ML20003C275

Forwards Amend 1 to CESSAR-F.Application to Review Amend & Responses to NRC 790323 Questions Encl

ML20003C275
Person / Time
Site:	05000470
Issue date:	02/24/1981
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20003C276	List: ML20003C275 ML20003C277
References
LD-81-011, LD-81-11, NUDOCS 8102270514
Download: ML20003C275 (66)

Text

C-E Power Systems Tel 203/6881911 Comeusbon Engineenng <nc Te te x 99297 1000 Prcspect Hill Acac Windscr. Connect: cut C6095 POWER i SYSTEMS February 24, 1981 LD-81-011 Mr. James R. Miller, Branch Chief Standardization and Special Projects Branch Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

CESSAR-F Amenument No.1

Reference:

NRC letter from C. J. Heltemes to A. E. Scherer, dated March 23, 1979.

Dear Mr. Miller:

Combustion Engineering, Inc. hereby submits for your review seventy (70) copies of Amendment No.1 (non-proprietary) to CESSAR-F. Also e. closed are responses to each of the questions provided by the above reference.

Where changes to CESSAR-F were needed, the responses reference the section within Amendment No. I which provide the requested information.

If we can be of any additional assistance, please feel free to contact either myself or Mr. T. J. Price of my staff at (203)688-1911, Ext. 2803.

Very truly yours, COMBUSTION ENGINEERING, INC.

A.EMcherer

. Director Nuclear Licensing AES:dac 1.;:i n

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of: )

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Combustion Engineering, Inc. ) DOCKET NO. STN 50-470F

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Standard Plant )

APPLICATION FOR REVIEW 0F

" COMBUSTION ENGIPEERING STANDARD SAFETY ANALYSIS REPORT,"

AMENDMENT NO. 1 J. M. West, being duly sworn, states that he is Vice President, Nuclear Power Systems, Combustion Engineering, Inc. ; that he is authorized on the part of said corporation to sign and file with the Nuclear Regulatory Commission this Amendment; and that all statements made and matters set furth therein are true and correct to the best of his knowledge, information and belief.

COMBUSTION ENGINEERING, INC.

Subscribed and swern to before me d

By UN this 74 day of Fel- - 19 c / J. M. Vest bem -

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Vice President Nuclear Power Systems l CAREY[WEY.-- ZEL, NOTARY PUBLIC State of Coor.ecticut No. 59962 Cecmission Expires March 31,1985

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000.3 Amend Section 1.3 of the CESSAR rSAR by identifying all deviations from (1.3) the revision of NUREG-75/087, " Standard Review Plan for the neview of

Safety Analysis Reports for Nuclear Power Plants" which was in effect six months prior to the tendering of the CESSAR FSAR.

Resconse i... SRP's were " prepared for the guidance of the staff reviewers in the Office of Nuclear Reactor Regulation in performing safety reviews...".

They were intended to be internal NRC management dccuments, provided to the industry and public so that we could have knowledge of the staff's ctview mechanism and some assurance that uniform review' criteria would be emoloyed. To then take the SRP's and establish them as a standard against which the applicant must evaluate himself is a fundamental and significant change in policy.

In fact, this issue is the subject of a Notice of Proposed Rulemaki g, 45 FR 67099, October 9,1980. Combustion Engineering has objected to the plan in a letter to the Secretary of the Ccmmission, dated November 24, 1980.

Combustion Engineering will therefore await the Commissioners decision on the propcsed rulemaking.

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000.4 The information contained in Appendix A to the CESSAR FSAR was (Appendix A) submitted for PDA extension purposes; therefore, said informa-tion is at the PSAR level of detail. The staf f has found the information contained in Appendix A sufficiently complete for PDA extension purposes; however, the information is not suffi-ciently detailed for FDA purposes. C-E's responses to the Category II, III and IV matters that have emerged subsequent to the regulatory requirements cut-off date for the CESSAR PSAR must, therefore, be addressed at a level of detail commensurate with an Operating License application.

Response

C-E prepared Appendix A with the intention that it would be used primarily for the FDA review. Our responses were prepared as -

position statements on each of the NRC guidelines and were not intended to describe the NSSS design features. The NSSS description is already provided in the text of CESSAR-F, and together with the Appendix A position statements, provides a level of detail consis-tent with other FSAR's concerning conformance to NPC guidelines.

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010.1 Provide an interface which includes a list of Combustion Engineering (3.0) supplied equipment located outside of containment that must be pro-tected from wind and tornado loadings (Section 3.3), floods (Section 3.4), and from internally and externally generated missiles (Section

, 3.5).

Response

Sections 3.3, 3.4, and 3.5 of CESSAR-F have been revised in Amendment No. I to provide the requested information.

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010.2 Provido an interface discussion for the 30P designer of the possible (3.5) missile sources from NSSS equipment located outside containment, and

include a table of missiles and their characteristics. l

Response

i C-E is presently preparing a response to this question. The response will be submitted by Second Quarter 1981.

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010.3 provide the heat load interface requirements the balance of plant (9.1.3) designer needs to design the spent fuel pool cooling system as follows:

1. Decay heat load from a refueling batch (1/3 core) within 150

, hours after reactor shutdown.

2. Decay heat load from an emergency unload of the reactor (1 core)

., 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown plus the decay heat lead of 1/3 core placed in the pool 30 days prior to the emergency core 'inload.

l The 1/3 core was placed in the pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after rea: tor shutdown. Confirm that this heat load is the maximum heat load attainable which must be considered by the 80P designer in the design of the spent fuel pool cooling system. -

In calculating the infonmation requested under items 1 and 2 above, use the method set forth in Branch Te-5nical Fosition 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", tttached to Section 9.2.5 of the Standard Review Plan.

Resoonse This interface is not within CESSAR scope. The applicant calculates the heat loads based on the total core heat output provided in Tabia 4.4-1 of CESSAR-F.

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-010.4 Your response to the Category I item in Appendix A to the CESSAR FSAR, i regarding conformance with Regulatory Guide 1.29 (Revision 2), reads, in part, as follows:

...the RCP bearing oil system, including component cooling water equipment associated with oil coolers need not be seismic Category I."...

