ML17297A517

Forwards marked-up Draft SER Input for Mechanical Engineering.Also Forwards General Questions.Requests Agenda for 3-day Meeting in Norwalk,Ca to Discuss & Resolve Open Issues

ML17297A517
Person / Time
Site:	Palo Verde
Issue date:	06/22/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
References
NUDOCS 8106260375
Download: ML17297A517 (72)

Text

JUN 2 2 1981 Docket Nos.:

STN 50-528/529/530 Nr. E. E.

Van Brunt, Jr.

Vice President - Nuclear Projects Arizona Public Service Company P.

O. Box 21666 Phoenix, Arizona 85036

Dear Nr. Van Brunt:

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SUBJECT:

DRAFT NECHANICAL E(ENGINEERING INPUT TO PALO VERDE SER The NRC and tts contractor, Pacific Northwest Laboratory, have completed the review of the Palo >Verde FSAR through Amendment 3.

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

You are lequested to prepare an agenda for a meeting in which we can discuss and resolve the open issues in our revi'ew.

We anticipate this meeting being held over a 3-day. period at the Bechtel Offices in Norwalk. Califoinia.

After this meeting and any necessary follow-up, we will update the SER in-put into a form sufficiently clean for publication.

We expect this extended meeting to resolve almost all of these open issues.

Therefore, you should bring the,AE and utility people necessary to both discuss technical details and make binding commitments.

The enclosed draft SER contains those sections of 3.2, 3.6, 3.7, and 3.9::.."

appli'cable to the Nechanical Engineering Branch's scope of responsibility.

Section 5.2 will be reviewed separately at a later date.

Please contact us if you have any additional questions en this matter.

Sincerely, Original signed by Robert L Tedesoo Rober t L. Tedesco, Assistant Director for Licensing Division of Licensing cc:

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Docket Nos.:

STN 50-528/529/530 Mr. E.

E.

Van Brunt, Jr.

'Vice President - Nuclear Projects Arizona Public Service Company P. 0.

Box 21666

Phoenix, Arizona 85036

SUBJECT:

DRAFT MECHANICAL ENGINEERING INPUT TO PALO VERDE SER The NRC and its contractor, Pacific Northwest Laboratory, have completed the review of the Palo Verde FSAR through Amendment 3.

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

You are requested to prepare an agenda for a meeting in which we can discuss and resolve the open issues in our review.

We anticipate this meeting being held over a 3-day period at the Bechtel Offices in Norwalk, California.

After this meeting and any necessary follow-up, we will update the SER in-put into a form sufficiently clean for publication.

We expect this extended meeting to resolve almost all of these open issues.

Therefore, you should bring the AE and utility people necessary to both discuss technical details and make binding commitments.

The enclosed draft SER contains those sections of 3.2, 3.6, 3.7, and 3.9 applicable to the Mechanical Engineering Branch's scope of responsibility.

Section 5.2 will be reviewed separately at a later date.

Please contact us if you have any additional questions on this matter.

Sincerely, (Qi 5, ~

Robert L. Tedesco, Assistant Director for Licensing Division of Licensing cc:

See next page w/enclosure

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Mr. E. E.

Va'n Brunt, Jr.

Vice President - Construction Projects Arizona Public Service Company P. 0.

Box 21666

Phoenix, Arizona 85036 PALO VERDE jgg 2 2 1981 CC:

Arthur C. Gehr, Esq.

Snel1 5'Wilmer 3100 Valley Center

Phoenix, Arizona 85073 Charles S. Pierson Assistant Attorney General 200 State Capitol 1700 West Washington Phoenix, Arizona 85007 David N. Barry, Esq.,

Senior Counsel Charles R. Kocher, Esq., Assistant Counsel Southern California Edison Company P. 0.

Box 800

Rosemead, California 91770 Margaret Walker, Deputy Director of Energy Programs Economic Planning and Development Office 1700 West Washington Phoenix, Arizona 85007 William Primm Assistant Attorney General Bataan Memorial Building Santa Fe, New Mexico 87503 Resident Inspector Palo Verde/NPS U.S. Nuclear Regulatory Commission P. 0.

