ML20132C559

Forwards Draft SER Sections Which Require Rev as Result of Review of Applicant Responses to Open Items & Info in Amends 7 & 8 to Fsar.Salp Input Also Encl

ML20132C559
Person / Time
Site:	Beaver Valley
Issue date:	11/21/1984
From:	Houston R Office of Nuclear Reactor Regulation
To:	Novak T Office of Nuclear Reactor Regulation
Shared Package
ML19283C868	List: IA-85-926, Forwards Rept of 840726-27 in-progress Audit Re Detailed Control Room Design Review (Dcrdr).Dcrdr Proceeding Satisfactorily.Concern Expressed Re Differences Between Unit 1 & 2 Control Rooms & Role of Human Factors Specialist ML20132B767 ML20132B791 ML20132C010 ML20132C076 ML20132C105 ML20132C131 ML20132C209 ML20132C245 ML20132C247 ML20132C253 ML20132C285 ML20132C290 ML20132C295 ML20132C301 ML20132C310 ML20132C323 ML20132C334 ML20132C338 ML20132C344 ML20132C365 ML20132C379 ML20132C383 ML20132C389 ML20132C394 ML20132C421 ML20132C427 ML20132C433 ML20132C442 ML20132C453 ML20132C459 ML20132C464 ML20132C468 ML20132C474 ML20132C479 ML20132C484 ML20132C487 ML20132C493 ML20132C499 ML20132C512 ML20132C518 ML20132C536 ML20132C542 ML20132C559 ML20132C569 ML20132C575 ML20132C633 ML20133M691 ML20133P834
References
FOIA-84-926 NUDOCS 8412030159
Download: ML20132C559 (17)

Text

.

enr1nsHRE 4 NOV 211984 MEMORANDUM FOR:

Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:

R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration

SUBJECT:

REVISION OF ICSB INPUT TO SER - BEAVER VALLEY UNIT 2 Plant Name: Beaver Valley 2 Docket No.: 50-412 Licensing Status: OL Responsible Branch: LB #3 Project Manager: B. K. Singh Review Branch: ICSB Review Status:

• Incomplete In our memorandum dated October 22, 1984, we provided ICSB's input to the SER for Beaver Valley Unit 2. Per discussions with M. Ley, we are providing, as Enclosure 1, marked-up copies of SER sections which require revision as a result of our review of the applicant's recent responses to several open items and infonnation presented in Amendments 7 and 8 to the Beaver Valley Unit 2 FSAR. A SALP input is also provided as Enclosure 2.

InalSigned By N R

y . ayngpustos R. Wayne Houston, Assistant Director -

For Reactor Safety Division of Systems Integration

Enclosures:

Distribution:

As stated Docket File ICSB Rdg.

cc: R. Bernero F. Burrows (PF)(2)

D. Eisenhut T. Dunn.ing G. Knighton F. Rosa B. K. Singh ADRS Rdg.

Beaver Valley Subject File

Contact:

F. Burrows, ICSB X29452 FC :ICSB/DSI :ICSB/DSI ICSB/ I :ADRS/. 1  :  :  :

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and on those areas that have been of concern during reviews of other similar plants.

A meeting was held with the applicant and the NSSS and BOP designers to clarify the design and to discuss concerns the staff has with the design.

Detail drawings--including piping and instrumentation diagrams, logic diagrams, control wiring diagrams, electrical one-line diagrams, and electrical schematic diagrams--were audited during the review.

7.1.3 General Conclusion The applicant has identified the instrumentation and control systems important to safety and the acceptance criteria that are applicable to those systems as identified in the SRP. The applicant has also identified the guidelines--

including the regulatory guides and the industry codes and standards--that are applicable to the systems as identified in FSAR Table 7.1-1.

Based on the review of FSAR Section 7.1, the staff concludes that the implemen-tation of the identified acceptance criteria and guidelines satisfies the requirements of GDC 1, " Quality Standards and Records", with respect to the design fabrication, erection, and testing to quality standards commensurate with the importance of the safety functions to be performed. The staff finds that the NSSS and the BOP instrumentation and control systems important to safety, addressed in FSAR Section 7.1, satisfy the requirements of GDC 1 and, therefore, are acceptable.