In contract, Section 5.4.1.3 of the CESSAR FSAR contains the follcwing statement:

" Lubricating oil is cooled by cooling coils submerged in the oil sumns.

Both sumps and cooling coils are internal to the motor structural frame and are designed for seismic Category I operation, and use the intent of the ASME Boiler and Pressure Vessel Code,Section III, Class 3 as a guide for design and construction."

In light of the above, amend either Section 5.4.1.3 of the CESSAR FSAR or Appendix A to the CESSAR FSAR, as appropriate, so as to reconcile this apparent discrepancy.

Response

The CESSAR-F Appendix A position statement on Regulatory Guide 1.29 was modified .n the docketing version of CESSAR-F to be consistent with Section 5.4.1.3.

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010.5 Provide the necessary information that shows how the CESSAR design meets (9.5.1) applicable portions of our Branch Technical Position ASB 9.5-1, attached to Standard Review Plan Section 9.5.1. This information should include the following: ,

1. A list of your equipment that recuires the use of flamable material such as lube oil or resins, and the amount of material contained by the equipment or is necessary to be stored or supplied for the equipment.

2. An evaluation of the potential fire hazards covering all equipment within the CESSAR design, e.g. , pumps with lube oil supply, electri-cal cabinets, electrical cables. Identify those safety related control room cabinets that contain cables from redundant safety divisions.

Provide an analysis of the consequences of a fire in each of these cabinets, with respect to the controls and wiring of one division of the cabinet performing the system's safety function following a fire.

Also describe the separation between divisions and the need for in-stalled smoke detectors in the cabinets. Provide any interface require-ments necessary as a result of your analysis.

3. Rating of flame spread, smoke ar.J fuel contribution for materials used throughout the CESSAR System 80 design.

4. A discussion of the CESSAR System 80 cable design showing how the cables meet the applicable guidelines of position B.3 of Branch Technical Position ASB 9.5-1; and

5. The interface requirements imposed on the dehign of the fire detection and suppression systems and on the separation of CESSAR System 80 systems and components in order for the o.erall fire protection sys-tems design to conform with the guidelines of Branch Technical Position ASB 9.5-1.

Response

The content of BTP 9.5-1 is also addressed in Regulatory Guide 1.120. C-E position statements on both of these documents are provided in Appendix A to CESSAR-F, since they were issued after the CESSAR-P regulatory guide cutoff date. With the clarifications pro'vided in those position statements, BTP 9.5-1 is not within the licensing scope of CESSAR-F. C-E provides to the Applicant all necessary infomation on flammabla materials within CESSAR-F scope, but the fire hazard analysis is in the Appl l cant's scope.

4 010.5 Interface Requirements for fire protection of CESSAR-F safety systems are i

(9.5.1)-

Cont. listed under "Related Services" for each of those systems. Separation of

( safety related circuits is discussed in Section 7.1.2.10 of CESSAR-F, and t -

i interfaces on separation are provided in Section 7.1, 7.2 and 7.3.

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03].03 Our review of the design criteria presented in the FSAR Tendering Draft

. (7.1) indicated that the Instrumentation a..J Control Systams Branch Technical Positions were not part of the design criteria. Amend the FSAR to demonstrate how the design conforms to the Instrumentation and Control Systems Branch Technical Positions defined in Table 7-1 of NUREG-75/112 and in NUREG-75/087, (Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," LWR Edition, September, 1975).

Resoonse J

Conformance with licensing guidelines will be discussed in the upcoming Independent Design Review (IDR) for Instrumentation and Control Systems.

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030.04 In a review of the design criteria, we detected several deficiencies.

(7.1) In comparing the design criteria stated in the FSAR Tendering Draft to the design criteria of Table 7.1 of NUREG-75/112, we noted that the design is to conform to IEEE 344-1971 instead of IEEE 344-1975.

We require that the designs presented in Chapter 7 of the FSAR Tendering Draft conform to the criteria contained in IEEE 344-1975.

Amend the FSAR to demonstrate that the designs conform to IEEE 344-1975.

Resconse IEEE 344-1975 was issued after the PDA cutoff date established for CESSAR-P and thus was not listed as criterion for the PDA and, according to the Commission's Policy Statement of August 31, 1978, should not be .

listed as criterion for the FDA.

In addition to IEEE 344-1971, additional requirements were imposed at the. tine of CESSAR-P review; C-E has addressed these requirements in the seismic qualification of instrumentation and control equipment.

As stated in Appendix A of CESSAR-F (in the. Category I position state-ment for Regulatory Guide 1.100), we believe that conformance to IEEE 344-1971, plus the additional requirements listed in Section 3.10 of CESSAR-F, is consistent with the guidance of IEEE 344-1975.

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1 030.05 In Section 7.1.1.7, Design Comparison, it is stated that the Reactor (7.1) Protection System (RPS), of which the Core Protection Calculators (CPCS) are a part, will be " functionally identical (with specified exceptions) to the system provided for the Arkansas Nuclear One -

Unit 2 (ANO-2) plant. Please define " functionally identical".

Also, we were recently informed by our office of Inspection and Enforcement that Systems Engineering Laboratories (SEL) is no longer a supplier to Combustion Engineering. The Universal Multiple Acquisition Control System, a SEL product, is a major component of the Core Protection Calculator System licensed for ANO-2. If the change has been made in the design, it has not been -defined in the FSAR. Amend the FSAR to provide the final design of the Reactor Trip System, including the hardware design

, and software design of the Core Protection Calculator System.

Describe in detail how this design differs from the RPS design approved for ANO-2. Also identify any R&D effort ongoing in this area.

Response

The ANO-2 CPCS and the System 80 CPCS are functionally identical based on Section 7.2.1.1.2.5 of CESSAR-F.

