ML20040A322

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Requests Response to Encl Regulatory Positions & Requests for Addl Info for Continuing Review of BSAR-205.Complete Response Requested by 761115
ML20040A322
Person / Time
Site: 05000561
Issue date: 10/01/1976
From: Parr O
Office of Nuclear Reactor Regulation
To: Suhrke K
BABCOCK & WILCOX CO.
Shared Package
ML111090060 List: ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8201200761
Download: ML20040A322 (5)
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D cket File R. Vollmer i

OCT 9 1 E76 NRC PDR M. Ernst Local PDR W. Gannill LWR f3 File W. Mcdonald. MIPC i

R. C. DeYoung ELD Docket'No. STN 50-561 k.

D. B. Vassallo IE (3)

F. J. Williams ACRS(16) s D. Parr 4. Cox Babcock & Wilcox Company M. Rushbrook bec:

J. Bitchanan ATTH: Mr. Kenneth I. Suhrke R. Heineman T. Abernathy Hanager. Licensing D. Ross Nuclear Powcr Generation J. Knight P. O. Box 1260 R. Tedesco

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Lynchburg, Virginia 24505 H. Denton V. Moore Gentlemen:

ROUND 2 POSITIONS AND REQUEST FOR INFOR!CTION, BSAR-205 As a result of our continuing review of the Babcock & Wilcox Standard Safety Analysis Report BSAR-205, your response to certain Regulatory staff positions and requests for infomation is required. The specific infomation required is detailed in Enclosure 1.

Regulatory Positions are identified by (RSP) underneath the position number shown in

' Enclosure 1.

The requests herein concerning containment systems

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design were discussed in a meeting with your staff in Bethesda on September 27. 1976. We are prepared to meet with you to discuss further any of our positions to assure complete understanding of the factors at issue and the bases for our positions; however, we do not believe extended or iterative debate would be useful.

In order to maintain our licensing review schedule, we need your complete responses to the Enclosure 1 items by Hovember 15, 1976.

Please infom us within seven days af ter receipt of this letter of the date on which you plan to respond so that we may revise our i

schedule if necessary.

If you plan to appeal to licensing managsaent on any of these positions, please advise us of your intentions within two weeks.

Please contact us if you have any questions.

Sincerely,

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l DrIgInal signed b'y "

1. W. Dromorick s

Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management

Enclosure:

Positions and Requests f( r Additional information LWR #3 LWR 3 cc:

Se'e~neithig TCox/LL dbb b 9/ 3 0 /76 9/ g /76 8201200761 010403 PDR FOIA

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Sabcock & Wilcox Comparty OCT e1576 L

cc: Washington Public Power Supply System ATTN: Mr. J. J. Stein Managing Director P. O. Box 968 3000 George Washington Way Richland, Washington 99352 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox T4uclear Power Generation Division Suite 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 B. G. Shultz Project Engineer Stone a Webster Engineering Corp.

P. O. Cox 2325 Boston, Massachusetts 02107 Mr. A. H. Monteith Ohio Edison Company l

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47 tiorth itain Street Akron, Ohio 44300 Mr. W. E. Kessler Comonwealth Associates, Inc.

209 East Washington Jackson, Michigan 49201 1

P.obert J. Kafin, Esq.

115 Maple Street Glen Falls, t w York 12801 orrec e >

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h ABC 388 (Rev. >$3) ABC3E SMS W de os sovsamusst Pauenne omca vers.aee-see

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t-P, CNCLOSURE M POSITIONS AND RE0 VESTS FOR ADDITIONAL INFORMATION I

BSAR-205 DOCKET NO: STN 50-561 020.0 CONTAllMENT SYSTEMS i

022.6 The response to Question 222.1 is incomplete. Provide the mass and energy release data requested for the containment subcompartment analysis.

022.7 The response to Question 042.4 is incomplete. Justify that the double-ended rupture of a main steam line will result in the design basis main steam line break accident, or provide the mass and i

energy release data for a spectrum of main steam line breaks.

i 022.8 Provide the mass and energy release data for a postulated main feedwater line break.

022.9 The maximum environmental temperature that safety-related mechanical (RSP) and electrical equipment will be designed and qualified to is given in Table 3.11-2 as 300 F.

Main steam line break accident analyses I

have indicated that the containment atmosphere temperature may exceed 420'F. Therefore, it is our position that Table 3.11-2 must be revised to reflect a conservative upper bound on the maximum temperature expected inside the containment following a postulated main steam line break accident, or that a commitment be made to qualify the safety related equipment for the containment environmental s.

conditions calculated by the B0P designer.

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! i 022.10 The response to Question 022.3, which requested additional information (RSP) about the containment isolation system within the 8-SAR-205 scope of design, is unacceptable.

It is our position that the requested information should be included in B-SAR-205, and the interface requirements for the BOP designer should be identified.

022.11 The Engineered Safety Features Actuation System (ESFAS) provides (RSP) the signals for ESF actuation, including containment isolation.

For the containment isolation and cooling (CIC) signal, the ESFAS only monitors containment pressure. It is our position that there should be diversity in the parameters sensed for the

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initiation of containment isolation. Therefore, propose.other

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ESFAS signals that satisfy this position; e.g., safety injection signal.

022.12 Specify the maximum leakage rate and type of fluid for all containment isolation devices provided with the scope of B-SAR-205.

This infonnation is needed by a B0P designer to determine the potential bypass leakage for a dual containment plant.

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022.13 With regard to the combustible gas production and accumulation analysis, provide the following information:

1.

Specify the amount of hydrogen assumed to be in the primary coolant; l

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2.

Provide a table of the alurninum und zinc components in the (Cont.)

B-SAR-205 scope of supply, including a description of the components, and their weight, thickness, and exposed surface area.

3.

Figure 6.2-2 shows the total hydrogen generated by the corrosion of aluminum, Provide an accompanying analysis, and ~ justify the metal corrosion rates for the assumptions that were used.

4.

Specify the temperature and pressure used to calculate standard cubic feet.

