ML20011D735
| ML20011D735 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1989 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20011D723 | List: |
| References | |
| NUDOCS 8912280337 | |
| Download: ML20011D735 (16) | |
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TABLE OF CONTENTS-Chapter 1 f
1.1 INTRODUCTION
AND PURPOSE OF THIS REPORT
1.2 BACKGROUND
AND PLANT DESCRIPTION 1.
2.1 BACKGROUND
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1.2.2 PLANT DESCRIPTION 1.3 SCOPE, CHARACTER, AND CONCLUSIONS OF THIS REPORT j
1.3.1 SCOPE 1.3.2 CHARACTER i
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3.3 CONCLUSION
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1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS I
i 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5.1 INTEGRATED ASSESSMENT OF ISSUES.
1.5.2 PROBABILISTIC RISK ASSESSMENT (PRA) 1.5.3 SYSTEMATIC EVALUATION PROGRAM (SEP) i 1.6 MATERIAL INCORPORATED BY REFERENCE 1.7 DRAWINGS AND OTHER DETAILED INFOKMATION
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CHAPTER 1 INTRODUCTION AND CONCLUSIONS
1.1 INTRODUCTION
AND PURPOSE OF THIS REPORT This report presents information to give reasonable assurance that-the nuclear power plant as described will be operated by Consumers Power Company at Big Rock Point without undue risk to the health and safety of the public.
The updated information is presented to support Consumer Power Company's facility operating license No. DPR-6 for the Big Rock Point nuclear reactor in accordance with requirements established by Paragraph 50.71(e) of 10CFR, Part 50.
The function of the report will be to present the design bases and limits on facility operation, describe the facility, and present the safety analyses of selected structures, systems, and components.
The primary purpose of this report is to demonstrate, on the basis of existing technology and experience, that the plant can be operated within the scope of the operating license without undue risk to the health and safety of the public.
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1.2 BACKGROUND
AND PLANT DESCRIPTION 1.
2.1 BACKGROUND
The general features of the Big Rock Point nuclear power plant including the pertinent details of the site are described in this report and were described in the following previous reports:
- 1) " Big Rock Point Nuclear Power Station, Application to U.S.
Atomic Energy Commission for Reactor Construction Permit and Operating License Part B, Preliminary Hazards Summary Report,"
January 14, 1960.
- 2) " Amendment No. 2 to Application for Reactor Construction Permit and Operating License, ' Revised Hazards Summary Report,'"
October 14, 1960.
- 3) " Amendment No. 3, to Application for Provisional Operating License, ' Revised Final Hazards Summary Report,'" November 14, 1961 as revised 3/12/62, 3/19/62, and 3/23/62, Revision 1.
On the basis of submittal (1), the U.S. Atomic Energy Commission issued a construction permit (No. CPPR-9) on May 31, IMO.
g On the basis of submittals (2 and 3) as supplemented and amended,
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Operating License No. DPR-6 was issued and remains in effect.
Events of significance are:
Provisional Operating License issued August 30, 1962 Initial Criticality was achieved September 27, 1962 Initial Power Operation was achieved December 8,1962 The Date of Commercial Operation was March 29, 1963 Power level was increased from 157 MWt to 240 MWt in May,1964 The Full Term Operating License was as of May 1, 1964 In accordance with the above authorization, Consumers Power Company is operating a direct cycle, forced circulation boiling water reactor at the Big Rock Point site, which is located in Charlevoix County, between the towns of Charlevoix and Petoskey, on the northern shore of Michigan's lower peninsula.
1.2.2 BIG ROCK POINT PLANT DESCRIPTION This brief BRP plant description is intended to make the reader aware of the major design differences between BRP and the more (n
recent model boiling water reactors. (Refer to Figure 1.1) 1.2-1 MIO587-0194A-BX01
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I BRP is the second oldest commercial nuclear power plant still operating in the U.S and at a rated electrical output of 75 MWe, it is the smallest.
BRP is a direct cycle, forced circulation boiling water reactor capable of producing 240 megawatto thermal at a nominal operating l
pressure of 1350 psia. The electric generating capacity is 75 megawatts at this thermal output.