1. Both systems use the same inputs:

a. excore nuclear instrumentation flux power,

b. pressurizer pressure,

c. reactor coolant pump speeds,

d. CEA position, and

e. hot and cold leg temperatures

2. Both systems calculate the following intermediate variables in either the CPC or the CEA Calculator:

a. CEA deviations;

b. Correction factor for excore flux power for shape annealing and CEA shadowing;

c. Reactor coolant flowrate from reactor coolant pump speeds and temperatures;

d. aT power from reactor coolant temperatures, pressure, and ficw information; I

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030.05 e. Excore flux power: excore flux power signals are summed and (7.1) corrected for CEA shadowing, shape annealing, and cold leg Cont. temperature shadowing. This corrected flux power is period-ically calibrated to the actual core power measured indepen-dently of the Reactor Protection System. Tnis calibration does not modify the inherent fast time response of the ex-core signals t:; power transients:

f. Axial power distribution from the corrected excore flux power signals;

g. Fuel rod and coolant channel planar radial peaking factors, selection of predetermined coefficients based on CEA positions.

h. DNBR;

i. Comparison of DNBR with a fixed trip setpoint;

j. Local power dens 'ty;

k. Comparison of . local power density with a fixed trip setpoint; and

3. They have the same outputs as given in Section 7.2.1.1.2.5:

a. JNBR trip and pretrip;

b. DNBR margin (to control board indication);

c. Local power density trip and pretrip;

d. Local power density margin (to control board indication);

e. Calibrated neutron flux power (to control board indication); and

f. Control Elemenc Assembly Withdrawl Prohibit (CWP);

i j , g. CEA deviation alann.

Since the ANO-2 CPCS and the System 80 CPCS have the same inputs, inter-l l mediate calculations and the same ou? uts, the systems are said to be l'

functionally identical.

Appropriately qualified hardware will be provided which will replace the l

Universal Multiple Acquisition Control System that was supplied by Systems l-l Engineering Laboratories for ANO-2.

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030.05 The final design of the Reactor Trip System, including the CPCS, is (7.1)

Cont. provided in Section 7.2 of CESSAR-F and will be discussed in the up-coming Independent Design Review (IDR) for the Instrumentation and Control Systems.

All R&D necessary for the completion of the System 80 RPS design ha3 been completed.

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030.07 In Section 7.7, the Re ctnr Power Cutback System (RPCS) is defined and (7.7) discussed. It is stated that the objective of the system is to achieve

" step" reductions in reactor power to accomodate certain types of power imbalances. This is achieved by simultaneous dropping of one or more preselected groups of full length regulatory CEAs into the core.

The Reactor Power Cutback System is defined as a control system not required for safety. For this type of system, the Standard Format (Regulatory Guide 1.70) requires the following analysis:

" Provide analyses to demonstrate that these systems are not required for safety. The analyses should demonstrate that the protection systems are capable of coping with all (in-ciuding gross) failure modes of the controls systems."

Amend the FSAR to provide the above stated analysis required by the Standard Format.

Resconse The Reactor Power Cutback System is designed to rapidly reduce the reactor power by 1) dropping preselected Regulating CEA group (s) and 2) running back the turbine in response to either a large load rejection or the loss of a feedwater pump from initially high power levels. The RPCS is designed to control NSSS variables to within operational limits to improve plant availability only. The system is not required for safe shutdown nor given any credit for mitigation of the consequences of events in the safety analyses.

The Reactor Protective System (RPS) analysis of Section 7.2.2 encompasses the failure modes of. control systems and demonstrates that these systems are not required for safety. The' safety analyses of Chapter 15.0 do not require these control systems to remain functional.

The Core Protection Calculator System provides CEA deviation protection for CEA related cesign basis events. These events and subsequent acceptable

030.07 results are specified in Chapter 15 of CESSAR-F. Initiating events (7.7) -

caused by postulated failures of the RPCS (i.e., decreased Main Steam Flow and Uncontrolled Negative Reactivity Insertion events) were con-sidered in the generation of Chapter 15. The consequences of these initiating events are less severe than those identified in Sections 15.2,1 and 15.4.1. In addition, failures in the RPCS considered in combination with other initiating events are listed in Table 15.0.6.

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030.09 Modify the FSAR to provide a readable copy of Figure 7.2.13, Plant (7.2) Protection System Interface Logic Diagram.

Resconse A revised figure is provided in the docketed version of CESSAR-F.

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040.1 Your interface requirements presented in Section 8.1.1 of the FSAR is (8.1) incomplete. We require the following Standards, Regulatory Guides and Branch Technical Positions be included in your interface requirement criteria.

1. IEEE Std. 334

2. IEEE Std. 382

3. IEEE Std. 383

4. Reg. Guide 1.40

5. Reg: Guide 1.70

6. Reg. Guide 1.73

7. Reg. Guide 1.89 S. Reg. Guide 1.93

9. PSB Branch Technical Pos1Lisa Nos. 2, 6, 8, 11, 15, 17, 18 and 21.

Resocnse Regulatory Guide 1.93 was issued after the PDA cutoff date established for CESSAR-P and thus was not listed as a criterion for the PDA and according to the Commission's Policy Statement of August 31, 1978, should not be listed as a criterion for the FDA.

Sections 8.1 and 8.3 of CESSAR-F are revised in Amendment No. I to include the following:

IEEE Std. 334-1971 IEEE Std. 382-1972 IEEE Std. 383-1974 Reg. Guide 1.40 Reg. Guide 1.73 Reg. Guide 1.89 We feel it inappropriate to impose Br.ach Technical Positions as inter-face requirements for the following. reasons:

1. BTP's have not been subjected to rigorous review by .the NRC or the public.

2. No value/imoact assessment has been made on these BTP's to determine their relative safety gain versus cost.

3. These BTP's have not yet been reviewed by the Regulatory Requirements Review Committee (or its ecuivalent) to determine their significance to safety, a

040.1 We also feel that it is inappropriate to impose the SAR Guide, R.G.1.70, (8.1)

Cont. as an interface requirement. The NSSS design described in CESSAR-F is not impacted by the format of the referencing applicant's SAR.

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040.2 Section 3.3 of the FSAR should contain an explicit staterant as to (8.3) whether CESSAR design has any direct dc loads. If so, specifications covering ripple and other distortions should be added to the inter-face criteria requirements for the de sources.