, 7.1.4 Specific Findings 7.1. 4.1 Open Items The staff's conclusions apply to the instrumentation and control systems important to safety with the exception of the open items listed below. The staff will reveiw these items and report their resolution in a subsequent '

version of this report. The applicable sections of t51s report that address these items are indicated in parentheses following each open item.

10/10/84 7-2 -

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f. Service Water System Isolation on Low Header Pressure (7.3.3.4)

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J. Steam Generator Level Control and Protection (7.3.3.12)

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3 Bypass and Inoperable Status Panel (7.5.2.4)

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_.--.. ._ _:7: s. ._._7 7.1. 4. 2 Confirmatory Items E

In a number of cases, the applicant has c,ommitted to provide additional documen- .

tation to address concerns raised by the staff during its review. Based on information provided during meetings and discussions with the applicant, the technical issue has been resolved in an acceptable manner. However, the applicant must formally document his commitments for resolution of these items. The sections of this report that address these items are indicated in

, parentheses.

1.

Design Modification for Automatic Reactor Trip Using Shunt Coil Trip Attachment (7.2.2.3) 2.

Service Water System Isolation on Low Header Pressure (7.3.3.4) 10/18/84 7-3

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3. Automatic Opening of Service water System valves M0v 113C and 1130 (7.3.3.10) 4 IE Bulletin 80-06 Concerns (7.3.3.13)

[ Remote Shutdown Capabilty (7.4.2.1) d NUREG-0737 Item II.F.1 Accident Monitoring Instrumentation Positions (4),

(5), and (6) (7.5.2.2) 7 Bypass and Inoperable Status Panel (7.5.2.4)

83. Cold Leg Accumulator Motor-Operated Valve Position Indication (7.6.2.4)

T3 NUREG-0737 Item II.K.3.9, Proportional Integral Derivative (PID) Controller Modification (7.7.2.1) ,

les GugeeqkhcalSpecificationItemsSv3Mgnau 7.1. 4. 3 Tec Outo gy Nlntree.

Items to be included in the plant Technical Specifications and information to be audited as part of the effort to issue Technical Specifications are discussed in the following sections:

1.

Lead, Lag, and Rate Time Constant Setpoints Used in Safety System Channels

. (7.2.2.1)

2. Turbine Trip Following A Reactor Trip (7.2.2.2)

3. Trip Setpoint and Margins (7.2.2.4) 4.

NUREC-0737 Item II.K.3.10, Proposed Anticipatory Trip Modification (7.2.2.5) l S.

Undetectable Failure in Online Testing Circuitry for Engineered Safeguards Relays (7.3.3.3) 10/10/84 7-4 l BEAVER VALLEY 2 SER SEC 7 INPUT

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that automatic shunt trip actuation would not provide substantial, additional protection if incorporated into the plant design. The staff found this response unacceptable. In a September 7, 1984 response, the applicant committed to provide the Westinghouse Owners Group generic design modification. The staff 9 ren finds this acceptable, but considersa." canmar,4r zrcm

,_ '_ until the modifi-cation installation is completed. -

i In addition, the staff has identified additional information required on a plant specific basis as part of the acceptance c4 the generic modification.

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The staff considers the submittal of the require'd information and an FSAR revision covering the modification to be a confirmatory item.

i 7.2.2.4 Trip Setpoint and Margins I The setpoints for the various functions in the reactor trip system are deter-mined on the basis of the accident analysis requirements.. As such, during any anticipated operational occurrence or accident, the reactor trip maintains.

system parameters with the following limits:

. g (1) minimum departure from nucleate boiling ratio of 1.30. 'l (2) maximum system pressure of 2750 psi (absolute).

