMONY, 1.filf,P, ALIA: AS SUPPORT FOR THIS CRITERIA, THE NUCLEAR REGULATORY COMMISSION REGULATORY GUIDE 1 68, REv. 2, AuG. 1978 STATES, AND I QUOTE: "SECTION XI, TEST CONTROL, OF APPENDIX 8, QUALITY ASSURANCE CRITERIA FOR NUCLEAR POWER PLANTS AND FUEL REPROCESSING PLANTS, TO CFR PART 50 REQUIRES THAT A TEST PROGRAM BE ESTABLISHED TO ENSURE THAT STRUCTURES, SYSTEMS, AND COMPONENTS WILL PERFORM SATISFACTORILY IN SERVICE. SINCE ALL FUNCTION $ DESIGNATED IN THE GENERAL DESIGN CRITERIA ARE IMPORTANT TO SAFETY, ALL STRUCTURE, CYSTEMS, AND COMPONENTS REQUIRED TO PERFORM THESE FUNCTIONS NEED TO BE TESTED TO ENSURE THAT THEY WILL PERFORM PROPERLY. IHESE FUNCTIONS, AS NOTED THROUGHOUT THE SPECIFIC GDC, ARE THOSE NECESSARY TO ENSURE THAT SPECIFIED DESIGN CONDITIONS OF THE FACILITY ARE NOT EXCEEDED DURING ANY CONDITION OF NORMAL OPERATION, INCLUDING ANTICIPATED OPERATIONAL OCCURRENCES, OR AS A RESULT OF THE POSTULATED ACCIDENT CONDITIONS.* THIS GUIDE HAS BEEN ATTACHED OR 1 SHOULD SAY, HAS BEEN MARKED AS EXHIBIT NO. 2, ACCORDINGLY AND ITEMS 1 AND 2 ARE DESIGNED TO ENSURE THAT, PRIOR TO THE DECLARATION OF COMMERCIAL OPERATION, THE PLANT OPERATES RELIABLY, AND AT THE OUTPUT LEVELS SPECIFIED BY CONTRACT. 8 m.
t IO BE M00E TECHNIC ALU PRECllt, l TEM NO. 1 (Q$URES VHAT ALL $YSTEM COMPONENil HAVE BEEN INDlVIDUALLY TE$TED TO PF 3RM PUR$UANT TO THE APPLICABLE PURCHASE
$F6ClFICAfl0N AND/0R TEST $PECIFICAfl0N. WITHOUT $UCH C TESTS, ONE CANNOT BE REA$0NABLY AS$URED THAT THE COMPONENTS WILL PERFORM OVER THE RANGE OF QPERAfl0N$
CONTEMPL ATED IN THE COMPONENT SPECIFICATION, OR THAT ANY DEFIClEICIES IF PRESENT, WOULD BE CORRECTED.
. ( I R . , W E B F. R , PP. 63-64).
17 MR. WEBER TESTIFIED ON $EHALF QF THE COMMllSION STAFF lN SUPPORT OF ST AF F (RITERl0N NO. 2 A$ FOLLOWS, jaita ALLO FURTHER, BY CONDUCTING THE NSSS PERFORMANCE TEST ADv0CATED IN ITEM 2, wt WILL AS$URE ACHitvEMENT OF THE i BENCHMARA 0F ACCEPT ABLE PERFORMANCE CONTEMPLATED BY THE PAR'l[$ WHEN EXECUTING THE CONTRACT. FOTH PARTIES AGREED THAT SUCH A DEMONSTRATl]N WA$, AFTER ALL, ACHIEVABLE. IHE NSSS CONTRACT SPECIFICALLY PROVIDES ON PAGE 10 Al FOLLOWS, AND l QUOTE "IHC PERFORMANCE WARMNTY FOR T1E NSSS WILL BE SAfiSFitD BY COMPLETION OF THE PERFORMANCE TEST OF 250 HOURS AT THE NSSS WARRANTED OUTPUT. IHl$ TEST WILL BE PERFORMED BY THE OWNER OPERATING THE EQUIPMENT WITH THE TECHNICAL ASSISTANCE OF WESTINGHOUSE. UETAILS OF THE TEST SCHEDULE AND PROCEDUnE WILL BE ESTABLISHED LY MUTyAL AGREEMENT BETWEEN WESTINGHOUSE AND THE OWNER. HE FERFORMANCE TEST WILL BE CONDUCTED AS $00N AS PRACTICABLE UPON REACHING FULL POWER. lN CASE OF INTERRUPTION OF THE TEST CAU$ED BY EQUlPMENT OR MATERIALS FURNISHED BY OTHER THAN WESTINGHOUSE OR FOR ANY OTHER CAUSE NOT THE FAULT OF EQUlPMENT OR MATERlALS FURNISHED BY WESTINGHOUSE, THE PERFORMANCE TEST WILL BE PERMITTED TO CONT!NUE, INSOFAR AS SAFETY AND GOOD C ENGINEERING PRACTICES PERMlf, UNTil 250 HOUR $ OF JPERAfl0N HAVE BEEN ACCUMULATED, INCLUDING AT LEAST 100 HOLAS OF CONTINUOU$ OPERATION AS PERMITTED BY $YSTEM LOAD DEMAND OR BALANCE OF PLAV .eraBILITY. lN C ASE OF INTERRUPil0N OF THE TEli CAU$ED BY EQUIPMENT, MATERIALS, OR SERVICES FURNISHED BY WESTINGHOUSE, THE PERFORMANCE TEST WILL BE REPEATED UNTIL 250 HOUR $ OF CONTINUOU$ OPERAtl0N OF THE NSSS HAVE BEEN DEMONSTRATED. IF THE COMPLET,'N OF THE PERFORMANCE TEST IS DEL AYED BEYOND THE $CHEbul.' OPERAfloh DATE BY MORE THAN A TOTAL OF 24 MONTHS DUE Th CAUSES OTHER THAN THOSE ATTRIBUTApLE TO WESTINGHOUSE UNL 4 THl$ CONTRACT OR THE NUCLEAR rVEL CONTRACT, THE FERF0 1ANCE TE$T mlL BE CONSIDERED A$ HAVl;eG BEEN SATISFACT06iLY COMPLETED.( APPARENTLY, KO&E AND KCP&L THOUGHT T' 4T THE STANDARD BolLER PLATE $PECIFICATION DEMONSTdATION OF 100 HOUR $ WAS INSUFFICIENT TO ASSURE BOTH RELIABILITY AND ADEQUATE DEMONSTRATION OF PERFORMANCE OF THE NSSS. CONSEQUENTLY, THE OWNER $ CONTRACTED FOR A 2SU* HOUR PERFORMANCE DEMONSTRATION TEST. IHE COMMIS$10N SHOULD NOT BE EXPECTED TO ACCEPT A LES$ $TRINGENT CRITERIA THAN THOSE CONTEMPLATED BY THE OWNER $. ADDITION ALE Y, THil TEST EXPOSES ALL OF THE OPLRATING SYSTEMS TO THE RlGOR$ OF FULL POWER STRE$$ES, TEMPERATURES, VIBRATIONS, TORQUES, CURRENTS, AND OTHER FULL LOAD CONDITIONS WHICH ARE NOT NECE$$ARiLY EXPERIENCED AT LOWER POWER LEVELS. (IR., WEBER, P. 64*b6). L 9
2 MR. NOVIN TE$flFIED 04 BEHALF OF THE COMMIS$10N $TAFF IN
$UPPORT OF ST AFF CRITERl0N h0 2 AS FOLLOWS, MLM ALJAt INE NUCLEAR STEAM SUPPLY SYSTEM CONTRACT llETWEEN K68[, '
KC P!l. AND WESTINGHOUSE ELECTRIC CORPORATION C L E A Rl, Y STATES THAT IN ORDER FOR THE WARRANTY OF NSSS EQUlPMENT
. TO TAKE EFFECT, A PERFORMANCE TEST WlLL BE CONDUCTED FORPOWER.
