ML20213D019

Forwards Request for Addl Info Required to Complete First Round Questions for Evaluation of Licensees FSAR

ML20213D019
Person / Time
Site:	Columbia
Issue date:	07/20/1979
From:	Pawlicki S Office of Nuclear Reactor Regulation
To:	Rubenstein L Office of Nuclear Reactor Regulation
References
CON-WNP-0282, CON-WNP-282 NUDOCS 7908240612
Download: ML20213D019 (7)

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Docket No.450-397 MS 12-12 MEMORANDUM FOR:

L. Rubenstein. Acting Chief Light Water Reactors Branch No. 4 Division of Project Management FROM:

S. S. Pawlicki. Chief Matarials Engineering Branch 4

SUBJECT:

WPPSS !#! CLEAR PROJECT NO. 2 (BWR/5)

Plant Name: WNP-2 Suppliers: General Electric; Burns and Roe Licensing Stage: OL Docket Number: 50-397 Responsible Branch and Project Manager: LWR 4; M. D. Lynch Reviewer:

M. L. Boyle Description of Task: Q-2 Review Status: Additional Infomation Requested The Materialt Integrity Section of the Materials Engineering Branch, Division of Systems Safety, has reviewed the FSAR for WHP-2 through Amendment 3, including the preliminary response to our Q-l's. Based on our review of this information, we have identified additional infonnation (see attacivnent) that must be addressed in the WNP-2 FSAR before our evaluation may be completed.

Due to the incomplete response to our Q-l's, we may need a supplemental Q-2.

S. S. Pawlicki, Chief Materials Engineering Branch Division of Systems Safety Office of Nuclear Reactor Regulation

Enclosure:

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121-1 i

121.0 MATERIALS ENGINEERING BRANCH - MATERIALS INTEGRITY SECTION 121.11 The response to Question 121.2 is incomplete. Appendix G, 10 CFR (5.2.3)

Part 50 explicity specifies minimum fracture toughness requirements for all ferritic materials of pressure-retaining components of the reactor coolant pressure boundary, not just the reactor vessel beltline materials. Therefore, provide the unfrradiated mechanical properties (as described in Question 121.2) for all materials required by Appendix G, 10 CFR Part 50. Also provide the minimum thickness of the reactor vessel shell material.

121.12 Sections 5.2.3.3.1.1 and 5.3.1.5.1.4 of the WNP-2 FSAR indicate that (5.2.3) the minimum metal temperature for significant pressurization shall '

(5.3.1) be RT 60*F, where RT react $kv+essel beltline r$hIo=.40*F for areas remote from the n

Section 5.3.1.5.2.2, supplied in response to Question 121.3, indicates that the minimum metal temperature for pressurization is based upon an RT of 20*F the reactor vessel closure flange. Clarify this dNrepancy. for 121.13 In Section 5.3.1.5.2.2 of the WNP-2 FSAR, supplied in response to (5.3.1)

Question 121.3, it is stated that a factor of 2*F per ft-lb is used to convert longitudinal Charpy V-notch impact data obtained at the 30 ft-lb level to estimates of the values at the 50 ft-lb level.

It is stated that these estimates, plus a 30*F adjustment for specinen orientation, are based upon information tabulated in WRCB-217, " Properties of Heavy Section Nuclear Reactor Steels," and from other fracture toughness tests.

Explicitly state the procedures used to verify this factor, includ-ing a sample calculation, and any data other than that in WRCB-217 used as a basis for this estimate.

121.14 To demonstrate compliance with Appendix H to 10 CFR Part 50, (5.3.1) include in the WNP-2 FSAR and Technical Specifications a table that provides the following infonnation for each surveillance specimen j

capsule:

i (1) The actual surveillance materials in each capsule, (2) The beltline material from.which each surveillance material was obtained, (3) The test specimen type (s), and their orientation, for each surveillance material, (4) The actual location of each capsule in the reactor vessel, t

(5) The lead factor for each capsule calculated with respect to the k wall thickness location.