1.2.2.1 Reactor Vessel and Steam Drum The BRP reactor pressure vessel is 30 feet in overall length and 106 1
inches in diameter. The reactor pressure vessel has 61 penetrations, the largest two being 20 inches in diameter.' The active fuel length is 70 inches and there are 32 bottom-entry control rods.
Steam separation and feedwater addition occur in a separate steam drum rather than in the reactor pressure vessel.
Subcooled liquid enters the reactor vessel from two constant flow recirculating water pumps near the bottom of the vetsel.
Each recirculating pump is capable of pumping six million pounds per hour of coolant. As the coolant passes through the core, it is heated to a steam-water mixture.
Steam baffles, located approximately six feet above the top of the active fuel, force the steam-water mixture out six 14-inch risers. The steam-water mixture travels up the risers to a
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steam drum located approximately 30 feet above the top of the reactor pressure vessel.
The steam drum is 40 feet long and 78 inches in diameter.
It provides three basic functions:
(1) steam separation from the steam-water mixture occurs in the drum with turboseparators and screens, (2) feedwater addition and mixing and (3) it supplies a net positive suction head for the reactor recir-culating pumps located 65 feet below the drum. The coolant flows from the steam drum to the recirculating pump suction via four 17-inch downcomers. There are no jet pumps in the BRP reactor pressure vessel and the downcomers are external to the vessel.
Overpressurization of the primary syctem is prevented by six spring-operated relief valves located on the steam drum.
The first relief valve opens at 1,535 psig and the remaining valves open in 10 psi increments.
The safety relief valve discharge is unpiped and relieves directly into the steam drum cavity.
1.2.2.2 Containment The BRP reactor containment building is si nificantly different from t
the current BWR 3, 4 and 5, but similar in design to Dresden 1.
The BRP containment is a spherical steel vessel 130 feet in diameter.
The sphere extends 27 feet below grade and 103 feet above grade The BRP sphere is designed for 41.7 psia internal pressure with the design basis loss of coolant event pressure rating of 37.7 psia.
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The sphere free volume is approximately one million cubic feet.
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BRP containment is not inerted and has a continuous flow of air.
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1he air enters and leaves the containment through two 24-inch supply and exhaust lines.
Both the supply and exhaust lines have two isolation valves that close automatically on all Reactor Protection System operations. These valves also close on high-radiation alarms from two area monitors located near the fuel handling areas. There I
are two conditions that cause the containment to isolates (1) containment high pressure, and (2) reactor low water. During power operation, the containment is habitable and routine inspections of equipment located in containment are made. Routine maintenance is also performed in containment during power operation. There are three means of ingress to and egress from the containment: the personnel, equipment and escape locks. These locks are double-door interlocked systems which allow only one door open at a time, t
1.2.2.3 Emergency Core Cooling System (ECCS) t The ECCS is designed to provide cooling through the ring or nozzle spray lines in the event of a Loss of Coolant Accident (LOCA). The ring and nozzle spray Ifnes discharge water on top of the core when the associated spray valves are opened. A coincident trip signal from reactor low pressure and reactor low water level actuation channels initiates the system.
The ECCS consists of diesel and electric fire pumps, a fire /ECCS rg Water Supply Distribution System, a nozzle spray line, and a ring
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spray line. Either one of the two fire pumps can provide the ECCS specified header pressure and flow rate.
Each spray line is designed to provide the required core spray flow under LOCA conditions to the reactor vessel.
1.2.2.4 Post-Incident System (PIS)
Since there is no suppression pool or wet well in the dry containment, a Post-Incident System is required for long-term cooling. This syFtem provides a Capability for recirculation and Cooling of the water accumulated in the lower portion of the containment after it rises above a certain level.
This core spray recirculation mode will automatically add emergency makeup water to the spent fuel pool.
The Post-Incident System also includes enclosure sprays which reduce the temperature / pressure of the containment sphere in the event of rupture in the reactor vessel or primary system.
