ML17333A011

Forwards Mechanical Engineering Branch Draft SER Input, Based on Review of FSAR Through Amend 15.Requests Agenda for Meeting,Tentatively Scheduled During Wk of 810914,to Resolve Open Issues

ML17333A011
Person / Time
Site:	Columbia
Issue date:	07/23/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
NUDOCS 8108030470
Download: ML17333A011 (76)

Text

,0 8108030470 810723 PDR ADOCK 05000397 E

PDR Docket No. 50-397 JUL 7>3 fget Hr. R. L. Ferguson JJanaging Director llashington Public Supply System 3000 George ltashington May Richland, Mashington 99352

Dear Hr. Ferguson:

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DRAFT INPlJT TO MNP-2 SER The Mechanical Engineering Branch has completed the review of FSAR through Amendment 15.

lte have chosen not to develop a second round of questions but to proceed directly to a draft SER input.

Enclosed is a draft SER with questions generated in each section during our review and at the end is a sumIary of all these questions.

You should prepare an agenda for a meeting in which we can discuss and resolve the open issues.

Me anticipate this meeting being held over a 3-5 day period at a mutually agreeable site.

Me have tentatively scheduled this meeting for the week of September 14, '1981.

After this meeting and any necessary follow-up, we will update the SER input.

At this meeting we expect to resolve almost all the open i'ssues.

Therefore, you should have the NSSS, AE, and HPPSS people necessary to both discuss technical details and make binding commitments present at the meeting.

Me suggest the meeting be held at the Burns and Roe, Inc. offices in New Jersey.

If you have any questions, call R. Auluck, Project Manager at (301) 492-7702.

Sincerely.

Enclosure:

Draft SER AC@~ g@g

@abet rgg~II Robert L. Tedesco, Assistant Director for Licensing Division of Licensing cc w/enclosure:

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Do et No. 50-397 Nr. R. L.

erguson Managing D

ector Washington Pu lic Supply System 3000 George Wa ington Way Richland, Hashi ton 99352

SUBJECT:

DRAFT INPUT 0 WNP-2 SER DISTRIBUTION:

Docket File LB¹2 File

Attorney, OE DEisenhut/R rple RTedesco ASchwenc r RAuluc YLi RBo ak Hg ammer Service IgtE (3)

BCCS:

Local PDR NRC PDR NSIC TERA TIC ACRS (16}

The Mechanical Engineerin Branch has corn eted the review of the WNP-2 FSAR through Amendment 15.

)e have chos n not to develop a;second round of questions but to proceed rectly t a draft SER input.

Enclosed is a draft SER with the summary o

quest ons generated during our, review.

You should prepare an agenda for eting in which we can discuss and resolve the open issued.

He antic'te this meeting being held over a 3-5 day period at a mutually agr a

e site.

We have tentatively schedul-

'd this meeting for the week of epte er 14, 1981.

After this meeting and:any necessary follow-up, v will u ate the SER input.

At this meeting we expect to resolve almost 1 the open ssues.

Therefore, you should have the

NSSS, AE, and WPPS people necess y to both discuss technical details and make binding mmitments present at the meeting.

We suggest the meeting be held at t e Burns and Roes, In offices in New Jersey.

If you have any quest ns, call R. Auluck, Projec tfanager at (301) 492-7702.

Sincerely, cc:

e next page Robert L. Tedesco, Assist t Di -:ector for Licensing Divj'sion of Licensing OFHCE P SURNAME/

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NRG FORM 318 O040) NRCM 0240 OFFlCIAL RECOR D,COPY USGPO1 1951~960

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Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. 0.

Box 968 3000 George Washington Way Richland, Washington 99352 ccs:

Nicholas Reynolds, Esq.

Debevoise 8 Liberman 1200 Seventeenth'treet, N.W.

Washington, D.

C.

20036 Richard g. guigley, Esq.

Washington Public Power Supply System P. 0.

Box 968

Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympis, Washington 98504 Mr. Albert D. Toth Resident Inspector/WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. 0.

Box 69

Richland, Washington 99352 Mr. G.

C. Sorensen, Licensing Manager Washington Public Power Supply System P. 0.

Box 968

Richland, Washington 99352 Mr. 0.

K. Earle, Project Licensing Supervisor Burns and Roe Incorporated 601 Williams. Boulevard

Richland, Washington 99352

MNP-2 NUCLEAR PROJECT NO.

2 DRAFT SER 3.6.2 Determination of Break Locations and D namic Effects Associated with the Postulated Ru ture of Pi in The review performed under this section pertains to the applicant's program for protecting safety-related components and structures against the effects of postulated pipe breaks both inside and outside containment.

The effect that breaks or cracks in high and moderate energy fluid systems would have on adja-cent safety-related components or structures are required to be analyzed with respect to jet impingement, pipe whip, and environmental effects.

Several means are normally used to assure the protection of these safety-related items.

They include physical separation, enclosure within suitably designed structures, the use of pipe whip restraints, and the use of equipment shields.

Our review under Standard Review Plan Section 3.6.2, "Determination of Break-Locations and Dynamic Effects Associated with the Postulated Rupture of Piping",

was concerned with the locations chosen by the applicant for postulating piping failures.

Me also reviewed for the size and orientation of these postulated failures and how the applicant calculated the resultant pipe whip and jet impingement loads which might affect nearby safety-related components.

The following discusses several open issues in our review and concludes with our findings which are contingent upon resolution of these open issues.

a.

In order for us to complete our 'review, the applicant should provide a

summary of the data developed to select postulated break locations including, for each point, the, calculated stress intensity, the calculated cumulative usage factor, and the calculated primary plus secondary stress range.

This data is required for review to ensure that the pipe break criteria have been properly implemented.

This data has not been submitted.

Figures 3.6-11 through 3.6-36 are not completed.

Therefore, review of these areas remains an open area.

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b.

Paragraph 3.6.2.1.1.1.b (2)(b)(Page 3.6-25) implies that the cumulative usage factor limit of 0.1 is considered only when high stress occurs.

It is the staff's position that breaks must be postulated at any location where the cumulative usage factor exceeds

0. 1.

At these locations both circumferential and longitudinal pipe breaks should be postulated, unless it can be clearly shown that the high usage factor is due primarily to stresses in only one principle direction.

The applicant response to question 110.012 states that the rules set forth in 3.6.2. 1.4. l.e (1) and (2) exempt certain break orientations based solely on stress and are inde-pendent of calculated cumulative usage factor.

Clarification of this area is required.

C.

Paragraph 3.6.2. l. 1. 1.b (2)(c)(Page 3.6-26) implies that breaks are postulated when the stress ranges as calculated by Equations 12 or 13 of the code exceed 2.4 S

and Equation 10 exceeds 3

S

. It is the staff's m

m'osition that if Eq. (10),

as calculated by Paragraph NB-3653, ASME Code Section III, exceeds 2.4 S, then Eqs.

(12) and (13) must be evaluated.

If either Eq.'12) or (13) exceeds 2.4 S, a break must be postulated.

In m'ther

words, a break is postulated if or Eq. (10)

> 2.4 S

and Eq.

(12)

> 2.4 S

Eq.

(10)

> 2.4 S

and Eq.

(13)

> 2.4 S

d.

For those portions of ASME,Section III, Class 1 piping discussed in FSAR Sections 3.6. 2. 12. 1 and designed to seismic Category I standards and included in the break exclusion area breaks need not be postulated providing all of the following criteria are met.

(1)

Eq.

(10) as calculated by Paragraph NB-3653, ASME Code,Section III does not exceed 2.4 S

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(2) If Eq.

(10) does exceed 2.4 S, then Eqs.

(12) and (13) must be m'valuated.

If neither Eq.

(12) or (13) exceeds 2.4 S, a break need m'ot be postulated.

In other words, a break need not be postulated if:

and Eq.

(10)

> 2.4 S

and Eq.