Box 21324

Phoenix, Arizona 85001 Ms. Patricia Lee Hourihan 6413 S. 26th Street Phoenix, Arizona 85040 Bruce Meyerson Arizona Center for Law in the Public Interest 112 North Fifth Avenue Phoenix, Arizona 85003

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PALO VERDE SER

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3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS 3.2.1 Seismic Classification Criterion 2 of the General Design Criteria requires that nuclear power plant structures,

system, and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety function.

These plant features are those necessary to assure:

1) the integrity of the reactor coolant pressure boundary;

2) the capability to shutdown the reactor and maintain it is a safe condition; or 3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to 10 CFT, Part 100 guideline exposures.

Structures,

systems, and compents important to safety that are required to be designed to withstand the effects of the safe shutdown earthquake and remain unctional must be classified as seismic Category I items in accordance with Regulatory Guide 1.29, "Seismic Design Classification".

All other structures,

systems, and components that may be required for operation of the facility have been designed to other than seismic Category I requirements.

This includes those portions of seismic Category I systems such as vent, drain and test lines on the downstream side of isolation valves which are not required to perform a safety function.

We have several questions regarding the seismic classification of systems and components.

Why is the liner for the spent fuel pool given a non-seismic classifi-cation?

Provide justification for the non-seismic classification of radioactive pumps and valves.

For the isolation valves on the HVAC system provide justification for their non-seismic classification.

Why are valves in "III-1" systems given a quality Group classification of B?

Based upon our review of FSAR Section 3.2.1 and subject to the satisfac.ory resolution of the open items, our findings will be as follows:

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Structures,

systems, and components important to safety that have been designed to withstand the effects of a safe shutdown earthquake and remain functional are identified in an acceptable manner in Table 3.2-1 of the Final Safety Analysis Report.

The basis for acceptance in our review has been conformance of the applicant's

designs, design criteria and design bases for structures,

systems, and components important to safety with the Commission's regulations as set for th in General Design Criterion 2 and in

'Regulatory Guide 1.20, "Seismic Design Classification," our technical positions and industry codes and standards.

We conclude that structures,

systems, and components impor'tant to safety that are designed to withstand the effects of a safe shutdown earthquake.

and remain functional have been properly classified as seismic Category I items in conformance with the Commission's regulations, the applicable regulatory

guides, and industry codes and standards and are acceptable.

Design of these items in accordance with seismic'ategory I requirements provides reasonable assurance that in the event of a safe shutdown earthquake, the plant will perform in a manner providi.ng adequate safeguards to the health and safety of the public.

3.2.2 S stem ualit Grou Classification Criterion 1 of the General Design Criteria requires that the nuclear power plant systems and components important to safety shall be designed, fabricated,

erected, and tested to quality standards commensurate with the importance of the safety function to be performed.

Fluid system pressure-retaining components important to safety must be designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety function to be performed.

Based upon our review of FSAR Section 3.2.2 and contingent upon satisfactory resolution of the open issues, our findings will be as follows:

The applicant has identified those fluid-containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems:

1) to

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prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure

boundary,

2) to permit shutdown of the reactor and maintain it in a safe shutdown condition, and 3) to contain radioactive material.

These fluid systems have been classified in an acceptable manner in Tables 3'-1 and 3.2-3 of the Final Safety Analysis Report and on system piping and instrumentation diagrams in the Final Safety Analysis Report based on conformance with Regulatory Guide 1.26, "guality Group Classification an'd Standards."

The applicant has applied guality Groups A, 8, C, and 0 in Regulatory Guide 1.26, "guality Group Classifications and Standards,"

to the fluid system pressure-retaining components important to safety.

Those components that are classified guality Group A, 8, C, or 0 have been constructed to the codes and standards identified in Table 3.2-2 of the Final Safety Analysis Report.

3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 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 or moderate energy fluid systems would have on adjacent safety-related components or structures has been analyzed with respect to jet impingement, pipe whip, and environmental effects.

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

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

3.6.2 Determination of Break Locations and D namic Effects Associated with the Postulated Ru ture of Pi in Our review under Standard Review Plan 3.6.2 was concerned with the locations chosen by the applicant for postulating piping failures.

We also

reviewed 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.

Standard Review Plan 3.6.2 also sets forth certain criteria for the analysis and subsequent in-service inspection of high energy piping in the break exclusion area of containment penetration.

