ML20198G276
| ML20198G276 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 01/28/1975 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1036 WNP-1036, NUDOCS 8605290391 | |
| Download: ML20198G276 (16) | |
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,J V. A. Moore. Assistant Director for LWRs. Group 2. ONRR WAS111ETON PUBLIC POWER SUPPLY SYSTEM. UNITS 1 & 4 INSTRUMENTATION.
CONTRCL AND ELECTRICAL POWER SYSTEMS SAFETY EVALUATION REPORT Plant Name: NPPSS 1 & 4 Licensing Stage: Construction Permit Docket. Webers: 50-460/613 Responsible Branch ey.d Project Manager: LWR 2-3. T. Cox Requested Completion Date: January 23. 1975 Applicant's Resp 6nse Date: NA Description of Response: NA Technical Revfew Branch involved: EILCS Branch Review Status: Safety Eva1 nation Report Ccaplete The er. closed SWety evaluaticn tcport was prepared by the L:RS.
Elect.rical. Instrumentation and Control Systems. Branch. This evaluation covars our review of the PSAR through Amendment 13.
In my letter to ycu of N6venber 5.1974, transmitting the second set of questions and posittant, we pointed our the need for a third set of questions and pcsitions. The reason for a third set of questions and positions was that the applicant dhanged the plant site and added another unit after the first set of guestleas ard positions on the original PSAR was issued.
At tha request of Project,, va have issued the SER without the
'nenefit of anofber set of questions end positio s t ca su e of n
,e schedule related considerations for 'Tlue Book" starred plants.
Therefore, this SER contains a larse n'anber of outstanding items that will have to be resolved and reported in the Supplement to the SER. Because of the large nuster of outstanding items, additional 'eview tirae for the Supplement to the SER will be needed than that originally allocated. Therefore, appropriate schedule provisions snould be made.
The outstandirg items are clearl/ identified in the SER as well as our reconmended positions regarding them, where applicable.
.j. x It should be pointed out that the Rscctor Prota: tion System (RPS) i for WPPSS 1 & 4 represents a significant departare from that of 3
previously approved plants such as North Anna 3 & 4. a reference I
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plant for WPPSS e1 & 4.
The concept of using minicomputers in the i
Reactor Protection System has not been evaluated by the staff.
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1s presently under review in conjunction with the review of B&W Topical Report BAW-10085. Pending evaluation of this report, the j
RPS for WPPSS 1 & 4 remains an unreviewed and outstanding item.
The following is a sunnary of the outstanding items.' the areas I
where additional inforestion and documentation is required, and potential prob 1cm areas.
I A.
Outstanding Items i
1.
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RPS-II is presently under generic review under i '
TAR-696. The applicant has made the cournitment that in the event that RPS-II is not approved by the time of. the FSAR submittal, the RPS design will be such that it will achieve at least the l
same degree of safety that is provided by the previously reviewed and accepted RP54I system. (Section 7.2) 2.
Transfer from the Injection to the Recirculation M5de of ECCS Operation
.The. applicant agreed to provide fore the automatic opening of the sump suction valves 'on low water level signal in the BWST.
In order to establish that for WPPSS 1 & 4 this is an acceptable degree of automation of the transfer from the injection to tne recirculation m:de'.of ECCS operation we need the information described in Section 7.3.4.
S 3.
_ Decay Heat Removal (DHR) System l
The DHR System does not meet the single failure criterion with respect to protecting against overpressurization.
Appropriate design changes should be made to satisfy the single failure criterion. (Section 7.4.1)
- 4. - Auxiliary Feedwater System l
The Auxiliary Power and Conversion Systems Branch has requested additional infonnation to evaluate the significance of required operator action five (5) minutes following a DBA. Acceptance of the Auxiliary Feedwater
' System is pending review of this additional information.
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Cable Separation Criteria The reconmendations of Regulatory Guide 1.75 are not complied with fully. We found the applicant's exemptions unjustified.
(Section 7.7) 6.
Environmental Qualification Compliance to the reconnendations o'f IEEE Std 323-1974 and Regulatory Guide 1.89 remain an open item. The applicant stated that a future B&W topical report
._. lf (BAW-10082) will be the basis for safety rela.ted i
equipment and cabling qualification. (Section 7.8) 7.