(12)

< 2.4 S

Eq.

(13)

< 2.4 S

(3)

The cumulative fatigue usage factor is less than 0. 1.

(4)

For plants with isolation valves inside containment, the maximum

stress, as calculated by Eq. (9) in ASHE Code Section III, Paragraph NB-3652 under the loadings of internal pressure, dead-weight and a postulated piping failure of fluid systems upstream or downstream of the containment penetration areas must not exceed 2.25 S

The above criteria areevaluated under loadings resulting from normal and upset plant conditions including the OBE.

In addition, augmented inservice inspection is required on all ASHE Class 1,

2 and 3 piping in the break exclusion area.

It is not clear whether footnote (a) on Page 3.6-28 of the FSAR is applicable to Section 3.6.2.1.2.2.

The applicant must provide assurances that their criteria for piping in the break exclusion areas complies with the requirements outlined above and those of Standard Review Plan 3.6.2.

A list of all systems included in the break exclusion areas must be included in the FSAR.

In addition, break exclusion areas should be shown on the appropriate piping drawings.

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e.

Any instances with limited break openings or break opening times exceeding one millisecond must be identified.

Any analytical

methods, representing test results or based on a mechanistic

approach, used to justify the above must be provided and explained in detail.

This applies to containment and annulus pressurization as well as general pipe break.

Paragraph 3.6.2.5.4. 11 c (Page 3.6-70) states, "A pipe break in one of the six lines, if unrestrained, may result in pipe whip impact with adjacent isolation valves, possibly rendering them inoperative.

Furthermore, unre-strained motion may cause impact with other lines, which may result in escalation of pipe breaks.

Such a condition may unacceptably increase the severity of the initial pipe break."

The way this paragraph is written, it is not apparent that sufficient protection has been provided to preclude the failure conditions discussed or whether these are failure conditions for which the protection was provided.

Clarification of this area is requested.

Subject to resolution of the above open issues, our findings are as follows:

The applicant has proposed criteria for determining the location, type and effects of postulated pipe breaks in high energy piping systems and postulated pipe cracks in moderate energy piping systems.

The applicant has used the effects resulting from these postulated pipe failures to evaluate the design of systems, componenets, and structures necessary to safely shut the plant down and to mitigate the effects of, these postulated piping failures.

The applicant has stated that pipe whip restraints, jet impingement bar riers, and other such devices will be used to mitigate the effects of these postulated piping failures.

We have reviewed these criteria and have concluded that they provide for a

spectrum of postulated pipe breaks and pipe cracks which includes the most likely locations for piping failures, and that the types of breaks and their effects are conservatively assumed.

We find that the methods used to design the pipe whip restraints provide adequate assurance that they will function

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properly in the event of a postulated piping failure.

Me further conclude that the use of the applicant's proposed pipe failure criteria in designing the

systems, components, and structures necessary to safely shut the plant down and to mitigate the consequences of these postulated piping failures provides reasonable assurance of their ability to perform their safety function following a failure in high or moderate energy piping systems.

The applicant's criteria comply with Standard Review Plan Section 3.6.2 and satisfy the applicable portions of General Oesign Criterion 4.

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3.7.3 Seismic Subs stem Anal sis The review performed under Standard Review Section 3.7.3 included the applicants dynamic analysis methods for all Seismic Category I systems, components, equipment and their supports.

It included review of procedures for modeling, use of floor response

spectra, inclusion of torsional effects, and determina-tion of composite damping.

The review has included design criteria and procedures for evaluation of the interaction of non-Category I piping with Category I piping.

The review has also included criteria and seismic analysis procedures for reactor internals and Category I buried piping outside contain-ment.

In addition to operating transient

loads, the analysis also considers abnormal loading such as an earthquake.

Piping was idealized by the applicant as a mathematical model consisting of lumped masses connected by massless elastic members.

The stiffness matrix of the piping system was determined using the elastic properties of the pipe.

The model included the effects of torsional,

bending, shear, and axial deformations as well as the change in stiffness due to curved members.

The dynamic response of the piping system was calculated by using the response spectrum method of analysis.

For a piping system which was supported at points with different dynamic excitations, the response spectrum analysis was performed using the envelope response spectrum of all support points.

Alternately, the multiple excitation analyses methods may have been used where separate acceleration time-histories or response spectra were applied to each piping system support points.

Relative displacement between anchor points was determined from the dynamic analysis of the associated structure.

The relative anchor point displacements were then applied to the piping model in a static anaylsis in order to deter-mine the secondary stresses caused by relative anchor point displacements.

Modal response spectrum multidegree of freedom and time history methods form the bases for the analyses of all major Category I systems and components.

Mhen the modal response spectrum method is used, governing response parameters are combined by the square root of the sum of the squares rule.

However, the absolute sum of the modal responses are used for modes with closely spaced frequencies.

P 4

The applicant's procedures for the dynamic analysis of Category I systems, components, equipment and their supports have been reviewed by us and found to be generally acceptable.

However, the following open issues must be resolved before we can report our findings.

a.

Paragraph 3.7.2. 1.8.2 (Page 3.7-15) stated that for the equivalent static load method, a minimum load factor of 1. 15 is applied to building acceler-ations to include the effect of higher modes of vibration.

The acceptance criteria of SRP 3.7.2 for the equivalent static load method is to apply a

load factor of 1.5.

A factor of less than 1.5 may be used if adequate justification is provided.

Justification for utilizing this reduced factor is required.

b.

Paragraph

3. 7. 3. 2. 1 of the FSAR states that "Based on Reference

3. 7-10 (BMR/6 General Electric Standard Safety Analysis Report, Volume 1, General Electric Company, 4/30/74),

which summarized data related to seismic histories presented in PSARs for many plants, it is conservatively assumed that combined effects due to seismic events of an intensity less than or equal to OBE intensity may be considered equivalent to two earthquakes of OBE intensity.

Therefore, the lifetime number of earthquake cycles may range from 200 to 600 assuming 30 seconds of strong motion earthquake acceleration for each seismic event."

Please provide clarification of this statement.

C.

Paragraph 3.7.3.2.2 arrives at only one OBE intensity earthquake for design of the NSSS systems and components.

Justification is required for this conclusion.

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3.9 Mechanical S stems and Com onents The review performed under Standard Review Plan Sections 3.9. 1 through 3.9.6

'ertains to the structural integrity a'nd functional capability of various safety-related mechanical components in the plant.

Our review is not limited to ASME Code components and supports, but is extended to other components such as control rod drive mechanisms, certain reactor internals, supports for ventila-tion ducting and cable trays, and any safety-related piping designed to industry standards other than the ASME Code.

Me review such issues as load combinations, allowable stresses, methods of analysis, summary results, pre-operational

testing, and inservice testing of pumps and valves.

Our review must arrive at the conclusion that there is adequate assurance of a mechanical component performing its safety-related function under all postulated combina-tions of normal operating conditions, system operating transients, postulated pipe breaks, and seismic events.

3.9. 1 S ecial To ics for Mechanical Com onents The review performed under Standard Review Plan Section 3.9.. 1 pertains to the design transients, computer programs, experimental stress analyses and elastic-plastic analysis methods that were used in the analysis of seismic Category I ASME Code and non-Code items.

The following discusses several open issues in our review and concludes with our findings which are contingent upon resolution of these open issues.

a.

In general, the transient conditions were reviewed and appear to be lacking with respect to the seismic transients.

No seismic transients are specified for the majority of the components and components for which they are specified require only one OBE cycle.

SRP 3.7.3 specifies that a minimum of 5 OBEs should be assumed.

b.

Paragraph 3.9.1.1 (Page 3.9-1) states, "The cycles due to SSE and OBE used in the fatigue analysis are shown in Table 3.7-4."

The title of

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Table 3.7-4 is "Reactor Building-Seismic Analysis Natural Frequency and Natural Period."

Reference to this Table appears to be in error.