FULL 250 HOURS OF CONTINUOUS OPER4Tl0N UPON REAQHING
- IHE MAJOR COMPONEN15 0F NUCLEAR STEAM SUPPLY SY$ TEM ARE Al FOLLOWS NUnBER ONE, REACTOR VES$(LJ NUMBER TWO,
$1EAM GE NE R ATOR$ J NUMBER THREE, HEACTOR C OOL A N T PVMP$ AND DRIVESJ FOUR, PRESSURIZER INCLUDING HEATER $J NUMBER FIVE, VESSEL INTERNAL $J S!X, CONTROL ROD drive MECH ANISMS J SEVEN, SAFETY INJECTION PUMPl AND DRIVESJ ElGHT, REllDUAL HEAT REMOVAL PUMPS AND DRIVESJ NINE, BORON INJECTICN REclRCULATION PUMP AND DRIVE $J TEN, CHARGING PUMPS.
IHE OPERATION OF EACH EQUIPMENT MENil0NED ABOYE 15 AS (NPORTANT SY$ FEM ll 10 TM TOENilRE THE NSSS PLANT. SYSTEM Al OPERAfl0N OF NSSS NORMALLY, THE EQUIPMENT THAT WOULD BE OPERATING AT THE FULL POWER LEVEL WILL OPERATE AT $0 PERCENT POWER LEVEL. HOWEVER, EQUIPMENT OPERATING AT THE 50 PERCENT POWER LEVEt, WILL NOT EXPERIENCE MAXIMUM $ TRESS E.0ADING AT THE 100 POWER LEVEL. AT 100 PERCENT POWER OR FULL POWER, MORE STEAM WouLD BE GENERATED IN THE STEAM GENERATORS. IHEREFORE, THE STEAM GENERATORS HAVE GONE THROUGH THE14 ULTIMATE STRESS LOADING TEST.IURBINE VALVE 5 WOULD THEREFORE, GENERATE BEMORE FURTHER POWER. OPEN TO ADMIT NORE STEAM AND lHUS, THE TURBlNE HAS GONE THROUGH MORE RIGOROUS WORK AND $fRESS. C CONDENSATE WOULD BE PUMPED BY CONDENSATE PUMPS, MAIN FEED PUMPS AND HEATER liORE CRAIN PUMPS. lHUS, THE AFOREMENil0NED. LOAD CARRYING STAGEle EQUlPMENT HAS G0NE THROUGH ITS MAXIMUM L APPROPRIATE AT THE SELECTEDHo'MILESTONES D P0lNTS SHOULD BE E S T A BL I S HED BY K6&L THROUGHOUT THE POWER ASCENSION TEST PHASE TO ENSURE THAT RELEVANT TEST RESULTS ARE EVALUATED AND APPROYED BY PERSONNEL OR GROUPS DEllGNATED BY THE UTILITY PRIOR TO PROGAESSING WITH THE POWER ASCEN110N TEST PHASE. HOLD AS A MINIMUM, POINTS SHOULD BE ESTABLISHED FOR FWHS AT APPROXIMATELY 2S PERCENT, 50 PERCENT AND 76 PERCENT POWER LEVEL KEGULATORY TEST 6UIDE CONDITIONS 1 68 P.4. (IR., NOVIN,IN ACCORDANCE WITH NKC P. 29*31) 18 N. NOVIN TE$TIFIED IN SUPPORT OF STAFF CRITERION NO. 3 A S F OLL O WS , DilM Ak1A t
$1N!e 'HE RATEPAYERS OF KANSAS WILL BC PAYING FOR THE POWE-RATES, THEIR OENERATED FROM WOLF CREEK POWER PL AN T THROUGH IT l$ AN ABSOLUTE NECES$1TY FOR THE C0"PANY EL!CTRICAL TO CAPACITY BE *BLE TOToDELIVER 11$
THE TOTAL DES 10N NET KANSAS .1 A T E P A Y E R S . CERT IF IC A ll0N AND ANY OTHER REQUIRJD DOCUMENTATION SHOULD EVALUATIONBEAT PRESENTED THE TIME TOOF THE COMMI5510N S $TAFF FOR THEIR PLANT'S DEDICAll0N OF COMMERCIAL OPERATIONS. tid, h0VlN, P. 3h*j/). 10 _ _ _ _ - - _ _ _ - - - ^^ I
b 4 ADDITIONALLY, NR. WEBER $UPPORTED STAFF CRITER10N NO. 3 BY Hll F 0t.L O W I N G T E S T I MON Y , 131Lft e,LIAI OBvl0V$LY GENERATOR #lF$ THERE'$ OUTPUT NO TO THEME ANS OF ECONOMIC UTILITIE5' RATEPAYER DELIVERY
$, 17 OF A DOE $ THEM LITTLE GOOD. CONSEQUENTLY, THE LACK OF A DELIVERY SYSTEM $HOULD PRECLUDE A DFCLARATION OF COMMERCIAL OPERAfl0N. IHE TRANSMIS$10N F ACILITIES MUST BE DEllGNED TO ECONONICALLY HANDLE NOT ONLY THE FULL
$fCADY STATE ELECTalc LOADS, BUT AL$0 CERTAIN TYPICAL CONTINGENCIES ANP TRAN$ LENT NON* STEADY STATE CONDITlONS.