Integrated Control System (ICS).
We are not in a position to accept the classification i
of the ICS as a control system not required for j
safety. Question 223.39 transmitted to Projects by our letter dated Noveder 5.1974 which was intended I
to 9anerate additional information for a comprehensive review of the ICS was not transmitted to the applicant.
We will pursue this issue on a generic basis with 2
B&W. (Sections 7.9".l. 7.3.2) 8.
Nuclear Ir.strumentation (NI) System t
Acceptability of the NI system is contingent on the results of the RPS-II and Incore Monitoring System i
presently under review. (Section 7.9.2) 9.
Offsite Power System We require additional infonnation to evaluate the offsite i
power system's conformance to GDC-17.
(Section8.2) l I
B.
Requirements for Additional Infonnation and Documentation i
1.
There are numerous inconsistencies in the PSAR in reference to IEEE Std 323. The applicant agreed to review the PSAR and correct all references to this standard to read IEEE Std 323-1974.
2.
Table 8.1-1 lists a total 8.341 kW of emergency bus loads.
It is stated in'the PSAR however, that 6500 kW diesel-generator sets will be procured. The applicant will provide additional infonnation to reconcile this i
cifre rence.
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The applicant's responses to Questions 7.7 and 7.52 are in direct conflict, however, the response to 7.52 is intended to be operative and we accepted it.
Appropriate corrections and cross-referencing will be provided in a future Amendment to the PSAR.
I 4.
We prefer that the results of a failure mode and effects analysis be documented as part of the PSAR and not the FSAR as the applicant proposes. (Q7.8) i 5.
I The applicant was asked orally to analyze, and document i
the results of such analysis, the safety implications of the operator resetting the ESF actuation signal following a LOCA without initial loss of offsite power.
but with subsequent ' loss of offsite power and after resetting the ESF actuation signal.. This requirement for additional infomation and documentation should be transmitted formally to the applicant 1 l
6.
The applicant agreed to explicitly state in Chapter 7 l
or on p. 8.3-12 that safety related instrumentation
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will be. designed for protection against lightning.
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7.
The applicant stated orally that with respect to thermal i
overload protection of motors of motor-operated valves, j
the bypass option will be incorporated in the ESF design instead of that requiring periodic testing. This i
should be documented in the PSAR.
i 8.
Periodic Testing of certain components and functions only during shutdown periods, as proposed by the applicant, l
is not fully justified. The applicant will review these components and functions for the purpose of minimizing their number.
(Section 7.3.3. Q7.33) i 9.
The applicant should document that the Essential Control l
and Instrumentation (ECI)' System will cogly to the requirements of IEEE Std 279-1971.
(l(7.5.2) i i
10.
The applicant has made the comitment to amend Chapter 7 by classifying tha pressurizer leater and spray control systems under systems not required for safety.
i 11.
The applicant agreed to review Table 7.4-2 and report if any non-Category I or non-Class IE equipment are -
listed as required to achieve and maintain a safe diitritma rt l
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- 12. A number of less significant items discussed during two i
telephone conferences of January 14 and 22,1975 will i
have to be documented or clarified.
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Potential Problem Areas 1.
The environmental qualification according to IEEE Std j
323-1974 is expected to be a topic of continuing i
discussion.
2.
The diesel-generator set qualification testing might present some first-of-a-kind problems because of the large size of the proposed sets. This size of d-g i
sets has not been qualified before for nuclear power plant service.
3.
Acceptance of RPS-II remains uncertain, both in terms i
of the review of BAW-10085 as well as its integration into the WPPSS 1 & 4 design.
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Or:f n il S'aned by T. A. I;;Mila
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Victor Stello, Jr., Assistant Director for Reactor Safety l
Office of Nuclear Reactor Regulation i
Enclosure:
Safety Evaluation Report i
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W. Mcdonald
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S. Hanauer i
F. Schroeder A. Giambusso' S. Varga A. Schwencer j
T. Cox i
T. Ippolito i
i F. Rosa l
j D. Basdekas I
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GT0ti PUBLIC POWER SUPPLY SYSTE" UilITS 1 & 4 DOCKET I;05. 50-450/513 SAFETY EVALUATI0ft REPORT 7.0 IfiSTRUMEliTATI0tl AliD C0 iTROLS 7.1 General The Commission's General Design Criteria (GDC)
IEEE Standards including IEEE Criteria for Protection Systems for Nuclear Power Generating Stations (IEEE Std 279-1971),'and applicable Regulatory Guides for Power Reactors have bean utilized as the bases for evaluating the adequacy of the protection and control systems.