Resoonse As requested, the DC sources' interface requirements are added to Section 3.3.2 of CESSAR-F, DC POWER SYSTEM, in Arancment No.1. DC loads are added to Table 8.3.1-1 in Amendment No. 1.

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040.3 Section 3.11 does not address the potential for submergence of electrical (3.11) equipment within the containment as a result of an accident. State whether your equipment qualification program will include submergence.

If not included, we require that an interface be provided to ensure this equipment will be able to perform the safety functions as required.

Response

Submergence testing is not included in the equipment qualification program.

Environmental interface requirements are listed within the CESSAR-F sections for each of the safety systems. They state that safety equip-ment listed in Section 3.11 must be located in an environment that is consister.: with the environments listed in Section 3.11.

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j 040.4 With regard to essential safety feature pump motors and their supoorting i (8.3) auxiliary system pump motors which are in CESSAR scope, provide the

. following information:

i a) Specify the minimum voltage required at the motor to successfully accelerate its pump load within the required period. Correlate i these requirements with the recomendations contained in Regulatory i Guide 1.9 for sequencing safety system loads on an offsite power system.

b) State the minimum margin of motor torque allowed over the pump load

! torque during the accelerating period of the pump load, and provide criteria for selecting this minimum value.

s j c) Describe the features provided to monitor the temperature rises in i large horsepower motor components to evaluate any ingress into their

! design temperature rise limitations, when a motor fails to accelerate its pump load within the number of starts crescribed by NEMA MG-1.

Response

a) The requested information is provided in Subsection 6.3.2.2.2 (LpSI pumps), Subsection 6.3.2.2.3 (HpSI pumps) and Appendix 6A, Subsection 3.2.1 (containment spray pumps).

The pump motors require 75% voltage initially, increasing linearly to 90% in two seconds, then to 100% in the next two seconds. This information is provided to the Applicant for its use in sequencing safety system loads. Regulatory Guide 1.9 is not within the licensing scope of CESSAR-F.

b) ~ The starting time and voltage requirements provide adequate torque during the accelerating period of pump load. It should be noted that pump motors are normally procured 'by the pump vendor. The pump vendor, therefore, is responsible for ensuring ~ that the. motor has adequate torque. C-E reviews the vendor " pplied data of motor and pump torques to ensure that the acceleration characteristics are acceptable.

i 040.4 c) The designs for large horsepower motors include stator winding (8.3) 4 Cont. temperature sensors. These sensors can be used to provide steady-state temperature monitoring of the stators. Use of these sensors to monitor starting (or other high slip conditions) is considered impractical. During these transients, peak temperatures are experienced in the rotor and not in the sensor areas. Motor i

starting circuits, which provide short term protection for starting, are in the Applicant's scope.

As a minimum, C-E motors are designed to allow the NEMA prescribed number of starts without degradation. Adherence to the limitations I

on the number of starts will provide adequate protection from thermal damage due to starting.

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040.5 Provide a list of emergency load requirements of loads used to shutdown '

(8.3) the reactor and acting to limit the release of radioactive material for

the following conditions.

1. Loss of Coolant Accident

2. Steam Line Break

3. Blackout (loss of offsite power) -

r In addition, provide the requirements for the safeguards actuation sequence and a load sequence internal for the abovi conditions.

Resconse The emergency load requirements for components in the Safety Injection System (SIS), Chemical and Volume dentrol Systems (CVCS), and the Shutdown Cooling System (SCS) used to shutdown the reactor and acting to limit the release of radioactive material are specified in Table 8.3.1-3 of CESSAR-F. These essential components are powered from the standby generators so that a loss of offsite power would not prevent their operation.

The safeguards actuation sequence interface requirements are given in Section 8.3.1 and in Table 8.3.1-4 of CESSAR-F. Lead sequence intervals -

for the standby generators are discussed in the Applicant's Safety Analysis Report.

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! 131.1 The response spectra in Figures 3.7.1-1 through 3.7.1-4 do not have (3.7.1) adequate description and specification. Finer gradation for both l horizontal and vertical scales shor.id be provided for better inter-pretation. Numerical values for 74ximum, minimum and other control i

points of spectra should be supplied.

. Response '

As requested, horizontal and vertical scales are added to the above figures via Amendment No. I to CESSAR-F.

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131.2 The Section 3.7.1.4 " Supporting Media for Seismic Category I (3.7.1) Structures" should be addressed.

Resoonse Supporting media for seismic category I structures is not within the scope of CESSAR-F. The section is added to CESSAR-F in Amendment No.1, and refers to the Ar'licant's SAR.

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I 131.3 Interaction of Non-Category I structures with Seismic Category I (3.7.2) structures should be addressed.

Resconse Interaction of non-category I structures with seismic category I structures is not within the scope of CESSAR-F. The section is added to CESSAR-F in Amendment No.1, and refers to the Applicant's SAR.

Note that the interaction of other piping with seismic category I piping is addressed in Secticn 3.7.3.13.

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131.4 Informat on on " Analysis Procedure for Damping" should be adequately i

(3.7.2) prov" -

Resconse Unifcrm modal damping factors are used in the analysis of the coupled components of the reactor coolant system. In Amendment No.1, Section 3.7.2.15 is revised to identify this procedure, Section 3.7.2.1.1 is revised to clearly identify the range of damping values used for analysis of CESSAR-F systems and subsystems, and Table 3.7.2-1 is revised to delete damping factors for items not within the scope of CESSAR-F.

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i 131.5 In Section 3.7.3.4, describe the criterion to modify the restraint (3.7.3) and supports systems in order to maintain the fundamental frequen-cies of equipment r-d - Osystems sufficiently removed from the resonance range.

Response

The criterion is that stresses and deformations obtained frem analysis of the preliminary design do not exceed established acceptable design limits. The subsystem supports design is sufficiently adaptable such that, dependent on the subsystem involved and the quantitative frequency change required, modifications can be made either by i changing the stiffness of existing support assembly components or by adding additional support system restraints te the equipment whose response otherwise exceeds established limits. Section 3.7.3.4 is revised in Amendment No. I to more clearly define the criterion used and, when required, how the modification is accomplished. For

, Applicants which reference CESSAR-F, the analysis has been completed and fundamental frequencies of equipment are all sufficiently removed

, from the resonance range.

