ML20028H393

Forwards Request for Addl Info on DC Application for C-E Sys 80 Design Project 675.Response Required within 90 Days of Reciept of Request

ML20028H393
Person / Time
Site:	05000470
Issue date:	12/24/1990
From:	Wambach T Office of Nuclear Reactor Regulation
To:	Erin Kennedy ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
References
PROJECT-675A NUDOCS 9101020300
Download: ML20028H393 (12)

Text

C December 24, 1990 Mr. E. H. tenhedy, Manager Nuclear Systems Licensing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Conntcticut 00095 L

Dear Mr. Kennedy:

l SULJECT:

REQUEST TOR ADDITIONAL INFORMATION ON CESSAR DC, SYSTEM 8: +

Enclosed is a request for additional information in addition to the request dated Decerber 21, ISSO.

This request is bestc on a prelianinary review by the Reactor Systems Brench of Chapttrs 1, 6,16 and Appendix A & B of CESSAR-DC.

There will be additional questions in these areas as we progress in our dete11ed review.

To allow us to establish a realistic schedule, you must rtspond within 90 days of receipt of this request. The reporting and/or recordkeeping requirements i

contained in this letter affect fewer than ten respondents; therefore, OMB

-clearance is not required under P.L. 96 011.

Sincerely, p/

I Thom6s V. Wambach, Project Manager Standardization Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated DISTRIBUTION:

4entrabme TMurley/FMiraglia,12G10 NRC PDR Chiller PDST R/F WTravers Pshea.

DCrutchfield, 13A2 TWambach JTaylor, 17G21 EJordan, MNBB3701 EJordan, MNBB3701 000.-1EB18 JPartlow, 12G1B ACRS-(10), Phillips

.S]Of f

LA:P0( '

PM!fDST D:PDST PShea TW6mbpchicw Chille M U 12/A7/90 12p/90 12/*/90 DOCUMENT NUIE:

LTR TO MR. E. KENNEDY 12/24 we oe

,,m _ em m 4

3100F

~

I pl*Alop4 D

l .,,..g ;i UNITED STATES 5c

-f NUCLEAR REGULATORY COMMISSION 9* v[B>f,f.

W ASHINGTON, D C. 20M6 g

December 24, 1990 Mr.E.H.LennedyIcensing Manager Nuclear Systems L Corbustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095

Dear Mr. Kennedy:

{

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON CESSAR-DC, SYSTEM C0+

Enclosed is a request for addition 61 information in addition to the request d6ted December 21, 1990. This request is based on a preliminary review by the Reactor Systems Branch of Chapters 1, 6,15 and Apptindix A & B of CESSAR DC.

There will be additional questions in these areas as we progress in our detailed review.

To allow us to establish a realistic schedule, you must respond within 90 days of receipt of this request. The reporting end/or recordleeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not requind under P.L.96-511.

Sincerely, k

W )'

W ltf f Thomas V. W6mbach, Project Manager Standardization Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation i

Enclosure:

As stated s

l

Ccr.bustion Engineering, Inc.

Project No. 675 cc: Mr. A. E. Scherer, Vice President Huclear Quality ABB Combustion Engineering Nuclear Power 1000 Prospect flill Road Fett Office Bcx 500 Windsor, Connecticut Of095-055 Mr. C. B. Crinknan, tianager Washington Nuclear Operations Combustion Engineering Inc.

12300 Twinbrook Parkway Suite 330 Rockville, Maryland 20852 Mr. Stan Pitterbusch Nuclear Licensing Combustiun Engineering 1000 Prospect 11111 Poad Post Office box 500 Windsor, Connecticut 06095-0500

REQUEST FOR ADDITIONAL INFORMATION ON THE DC APPllCATION FOR

]

COMBUSTION EHGINEERING SYSTEM 80+ DESIGN FROJECT NUMBER 675 CESSAk-DC REACTOR SYSTEMS BRANCH 440.10 Section 1.13 of the SAR states that CESSAR DC will take into account severe accident acceptance criteria from EPRI ALWR & DOE ARSAP Topic Papers. The referenced documents do not provide staff acceptance criteria for severe accident issue retolution. Such criteria are documented in the staff Draft Safety Evaluation Report on Chapter 5 of the EPRI Requirements Document and applicable SECY Papers such as90-016. The SAR write-up should be clarified to show compliance with approved staff positions.