Specific docunents employed in the review are listed in the Appendix to this report.
The review of WPPSS 1 & 4 protection and control systems was accomplished by comparing the designs with those of the North Anna 3 & 4 and Bellefonte Plants.
Our revie's concentrated on those areas of design which are unique to the WPPSS 1 & 4 plants, for which new information has been received, or which have remained as continuing areas of concern during this and prior reviews of similarly designed plants.
During the course of our evaluation, we had one meeting with the applicant, Washington Public Power Supply System and its contractors.
7.2 Reactor Protection System (RPS)
The WPPSS 1 & 4 RPS design, designated as RPS-II is a new concept utilizing a minicomputer or " calculating module" in each RPS
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channel for calculating a power trip envelope, It is the same concept proposed in the Bellefonte application but it represents a basic departure from the North Anna 3 & 4 design, designated as RPS-I. RPS-II is presently under review by the staff in conjunction with the review of 3&W topical report BAW-10085.
It represents an entirely new approach to RPS design by virtue of introducing digital computer modules in each of the four RPS channels and remains an unreviewed and outstanding item for WPPSS 1 & 4 pending the evaluation of topical report BAU-10085.
The applicant has made the commitment that in the event that we do not approve RPS-II by the time of the FSAR submittal, an LPS design will be provided such that it will achieve at least the same degree af safety that is provided by the previously reviewed and accepted A:5-I design.
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7.2.1 Hich Containnent pressure Reactor Trio The U?PSS 1 & 4 design provides for reactor trip an low pressurizer level as well as a diverse reactor trip on low reactor coolant pressure. The ECCS, on the other nand, is actucted by low reactor coolant pressure and high containment pressure signals. The staff's position has been that since a reactor trip is assumed in the analysis of design basis accidents performance of the ECCS, the same signals actuating the ECCS should trip the reactor. The applicant maintains that low pressurizer level in lieu of high containment pressure will provide an equivalent degree of assurance that the reactor will trip prior to, or coincident with, the time assumed in the accident analysis to be performed at the FSAR stage of review. The applicant has requested to be allowed to retain the option to modify the reactor protection system design to include a high containment pressure signal to trip the reactor, if it cannot be demonstrated at the FSAR stage of ' the review that pressurizer water level signal will perform adequately to protect the reactor core from exceeding design limits, and the reactor coolant system and containment from overpressurization.
We considered this request and we agreed to evaluate the aoplicant's analysis at the FSAR stage of review.
7.3 Engineered Safety Features Systems (ESFS)
The Engineered Safety Features Actuation System (ESFAS) is functionally comparable to that of Morth Anna 3 & 4 and Bellefonte plants. Our review encompassed all aspects of the ESFAS that actuate and control the operation of the ESFS and their vital auxiliary supporting systems. The following sections identify those aspects of the design that were not acceptable to us and were changed as a result of our review; and those items of concern that have been identified during this and previous reviews of similar plants.
7.3.1 Containment Soray System Manual initiation of the containment spray system is accomplished by two manually operated switches per train. One of the switches actuates the spray header valves and the other switch actuates the spray pumps. This arrangement, as opposed to a single switch or manual system level initiation minimizes inadvertent actuation of the containment spray system. We find this design feature acceptable, even though it is not the ultimate in system level initiation. We believe that the intent of Section 4.17 of IEEE Std 279-1971 is satisfied and that the health and safety of the public is protected.
We agree with the applicant's contention that no single operator error should cause the containment spray actuation.
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. 7.3.2 Safety Fen:tions Involvino the Integrated Control System (ICS) and the Control Rod Drive Control System (CROCS)
The WPPSS 1 & 4 design of the ICS and CRDCS provides for both of these systems to perform safety related functions for which, the applicant maintains, there is no credit taken in the analysis of the respective accidents and anticipated operational occurrences identified in Chapter 15.0.