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212.3 As required by Regulatory Guide 1.70, Section 5.4.7.2.2 should be amended (5.4.7) to provide the bases for the sizing of RHR relief valves.

Resoonse The bases for sizing of the RHR relief valves are provided in Section 5.4.7.2.2.4.a.

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212.4 As required by Regulatory Guide 1.70, provide a discussion in Section (5.4.7) 5.4.7.2.6 of manual actions cutside the control room required for normal operation (heatup and cooldown) of the shutdown cooling system (SCS).

Provide a discussion of manual actions required outside the control room in the event of single active failures to the SCS.

ResDonse As stated in Section 5.4.7.1.2 of CESSAR-F, "... it is a functional design basis for the shutdown cooling system that no single active failure prevents at least one complete train of the shutdown cooling system from being brought on line from the control room whether this is during normal plant cooldown or following a design basis event."

Therefore no manual actions are required outside the control room.

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212.5 As required by Regulatory Guide 1.70, provide a listing of the design

, (5.4.7) pressure and temperature of eacn ccmcenent in the 2.HR system anc  ;

1 explain the bases for the selection of these design values.

Resconse

Section 5.4.7.2.2 provides the design pressures and temceratures for i i

the components of the shutdcwn cooling system and provides the cases 1

for selection of these values. Tc avoid duplicating informatie.

t p portions of Chapter 6 are referenced to provice descriptions of those ccmeonents wnich are shared with the Safety Injection System.

l Section 5.4.7.2.2 and the referenced portiens of Chapter 6, together,

! provice the requested infor .ation.

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, 212.6 As required by Regulatory Guide 1.70, identify each component of the i j (5.4.7) SCS that is also a portion of some other system (e.g., ECCS). '

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j The shutdown cooling system is shown in the Reactor Coolant System '

P&ID (Figure 5.1.2-1) and the Safety Injection System (SIS) P&ID (Figure 6.3.2-1A, IB). The SIS is the only system with which SCS components are shared. Tne shared components are shown in the i SIS flow diagrams for the shutdown cooling mode, Figures 6.3.2-1J, 1K. and IL.

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212.7 Your discussion of plant equipment is too general in Section 6.3.5.

(6.3.5) Identify each piece of equipment you reference by its number. For example, in Section 6.3.5.3.3.6, provide valve numbers for valves which will have position indication in the control room.

Resconse The information requested is provided by the Safety Injection System P&ID, Figure 6.3.2-1. Section 6.3.5.3 is revised in Amendment No. 1 of CESSAR-F to reference the P&ID.

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i ' Response The requested information is provided in Section 6.3.2.2 of CESSAR-F.

! This section discusses ECCS isolation vcives and associated interlocks.

Section 15.3.2.2 references Section 7.6 tir more detail of the control e

circuitry.

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212.9 As required by Regulatory Guide 1.70, provide a failure mode and effects (6.3) analysis for the ECCS. Include in your analysis the effect of potential passive failures of fluid systems during long-term cooling and potential failures which might effect boron precipitation during long-term cooling.

Resoonse A Failure Modes and Effects Analycis (FMEA) is presented in CESSAR-F as Table 6.3.2-2. The FMEA addresses failure modes of the components of the safety injection system as described by the Piping and Instrument Diagram presented as Figures 6.3 1A and 18.

The actuated components are addressed in line items 1 through 30. The FMEA is organized so as to paralled the system operating modes described by Figures 6.3.2-1C through II. The failure modes of the actuated components are addressed as follows:

FMEA MODE FLOW DIAGRAM LINE ITEMS Injection Figure 6.3.2-1C, D 1 to 15 Short Terms Figure 6.3.2-1E, F 16 to 20 Recirculation Long Term Figure 6.3.2-1G, H 21 to 22 Recirculation Shutdown Figure 6.3.2-11 23 to 30 Cooling Additionally, failure modes of unactuated components, including passive failure modes, are addressed in line items 31 through 42. Boric acid precipitation during long term co,oling is precluded by realigning the flow from the HPSI pumps (after two hours) to divide flow equally between the hot and cold legs. This manual operation is accomplished within the constraints of the single failure criterion and is specif-ically addressed in line item 22 of the FMEA.

212.13 As required by Regulatory Guide 1.70, provide assurance that ECCS valve (6.3) motor operators and safety equipment vital to LOCA mitigation located inside containment will not be submerged following a LOCA.

Response

Interface Requi ement 6.3.1.3.Q of CESSAR-F requires that the Safety Injection System "...shall be provided with an independent control system such that the safety related equipment in each train operates within the environmental design limits specified in Section 3.11."

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Therefore, it is in the Applicant's scope to protect the specified equipment from submergence.

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212.11 Provide identification of all process instrumentation (see Section (6.3.2) 6.3.2.7) available to operators in the centrol room to assist in assessing post-accident conditions. Discuss the information avail-able to the operatar, the time delays during which his failure to act properly will nave no unsafe consequences, and the consequences if such action is not performed at all. .'

Resoonse 4

Post Accident Monitoring Instrumentation is discussed and listed in Section 7.5 and Table 7.5-3. All of this instrumentation is Class IE and its output is available in the control room. In addition, the -

Sequence of Events Diagre.m presented in Figures 6.3.3.5-1A to IF indicate the instrumentation used in any given event and the operator action required.

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212.12 As requested by Regulatory Guide 1.70, provide a list of the accidents (6.3.3) in Chapter 15 which result in ECCS operation and sunmarize the con-clusions detailed in the accident analyses.

Response

Section 6.3.3.7, Chapter 15 Accident Analyses, is added to CESSAR-F in Amendment No. 1.

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212.13 As required by Regulatory Guide 1.70, state the bounds within which (6.3.3) principal system parameters c.ust be maintained in the interest of constant standby readiness. For example, the minimum boron con-centrations in borated water sources, the maximum number of inoper-able components and the maximum allowable period a component may be out of service should be discussed.

