ML20040A630

Forwards Request for Addl Info Re Reactor Sys,Radiological Assessment & Structural Engineering Needed for OL Application Review

ML20040A630
Person / Time
Site:	River Bend
Issue date:	01/06/1982
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	William Cahill GULF STATES UTILITIES CO.
References
NUDOCS 8201210318
Download: ML20040A630 (20)

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DISTRIBUTION:

' Docket Files"-- 50-458/459 JAN 6 1382 LB#2 File Dewey, OELD DEisenhut/RPurple RTedesco Docket flos. 50-458 ASchwencer t

and 50-459 RPerch I&E IE Regional Offic m

LB#2 LA 4

b Mr. William J. Cahill, Jr.

Senior Vice President C

1 River Cend Nuclear Group 0

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Gulf States Utilities Company g-4 r8 Post Office Box 2951 e

Beaumont, Texas 77704 g

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Dear fir. Cahill:

g 47

Subject:

River Bend Station Unit Nos. I and 2 - Request for Addit Information As a result of our review of your application for operating licenses for the River Bend Station, we find that we need additional.information in the Reactor Systens, Radiological Assessment, and Structural Engineering areas of review.

The specific information required is listed in enclosures (1), (2) and (3).

Responses to the requested additional information should be provided as soon as possible.

Please contact the flRC project manager for your facility within seven days of receipt of this letter to provide your planned response date.

Sincerely, I

Odg)nal signd M A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing

Enclosures:

As stated cc w/ enclosures:

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NRC FORM J18 0}80) NRCM om OFFICIAL RECDiiD COPY usa mssw

Mr. William J. Cahill, Jr.

JEEI O 1982 Senior Vice President River Bend Nuclear Group

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Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas.77704 cc:

Troy B. Conner, Jr., Esquire Conner and Wetterhahn 1747 Pennsylvania Avenue, N. W.

Washington, D. C. 20006 Mr. J. E. Booker Manager -Technical Programs Gulf States Utilities Company Post Off. ice Box 2951 Beaumont, Texas 77704 Stanley Plettman, Esquire Orgain, Bell and Tucker

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Beaumont Savings Building Beaumont, Texas 77701

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Karin P. Sheldon, Esquire Sheldon, Harmon & Weiss 1725 I Street,' Ns W.

Washington, D. C. E0006 William J. Guste, Jr., Esquire Attorney General State of Louisiana Post Office Sox 44005 State Capitol Baton Rouge, Leuisiana 70804 Richard fl. Troy, Jr., Esquire Assistant Attorney General in Cnarge State of Louisiana Department of Justice 234 Loyola Avenue New Orleans, Louisiana 70112 A. Bill Beech, Resident Inspector s ^

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P. O. Box 1051 St. Francisville, Louisiana 6

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Reactor Systems Brancj),

-. Scram time characteristics shown in figure 5.2-3 are inconsistent wi1hithe ~

440.8 i

(5.2.2).

scram time'ch'aracteristics.shown in figure 15.0-3.

Explaim why they are l

different?

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440.9 Confinn that.the~ overpressure analysis includbs the effects of the ATWS reactor (5.2.2) recirculation pump. trip:on high reactor pressure.

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440.10 Page 5.2-13 discusses ADS valve actuations and states that accumulator capacity

.(5.2.2) is sufficient _ for.each' ADS valve to provide two actuatiorts against 70% 'of the maximum drywell -design ' pressure. -

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TMI Action Plan Item II.K.3.28 identified the need to assure that air or nitrogen accumulators for the ADS valves are.provided with s,.ufficient capacity to cycle the valves open sufficient times at design pressures. It should also be assured that the long-term air supply is designed,to withstand a hostile environment and still perform its function 100 days after aE accident.

Finally, i: 0;f d be verified it.at no sir.gie active failure can disable the long-terr. air supply (LRG II issue 8 RSB)/ We require additional information addressing the a t:ve ::ncerns.

M C.11 Provide all system and core parameter initial values assumed in the overpressure (5.2.2) analyses.

Include their nominal operating range with uncertainties and Technical I

S,pecificati'on limits.

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440.12 Provide the calculations or data to support your relief valve discharge coefficients l(5.2.2)

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and flow capacities.

dl3 Sensitivity studies showing the efiect of initial operating pressure on the t.s.e.2 ) peak transient pressure attained during a limiting overpressure event have not been provided.

Therefore, either:

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Pag 2 2 of 4 -

1) Provide the sensitiv'ity study.which shows that increasing thf,

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initial operating pressure (up to the maximum permitted by'the

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high pressure trip'setpoint) will have a negligible effect on the

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peak transient pressure, or

2) Propose a Technic ~al Specification which will assure that the

' reactor operating pressure will not exceed the initial pressure assumed in.the, overpr. essure analysis.