(3) fuel rod maximum linear power of 18.0 kW per foot.

l The staff requested detailed information on the methodology used to establish the technical specification trip setpoints and allowable values for the Reactor l

Protection System (including Reactor Trip and Engineered Safety Feature channels) assumed to operate in the FSAR accident and transient analyses. This includes the following information:

1 (1) The trip setpoint and allowable value for the Technical Specificatifons.

(2) The safety limits necessary to protect the integrity of the physical barriers which guard against uncontrolled release of radioactivity.

10/10/84~ 7-13 BEAVER VALLEY 2 SER SEC 7 INPUT

I C O Y--

(1) The 4/4 logic, although redundant in each RPS train, has four input channels developed from position switch contacts on the four turbine stop I

valves. The installation of the stop valve position contacts and their cable routing to the RPS input cabinets do not preclude a single failure from preventing either train from performing its safety function.

(2) The sensors and stop valve contacts are not qualified to operate in a seismic event.

In response to the staff's first concern, the applicant stated, in a. February 2'1, 1984 letter, that the reactor trip on turbine low auto stop oil pressure provides a diverse b.ackup for the trip on stop valve closure. The applicant also reiterated that this trip is anticipatory, is included for the protection of the turbine equipment, and no credit is taken for this trip in any FSAR  !

i Chapter 15 accident analysis.

The staff finds the applicant's response accept-able and considers this concern resolved.

In response to the staff's second concern, the applicant stated in an August 9, 1984 letter, that the pressure sensors and stop valve contacts fail in a safe direction (provide the trip) if they fail due to a seismic event. The staff finds the applicant's response acceptable and considers this issue closed.

7.2.3 Conclusions We have conducted an audit review of the Reactor Trip (RTS) for conformance to guidelines of the applicable regulatory guides and industry codes and standards as outlined in the Standard Review Plan, Section 7.2, Part II and III. In Section 7.1 of this SER, we concluded that th'e applicant had adequately identi-fied the guidelines applicable to these systems. Based upon our audit review of the design for conformance to the guidelines, we find that ,_ ___:..' _ ., .

there is reasonable, assurance that the systems will conform to the applicable guidelines.

Our review has included the identification of those systems and components for the RTS which are designed to survive the effects of earthquakes, other natural phenomena, abnormal environments, and missiles. Based upon our review, we 10/10/84 7-15 BEAVER VALLEY 2 SER SEC 7 INPUT

G*Jr Based on our review of the interfaces between the RTS and plant operating control systems, we conclude that the system satisfies the requirements of IEEE-279 with regards to control and protection system interaction. Therefore, we find that the RTS satisfies the requirements of GDC-24, " Separation of Protection and Control Systems."

Based on our review of the Reactor Trip System, we conclude that the system satisfies the protection system requirements for malfunctions of the reactivity control system, such as accidental withdrawal of control rods. Section 15 of the SAR addresses the capability of the system to assure that fuel design limits are not exceeded for such events. Therefore, we find that the RTS satisfies the requirements of GDC-25, " Protection System Requirements for Reactivity Malfunction."

Our conclusions, noted above, are based upon the requirements of IEEE-279 with respect to the design of the RTS. Therefore, we find that the RTS satisfies the requirement of 50.55a(h) with regards to IEEE-279.

Our review of the RTS has examined the dependence of this system on the avail-ability of essential auxiliary support (EAS) systems. Based on our review, '

we conclude that the design of the RTS is compatible with the functional performance requirements of EAS systems. Therefore, we find 'the interfaces between the RTS design and the design of the EAS systems to be acceptable.