Response

The requested information is provided in Chapter 16, Section 3/4.5.

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, 212.14 Appendix A to the CESSAR FSAR addresses C-E's response to various (Appendix A) Category II and III items which were noted by the RRRC subsequent to the regulatory requirements cut-off date for the CESSAR PSAR.

C-E's responses to these items, however, are at the Construction

Permit level of detail rather than at the Operating License level.

1 C-E's responses to the Category II, III and IV items which pertain to the RSS area of responsibility should be addressed at an Operating License level of detail. (Note: At the time this question was pre-

pared, the Category IV items were not addressed in the appendix.)

Resconse Please refer to the response to Question 000.4.

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i 221.1 provide a more detailed discussion of the test data from the (Appendix 4B) hydraulic tests on the 3/16-scale reactor vessel model. The discussion should include the results of measurements of .

pressure loss from the inlet to the outlet of the vessel and between significant intermediate points, and the results of

measurements of the fluid mixing that occurs between the vessel inlet nozzles and the core inlet, and between the inlet and ,

j outlet of the core. Further information supplied should in-

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clude quanitifcation of uncertainties in the test data.

Response '

Additional information is provided in Appendix 4B in Amendment i

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No. 1. Table 4.4-4 is also revised in Amendment No. I to pro-j vide data consistent with Appendix 48.

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221.2 The staff is performing a generic study of the hydrodynamic sta-(4.4.4.5.3) bility of light water reactors, including the evaluation methods for CESSAR System 80. The results of the staff study will be applied to the acceptability of the stability methods now in use by reactor vendors. Combustion Engineering shoulo provide a dis-cussion of any hydraulic stability test program planned to verify the methods 'used.

Response

Combustion Engineering believes that the test data discussed in Section 4.4.4.5.3 of CESSAk-F provides verification of the ficw stability methods used by C-E. Because of these supporting data and the analytical trends, also discussed in Section 4.4.4.5.3, indicate that flow instabilities will not adversely 2ffect the System 80 core thermal margin during normal operation or antici-pated operational occurrences, no additional hydraulic stability tests are required.

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221.3 Combustion Engineering must provide a commitment to obtain reactor data on rod bow for the purpose of verifying the rod bow model used in the CESSAR System 80 Design.

Response

In accordance with our agreements with both DOE and EPRI, Combustion Engineering is ccomitted to obtaining rod-to-red channel width data from fuel in the initial core loading of Arkansas Nuclear One Unit 2.

These data will, as a minimum, be obtained at the end of cycle 1, at the.end of cycle 3, and following extended burnups to be achieved in cycle 4. Because of the similarity in fuel design, these 16 x 16 data will provide a basis for verifying the rod bow model used in the CESSAR-F System 80 design.

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231.1 Our review of the Systems 80 CESSAR FSAR will be conducted in ac sedance to the recent Revision 1 of Section 4.2 of the Standard Review Plan. We i have ccmpared the information contained in the CESSAR FSAR to that infor- t mation described in the SRP and have determined that there are many issues  !

that are not now addressed but will ultimately need to be addressed in the FSAR. We therefore recommended that C-E review Revision 1 of Section 4.2 of the Standard Review Plan for information that is currently missing in the FSAR.

Response

C-E has reviewed Revision 1 of Section 4.2 of the Standard Review Plan.

Coments developed by our review were forwarded by C-E letter, LO-79-058, r

A. E. Scherer to H. R. Denton, dated September 17,1979, "Cocunents on Standard Review Plan SRP 4.2". Consistent with those coments, we have reviewed CESSAR-F and believe that Section 4.2 provides all of the necessary information.

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231.2 An unexpected degradation of guide tubes that are under Control Element Assemblies (CEAs) has been observed in irradiated fuel assemblies taken from operating Combustion Engineering reactors. Apparently, coolant turbulence is responsible for inducing vibratory motion in the normally fully withdrawn control rods. When these vibrating rods are in contact with the inner surface of the guide tubes, wearing of the guide tube wall has taken place. Significant wear has been found to be confined to the relatialy soft Zircaloy-4 guide tube because th Inconnel-625 cladding on the control rods is a relatively hard wear surface. The extent of the observed wear has appeared to be plant dependent, but has in some cases extended completely through the guide tube wall. Combustion Engineering

, shculd provide the staff with an analysis of the method by which they will deal with this praclem.

Additional design drawings chauld also be incorporated into the FSAR. These drawings should detail penetrations in the fuel slignment plate, CEA shrouds with any scuppers, and drawings of other relevant components (depending on the particular solution that is proposed by C-E).

Resoonse The System 80 fuel assemblies, control rod assemblies, and reactor internals include features which are different from those in Combustion Engineering plants which have previously experienced guide tube wear. As can be seen in Figures 3.9.5-1 and 3.9.5-6 of CESSAR, the lower portion of the upper guide structure (UGS) is comprised of a tube array. Each tube shrouds each individua? control rod element and protects the CEA from high coolant flow rates in the core exit region. We are testing the System 80 twelve 1

rod CFA (along with a portion of the UGS and fuel assemblies) in our hot-

-flow test facility.. The test was described in an April 30, 1980 meeting with the NRC staff on the subject of guide tube wear. Slides from the meeting (proprietary) were t'ransmitted to the staff in a letter from A. E. Scherer to Robert A. Clark, April 30, 1980, LD-80-019. Following completion of the tests and evaluation of the results, we will advise the NRC staff as to whether any changes in the System 80 design are warranted.

This information should be available by mid-1981.

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231.3 In various places the CESSAR FSAR references CENPD-178, " Structural Analysis of Fuel Assemblies for Combined Seismic and Loss of Coolant Accident Loading." This topical report is referenced as the method-ology by which analyses are performed to show that the System 80 fuel assemblies can withstand the combined seismic and LOCA mechan-ical loads. We have notified (D. F. Ross letter to A. E. Scherer dated February 2,1978) C-E of shortcomings in this report and requested tnat C-E revise the topical report and provide additional information and analyses. Additionally, an NRC staff task force is working to establish acceptance criteria for this type of analysis.