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440.14 The performance of essentialTy all types of safety / relief valves has been less (5.2.2) than expected for a safet'y ' component. Because of Peportable events involving

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malfunctions of these valves on operating BWRs, the staff is of the opinion that significantly better safety / relief valve performance should.be required of.new pl a nts. Provide a detailed description o'f improvements between your plant and

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presently operating plants in the areas listed below.

In addition, explain why the noted differences will provide the required performance improvement.

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1) Valve and valve operator type and/or design.

Include discussion of improvements in the air actuator, especially materials used for cc ponents such as diapnragms and seals.

Discuss the safety margins and confidence levels associated with the air accumulator design.

Discuss the capability of the operator to detect 1.ow pressure in the accumulator (s).

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2) Specifications. What new provisions have been employed to ensure that

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valve and valve actuator specifications include design requirements for -

l operation under expected environmental conditions (esp. temperature, y

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humidity, and vibration)?

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3) Testino.

Prior to installation, safety / relief valves should be proof-

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U2.2) tested under environmental conditions and for time periodsnepresentative -

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of the most severe operating conditions to which they rnay be subjected.

4) Quality Assurance. What new programs have been instituted to assure that valves are manufactured to specifications and will operate to.

g,ecifications? For example, what tests are performed by the applicant

-to assure that the blowdown capacity is correct?:

.-'. ' 5) Valve Operability.

Provide your surveillance program to monitor.the

2.2 performance of the safety / relief valves.

Identify the information that will be obtained and how these data will be utilized to improve the operability of the valves.

For example, how will this ' program reduce the malfunctions that have occurred in operating reartnrs?

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6) Valve Inspection and Overhaul.

The FSAR states that one half of the safety / relief valves will be bench checked and visuaTly inspected

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e. e r;..efueling c a ge.

Mc. ever, depending on operating cycle l ength, this may result in 'several years between inspections.

Ersting experience has sho,in that safety / relief valve failum may be caused by exceeding the n;anufacturer's recommended service life for the internals of the safety / relief valve or air actuator.

At what frequency do put intend to visually inspect and' overhaul the ADS portion of the safety / relief valve 7 For both safety /

I relief and ADS modes, what provisions exist to ensure that valve inspection and overhaul are in accordance with the manufacturer's recomendations and thht the

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design service lif,e is not exceeded for any component of the safety / relief valve?.

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n 44}.15 Participation in GE safety relief valve surveillance program - confinnation from j

.( 5. 2. 2 )

l the applicant is required of the applicant's participation in this surveillance program (NUREG-0512).

(LRG II Issue - 3 RSB) l l

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Paga 4 of 4 G40.16 5/R valves test requirements - TMI Item II.D.l.

The applicant states,that they '

(5.2.2)

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are working with BWR Owner's Group to develop a generic position. En~ce a generic position is already developed, we require detailed and core specific response.

Testing of SRVis to measure water flow rates is to be addressed.

440.17 TMI Item II.D.3 The SRV position indication design for RBS should be described '

(5.2.2) in the FSAR.

440.18 TMI Item II.B.1 Reactor Coolant System Vents. Your response is not acceptable.

[5.2.2)

Refer to BWR Owner's Group position and submit plant specific response in detail as given by other BWR owners.

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Radiological Assessment Branch 471.10 In Section 12.4.1 a nd 12.1.2.5s you stated that t h e.'o c c u p a t i on a l (12.4)

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~7-(12.1) dose assessment vill be provided in a la t er amesadm ent.,

Regulatory Guide 1.70s Section 12.4s specifies that you should-perform such a dose assessment.

You should provide a' copy of this assessment 'in~ accordance with Regulatory Gside 81,19 and a. Listing of plant improvements 'you wiLL make as a ' result of this revi.ew.

471.11 As specified in Regulatory Guide 1.70s Section 12.2.2, you (12.2) should provide.a tabulation. of the expected airborne radioactivit) concentrations in equipment cubicles, corridors and operat irig '

areas normally occupied by operating personneti including sources resulting from reactor vessel head removat and spent' fuet ~

op.erations.

In Section 12.2.2r ~ you s p e c if ied that such areas

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do not contain airborne s o.u r c e s.

O p e r a t i n g - e x pes-i e n c e has shown that equipment leaks have resulted in airborne radio-activity.

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471.12 As specified in Regulatory Guide 1.70, Section 12. 2.1 r you should (12.2) provide scurce descriptions _for the maj.or scurces in the plant.

Your FSAR did not i n c t, u d e information on the source t e rra -

for shielding the transvarse incore probe system You should

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provide this source term _information.

471.13 As specified in Regulatory Guide 1.70s S ec tion 12.1.2, wi th (12 1) regard to the radiation protection reviews to assure that ALARAc.-

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objectives are being met in the design process and during c

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construction (including field run pip.ing):

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Identify by title the individual (s) responsible for the radiation protection design reviewscarad desc' ribe "

how.they-relate to the individual responsible'.for the overalL design.