(IH, WEBER, P. 69) 19 MR. WEBER TESTlFIED IN SUPPORT 3F STAFF CRITER10N NO. 4 AS FOLLOWS, 131LR ALLA1 BECAU$E MOST UNITS, INCLUDING THis ONE, HAVE ALTERNATIVE FlRING MEANS FOR TE$flN3 AND $ FART *UP PURPOSES, A UNif CANNOT BE DEEMED, UNDER NORMAL CIRCUMSTANCES, TO BE IN COMMERCIAL OPERATION IF lT 15 RUNNING ON A BACKUP FUCL SOURCE. UNIT EFFICIENCIES WERE STRUCTURED AROUND THE PRIMARY FUEL SOURCE, AND CONSEQUENTLY, THE RATEPAYER AND lHE COMPANIES WOULD BE RECEIVING AN INFERIOR PRODUCT WERE THl3 TO BE ALL0wtD. (IK, WEBER, P. 69) 20 IHE PARTIES DlD NOT OPPOSE ADOPT ION OF STAFF CRITERl0N NO. 3 CONCERNING TRAN$Mi$$10N FACILITIES AND CRITERl0N NO. 4 CONCERNING THE PRIMARY FUEL $0URCE. (IR, P. 14). $EYERAL PARf!E$ OPPOSED THE ADOPfl0N OF STAFF CRITERl0N NO. 1 CONCERNING PREOPERATIONAL AND START-UP TE$ TING AND STAFF C R I T E R 10N NO. 2 CONCERNING THE NUCLEAR $7EAM $UPPLY $YSTEM PERFORMANCE TEST A$ INDICATED BELOW. B. CRITEnl0N PROPOSED BY KGtE AND KCPL. 21 KGtE CONTENDS THAT THE TERM " COMPLETED" IN K.S.A. 66-128 REFER $ TO COMPLEiloh 0F CONSTRUCTION. fURTHER, KGt6 CON-TENDS THAT THE PHRASE " DEDICATED TO COMMERICAL SERVICE" IN l K.S.A. 66-128 MEANS THAT THE OWNER ($) 0F A FACILITY HAYE DECLARED PROPERT" IN COMMERCIAL SERVICE SUBJECT ONLY TO A REvlEW OF REA50NABt.ENF$$ SY THE COMM15$10N. ;IX, P. Ib9*170, KG$E BRIEF, P.ll-15). , l l 11 1
O l 22 KGa[ PROPOSES THE ADOPfl0N OF THE FOLL0tflNG CRITERl0N! Al$UdlHG THAT THE NuCL E AR HEGUL ATORY COMMill10N HA$ NOT lMPOSED A LIMIT ON THE LEVEL OF P0wfR GENERAfl0N, KGtE C WILL CONSIDEH THE WOL F CREEK GENERATING $fAfl0N TO BE IN SERVICE, 1*E. COMMERCIAL OPERAfl0N, THE POWER USED, AND THE PLANT USEFUL UPON COMPLETION OF THE TE$f 5[QUENCE AT 50 PERCENT POWER, AS DESCRIBED IN ]H{ ATTACHED ' WOLF CREEK GENERATING STATION PHASE ll1 START-UP SCHEDULE', AND ANY SCHEDULED EXTENDED QUTAGE
- WHICH MAY OCCUR CL' RING POWER ASCEN$10N TE$flNO.
ALTHOUGH 1T l$ CL'NCEIVABLE THAT SOME CONDITION COULD ARl1E TO CAUSE THE NHC TO 00 $0, K6&E PRESENTLY HA$ NO REASON TO EXPECT THE NHL TO IMP 05E A LIMIT ON THE LEVEL OF POWER GENERAfl0N A? WOLF CREEK. IHE EXTENT AND DUR* AT10N OF THE P O W E lJ LIMITAT10N WOU(D DEPEND UPON THE CONDifl0N WHICH GAVE RISE TO 17 If l$ NOT POS$1BLE, THEREFORE, TO KNOW AT THl3 TIME WHETHER OR NOT A CON
- CElyABLE BUT UNEXPECTED POWER LIMITAfl0N WOULD AFFECT K6at,'S CON 51 DE R Af l0N OF WHEN WOLF CREEK WlLL BE IN COM*
MERCIAL OPERAfl0N. IN THE EVENT A P0wtR LIMjfATION WERE IMPOSED, KGtE Wouto IMMEDI ATtLY ADvist THE LOMMis-
$10N OF SUCH AND WOULD INDICATE WHAT EFFELT, IF ANY, SUCH LIMIT AT ION WOULD HAVE ON THE CONSIDERATION OF COM" MERCIAL OPERAfl0N.
(IN, RHODEE. P. 33*84). 23 dR. FORREST RHODES TESTIFIED IN $UPPORT of KGsE's C R I T E R lON A5 FOLLOWS, Jft11R ALIAt CONSTRUCTION OF WOLF CREEK WILL HAVE BEEN ESSENTIALLY COMPLETED PRIOR TO THE BEGlNNING OF THE INltlAL CORE L ADING TEST SEQU NCE. ( H, NHODES, P. 8) , , , I 4E NRC REQUIRES THAT EXTEN51YE TESTING BE UNDERTAKEN TO J'*URE THAT THE PLANT HAS BEEN PROPERLY CUN"
$1RUCTEu. INE START *UP PROGRAM At WOLF CREEK 15 DIVIDED INTO SEVERAL DIFFERENT TYPES OF PROGRAHIC FUNCTIONS. IHESE ARE FLUSHING, HYDROSTAfic TESTING, COMPONENT TESTING, PRE *0PERATIONAL TESTING, AND THE POWER ASCENSION TESTING. THE FLUSHING AND HYDROSTATIC TESTS ARE CONSTRUCTION COMPLETION TESTS TO VERIFY CLEANLINESS AND TO PRES 5URE TEST THE PlPING SYSTEM TO VERIFY STRUCTURAL INTEGRITY. PRIOR TO THE BEGINNING OF POWER A$CENSION, WE WILL HAVE COMPLETED ALL 608 0F THESE TESTS. INE COMPONENT TESTS ARE TO CHECK OUT A SPECIFIC PIECE OF EQUIPMENT. ALL OF THESE TESTS MUST BE COMPLETED PRIOR TO THE BEGINNING OF P0wtR ASCEN$10N. AT WOLF CREEK, THIS WILL INVOLVE $1,3/1 TESTS. (TR, RHODES, P. 87-68) e e e VPON COMPLETION OF TESTING AT THE $0 PERCENT POWER LEVEL, WOLT CREEK WILL BE A $AFE, RELIABLE, CONTINUOUS, AND $UBSTANTIAL SOURCE OF POWER AND ENLRGY FOR THE CUSTOMERS OF ITS OWNERS. IN THE CASE OF KG&E, FOR EX AMPL E , AT THE S0 PERCENT POWER LEVEL, WOLF CREEK WILL PROVIDE CAPACITY AND ENERGY NEARLY EQUAL TO Kb5E'S CAPACITY AND ENERGY FROM ITS SHARE OF Ts'0 UNITS AT THE JEFFREY ENERGY LENTER. (IK, RHODES, P. 9b*97).
L 12
- 24. KCPL INDICATED Al HEARING (IK, Pe }86) AND IN IT$ BRIEF THAT IT DEFERS IN THi$ MAITER TO THE Pollfl0N OF KGtE.