Breaks need not be postulated in those portions of piping that meet the requirements of the ASHE Code,Section III, Subarticle NE-1120 and the additional design requirements outlined in Branch Technical Position MEB 3-1.

Additional in service inspection is also required for those portions of piping.

The following discussed open issues found in our review of FSAR Section 3.6.2.

It concludes with our finding contingent upon resolution of all open issues.

Pipe whip need only be considered in those high energy piping systems having fluid reservoirs with sufficient capacity to develop a jet stream.

The means for determining high and moderate energy lines is found in NRC Regulatory Guide 1.46 "Protection Against Pipe Whip Inside Containment."

This criteria has been used correctly by the applicant.

Some additional information in required to clarify this section.

How it is determined that the "internal energy level associated with whipping i s insufficient to impair the safety function of any system or component to an unacceptable level?",Details should be provided of any flow restrictors used.

Methods used to determine fluid reservoirs with sufficient capacity to develop a

jet stream should also be provided.

Standard Review Plan 3.6.2 outlines the procedures for choosing locations for breaks and cracks in high and moderate energy piping.

There are several points that the applicant has used for determining break and crack locations that require further clarification or justification.

For postulating breaks in high energy piping it is the staf,'s position that a break is to be postulated if the stress intensity range (including the

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zero load set) as calculated hy equation (10) and either equation (12) or We9 Gm (13) in ASME Code Section III, Paragraph HB-3653 exceeds &44-e for normal and upset plant conditions, including the OBE.

Assurances are needed that the applicant's criteria for choosing pipe break locations in high energy lines is in compliance with that from the SRP.

If not, justificatioh must. be provided as well as a list of those locations not in compliance.

All references to 10Ã of the code allowable for fatigue sho'uld be changed to a cumulative usage factor of.l.

Are there any Class 1 lines that penetrate containment?

Are the piping penetrations designed to the ASME Code?

Any break exclusion areas must be identified on the drawings and a list of the systems provided.

"A main st/earn line pi pe rupture is not postulated to occur between the containment penetration and the MSIVs nor between the double wall (designed as a pipe whip restraint) downstream from the MSIYs."

Is this part of the break exclusion area?

Any locations where piping restraints or supports are welded directly to the pipe need to be identified.

Provide an example of the analysis done on these attachments.

"Pressure-temperature

analysis, assuming a single non-mechanistic break of the main st)(earn lines, was performed to establish design parameters for the main stream support structure."

This needs to be expanded to give details of what actually has been done.

Are there any Class 1 and/or high energy lines outside of containment?

Justify why breaks at terminal ends of piping are postulated only if the piping is "located within the orotected structure"?

~ (afar'.l The SRP requires~separation of at least one pipe diameter for cicumferential breaks.

For longitudinal breaks the area must be at least equal to the cross sectional flow area of the pipe at the break location.

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Justification is required for any break. areas that do not comply with the above criteria.

In addition, any locations where break opening times greater than one millisecond have been used must be identified and justification provided.

In discussion on high energy piping other than the RCS main loop reference is made to an "appropriate ductility factor."

How is it used?

How is it determined that it is "appropriate'?"

On what systems has plastic analysis been used?

Provide an example of the plastic analysis and how it is done.

Any references to ".5 ultimate strain" should be changed to 50K ultimate strain.

The applicant references CENPD-168.

Assurances must be provided that break locations and pipe restraint locations, for the reactor coolant system, are the same as those in CENPD-168A.

The applicant must also provide assurances that their plant is designed within the parameters used in this report.

Are the pipe stops described in Section 4.2.1 of CENPD-168A the same as those used at Palo Verde?

Are the component supports used at Palo Verde the same as those discussed in Section 2.0 of CENPD-168A?

The forcing functions used in the analysis of the reactor coolant system for Palo Verde must be justified.

The applicant should also provide assurance that its procedure for choosing intermediate break locations in systems other than the RCS is in compliance with the SRP.

If not, additional justification will be required.

In the design of piping restraints for other than the RCS the amount of strain hardening used needs to be identified.

Any instances where an increase in yield strength of more that 10'. is used must be justified.

Provide a reference and a justification for the ductility ratios used.

Provide justification for the "dynamic increase factor."

"In other fluid system piping the energy absorbed by the pipe is taken into account only when it is significant."

More explanation of what is done here is needed.