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Provide a breakdown by title of radiation.protectiosa personnel who have been or wiLL be p'articip'ating in.these reviewse tabulating the health physics education and--

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3-experience required of, each.

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Describe f ormal arrangements and procedures for assuring' c.

' hat adequate independent radiation protection reviews. are t

performed throughout the design and construct-ion processesc and that adequate records are.kept to document the completion of each such review.

471.14 Provide additional information on how your exposure-tracking (12.1 )

(12.5) and exposure reduetion'programe incl.udes' the el'ements of Regulatory Guide'1.70, S e c t i o~n 12.1.3 and 12.5.3, and Reguletory Guide 8.8r Section C.3.9(8)Cj), C.3.8(2), and C 3.cC2)(5)<-

i n c l.u d in g rem-tracking, self readina pocket dosideter use, post-taske actual exposure evat.uatione hnd.how N ese results are used to make changes in future work.

Verify that annasa l ex p osu r e _

management and that.these are.

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reviews are performed by plant

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used to ide3tify groups with the highest exposure'in order t o* i-p assure that doses are ALARA.

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. 3. 2 '- and (12.3)

Pegulatory Guide 8.8, Section C.2.a.11, you should provide a' description of aLL design f.eatures, including shielding a n d.-

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. int e rloc k s u s ed t o r edu.c e. rad i ati on in-the.a reas of.(. th'e spent.

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fuel transfer tube.

Include 'the estimated radiation levels.

outside such shielding during fuel transfer.

It is'our Y.

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all accessible port. ions of.th.e spent fuel trans-7

-. position that

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.fer. tube and or canal must be shield 6d during fuel transf er.

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This Use of removable shielding f or this purpose is acceptable.

,.7 shielding shall be such that the resultant contact radiation.

levels.shaLL, be' no greater than 100 rads per hour.

ALL access-

.ible pori: ions of the spent fuel transf er ' tube shaLL' be clearly rarked with a sign stating that potentially lethat radiation fields are possible during fuel transfer.

If removable shield-ing is used for the fuel transfer tubes, it must also be r.

explicitly marked as above.

If other than per:anent shielding is used, local audible and-visible alarming radiation monitors must be installed to a.lert personnel if t esip o ra ry fuct transfer -

tube shielding is removed during f,uel transf er operations.

471.16 Provide a description of how the airborne _radioactjvity monitors (12.3)

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. described in Section 12.3.4 (which consist of noble gas'.and ~-

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part{culat6 op only noble gas monitors', can haVe a minimum c I-41-1

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c apabi.L i t y ;of det ec ting 10 mpc-hou r.4 of particulate or iodine

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radioactivity,.as specified in NUREG-0800 Section 12.3.

You should note that the monitors should be able to detect the 10 mpc-hours from any compartment that has the possi.bility of-contain.ing a.irborne.. radioactivj ty. and which.snormally may be-o c c u p i e_d e. t a k.i n g.into. acc ount. d.i.lution in' the venti tatibn system.

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'.... In. accordance with Regulatory Guide 1.70, Section.12.3.ir you.

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471.17 (12.3)

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design should' d'escribe' the' Yeatures in~corporated in the plant, to maint'aih occ~up.atiohat 'adiation ex posure At. ARA by minimizing. '

r and c~ntrollihg th'e buildups transports and dispositsion of o

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a c t iva t ed' c6r r osi'on p'r odu'ct s in reactor coolant and, auxiliary system'.

In additione you should ou.tline the chedistry control program to reduce and control the transporti a c t i v a t i on and deposition l

- corrosion produc'ts in normally radioactivs? systems.

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'471.18 Your L"i s t i n g o f 'a r'e a-r a d i a t i on m o'n i t o r s in Sect i on 12.3.4 (12.3) 4 shows only one monitor with a range of 10 R/hr or above.

l; Your listing of portab'le radiation survey instru&ents in 4

Table 12.5-3.shods no instrument s with. a r.ance of 10 R/hr.

r Regulatory Guide 1.97 (Rev. 2) specifies that portable su r'v e y.,,

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_ meters and the area radiation monitors in areas requiring

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.....ess aftp an accid _nt should have a range up to '104 R/hr.H '-

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p You should Provide a commitment in,ybur FSAR to have such y,

portable instruments'and.should specify locations of area

radiation monitors in areas requiring access after an accident.

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471.19 Table 12.5-2 'shods th'e number 'of radiation detection instru

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ments whic'h sill be available for both nits.

It is our -

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mentsi (especially those which are most^ f requently. used.by-'::f ~

radiation protection personnele 0-5000 mR/hr) should be adequates

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1or.the fotLowing: situations.. (a) both units'can be shut-s-- -y.y. g.,; _.g:;. g-4g.,. r _:.-,; - y.;..