3 In summary, the staff concludes that the design of the Reactor Trip System (RTS) and the design of the essential auxiliary support (EAS) systems are acceptable and meet the relevant requirements of General Design Criteria 2, 4, l 20, 21, 22, 23, 24, and 25, and 10 CFR Part 50, 50.55a(h),. Z :__ _ _ _ J r-a + u n.. + - z w+ < < z a -

l 7.3 Engineered Safety Features Systems '

7.3.1 Engineered Safety Features Actuation System (ESFAS) l The ESFAS is a portion of the plant protection system that monitors selected plant parameters and, on detection of out-of-limit conditions of these parameters, 10/10/84 7-17 BEAVER VALLEY 2 SER SEC 7 INPUT

9 feedwater isolation provided by this signal is not assumed in Chapter 15 of the FSAR and is not necessary for safety and is therefore not required to be redundant. The staff has reviewed the applicant's response and finds that since there is no current basis to apply additional regulatory requirements, the design is acceptable and considers this issue closed.

dM Additionally, FSAR Figures 7.2-1 (Sheet 13) and 7.3-18 de not agree with the information (discussed above) provided by the applicant. T M r..ff . '"- ;

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7.3.3.8 Control Room Isolation The applicant had indicated that the design of the control roon. and pressurization system was incomplete during a December 1983 meeting. Based on our review of preliminary information, the staff expressed a concern that the design, which is integrated into the current control room isolation and pressurization system, may not meet the requirements of GDC-5, " Sharing of Structures, Systems) and Components."

The staff requested detailed schematic drawings be provided for this system when the design was finalized. In a September 7, 1984 letter, the applicant provided information on the design and the interrelationship' between Unit 1 and Unit 2. The staff h reviewed this information and required additional information covering testability of the system. %....,....g '"- % 4 suSSE@Enir p Kcus snest mamme THE AtPucur awant- oken ries em o adre:renicaro ryc. srY$rfet eetuD BC 'frSTED M N"'#6" ss Errrers 7k srnip cusiacas mos issac ewsts.

7.3.3.9 Control Room Isolation on High Radiation Signal During the staff's review of the control room isolation system, a conflict was found between the plant schematics and the information p ovided by FSAR Figures 7.2-1 (Sheet 8) and 7.3-13. These figures show that the control room was wAs

.M isolated by a high radiation signal which pr, according to the applicant, in error. E : . 5 ' ;;,,f;, ,,,g,, g _ ,

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. c' system and initiates operation of these systems. The ESF control system regulates the operation of the ESF system following automatic initiation by the protection system or manual initiation by the plant operator.

We have conducted an audit review of these systems for conformance to guidelines of the applicable Regulatory Guides and industry codes and standards as outlined in the Standard Review Plan, Section 7.3, Parts II and III.

In Section 7.1 of this SER we concluded that the applicant had adequately identified the guidelines applicable to these systems. Based upon our audit review of the system design for conformance to the guidelines, we find that upon satisfactory resolution of the open items identified in Sections 7.3.3.4M_ and 7.3.3.12 there is reasonable assurance that the systems conform to the applicable guidelines.

Our review has included the identification of those systems and components for the ESFAS and ESF control systems which are designed to survive the effects of earthquakes, other natural phenomena, abnormal environments and missiles.

Based upon our review, we conclude that the applicant has identified those systems and components consistent with the design bases for the systems.

Sections 3.10 and 3.11 of this SER addressed the qualification programs to -

demonstrate the capability of these systems and components to survive appli-cable events. Therefore, we find that the identification of the systems and components satisfies this aspect of the GDC-2, " Design Bases for Protection Against Phenomena," and GDC-4, " Environmental and Missile Design Bases."

e Based on our review, we conclude that that ESFAS conforms to the design bases  ;

requirements of IEEE-279. The system includes the provisions to sense accident

) conditions and anticipated operational occurrences to initiate the operation i of ESF and EAS systems consistent with the analyses presented in Chapter 15 of the SAR. Therefore, we find that the ESFAS satisfies the requirements of GDC-20, " Protection System Functions."

The ESFAS adequately conforms to the guidance for periodic testing in Regulatory Guide (RG) 1.22 and IEEE-338 as supplemented by Regulatory Guide 1.118. The bypassed and inoperable status indication adequately conforms 10/10/84 7-38 BEAVER VALLEY 2 SER SEC 7 INPUT

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requirements of EAS systems. Therefore, we find the interfaces between the ESFAS and ESF control systems and the EAS systems to be acceptable.