Explain in detail and provide an example of a cantilevered compression restraint.

Also provide an example of a jet impingement restraint.

Show how they are designed and analyzed.

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"A summary of the dynamic analysis applicable to the RCS main loop piping and component supports that determine the loading resulting from the postulated RCS pipe break is covered in CESSAR Section 3.6.2.5."

The summary is not there.

Based on our review of FSAR Section 3.6.2 and subject to the satisfactory resolution of the identified open items, our findings will be as follows:

~~go go log (es The applicant has proposed a~i+ 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, components, 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 barriers, and other such devices will be used to mitigate the effects of these postulated piping failures.

~~+odatogies We have reviewed these ~~~

and have concluded that they provide for a spectrum of postulated pipe breaks and pipe cracks which includes the most likely locations for pi ping failures, and that the types of breaks and thei r effects are conservatively assumed.

We find that the methods used to design the pipe whip restraints provide adequate assurance that they will function properly in the event of a postulated piping failure.

We further p cMc4rlag ic S conclude that the use of the applicant's proposed pipe failure z~~ 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.

~eWod o (ops cs The applicant's ~~ comply with Standard Review Plan Section 3.6.2 and satisfy the applicable portions of General Design Criterion 4.

3.7.3 Seismic Subs stem Anal si s The review performed under Standard Review Plan Section 3.7.3 included the applicant's dynamic analysis of all seismic Category I piping systems.

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. ~ke In addition to loads this analysis also considers opc,rWln. +~~)~

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loadings such as For the dynamic analysis of seismic Category I piping, each pipe line was idealized as a mathematical model consisting of lumped masses connected by elastic members.

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

This includes the effects of torsional,

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

Next, the mode shapes and the undamped natural frequencies were obtained.

The dynamic response of the system was caluclated 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 of response sepctra were applied to each piping system support points.

The following discusses open issues found in our review of FSAR Section 3.7.3.

It concludes with our findings which are contingent upon the resolution of all open issues.

How many cycles of earthquake were used in the BOP scope of analysis?

The FSAR references BP-TOP-1 for combination of modal frequencies.

BP-TOP=1 states that closely spaced modes were only combined if they were associated with the same part of the system and in phase.

Provide an explanation of how phasing of modal responses can be obtained from a l

response spectrum analysis.

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Reference is made to the Component Factor Method.

In what instances was this method used?

How are modal frequencies combined before using this method.

Provide justification for the use of the Component Factor Method.

8ased on our review of FSAR Section 3.7.3 and subject to the satisfactory resolution of the identified open items, -our findings will be as follows:

The scope of the review of the seismic system and subsystem analysis for the Palo Ve'rde plant included the seismic analysis methods for all Category I systems and components.

It included review of procedures used for modeling and evaluating Category I systems and components.

The review included design criteria and procedures for evaluation of the interaction of

.non-Category I piping with Category I piping.

The review also included d

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p piping outside containment.

The system and subsystem analyses are performed by the applicant on an elastic basis.

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

When 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 frequenci es.

The square root of the sum of the squares of the maximum codirectional responses i s used in accounting for three components of the earthquake motion for both the time history and response spectrum methods.

We conclude that the seismic system and subsystem analysis procedures are criteria proposed by the applicant provide an acceptable basis for the seismic design of systems and components.

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3.9 MECHANICAL SYSTEMS AND COMPONENTS The review performed under Standard Review Plan Sections 3.9.1 through g ~+~ra I 3.9.6 pertains to the s+makeM integrity and operability 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 compone ts

&tnt R A4A such as control rod drive 'mechanisms, certain reactor internals, 4uc,4~

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, and any safety-related piping designed to industry standards other than the ASME Code.

We review such issues as load combinations, allowable stresses, methods of analysis, summary results, and pre-ooerational testing.

Our review must arrive at the conclusion that there is adequate assurance of a mechanical component performing its safety-related function under all postulated combinations 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 revieN 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.

Additionally, we have contracted with Pacific Northwest Laboratories to perform an independent analysis of a sample piping system in the Palo Verde plant.

This analysis will verify that the sample piping system meets the applicable ASME Code requi rements.

We will report the results of this independent piping analysis in a supplement to this Safety Evaluation Report.

Computer programs were used in the analysis of specific components.