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d own f o r.. r..,e pa.i r s a t the same time (maximum use.of surveyT,.c'- -

,1 instr.uments)r (b),a number of" instruments out of service (in

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need of calibration or repairs), and C.c ). a number of spare

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operational instruments should be always available f or use in

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unusual. occurrences.- You should evaluate. the quantities of the above and ' revise-your

....<.c instruments anticipated considering inventory as appropriate.

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As specified.in Regulatory Guide 1.70s Section 12.5.2, f ou should specify in Table 12.5 -4, the =inimum quantities of respi.ratory p r o t e c t i on e qu iptr e n t that wiLL be available.

State if quantitative respirator fit test wiLL be available for both testing and for training of personnel in'the usage of y,

respirators'.~~

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471.20 Provide the information requested in II.B.2r II.F.1(3) and.';.

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II.F.1(3);; III.D.3.3 of NUREG-0737, " Clarification of TMI Action.PlanHAy

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471.21 As 'specified in Regulatory Guide 1.70s Secti on 13.1.1.3 cand (13.1)

. NUREG-0737, you should provide an cut tine of the qualifications T

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'It.is otw ~ position that' - T

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RPG have the qualifications specified in. Regulatory Guice 1.8.-

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for Radiation Protection Manager.

The December 1979 revision T;;

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xn r-( d ;M. f, of ANSI 3.1's;secifies that individuals temporarity. fi Lling the

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degree.in science or' engineer-i s..-

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RPM position should have a B.S.

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. which should be. nuclear power plant experience, six months

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of. which sl$ould.be on site.

It is our position that su c h.... ; -

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' ' ;r experience should be prof essional ex;ierience.

Identify a'nd.

provide en outline of the qualification of the individual who wiLL act'as the backup for the RPM in his. absence.

M 471.22 In accordance with Regu'atory Guide 8.8r it is dur position (13.1) that the Radiation Pr'otect. ion Supervisor (KPS) should have direct access to the Plant Manager in all radiation protection matters. 2 In matters related to radiological health and safety, the RPS has direct' responsibility to both emplo'yees' and management t h a t' -. -

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can be'sj be fulfilled.if he is independent of station divisio'ns,

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technicat..... support,

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maintenance or a

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?:- E prime respon'sibility.i.<s con.tinuity or improvement o f ' s t a t i o n r- -;P.

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Your FSAR r e-v i s ed to outline how your:l.".J y.

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Concurrent to the changes r eques t ed abov e, N gure 13.1.2

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should show that on radiation protection matters, the RPS-has -

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access.to the Plant Manager...

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ENCLOSURE (3) d N.

i.h 220.0 STRUCTURAL ENGINEERING BRANCH

...C,

~J~

' ' ~

22b.12.. _

The table referred to fn paragraph 3.3.2.2.3 is an d' -

~

(3.3.2.2.3) incorrect reference. Please provide the correct re-..,

ference for tornado missiles. ;

g 220.13 Provide an explanation of your r.easons for not providing.~.'

(3.3) tornado protection foe the radwa'ste building above grade-te noted in footnote (27). of Table 3.2-1 of the FSAR.;n...

g-

~ _

3 Tif '

220.14 (3.3.2.2.4)

The tornado load c'o'mbin'ations.l.is'ted in parag~raph 3.3'.2.2.4 of the FSAR Tdb 'n'ot' rEnta'in the folloaring as explicit combinat'i~ons:'

Z

~,

"~

W

=W 1:f

.~:. '

~

L....

^

~

t w

W

=W t

m

. ~

W

=W to.5 W'p'

~ ~ ~

~

t w

Provide justification, for this: omission or include the above combinations in :your an'alysis.

220.15 Table 3.4-1 indicates that the design basis fland level (3.4.2)'

(DFBL) is at 95' - 1" above mean sea level (MSL). Table 1.3-8 indicates that the DFBL is at 98' - 6" above MSL.

Which is the correct value.

If the higher level is correct what are the consequences.

220.16 Table 3.4-1 ind-icates that the design basis ficad level (3.4.2)

(DFBL) for ground water is 70' - 0" above mean sea level (P.5 L ).

It is stated in the first entry of table 1.3-8

• hat the ground water will rise a maximum of 13' - 0" above the normal ground water level of 57' - C'".

Assuming that the 70' - 0" MSL was used to determine lateral earth pressures, how was it assured that the ground water level will not rise above the 70' - 0" MSL under some circumstances.

220.17 Provide an explanation and derivation for the expressions.

(3.4.2) listed in Figure 2.5.79 and, in particular, the asterisked footnote as well as the pressure diagram at the extreme right referring to compaction lo' ads.

220.18 Provide a comparison of the tornado missile barrier thick.

.,[..

' # (3.5.3) ness used for all Category 1 concrete structn es at the "

C

~-

plant and those listed in the NRC S.R.P., NUREE 0800,.