022.14 Section 6.2.1 lists the criteria to assure that the structural (RSP) integrity of the containment will be maintained. Item 3 in this g

section implies that the only accident that must be considered is the LOCA. It is our position that containment and interior structures be designed to withstand the effects of a postulated main steam line and feedwater line accident.

Include these accidents in your criteria.

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UNITED STATES

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l'g NUCLEAR REGULATORY COMMISSION q

WASHINGTON, D. C. 20666 j/

AUG 2 71976

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Docket No. S'It150-561 Babcock & Wilcox Company ATIN:

Mr. Kenneth E. Suhrke Manager, Licensing Nuclear Power Generation P. O. Box 1260 Lynchburg, Virginia 24505 Gentlemen:

POUND 2 POSITIONS AND REhUEST EOR INEDRMATION As a result of our continuing review of.the Babcock & Wilcox Standard Safety Analysis' Report BSAR-205, your response to certain Regulatory staff positions and requests for information is required. 'Ibe specific information required is detailed in Enclosure 1.

Please note that the subject matter of the enclosed positions and requests in the 012.XX series i

was discussed with your staff on August 6,1976. at a meeting in our offices,

in Bethesda. Regulatory Positions are identified by (RSP) underneath the ___ _.

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position nuncer shown in Enclosure 1.

We are prepared to meet with you to discuss further any of our positions to assure emplete understanding of the factors at issue and the bases for our positions, however, we do not believe extended or iterative debate would be useful.

In order to maintain our licensing review schedule, we need your complete responses to the Enclosure 1 items by October 18, 1976. Please inform us within seven days after receipt of this letter of the date on which you plan to respond so that we may revise our schedule if necessary. If you plan to appeal to licensing management on any of these positions, please advise us of your intentions within two weeks.

Please contact us if you have any questions.

Sincerely, A lo./L Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management Positions and Request for Additional Information cc: see next page fx Q

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Babcock & Wilcox Company 2-

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cc: Washington Public Power Supply System

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ATTN:

Mr. J. J. Stein Managing Director j

P. O. Box 968 3000 George Washington Way Richland, Washington 99352 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox Nuclear Power Generation Division Suite 5515;'7735,01d Georgetown Road Bethesda, Maryland 20014 B. G. Shultz, Project Engineer Stone & Webster Engineering Corp.

P. O. Box 2325 Boston, Massachusetts 02107 Mr. A. H. Monteith Ohio Edison Company 47 North Main Street Akron, Ohio 44308 Mr. W. E. Kessler Commonwealth Associates, Inc.

209 East Washington Jackson, Michigan 49201 Robert J. Kafin, Esq.

115 Maple Street Glen Falls, New York 12801 l

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AUG 2 71976 j

( 3 ENCLOSURE 1

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POSITIONS AND REOUESTS FOR ADDITIONAL INFORMATION

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BSAR-205 l-DOCKET NO: STN 50-561

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010.0 AUXILIARY AND POWER CONVERSION SYSTEMS 012.43 We find that in several P&ID systems interface tables, the (General) references provided under the column, " Interface Requirements References," for each uniquely identified interface, are not all included in the " Interface Requirements Section " or the reference is incorrect. For example, on interface table figure i

10.0-1, sheet i for interfaces No. SP1 and SP2, the table references category 10.5.9, " System / Heat removal," sub-items 14 to 18 in section 10.5, " Interface Requirements." Sub-item 18 is not listed in section 10.5 under category 10.5.9.

On the same table for interfaces No. SP3 and SP4, the table references category 10.5.15,

" System / Component Arrangement," sub-items 1 to 4.

Sub-item 1 discusses the main steam isolation valves. The interface requirement on the' maximum closing time for the MSIV in the event of a steam

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line rupture to prevent blowdown of more than one steam generator is necessary. For the same interface Nos. SP3 and SP4 category 10.5.17, " Overpressure Protection," only sub-item I is referenced.

We believe sub-items 2 and 3 should also be referenced. Review and check al: P&ID's, interface tables, and Interface Requirements sections for inconsistencies, and make corrections where necessary.

012.44 Your response to our mquest 012.29 is not complete. Expand (9.1.5) section 9.1.5, " Interface Requirements " to provide interface criteria for safety related items specified by B&W and required by the B0P designer for safety related equipment installed in the spent fuel pool. Include for the spent fuel handling bridge crane, e.g., span length, rail loads from the crane, and electrical service required.

012.45 Your response to our request 012.30 is not complete. In addition (9.1.5) to information provided in section 9.1.5, " Interface Requirements,"

(RSP) category 8, component cooling and heating, provide the numerical values of the heat loads for the conditions stated in our request 012.30.

012.46(RSP)

Your response to our mquest 012.31 is not complete. Our position (9.1.5) is that a minimun water depth above the top of the spent fuel array must be listed as an interface requirement.

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~012.47 (RSP)

Your response to our request 012.32 is not complete.

It is our (9.1.5 )

position that load input to B0P structures and the locations of these loads relative to attachment interfaces on the new and spent fuel racks must be provided.

012.ap (RSP)

Your response to our request 012.36 is not acceptable. You must (5.2<2, demonstrate that decay heat loads are calculated in accordance with BTP APCSB 9-2.

012.49(RSP)

Section 9.3.1, compressed air requirements for the various NSSS (9.3.1) systems " Interface Requirements" sections do not provide specific compressed air quantities, i.e., cfm for safety related air operated

. valves which are part of the NSSS design secpe.

It is our position that the quantity of compressed air must be supplied for all safety related air operated valves in the NSSS design scope. The interface requirement may be included in the " Interface Raquirements" section at the end of each system or they may be tabluated in section 9.3.1 listing valve identification number, P&ID number, and air requirements.

