ML20054F720

Forwards Draft Safety Evaluation of SEP Topic III-1, Quality Group Classification of Components & Sys

ML20054F720
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	06/07/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Linder F DAIRYLAND POWER COOPERATIVE
Shared Package
ML20054F721	List: ML20054F720 ML20054F725
References
TASK-03-01, TASK-3-1, TASK-RR LSO5-82-06-016, LSO5-82-6-16, NUDOCS 8206170248
Download: ML20054F720 (7)

Text

l n=

c w

June 07, 1982 Docket flo. 50-409 LS05-82-06-016 fir. Frank Linder General 11anager Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601

SUBJECT:

SEP TOPIC III-1, QUALITY GROUP CLASSIFICATI0tl 0F C0ftP0NEllTS AllD SYSTEftS - LACROSSE BOILIllG WATER REACTOR Enclosed is the staff's draft safety evaluation of SEP Topic III-1 for the Lacrosse Boiling Water Reactor. Our evaluation (Enclosure 1) is based upon our contractor's final evaluation (Enclosure 2) of this topic.

This assessment compares your facility with the criteria currently used for licensing new facilities.

You are requested to examine the facts upon which the staff has based its evaluation and respond either by confirming that the facts are correct, or by identifying errors and supplying the correct information.

The staff was unable to complete this topic due to the lack of informa-tion of original design requirements for various components. The need to supply additional information and to complete the evaluation for this topic will be determined during the integrated assessment.

We have concluded, for those camponents where a comparison of codes was possible, that the changes in the codes since the original design do not significantly affect the safety of the plant.

Based on our sampling

$60Y of code comparisons to date we do not expect the rematring items to pose

&}

a significant hazard from continued plant operation.

Yoe: response is requested within 30 days of receipt of this ev

)

If no response is received in this time, we will assume the evaluation l

1s correct.

Apoi f

r, /Aha &IS I

Sincerely.

h:DL 0 ' N A

Dennis !1. Crutchfield, Chief Dr

.chfield GL inas 6/tI/82 6/ /82 Operating Reac. tors Branch flo. 5

\\

Division of Licensing

• See previo,us yellow for additional concurrences.

r__,__.___

~

mu-m m. m o. o.

. SERB:DL.

SEP8 ;DL-...-.SEPBo0L..- - DRB & pH- --

omcE >

SURN1.ME)

..R 5........

.(.

b. *

• bbO R 0 N d lOM.** * ** * * ** *

$$fk[o!

o!$0No9

. 6/.d./.8 2.........

.5/.19/32.~.

5/48/83...-- 5/20/.82- - -

o.

snc rond -.mu -

vr'FICIAL RECORD COPY usom ini-m

e 9

+7 Docket No. 50-409 LS05 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601

SUBJECT:

SEP TOPIC III-1, QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS - LACROSSE BOILING WATER REACTOR Enclosed is the staff's draft safety evaluation of SEP Topic III-1 for the Lacrosse Boiling Water Reactor. Our evaluation (Enclosure 1) is based upon our contractor's final evaluation (Enclosure 2) of this topic.

This assessment compares your facility with the criteria currently used for licensing new facilities.

You are requested to examine the facts upon which the' staff has based its evaluation and respond either by confiming that the facts are correct, or by identifying errors and supplying the correct infornation.

The staff was unable to complete this topic due to the lack of infoma-tion of original design requirements for various components. We have concluded, for those components where a comparison of codes was possible that the changes in the codes since the original design, do not signifi-cantly affect the safety of the plant. Based on our sampli.ng of code comparisons to date we do not expect the remaining items to pose a significant hazard to safe plant operation.

Your response is requested within 30 days of receipt of this evaluation.

If no response is received in this time, we will assume the evaluation is correct.

Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing ORB #5 AD:SA:DL

Enclosures:

RDudl y Glainas g g /82 5/ /82 As stated cc w/ enclosures:

SEPB :DLIF.SL-SEPB :Dh~

L. SEPB:DL C-0RB #5 :DL C-

....~......~...nf

• ~ ~~ ~"""""

omcE >

~ ~... ~

...Seg.'tf uX t..p sYe...........

MBoyl e:dn C..G..r im.e..s..nu WRus. sell DCrutchfield

$URNAMEh au- - ~ " " * -

..u

............n...

.. ![kI....