C. CRITER10N PROP 0$tD er KEPCO. 25 KCPCo PRESENTED NO EYlDENCE AT HEARING, HOWEVER, IT PRESENTED ITS POSITION THROUGH THE FOLLOWING STATEMENTS OF IT$ COUNS EL ,1Mia ALIAt flR . CHAIRMAN, KEPCO'$ POSITION ON THis 18, THAT WE FEEL THAT WOLF CREEK SHOULD OPERATE AT 100 PERCENT BEFORE IT
. IS DECLARED COMMERCI ALLY OPERABLE. WE FEEL LIKE THIS IS A NECESSARY RELI ABILITY CRITERIA AND WE BASICALLY FEEL THAT li SHOULD OPERATE At 100 HOUPS At 100 PERCENT OF FULL POWER AND THEN BE DECLARED COMMERCIALLY OPERABLE RATHER THAN THE 250 HOUR TEST THAT THE STAFF IS PROPOSING. WE FEtt, L.lKE THE 100 H0un$ is SUFrICIENT AND THAT THAT GIVES A GOOD RELIA 81LITY MEASURE AND THAT 11 KCPCO'S Posifl0N IN THis MATTER. (IR, P.25)
D. Crit <.RIA pro s: 70 av INTERvENORS 26 KNRC PRESET TED NO EVIDENCE AT HEARING, HOWEVER, IT ( INDICATED THROUGH STATEMENTS OF COUNSEL AND IN ITS BRIEF THAT IT SUPPORTS THE CRITERIA PROPOSED BY THE COMMIS$1ON STAFF. IV. nuuta 27 THE COMMISSIC-N BELIEYES THAT K.S.A. bb*128, AS AMENDED, ESTABLISHES A T W O - F O L (- TEST FOR RATE BASE INCLUSION OF PUBLIC UTILITY PROPERTY. IHE FIRST TEST ll THAT PUBLIC UTILITY PROPERTY MUST BE GQtLL11LA AND DEDICATED TO COMMERClAL SERVICE PRIOR TO RATE BASE INCLUSION (WITH AN EXCEPTION NOT APPLICABLE HERE). THE SECOND TEST, WHICH 15 REACHED ONLY IN THE EVENT THE FIRST TEST IS MET, 15 THAT PUBLIC UTillTY PROPERTY MUST BE USED AND REQUIRED TO BE USED AT THE TIME OF RATE BASE INCLUSION TO PROVIDE SERVICE TO THE PUBLIC WITHIN THE STATE OF KANSAS. IHE COMMIS$10N'S AUTHORITY, IN FACT OBLIGATION, TO ADOPT COMMERCIAL OPERATION OR
- L 13
'lN*$[RVICt' CRli(RIA is D(RIVID FROM THE $fATUTORY PROHlBill0N OF RAIL BASC TREAIMENT FOR C O N11 A UC T 10N WORK lN PROGRill. INC AD0Pfl0N Of $UCH CRl1CRIA l$ RLQUIR(D IN ORDER TO Dt f ERMINt THAT TH[ FIR $i TEST HAS BE[N MET.
28 IHE TERM *COMPLfitD* l$ NOT DlflNfD IN K. S. A. bb*l28, A$ AMEND (D, HOWEV(R, WORDS IN COMMON U$ AGE ARE 10 BC GivtN THClR NATVRAL AND ORDINARY MCANING IN ARRlVING af A PROPER CON $fAUCTION OF A sfATUTE. CULQL LLNLult_ (11LQfJLAIML, 233 KAN. 159, 6bl) P.201368 (1983). IHC TLRM *COMPLif f D* 15 DEFINf D IN OLACK'$ LAW Dicil0 NARY (RtylSED FOURTH [D., 1968) A f,1 flNilHEDJ MQ.t H j N Q . $ y $ $ 1M11 A L_J L MA LN j y Q _10_ B ( All J
$ TATE Of A THlNG THAT HAl BCCN CREATED, LRtCTED, CON $fRUCTED OR D 0 4 L__$ V t 11AhllA L LT _ _lC C Q R D1 Q __1Q A.C l. (CilAfl0N$ OMITTID, EMPHA$15 ADDED, D. AT SIMILARLY, THE TERM *COMPLift" Il DErlNED AT ISA C.J.S COMetttt 118 A$ FOLLOW $l 81!10LMllLL _. f_lN1$_dLQJ COMPLEftD OR CONCLUDED 1 COMMEN$UNATE, ENilRE, F lLL C D UP J ff L( EI10.OLflCl[hC.fu ELREEC1J INCLUDING CVERY lT(9 OR (LIM [Ni 0F THE THING
$POKEN OF, WifHOUT OMIS$10NG OR DCFIClLNCIESJ WHOLEJ
( LACKING NOTHINGJ W11H NO PART, ITEM OR ELEMENT LACKINGJ HAvlNG ALL NCEDED OR NORMAL paris, ELCMENil OR DETAILS (CITAfl0NS OMiffED, (MPHAll$ ADDED). I Sn AL 50, MOLLu11L_Y t.J1LLLu, 661 P. 2D 1145, 1146, 135 ARiz. , 444 (1983)J AND ((Lhk[Y V._(11LQL,1,qlMQiLM, 2b8 P. 2D 12,18, 124 C.A. 20 71 (1954)). THE COMMl5$10N BELitVgl THAT UNill $UCH TIME As THERE Exists *NOTHING $U8$fANTIAL REMAINING
- 10 BE DON [
' ACCORDING TO CONTRACT
- DUC 10 THE F ACT THE WOLF CREEK GENERATING STAfl0N l$ "ABSOLUT[LY FINISHCD,* *ENilRt*, AND
- FREE FROM DEFICitNCY', THE PLANT CANNDT BE DEEMED
- COMPLETED
- FOR RATEMAKING PURPOSES.
29 THE NUCLE AR NE GULATORY COMMi$S ION REGULATORY 6UIDE 1.h8 i (REY. 2, AuG. 1978) (EXHIBIT 2) at0Vint$ A PRL OPER Afl0N AL AND
$fARTUP Tt1T PROGRAM TO PROVIDE A$$URANCE INAT CONSTRUC110N HA$
l BEEN ACCOMPLl$HED IN A C C O R D A k r.E WITH DEslGN REQUIREMENT $1 IHE APPLICANT FOR A CON $1RUCil0N PERMIT OR OPERATlNG LICENSC !$ RESPON51BLE FOR ENbuRING THAT A $UITABLE l 14