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t 122.0 MATERIALS ENGINEERING - METALLURGY 122.9 Branch Technical Position MTEB 5-1 requires production testing of (Reg.

austenitic stainless steel welds to verify that adequate delta Guide ferrite levels are present. Your response to request No. 122.4 is 1.31) inadequate to justify no production testing. The following procedure

  1. 6 's an acceptable alternative to production testing of austenitic

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h Prior to production usage, the delta ferrite content of each lot and heat of weld filler metal procured should be verified for each pro-cess to be used in production. This delta ferrite verification should be made by determinations on undiluted weld deposits using magnetic measuring devices. For submerged arc welding processes, the verification tests for each wire-flux combination may be made on a production weld. All other delta ferrite measurements should be made on weld pads which will provide an undiluted layer of weld metal. The " Welding Research Council Recommended Procedure for Pad Preparation and Ferrite Measurement of Covered Electrode Deposits,"

i is considered acceptable for use.

The undiluted weld deposit should show an average Ferrite Number from 5 to 20.

The upper limit of 20 may be waived for:

(a) welds which do not receive post veld stress-relief heat treatment, or when such post weld stress-relief treat-ment is conducted at temperatures less than 900*F, (b) welds which are given a solution annealing heat treatment, and (c) for single pass welds which employ consumable inserts.

It is our position that

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you should provide your commitment to the alternative procedure described, or to the BTP MrEB 5-1.

122.10 he lack of testing requirements needed to determine the suscepti-(Reg bility to intergranular corrosion of unstabilized austenitic Guide stainless steel is not acceptable. Our position is that nonsensi-1.44 tization of unstabilized austenitic stainless steel should be veri-fdh) fied using ASTM A 262-70, " Recommended Practices for Detecting Susceptibility to Intergranular Attack in Stainless Steel" Practices A or E, or another method that can be demonstrated to show nonsensi-tization in austenitic stainless steel.

122.11 The nondestructive examination of tubular products performed in (Reg.

accordance with the requirements of Section III of the ASME Code, Guide Summer 1974 Addenda is acceptable. However, in order to eliminate 1.66 any misinterpretation as to the requirements for the number of

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directions of ultrasonic testing the Winter 1975 Addenda of the ASME Code should be specified.

122.12 Provide a list of all ME Class 2 components within BSAR-205 scope p \\ \\g that have thicknesses uceeding 2-1/2 inchee.

Identify the mater-ials of consttuction and indicate the technical basis for not im-g posing fracture toughness requirements on those components.

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ANALYSIS _

We have determined that the MOODY correlation as used in the CRAFT 222.10 code for calculation of critical flow from the primary system for i

sub-compartment analysis is not conservative for subcooled flow Accordingly, provide mass and energy release data and conditions.

the noding sensitivity study discussed in our Question 222.1 using an acceptable critical flow correlation such as the modified Zaloudek correlation in the CRAFT-2 code. Note that the Zaloudek correlation is based on stagnation properties at the break. The safeguards power level of 3876 MWt should be used in these analyses.

Your response to Question 222.9 parts (1) and (4) concerning the 222.11 (RSP) assumptions for calculating the mass and energy release to the containment following a postulated main steam line break is We require that you submit and justify all input s.

inadequate.

O assumptions for steam separation and heat transfer for the steam Your discussions should include the following generators.

considerations:

(1)

Provide and justify by comparison with an appropriate correlation the heat transfer coefficient from the secondary liquid to the tube walls.

(2)

Provide and justify by comparison to an appropriate condensing heat transfer correlation the input value for heat transfer coefficient from the secondary steam to the tube walls.

(3)

Two options for steam separation are discussed on page 15B-8. Provide and justify the option and input values used in your analyses.

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.N Provide the mass and associated energy for steam in the steam line piping and turbine plant that would be added to the containment 4

following a steam line break asstaning failure of one of the steam line isolation valves.

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Following i postulated steam line break, provide the closure time

.13 of the main feedwater isolation valves assumed in your calculation g

Provide the flow of mass and energy release to the contairnent.

rate and enthalpy of the main feedwater flowing into the ruptured steam generator before isolation.

The following questions relate to the methods described in Appendix 2.14 6A used to calculate mass and energy releases to tne containment following a postulated LOCA:

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On page 6A-10 you discuss the results of steam-water mixing (1) tests at Battelle Columbus Aero,1et Nuclear Corporation and

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Creare. Provide references for these test results.

(2) Provide a comparison of the 100 ft/sec entrainment limit that is used in your steam quenching calculation with the low steam flow data in EPRI report No. 294-2, " Mixing of Emergency Core Cooling Water With Steam: 1/3-Scale Test and Summary," June 1975.

(3) Pages 6A-21 through 6A-40 describe the REFLOOD code as used in containment mass and energy release calculations. Indicate which portions of the discussion represent changes to the REFLOOD code description in BAW-10093.

(4) On page 6A-34 you state that a heat transfer coefficient of 1000 BTU /Hr.-Sq.Ft.- F is used to calculate heat flow from the steam generator tubes to the primary coolant. On page 6A-35 you state that the hcat transfer coefficient is 5000 BTU /Hr.-Sq.Ft. OF

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222. 14 for hot leg breaks. Provide justification for these values by (Continued) emparison to an appropriate nucleate boiling heat transfer correlation and discuss the basis for selecting two different values.

(5) Provide justification for the assumption that film boiling occurs at a quality of 50% in the steam generator tubes by comparison to an appropriate critical heat flux correlation.

This asstanption is made on page 6A-35.

(6) Provide justification for the assumption that the flow split of ECCS water between the core and ruptured steam generators may be calculated from the area ratios. Justification should consist of flow calculations for the core and steam generator

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based on the pressure drops calculated by the REFLOOD code.

This assumption is made on page 6A-36.

(7) On page 6A-4 you state that the end of the reflooding period is detennined when themal equilibrium occurs between the primary and secondary systems and the core is quenched. For each break size and location analyzed in Appendix 6A, indicate the time when each of these conditions is reached.

(8) The long tenn post reflood mass and energy releases to the containment are calculated using the CONTEMPT code. Provide-a description of all differences between the code version used in these analyses and the version described in BAW-10095A.

(9) In part 2 of our question 222.5 we requested a sensitivity study showing the effect of the containment pressure on the mass and energy release calculated during the reflooding period. Provide l

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. 2.14 justification for the value of entrainment threshold velocity (Continued) selected for use with each containment pressure in this study.