Section 3.5.3, Table 1, Revision 1 dat.ed July 3S81. Where

'r * "

. wall or roof thicknesses are less than those noted in the S.R.P., provide an explanation.

4

y.:i.;,.

~.

. ',; d -

220.19 Are there any openings in the. walls or roofs af Categorf 1; -

3]%.g,n.

(3.5.3)

Structures (for example, for ventilation which could allow e t. '.-

a tornado missile to pass?

If so, what protection is pro..-

-Of

.I Vided to protect targets in way of the openings?

.g

. f.. _

. q~~ :

. vq.

220. 20 For concrete structural components designed to resistb-^

'- ~

(3.5.3) impactive or impulsive loads, provide a comparison of the

~

. 4. c.c.

s design criteria you used for allowable ductility ratios.3p

, t'~?

and the criteria outlined in Appendix C of ACL349 as :,..,/.

modified by USNRC Regulatory. Guide'1.142.. Provide an M W.....C.

5. M.E.. 3 explanation for any unconservative differences.

{

220. 21 For steel structural components. designed to resist imOf (3.5.3) pactive or impulsive loads, provide the design criteria ec

v. i for allowable ductility ratios and technical basis. Com.

L pare your criteria with that found in Appendix A of USRIRC::.

~ 4. ~.

.r..

S.R.P. 3.5.3 and provide an explanation for any unconser-vative differences.

4-

-g,

x a

220. 22 Provide justification for adopting the methods of Appendix C (3.5.3) of SWECO 7703 for the evaluation of overall response of barriers to missile impact.

It is. understood tfaat this method is somewhat.less conservative than that recommended' by Williamson & Alvy.

220.23 Demonstrate that the frequency intervals, at tdsich spectra (3.7.1.2A) values are calculated from' the design time histeny, are small enough such that any reduction in these intervals does not result in more than ICS change in the cc:puted spectra.

220. 24 Is the seismic analysis method used for Category 1 Structures (3.7.2.1.A) at River Bend (response spectra acdal analysis) censervative with respect to a finite bcundary approach in which the soil structure is modeled as a finite element mesh?

Provide the basis for your answer.

220. 25 You state that:

"The number of mass points is then in-(3.7.2.1.1.2A) creased until additional mass points do not apbreciably change the dynamic characteristics of the model:." The staff cr.iteria is that the number of mass points.' included.

in the model is adequate if the inclusion of additional masses will not increase responses by more than 15.

Indicate if such is the case and provide justification and further details if it is not.

220 26 Indicate _if the number of modes considered in your analyses ~

Y*7-9 3.7.2.1.1.2A) of Category 1 Structures.is such that the consideration of additional modes will not. result in more than a IM in-

'r O crease in responses. Provide further details amad,iusti-

~

fication if such is not'the case.

S h

.EY ',.

.7.*

220.27 Provide a summary of natural frequencies mode shapes and -2 C.

(3.7.2.1.1.3A) responses for the Standby Service Water Ceoling Tower and Standby Service Water Pump House.

220.28 Provide a tabulation of the " rattle space" surrounding (3.7.2.1.1.3A)

Category 1 Structures along with 'an adjacent tabluar listing of the Worst-condition gaps between the structures. -

M.

If any excursions greater than the "rattlespace!" are ini..

~

dicated, provide an exp,lanation.r.

.... y;c::::. --.' ~- Q,

30. af

~~

220.~ 29 Discuss your approach to soil laysr modeling b your (3.7.2.4A) analysis of the plant structures for seismic Toads and the construction of soil springs.

-a.;,

. ~,-

220.30 Explain why the off-diagonal terms of the dawing matrix-..

(3.7.2.5A)

C

...are ignored with no significant loss of accuracy'..."

~

220.31 Provide an equation to describe the method used to combine

.(3.7.2.7A) closely spaced modal responses.

220.32 Will the collapse of any non-Category 1 Structurle impair (3.7.28A)-

the integrity of any Seismic. Category 1 structurv Gr component? If so,'is the non-Category.1 Structure designed to Category 1 standards? Provide an explanation if necessary, 220.33 In your consideration of torsional effects, in the dynamic

~

(3.7.2.11A) analysis of Categcry 1 Structures, was an additional eccentricity of 5% of the maximum building er structure dimension added to account for inaccuracies in determining torsional effects?

If not, explain your reason for not adding in such a conservatism.

220.34 The values in Ta',le 3.7A-7 do not appear to der.cmstrate (3.7.2.12A) approximate equivalency of the results of the time history analysis and response spectrum analysis. Please explain.

Also, what are the units of the entries in the table?

220.35 Show the formulation by which the equivalent m6dal damping (3.7.2.15A) ratios are defined. Also, provide an-explanation of your (3.7.2.5A) statement, quoted as follows, and the. method. referred to therein:

"The modal damping is determined by the ratio of

~

dissipated energy to strain energy for each mode' shape, a method which provides realistic estimates l

of damping, especially in~ modes whie foundationtranslationsorrotationsf57.""*'i"I"#98 M

5.'220.36 Explain the reference, in the last para' graph of this (3.7.2.15A) section, to "the stress intensities given in Regulatory l

Guide 1.61."