For those cases where it is has been that the worst case analysis for a given accident or anticipated operational occurrence is the case with the associated control system not functioning, the results of such an analysis are acceptable.
In certain cases, however, where the analysis is based on unrealistic assumptions, or unacceptable computer codes., and the staff's determination is that safety limits of the core or contain-ment are likely to be exceeded, then the safety actions of the ICS and CRDCS do require that these systems, or their appropriate parts, meet the requirements of systems required for safety. For additional discussion of ICS see Section 7.9.
For the cases of an inadvertent single control rod withdrawal, insertion, or misalignment by a malfunction or operator error, the present B&W design relies on the CRDCS and ICS to sense the occurrence and initiate corrective actions. The applicant's response to our concern with regard to the adequacy of core protection provided to mitigate the consequences of such an occurrence has been inadequate for substantiating the claim that the design meets the requirements of GDC - 25. We will require that the WPPSS 1 & 4 plants be designed and operated so that safety limits such as DNBR of 1.32 and fuel centerline melting limits not be violated for any single rod withdrawal, or, in case the above limits are shown by analyses to be violated, appropriate Reactor Protection System design changes and/or additions be incorporated to terminate the rod withdrawal before these limits are violated. Furthermore, it is the staff's position that since the operator has the capability to individually withdraw any rods, the single rod withdrawal incident be treated and analyzed as an anticipated occurrence with, as well as without, reactor scram.
A 7.3.3 Periodic Testing of Reactor Protection and Engineered Safety Feature Systems The applicant's design criteria provide for testability of individual channels, logic and final actuating devices satisfy the requirements of IEEE Std 279-1971 and Regulatory Guide 1.22, except Regulatory l
Position D-4.
The applicant has identified as non-testable during plant operation, as provided by Regulatory Position D-4, a number of functions and components that we consider can be tested during plant operation without jeopardizing plant operation. These include main steam isolation valves and main feedwater isolation valves.
We asked the applicant to review each of these cases for the purpose of reducing their number to a minimum.
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7.3.4 Transfer from the Injection to the Recirculation Mode of ECCS 0;eration In tne design or'iginally proposed, changeover from injection to the recirculation mode and the cross-cver mode (using LPI pumps as boosters for the HPI punps) of operation following a loss of coolant accident was accomplished by the operator in accordance with established procedures which included a series of manual actions. The complexity of the proposed changeover p'rocedures to be followed during the worst possible operating conditions (LOCA) did not appear to provide adequate assurance that the operator would correctly perform the required actions.
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We asked the applicant to provide design changes so that the transfer from the injection to the recirculation mode of ECCS operation is performed automatically, at least in part, with the capability for manual transfer at the systen level. The applicant has complied with this position and provided for the automatic opening of the pump suction valves. We consider this an acceptable way of increasing the amount of time available to the operator to complete the transfer from the injection to the recirculation mode of ECCS operation including the cross-over mode wnereby the high head ECCS pumps are aligned to take suction from the low head ECCS pumps. We will require that the applicant submit an analysis of the entire sequence of operations showing the time available to the operator for their completion both individually and collectively. We will also require that three safety grade BWST water level instrumentation channels supplying a two-out-of-three logic should be provided with all channels having readouts in the control room. We expect the design to be similar to the one we have accepted for the Bellefonte plant and we will report on the final resolution of this item in the supplement to this SER.
7.4 Systems Required for Safe Shutdown We have reviewed the instrumentation.. control and electrical systems provided for safe shutdown as well as the design provisions to place and keep the plant in a safe shutdown condition in the event that access to the main control recm is restricted or lost. We have concluded that the applicant's preliminary design and design criteria are acceptable. The following sections identify those aspects of the design that were not acceptable to us and were changed as a result of our review, or remain unresolved.
7.4.1 Decay Heat Removal (DHR) System The DHR design provides two serially connected motor-operated valves on the suction line of each DHR oump.
Both serially connected valves in each set are powered frem one emergency bus with each set of valves powered from a different bus. The design, as presently proposed, satisfies the single failure criterion, from the standpoint of assuring decay heat removal, hcwever, it does not satisfy the single failure c<iterion with respect to protecting against over-
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. pressurization of the system (failure of one power source will prevent automatic closure on increasing pressure of both valves in one line.) Diverse interlocks, however, are provided which prevent the valves from being opened unless RC pressure is below the DHR system design.