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J.0 EFFLUENT TREATENT SYSTEMS 320.9 Your msponse to Question 320.7 states that expected releases from (11.1.5) the boron recovery system am accomplished by bleeding to the i

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liquid waste system a portion of the recovered distillate gW downstream of the distillate test tanks. It is our position that i

you must provide the expected and maximum volumes, radioactive concentrations, and chemical content of this stream as interface infonnation to the B0P designer.

320.10 In Amendment 2, you state that the airborne tritium concentration (11.1.2.2) is based on a 175 lb/hr evaporation rate from the refueling canal during the 14-day refueling periods. Provide the bases for this rate, preferably citing operating experience.

to\\h Provide the bases for the tritium concentration given in Figures 11.1-1 thru 11.1-5, including assumed mixing between reactor water and fuel pool during refueling. Cite previous pertinent experience from operating reactors. You should also consider the f

use of Li-OH containing less than 99.99% lithium.

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RADIOLOGICAL ASSESSMENT i

331.11 Your msponse to question 331.9 is incomplete. In order to (RSP) assure that occupational radiation exposures will be maintained as low as is reasonably achievable, the buildup and transport of activated corrosion products in the reactor coolant system should be minimized.

Inconel steam generator tubing material is generally available at less than 0.1 waight percent cobalt, and several comonly used austenitic steelt used for reactor coolant system materials are generally available at less than 0.1 weight percent.

It is our position that the B&W specification of 0.2 weight percent residual cobalt content of material in contact with reactor coolant should apply selectively only to that Inconel and steel for which T

lower cobalt bearing alloys are not generally available. Provide a

\\g detailed breakdown for materials which have a greater than 0.1 w/o residual cobalt content, and are in contact with reactor coolant, including steam generator tubing, major pressure vessels, coolant piping, core structural material, pressurizer heaters, and valve and pumps materials in contact with coolant. Provide a justification for their use including an estimate of the surface area exposed to the coolant and reason for selection. Describe the expected exposure reduction benefit from using stellite instead of other alloys which may increase maintenance or coolant leakage. Describe an example of the process used to detemine that radiation levels and radiation exposure will be minimized by use of the stellite materials.

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Section," or the r,dference is incorrect. For example, on Interface table figure 10.0-1, sheet 1 for interfaces No. SPI and SP2 the table references category 10.5.9, " System / Heat removal "

sub-itemhl4to187'Tnsection10.5,"InterfaceRequirements."

Sub-item 18 is not listed in section 10.5 under category 10.5.9.

On the same table for interfaces No. SP3 and SP4, the table referen Les categ3ry 10.5.15 " System / Component Arrangement,"

sub-items (1 to 4f. Sub-item 1 discusses the main steam isolation valves. The interface requirement on the maximum closing time for the MSIV in the event of a steam line rupture to prevent blowdown of more than one steam generator is necessary.

For the same interface Nos. (SP3 and SP4) category 10.5.17, " Overpressure Protection," only sub-item 1 is referenced. We believe sub-items 2 l

and 3 should also be referenced Review and check all P&ID's, l

interface tables, and Interfacehquirements sections for in-consistencies, and make corrections where necessary.

.g 012.44 Your response to our request 012.29 is not co, ete.

Expand (9.1.5) section 9.1.5, " Interface Requirements," t rovide interface criteria for safety related items c by B&W and required g

by the B0P designer for safety related equipment installed i - /L i

in the spent fuel pool.

Include for the spent fuel handling Adge crane S, span length, rail loads from the crane, 4.9.;

and electrical service / required.

012.45 Your response to our request 012.30 is not complete.

In addition (9.t.5) to information provided in section 9.1.5, " Interface Requirements,"

(RSP) category 8, component cooling and heating, provide the numerical values of the heat loads for the conditions stated in our request 012.30.

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012.46 (RSP)

Your resp 1se to our request 012.31 is not complete. Our position (9.1.5) is that minimum water depth W. be listed as an interfar.e requirement.

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012.47 (RSP) Your response to our request 012.32 is not complete.

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(9.1.5) position that load input to BOP structures and the locations of f

these loads relative to attachment interfaces on the new and spent fuel racks must be provided.

pg 012.48(RSP) Your response to our request 012.36 is not acceptable. Ap ance with BTP APCSB 9-2.

4-(9.2.2) decay heat lo ga 012.49 (RSP) Section 9.3.1, compressed air requirements for the vari (9. 3.1 )

systems " Interface Requirements" sections do not de specific compressed air quantities, i.e., cfm for saf +

related air operated It is our valves which are part of the NSSS ::: : - '

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position that the quantity of compressed air must be supplied for all safety related air operated valves in the NSSW designa.

The interface requirement may be included in the "IWn er ace Require-L ments" section at the end of each system or they may be tabulated in section 9.3.1 listing valve identification number, P&ID number, and air requirements.

012.50 In your meeting wi

'tlie staff on August 6,1976, yeti indicated ireProtectiop[dedinAmendmenyto.2under

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thp.t' interface i ormation provi included an grror in the

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ection 9.5.1 specified o' volume in thyreactor coolanVpumps. Pro e the f correct ormation.

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C We have detennined that the MOODY correlation as used in the CRAFT I

AM' 8 code for calculation of critical flow from the primary system for sub-compartment analysis is not conservative for subcooled flow 1

conditions. Accordingly, provide mass and energy release data and the noding sensitivity study discussed in our Question 222.1 using an acceptable critical flow correlation such as the modified Zaloudek correlation in the CRAFT-2 code. Note that the Zaloudek correlation is based on stagnation properties at the break. The safeguards power level of 3876 MWt should be used in these analyses.

i 42a..ll W Your response to Question 222.9 parts (1) and (4) concerning the SN assumptions for calculating the mass and energy release to the containment following a postulated main steam line break is inadequate. We require that you submit and justify all input assumptions for steam separation and heat transfer for the steam genera tors.

Your discussions should include the following considerations:

a.) Provide and justify by comparison with an appropriate correlation the heat transfer coefficient from the secondary liquid to the tube walls.

b.) Provide and justify by comparison to an appropriate l

condensing heat transfer correlation the input value for heat transfer coefficient from the secondary steam to the l

tube walls, c.) Two options for steam separation are discussed on page s.