Clarification is requested.

c.

Paragraph 3.9. 1. l. 13 (Page 3.9-14) states that the applicable seismic cyclic loading for operating basis earthquake is shown in Table 3.9-15.

This Table has not been completed and therefore remains an open item.

d.

Computer programs were used in the analysis of specific components.

A list of the computer programs that were used in the dynamic and static analyses to determine the structural and functional integrity of these components is included in the FSAR along with a brief description of each program.

Oesign control measures, which are required by 10 CFR Part 50, Appendix B, require that verification of the computer programs be included.

While the requir ed verification is provided for most computer programs, it is lacking for several.

The applicant must provide methods of verification for all of the listed computer programs.

e.

The computer code utilized in the analysis of the ECCS Pump Motor Rotor Shafts addressed in paragraphs 3.9. 1.2.4, ECCS Pumps and Motors, is not identified.

This code should be identified and data. presented for the validity and applicability for use of this code.

The Orificed Fuel Support experimental stress analysis discussed in Paragraph 3.9. 1.4.2.5 (Pages 3.9-19 and 20) is not adequate to establish the validity of this program.

Additional details concerning this test program are required.

In addition, it states that the allowable stress limits were arrived at by applying a 0.65 quality factor to the ASME Code allowables of 1.5 S

for upset.

The basis for the 0.65 factor is m

requi red.

g.

The statement is made in Paragraph 3.9. 1.4. 1.2, Hydraulic Control Unit, that "These stresses were obtained by assuming that two HCUs were braced

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together back to back..."

Are the units actually tied together as assumed?

Additional details are required.

Subject to resolution of these open'issues, our.findings are as follows:

The methods of analysis that the applicant has employed in the design of all seismic Category I ASME Code Class 1, 2, and 3 components, component supports, reactor internals, and other non-Code items are in conformance with Standard Review Plan Section 3.9. 1 and satisfy the applicable portions of General Design Criteria 2,4,14; and 15.

The criteria used in defining the applicable transients and the computer codes and analytical methods used in the analyses provide assurance that the calcu-lations of stresses,

strains, and displacements for the above noted items conform with the current state-of-the-art and are adequate for the design of these items.

3.9.2 D namic Testin and Anal sis The review performed under Standard Review Plan Section 3.9.2 pertains to the criteria, testing procedures, and dynamic analyses employed by the applicant to assure the structural integrity and operability of piping systems, mechanical equipment, reactor internals and their supports under vibratory loadings.

3.9.2. 1 Prep erational Vibration and D namic Effects Pi in Tests The preoperational vibration test program will be conducted during startup and initial operation.

The purpose of these tests is to confirm that the piping, components, restraints, and supports have been designed to withstand the dynamic loadings and operational transient conditions that will be encountered during service as required by the ASME Section III Code and to confirm that no unaccept-able restraint of nromal thermal motion occurs.

Me have identified the following open issues in our review.

10

a.

The applicant should provide a commitment in the FSAR stating that all required piping restraints, components and component supports have been installed in the piping system prior to testing.

b.

The applicant's proposed preoperational test program covers the vibration and dynamic effects.

However, the thermal expansion effects required in SRP 3.9.2. II-l.d, e, and f are not adequately addressed.

The thermal motion monitoring program should deal specifically with verification of snubber

movement, adequate clearances and gaps to allow free movement of the pipe during heat-up and cooldown and should include acceptance criteria and test procedure.

Additional information on this program is required.

c.

The applicant has not given a clear description of the acceptance criteria for steady-state piping vibrations.

The staff's position is that acceptance limits for vibration should be based on half the endurance limit as defined by the ASEM Code at 10'ycles.

d.

Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety-related systems and components, it is requested that the operability program for snubbers should be included and docu-mented by the preservice inspection and preoperational test program.

Me will require the applicant's response to the letter from R.

Tedesco to R.

Ferguson, "Preservice Inspection and Testing of Snubbers,"

dated triarch 6, 1981.

Subject to resolution of these open issues, our findings will be as follows:

The vibration, thermal expansion, and dynamic effects test program which will be conducted du~ing startup and initial operation on specified high and mod-erate energy piping, and all associated

systems, restraints and supports is an acceptable program.

The tests provide adequate assurance that the piping and piping restraints of the system have been designed to withstand vibrational dynamic effects due to valve closures, pump trips, and other operating modes associated with the design basis flow conditions.

In addition, the tests provide assurance that adequate clearances and free movement of snubbers exist 11

b

for unrestrained thermal movement of piping and supports during normal system heatup and cooldown operations.

The planned tests will develop loads similar to those experienced during reactor operation.

This test program complies with Standard Review Plan Section 3.9.2 and constitutes an acceptable basis for fulfilling, in part, the requirements of General Design Criteria 14 and 15, 3.9.2.4 Flow Induced Vibration Testin of Reactor Internals Flow-induced vibration testing of reactor internals should be conducted during the preoperational and startup test program.

The purpose of this test is to demonstrate that flow-induced vibrations similar to those expected during operation will not cause unanticipated flow-induced vibrations of significant magnitude or structural damage.

Reactor internals for WNP-2 are substantially the same as the internals design configurations which have been tested in prototype BWR/4 plants.

The only exception is the jet pumps, which are of the BMR/5 design.

The vibration measurement and inspection program has been conducted in the Tokai-2 plant, to verify the design of the jet pumps with respect to vibration.

WNP-2 reactor internals will be tested in accordance with provisions of Regulatory Guide 1.20, Revision 2 for nonprototype, Category IV plants using Tokai-2 as the limited valid prototype.

The applicant has referenced G. E. Topical Report "Assessment of Reactor Internals Vibration in BMR/4 and BWR/5 Plants" NEDE-24057-P (Class III) and NEDO-24057 (Class I), October 1977 which also contains information on the jet pump vibration measurement and inspection programs performed in the Tokai-2 plant.

Me have reviewed this report and find it to be acceptable.

The preoperational vibration program planned for the reactor internals provides an acceptable basis for verifying the design adequacy of these internals under test loading conditions comparable to those that will be experienced during operation.

The combination of tests, predictive analysis, and post-test inspection provide adequate assurance that the reactor internals will, during their service lifetime, withstand the flow-induced vibrations of reactor operation without loss of structural integrity.

The integrity of the reactor 12

III

internals in service is essential to assure the proper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit safe reactor operation and shutdown.

The conduct of the preoperational vibration tests is in conformance with the provisions of Regulatory Guide 1.20 and Standard Review Plan Section 3.9.2, and satisfies the applicable requirements of General Design Criteria 1 and 4.

3.9.2.5 D namic Anal sis of Reactor Internals under Faulted Conditions The applicant has presented inadequate data to verify the mathematical models for the dynamic analysis.

Specifically an explanation of the dynamic model is requested and justification of the statement that "Only motion in the vertical direction will be considered her e;

hence, under structural member can only have an axial load."

3.9.3 ASME Code Class 1

2 and 3

Com onents Com onent Su orts and Core Su ort Structures Our review under Standard Review Plan Section 3.9.3 is concerned with the structural integrity and functionability of pressure-retaining components, their supports, and core support structures which are designed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel

Code,Section III, or earlier industry standards.

3.9.3. 1 Loadin Combinations Desi n Transients and Stress Limits a ~

The loading combinations and stress limits used in the design of (1) all ASME Class 1, 2, and 3 systems, components, equipment and their supports, (2) all reactor internals, and (3) control rod drive components need to be clarified in the FSAR.

Section 3.9.3. 1 and the majority of Tables 3.9.2 (a) through 3.9.2 (a c) in the FSAR do not clearly define the loading combinations and stress limits.

We will require a concise summary (prefer-ably in table form) of this information.

This summary should include a

listing of all the loads which were considered for each service condition or load case plus the acceptance criteria.

Appendix 110-1 to NRC guestion 110.27 contains loading combinations and acceptance criteria 13

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applicable to all of the above

systems, components, equipment and supports.