(6) The proposed loading schedule of the capsules into the WNP-2 reactor vessel, and

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121-2 (7) The proposed time of capsule withdrawal (calender years and effective full power years).

i 121.15 Table 5.3-la, supplied in response to question 121.3', indicates (5.3) that the welds, which were required to be impact tested according to paragraph II.C.2 of Appendix G,10 CFR Part 50, were not made from the same heat of base plate. Identify the base material used and confim that the weld wire, flux and weld procedure used to make weld metal test specimens are the same as those used to fabricate the reactor vessel beltline welds.

121.16 Appendix G,10 CFR Part 50, requires that RT be adjusted for (5.3.1) irradiation effects by adding to RT the t$herature shift in the Charpy V-notch curve for irradiNNd material relative to that for unirradiated material, measured at a specific energy or mils 2

lateral expansion level using test specimens oriented in the transverse direction.

In response to Question 121.3, the statement is made that, based on GE experience, the amount of shift in the RT measured on irradiated longitudinal test specimens will be UNentially the same as the shift in equivalent transverse specimens. Provide the Charpy V-notch impact data to demonstrate that the shift in RT NDT using longitudinally or transversely oriented specimens are equivalent. Also demonstrate that the decrease in upper-shelf energy caused by neutron irradiation can be equally measured by longitudinal or transverse specimens.

121.17 The response to Question 121.5 is inadequate.

It is our position (5.3)RSP that the construction and inspection of the reactor surveillance capsule attachments be done according to the requirements for permanent structural attachments to reactor vessels specified in i

ASME Code Sections III and XI.

121.18 The response to Question 121.7 is inadequate. The reports (10.2.3) referenced (Reference 10.2-1 and 10.2-2) are not on file with the l

NRC. Provide these reports so that a complete evaluation of the l

response to Question 121.7 can be m:de. Also identify the material specifications for the rotor shaft and discs of the turbine-generator.

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_s 0M L. Rubenstein cc w/ encl:

R. J. Mattson D. G. Eisenhut D. M. Crutchfield

.J. P. Knight S. S. Pawlicki R. J. Bosnak H. F. Conrad R. M. Gamble M. D. Lynch C. O. Thomas S. Miner G. B. Georgiev S. J. Bhatt J. M. Grant J. Halapatz M. R. Hum F. B. Litton C. D. Sallers M. L. Boyle cc w/o encl:

D. B. Vassallo W. J. Pike

Contact:

M. L. Boyle X27255 1

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121-1 i

121.0 MATERIALS ENGINEERING BRANCH - MATERIALS INTEGRITY SECTION 121.11 The response to Question 121.2 is incomplete. Appendix G, 10 CFR (5.2.3)

Part 50 explicity specifies minimum fracture toughness requirements for all ferritic materials of pressure-retaining components of the reactor coolant pressure boundary, not just the reactor vessel beltline materials. Therefore, provide the unirradiated mechanical propertfes (as described in Question 121.2) for all materials required by Appendix G, 10 CFR Part 50. Also provide the minimum thickness of the reactor vessel shell material.

121.12 Sections 5.2.3.3.1.1 and 5.3.1.5.1.4 of the WNP-2 FSAR indicate that (5.2.3) the minimum metal temperature for significant pressurization shall -

(5.3.1) be RT 60*F, where RT reactHFv+ssei beitiine ruBTo=.40*F for areas remote from the e

n Section 5.3.1.5.2.2, supplied in response to Question 121.3, indicates that the minimum metal temperature for pressurization is based upon an RT of 20*F the reactor vessel closure flange. Clarify this dNrepancy. for 121.13 In Section 5.3.1.5.2.2 of the WNP-2 FSAR, supplied in response to (5.3.1)

Question 121.3, it is stated that a factor of 2*F per ft-lb is used to convert longitudinal Charpy V-notch impact data obtained at the 30 ft-lb level to estimates of the values at the 50 ft-lb level.

It is stated that these estimates, plus a 30*F adjustment for specimen orientation, are based upon information tabulated in WRCB-217. " Properties of Heavy Section Nuclear Reactor Steels," and frcm other fracture toughness tests.

Explicitly state the procedures used to verify this factor, includ-ing a sample calculation, and any data other than that in WRCB-217 used as a basis for this estimate.