158-8. Provide and justify the option and input values used in your analyses.

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i 222,r2. $ Provide the mass and associated energy for steam in the steam line l

piping and turbine plant that would be added to the contairunent i

following a steam line break assuming failure of one of the steam line isolation valves.

.222. l'b 4$ Following a postulated steam line break, provide the clo'sure time of the main feedwater isolation valves assumed in your calculation of mass and energy release to the containment. Provide the flow rate and enthalpy of the main feedwater flowing into the ruptured steam generator before isolation.

.2n,14 5$ The following questions relate to the methods described in Appendix 6A used to csiculate mass and energy releases to the containment following a postulated LOCA:

a.) On page 6A-10 you discuss the results of steam-water mixing tests at Battelle Columbus, Aerojet Nuclear Corporation and Crea c.

Provide references for these test results, b.) Provide a comparison of the 100 ft/sec entrainment limit that is used in your steam quenching calculation with the low steam flow data in EPRI report No. 294-2, " Mixing of Emergency Core Cooling Water With Steam:

1/3-Scale Test and Summary," June 1975.

c.)

Pages 6A-21 through 6A-40 describe the REFLOOD code as used in containment mass and energy release calculations. Indicate which portions of the discussion represent changes to the REFLOOD code description in BAW-10093.

d.) On page 6A-34 you state that a heat transfer coefficient of 1000 BTU /Hr.-Sq.Ft.- F is used to calculate heat flow from the steam generator tubes to the primary coolant. On page 6A-35 you

. state that the heat transfer coefficient is 5000 BTU /Hr.-Sq.Ft. or

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i AA2 'T for hot leg breaks. Provide justification for these values by (c&s~ 4) comparison to an appropriate nucleate boiling heat transfer i

correlation and discuss the basis for selecting two different values.

~

e.) Provide justification for the assumption that film boiling occurs at a quality of 50% in the steam generator tubes by comparison to an appropriate critical heat flux correlation.

This assumption is made on page 6A-35.

f.) Provide justification for the assumption that the flow split of ECCS water between the core and ruptured steam generators may be calculated from the area ratios. Justification should consist of flow calculations for the core and steam generator based on the pressure drops calculated by the REFLOOD code.

This assumption is made on page 6A-36.

g.) On page 6A-4 you state that the end of the reflooding period is determined when thermal equilibrium occurs between the primary and secondary systems and the core is quenched. For each break size and location analyzed in Appendix 6A, indicate the time when each of these conditions is reached.

h.) The long term post reflood mass and energy releases to the containment are calculated using the CONTEMPT code. Provide a description of all differences between the code version used in these analyses and the version described in BAW-10095A.

i.) In part 2 of our question 222.5 we requested a sensitivity study showing the effect of the containment pressure on the mass and i

x.

energy release calculated during the reflooding period.

Provide

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I a a.a. if (tMe-4) justification for the value of entrainment threshold velocity l

. selected for use with each containment pressure in this study.

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t-320.0 EFFLUEhT TREATMENT SYSTEMS V

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Your response to Question 320.7 states'that expected n ases 320.10 (11.1.5) from the boron recovery system are accomp1 shed by bleeding (R$p) to the liquid waste system a portion of t,, recovered distillate downstream of the distillate test tanks.,rrovide the expected and maximum volumes, radioactive concentrations, and chemical content of this stream as interface information to the BOP designer.

320.11 In Amendment 2, you state that the airborne tritium concentra-(11.1. 2. 2) tion is based on a 175 lb/hr evaporation rate from the re-fueling canal during the 14-day refueling periods. Provide p

the bases for this rate preferably citing operating experience.

4 3

i Provide the bases for the tritium concentration given in Figures 11.1-1 thru 11.1-5, including assumed mixing between reactor water'and fuel pool during refueling. Cite previous pertinent experience from operating reactors. You should also consider the use of Li-0H containing less than 99.99%

lithlum.

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33I.O fhowtosteAL AE N 33 L.11 Your response to question 331.9 is incomplete. In order

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. to. assure that occupational radiation exposures will be maintained 'as low as is ' reasonably achievable, the build-up and transport of activated corrosion products in the reactor coolant syste= chould be minimized.

Inconel steam generator tubing material is generally available at less than 0.1 weight percent cobalt, and several commonly uded austenitic steels used for reactor coolant system caterials are generally availabic at less than 0.1 weight percent.

L_..

'- : the B & W specification of 0.2 gqightperc'entresidualcobaltcontentofmaterialincontact

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with primary coolant should apply selectively only to that Ideonel and steel for which lowet cobalt bearing albys are nSt generally available. Provide a detailed breakdown _

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0 N 8*

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1. E tc.1; M for materials [n 4 o4 Wg M cbntact with reactor coolant, including steam generator tubing, major pressure vessels, coolant piping, core

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a structural material, pressurizer heaters, and valve and pumps materials in contact with coolant.

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_ Srovide a justification for

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their use including an estimate of the surf ace area exposed to the coolant and reason for selection.

Describe the c::pected exposure reduction benefit from using stellite I4s/eedof L

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.:-. --_; other alloys which may increase maintenance or coolant leakage.

Describe an example of the process used to dc.termine that radiation 1cvels and radiation expcsure will be minimi=ed by use of the n gaterials.

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t Docket No. S7tl 50-561-um Babcock & Wilcox Company

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ATnt:

Mr. Kenneth E. Suhrke Manager, Licensing Nuclear Power Generation P. O. Box 1260 Lynchburg, Virginia 24505 Gentlemen:

.< p IOUND 2 IOSITIONS NJD REQUEST FOR INFORMATION As a result of our continuing review of the Babcock & W leox Stm.c u.

Safety Analysis Report BSAR-205, your response to certain Regulatory j

staf f positions and requests for information is required. The specific information required is detailed in Enclosure 1.