Table 3.6-5 of the WNP-2 "Plant Design Assessment for SRV and LOCA Loads" presents information which is not completely acceptable.

We will require a commitment to the Appendix 110-1 mentioned above.

In addition, we will require a clarification of the applicability of Table 3.6.5, i.e.

are all of these loading combinations and acceptance criteria applicable to all of the systems, components, equipment, etc.

discussed in the first paragraph I

above?

b.

Several references are made in Table 3.9.2 (a) through 3.9.2 (ac) to allowable stresses for bolting.

Specifically, what loading combinations and allowable stress limits are used for bolting for (a) equipment anchorage, (b) component

supports, and (c) flanged connections?

Where are these limits defined?

c.

The applicant has not yet responded to guestion 110.27, Appendix 110-2, "Interim Technical Position Functional Capability of Passive Piping Components. "

d.

The methods of combining responses to all of the loads requested in a.

above is required.

Our position on this issue for Mark II plants is outlined in NUREG0484, Revision 1, "Methodology for Combining Dynamic Responses."

However, since the primary containment for the WNP-2 plant is a free-standing steel pressure vessel and the plant is in a higher seismic

zone, the staff will requi re that the criteria in Section 4 of NUREG-0484, Revision 1, "Criteria for Combinations of Dynamic Responses other than those of SSE and LOCA" be satisfied if the square root of the sum of the squares method of combining these responses is used.

(Reference Regulatory Position E (2) in the enclosure to a letter from J.

R. Miller, NRC to Dr.

G.

G.

Sherwood, G.E.,

"Review of General Electric Topical Report NEDE-24010-P,"

dated June 19, 1980.)

The conclusions of NUREG-0484 Revision 1 are based on the studies performed by GE in NEDE-24010-P and BNL in NUREG/CR-1330.

The applicant must demonstrate that an SRSS combina-tion of dynamic responses achieves the 84K non-exceedance probability level because of the differences in containment and seismic level which were not included in the earlier studies.

I

e.

The note in Table 3.9-2 (a) of the FSAR states that NSSS components designed to the upset plant condition (normal operating loads

+ upset transients

+.5 SSE) will meet the upset design condition limits without a fatigue analysis.

It is the staff's position that for all ASME Class 1

components a fatigue analysis shall be performed for all loading condi-tions.

The basis for deviating from this position should be provided for WNP-2.

If the WNP-2 position on this issue is implicit in the letter from W.

Gang to R.

Bosnak, "G.E. Position on Fatigues Analysis," dated January 15, 1981, provide the information requested in the letter from R.

Bosnak to W. Gang dated February 19, 1981.

f.

The safety relief valve discharge piping and downcomers are ASME Class 2

and 3 components, a fatigue analysis is not required in their design by the ASME Section III Boiler and Pressure Vessel Code.

A through wall leakage crack in these lines resulting from fatigue caused by SRV actuations and small LOCA conditions would allow steam to bypass the pressure suppression pool.

This could result in an unacceptable overpres-surization of the containment.

We, therefore, require that the applicant perform a fatigue evaluation on these lines in accordance with the ASME Class 1 fatigue rules.

g.

Table 3.9-1 specifies one OBE with 10 maximum load cycles per event in the table of plant events.

SRP 3.7.3 requires the use of 5 OBEs with 10 maxi-mum load cycles per event.

Justification of this reduced number of OBEs is requested.

Note - This justification was al,so requested in the review of Section 3.7.3.

h.

Table 3.9-2 (a) lists the allowable general membrane stress for the emergency loading conditions as 1.5 S

ASME Section III Figure 3224-1 m'pecifies this limit as the greater of 1.2 S

or S

What is the validity m

y of the usage of 1.5 S

Also, the 1.5 S

listed is 42300 psi.

1.5 x 26700 = 40050.

This table also specifies one of the loads for the emergency condition as maximum credible earthquake (Design Basis Earthquake) and one of the loads for faulted conditions as maximum credible earthquake.

These terms have 15

~ I tt A

Ili tt

not been previously defined and utilized.

Are these loadings the SSE loadings?

In Table 3.9-2 (a), it is noted that the support skirt and the shroud support legs have been evaluated for buckling, but the buckling limits are not specified.

The applicant should discuss the applicability of the criteria in FSAR Section 3.9.3.4, "Component Supports" to this table.

It is stated in Table 3.9-2 (a) that for the RPV Support (Bearing plate),

the allowable stress for emergency conditions is 1.5 x AISC allowable stresses and for faulted conditions

1. 67 x AISC allowable stresses.

The applicant should provide the basis for these numbers.

For the RPV stabilizer, the allowable stresses are also based on the AISC specification.

The allowable stress for the ROD is shown as 84,000 psi.

Mhat is the basis for this number?

For the faulted loading condition, the allowable stress is shown as the material yield strength.

Mhy is the difference from the previous faulted allowable stress of 1.67 x AISC allowable stress?

k.

Table 3.9-2 (b) shows the general membrane plus bending allowable stress for emergency conditions as l. 5 SA where SA = 1.5 S

and for faulted conditions as 2 SA.

What is the basis for these numbers?

The ASME Section III code Figure NB3224-1 specifies 1.8 S

or 1.5 S

for emergency m

y and Table F1322.2-1 specifies, 2.4 S

or 0.7 S

for components and 1.5 S

m U

m or 1.2 S

for component supports, for faulted conditions.

Y l.

Table 3 ~ 9-2 (e) shows the allowable for the emergency condition as P

< 3.0 S

Mhat is the significance and validity of this equation?

Table 3.9-2 (i) Item 9, Hanger Bracket Combined Stress.

In the method of analysis, it is stated that the load = (W

+ M

+

W ).33 and that the multiplier (.33) is added as a safety factor specified on the purchase part drawing.

Mithout being able to evaluate the intent of this analysis in detail, it appears that this factor,results in using only 0.33 of the 16

l]

I'

total weight to determine the stresses.

Additional details. of this analysis are requested.

n.

Table 3.9-2 (n) lists the calculated stresses and allowable stress for the ECCS Pumps.

The actual stress exceeds the allowable for the RHR suction nozzle.

Mhile the excess is small, it is not noted what stresses, normal, upset, emergency or faulted, are being computed, and what loads were considered in determining these stresses.

Additional information on the. stresses in this area is requested.

o.

In the discussion of the nozzle loads for the RCIC Pump on Page 3.9-50, it is not clear how the equation, F

~

+ M.

1 1< 1 F

M 0

0 is to be applies.

Is F. to be the maximum of F F

and F

and M. to be 1

x' z

i the maximum of M, M

and M?

Clarification is requested on this point.

x' z

p.

Table 3.9.2(s).

Justification is required for the usage of the AISC for the source of the allowable stresses and the source of the 1.6 S factor as the allowable stress.

An explanation is also requested for the allowable stress of 0.7 ULT being equal to 35000 psi.

If the material is 6061-T6'luminum as noted in note a, the ultimate strength per ASTM 8308 is 38000 psi so the allowable would be 0.78(38000)

= 26600 psi.

q.

Table 3.9-2(w).

An explanation is requested for the 1.5 S

and 2.25 S

m m

emergency stress limits and the 2

S and 3

S faulted stress limits.

m m

r.

Table 3.9-2(y) does not present adequate information for evaluation.

What is meant by stress limits for VI and VII, and what are the stresses being evaluated?

s.

Table 3.9-2(aa).

The stresses evaluated are the Normal and Upset and the faulted loading condition.

Mhy is there no emergency loading condition for this component.

17

l h,

'l f

We have contracted with the Energy Technology Engineering Center to perform an independent analysis of a sample piping system in the WNP-2 Plant.

This analy-sis will not only verify that the sample piping system meets the applicable ASME Code requirements, but will also provide a check on the applicant's ability to correctly model and analyze its piping systems.