I N I T I A L, (PRE 0PER A110NAL AND $f ARTUP) i[5t PROGRAM WILL BE CONDUCit0 FOR THE FACILITY. IHf PRIMARY 08J(CflVE$ OF A SUIT ABL E PROGRAM ARE (1) IQ lipijfLAtDjll0RAL Al$MAMC,L_ISAL.IEL_f> CIL L 1 Y H AL3 LLM_) D L2uALLLL_. C DLiishQ AND, 10 THE EKf(NT PRACilCAL, 10 VALIDA11 THE ANALfilCAL MODELS AND TO VERIFY THE CORRfCTNESS OF CONSERVATION OF A$$UMPfl0NS U$[D FOR PREDICTING PLANT RESPONSES TO ANT (CIPATED TRAN51[ Nil AND PO$TULAf(D ACCIDENT $ AND (t) IQ,__Plgy1Di _ &1)VRAA( L _I)1A L _ CQAllAVC11QM AND INSTALLAil0N OF COUIPMENT IN THE FACiLIiY HAYE Bf :N A CCMPL11HLLILA(SQ1213(LWjllL g(1113 (12. AT , . 6 8 -(l, LMPHAll8 ADDED). 30 MR. RHODES titTIFitD THAT POWER A$C[N$10N illTING ll At0UIRED TO 'XLRlil' THAT THE PLANT HAS BEEN PROP [RLY CONSTRUCTEDI Q. ARE ANY TEST $ UNDERTAKEN 'T O YLIL1fl._ lMA1 COM$1AVCittM OF SUCH A COMPL(X FACILjTY A$ THE NOLF LRLEK GENERATING $ fail 0N 11A1 IN FACl;_EL{3 CORLi1 G7 A. YEs. IHE NRC REQUIRES THAT f EILM$lyLIL$1131 Bt UNDERTAKEN lL.A$ $ V.RLlH,4LlHil&hLH A $jlih PA Q P.L* LLCO N 51AV C 11 D.. lHE START UP PROG 4AM Al WOLF CREEA 15 DivlDED J NTO $EV,[R AL DIF F E REN T TYPES OF PROGRAMIC FUNCTIONS. lHESE ARE FLUSHING, NYDROSTAilC TElflNG, COMPONENT TESTING, PRE *0PERAil0NAL TESTING, AND THE POWER A$CENSION ftsilNG. . . (IK., RHODES, P. 87, EMPHA$ll ADDED). ( 31 PROBLEMS ENCOUNTERED DURING THE POWER A1CEN510N TESTING PROCtll COULD RE$ ULT IN POWER REDUCil0N (INCLUDING TAKING THE UNIT OFF LINE) AND ADDifl0NAL CONSTRUCTION. IHE RELEVANT Guts f l0' IS WHETHER THE UNIT WILL OPERATE REllABLY IN ACCORDANCE WITH DEllGN REQUIREMENTS AT ANY POWER LEVEL PRIOR TO SAfl$ FACTORY COMPLtil0N OF PRE 0PE R Af l0N AL AND STARTUP TElilNG (INCLUDING THE NUCL(AR STEAM $UPPLY SYSTEM ACCEPTANCE TEST). INE AN$wtR TO THE QUESTION l$ UNCERTAIN A5 DEMONSTRATED BY THE F OL LOW I NG TESTIMONY OF f0RRilf RHODES: Q. MR. KHODES, YOU CANNOT, 1 ASSUME, A55URE THE COMMIS$10N THERE WILL BE NO SY$ FEM FAILURE RESULTING FROM FULL STRE$$ TESTING AT 100 PERCENT POWER 7 A. l CANNOT THE C0HH15510N THERE WILL BE NO FAILURES A$$UR[00 FROM 2 rtsCENT POWER TESTING. IHAT'S CORRECT. (IR., RHODES, P. 147). e e e Q. YOU WOULD AGREE IT l$ PO$51BLE THAT FUNCTIONAL TESYING At 100 PERCENT POWER COULD C AUSE A PROBLEM, REQUIRING THAT THE PLANT BE BROUGHT DOWN To BELOW 15
b
. 50 PERCEnf power r0R REPAlR$? THAT 15 s0METHING THAT l$ PO$$l>LE, l$N'T IT7
( A. Y $, II'$ POS$1BL ( R., RHODES, P. 9, EMPHAll$ ADDED). O. WHAT ARE THE CONSEQUENCE $ OF TE$f FAIEURE FOR THE PLANT TRIPPED FROM 100 PERCENT POWER 1EST7
- ACAlN, if WOULD DEPEND ON WHAT THE F AILURE 15 A. UNE OF THE IfiMS IN OUR DEllGN 18 10 GIVE REDUNDANT TYPE $ OF TR 50, IF ONE PARTICULAR TRIP FUNCTION DlDNlPl.T wCRK, THEN THAT TRIP FUNCil0N WOULD HAVE TO BE fixed. So, AGAIN, THE CONSEQUENCE $ OF FAILING THE TRIP WHEN A GlVEN BUTTON WAS PU$hED WOULD HAVE TO BE ANALY2ED. AND THAT PARTICULAR BUTTON ClkCUITRY woVLD HAVE TO BE CHECKED OUT AND REPAlRED.
Q. NIGHT THAT KIND OF PROBLEM REQUIRE THAT THE PLANT BE $ HUT D0wN FOR REPAlR7 A. lT WOULD ALL DE PE ND ON wHAT THE PROBLEM WA1 No$f 0F OUR TRIPS OCCUR **NO, NOT NECE$$ARILY. Most OF OUR TRIPS DCCUR IN $EVERAL CHANNELS. lF ONE CHANN(L GOES BAD THAT'$ NOT $1GNiflCANT, wE CAN CHANNEL IN BYPA$$ AND Fit IT AND G0 ON, WITH NO EFFECT ON THE OTHER CHANNEL $. $0, if DEPEND $ ON THE $PECIFIC TRIP FUNCil0N AND WHETHER OR NOT THAT WOULD C AUSE A PROBLEM OR WHETHER THE PROBLEM THAT l $ C AU$E D RLQQlRL $_IELW(LL9fj_lJM. IT MIGHT OR M GHT NOT BE THE CASE. ( R, WHODES, P. 145, EMPH A$ll ADDED). 32 IN LIGHT OF THE EXTEN51YE RECORD IN THi$ MATTER, THE COMMl5510N FlRMLY BELIEVES THAT IF A NUCLEAR
- FUELED ELECTRIC GENERATING PLANT l$ INCAPABLE OF OPERAfl0N IN ACCORDANCE WITH DEllGN REQUIREMENTS, if CANNOT REA$0NABLY BE CON 51DERED " FREE FRCM DEFIClENCY" OR
- ABSOLUTELY FINISHED" AND, THU$,
"C OMPL E T E D* . PR E
- 0 PE R A T I O N AL AND $TARTUP TESTING ll A REQUl$1TE TO THE DETERMINATION THAT A NUCLEAR GENERAilNG UNIT 15 CAPABLE OF OPERATION IN ACCORDANCE WITH DE$lGN REQUIREMENTS. ADDITIONALLY, THE COMM11$10N BELIEVES THAT IF A NUCLEAR
- FUELED ELECTRIC GENERAflNG PLANT ll INCAPABLE OF $UCCE$$FUL COMPLETION OF PERFORMANCE TESTS WHitH ACTIVATE WARRANTIES APPLICABLE TO THE NUCLEAR STEAM $UPPLY $Y$ FEM, IT CANNOT REASONABLY BE CONSIDERES
' FREE FROM DEFICIENCY" OR
- ABSOLUTELY FINi$HED' AND, THU$,
" COMPLETED
- AND THAT Sail 5 FACTORY COMPLETION BY ACTUAL TEST DEMON $fRATION OF THE 260 MOUR FULL WARRANTED OUTPUT PERFORMANCE TEST OF THE NUCLEAR $ TEAM $UPPLY $Y$ FEM l$ A REQVl$1TE TO THE
! 16
e
. 2 DETERMINAfl0N THAT A NUCLEAR GEhERATING UNif !$ CAPABLE OT OPERAfl0N IN ACCORDANCE WITH DEllGN AND CONTR AC TUAL REQUIREMENTS, l$ " FREE FROM DEFICIENCY", " ABSOLUTELY FINISHED
- AND
" COMPLETED".