1.2.2.5 Reactor Depressurization System (RDS) i To allow the use of the Low-Pressure Core Spray System during transients caused by small and intermediate LOCAs, the Reactor Depressurization System (RDS) was installed in 1976. The RDS is similar in concept to the Automatic Depressurization System used in
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modern plants. However, it varies significantly in design. The RDS k- /
consists of four trains of two valves in series, one isolation valve 1.2-3 MIO587-0194A-BX01
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k and one depressurization valve in each train.
Both the isolation s-valve and depressurization valve are normally closed. RDS blowdown is initiated by falling steam drum level and activated by coincident signals of low reactor water level and sufficient fire main pressure available after a two minute timer has timed out to open the valves.
The RDS discharges directly into the steam drum cavity as there is no suppression pool, t
1.2.2.6 Emergency Condenser Big Rock Point is equipped with an emergency condenser which is similar in concept and design to the isolation condenser used in the BWR 3 and 4 models.
The emergency condenser is automatically actuated at ~1,435 psig (operating pressure plus 100 psi) and is capable of removing ~10% of rated heat with both tube bundles.
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1.2.2.7 Feedwater and Main Steam Feedwater is supplied to the steam drum by two centrifugal pumps.
Each pump will deliver ~1,600 gpm of 260*F water and is driven by a 1,500-horsepower induction motor.
The Feedwater System and control rod drive pumps are the means of supplying high-pressure water to the primary system.
f-~g The Main Steam System supplies steam from the steam drum to the l
t turbine and condenser. Unique features at BRP include a single
'j bypass valve that is designed for 100% steam flow and a single main steam isolation valve located inside containment. The main steam isolation valve is backed up by non-automatic valves located outside containment.
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1.3 SCOPE CHARACTER. AND CONCLUSIONS OF THIS REPORT 1.3.1 SCOPE This report constitutes an update to the November 14, 1961 Final Hazards Summary Report as revised 3/12/62, 3/19/62, and 3/23/62, i
Revision 1, and is now titled " Updated Final Hazards Summary Report (FHSR)." The information presented is based upon the plant in its present as-built condition and reflects the current Safety Analysis l
Design Bases and operating requirements.
This report is a unique document which stands alone as the BRP Updated FHSR, which can serve as the baseline for future periodic changes.
The Updated FilSR is an integrated document containing or referencing the latest information developed in response to NRC Requirements.
The information presented is not based upon the descriptions or degree of detail required to meet Standard Review Plan content for modern plant Safety Analysis Reports (FSAR), however, this update is intended to be similar or comparable to an FSAR for those items, areas, or activities evaluated or referenced within this report.
The information contained herein is current as of F
.e of this Report.
1.3.2 CHARACTER This report consists of 18 chapters which present a complete description and safety evaluation of the nuclear plant or provides specific reference to direct the users of this Report to appropriate detailed information where necessary to fully explain the Design Bases, Operating Requirements, or other Analyses, Activities, Plans, Hanuals, Programs, or Reports pertinent to this FHSR Update.
1.
3.3 CONCLUSION
S As a result of the information presented or referenced in this report, Consumers Power Company considers that the Big Rock Point nuclear reactor and facilities have been designed and will continue to be operated with adequate protection against all potential accidents and of other possible hazards and shows reasonable assurance that the Big Rock Point nuclear power plant can be operated by Consumers Power Company without undue risk to the health and safety of the public.
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l 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS i
Bechtel Corporation was the prime contractor, and the General Electric Company furnished the reactor with its associated. nuclear-steam supply system, turbine-generator, and selected electrical apparatus.
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1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION Consumers Power Company - Big Rock Point Plant - utilizes unique methods for providing further technical information to the NRC, these are through semi-annual updates to the Integrated Assessment of Open Issues and by use of Probabilistic Risk Assessment (PRA) techniques for these issues.
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1.5.1 INTEGRATID ASSESSMENT OF ISSUES (Reference 1)
Consumers Power Company conducted a comprehensive integrated assessment of all open issues for Big Rock Point Plant. The assessment was considered comprehensive since it covered open safety issues (including those identified by the NRC such as SEP topics and TMI requirements and issues identified by the Plant Review Committee) as well as non-safety issues which may affect plant operability and reliability.