0

~,,

...~.

g;.

. _ 5 :.,=. w

.{.-f: % 2

., ~

.. ~ b, ?

- 9,.Jli, ~.

., h.

P #.t'. "k Provide an explanation and justification for: choosing ad;//e*.

220.~37

-Y=r.

1

-(3.~7. 2.15A) 10% damping. ratio:for subgrade. ~ components as apposed to%Ei N-fM.E some other values; hfine the phrase "subgrade components."J '- f.y 3;is

. 2.=- 5..

=: i-.LD ' 1 method for applying the earthquake input to the radmaste.YM. MMS Provide a detailed, step by step explanatiorr'ef the? J.M'EN7 220.38

~

(3.7.2.16A) building; i.e... clarify and expa'nd the last paragraph Q -f iy,]-jg@

~

in_.this section. 3

" d +h

^

~~.x

- -, - m-a,.g..; a,gg For the static analysis method.&.,7 provide justWication. fore..

-220.39 (3.7.-3.1.1.1A) applying a static' coefficient to the peak.acri'leration of 3{g.c -ue$. w.. ' 'f;-

M.,

1.3.rather than the usually accepted value of 1.5.

Also,2

yM^

provide justification for not applying the static coefficient. :. 5.Z-in a11 sitac.tions.

.6

. ~#.'.:-.:.x ?.-J.C; :R.' JF

..c.w n... %.z.

n. _=-

220.'40' Show that the static analysis will always prourideNYGk}[T'8.N (3.7.3.1.1.1A).

servative results when compared to a dynamic analysis. f...'

~.c.4.

22b. 4l For.tfie dynamic analysis of Seismic Category 1' Subsystems

~N.-

. (3.7.3.1.1.2A) provide an additional explanatico of the method and in. T dicate if the following items have been considered..

(a)

Consideration of tne torsional, rocking, and trans- '

~~

1ational responses of the structures and their foundations.

(b)

Use of an adequate number of masses or degrees of freedom in dynamic modeling to determine the response of all applicable components and plant ecripment.

The number is considered adequate when ad.ditiona').

degrees of freedom do.not result in more than a 10%

increase in responses. Alternately, the number of degrees of freedom may be taken equal to twice the number of modes with frequencies less than 33 cps.

(c)

Investigation of a sufficient number of undes.to assure participation of all significant cmdes. The criterion for sufficiency is that the inci3usion of additional modes -does.not result in moreJhan a 10%

increase in responses.

(d)

Consideration of maximum relative displagsments among supports of structures, systems, and components.

(e)

Inclusion of significant effects such as piping

~-

interactions, externally applied structural re-straints, hydrodynamic (both mass and stiffness. x --~

x c e i effects) loads,andnonlinearresponses2

• N 'W

~ ~ ~ ~ '

x i

.g.

G e

.. % -Z

^

. %}.;;.,

. _Q, -

~

~

.s ; - O c.: 1..

. g,. 9 y =. %.... :...:s s..

. g--

s._

,c ; ;.

A W:

Describe in~ detail the methods 'ofI seismic design and ;E;:.*' g;.. g@3.:-

n...

~

.n? -

.lT 220.47.

(3.7.3.12A) analysis, pertinent design criter-ia and results of you

.fWt design for buried structures, other than piping..whichf? w.....,e.. '.W.6 I house safety related components.or systems. Provide a J.;;~

._ g, r.jp:.- i

?.h diagram of the site' showing al1-such seismic Category ifr.~t' V... 3'Ap K l

_;;.r?..;.. J;._s : _ ;

._.... i #.

buried structures.,

. :-h.;

..49:b. -QT'q::5 c '.7%f.. :-n. ::]jfn-l....

m.

.m_

45 3.M 220.43 Provide a list of all structures ~and/or components of E M M

~ -:e'I.&

~

. 4 - c,.

structures which are designed in'a'ccordance with the.4p M.-. Mm -

(3.78)

.m.

. requirements of Section 3.7B'of the FSAR. For those'-$.

I.2.3k.?

structures and/or components of structures designed in g KT?-Q-j ?-#

efij accordance with Section 3.7B', de'ssribe the locacion of'.

T di the' interface with other~ strtrettfres and explain why theE:

interface was chosen in that-pTic~e- (unless it is otherwise~ ~ 6N

,M.T, i

~. '. _

~

readily apparant). For-any structures designed in.accor.;..j - '.??S

_., ~

e dance with 3.78,. discuss the differences in methods'ofM:N" W.... -e:.

design between 3.7A and 3.78, ~particularly with respect to

~

77F

.the degree of conservatism of;the design.

e-4.