33 OUR FINDING $ APPE 8,R TO 8E CON $llTENT WITH DEClll0NS IN DIJ_LCL 0L{.QX1MMRL'_.121*il E LY . PugLLLittt(IL1LL._fonstittaLg1 Onia, 391 N.E. 2D 311, 58 OHIO ST. 2D 499 (1979) AND 3L_1QQlMLQ (AL 1 F 0 RN.lA_(Q11QfL1QM P A(Y., 55 P.U.R. 4TH 537 (1983). IN QUIC1_, OL1Q11ELB$ '_ {1VliLEL h P.V2LLLV.ULLLILLCAMLiLL2LQf_ CMLQ,
. LQLM, THE SUPREME COURT OF OHIO OVERTURNED A COMMIS$10N ORDER WHICH $ ELECTED THE INITIAL SYNCHRONilAfl0N OF THE DAYll-dt$$1E UNIT NO. 1 NUCLEAR STAfl0N TO THE IOLEDO EDISON COMPANY TRAN$ MIS $10N SYSTEM A$ THE APPROPRIATE COMMERCIAt. OPERAfl0N OR "lN-SERVICE' CRITER10N. INE COUNT FOUND THATI WHILE THE INITIAL SYNCHRONIZATION OF A NUCLEAR GENERATING UNIT 10 ITS TRANSMIS$10N $YSTEM PRESENTS SOME INDICATION THAT A GENERAT]NG FACILITY l$ USEFUL FOR PURPOSES OF SUPPLYlNG $ERVICE TO RATEPAYERS, AS THE COMMiss10N FOUND, WE CCNCLbDE THAT, UNDER THE P'ACT$ AND CIRCUMSTANCES AT BAR, THE MANIFEST WElGHT OF THE
(' EYlDENCE DEMONSTRATE $ THAT, AT THAT DATE CERTAIN, THE UNIT IN QUE$Tl0N WAS VaQRqaliL11 AII,RL,.yP W A S N QL(QRELi1LQ U N T I L NOVEMBER l 1777. E111h2 M1LLL WHICH lM61 ILL_LLw ALu1EXQELtdL1HEA_.IMLVRLL' L11111HLtQ.VLQ LuntlLQA LLAN 1M1EGRALUL MX1LLMQ_ C21LLWL10_aQ j2 LN THE PROXIM ME FUTUqi. Q2. AT 314, EMPHA$15 ADDEDi. 1T WOULD BE 11L2V111 ELL 10. P R EM AT V R.1LI._,_$d I F T THE EL11 7 DF PLATT FAILQRL FROM THE UTILITY $ INVESTORS TO THE RATEPAYERS BY THE INCLUSION IN THE RATE B ASE QL4tGhL1 CanELEX Aha_.LNJt01A_LLtL1LthiQiaGY WHICH HAS N Q.Lt1LS ER0 YEN T Q,tLR EA LQJi A.) 4F R EL1R_0Q tLi.lG3 LLL(A Ml_Q L $.LQ3_QE CQNLLRJLCIl0N DE F LC11 lHE INITIAL RISK OF FAILURE IS APPROPRIATELY BORNE BY THE INVESTORS, WHO HAVE UNDERTAKEN THE PROJECT AND WHO WILL ULTIMA 1ELY PROFif FROM 175 SUCCESS. IT IS ONLY PROPER THAT THilR VENTURE BE FOUND OPERAtl0NAL BEFORE THEY COMMENCE TO RECOUP THEIR CAPITAL OUTLAYS FROM THE CONSUMERS. QQ. AT 315, EMPHAS15 ADDED). lN EL.JAQJEE RLL,_[ AL1LQX41 A (DLiDN. COMPANY. SUPU., THE CAtir0RNIA COMMIS$10N HELDI ON JUNEH.C00gCOMMERClatUPERATION 14, 1 82, THE ADMINISTRATIVE LAW UATE] JUDGE CRITERIA ISSUED A RULING IN THis PROCEEDING THAT THE COMMERCI AL OPERATING DATE WOULD OCCUR WHEN ALL INITIAL START *UP TESTING INCLUDING THE WARRANTY AUN, HAD BEEN SUCCESSFULLY COMPLETED. lN DECISION 82-09-111, IN RESPONSE TO 17
*s 4 EDi$0N'S PETITION FOR REHEARING, UE AFFIRMED THE A.L.J.'$ RULING. IT MUST BE DEMON $(RAIED TO THE STAFF's $ATISFACTION THAT THE PLANT WILL MEET THE CRITERIA IN AN ACCEPTANCE TE$f 0F 200 HOUR $ OF C CONTINUOU$ OPERATION (THE WARRANTY RUN).
DEMON $TRATION IN THE ACCEPTANCE TEST, THAT THE PLANT WE ylEWED THE COULD MEET THE REQUIRED CRITERIA, TO BE THE BEST INDICAfl0N OF THE COV. AT THE JULY 15, 1983, ORAL ARGuMcNT, EDil0N ONCE AGAIN RENEWED ITS REQUEST THAT THE COMMIS$10N RECON $lDER ITS C00 CRITERIA. EDI$0N ARGuto THAT THE RioVIREMENT THAT THE PL ANT COMPLETE IT$ 100 PERCENT POWER LEVEL TESTING AS WELL AS THE 200" HOUR WARRANTY RUN l$ T00 RIGID AND UNRELATED 70"wHAT $HOULD BE gDNSIDERED FOR DECLARING A ANT TO BE U$ED AND USEFUL . LDi$0N STATED ON JULY P(5, 1 1983, THAT IT HAS RESUMED POWER TESTING AT THE 100 PERCENT POWER LEVEL AND THAT 17 WOULD $00N COMPLETE ALL Of ITS TEST $ AT THE 100 PERCENT POWER LEVEL Al WELL AS Th4 WARRANTY RUN. lN RENEWING ITS REQUEST FOR RELON$1DERAfl0N OF THE CVU CRITERI A AT THIS LATE DATE EDISON MUST BE CONCERNED ABOUT THE PRECEDENT!AL NATURE OF OUR CUD CRITERI A. WE AGREE THAT CUR CRITERIA ARE RIGIDJ HOWEVER, W EJELLUE IT WAS AND IS M TH( JLAT E P A Y E P S ' I N REL11._LQ_ill H I G1Lil AED AMLl0_ &i1RL QURSELVE1 THAT T 4Lf'LA N T w Qy(D_lLL(AflgkLQ F P R01Q(hq POWER AS f1)LM LQJ EVEN THOUGH THE DELAY IN PLACING THE PLANT INTO SERYlCE MAY RESULT IN $0MEWHAT HIGHER COSTS TO RATEPAYER $ IN THE LONG RUN. M j (LIR L J1_w,,QQLQ tL(y.L jl R N M0RE DET RIMigi AL TqAyLALLaw EDELf1L 3 MlE BA$E Af1EjLltLL _. Id P E R _ CLN T P QW{RJLill,,,QQL P E R C E N Lf_Q wlfL T E 5 T $ H A D.liEh._f,.Q3f1LRQ ,AgjL1, 4 E N T 03Ay1 lukiL.10EiLihill._n 1 T H A SQLp er rLnT w?' m jlf.CAkiL QF THE N(LL1Q..AE.P A 1 R F L A w 1_qfLALEl(li_JfL,lH,L 2.LAll . WHILE THE IMPOSITION OF OUR RIGID CRITERIA IN C NO WAY GUARANTEE $ THAT THE PLANT WILL OPERATE WITHOUT TROUBLE, IT DOES INDICATE THAT WE HAVE TAKEN REASONABLE STEPS TO ASSURE ALL P ARTIES THAT FLAWS AND DEFECTS WILL HAVE BEEN ELIMINATED AND CORRECTED TO THE EXTENT POS$1BLE. WITH A JATISFACTORY EXPERIENCE ON SONGS 2, WE MAY INDEED AGREE THAT THE CRITERIA SHOULD BE RELAXED. WE WILL BE WILLING TO HAVE APPLICANTS REOPEN THis ISSUE IN CCNNECTION WITH SONGS 3 OR ANY OTHER NUCLEAR POWER PLANT PROJECT. (1Q. AT 573, EMPHA$l$ ADDED). 34 INE COMMISSION Bell;VES THE RI$K THAT A PLANT CONSTITUTING NEW AND COMPLICATED TECHNOLOGY WILL BE INCAPABLE OF OPERATION IN ACCORDANCE WITH DE$1GN REQUIREMENTS IS A RISK PROPERLY BORNE BY THE APPLICANTS SHAREHOLDERS. IHE EVIDENCE PRESENTED BY THE COMMISSION STAFF IN SUPPORT OF ITS PROPOSED CRITERIA NUMBERED 1, 2, 3 AND 4 l$ CONVINCING AND THE CCMMISSION FINDS EACH OF THE FOUR CRITERl0N TO BC BOTH REASONABLE AND APPROPRIATE FOR APPLICATION IN THESE DOCKETS. II 15, THEREFORE, BY THE COMMISSION ORUEKE0 THAT: 1 IHE FOLLOW!NG FOUR "IN*SERvlCE" OR COMMERCIAL OPERATION CRITERIA PROPOSED by THE COMMISSION STAFF $ HALL BE APPLIED IN THE l 18
D l CAPTINED DOCKET S FOR THE PURPOSE OF DETERMINING, FOR RATEMAKING PURPOSES, WHETHER THE WOLF CREEK GENERATING STATION HAS BEEN
" COMPLETED AND DEDICATED TO COMMERCIAL $[RYlCE" AND MAY PROPERLY BE DEEMED TO BE IN COMMERCIAL OPERATION IN ACCORDANCE WITH THE TERMS OF K.S.A. 66-128, AS AMfHDED:
A. ALL NECESSARY P R EO PE R A t l 0N AL AND START *U* COMPONENT TESTING MUST BE DEMONSTRABLY CCMPLETEDJ B. THE Two*HUNDRED FIFTY HOUR FULL WARRANTED OUTPUT PERFORMANCE TEST OF THE HUCLEAR STEAM SUPPLY SYSTEM SHALL BE SAflSFACTORILY COMPLETED BY ACTUAL TEST DEMONSTRAtl0N AS REQUIRED BY CONTRACTJ
. C. IHERE MUST EXlST DEMON;TRABLY SUFFlCIENT TRANS*