Pleas-

% da ripct mat-ter--of4 m 1md-posi-tJene-end--r~;ue'- i n '

"'c-@m. a are:::dincu x0 i" yc,m -owk-v.eAupA 6,-19%r-et-e.c.eeting=irm-office

.m Dethmsde:. Regulatory Positions are identified by (ESP) underneath the j 'p position nunber shown in Enclosure 1.

We are prepared to meet with you to discuss further any of our positions to assure complete understanding i ll of the factors at issue and the bases for our positions; however, we do I<

not believe extended or iterative debate would be useful.

In order to maintain our licensing review schedule, we need your complete responses to the Enclosure 1 items by October 18, 1976. Please inform us e

within seven days after receipt of this letter of the date on which you i,a plan to respond so that we may revise our schedule if necessary.

If you 8

plan to appeal to licensing management on any of these positions, please 4

,d advise us of your intentions within two weeks.

a I

Please contact us if you have any questions.

4 l

Sincerely, l

1 t

h Olan D. Parr, Chief Light Water Reactors Branch No. 3 j

Division of Project Management Enclosute 1 j

Positions and Request for Additional Information cc; see next page F

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A Babcock & Wilcox Company -

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cc: Washington Public Power Supply System ATIN:

Mr. J. J. Stein Managing Director P. 0. Box 968 3000 George Washington Way Richland, Washington 99352 i ['

Mr. Robert Borsum I--

Bethesda Representative

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Q Babcock & Wilcox Nuclear Power Generation Division Suite 5515;"7735,0ld Georgetown Road Bethesda, Maryland 20014 B. G. Shultz, Project Engineer Stone & Webster Engineering Corp.

P. O. Box 2325 l

Boston, Massachusetts 02107 h

Mr. A. H. Monteith Ohio Edison Company p

J 47 North Main Street Akron, Ohio 44308 I

Mr. W. E. Kessler Connonwealth Associates, Inc.

209 East Hashington

.i!,.

Jackson, Michighn 49201 Robert J. Kafin, Esq.

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,;7 115 Maple Street m

Glen Falls, New York 12801 h

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ENCLOSURE l_

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POSITIONS AND REOUESTS FOR ADDITIONAL INFORMATION g

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l BSAR-205 DOCKET NO: STN 50-561 (Ph 110.0 4 DICAL I

110.38 The response to Question 110.18 is not completely acceptable. The (3.6.3.1) statement that subsection NF of the ASME Code will be the basis t

(110.18) for operability tests is unclear, since subsection NF addresses only structural integrity. The staffs primary concern regarding snubbers is that the structural aspects of snubber utilization is_ fully evaluated._ We will require that the following information concerning snubbers utilized in B-SAR-205 systems be provided in the FSAR.

.l 1.

Snubber design specifications.

t 2.

Description of snubber suppliers performance qualification tests and load tests.

3.

System and component structural ~ analysis showing:

a.

Structural analysis model f'-

b.

Description of the characterization of bydraulic snubber

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mechahical'p'roperties used in the structural analysis

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including considerations such as(i) diff erences in tension and compression spring rates, (ii) effect of entrapped air and temperature on fluid properties, (iii) other factors affecting snubber characteristics.

c.

List load conditions and transients analyzed.

d.

Maximum snubber loads, corresponding piping or component stresses.

e.

Comparison of computed loads and stresses with rated snubber load and stress intensity limits.

4.

Discuss design provisions for a:cessibility for inservice

-inspection and possible removal for operability testing and repair or replacement of snubbers.

Provide a commitment to include all of the above information in the B-SAR-205 FSAR.

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110.39 (RSP)

The informatimiin Section 3.9.1.1, " Vibration Operational Test (3.9.1.1)

Program".is not completely acceptable.

In addition to the information presented, it is the staff's position that in the PSAR, a commitment should be made to conduct preoperational piping vibrational and dynamic effects testing in accordance with NRC Standard Review Plan, Section 3.9.2 on all of the following classes of piping systems if they are within the B-S_AR-205 scope of design:

1.

All ASME Class 1, 2 and 3 piping systems.

2.

All high energy piping systems outside containment.

3.

All Seismic Category 1 portions of moderate energy piping l

systems outside containment.

7 I

110.40 (RSP)

Questions 110.19, 110.20 and 110.24 are related to seismic (3.9.1, qualification of mechanical and electrical equipment and 3.9.2, instrumentation. Respontes to these questions in amendment 2 3.10) to the B-SAR-205 PSAR are not entirely in accord with current (110.19, NRC requirements. The responses cite seve'ral documents, IEEE-344-

~

,110.20) 71, IEEE-382-72, and topical report BAW-10082, which are not c'

21, currently acceptable references.

. 4) lt. is recognized that much B & W designed equipment has been

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previously qudlified by methods and to standards which may or may not be acceptable by current standards. The principal items of concern are those vital appurtenances necessary for the actuation and continued operation of safety related pumps and valves during accident conditions. The operability of such appurtenances as electrical switching gear, motors and valve operators subjected to complex oscillatory motions is dif ficult to verify analytically. Consequently, some such components may have to be retested to current seismic; qualification standards

'6 specified in section 3.10 of the NRC standard review plan.

Our position is that you should provide, in the BSAR 205 document, i

I one of the following:

I i

1.

revised sections 3.9.1, 3.9.2 and 3.10 of the BSAR-205 to be entirely consistent with the corresponding sections of the NRC Standard Review Plan or L

2. A 9 5 4 your commitment to conform to the final generic resolution by the NRC seismic qualification task group and B & W of out-standing questions pertaining to the seismic qualification of B & W equipment.

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lO 110.41 The response,to Question 110.24 in. Amendment 2 to the PSAR is (3.9.2.4) not completely acceptable as a response to Question 110.21.

,It is the staffs' position that the manufacturers of all ASME Class 1, 2 and 3 active pumps and valves aust be required to demonstrate that the pump or valve will operate normally when subjected to all 8

loads and other environmental conditions associated with a faulted condition. These loading conditions should be clearly defined to the pump or valve manufacturer. Provide a. specific commitment to this position in Section 3.9.2.4 of the PSAR.