The results of the above evaluations will be presented in a future supplement to this report.

Subject to resolution of the above open issues, our findings are as follows:

The specified design and service combinations of loadings as applied to ASME Code Class 1, 2, and 3 pressure retaining components in systems designed to meet seismic Category I standards are such as to provide assurance that, in the event of an earthquake affecting the site or other service loadings due to postulated events or system operating, transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the materials of construction.

Limiting the stresses under such loading combinations provides a conservative basis for the design of system components to withstand the most adverse combination of loading events without loss of structural integrity.

The design and service load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 components comply with Standard Review Plan Section 3.9.3 a'nd satisfy the applicable portions of General Design Criteria 1, 2, and 4.

18

L II fi

3.9.3.3 Desi n and Installation of Pressure Relief Devices We have reviewed the design and installation criteria applicable to the mounting of pressure relief devices used for the overpressure protection of ASME Class 1, 2, and 3 safety and relief valves.

We have specifically reviewed the applicant's compliance with SRP 3.9.3.

The response to guestion 110.031 in the

FSAR, Amendment 9 does not comply with the quidelines in Regulatory Guide 1.67, "Installation of Overpressure Devices" concerning dynamic load factor.

Paragraph 3.9.3.3.2, "Open Relief Systems,"

implies that there may be pressure relief devices of the WNP-2 plants which relieve to open discharge systems.

More information on what dynamic load factor was used and how it was determined is required.

In addition, the applicant is requested to provide a commitment that all of the information in Sections 3.9.3.3.2 and 3.9.3.3.3 of the FSAR are applicable to both NSSS and BOP supplied components.

Based upon our review of FSAR Section 3.9.3.3 and contingent upon the satisfactory resolution of the open items, our findings will be as follows:

The criteria used in the design and installation of ASME Class 1, 2, and 3

safety and relief valves provide adequate assurance

that, under discharging conditions, the resulting stresses will not exceed allowable stress and strain limits for the materials of construction.

Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a conservative basis for the design and installation of the devices to withstand these loads without loss of structural integrity or impairment of the overpressure protection function.

The criteria used for the design and installation of ASME Class 1, 2, and 3 overpressure relief devices constitute an acceptable basis for meeting the applicable requirements of General Design Criteria 1, 2, 4, 14, and 15 and are consistent with those specified in Regulatory Guide 1.67 and Standard Review Plan Section 3.9.3.

3.9.3.4 Com onent Su orts We have reviewed information submitted by the applicant relative to the design of ASNE Class 1, 2, and 3 component supports.

Our review included an assess-ment of the structural integrity of the supports and the effect of support deformation on the operability of active pumps and valves.

Our review has resulted in the following open issues:

a.

The applicant's response to NRC guestion 110.29 is not completely acceptable.

The revised paragraph 3.9.3.4 states, "In design of the reactor vessel support skirt as a plate and shell-type component support, the allowable compressive load was limited to 90 percent of the load which produces a stress equivalent to yield stress in the material, divided by the safety factor for the plant condition being evaluated.

The safety factor for the faulted condition was

1. 125.

The effects of fabrication and operational eccentricity were included in stress calculations."

This Implies that the reactor vessel support skirt was designed to an allowable compressive load of.8 material yield stress.

It is not clear how the applicant's design would meet the staff's acceptable allowable load of two-thirds of critical buckling load.

In addition, the applicant has assumed the critical buckling stress as the material yield stress at temperature.

This definition could result in a non-conservative value for critical buckling stress.

Critical buckling stress depends upon the configuration (including manufacturing effects) and the material proper-ties (elastic modulus, E and minimum yield strength S ) of the load bearing number.

Because both of these material properties change with temperature, the critical buckling stress should be calculated using the values of E

and S

at the temperature.

Y The applicant will be required to provide the basis for using the critical buckling stress as defined in the FSAR and to clarify how the design of the reactor vessel support skirt meets the staff's acceptable allowable load of two-thirds of the critical buckling load.

'20

b.

The applicant has supplied information concerning the design of the bolts and the baseplates as a response to our office of Inspection and Enforcement Bulletin 79-02.

The review of this information is being performed jointly by our office of Inspection and Enforcement and our office of Nuclear Reactor Regulation.

Me will report the results of our review in a supplement to this Safety Evaluation Report.

Subject to resolution of the above open issues, our findings are as fo11ows:

The specified design and service loading combinations used for the design of ASHE Code Class 1, 2, and 3 component supports in systems classified as seismic Category I provide assurance that', in the event of an earthquake or other service loadings due to postulated events or system operating transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the materials of construction.

Limiting the stresses under such loading combinations provides a conservative basis for the design of support components to withstand the most adverse combination of loading events without loss of'tructural integrity or supported component operability.

The design and service load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 component supports comply with Standard Review Plan Section 3.9.3 and satisfy the applicable portions of General Design Criteria 1, 2, and 4.

3.9.4 Control Rod Drive S stems Our review under Standard Review Plan Section 3.9.4 covered the design of-the hydraulic control rod drive system up to its interface with the control rods.

We reviewed the analyses and tests performed to assure the structural integrity and operability of this system during normal operation and under accident conditions.

We also reviewed the life-cycle testing performed to demonstrate the reliability of the control rod drive system over its 40 year 1 ife.'1

/l I'

The information presented in the FSAR, pertaining to the test programs which were conducted to verify the design, is inadequate to arrive at a conclusion as to whether the drives will function over the full range of temperatures,

pressure, loadings and misalignments as required.

Areas for which additional information is requested are:

a.

Paragraph 3.9.4.3 (Page 3.9-73) states that deformation is not a limiting factor in the analysis of the CRO's components since the stresses are in the elastic region.

This statement is not necessarily valid. It seems that elastic deformations and thermal deformations could possibly result in critical displacements.

Have these areas been considered in the analysis?

b.

Table 3.9-2(v)

(Page 3.9-167) lists the stress limit for faulted conditions as:

Sl..t = 1.2 S

= 1.2 x 16660 = 20000 psi., with a note:

Analyzed to emergency conditions limits.

Then in the column of Allowable Stress is listed 24990 psi.,

and a calculated stress of 22030.

The calculated stress is within the limits for an allowable stress of 24990 but not for an allowable stress of 20000 psi; Clarification is requested of this area (Reference Section 3.9.3. 1(a) of this Draft SER).

Subject to resolution of the above open issues, our findings are as follows:

The design cr iteria and the testing program conducted in verification of the mechanical operability and life cycle capabilities of the control rod drive system are in conformance with Standard Review Plan Section 3.9.4.

The use of these criteria provide reasonable assurance that the system will function reliably when required, and form an acceptable basis for satisfying the mechanical reliability stipulations of General Design Criterion 27.

22

~ ~

J t

3.9.5 Reactor Pressure Vessel Internals Our review under Standard Review Plan (SRP) Section 3.9.5 is concerned with the load combinations, allowable stress limits, and other criteria used in the design of the WNP-2 reactor internals.

7 Our review has resulted in the following open issues.

a.

Table 3.9-13 establishes stress intensity limits for the core support structure faulted loading conditions.

As this table is somewhat different than the limits from Section III Appendix F, what is the basis and justi-fication for Table 3.9-13?

Would the computed stresses be in compliance with the faulted condition limits of Section III Appendix F?

b.

It is the staff position that all BWRs under construction should document their actions being taken with respect to the problem of cracking of jet, pump holddown beams.

Me will require the applicant's response to the letter from R.

Tedesco to N. Strand, "Cracking of BWR Jet Pump Holddown Beam," dated August 5, 1980.

c.

We will require the applicant to provide a commitment to NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking."

Subject to resolution of these

issues, our findings are as follows:

The specified transients, design and service loadings, and combinations of loadings as applied to the design of the WNP-2 reactor internals provide reasonable assurance that in the event of an earthquake or of a system trans-ient during normal plant operation, the resulting deflections and associated stresses imposed on these reactor internals would not exceed allowable stresses and deformation limits for the materials of construction.