440.11 Section 1.2.1.3 of the SAR states that for multi plant sites, inde-pendence of safety related systems is maintained between individual plants. The ability to cross tie systems (between units) in an emergency may enhance safety system functional availability.

Has this capability been considered, if so, what are the competing benefits and risk associated with such cross ties?

-440.12 Section 1.9 is identified as design interface criteria.

No criteria are.specified.

Identify all site specific criteria which an applicant will have to comply with, or indicate that no such criteria exist.

l 440.13 WithrespecttoAppendixB(ProbabilisticRiskAssessment), identify all system and component design assumptions utilized in the CESSAR-DC System 80+ PRA which are in systems beyond the 80+ design scope or j-if in the design scope, not yet developed in complete detail.

Propose a reliability validation program which will ensure that all assumptions which went into the PRA are satisfied by an as-built plant.

I 4

9

)

440.14 Table Bl.S 1 of Appendix B provices comparisons of severe accident frequency for System 804 verses System 80.

Recent scoping studies haveindicatedthateventswhennotatpower(i.e. shutdown)can provide significant risk contribution, Have these events been analyzed and incorporated into the PRA, and if so provide relevant information in the sumary table?

If not i

evaluated, provide a schedule for submitting this additional analysis.

440.15 (ATWS) i Provide discussion on any analysis performed relative to peak primary system pressure, fuel performance and radiological consequences following a limiting ATWS transient for System 60+ design. As an alternative, discuss the applicability of the previously performed analysis for CE plants to the System 80+ design.

440.16 (Shutdown Operations)

(a) GenericLetter(GL)87-12requestedinformationregarding l

lowered RCS inventory operation, please provide a response to the generic letter with respect to the System 80+.

(b) Please describe instrumentation provided to the operator during shutdown operations which characterize the state of the reactor coolant systeni (RCS).

Include RCS level, RCS temperature, and shutdown cooling system (SCS) performance and provide a description of the appropriateness and accuracy of each instrument with respect to its intended function.- Also, include identification of audible and visual alarms used to delineate out-of-range conditions, including the values which-constitute those conditions.

I i

i

\\

3 (c) The staff has identified that Diablo Canyon, Unit 2, was in a condition not previously analyzed by the kRC staff during the loss of RHR event of April 10,1987 (NUREG 1269).

Please describe the steps that have been taken to alleviate this situation f or the System 80+.

l (d) NUREG-1269 contains the statement " Design of the nuclear steam supply system (NSSS) did not appear to provide dttailed provisions for mid-loop operation." please address this identified deficiency in PWR design with respect to the System 80+ design.

Include identificction of and discussion of each of the design changes in the System 80+ which represents an improvement over existing designs and establish the I

adequacy of the System 80+ design for lowered RCS inventory operation.

(e) NUREG-1209 identified that conteinnent was open throughout the April 10, 1987 event at Diablo Canyon, and there were no procedures to reasonably assure containment closure in the event of progression of the accident to a core damage condition. Address this situation with respect to the System 80+ design and the anticipated methods that will be used to operate the plant.

Include such design considerations as the need for removal of the equipment hatch and improvements in the System 80+ design which facilitate rapid replacement of the hutch should the need arise. Simi16rly address other containment penetrations and potential bypass paths.

(f) The Diablo Canyon event and subsequently obtained information have shown operating procedures to be inadequate for lowered RCS inventory operation. What plans exist for recommending improved procedures and administrative controls to System 80+

owners / operators so that this situation is eliminated.

What changes will be made to the Emergency Procedure Guidelines (EPGs) (CEN-182) to accomodate this need?

4 (g) What equ1paent exists in the System 80+ that can be used to assure adequate core cooling in the event of a complete loss of SCS?