MIS $10N CAPACITY IN PLACE (EITHER OWNED OR OTHERwlSE OBTAINED) TO CARRY THE TOTAL DESlGN NET ELECTRICAL CAPACITY OF THE RESPECTIVE OWNERS FROM THE GENERATING STATION TO THE DIS 1RIBUT10N SYSTEM OF THE OWNER UTILITIES) AND 0 THE UNIT MUST BE FIRED BY ITS PREDOMINANT FUEL SOURCE. 2 EACH OF THE SPECIFIC FINDINGS OF FACT ABOVE STATED ARE HEREBY ADOPTED AS ULTIMATE FINDINGS AND CONCLUSIONS OF LAW BY THE (.0 MM I S $ 10 N . IHE COMMi$SION RETAINS JURISDICTION OF THE SUBJECT MATTER AND THE PARTIES FOR THE PURPOSE OF ENTERING SUCH FURTHER ORDER OR ORDERS-AS DEEMED NECESSARY AND PROPER. BY THE COMMISS10N IT IS SO ORDERED. DATEU: OCTOBER 9, 1984 LENNEN, CHMN.: LOUX, CUM.: HENLEY, COM.
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. MAR 1113N Hr. Glenn L. Koester Vice President - Nuclear Kansas Gas and Electric Company 20] North Market Street Post Office Box 208 Wichita, Kansas 76201
Dear Mr. Koester:
Subject:
Issuance of Facility Operating License NPF-32 Volf Creek Generating Station, Unit ! The U.S. Nuc1*ar Regulatory Comission (NRC) ha$ issued the enclosed r Operating License NPF-32, together with Technical Specifications and Environ-ecility .. mental Protection Plan for Wolf Creek, Unit 1. License No. NPF-32 au;horires operation of Wolf Creek - 3411 tr.egawatts thenel (,100T power). Unit 1 at reactor core power levels not in e Pa n d i n glo2rnhsiqrta pe r.o ral.m etAliv is-restricted-to7cwer-levelnnrto megawatts themal). exceed 5 percent of full power (170 ( Enclosed is a copy of a related notice, the original of which has been for-warded to the Office of the Federal Register for publication. . Four signed copies of Amendment No. 2 to indemnity Agreement Nn. B-99 which covers Please sign the all activities copies andauthorized return oneunder to this License office. No. Nff-32 are also enclosed An FinalAssessment Environmentalof the Effect Statement of License for the Wolf Crock Duration (;enerating onStation, Hatters UnitDiscus:e 1 is enclosed as Enclosure 4 Sincerely, i ! f A. J ! Nuph . Thompsey J f rec tor Dit 4., ion of LicV.ns 's Office of Nuclear Reactor Regulation
Enclosures:
- 1. Facilitl< Operating License HPF-3?
- 2. Federal Reoister Notice 3
Eendment No, fto Indemnity Agreement No. B-99 !( 4 Assessment of the.Effect of License Duration on Hatters Discussed in the FES , cc w See ext n/ page enclosures: A
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(( g- m%, 'a NUCt.C AR REGUL ATORY COMMISSION p , - < WAs motow o. c.nossi e, g . .L< ... g}NSAS GAS AND ELECTRIC COMPANY KANSAS CITY POWER A LIGHT COMPANY KANSAS CLECTRIC POWER C00pfRATIVE, INC. DOCKET NO. STN 50-482 WOLF CREEK GENEPAllNG STATION, UN!_7 NO. 1 TACILITY OPERATING l.! CENSE License No. NPF-3?
- 1. The Nuclear Regulatory Comission (the Comission) has found that:
A. The application- for liceny filtd_by_lansas Bas _and Clectr4c-{ompany, KanrWCity~ Power & Light company, and Kansas Electric Power Cooperative, '~ Inc. (licensees). complies with the stantiards and requirements of the Attcic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in -10 CFR Chapter I, and all required notifica-tions to other agencies or bodies have been duly made; B. Construction of the Wolf Creek Generating Station, Unit No.1 (the facility) has been substantially completed in conformity with Con-struction Permit No. CPPR-147-and the application. as amended, the provisions of the Act, and the regulations of the Comission; C. The facility will operate in confonnity with the application, as amended, the provisions of the Act, and the regulations of the Com - mission, (except as exempted from compliance in Section 2.0 below); D. There is reasonable assurance: (1) that the activities author 12ed by this operating license can be conducted without endangering the health and safety of- the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in CFR Chapter 1, (except as exempted from conpliance in Section20below); . E. Kansas Gas and Electric Company
- is technically qualified to engage in the activities authorized by this licerse in accordance with the Comission's regulations set forth in 10 CFR Chapter I;
~(- _ _ _ . Kansas Gas and flectric Company is authorized to act as eoent for the Kansas City-Power & Light Company and the Kansas Electric Power Cooperative, Irf., and has exclusive responsibility and control over the physical construction, opere. tion and maintenancejof the facility. 4 ma~ A$deeeAeema.m in s a # Aa e
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P. The licensees have satisfied the applicable provisions of 10 CFR part 140 " financial Protection Recuirements and Indemnity A9 rcements," of the Comission's regulations; G. The issuance of this licorse will not be inimical to the comon defense and security or to the health and safety o' the publici H. After weighing the envirnnmental, econorric, technical and other bene-fits of the facility against environmental and other costs and con-sidering available alternatives, the issuance of this facility Oper-ating t,icense No. NPF 32, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CTR Part 51 of the Comission's regulations and all applicable requirements have bun satisfied; and 1. The receipt, possession, and use of source, byproduct and speci41 nu-clear material as authorized by this license will be in accordance with the Comission's regulations in 10 CTR Parts 30, 40 and 70, 2. Based on the foregoing findings regarding this facility, facility Operating license No. NPF-30 is hereby issued to Kansas Gas and Electric. Company,--- -. Kansas-City Power &-Light"Cenpan DFid Kansas ~ Electric Power Cooperative, ( Inc. (the licensees) to read as follows: A. The license applies to the Wolf Crock Cenerating Station Unit No.1, a pressurized water nuclear reactor and associated equipment (the fa-cility), owned by kansas Gas and Electric Company, Kansas City Power
& Light Company, and Kansas Electric Power Cooperative, Inc. The facility is located in Coffey County, Kansas, approximately 28 miles cast-southeast of Emporia, Kansas, and is desc.15ed in the licensees'
" Final Safety Analysis Report", as supplemente. 1 amended, and in the licensees' Environmental Report, as supplemeented and amended.