Our review of the ESF control systems included conformance to the requirements for testability, operability with onsite and offsite electrical power, and single failures consistent with the General Design Criteria applicable to these ESF systems. We conclude that the ESF control systems are *astable and are operable on either onsite or offsite power (assuming only or.e source is available) and that the controls associated with redundant ESF systems are independent and satisfy the requirements of the single failure criterion.

Therefore, we find the ESF control systems meet the relevant requirements of GDC-34, " Residual Heat Removal," and GDC-35, " Emergency Core Cooling," GDC-38,

" Containment Heat Removal," and GDC-41, " Containment Atmosphere Cleanup."

In summary, the staff concludes that the ESFAS and the ESF control systems will be acceptable and meet the relevant requirements of General Design Criteria 2, 4, 20 thru 24, 34, 35, 38, and 41 and 10 CFR Part 50.55a(h) subjecttoresolutionoftheopenitemsidentifiedinSections7.3.3.4g M and 7.3.3.12 of this report.

7.4 Systems Required for Safe Shutdown 7.4.1 System Description

, This section describes the equipment and associated controls and instrumentation of systems required for safe shutdown. It also describes controls and instru-j mentation outside the main control room that enable safe shutdown of the plant in cas.e 'the main control room needs to be evacuated.

7.4.1.1 Safe Shutdown Systems l

Securing and maintaining the plant in safe shutdown condition can be achieved I by appropriate alignment of selected systems that normally serve a variety of operational functions. The functions which the systems required for safe shutdown must provide are:

10/18/84. 7-40 BEAVER VALLEY 2 SER SEC 7 INPUT

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Review Plan (SRP) Section 7.4 interprets the GDC-19 requirements. The design should provide redundant safety grade capability to achieve and maintain safe  !

shutdown from a location or locations remote from the control room, assuming  !

no fire damage to any required systems and equipment and assuming no accident has occurred. The remote shutdown equipment should be capable of maintaining functional operability under all service conditions postulated to occur ,

including the seismic event. The remote shutdown station and the equipment used to maintain safe shutdown should be designed to accommodate a single f failure.

In the FSAR Section 7.4.1.3, the applicant states that the design basis for i control room evacuat. ion does not consider a single failure. The staff found '

the applicant's design basis for remote shutdown capability unacceptable and required that the applicant clarify the design criteria for remote shutdown and ,

address the isolation, separation, qualifiction and transfer / override pro- l visions of the remote shutdown equipment in Section 7.4 of the FSAR.

i In a June 13, 1984 response, the applicant stated that the next FSAR amendment  ;

will indicate that the design criteria for the control room evacuaton includes the single failure criterion and coincident loss of offsite power. Additionally ~{

the applicant stated separation of redundant train-related and non-1E circuits l is maintained by barriers or appropriate air space, all Class 1E control i equipment (other than indicators) meet the requirements of IEEE-STD-344-1975 t and IEEE-STD-323-1974, and that transfer to the ESF is accomplished by push-

buttons and switches on the shutdown panel. I fouso The staff temsreviewed the applicant's response and m uda-it acceptable with the exception of the seismic qualification of indicators of the ESP. The staff ten requested additional information on this issue [E Y-WY.Y 37 _ , yy 3

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? Additionally, t e staff considers the applicant's pending FSAR amendment to include the information provided in the June 13, 1984 response to be a confir-matory item.

10/10/84 7-45 BEAVER VALLEY 2 SER SEC 7 INPUT

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• during shutdown including a shutdown following an accident. Equipment at tppropriate locations outside the control room has been provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitabe procedures. Therefore, we conclude that the systems required for safe shutdown satisfy the requirements of GDC-19,

" Control Room."

Our review of the instrumentation and controls required for safe shutdown has examined the dependence of these systems on the availability of essential auxiliary support (EAS) systems. Based on our review and coordination with those having primary review responsibility for the 'EAS systems, we conclude that the design of EAS systems are compatible with the functional performance requirements of the systems reviewed in this section. Therefore, we find the interfaces between the design of safe shutdown systems and the design of EAS

, systems to be acceptable.