We will require that the applicant modify the electrical design to meet the s:ngle failure criterion with respect to both the decay heat removal function and the prevention of overpressurization function. Also, we require that the instrumentation, control and electrical equipment pertaining to this system be designed to conform to IEEE Std 279-1971 and IEEE Std 308-1971. The applicant has been advised of our position. We will report the resolution of this positico in the supplemental safety evaluation report.
7.4.2 Auxiliary Feedwater and Main Steam Line Break Instrumentation and Control Systems Our review of the auxiliary feedwater control system and the main steamline break instrumentation and control system revealed that operator action may be needed within five (5) minutes of the accident if the system were to perform its function while sustaining a break inside containment and a single active (electrical) failure anywhere in the system. The applicant was asked to submit additional information so that we can evaluate the potential problem in this respect.
Furthermore, the control ~ power to the motor-driven pumps and the steam turbine driven pump is provided from d-c power sources thus not meeting the diversity requirement in this respect. We will require that diverse power sources be provided for the control functions associated with the motor-driven pumps as opposed to the steam turbine driven pump. We also determined that when the Auxiliary Feedn-ter System is to be used to achieve a safe shutdown through the Essential Control and Instrumentation (ECI) system,this system must conform to Section 4.12 (operating bypasses) of IEEE Std 279-1971. We will require that the applicar,t modify the design to be in full conformance to Section 4.12 of IEEE Std 279-1971.
We will report on resolution of all concerns in this area in the supplement to the SER.
7.5 Safety-Related Disolay Instrumentation He have reviewed the design of the instrumentation systems that provide information (1) to enable the operator to perform required manual safety functions and (2) those required for post-accident monitoring. For each plant parameter required for these functions, two instrumentation channels, with at least one recorded will be provided. Also, they will be pcwered from the emergency power buses and designed to meet the requirements of IEEE Std 279-1971.
Further-more, the applicant agreed to provide recorders capable to withstand a design basis earthquake. We find these commitments acceptable.
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6-7.6 All Other Systems Required for Safety We have reviewed the design criteria for all other systems required' for safety and find that they conform to the applicable safety standards and Regulatory Guides and are therefore acceptable.
7.7 Cable Se03 ration and Identification Criteria for Protection and Energency Po.er Systems We reviewed the applicant's separation criteria for preserving the independence of redundant parts of reactor protection systems, engineered safety feature systems, Class IE electric systems, and other systems required for safety. The applicant has saught exemptions from some of the separation provisions of Regulatory Guide 1.75. We found some of these exemptions unjustified as presently stated in the PSAR. The applicant stated orally that these exemptions will be amended in the near future. We will review these amended exemptions and report in the supplement to the SER.
We reviewed the applicant's program for identification of protective system equipment and cabling and found it acceptable.
7.8 Seismic, Radiation and Environmental Qualifications The applicant stated that seismic qualification of all safety-related components will conform to the requirements of IEEE Std 344-1971 and the supplementary Regulatory Position on Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations. We conclude that these criteria are acceptable.
For all safety-related equipment and cabling inside containment within the scope of B&W supply, which must function subsequent to a DBA in a post-accident environment, the applicant is proposing to perform the qualification testing in accordance with Topical Report BAW-10082 to be submitted for our review in the near future.
This qualification comnitment has not been documented to cover equipment outside the scope of B&W supply. We will require that this be done.
The applicant maintains that Resistance Temperature Detector (RTD)
Instrumentation channels located inside a containment need not be qualified because theyf are not required for protective actions related to a LOCA. However, a hostile environment similar to that produced by a LOCA or steam line break may develop and preidl inside containment for sometime before a RTD generated reactor trip needs to be initiated. Therefore, we will require that safety-related RTD instrumentation channels be qualified to withstand the most severe hostile envircrment postulated for the instruments.
The applicant listed the worst-case hostile enviror, ment inside 6.44 x 10}nment to inc!ude 271*F, 42.2 psig,1005 humidity end the conta Rad integral radiation dose. These numbers represent
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a reduced severity environment from those submitted originally, nana:y, 256:F, 60 psig, 1003 hu'idity and 103 r.*d integral dose.