Limiting the stresses and deformations under such loading combinations provides an acceptable basis for the design of these reactor internals to withstand the most adverse loading events which have been postulated to occur during service lifetime without loss 23

f ki P

il

of structural integrity or impairment of function.

The design procedures and criteria used by the applicant in the design of the MNP-2 reactor internals comply with Standard Review Plan Section 3.9.5 and constitute an acceptable basss for satssfying the applicable requirements of General Oesign Criteria 1, 2, 4, and 10.

3.9.6 Inservice Testin of Pum s and Valves In Secti ons 3. 9. 2 and 3. 9..3 of thss Safety Evaluation Report we discussed the design of safety-related pumps and valves in the MNP-2 facility.

The design of these pumps and valves is intended to demonstrate that they will be capable of performing their safety function (open, close, start, etc.) at any time during the lant lif g

p life.

However, to provide added assurance of the reliability of these components, the applicants will periodically test all its safety-related pumps and valves.

The se tests are performed in general accordance with the rules of Section XI of the ASNE Code.

These tests verify that these pumps and valves operate successfully when called upon.

Additionally periodic measurements are made of various parameters and compared to baseline measure-ments in order to detect ion t g erm degradation of the pump or valve performance.

Our review under Standard Review Plan Section 3.9.6 covers the applicant's program f'r preservice and inservice testing of pumps and valves.

Me give particular attention to those areas of the test program for which the applicant requests relief from the requirements of Section XI of the ASME Code.

The applicant must provide a commitment that the inservice testing of ASNE Class 1

2 and

, 2, and 3 components will be in accordance with the revised rules of 10 CFR, Part 50 Section 50.55a, paragraph (g).

The applicant has not yet submitted its ro ram f r p

g o

the preservice and inservice testing of pumps and valves there o

erefore, we have not yet completed our review.

Any requests for relief from ASME Section XI should be submitted as soon as possible.

There are several safet y systems connected to the reactor coolant pressure boundary that have desi n

g pressure below the rated reactor coolant system (RCS) pressure.

There are also s ome systems which are rated at full reactor pressure

il

),

)l

on the discharge side of pumps but have pump suction below RCS pressure.

In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems.

The leak tight 'integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure

systems, thus causing the inter-systems LOCA.

Pressure isolation valves are required to be Category A or AC per IMV-2000 and to meet the appropriate requirements of IN/-3420 of Section XI of the ASME Code except as discussed below.

Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which wi 11 require corrective action; i. e.,

shutdown or system isolation when the final approved leakage limits are not met.

Also surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.

Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and for systems rated at less than 50K of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.

The testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, et'c.

The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of -valve degradation over a finite period of time.

Significant increases over this limiting value would be an indication of valve degradation from one test to another.

Leak rates higher than 1

GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes 25

x la l

i,f

measuring 1

GPM with sufficient accuracy.

These items will be reviewed on a case by case basis.

The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.

In cases where pressure isolation is provided by two valves, both will be independently leak tested.

When three or more valves provide isolation, only two of the valves need to be leak tested.

Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves.

Also discuss in detail how your leak testing program will conform to the above staff position.

We will report the resolution of these issues in a supplement to the Safety Evaluation Report.

26

t r']

SUMMARY

OF QUESTIONS FROM WNP-2 ORAFT SER Determination of Break Locations and Dynamic-Effects Associated with the Postulated Rupture of Piping In order for us to.complete our review, the applicant should.provide a

summary of the data developed to select postulated break locations including, for each point, the calculated stress intensity, the calculated cumulative usage factor, and the calculated primary plus secondary stress range.

This data is required for review to ensure that the pipe break criteria have been properly implemented.

Figures 3.6-11 through 3.6-36 are not completed.

It is the staff's position that breaks must be postulated at any location where the cumulative usage factor exceeds O.l.

At these locations, both circumferential and longitudinal pipe breaks should be postulated, unless it can be clearly shown that the high usage factor is due primarily to stresses in only one principle direction.

The applicant's response to g.110.012 states that the rules set forth in 3.6.2.1.4.1e (1) and (2) exempt certain break orientations based solely on stress and are independent of calculated cumulative usage factor. Clarification of this area is required.

For.ASME,Section III, Class 1 piping designed to seismic Category I standards, breaks due to stress are to be postulated at the following locations:

(1) If Eq. (10),

as calculated by Paragraph NB-3653, ASME Code Section III, exceeds 2.4 Sm, then Eqs.

(12) and (13) must be evaluated.

If either

, Eq.

(12) or (13) exceeds 2.4 Sm, a break must be postulated.

In other

words, a break is postulated if Eq. (10)>2.4 S

and Eq. (12))2.4 S

or Eq. (10)~2.4 S

and Eq. (13)~2.4 S

(2)

Breaks must also be postulated at any location where the cumulative usage factor exceeds O.l.

The above criteria is evaluated under loadings resulting from normal and upset plant conditions including the OBE.

Any deviations from the above criteria must be justified.

For those portions of, ASME,Section III, Class 1 piping discussed in FSAR Section 3.6.2.1.2.1 and seismic Category I standards and included in the break exclusion area breaks need not be postulated providing all of the following criteria are met.

(1)

Eq.

(10) as calculated by Paragraph NB-3653, ASME Code,Section III, does not exceed 2.4 Sm.

(2) If Eq. (10) does exceed 2.4 Sm, then Eqs.

(12) and (13) must be evaluated.

If. neither Eq.

(12) or (13) exceeds. 2.4 Sm, a break need not be postulated.

In other words, a break need not be postulated if:

Eq. (10)~2.4 S

and Eq. (12)<2.4 S

and Eq.

(13)C2.4 S

H k

k I

4 5.

6.

(3)

The cumulative fatigue usage factor is less than 0.1.

(4)

For plants with isolation valves inside containment, the maximum stress, as calculated by Eq,.

(9) in ASHE Code Section III, Paragraph NB-3552 under the loadings of internal pressure, deadweight and a postulated piping failure of.fluid systems upstream or downstream of the contain-ment penetration area must not exceed 2.25 S

The above criteria is evaluated under loadings resulting from normal and upset plant conditions including the OBE.

In addition, augmented inservice inspection is required on,all ASME Class 1,

2 and 3 piping in the break exclusion area.

It is not clear whether footnote (a) on page 3.6-28.of the FSAR is applicable to Section 3.6.2.1.2.2.

The applicant must provide assurances thattheir criteria for piping in the break exclusion areas complies with the requirements outlined above and those of Standard Review Plan 3.6.2.

A list of all systems included in the break exclusion areas must be included in the FSAR.

In addition, break exclusion areas should be shown on the appropriate piping drawings.

Any instances with limited break openings or break opening times exceeding one millisecond must be identified.

Any analytical

methods, representing test results or based on a mechanistic

approach, used to justify the above must be provided and explained in detail.

This applies to containment and annulus pressurization as well as general pipe break.

Expand paragraph 3.6.2.5.4.11c to provide assurance that sufficient protection has been provided to preclude the pipe break damage for main steam and reactor feedwater piping inside the main steam tunnel.

3.7.3 Seismic Subsystem Analysis 7.

8.

9.

Provide justification for.utilizing the load factor of 1.15 for the equivalent static load method.

The acceptance criteria of SRP 3.7.2 for the equivalent static load method is to apply a load factor of 1.5.

Provide clarification of the statement. in Paragraph 3.7.3.2.1 of the

FSAR, "Based on Reference 3.7-10, which summarized data.related to seismic histories presented in PSARs for many plants, it is conservatively assumed that combined effects due to seismic events of an intensity less than or equal to OBE in-tensity may be considered equivalent to two earthquakes of OBE intensity.

Therefore, the lifetime number of earthquake.cycles may.range from 200 to 600 assuming 30 seconds of strong motion earthquake acceleration for each seimsic event."