Our review of the instrumentation and control systems required for safe shutdown included conformance to the requirements for testability, operability with onsite and offsite electrical power, and single failures consistent with the General Design Criteria applicable to safe shutdown systems. We conclude that these systems are testable, and are operable on either onsite or offsite electrical power, and that the controls associated with redundant safe shut-

, down systems are independent and satisfy the requirements of the single failure criterion. Therefore, we find that these systems meet the relevant require-ments of GDC-34, " Residual Heat Removal," GDC-35, " Emergency Core Cooling," and GDC-38, " Containment Heat Removal."

In summary, the staff concludes that the systems required for safe shutdown are acceptable and meet the relevant requirements of General Design Criteria 2, 4, 13, 19, 34, 35, and 38,: dj:;t t: ::"-'r"; m :it? :. :'

10/10/84 7-47 BEAVER VALLEY 2 SER SEC 7 INPUT

PORV. If the staff tioes not accept the Westinghouse conclusions (under II.K.3.2 review), we will address this item in a supplement to this report.

7.6.2.2 Reactor Coolant System Loop Isolation Valve Interlocks The FSAR 7.6.6 describes the reactor coolant system loop isolation valve interlocks. The description was incomplete and additional information was required to clarify that the design is in conformance with IEEE STD-279 require-ments. Additionally, the staff was concerned that, during operation with N-1 loops, the criteria for testing and single failure may not be met due to" reduced protection logic.

In a July 12, 1984 letter, the appl.icant responded to this issue. The staff has reviewed the applicant's response and will pursue this issue as part of plant Technical Specifications review.

7.6.2.3 Primary Component Cooling Water Isolation from Reactor Coolant Pump i

, Thermal Barriers The FSAR Section 9.2.2 describes the isolation of the reactor coolant pump thermal barrierr from the primary component cooling water system. A check 1 valve is installed in each inlet cooling water line to the thermal barrier  !

c,ooling coil and an air-operated isolation valve is installed in each outline line. Each isolation valve closes on signals developed from a corresponding line's pressure or flow sensor. Because the FSAR sloss dM

, not provide the design basis for this isolation, the staff 3r concerned about its safety significance. i Therefore, the staff requestfthe applicant provide information about the design basis for this system and a discussion on the consequences of either I the check valve or the air-operated isolation valve failing to close under l

conditions related to the design basis. F-O_ x Mn i

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.During the staff's review of the power lockout circuitry, a conflict was found 4 between plant schematics and the information provided by FSAR Section 6.3.5.5.

The F,SAR states that the valve position indicating lights are powered by the

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7 Based on our review of the interlock systems important to safety, we conclude that their design bases are consistent with the plant safety analysis and the systems' importance to safety. Further, we conclude that the aspects of the design of these systems with respect to single failures, redundancy, indepen-dence, qualification, and testability arc adequate to assure that the functional performance requirements will be met.

Our review has included the identification of the systems and components of interlock systems important to safety which are designed to survive the effects l

of earthquakes, other natural phenomena, abnormal environments, and missiles. I Based upon our review, we conclude that the applicant has identified the systems and components consistent with the design bases for the interlock systems. Sections 3.10 and 3.11 of this SER address the qualification programs to demonstrate the capability of these systems and components to survive applicable events. Therefore, we find that the identification of the systems and components satisfies this aspect of the GDC-2, " Design Bases For. Protection Against Natural Phenomena," and GDC-4, " Environmental and Missile Design Bases."

In summaiy, the staff concludes that the interlock systems important to safety are acceptable,: 2j_Z ^ ^ '

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7.7 Control Systems O-The general design objectives of the Plant Control System are:

(1) To establish and maintain power equilibrium of the primary and secondary system during steady state unit operation;

, (2) To constrain operational transients so as to preclude unit trip and re-establish steady-state unit operation; and (3) To provide the reactor operator with monitoring instrumentation that indicates all required input and output control parameters of the systems and provides the capability of assuming manual control of the syst'em.