The new numbers are under review by the staff and have r.ot been accepted yet. We will report on this matter in the supplement to the SER.
We will require that safety-related equipment be qualified in accordance with the folloaing, (1) IEEE Std 317-1971, Electrical Penetration Assemblies, in Containment Structures for f;uclear Fueled Power Generatir.g Stations, (2) IEEE Std 323-1974 Qualifying Class IE Electrical Equipment for Nuclear Power Generating Stations, (3)
IEEE Std 382-1972, Trial-Use Standard: Guide for Type Tests of Class IE Electric Operated Valve Actuators for !;uclear Power Generating Stations, (4) IEEE Std 344-1971 Type Test of Continuous Duty Class IE _ Motors Installed Inside the Containment of Nuclear Power Generating Stations, (5) Regulatory Guide 1.40, Qualification Tests of Continuous Cuty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants, (6) Regulatory Guide 1.63, Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Po<:er Plants and, (7) Regulatory Guide 1.89, Qualification of Class IE Equipment for fluclear Poner Plants.
The applicant's qualification test progran acceptability is subject to final review and acceptance of the referenced topical report and related additional information and qualificaticn test da'ta to be submitted in the FSAR.
7.9 Control Systems Not Required for Safety.
7.9.1 Integrated Centrol System (ICS)
We consider the classification of the ICS as a control system not required for safety an unresolved item. However, we have decided that it would be more appropriate and effective to pursue its review on a generic basis directly with B&W. The functions performed by the ICS during anticipated transients as well as accidents may be safety-related under certain conditions of the plant, and mal-functions and failures of the ICS. This item will remain unresolved until we receive and evaluate additional information from'B&W.
(See Section 7.3.2).
7.9.2 Nuclear Instrumentation (flI) System The NI system is designed to provide neutron-flux information over the full range of reactor operation utilizing three ranges of nuclear detectors; source range, intermediate range, and power range. The power range nuclear instrumentation is an integral part of the reactor protection system (RPS). Calibration of the NI power range instrumentation channels and setpoint calculations are made using information derived from the incore monitoring system. We are presently reviewing the RPS and the use of incore monitoring system infermation on a generic basis. Acceptance of the NI system is contingent on the results of these reviews.
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8.0 ELECTRIC FCMER 8.1 General The Ccmmission's GDC 17 and 18, IEEE Standards including IEEE Criteria for Class IE Electric Systems for nuclear Pcwer Generating Stations (IEEE Std 303-1971), and Regulatory Guides (RG) for Power Reactors including RG 1.6 and 1.9, served as the bases for evaluating the adequacy of the electrical power system. Specific documents in the review are listed in the Appendix to this report.
8.2 Offsite Power System l
The WPPSS 1 & 4 units will be connected to the Bonneville Power Administration (BPA) 500 kV transmission system at the H. J. Ashe switchyard, located about one mile Northwest of the plant, by means of one transmission line through a " breaker-and-a-half" bus arrangement. They will also be connected to the 230 kV BPA transmission system at the H. J. Ashe switchyard by means of one transmission line through a " radial" bus arrangement, connected to the AEC Hanford 230 kV loop, which provides immediate access to the transmission system as the backup offsite power supply.
Each main generator feeds through a 25 kV isolated phase bus to the generator step-up and three station auxiliary transformers (SAT). Three back-up auxiliary transformers (BAT) are connected to the 230 kV transmission lines. Two SATs and two BATS are connected to the a-c distribution system at the 4160 V level and one SAT and one BAT are connected to the a-c distribution system at the 13.8 kV level.
When generation of power at the station is interrupted because of a fault in the generator, the 25 kV system and connected transformers, or the 500 kV p1werhouse line to the switchyard, the generator load break switch (LSS) will not be opened, thus resulting in the loss of the 500 kV source of offsite power. On loss of the 500 kV source the BATS will be automatically connected to the Class IE emergency buses by way of the non-Class IE 4160 volt buses by means of a dead-bus transfer.
In the course of our review, it was determined that the 500 kV lines emanating from the H. J. Ashe switchyard to the BPA distribution system cross tne 230 kV lines a numter of times, thus jecpardizing the independence between the 500 kV and 230 kV sources of offsite power to the station. The applicant agreed (orally) to analyze on a case by case basis the consequences of a 500 kV line failing an adjacent 230 kV or 500 kV lines.