Provide justification for utilizing one OBE intensity earthquake for design of the NSSS systems and components in'aragraph.3.7.3.2.2.

Specifically, provide. justification that the information in Paragraph 3.7.3.2.2 is applicable to the WNP-2 site.

k, ij, 3.9 3.9.1 10.

12.

13.

14.

15.

16.

17.

Mechanical Systems and Components Special Topics for Mechanical Components No seismic transients are specified for the majority of the components and the components for which they are specified require only one OBE cycle.

Justification is required.

Paragraph 3.9.1.1, Design Transients, referring to Table 3.7-4, "Reactor Building-Seismic. Analysis Natural Frequency and Natural Period,"

appears to be in error.

Clarification is required.

Table 3.9-15, Applicable Seismic Cyclic Loading, is indicated as "Later."

Provide a schedule for its inclusion in the FSAR.

Methods. of verification are required for all NSSS computer codes used in the analysis.

All computer programs used in the design.

and analysis of systems and components within the BOP scope must be listed.

Methods of verification are required for all BOP programs.

The computer code utilized in the analysis of the ECCS Pump Motor Rotor Shafts addressed in Paragraph 3.9.1.2.4, ECCS Pumps and Motors, is not identified.

This code should be identified and data presented for the validity and appli-cabilityy for use of the code.

Provide additional details concerning tbe test program performed on the orificed fuel support to establish the validity of the program.

In addition, provide justification for using the allowable stress limits by applying a

0.65 quality factor to the ASME Code allowables of 1.5 Sm for upset condition.

Expand Paragraph 3.9.1.4.1.2 (page 3.9-18) to describe the actual mounting of the hydraulic control units and to justify the validity of the assumption utilized in the FSAR.

3.9.2 Dynamic Testing and Analysis 3.9.2.1 Preoperational Vibration and Dynamic Effects Piping Tests 18.

Provide a coamitment in the FSAR stating that all required piping restraints, components and component supports have been installed in the piping systems prior to testing.

19.

The applicant's preoperational test program covers the vibration and dynamic effects..

However, the thermal expansion effects required in SRP 3.9.2.II-l.d, e, and f are not adequately addressed.

The thermal motion monitoring program should deal specifically with verification of snubber

movement, adequate clearances and gaps to allow free movement of the pipe during heat-up and cooldown and should include acceptance criteria and test procedures.

Additional information on this program is required.

f, t

'r, f

I I

~ ~ 20.

21.

3.9.2.5 22.

The applicant.has not given a clear description of the acceptance criteria for steady-state piping vibrations.

The staff's position is that acceptance limits for vibration s)auld be based on half the endurance limit as defined by the ASME Code at 10 cycles.

Provide a response to the letter from R. Tedesco to R. Ferguson, "Preservice Inspection and Testing of Snubbers,"

dated March 6, 1981.

Dynamic Analysis of Reactor Internals under Faulted Conditions The applicant has presented inadequate data to verify the mathematical models for the dynamic analysis.

Specifically,.an explanation of the dynamic model is requested and justification of the statement that, "only motion in the vertical direction will be considered here;

hence, each structural member can only have an axial load."

3.9.3 ASME Code Class 1,

2 and 3 Components, Component

Supports, and Core Support Structures 3.9.3.1 23.

24.

25.

26.

Loading Combinations Design Transients and Stress Limits The loading combinations and stress limits used in the design of (1) all.ASME Class 1,

2 and.3 systems, components, equipment and their supports, (2) all reactor internals and (3) control rod drive components need to be clarified in the.FSAR.

Section 3.9.3.1 and the majority of Tables 3.9.2(a) through 3.9.2(ac) inthe FSAR do not clearly define the loading combinations and stress limits.

We will require a concise summary (perferably in table form) of this information.

This summary should include a listing of all the loads which were considered for each service condition or load case plus the acceptance criteria.

.Appendix 110-1 to NRC guestion 110.27 contains loading combinations and acceptance criteria applicable to all of the above

system, components, equipment and supports.

Table 3.6-5 of.the WNP-2 "Plant Design Assessment for SRV and LOCA.Loads" presents information which is not completely acceptable.

We will require a commitment to the Appendix 110-1 mentioned above..

In

addition, we will require a clarification of the applicability of Table 3.6-5, i.e. are all of the loading combinations and acceptance criteria in Table 3.6-5 applicable.to.all.of the systems, components, equipment, etc. discussed in the first paragraph above Several references are made in Table 3.9.2(a) through 3.9.2(ac),to allowable stresses for bolting.

Specifically, what loading combinations and.allowable stress limits are used for bolting for (a) equipment anchorage, (b) componet

supports, and (c) flanged connections.

Where are these limits defined?

The applicant has not yet responsded to guestion 110.27, Appendix 110-2, "Interim Technical Position-Functional Capability of Passive Piping Components."

The methods of combining-responses to all of the loads requested in (a) above is required.

Our position on this issue for Mark II plants is outlined in NUREG-0484, Revision 1,."Methodology for Combining Dynamic Responses."

However, since the primary containment for.the WNP-2 plant is a free-standing steel pressure vessel and the.plant is in a higher seismic zone, the staff will require that the criteria in Section 4 of NUREG-0484, Rev.

1, "Criteria for Combination of Dynamic Responses other than those of SSE and LOCA" be satisfied if the square root of the sum of,the squares method. of combining these responses is used.

(Reference Regulatory Position E (2) in the enclosure

\\ ~

I to a letter from J.

R. Miller, NRC to Dr. G. G. Sherwood, G.E.,

"Review of General Electric Topical Report NEDE-24010-P",

dated June 19, 1980).

The conclusions of NUREG-0484 Rev.

1 are based on the studies performed by GE in NEDE-24010-P and.BNL in NUREG/CR-1330.

The applicant must demonstrate that.an SRSS combination of dynamic responses achieves the 84K non-exceedance probability level because of the.differences in containment and seismic level which were not included in the earlier studies.

27.

28.

29.

30.

31.

32.

The note in Table 3.9-2(a) of the FSAR states that NSSS components designed to the upset plant condition (normal operating loads

+ upset.transients

+

.5 SSE) will meet the upset design condition limits without a fatigue analysis.

It is the staff's position that for all ASME Class 1 components a fatigue analysis shall be performed for all loading conditions.

The basis for deviating from this position should be provided for WNP-2.. If the WNP-2 position on this issue is implicit in the letter from W. Gang to R. Bosnak, "GE Position on Fatigue Analysis", dated January 15, 1981, provide the information requested in the letter from R. Bosnak to W. Gang, dated February 19, 1981.

The safety relief valve discharge piping and downcomers are ASME Class 2 and 3

components, a fatigue analysis is not required in their.design by the ASME.

Section III Boiler and Pressure Vessel Code.

.However, a through wall,leakage crack in these lines resulting from fatigue caused by SRV actuations and small LOCA conditions would,allow steam to bypass the pressure suppression pool.

This could result in an unacceptable overpressurization of the.containment.

We, therefore, require that the applicant perform a fatigue evaluation on these lines in accordance with the ASME Class 1 fatigue rules.

Provide justification for utilizing one OBE with 10 maximum load cycles specified in Table 3.9-1.

Provide the basis for utilizing the allowable general membrane stress for the emergency loading conditions as 1.5 Sm.in Table 3.9-2(a).

ASME Section III Figure 3.2.2.4-1 specifies this limit as the quater of 1.2 Sm or Sy.

This table also specifies one of the loads as maximum credible earthquake which has not been clearly defined.

In Table 3.9-2(a), it is noted that the support skirt and shroud support legs have been evaluated for buckling,,but the buckling criteria are not specified.

The applicant should discuss the applicability of the criteria in FSAR Section 3.9.2.4, "Component Supports" to this table.

Provide the basis.for utilizing.the allowable stress for emergency condition of 1.5xAISC allowable stresses and for faulted conditions of 1.67xAISC allowable stresses for the RPV support (bearing plate).