10/10/84 7-61 BEAVER VALLEY 2 SER SEC 7 INPUT

t 7.7.2.3 control System Failure caused by Malfunctions of Common Power Source or Instrument Line To provide assurance that the FSAR Chapter 15 analyses adequately bounds Gvents initiated by a single credible failure or malfunction, the staff has asked the applicant to identify any power source or sensors that provide power or signals to two or more control functions, and demonstrate that failures or malfunctions of these power sources or sensors will not result in consequences core severe than those of Chapter 15 analyses or beyond the capability of cperator or safety systems. .

The staff Juus reviewed (e wo the applicant's response, contained in Amendment-4 to ,

the FSAR, and taums,it nee further clarification in the following areas:

(1) The applicant's response S. based on the satisfactory review of this  !

issue on other Westinghouse plants. A statement ,%sasas needed to address the },

similarity of BV-2 to the other referenced plants as pertaining to this issue. '

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^(2) Where similarity does not exist, further analysis shouldet provided to properly address this issue.

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The control systems used for normal operation that are not relied upon to p2rform safety functions but which control plant processes having a signifi-cant impact on plant safety, have been reviewed. These control systems include the reactivity control systems and the control systems for the primary and secondary coolant systems. The staff concludes that the control systems are acceptable and meet the relevant requirements of General Design Criteria 13,

" Instrumentation and Control," and GDC-19, " Control Room." This conclusion is based on the following':

10/10/84- 7-69 BEAVER VALLEY 2 SER SEC 7 INPUT

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Based on our review of the plant transient response to normal load changes and anticipated operational occurrences, such as reactor trip, turbine trip, upsets in the feedwater and steam bypass systems, we conclude that the control systems are capable of maintaining system variables within prescribed operating limits. Therefore, we find that the control' systems satisfy this aspect of GDC-13, " Instrumentation and Control."

Our review of control systems included the features of these systems for both manual and automatic control of the process systems. We conclude that the c features for manual and automatic control facilitate the capability to maintain plant variables within prescribed operating limits. We find that the control systems permit actions which can be taken to operate the plant safely during normal operation, including anticipated operational occurrences; therefore, the control systems satisfy GDC-19, " Control Room," with regards to normal plant operations.

The conclusions of the analysis of anticipated operational occurrences and accidents as presented in Chapter 15 of the FSAR have been used to confirm that plant safety is not dependent upon the response of the control systems.

We conclude that failure of the systems of themselves or as a consequence of supporting systems failures, such as power sources, do not result in plant conditions more severe than those bounded by the analysis of

. anticipated operational occurrences.

e Finally, we have confirmed that the consequential effects of anticipated operational occurrences and accidents do not result in control system failures that would cause plant conditions more severe than those bounded by the analysis of these events. We find that the control systems are not relied upon to assure plant safety and are, therefore, acceptable.

In summary the staff concludes that the control systems are acceptable,suggese.

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10/10/84 7-70 BEAVER VALLEY 2 SER SEC 7 INPUT

ENCLOSURE 2 .

ICSB SALP INPUT

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Beaver Valley 2 - - '

PLANT:

SUBJECT:

Safety Evaluation Report EVALUATION PERFORilANCE BASIS CRITERIA CATEGORY

1. Management N/A No basis for assessment.

Inv31vement l An understanding of the issues was frequently lacking. Resolutions are/were

2. a o 3 delayed due to the lack of understanding of the technical issues involved.

Technical Issues Draft SER contained 23 open items. SER now contains 3 open items.

3. Responsiveness 3 Considerable NRC effort and repeated submittals were needed and are still needed to obtain acceptable resolutions.

4. Enfcrcement N/A No basis for assessment.

History

5. Reportable Events N/A No basis for assessment.

P

6. Staffing N/A No basis for assessment. ,

7. Training N/A No basis for assessment.

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