110.42 It is the staffs' position that the applicant should verify that (3.9.2) all ASME Class 1, 2 and 3 systems, components and supports which (5.2.1) are required (1) to assure safe shutdown of the plant or (2) to prevent or mitigate the consequences of an accident and which are designed to emergency and faulted condition stress limits will i

function as designed..The stress limits listed in the various j

sections of the PSAR assures the structural integrity of the j

various systems, but does not necessarily guarantee their functional integrity. Provide your commitment in the PSAR that the stress limits used in the design of all of the above noted systems,.

components and supports will not result in inelastic deformations

_which would prevent the normal operation or function of the item.

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STRUCTURAL ENGINEERING l.

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Your response to request number 132.22 concerning decoupling l

(3.7.2 r

RSP) criteria for subsystem is incomplete.

It is the staff's position that the decoupling criteria for subsystem should be expressed in tems of mass ratio and frequency ratio, as stipulated in NRC Standard Review Plan 3.7.2-II,3.b.

Provide appropriate commitments to the SRP critoria, or describe and justify any exceptions taken I

to the SRP cM teria, i

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130. P Provide a commitn}ent to review the BOP design infomation and drawings i.

i (3.8 RSP ush I

Interface) for compliance your QNSSS design criteria.

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i 211.18 In your response to 211.13, valves V4, V2A and V2B on Figure 9.3-1, sheet 2 are apparently incorrectly classified as Quality Group C j

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components. Please correct as necessary.

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231 30 The response to 1st-round question 231.1 is incomplete because (4.2.1.1) it does not provide the requested information on experimental confirmation of the fuel handling and shipping design loads.

Please describe the extent to which these design loads have been confirmed experimentally.

231.31 The response to 1st-round question 231.1 is incomplete with (4.2.1.1) respect to the discussion of fretting, wear, and deflection.

Please cite the current design limits for these phenomena, outline the on-going or planned R&D programs which should yield confirmatory information on the specific design limits, and present fall-back positions. Discuss how deflection is accounted for in the summation of stresses in the fuel assembly (as suggested in the response to question 231.2).

231 32 The response to 1st-round question 231.4 lacks detail. Please (4.2.1.2) describe how the specified coolant temperature limits and associated cladding loading are used in the fuel rod fatigue analysis. Show by means of an example how the coolant tem-

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perature limits and associated cladding loading are used to

" identify the conservative conditions for input to the stress analysis," as asserted in the response to question 231.4.

231.33 The response to 1st-round question 231.5 requires amplification (4.2.1.2) regarding (1) the " conservative models" said to be used for rod differential growth and grid pressure drop and (2) the out-of-reactor flow tests and measurements which reportedly confirm the calculations that show that grid position is well-maintained throughout life. Please show in greater detail how these calculations and experiments provide support for the conclusion that the frictional force on the fuel rods is sufficient to mair.tain grid position throughout life.

231.34 The response to 1st-round question 231.6 does not provide the (4.2.1.2) requested information on dimensions, spring constants, and.

experimental observations of the upper and lower plenum springs. Please provide the requested information and, in addition, show quantitatively that the resistance to creep l

and relaxation of age-ha ened A-286 alloy is sufficient to i

withstand the worst post ted flux, temperature, and ' stress g

conditions, as asserted the 1st-round response.

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231.35 The response'to 1st-round question 241.7 does not provide the (4.2.1.3) request.ed design bases for Zirealoy-4 irradiation growth.

j Design " bases" are not synonymous with " values," as appears to be implied by the response. Please provide the design bases as requested, and briefly outline the data which sup-i port these bases.

231.36 The response to 1st-round question 241.8 requires clarifica-

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(4.2.1.3) tion because of an apparent confusion of terminology. The response appears to treat cladding strain and fuel rod de-flection as if they were synonymous. An intent of lat-round J

question 241.8, however, was to establish the displacement

[

limit of B&W fuel rods from a rod bowing viewpoint. Such a j

displacement limitation, when used in fuel design, should reflect a DNB correlation and power peaking factor calcula-tion. Provide the as-manufactured displacement limitation as well as the one imposed during operation. Discuss how one i

confirms that these limitations are not exceeded..

231.37 The response to 1st-round question 231.11 does not provide the (None) requested information on the currently used stiffness limita-tions on the spacer grid assembly and individual grid springs.

In addition to providing this requested information, please outline how the results of specific portions of the Mark C fuel assembly development program will be used to provide the s-information requested in 1st-round question 231.11.

t 231.38 The response to 1st-round question 231.12 requires amplifica-(4.2.1.3) tion regarding the procedure for limiting the recommended power startup rate in the 0-20% power range.

Please quantify this recommended limit in power startup rate and provide ex-perimental quantitative verification of the effect of reduction in power startup rate on defect propagation.

231.39 The response to 1st-round question 231.18 addresses the 15 strain (4.2.1.3) limit which is based on average cladding strain. The R-2 re-actor power ramp tests, referred to in the response, were, I

however, performed on low exposure rods which were still ductile and, therefore, only demonstrated the ability of the rods to withstand pure mechanical loading. Describe any research pro-grams on analytical modeling development currently in progress or planned to evaluate the effects of local cladding strain due to pellet cracking on ridging, cumulative damage, and stress corrosion cracking.

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s 231.40 The response'to lat-round question 231.21 indicated that in I

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(4.2.3.2) experiments where irradiated A1 0 -B C was exposed to high-23 4 temperature high-pressure water..the B C reacted with the 4

water to form H3B0. Thus, if the poison rod cladding were 3

perforated, the H B03 would be leached into the coolant.

3 Please discuss the potential safety implications of the re-activity insertion resulting from the loss of B-10 from the burnable poison rods by this mechanism. Describe the re-activity anomaly that would result if all the B C were 4

removed from (a) one rod and (b) all the poison rods early in life.

Provide rate equations for the hydrolysis of B C 4

and rate of loss from perforated rods, and calculate these' rates at (a) reactor coolant temperature and (b) local r

poison pellet temperature.