!I. I. - " - !I hl - ~~ ~ EI-I S ~~ -

~ ~ " ~ " - -

om)

OFFICIAL RECORD COPY usmaat-um J NRC FORM 318 00-80) NRCM ON3

Mr. Frank Linder cc Fritz Schubert, Esquire U. S. Environmental Protection Staff Attorney Agency Dairyland Power Cooperative Federal Activities Branch 2615 East Avenue South Region V Office La Crosse, Wisconsin 54601, ATTN:

Regional Radiation Representative 230 South Dearborn Street

0. S. Heistand, Jr., Esquire Chicago, Illinois 60604 Morgan, Lewis & Bockius 1800 M Street, N. W.

Mr. John H. Buck

~

Washington, D. C.

20036 Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Mr. R. E. Shimshak Washington, D. C.

20555 La Crosse Boiling Water Reactor Dairyland Power Cooperative Mr. Ralph S. Decker P. O. Box 275 Route 4, Box 190D Genoa, Wisconsin 54632 Cambridge, Maryland 21613 Mr. George R. Nygaard Charles Bechhoefer, Esq., Chairman Coulee Region' Energy Coalition Atomic Safety and Licensin~g Board c

2307 East Avenue U. S. Nuclear Regulatory Commission La Crosse, Wisconsin 54601 Washington, D. C.

20555 Dr. Lawrence R. Quarles Dr. George C. Anderson Kendal at Longwood, Apt. 51 Department of Oceanography Kenneth Square, Pennsylvania 19348 University of Washington 4

Seattle, Washington 98195 U. S. Nuclear Regulatory Commission Resident Inspectors Office James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III Rural Route #1, Box 276 Genoa, Wisconsin 54632 799 Roosevelt Road Glen Ellyn, Illinois 60137 Town Chairman Thomas S. Moore Town of Genoa Atomic Safety and Licensing Appeal Board Route 1 U. S. Nuclear Regulatory Commission Genoa, Wisconsin 54632 Washington, D. C.

20555 i

Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Alan S. Rosenthal, Esq., Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comnission Washington, D. C.

20555

ENCLOSURE SYSTEMATIC EVALUATION PROGRAM TOPIC III-l LACROSSE BOILING W.TER REACTOR III-1, Classification of Structures, Systems and Components TOPIC: -

(Seismic and Quality)

I.

INTRODUCTION SEP plants were generally designed and constructed during the time span from the 1950's to the late 1960's. The plants were designed to generally recognized codes, standards and criteria in effect at that time; however, the codes, standards and criteria have been periodically revised. There-fore, the SEP plants may have been designed ;and constructed to codes, standards and criteria no longer in effect or acceptable to the NRC.

The purpose of Topic III-l is the review of the classification of structures, systems and components of as-built plants compared to the current classifications required for seismic and quality groups in the codes, standards and criteria.

Since the review of seismic classifica -

tion is addressed in other SEP topics (See Section III of this evalua-tion), this topic has been limited to the evaluation of quality group classifications.

II.

REVIEW CRITERI A The review criteria for this topic are presented in Appendix A of Technical Evaluation Report C5257-437, " Quality Group Classification of Components and Systems - Lacrosse ~ Boiling W'ater Reactor," prepared for the NRC by Franklin Research Center (attached).

I III.

RELATED SAFETY TOPICS The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review are performed in related topics.

As stated previously, the seismic aspect of this topic has been deleted.

The quality aspect for the reactor vessel and steam generators (PWRs The related safety only) and the quality assuranca have been deleted.

topics, and the subject matter covered in the topics, that cover the aspects deleted in Topic III-l are identified below.

l III-6 Seismic Design Considerations III-7.B Design Codes, Design Criteria, load Combinations and Reactor Cavity Design Criteria V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity XVII Operational Quality Assurance Program The resolution of Topic V-8 is part of Unresolved Safety I'ssues A-3, A 4 i

l and A-5.

]

t

IV.

REVIEW GUIDELINES The review guidelines are presented in Section 3 of Report C5257-437 (attached).

V.

EVALUATION The basic input for this report is Table 4.1 in Section 4 of Report C5257-437. Table 4.1 is a compilation of all systems and components which are required to be classified by Regulatory Guide 1.26 and the original codes, standards and criteria used in the plant design.

After comparing the original codes, standards and criteria with those currently used for licensing facilities the following areas were identified where the requirements have changed.

1.

Fracture Toughness 2.

Quality Group Classification 3.

Code Stress Limits 4.

Radiography Requirements 5.

Fatigue Analysis of Piping Systems An evaluation of each of these areas is presented in Section 5 of Report C5257-437 with a detailed discussion in the Appendix of the report.

We have detennined that changes in the following areas have not signifi-cantly affected the safety functions of the systems and components reviewed in this report:

1.