The integrated assessment provided the basis for determining which issues should be incorporated into the living schedule and consisted of the following tasks:
- 1) Issue identification which consists of preparing a detailed scope of each issue which identifies the issue initiator's concern, assesses the impact on the plant, the impact on public p
safety, and provides an estimate of the resources necessary to resolve the concern,
- 2) Issue Ranking which will be accomplished by a Technical Review Group (consisting of approximately six Consumers Power Company employees that are familiar with the issues as well as plant design, operation, risk and licensing).
1.5.1.1 Purpose (Reference 2)
In an effort to ensure safe and economical future operation of Big Rock Point Plant, finite company resources must be directed first towards those issues for which resolution offers the greatest return on the investment.
The purpose for the Integrated Assessment (IA) was to rank the issues relative to one another based on perceived magnitudes of reduction in risk or increase in plant availability attributable to their resolutions.
Plant-specific probabilistic risk assessment (PRA) cost / benefit data, in terms of dollars per man-Rem eliminated, was used for many of the issues to aid in the evaluation of return associated with their proposed resolutions. In general, the living schedule for issue resolution was constructed by assigning resources first to those resolutions featuring the greatest return. As a result, issues featuring resolutions which offer little return were scheduled to be performed at a later date. Many of the issue resolutions consist of performing evaluations to determine the necessity for implementing future plant modifications.
O The TRG will assess the results of such evaluations and may re-rank O
the issue accordingly.
1.5-1 HIO587-0194A-BX01
g 1.5.1.2 Schedule For Issue Resolution (Reference 2)
The input for Consumers Power Company's living schedule for issue resolution was the list of issues as ranked by the TRG. The resulting ranked list of issues has been subdivided into categories which represent the level of significance assigned to each issue by the TRG as a result of evaluating the issue against the assessment criteria (ie, safe shutdown, radionuclide release, plant availability or personnel safety).
1.5.1.3 On-Going Effort To Maintain The Living Schedule (Reference 2)
It is Consumers Power Company intention to maintain a living schedule for the resolution of non-repetitive type issues.
In doing so, Consumers Power Company plans to update the living schedule and to revise the schedule to maintain it as current as reasonably achievable.
Consumers Power Company will continue monitoring the progress of the living schedule and provide the NRC with semi-annual status updates.
It is our opinion that such an on-going effort will provide both Consumers Power Company and the NRC with a flexible and reliable means to help manage and control the issues that confront Big Rock
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CPCo has developed a comprehensive program to assess, coordinate, (Vl and schedule resolution of issues related to the BRP Plant including changes mandated or proposed by regulators or others.
Details of the management methods to be utilized are included in the " Plan for BRP Integr.ated Assessment" (the Plan) submitted to the NRC June 7, 1985 as revised September 3, 1985 and supplemented December 4, 1985.
The Plan is now a requirement included in License DPR-6.
1.5.2 PROBABILISTIC RISK ASSESSMENT (PRA)(REFERENCE 3) 1.5.2.1 Motivation For Performance of a Probabilistic Risk Assessment The Big Rock Point (BRP) Nuclear Power Plant is unique relative to the majority of US commercial nuclear plants from the standpoint of size, design, operating experience and site location. The small spentonplantmodik) cations.
size of BRP (240 MW limits the capital which can be economically i
Regulatory requirements imposed on nuclear plants on a generic basis after the accident at Three Mile Island made continued operation of BRP an unattractive alternative from an economic perspective. However, on the basis of the long, successful operating experience of BRP, the relatively remote site location, and the fact that the plant was only halfway through its projected lifetime, Consumers Power Company elected to find a better way of evaluating and enhancing plant safety.
It was judged that use of Probabilistic Risk Assessment (PRA) would allow consideration of the above factors while focusing attention on those features of h
the plant where modifications are necessary and effective in reducing
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different modifications addressing the same problem so that the most cost-effective modification can be pursued. This is crucial if the capital expenditures are to be kept within economic limits. The Probabilistic Risk Assessment of BRP is the necessary and responsible course of action for Consumers Power Company to take.