(h) The Vogt'" loss vf all safety-related AC power event occurred inMarch1990andtheNRCincidentinvestigationteam(11T) identified many areas of concern involving shutdown operation and they bre documented in NUREG-1410.

please address these staff concerns with respect to the System 80+ design.

(1) EvidenceexiststhatccrtainTechnicalSpecifications(TSs)may not be optimum when consideration is given to operation during non-power conditions.

For example, requirements for SCS suction valve it.terlocks inpact upon SCS reliability SCS flow rate recuirements n$y overly restrict flow rate range end increase the likelihood of loss of SCS due to vortexing, and TSs written on the basis of the state of the NS$$ and/or evtainment may be impacted, plebse address this topic with respect to the System 80+ design and provide recommendations for iniprovement, particularly with respect to the unique design aspects of System 80+,

(M Safety analysis reports (SARs) typically concentrate on power operation when consideration is given to many of the potential operation 61 transients 4 The recent experience from the events in operating reactors indicated that further evaluation for plant operation at lower modes may be required.

Hence, it may be prudent to address non-power operation in more depth than has been traditional. What plans exist, if any, with respect to this topic bnd the System 80+prograta?

440.17 (IntersystemLOCA)

For future evolutionary ALWR oesigns, the design the low-pressure systems connecting to the primary coolant system must be designed to

l i!,

5 withstand full RCS pressure to the extent prtcticable, for those systems that could not be designed to withstand full RCS pressure, l

evolutionary ALWRS should provide (1) the capability for leak testing of the pressure isolation valves, (2) valve position irdication that is available in the control room when isolation valve operators are deenergized and (3) high pressure alarms to warn control room operators when rising RCS pressure systems and both j

isolation valves are not closed.

In considering low-pressure s

systems cor.necting to primary coolant system, all elements of the j

low-pressure system should be included (e.g. shutdown cooling system, letdown systeu, charging system, saft:ty injection 4

systems, instrument lines, pump stals, heat exchanger tubes, valve bonnets,etc.). Provide detailed discussion on how the Systtm 80+

design satisfies the above staff requireronts.

5 440.18 (Hydrogon Generation and Control)

In a lettcr dated August 28, 1990, you have indicated that CE will j

provide information to justify a system 80+ containment design l-consistent with the EPRI ALWR Requirements Document and NRC staff j

position on this issue.

The System 80+ design includes a hydrogen i

ignitor systen, (control grade) to assure :.% iance with 10 CFR' 50.34(f).

Provide detailed description t*d b J results of the-supporting atelysis to address the issue of hydrogen generation and q

control for System 80+ design.

Discuss the availability and func-tionality of the ignition system under potentional severe accident conditions; such as containment environment and initiating events such as station blackout.

440.19 (HighPressureCoreMeltEjection)

Section 6.7.1 states that the Safety Depressurization system will provide a capability to depressurize the RCS in response to a severe accident scenario.

Thetotalrapiddepressurization(bleed)flowis 4

designed to provide the capability to depressurize the RCS from 2500 L

j-

.-_,,u._.__..__.._.-_

l 6

to 250 psia within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a reactor trip and subsequent core darnage from a severe accident.

Prvvide the following additional information:

(A) What were the system objectives anc success criteria for rapid RCS depressurization during a severe accident scenerio?

(b) Details on the emergency operating procedures (E0Ps) developed for rapid depressurization of RCS to prevent high pressure core inelt ejection.

(c) Thermal hydraulic analysis performed to demonstrate that under the inost limiting severe accident scenario, the success criteria will be n.et using the E0P described in item (b).

Provide a sensitivity study showing impact of vent timing and operator response to sneeting the success criteria.

(d) What control room indications are used by the operators f or a successful rapid depressurization operation during a severe uccident?

440.20 (Equipment Survivability)

Provide discussions on "ograms developed for equipment survivability applicable to System 80+ design in light of the staff requirement addressed in NRC SECY-90-016.