The FSAR should be revised to reflect the revisions to CENPD-178, the forthcoming NRC acceptance criteria, and the application of the seismic and LOCA structural analysis to specific plants that reference the CESSAR FSAR. This issue must be resolved before Section 4.2 of the CESSAR FSAR can be approved. .

Resoonse Combustion Engineering is currently preparing a revised version of CENPD-178 which addresses current NRC requirements for demonstrating structural adequacy of tuel assemblies for seismic and LOCA loading.

This revised topical report is curren.tly scheduled to be submitted in the second quarter of 1981. Affected portions of the CESSAR-F will be revised following this submittal.

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231.4 The need for routine surveillance 's discussed in Revision 1 of Section 4.2 of the Standard Review Plan. The CESSAR FSAR does not provide for such a program. Accordingly, C-E should suggest a surveillance program, which the applicant may reference, that includes a description of (a) the on-line fuel rod failure detection method, (b) CEA integrity assurance, and (c) post-irradiation fuel surveillance plan.

Response

The detection of fuel rod failures during the operation of the plant is based on periodic measurement of the coolant activity. In accordance with technical specifications, the activity, due primarily to the iodine isotopes, is monitored and limited. In addition, changes in the levels of coolant activity within the level allowed by the technical specifications are useful indicators of any changes in the number of fuel rods leaking fission products to the coolant.

The basic design of fuel assemblies and Control Elenent Assemblies (CEAs) used in the CE3SAR plants is similar to the design used in Arkansas Nuclear One Unit 2. Surveillance conducted in ANO-2 will demonstrate the mechanical integrity of the design in advance of the operation of the first plant based on CESSAR.

However, routine surveillance is not in the scope of CESSAR-F.

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231.5 In Section 4.2.1.1.3, please state what dimensional, requirements are imposed on the fabricated length of guide tubes. Also, in Section 4.2.1.2.2.2, state what dimensional requirements are imposed on the length of the finished fuel rod (after welding of end caps). This information is needed to enable us to perform audit calculations on various design features, such as fuel rod growth allowances.

Response

Additional information, beyond the dimensions requested, should be required to calculate fuel rod growth allowances. This information is considered proprietary by C-E and would have to be sur3 lied separately. Since the question indicates that other types of audit calculations are planned, we request that the NRC staff identify the specific information needed.

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4 231.6 In Section 4.2.1.2.1, which dibcusses fuel cladding design limits, it is stated that C-E's criteria do not prohibit fuel rod pressures from exceeding the primary coolant pressure. Our acceptance criterion for fuel rod pressure is given in sub-section II-A-1, paragraoh (f) of i Revision 1 to the Standard Review Plan. The criterion states that fuel and burnable poison rod internal gas pressures should remain below the nominal system pressure during normal operatica unless otherwise just-ified. C-E should either show that rod pressures remain below system pressure or provide justification to excaed that pressure. Should C-E choose to justify higher pressures, a substantial amount of new infor-mation would have to be submitted (see, for example, WCAP-8964, " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis)."

Resoonse i The C-E fuel cladding design limit related to internal pressure states in part the the internal pressure will not cause the clad to creep outward from the fuel pellet surface during normal operation. Al though it is not .necessary that the internal fuel rod pressure remain below the primary system pressure to satisfy this fuel design limit, it is sufficient. Based on previous analysis of similar designs, it is expected that the internal fuel rod pressure for the standard C-E PWR 1

NSSS (Cycle I core burnup of 13,740 AWD/MTU) will remain below primary i

system pressure.

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310.1 Main steam line breaks (MSLB) are listed in Table 15.0-1 of the CESSAR (15.1.5) FSAR. However, only the small MSLB outside containment has been analyzed. Provide the analyses for a spectrum of main steam line breaks, and in particular that of a large MSLB, in accordance with the appendix of Section 15.1.5 of the Standard Review Plan (NUREG-75/087).

Response

As indicated in Table 15.0-1, both large and small steam 'line breaks were considered, however, Regulatory Guide 1.70 Revision 2 requires quantitative evaluation and presentation of only the limiting events in each type and frequency category. Large steam line breaks are not limiting with respect to the acceptance criteria when compared to other events in the same frequency categories, and therefore, are not pre-sented in Chapter 15. The bases used in the analyses and in calculating the radiological consequences are provided in Section 15.0.4. In addition, it is C-E's position that Chapter 15 should only address events with occurrence frequencies which are greater than 10-6 per reactor-year. Therefore, event combinations described in Section

.5.1.5 of the SRP which have calculated frequencies of occurrence less than 10 -6 per reactor-year were not presented. However, to meet additional NRC requirements for docketing CESSAR-F, Appendix 15C was provided which addresses the SRP Section 15.1.5 events which are not within the scope of Chapter 15.

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310.2 The control element assembly (CEA) ejection accident is analyzed (15.4.5.2) in Section 15.4.5.2 of the CESSAR FSAR with the assumption of the failure of a fast transfer of one-half of the= non-ESF loads such that half the reactor coolant pumps continue to operate. Provide an analysis of this accident assuming the total loss of offsite power in accordance with the appendix of Section 15.4.8 of the Standard Review Plan.

Resconse It is C-E's position that Chapter 15 should only address events

- with occurrence frequencies which are greater than 10 ~0 per reactor-year. The calculated frequency of occurrence of a control element assembly ejection combined with a loss of offsite power is less than 10-6 per reactor-year. This event, therefore, is not included in the CESSAR-F Chapter 15.

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310.3 Provide an analysis of the steam generator tube rupture in Section (15.6.2) 15.6.2.2 of the CESSAR FSAR in accordance with the assumption and acceptance criteria contained in Section 15.6.3 of the Standard

, Review Plan.