B. Subject to the conditions and requirements incorporated herein the Com-mission hereby licenses Kansas Gas and Electric Company (kG&E), Kansas , City Power & Light Company (KCPL) and Kansas Electric Power Cooperative, Inc.(KEPCO). (1) Pursuant to Section 103 of the Act and 10 CFR Part 50 "Oomestic Licensing of Production and Utilization racilities," KG&f, to possess, use and operate the facility at the designated location in Cof fey County, Kansas, in accordance with the procedures and limitations set forth in this license; (2) KCPL and KEPC0 to possess the facility at the designated location in Coffey County, Kansas, in accordance with the procedures and limitations set forth in this license; 1 s
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. (3) and use at any fJme special nucicar materia accordance with the lirnitations for storage and amounts reovired for reactor Report, operation, as supplemented andas described amendedt in the Final Safety Analys (4) ceive, possess, and use at any time any byprodu source and special nuclear material as sealed neu'.ron sources, for reactnr startup, sealed sources for reactor instruthentation ard radia.
tinn in amounts tnonitorini, eculpment as required; calibration, and as fission detectors (5) KG&E, pursuant ceive, possess,to the Act and 10 CFR Parts 30, 40 and ?O, to re-and use in amounts As required any byproduct, source or special nuclear material without restrictinn to r. hem. ical or physical form. for sample analysis or instrument cali-bration and or associated with radioactive apparatus or components (6) LG6E, pursuant to the Act but not separati.-~siich~bfp~end_10..CERJort:; 30, AO and 70rto possess - i roduct and special nuclear materials as may be produced by the operation of the facility, C. specified in the Ccmission's regulations set forth and is subject to all applicable provisions of the Att and to the roles, regulations, and orders of the Cornission now or hereafter in effect; is subject to the additional conditions specified or incorporated belo (1) Maximumfewer level KG&E is authorized to operate the facility at reactnr core power levels not in excess of 3411 megawatts themal- (100t power) ih accordance with the conditions specified herein and in Attach-ment 1 to this license. arid be completed other as items identified in Attachtrent 1 to this license s specified. Attachment 1 is hereb into this license. Pending Comission approval, ythis incorporated liter.se is restricted to pnwcr levels not to exceed 5 percent of fv11 power (170 megawatts thernal); (2) Technical Specifications and Environrnental protection Plan . The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which license.are attt.ched hereto, are hereby incorporated into this l KG&f shall operate the facility in accordance with Plan;the Technical $pecifications and the Environmental prntection . i n e t nrMoc ry t te n t oC6f!\
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e 4. (3) Antitrust _ Conditions 4 ' Kansas Ces & Electric Cottpany and Kansas City Power & Light Company shall comply with the antitrust conditions delineeted in Appendix C to this license, (4) Environmental Ovali "r.ation (Section 3.112.5_SEp #4,_Section y - All electrical couipment within the scope of 10 CFR 50.49 shall be qualified by November 30, 1905. (5) Seismic end Dynamie.0ualificatinn (Section 3.10,_55,fR 4) Prior to exceeding five percent of rated power. KG&E shall, for that equipment which is not completely qualified, complete such qualification or submit iustification for safe operation at power levels greater than five percent. (6) fire Protection (Section_9.5.1. S[R, Section 9.5.1.8, SSEA d5)
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(a) KG&E shall maintain in effect a11 provisions of the approved I fire protection program as described in the SNUPpS final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, and as approved in the SER through Supplement 5, subject to provisions b & c below. (b) KG&E may make no change to the approved fire protection program which would decrease the level of fire protection in the plant without prior approval of the Ccanission. in rale such a change the lir7nsee must submit an application for license amendment pursuant to 10 CFR 50.90 ,. (c) KGAE may make changes to features of the approved fire pro-tection program which do not decrease the level of fire pro-tection without prior Cornission approval, prnvided: (i) such changes do not otherwise involve a change in a license condition or technical specification ., or result in an unreviewed safety question (see 10 CFR 50.59). De parenthetical notation following the title of many license conditions ( wherein the license condition is discusseddenotes the section of the S , i e
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- $- i (ii) such changes do not result in failure to complete !
the fire protection program approved b Comission prior to license issuance. y (he KGAE shall maintain in an auditable form, a current record of all such changes, including an analysis of the effects of the change on the fire proter. tion program and shall rnate such records available to NRC inspectors upon request. All changes to the apprnved program tr.ade without prior Comission approval shall be reported annually to the Director of the Office of Nuclear Reactor Regulation, together with suppnrting analyses. (7) ' kselification of Personnel (Section 13.1.2, SSER Q, Sect _fon 18,
.SO.D.
KG&E shall have on each shift operators who meet the requirements described in Attachment 2. . (8) NUREG-0D7 Supplement 1 Conditions (Section ??, SER) KG&E shall completejheleqqirements_ described in. At.tachment tb~the'Tatd faction of the NRC. These conditions reference the appropriate items in Section 22, "THI Action Plan Requirements ( for Applicards for Operatin Report and Supplerents 1. ?g , 3,Licenses." 4, and 5 NUREG-0881. in the Safety Evaluation (9) Post-Fuel-loadin0 Initial Test Program (Section 14. SER Section 14, T50 i5) Any changes in the Jr.itial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change. (10) Inservice Inspection Prnora., (Sections 5.2.4 and 6.6, SER) Within nine months of the date Of this license, KGAf shall submit for staff review and approval, the inservice inspection program which confoms to the ASHE Code in et T,*ct 12 months prior to the date of issuance of this license, til) Emergency Planning (d in the event that the NRC finds that the lack of progress in completion of the procedures in the rederal Emergency Fanagement Agency's final rule, 44 CfR Part 350, is an irdication that a ma,inr substantive problem exists in achieving or maintaining an adequate state of emergency ( preparedness, the provi,sions of 10 CFR Section 50.54(s)(2) will apply. 1 i e a . n c e 6 w. w , . . , . . na e.* , e a * * *
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