440.21 (Testingofair-operatedvalves)

Recent experiences from operating reactors indicate that there are safety-related air-opetated valves that fail to perform their designed safety function wher, the safety-grade air backup system (nitrogen bottle)wasused. This is because for some plants the air-operated valves are only tested with supply air connected to ncn-safety grade normal air supply system. Discuss the test program of System 80+

1

- ~.

7.

design in light of the above staff concern.

440.22

'(feed and Bleed)

In a letter cated August 28, 1990, you stated that the initiation of feed and bleed decay heat removal on the System 80+

can be delayed f or up to 30 minutes following steam generatoi dry out.

Provide detailed discussion on the subject including the following:

(a) What were the success criteria for feed and bleed decay heat removal on the System 80 plus design?

(b). Discuss the emergency operating procedure (EOP) developed for feed and bleed decay heat removal to achieve the success operation.

(c) Provide thermal hydraulic analysis performed based on the plant configyration of the System 80+, to demonstrate that under the most limiting complete loss of secondary heat sink conditions, the success etiteria will be met using the E0Ps described in item (b).

(d) What control room indications are used by the operators for a successful feed and bleed operation?-

440.23 (Emergency Procedure Guidelines)

The currently available CE (EPGs) (CEN-152) may not be applicable to System 80+ design. Provide a discussion of the necessary modifications made to the existing-EPGs-applicable to System 80+.

440.F4 (Sixty-year Life)

In a -letter dated January 22, 1990, you stated that CE will identify the components and systems which are hffected for a 60-year

l 8

plant life.

Provide all necessary information in CESSAR-DC to support the staff review for a 60-year design life including information on fatigue, corrosion, thermal aging, reactor vessel enbrittlement, as well as all the components and systems which are affected for the extended plant life.

440.25 (PressurizerHeaters)

Provide discussions on the safety classification of the pressurizer heaters.

If they do not meet the requirements of safety grade standards (e.g. Seismic Category I, Class 1E power supplies, etc),

discuss how the CESSAR-DC plant could achieve cold shutdown without operation of the pressurizer heaters and meet the Branch Technical Position RSB 5-1.

440.26 The CE LOCA evalu tion model approved by the staff may not be applicable to System 80+ design with respect to plant specific configurations in node arrangement and control system. Confirm that a new LOCA evaluation model will be prepared for the System 80+ design.

440.27 (Turbinebypassvalvesfailure)

There were two incidents of Palo Verde Nuclear Generating Station relative to the turbine bypass valves failing open due to a single failure in the electrical systems. The consequences of these events ney not be bounded by the analyses documented in Section 15.1 of the System 80+ as an event with moderate frequency in occurrences.

Please discuss the design features that would prevent or reduce the frequency of these events happening or mitigate their consequences.

9 440.28 On page 6.3-7 of the CESSAR-F system 80 imposes a requirement that "The total volume in the piping from the reactor coolant system up to these salves shall be less than 30 cubic feet per line.

This volume shall be kept to a minimum 50 that the delay time for injection of borated water will be a minimum.* This provisico has been removed from CESSAP-DC for System 80+.

Explein why this, require-ment is no longer applicable.

440.29 Provide an evaluation of the ECCS design features and proposed operating procedures according to 10 CFR Part 50 General Design Criterion 4 as related to the dynamic effects associated with flow instabilities and loads (e.g., water hammer).

(

440.30 System 80+ relies on operation of one SI pump during pos+ '.0CA long term cooling operation (LTC).

Provide an evaluation of SI pump reliability for extended operation during post-LOCA LTC modt.

(Ref. CESSAR-DC Section 6.3) 440.31 The IRWST design criteria for ECCS in CESSAR-DC states ' Baffles and intake screens shall be installed to limit the particle size entering the IRWST to 0.09 inches in diameter in order to prevent flow blockage in SIS components anc oiping and in the reactor.

Provide additional information on the aCVJal design of the interface between the contain:aent sump and the IRWST in terms of drainage from the sump to the IRWST.

Include relevant diagrams.

(Ref.CESSAR-DC Section6.3) i 1

__ _ _