. :1: ::

-r

' 7.i... -

Q...

220.44 Describe the seismic instrumentat' ion' surveillance' scheme

~.

(3.7.4) for the plant. The staff's requirement for such a pro-y.

gram is as follows:

SEISMIC MONITORING INSTRUMENTATION SURVEILI.ANCE REQUIREMENTS.

CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALIBRATION TEST 1.

Triaxial Time-History Accelerographs M

R.

SA 2.

Triaxial Peak Accelerographs NA-

~R NA 3.

Triaxial Seismic Switches M

R SA 4.

Triaxial Response-Spectrum Recorders M

R SA Legend:

l l

'~

Monthly M

=

~

Refueling.

~

R

=

Once per 18 months SA

=

Not Applicable

. ->c.

NA

=

- :~- '

.Each of the seismic instruments shall be demonstrated operable by the performance of the channel check, channel calibration, and channel :..,

.k

.g g.~

functional test operations at,the inte'rvals specified aboveZ W ^ W W N-2-

Explain and justify any deviations from the above..

g_

e

.6 e

e me -

- 6.-

[.y55,..

T rc.

A.

In._ tab.le e3. 8.1 de.f,ine the D.e.sigii _'and Operati.ng.ca.te

~

220, 45 (3.8.2) listed.inj.t.he. f;ir.st. column -in. terms. of ASMit Code service.-;.. -

.:J.c..-r.:

limit.s as..well as operating / environmental conditions;----- '

~

~l

~

i.

.4

- =.'

.. ~J u.

220. 46 Provide,co.nstruction, drawings showing the dome / cylinder"

~.,.. 9 - ~.

(3.8.2) int.er. face,.of.the conta.inment and..the mat / cylinder inter-.

z, t..,...

f;a.c,tof.;thef.co.n.ta.inment.... m 3 y.

. ; g g,y-

,].

_. ~.. -

..g.339....

220. 4?

P'dov'fde',5 con's;tir'uc'tiWdrqwTng' 'oIf tih'e. pola'r 'Tiane support.,;. h, ;,.3;...h_[..

7 (3.8.2) an'd,1,ts, attach, ment.tq-the. con _t.afnmen_t. u 3 w..

.; i f -.; s.

- -+..

...,.e - - ;- - w s w'...

....., s vidtidos~t'os.ind-f'iffdi ma' tier'iils"u' sed for 'd6 ors s[sfse~al's, Priodidi46till'i of,lth'e ' mat'e'riaTs. hs'ed fod'd-fin 220. 48

='

I and structural E '..-.J ' -

~

~

(3.8) a'ppleic'itidns 46 an'd 'b'et'we'e'n 'al.1: Sefsmic' C'atM 4 -Structures. W-$.? ^

Hap propedgon'ii.<(ida.t,' ion b'eeng'iyen' to 'ch'oosing maiterials.

., p; 4= : ~,?

such, that these items will not be subject to deterioration

.f"~. -M:w ~. -

." n from. radiation.as well..as all.other environmental factors?

E. '..'.,'i*;

P.rovi.d.6.4..t.idd.s.d.i.on.7 ': i. ! -. ~.

~.,.

. s:4

~

n t-m :-

~

~ 22.0,.;,43 The staff's' fnter'im'cr'iteria' concerning shell buckling is (3. 8. 2.')

as follows:

STAFF POSITION

- 'I h j ' 'S'T[RLICT. URAL ENGINEERING BRANCH

[,.

SAFETY FACTOR FOR STEEL CONTAINMENT SHELL BUCKLING

....u Under norma'l 'op' erat'ing condition, the steel containment should maintain a minimum of 3.0 safety factor for all loading ccmbinations. The safety factor (S.F) is defined as follows:

S.F. = Bucklino strenath of the containment shell Buck 11ng loao imposeo on tne sneil bhen design bases accident loads are considered, the safety factor should be minimum of 2.0.

The stsff position supersedes the safety factor provided in the following:

1) NE 3133 (NB 3133)*

2). NE 3222*

3) NRC SRP 3.8.2

• ASME Section III, Summer 1977 Addenda ~

Discuss t_he design of the River Bend containment with re e m.e.-uggh%r.cp_

spect to compliance yith the above.

~

l e:.

l e-o

(

2 j

. -.." 3.,_,

Q.;*

.. yp.g.: i&Q4 z y jGyf.

220.50.

P'rovide additional details showing how the containmentMC

.mf.9-t~

(3.8.2) liner is fastened to the case slab (i.e., welded studs,~ M '

~ ^ ~.?;F if used) in the field of the liner plate.. -

_ y.p- -

-],.

1 It is noted in Section 3.8.2.3.i~of the FSAR 'thidi'somei?';-.3NU.q:

'220251 r

(3.8.2.3.1) stress limits will.be. higher than those. indicated in e'

.. X.T c-Regulatory Guide 1.D.