If these analyses show that such an event will result in the loss of the 500 kV and the 230 kV sources of offsite power to the plant, we will require design changes in the offsite power systems so that the requirements of GDC 17 are met. We.will report on the acceptability of the offsite power system in the supplement to the SER.
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-g-8.3 Onsite Power Sy_ stems 8.3.1 A-C Power System The safety loads for each unit are assigned to two independent and
-fully redundant 4 kV Class IE emergency buses, either capable of supplying minimum engineered safety features and shutdown loads.
Each of these two buses is capable of receiving power from a SAT which is supplied by the main generator, a SAT which is supplied by offsite power, or a diesel-genarator.
Each 4 kV emergency bus supplies the safety loads of a 480 volt bus and other '.'CC's and lower voltage buses.
There are four separate 120 v a-c vital buses supplied by four dual inverters. Each inverter will be connected to a different station battery, and its output will be regulated. Power for each inverter will be taken from a point between the battery and its battery charger. Reactor protective channels will be powered from redundant vital bus distribution panels.
Two redundant diesel-generator sets of identical design and characteristics will be provided per unit to serve as the standby power supply. Each diesel generator set will be capable of a continuous output of 6800 kW which is expected to be sufficient, with some margin, to supply all required safety loads.
Each diesel-generator will be automatically started on loss of emergency bus voltage, or an ESF actuation signal. Under accident conditions with a loss of emergency bus voltage, the normal loads are disconnected and then the engineered safety feature loads are automatically sequenced as required to each diesel-generator.
The applicant has committed the diesel-generator system design to meet the recommendations of Regulatory Guide 1.9.
If the required type and size of diesel-generator set has not been previously qualified for nuclear power plant service, an acceptable qualifica-tion program will be provided and implemented prior to fuel loading.
Subject to the satisfactory completion of the qualification program, if needed, we conclude that the a-c emergency onsite power system is acceptable.
8.3.2 0-C Power System Onsite d-c emergency power is derived from the station battery /
charger system. Two redundant and separate 125 volt d-c systems are provided for each unit and are sized to carry required loads for two hours following loss of all a-c power. Each redundant system consists of two batteries, each with its own charger and d-c bus. A total of eight individual seismic &lly qualified batteries will be provided for the tuo units. Each battery will be mounted on Category I racks and 'ocated in a separate room in a Catc;ory I builcing. The ventilation for each room will be separate from that of other rooms.
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Separate and independent regulated battery chargers are provided for each Class IE station battery. Battery chargers are fed from the Class IE 4SO volt emergency tus of the same power channel as the bus and battery to which it is connected.
Each battery charger has the capacity to continucusly carry the steady state connected auxiliary loads required fcr plant full load operation, or the steady state post-DBA loads without assistance from the batteries. This capacity includes the capability to restore the battery from the design ninicua charge state to the fully charged state within eight hours regardless of the status of the plant during which this demand occurs. We find the onsite d-c emergency power system acceptable However, this is predicated on the satisfactory completion of the environmental qualification program as reported in Section 7.8
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.,a APPENDIN The following principal documents were used by D. L. Basdekas in the construction permit review of WPPSS 1 & 4 1.
Preliminary Safety Analysis Report (PSAR) through Amendment 13 for WPPSS 1 & 4.
2.
Sections 6,7 and 8 of the PSAR for North Anna 3 & 4.
3.
10 CFR Part 50 and Appendix A to 10 CFR Part 50.
4.
Regulatory Guides 1.6, 1.9, 1.11, 1.22, 1.30, 1.32, 1.40, 1.41, 1.47, 1.53, 1.62, 1.63,, 1.67, 1.75 and 1.89.
Institute of Electrical and Electronic Engineers (IEEE) 5.
Standards:
" Criteria for Protection Systems for Nuclear Power Generating Stations."
" criteria for Class IE Electrical Systems for Nuclear Power Generating Stations."
" Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations."
IEEE Std 323-1974 - "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."
" Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equip = ants During the Construction of Nuclear Pcwer Generating Stations."
" Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systa=s."
" Trial-Use Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations."
6.
Safety Evaluation Reports - Surry 3 & 4, Bellefonte 1 & 2.
.