121.14 To demonstrate compliance with Appendix H to 10 CFR Part 50, (5.3.1) include in the WNP-2 FSAR and Technical Specifications a table that provides the following' infomation for each surveillance specimen capsule:

(1) The actual surveillance materials in each capsule, l

(2) The beltline material from.which each surveillance material l

was obtained, (3) The test specimen type (s), and their orientation, for each surveillance material, (4) The actal location of each capsule in the reactor vessel, (5) The lead factor for each capsule calculated with respect to the k wall thickness location.

l (6) The proposed loading schedule of the capsules into the WNP-2 reactor vessel, and

121-2 (7) The proposed time of capsule withdrawal (calender years and effective full power years).

121.15 Table 5.3-la, suppited in response to Question 121.3, indicates (5.3) that the welds, which were required to be impact tested according to paragraph II.C.2 of Appendix G,10 CFR Part 50, were not made from the same heat of base plate.

Identify the base material used and confinn that the weld wire, flux and weld procedure used to make weld metal test specimens are the same as those used to fabricate tne reactor vessel beltline welds.

121.16 Appendix G, 10 CFR Part 50, requires that RT be adjusted for (5.3.1) irradiation effects by adding to RT the therature shift in the Charpy V-notch curve for irradiNNd material relative to that for unirradiated material, measured at a specific energy or mils lateral expansion level using test specimens oriented in the transverse direction.

In response to Question 121.3, the statement is made that, based on GE experience, the amount of shift in the Ri measured on irradiated longitudinal test specimens will be NNentially the same as the shift in equivalent transverse specimens. Provide the Charpy V-notch impact data to demonstrate that the shift in RT NDT using longitudinally or transversely oriented specimens are equivalent. Also demonstrate that the decrease in upper-shelf energy caused by neutron irradiation can be equally measured by longitudinal or transverse specimens.

121.17 The response to Question 121.5 is inadequate.

It is our position (5.3)RSP that the construction and inspection of the reactor surveillance capsule attachments be done according to the requirements for permanent structural attachments to reactor vessels specified in ASME Code Sections III and XI.

'121.18 The response to Question 121.7 is inadequate. The reports (10.2.3) referenced (Reference 10.2-1 and 10.2-2) are not on file with the NRC. Provide these reports so that a complete evaluation of the response to Question 121.7 can be made. Also identify the material specifications for the rotor shaft and discs of the turbine-generator.

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AUG 211979 Task Action. Plans A-8 and A-39

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Docket tros.

50-358I'50-352/353,' 50-367, 'n 374 ~, '

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50-387/338,'.50-410, 50-322, 50-391, MD10RANDUM FOR:

S. H. Hanauer, Director, Unresolved Safety Issues, NRR m

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C. J. Anderson, A-8 Task Mana'ger, Containment Systems Branch, ~D,SS

. APPLICANT:-

Members of Mark. II Owners Group

.S*JBJECT:

MEETING'WITH MARK.II CW'IERS CROUP TO DISCUSS THL MARK II LONG TERM PROGRAM (JULY 24 and 25, 1979)

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The purpose of this meeting was to review the total Mark II generic pool

' dynamic program including the program histor.<, status of the. supporting program, and identification of, the generic and plant unique efforts required to complete the program. ~

In addition, three tasks were discussed in ;reater detail; these included: Load Case 10, the extended 4T condensation.

oscillation tests and the, Mark 'II program efforts to monitor related ^

x suppression system tests conducted outside the Mark II owners supporting

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program.

An attendance lis.t-ind a copy of the meeting h'andouts are enclos'ed.

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A summary of the significant. topics discussed during this meeting is I

provided below.-

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Introduction and Program Histor9,

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' ~ Mr. Hedgecock, Chairman of the Mart II Owner's' Group presented an overview of the Mark II cuners supporting program including identification of the l

do.mstic member utilities, objectives, organization, and identification S

of program changes in the past year. He also discussed the role of plant

," unique programs versus the generic Mark II procr

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Mr. Sobon of General Electr;ic discussed the licensing background of n

the Mark 11 pool dynami p

history of the program.c program including a description of the early

l and a description of the major elements of program (i.e... Phase I < Dynamic Forcing Function Report. Phase II - Design.'7.J.- s.ud (

'...,7;. e' C~Analys.ls Report

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4 fr 9," related. key ' area land.'P_hase'IIIe 6 Supportin l5 iU

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.- O. NRC attention / resolution.CTh'ese include: SRSS loa'd combina 3-Combinations Load Case 10'and Ac'ceptance criteria '

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Mr. Smith.of, General Electric provided an overview of the;tas'ks cogrising

-f v ', the ' Mark IIJ owners supportYng' program. This included identification of the'

.;..- wi'ge,neric tasks pogrising the. supporting" program, interrelationship' of the

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' ' tasks task do.cumentation, 'and' status of th'e uncogleted tasks..The lead, "

t, plant program i. essentially complete. 2 The supporting program which.