Provide numerical comparisonsc_h

'..nG-R.:.2 ~

of hm much the guidanca of the Reg Guide is exceeded ~-E OJc.;n and state how you arried at your acceptance criteria.S-1.

_ ady,

values.-

. P 2.--w e'pw

..bw:.; #

~

220.52 For combining varicas dynamic loads, which may be applied -

- 3.33

?.y-T (3.8.2.4.1) simultaneously to the containment, it is the staff's w a..

. ? :9c position that the absolute sum mthod should be used'J.: '

. ~ ~:'! O i.e f J.

,;fE.f H.

employed.. If the latter method is to be adopted, de-@Q:.- '

~

~

.unless their actual time histories of occurrences are.

?

1;g.i..m.

tails of the method should be provided.

It is to be'.

noted that the method described in section 3 BA.B.4 of 1 GESSAR 238 has not been accepted by the staff. Discuss -

~

the compliance of the River Bend Containment design to the above.

22b.53 There is in the staff's position on MK III Containment -

(3.8.2.4.1) generic issues a fatigue analysis requirement far the

~

liner of concrete containment. For steel containment the consideration of' fatigue is specified in ASME Code Section o '

III Division 1.

However, the liner on the concrete foundation mat of a steel containment should be treated as the liner of a concrete containment. Since the staff position requires the pool liner to be designeci in accor-dance with the ASME boiic-r and pressure vessel code Division 1, subsection NE, it is. suggested that conditions and procedures to consicer f atigue of both the steel containment and the steel liner in the concrete contain-ment be established generically. State the prozedures and results by which fatigue was considered in the design and the containment ar.d mat liner.

220.54 Do you hahe any masonry walls at Riher Bend Station, the (3.8.4) failure of which could damage a safety related component or system?

If so, respond to this question.as. requested e

I in I&E Bulletin' 80-11 of May 8,1980, except that the

~

schedule for submittal is not applicable.

(The staff's

~

acceptance criteria for masonry construction is outs.

~

' ~

lined in Appendix A to SRP Section 3.8.4, NUREG-08000, Revision 0,datedJuly1981.)

t G-.

Ma:.xy~::.i.<N

~

mm The staff is aware th t the applicant intends to use a' "'"

y22b. 55 A

(3.8.4) particular type of drilled anchor bolt (Drillco - Maxi r1~-

Bolt) for anchoring all types of equipment to concrete structures throughout the plant. We are also aware that 1

l

.~

3 8

......: r:..

7

. +

w.

J.1.a' ~.+'.{.- Wf?> :.ts:'W3 W...

-.2-i.?. '

- ' 1-pf

.b, x $;_E 7 h Q h [

~..l.W^h. h-YS$heh$$6 N

• k

~

... h.. ^<

E-.h.... t. YK..'..

. :5...

the' applicant intends to use'a. lower factor-of-safety for.W~:cer...

these anchors than has been outlined in I&E Bulletin 79-022..;. t

_. 21 -

It is understood that the rationale for doing so is that-f#-

"'J

.J the Drillco Maxi Bolts are.100%. proof tested in place and;nk-

.y

~

also are backed by extensive laboratory tests and some'+M'M JEJ ? %

field experience. The staff's concern is that the proof;J,..&.. w..

.. O tests do not test the in-place failure cone of.the anchor #En "" :.%"-

bolts and that.the concrete, be'cause it is subject'to n.I ~ ~.W.

1. +

variability, of quality at any given place, is the weakih;3.V.

' 7:F+.

link. 1.aboratory tests cannot guarantee the behavior of3W'l. Cf" t in-sity concrete in a given location. - Accordingly, the? ?

applicant is requested to address the above concerns and :-r

.-1 provide' complete assurance that the anchor bolts will. M i.y. !~. _f }p.; -

perform as intended or increase the factor-of-safetyc'g 3.,,

f 3..'J.

for the anchor bolts.-

m. ~... % - : ~ 2. '

g-

. q.5 - g g.2: q.;r. w :.v 220.56 Provide a Desigo Report, as. outlined in Appendix.C to. 7 -

y

~

(3.8.2 SRP Section 3.8.4 (NUREG-0800, Rev. O, July 1981) for _

. (3.8.3 all Seismic Category 1 Structures. '

(3.8.4 (3.B.5 22b.57 Provide construction drawings of the spent fuel racks, (3.8.4) and the spent fuel pool and its liner.

220.58 The current staff position is that seisniic Category 1 (3.8.3)-

concrete structures, other than the containment, should (3.8.4) be designed in accordance with ACI 349 ano modified by URC Regulatory Guide 1.142, rather than ACI 318.

Indicate the~ instances where the design of the River Bend Seismic Category 1 concrete structures designs would be uncon-servative with respect to the staff's current criteria and provide an explanation.

i l

t e

1 a

A

~

~

y eF-

~

I e

s

.