For the RPV stabilizer, the allowable stresses are also based on the AISC specification.

The allowable stress for. the rod is shown as 84,000 psi.

What. is the basis for.this number?

For the faulted loading condition, the allowable stress is shown as the material yield strength.

Why is the difference from the previous faulted allowable stress of 1.67xAISC allowable stress'

I lt,

la 33.

34.

35.

36.

37.

38.

39.

Table 3.9-2(b) shows the general membrane plus bending allowable stress for emergency conditions as 1.5 S

where S

= 1.5 Sm and for faulted conditions as 2 SA.

What is the basis fir these kumbers7 The ASME Section III code Figure NB3224-1 specifies 1.8 Sm or 1.5 Sy for emergency and Table F1322.2-1 specifies, 2.4 Sm or 0.7 Su for components and 1.5 Sm or 1.2 Sy for component supports, for faulted conditions.

Table 3.9-2(e) shows.the allowable for the emergency condition as Pe 4 3.0 Sm.

What is the significance and validity of this equation7 Table 3.9-2(i ) Item 9, Hanger Bracket C'ombined Stress.

In the method of analysis, it is stated that the load

=

(W

+

W

+

W ).33 and that the multiplier (.33) is added as a safety factor s)ecif Ied on the purchase part.

drawing.

Without being able to evaluate the intent of this analysis in detail, it appears that this factor results in using only -0.33 of the total weight to determine the stresses.

Additional details of this analysis are requested.

Table 3.9-2(n) lists the calculated stresses and allowable stress for the ECCS Pumps.

The actual stress exceeds the allowable for the RHR suction nozzle.

While the excess is small, it is not noted what stresses, normal, upset, emer-gency or faulted, are being computed,.and what, loads were considered in determining these stresses.

Additional information on the stresses in this area is requested.

In the discussion of the. nozzle loads for the RCIC Pump on page 3.9-50, it is not clear how the equation,

+Cl Fi Mi Fo Mo is to be applied.

.Is Fi to be the maximum of Fx, Fy and Fz and Mi to be the maximum of Mx, My and Mz?

Clarification is requrested on this point.

l Table 3.9-2(s).

Justification is required for the usage of the AISC for the source of the allowable stresses and the source of the 1.6 S.factor as the allowable stress.

An explanation is also requested, for the allowable stress of 0.7 ULT being equal to 35000 psi. If the material is 6061-T6 aluminum as noted in note a, the ultimate strength per ASTM 8308 is 38000 psi so the allowable would be 0.78(38000)

= 26600 psi.

Table 3.9-2(w).

An explanation is requested for the 1.5 Sm and 2.25 Sm emergency stress limits and the 2

Sm and 3

Sm faulted stress limits.

Table 3.9-2(y) does not present adequate information for evaluation.

What is meant by stress limits for VI and VII, and what are the stresses being evaluated'0.

Table 3.9-2(aa).

The stresses evaluated are the Normal and Upset and the faulted loading condition.

Why is there no emergency loading condition for this component.

3.9.3.3 Design and Installation of. Pressure Relief Devices

'41.

The response to guestion 110.031 in the

FSAR, Amendment 9, does Oot comply with the guidelines in Regulatory Guide 1.67, "Installation of Overpressure Devices" concerning dynamic load factor.

'Paragraph 3.9.3.3.2 of the FSAR

I 3.9.3.4 42.

43.

"Open Relief, Systems",

implies that there may be pressure relief devices of the WNP-2 plants which relieve to open discharge systems.

More information on what dynamic load factor was used and how it was determined is required.

In addition, the applicant is requested to provide a commitment that all of the. information in Sections 3.9.3.3.2 and 3.9.3.3.3 of the FSAR are applicable to both NSSS and BOP supplied components.

Component Supports The applicant's response to NRC guestion 110.29 is not completely acceptable.

Paragraph 3.9.3.4 implies. that.the reactor vessel support skirt was designed to an allowable compressive load of.8 material yield stress.

,It is not clear how the applicant's design would meet the staff's acceptable allowable load of.two-thirds of critical buckling load.

In.addition, the applicant has assumed the critical buckling stress. as the material yield stress at temperature.

Provide basis for this assumption.

The applicant has supplied information concerning the design of not only the bolts but also the baseplates into which. the bolts are inserted and which the bolts connect to the underlying concrete or steel structures.

Thi s information has been submitted as a response to our Office of Inspection and Enforcement Bulletin 79-02, "Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts".

The review of this information is being performed jointly by our Office of Inspection and Enforcement and our Office of Nuclear Reactor Regulation; We will report the results of our review in a supplement to this Safety Evaluation Report.

3.9.4 Control Rod Drive Systems 44

'5.

Paragraph 3.9.4.3 (page 3.9-73) states that deformation is not a limiting factor in the analysis of, the CRD's components since the stresses are in the elastic region.

This statement is not necessarily valid.

It seems that elastic deformations and thermal deformations could possibly result in critical displacements.

Have these areas been considered in the analysis?

Table 3.9-2(v)

(page 3.9-167) lists the stress limit.for faulted conditions as:

Slimit = 1.2 Sm

= 1.2 x 16660 =20000 psi, with a note:

Analyzed to emergency conditions limits then in the column of Allowable Stress is listed 24990 psi, and a calculated stress of 22030.

The calculated sWess is within the limits for an allowable stress of 24990 but not for an allowable stress of 20000 psi.

Clarification is requested of this area (Ref. Section 3.9.3.1(a) of this draft SER).

3.9.5 Reactor Pressure Vessel Internals 46.

Table 3.9-13 establishes stress intensity limits for the core support structure faulted loading conditions.

As this table is somewhat different than the limits from Sect~on III Appendix F, what is the basis and justif-ication for Table 3.9-13?

Would the computed stresses be in compliance with the faulted condition limits of Section III Appendix F?

1 t

E 47.

It is the staff position that all BWR's under construction should document their actions being taken with respect to the problem of cracking of jet pump holddown beams.

We will require the applicant's response to the letter from R. Tedesco to N. Strand, "Cracking of BWR Jet Pump Holddown Beam,"

dated August 5, 1980.

48.

Provide a commitment to NUREG-0619, "BWR Feedwater Nozzle and Control Rod Orive Return Line Nozzle Cracking."

3.9.6 Inservice Testing of Pumps and Valves 49.

There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure.

There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.

. In. order to pretect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems.

The leak-tight integrity of these valves must be ensured by periodic leak testing.to prevent exceeding the design pressure of the low pressure systems thus causing an intersystem LOCA.

Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code except as discussed below.

Limiting Conditions for Operation (LCO) are required to be added to the technical specification which will require corrective action; i.e.,

shutdown or system isolation when the final approved leakage limits are not met.

Also, surveillance requirements, which will state.the acceptable leak rate testing frequency, shall be provided in the technical specifications.

Periodic leak testing of each.pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and,for systems rated at less than 50/ of RCS design pressure

.each time the.valve has moved from its fully,.closed position unless justification is given.

.The testing interval should average approximately one year.

Leak testing should also be performed after. all disturbances to the, valves are complete,.prior to reaching power operation following a refueling

outage, maintenance, etc.

The staff's.present position on leak rate limiting conditioos for operation must be equal to or less than

.1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.

Significant increases over this limiting value would be an indication of valve degradation from one test to another.

Leak rates higher than 1

GPM will be considered if the leak rate changes are below,l GPM above the previous test leak rate or system design precludes measuring 1

GPM with sufficient accuracy.

These items will be reviewed on a case by case basis.

I f

H t

The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves In cases where pressure isolation is provided by.two valves, both will be independently

')eak tested.

When three or more valves provide isolation, only two or the valves need to be leak tested.

Provide a list of all pressure isolation valves included.in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves.

Also, discuss in detail how your leak testing program will conform to the above staff possition.

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