Quality Group Classification 2.

Code Stress Limits 3.

Fatigue Analysis of Piping Systems In the remaining two areas, we have concluded the following:

1.

Fracture Toughness - The current code requires that pressure retaining materials be impact tested.

For 23 df 72 components reviewed, sufficient information was available to exempt them

~

.from this requirement.

2.

Radiography Requirements - We have determined that; a primary vessel designed to ASME Code Section VIII, for which a.

code cases 1270N and 1273N were invoked, would meet current full radiography requirements.

Secondary vessels designed to ASME Code Section VIII, for which code case 1270N was invoked, and which are presently categorized as class 2 or 3 vessels, meet current radiography requirements for material weld thicknesres exceeding 1-1/2 inches,

~

auttenitic stainless steel piping design to ASA B31.1, for which code case N7 was invoked, meets the current full radio-b.

graphy requirements.

Piping designed to ASA B31.7 meets the current full radiography requirement, and valves designed to ASA B31.1, for which code case N1, in combination with N7, N9 or N10, meet current full radio-c.

graphy requirements.

Our review has not identified any significant deviations from past codes.

However, we were unable to complete our evaluation due to insufficient information for the following:

Fracture Toughness - For 49 of 72 components there is insufficient 1.

The licensee lnformation on materials to complete our review.

should provide the necessary information using the format provided in items 1 thru 8 of Tables A4-4 through A4-6 in Appendix A of Report C5257-437. Table 5-1 of the report identifies those com-ponents for which this information is necessary.

Radiography Requirements - The licensee should provide the following 2.

~information:

Radiography requirements imposed on the control rod drive a.

housing.

Radiography requirements imposed on Class 2 and 3 vessels for

\\

which code case 1273N was not invoked,and with welded joint b.

thicknesses 'ess than 1-1/2 inches.

Radiography requirements imposed on Class 1 and 2 piping and l

c.

valves designed only to ASA B31.1.

(

Piping - Calculations similiar to those presented in examples 1 and 2 in Section 4.2, Appendix A of Report C5257-437, applicable to Lacrosse 3.

Plant design parameters, should be provided in order to assess the im-pact on the usage factor of gross discontinuities..in Class 1 piping systems for a medium and large number of cyclic loads.

Valves _ - Information should be provided by the licensee, on a sample' basis, regarding the design of valves in order to evaluate if they 4.

meet current body shape and pressure-temperature the current requirements.

j

5.

Pumps - None of the nine pumps reviewed'was designed to Section VIII of the ASME B&PV Code.

Information is needed on the pumps in order to evaluate if the current requirements are met.

Codes, code classes, editions, codes cases, and design calcu-lations should be provided for all pumps in the Lacrosse plant.

Proof of compliance with current fatigue analysis requirements for current Class 1 pumps (the recirculation system pumps)

~

should be provided.

6.

Storage tanks - (1) Atmospheric storage tanks should be evaluate'd to determine if they meet current compressive stress require-ments; (ii) O to 15 psig storage tanks should be evaluated to determine if they meet current tensile allowables for biaxial stress field conditions; (iii) additional information and calculations for storage tanks designed to AWWA D100 should be provided to determine if they meet current standards.

7.

Missing information - The following ir. formation, which is incomplete or missing from Table 4-1 or Tables 4-2(a), (b) and (c) of Report C5257-437 sho~uld be provided:

(i) any specifications or calculations used in designing pumps, valves, and tanks that may assist in conducting this evaluatin, (ii) confirmation of assumptions made regarding code editions (see Table 4-1),

(iii) provision of fatigue analysis calculati.ons for Class 1 piping.

systems similar to examples based on the Palisiides Specifications given in Section 4.2 of Appendix A; and g

(iv) clarification or additional information on notes 5 and 6 in Table 4-1.

A more detailed explanation of the information to be provided inay be found in Report C5257-437 (attached).

VI. CONCLUSION We have determined that for the following, changes between current and original code requirements for the Lacrosse Boiling Water Reactor will not significantly affect the safety functions of the systems and components reviewed:

1.

Quality Group, 2.

Code Stress, and 3.

Fatigue Analysis for Piping Systems.

We were unable to complete our review due to insufficient information regarding various other systems and components. The required information is discussed in Section V of this evaluation.

Based on our sampling of code comparisons to date, we do not expect the remaining items to pose a significant hazard to safe plant operation and,-

therefore, have determined that the schedule and need for providing the remaining information can be determined during the integrated plant safety assessment.