1.5.2.2 Objectives of the PRA There were two major objectives of the DRP PRA. The first was to quantify the risk to the public from operation of Big Rock Point.
i This would allow comparison of BRP with other US nuclear plants with regard to relative risk. The second objective was to define those design and procedural modifications to BRP which are most cost-effective from the standpoint of risk reduction. Completion of the second objective would allow Consumers Power to concentrate on making those modifications to the plant that will have the largest impact on increased safety.
1.5.2.3 Methodology To achieve the first of the above objectives, a complete PRA was undertaken for BRP. The approach used in conducting the PRA is detailed in Reference 3.
This approach is consistent with the recommendations of NUREG-0585, "TMI Lessons Learned Task Force Final Report" (dated October 1979), to perform a systematic reliability 7
evaluation of plant systems and to define a quantitative risk goal as a threshold for backfitting of new requirements to existing plants. Therefore, a comprehensive PRA for Big Rock Point utilizing probabilistic analysis methods similar to those employed in WASH-1400, Rasmussen Report (1975), has been completed.
i 1.5.3 SYSTEMATIC EVALUATION PROGRAM (REFERENCE 4)
The Systematic Evaluation Program was initiated in February 1977 by l
the U.S. Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when the supplement to the Final Integrated Plant Safety Assessment Report has been issued. Unlike previous SEP reviews, the review of Big Rock Point was expanded to address licensing requirements beyond those evolving from the original program.
The Big Rock Point Nuclear Power Plant has undergone a review under the NRC Integrated Plant Safety Assessment Systematic Evaluation _
Program. The Final Report by the U.S. NRC office of Nuclear Reactor Regulation was issued as NUREG-0828 dated May 1984, under Docket 50-155. NUREG-0828 documents the review of the Big Rock Point Plant, whi. is one of ten plants reviewed under Phase II of this-l l
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program. This report indicated how 137 topics selected for review I
under Phase I of the program were addressed.
It also addresses a majority of the pending licensing actions for Big Rock Point, which include TMI Action Plan requirements and implementation criteria for resolved generic issues. Equipment and procedural. changes have been identified as a result of the review.
These changes were scheduled for issue resolution via the Integrated Assessment Living Schedule l
as deffned an Reference 1 and summarized in 1.5.1 above.
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1.6 MATERIAL INCORPORATED BY REFERENCE
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The following provides a tabulation of topical reports, plans, programs, manuals, etc...which are incorporated by reference as part i
of License DPR-6, Docket 50-155 or this Updated THSR for the Big Rock Point Plant. These documents are updated and revised on schedules separate from this Updated FHSR report.
Docket 50-155 Big Rock Point Plant Facility Operating License DPR-6, Appendix A Technical Specifications i
Quality Assurance Program Description for Operational Nuclear
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Power Plants, CPC-2A
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Site Emergency Plan j
t Security Plan Fire Plan j
Integrated Assessment Plan t
Forty Year Master Inservice Inspection Plan Control Room Design Review Modified Program Plan, Volume 1 of the August 25, 1987 Final Control Room Design Review 1
Summary Report.
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DRAWINGS AND OTHER DETAILED INFORMATION I
Drawings such as Piping and Instrument Diagrams and Electrical, I
Instrumentation, and Control Drawings are referenced throughout this report.
These drawings are provided to the NRC as "Information Copies" as part of the distribution of BRP Volume 22 and as such are updated separate from this report and are not considered part of l
this report. Any unique drawings referenced in this report which are not included in BRP Volume 22.will be. included in this Updated i
FHSR and these drawings will be updated with revisions to this j
report.
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CHAPTER 1 REFERENCES
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CPCo Ictter dated March 18, 1983 - Big Rock Point Plant -
Integrated Assessment of All Open Issues...
2.
CPCo letter dated June 1,1983 - Big Rock Point Plant -
Integrated Assessment of Open Issues and Schedule for Issue Resolution...
3.
CPCo letter dated March 31, 1981 - Dig Rock Point Plant -
Submittal of the Probabilistic Risk Assessment...
4.
NUREG-0828, Final Report - May 1984 --Integrated Plant Safety Assessment - Systematic Evaluation Program - Big Rock Point Plant' O
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