231.41 The response to 1st-round question 231.17 on fuel rod bowing (4.2.1 3) refers to examination measurements on the Ocones

  • Mark B (15x15) assemblies which will be used as a basis for pre-dicting bowing in the Mark C-(17x17) assemblies. Please discuss how the bowing data from 15x15 Mark B assemblies will used for 17x17 Mark C, bowing predictions; i.e. how will 15x15 Mark B assembly data be related and applied to the 17x17 fuel?

Also provide the following information:

Status of the 15x'15 rod bowing data collection; (a)

(b) Schedule and scope of the 15x15 examination program; (c) Manner by which the 15x15-data and analysis will be reported to NRC, and approximate date for submittal of a topical report; (d) Plans for obtaining 17x17 fuel assembly bowing data; (e) Out-of-pile (if any) mechanical experiments which will provide input to a mechanistic bowinC model.

231.42 The treatment of the seismic and LOCA analyses for the Mark C (None)

(17x17) fuel assembly is inadequate. An in-depth safety analysis of the seismic and LOCA response of the Mark C (17x17) fuel assembly has been requested (letter, Ross to Schwencer, July 25, 1974) and a commitment to submittal of a topical report in early 1976 (at least one year prior to the filing of the first FSAR ingorporating the Mark C fuel assembly) was made by B&W (le.tter, MallD to

)(

Schwencer. September 3, 1974). Our evaluation of the B&W seismic and LOCA analyses for the Mark C assembly cannot be completed until the requested report has been received.

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232.26 The response to Question 232.17 is inadequate. Please (15.1.2) identify the 205 FA plant for which the analysis was l

performed.

232.27 The response to Question 232.11 -(as presented in the I

i response to Question 212.71) implies that power shapes I

with "large" negative offsets were used in the deriva-tion of the power range scram reactivity curve. Please i

confirm and indicate the range of negative offsets con-sidered. In particular was consideration given to a scram while in the recovery from a load following

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311.0 ACCIDENT ANALYSIS.c i

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311.6 (RSP) Respond to the following requests concerning inquir es 211.13 and 211.14.

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1.

Your response to 211.13 indicates that y'op have i

conducted radiological accident analyses to determine theseismicdesignclassificationoffedeborating demineralizer. The details of this/an.alysis should be provided including justificatien of the assumptions used.

Include an explanation o 'the assumptions associated with the 100jf~u,in e period prior to V

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system isolation.

A 2.

Provide similar informat on pertaining to the reactor coolant degasifier (21.14).

i 3

ummarize all simil analyses, if any, used to j'ostify component esign classification as non-se1\\mic Category indicates a limiting x/Q of 4.

The pgse 3.211.13 7.0 x 10

/m We do not agree that the x/Q is h be conservat e for anticipated Babcock-205 sites.

pf,I7 The staf hasdocumenged,inWASH1361,thatthe

.0 (1b-4) s/m id a limiting value for x/Q of

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only e half o the large number of sites examined.

Our osition is t t you must provide explicit

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identification of the limiting x/Q as an interface re,quirement on the BOP supplier.

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SEP 0 21976

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Project tio. P-566 lEMORANDU*i FOR:

K. V,niel, Chief Light Water Reactors Branch f2, DPM

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FROM:

D. F. Bunch, Chief. Accident alysis Branch, DSE SUDJECT:

B-SAR '05 Q-2 INPUT PLAliT NAME: Babcock & W Icox (B-SAR-205)

LICENSING STAGE: PDA (CP PROJECT NULL 3ER: P-566 MILESTONE liUMiiER: 12-31 RESPONSIBLE BRANCil: LUR #2e T. Cox, Pr1 REQUESTED COMPLETION DATE:

ugust

,1976 REVIEW STATUS: AAB Q-2 revie' com etc Ti.e Accident Analysis Branch h reviewed the responses to the Q-1 f.

inquiries and have determined at the applicant has supplied sufficient infomation for purposes of ndependent dose calculations.

As indicated in R. Kirkwood s merr to T. Cox (08/27/76) the AAC has agreed to follow up on th nadeq te responses to inquiries 211.13 and suggests.T. ore appropriate,s ion 311.y requests additional information and 211.14 The attached que accident i.wtcorology.

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W D. F.

unch, Chief Accide Analysis Branch Divisioig of Site Safety and Enviro ental Analysis

Enclosure:

Q-2 Input on BJ5AR-235

/

Distributioni S. Hanauer W. Mcdonald (w/o encl.)

K. Murphy H. Denton J. Panzarella (w/o encl.)

Socket File D. Muller P. Shuttleworth (w/o encl.)

NRR Rdg. File F. Miraglia D. Bunch DSE Rdg. File J. Miller G. Chipman AAB,Rdg. File S. Varga D. Vassallo AAB\\ Files l

R. Vollmer T. Cox M DS AAB

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.jdYdpma[g. unch 0 - t -76 09 L-76 09-N 6 yenn ABC.)ls (Rev. 9 5 5) AICM 0240

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Respond to the following requests concerning inquiries 211.13 and 211.14:

1.

Your response to 211.13 indicates that you have conducted radiological accident analyses to determine the seismic design classification of the deborating demineralizer.

The details of this analysis should be provided includingM^ "6 A justiff _ a ion of the assumptions used %4J- %N

& 4 s** 6 M W W + yts l**l*M.

wmhm m 2.

Provide similar information pertaining to the reactor coolant degasifier (211.14).

3.

Sumarize [s i r analyses, if any, used to justify ssification 'c= th2r AS"E E0:ti= !!!,- d HA 98ca

componen J.

bh us the e ana

. The response to 211.13 indicates a limiting X/Q of

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7.0 x 10-4 s/m3 We do not agree that the X/Q is h

' f:c;t conservative for anticipated Babcock-205 sites.*

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400.3 - 9 At a meeting with the staff on August 6, 1976, your Amendment 2 t

response to request 211.2 was discussed and certain commitments were made by B&W regarding additional information concerning the staff position and request for information embodied in 211.2, as well as several other numbered requests.

A sn= nary of the August 6,1976 meeting is enclosed for reference.

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Your adequate response to these issues within the time allotted for Round 2 responses is required to complete the BSAR-205 review effort within the review schedule presented in our letter of April 22, 1976 to Mr. K. Suhrke.

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