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..n. includes 101. tasks is about 70% complete.

The generic SRV related tasks

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iq& a ;.miscell,aneous; tasks. extend through the end of 1980 and 1981lres

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• i.' Mr. Sobon.of General Electric 'provided 'a summary ofethe Mart If owners

.c positions' regarding the' staff's NUREG.0487 acceptance criteria for. the '

. lead 'and'nonlead plants.~ Theistaff.. identified alternate criteria... 'J.

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' previously. proposed by the Mart'II ~ owners where the' staff hasco@leted.

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'. ' their revi'ew and is in ' agreement with the' Mark II owners.M These criter.iA ~

include item. I.'A, I.B.1(b).. I.B.3.(c),- III.A.1 and III.A.2 in the criterid tab.le. included i

.In addit.fon the staff' stated that an asymetric pool.n. the handouts.

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swell load ofl20% of_ the maxinum bubble pressure would n

o be acceptable.- The documentation of the staff's. evaluation cf.the'se l

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alternate criteria is presently scheduled to be released.in a.

.,..e A number offnQ'ue prog %f, v4 7 y [ 'sup'plement.td JiUREG,0487;f'n Novimber,1979.4 in,a pla t tasks fP:

"were identified by th'e Mark"II owners 'where n uniq ram !

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will be~ proposed.to justify a pla'nt uniq'ue. exception to the generic f

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criteria. The. staff expressed the concern that these plant unique.

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' efforts 1were poorly defined at this point and that an atteg t'should e.

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C. C be made'to ' consolidate these programs.to.the maximum extent feasible-; sh:e@.?,'

r,-t.l M o?ivot'd" delays' r'esulting"from limited HRC staff 'resou'rces.", i' j'.',~. 7,.'.S

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requested re

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.m.>. regdding the %uf r~ehNti'for lo'a'd case 10'eyaluation"of the'contaf nmeht.'

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i structure, piping and. equipment. This load case includes the combinationt _ ii

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i of LOCA i SSE and SRY loads.

T',he Mart II owners requested this relief

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on the grounds that this load combination is non-mechanistic.

In addit' ion, the probability for the spurious actuation of a single SRY is very small.

The results of a study for a BWR-5 plant to consider the potential for the mechanistic actuation of an SRV following a DBA uas presented by General Electric for a spectrum of breaks. The staff requested docu-mentation of this study along with the results of a similar study for.

a BWR-4 design. Arguments were presented relating to the low probability

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of this load combination resulting from spuricus SRV actuation. The staff l

requested that these arguments be documented along with the' mechanistic study. The staff stated that the arguments presented appeared a reasonable basis for providing relief from our requirement that this load case be '

considered in the evaluation of piping systems.

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4.

Extended 4T CD Tests -

Itr. Davis of General Electric presented a st'atus report for the extended 4T condensation oscillation tests.

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, This pre'sentat.f on included a description of the in'strumentation, test matrix, proposed <!ata reduction, proposed data interpretation, construction I

status and program-schedule. The staff expressed a concern related to the consideration given t'o vent mass flux and air content in establishing the test matrix. The staff's concern ' relates to the relatively high CO loads encountered in the FSTF tests during the high mass flux test. The Mark II. owners agreed 'to provida.the staff with nupporting information used by the owners to establish the test matrix.

The Mark II owners N.

m requested an accelerated review of-this informat ion by the staff and our consultants to avoid unnecessary delays in the 1est schedule.

5.

World Test Monitoring Prooram Mr. Davis of General Electric provided a ' status report of the Mark' II owners program to monitor related ~ pressure suppression LOCA and SRV tests conducted outside the Mark II. program. This presentation ini:luded identification of the sources contacted, the type of-information received and identification of s~ite visits conducted.

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Clifford Anderson, A-8 Task. Manager

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Enclosure:

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