Response

Regulatory Guide 1.70, Rev. 2, requires categorization of events by expected frequency of occurrence and requires quantitative evaluation and presentation of only the limiting events in each type and freq-uency category. With this as its basis, Section 15.6.2.2 of CESSAR-F presents the Steam Generator Tube Rupture (SGTR) event as the limiting radiological release event in the infrequent category. The expected frequency of occurrence of the SGTR event with Loss of Off-site Power (LOP), as defined in Section 15.6.3 of the Standard Review Plan, is classified as a limiting fault-3 (LF-3) event. In the uF-3 category, the Double Ended Break of the Letdown Line Outside Con-tainment-Upstream of the Letdown Control Valve (DBLLOCUS) event, not the SGTR with LOP event, is the limiting radiological release event. Thus, as per the guidelines in Regulatory Guide 1.70, Rev. 2, the DBLLOCUS, not the SGTR with LOP, is presented in Section 15.6.5.2 of CESSAR-F. Although the SGTR and DBLLOCUS release about the same

- inventory from the reactor coolant system during the transient, the DBLLOCUS releases this inventory to the auxiliary building while the SGTR release is to the steam generators. Since the decontamination factor is much lower for the auxiliary building (i.e., DF=2) than for the steam generator (i.e., DF=10), the resulting radiological release from the DBLLOCUS is higher. However, Appendix 150 is revised in

310.3 Amendment No. I to present a qualitative description of the SGTR with (15.6.2)

Cont. LOP event and a comparison of consequences between the SGTR with LOP and DBLLOCUS events.

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321.1 For all systems interfacing with the liquid, gaseous, and solid radwaste treatment systems, provide detailed interface information, including references to P&ID's, pressures, temperatures, flow rates, and expected volumes of waste input to each of the radwaste systems. Where appropriate, provide interface information for the Process and Effluent Radiological Monitoring and Sampling System.

Response

The requested information is provided in Section 11.1.9 in Amendment No.1.

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4 331.1 Describe system and equipment design considerations for the System 80 (12.1) NSSS that are directed toward ensuring that occupationa: radiation exposures are ALARA. Describe how experience from past designs and from operating plants is utilized to develop improved design for ensuring that occupational radiation exposures are ALARA. Describe how system and equipment design is directed toward reducing the need for maintenance of equipment. Describe what specific features are included to minimize dose to personnel during inservice inspection.

Include an indication of whether, and if so, how the design con-sideration guidance provided in Regulatory Guide 8.8 will be followed; if it is not followed, describe the specific alternative approaches to be used.

Response -

The requested information is provided in Section 12.1 of Amendment

No. I to CESSAR-F.

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4 331.2 - Describe the specific review functions and responsibilities of the (12.1) personnel who ensure that the features for maintaining occupational radiation exposures as low as is reasonably achievable are included in the design of the System 80 NSSS. Describe the procedures for assuring that adequate radiation protection reviews are performed throughout the design process.

i Response The requested information is provided in the last two paragraphs of Section 12.1.2 of Amendment No. I to CESSAR-F.

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! (12.1) occupational radiation exposure as low as is reasonably achievable during the eventual decommissioning of CESSAR reactor plants.

Response

l i The requested information is provided in the mid-section of Section 12.1.2 in Amendment No. I to CESSAR-F.

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331.4 Using calculational models and experience from operating reactors, (12.2) provide a listing of the maximum expected activities due to crud

. deposits on steam generator tubing and primary system piping for the System 80 NSSS.

Response

The information requested is provided in Section 11.1.2 of CESSAR-F -

specifically Tables 11.1.2-7 and 11.1.2-9.

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I-4 331.5 Supply a listing of the maximum specific source strengths (in (12.2.1.2.1) MeV/ gram-sec) in the shutdown cooling system (SOCS) at various 4 times after reactor shutdown. Provide these source strengths

for gamma energy groups between 0.2 and 4.0 MeV/gama.

Response j The requested information is provided in Table 12.2-11 of

4 Amendment No. I to CESSAR-F.

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331.6 For the following reactor plant (NSSS) components, describe the (12.3.1.2) specific design features which will be used to assure that occupational exposure due to operations and maintenance of the System 80 NSSS will be ALARA: valves, piping, evaporators, heat exchangers, demineralizers and sample stations.

Response

Section 12.3.1.2 of CESSAR-F is revised in Amendment No. I to provide the requested info nation.

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331.7 The information provided in the position presented on A-34 of Appendix A is insufficient.

Provide specific commitments with -d to:

(a) how system chemistry will be controlled to minimize buildup of activated corrosion products, including which parameters will be controlled and the nature of the limits to be applied du"ing operation.

(b) the limitations to be applied to cobalt content on reactor plant materials exposed to primary coolant.

(c) the specific surfaces within the primary coolant system to which stellite limitations will be applied.

(d) how erosion will be controlled by using favorable geometrics and lubricants, and by control of leakage purge across journal sleeves.

(e) cleanup systems for reme"*.1 of activation products from the prir.ary coolant during operr! con.

(f) which design moi . .ations provide for more efficient inspection of primary system welds.

Resconse The Appendix A position statement on Regulatory Guide 8.8 is revised in Amendment No.1 to provide the requested information.

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1 331.8 Describe any plans that you have to use removable insulation (12.3.1.2) sections over reactor vessel nozzle welds to facilitate in-service inspections and thereby reduce exposure times.

Response

Section 12.3.1.2 is revised in Amendment No. 1 of CESSAR-F to provide the requested information.

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'331.9 Field-run piping carrying radicactive material should be routed so (12.3) as to minimize personnel exposures. Describe how your design reflects this consideration.

Response

Section 12.3.1.2 of CESSAR-F is revised in Amendment No. 1. The requested infonnation is pruvided in Item G.

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423.1 Modify or expand Chapter 1 and Chapter 14 of your FSAR, as appropriate, (14.2) to address the following Regulatory Guides:

1.68 (Revision 2, 8/78), " Initial Test Programs for Water-Cooled Nuclear Power Plants."

1.68.2 (Revision 1, 7/78), " Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants."

1.79 (Revision 1, 9/75), "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors."

Provide technical justification for each exception to the regulatory positions in the guides. Also, modify the test descriptions in Chapter 14 to show that tests can be performed in accordance with these Reg-ulatory Guides.

Resconse These regulatory guides have been addressed in Appendix A of CESSAR-F.

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