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

Oesign control measures, which are required by 10 CFR Part 50, Appendix 8, require that verification of the computer programs also be included.

The applicant has not provided verification for all of the listed computer programs.

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Are there any Class 1 piping systems in the BOP scope of responsibility?

If so, provide a list of the design transients used.

The FSAR states that experimental stress analysis has been performed on some essential ducting.

A description of the analysis and a summary of the results is required.

"A component is assumed to be in the elastic range if yielding across a section does not occur."

How much yielding across a section is allowed?

Provide examples of where some net secti'on yielding has occurred.

Provide a list of those cases

where, using your criteria, component stresses exceed yield.

Provide a detailed example of the simplified methods of analysis used in those cases.

What computer code was used for the elastic-in-elastic time history analysis.

Reference is made to paragraph F-1223 of Appendix F of the 1974 ASNE Boiler and Pressure Vessel Code.

This paragraph is nonexistent.

What was actually used for the evaluation of faulted conditions for BOP systems and components?

Based upon our review of FSAR Section 3.9.1 and contingent on the satisfactory resolution of the open items, our findings will be as follows'he methods of analysis that the applicant has employed in the design of all seismic Category I ASHE Code Class 1,

2 and 3 components, component

supports, reactor internals, and other non-Code items are in conformance with Standard Review Plan 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 calculations 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.

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3.9.2 0 namic Testin and Anal sis of S stems, Com onents and E ui ment 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.

Seismic qualification of safety-related mechanical equipment will be reviewed by the Equipment gualification Branch.

Piping vibration, thermal expansion, and dynamic effects testing will be conducted during a preoperational testing program.

The purpose of these tests is to assure that the piping vibrations are within acceptable limits and that the piping system can expand thermally in a manner consistent with the design intent.

During the Palo Verde plant's preoperational and startup testing

program, the applicant will test various piping systems for abnormal steady-state or transient vibration and for restraint of thermal growth.

This test program must comply with the ASME Code,Section III, paragraphs NB-3622, NC-3622, and ND-3622 which require that the designer be responsible by observation during startup or initial operation, for ensuring that the vibration of piping systems is within the acceptable levels.

In addition, pipe whip restraint initial clearances will be checked, as will snubber response.

The test program should consist of a mixture of instrumented measurements and visual observation by qualified personnel.

The applicant will be required to provide a

summary of the results of this test, program upon its completion.

The applicant's discussion of the preoperational testing program in the FSAR is too general and should be redone.

More detail of what will actually be done must be provided.

What are the acceptance criteria for steady-state vibrations?

For transient vibration?

Will snubbers be checked?

To what transients will the piping be subjected?

Which lines, if any, will be instrumented?

If not instrumented, how will the visual observations be performed and on what size pipe lines?

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Will the preoperational. testing program for the Palo Verde reactor internals be in accordance with Regulation Guide 1.20 for prototype or non-prototype reactors?

Sased upon our review of FSAR Section 3.9.2.1 and contingent upon the satisfactory resolution of the open items, our findings will be as follows:

The vibration, thermal expansion, and dynamic effects test program which will be conducted during startup and initial operation on specified high and moderate 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 vi'brational dynamic effects due to va1ve closures, pump trips, and other operating modes associated with the design flow conditions.

In addition, the tests provide assurance that adequate clearances and free movement of snubbers exist 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 fulfillingthe applicable requirements of General Design Criteria 14 and 15.

The applicant has referenced CESSAR for FSAR Sections 3.9.2.3, 3.9.2.4, 3.9.2.5 and 3.9.2.6.

Our review of those sections can be found in the CESSAR SER.

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3.9.3 ASME Code Class 1, 2, and 3

Com onents, Com onent Su

orts, and'ore Su ort Structures Our review under Standard Review Plan Section 3.9.3 is concerned with h

their supports, and core supp'ort structures which are designed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, or earlier industrial standards.

This review is divided into three parts, each of'hich is discussed briefly below.

The first area of review is the subject of load combinations and allowable stresses.

The applicant has provided a commitment that all ASME Class 1, 2, and 3 components, component

supports, core support structures, control rod drive components, and other reactor internals have been analyzed or qualified in accordance with the referenced loading combinations.

There are however some areas requiring clarification.

What transients are used in the desing of Class 1 piping?

In the design of supports, how are reaction forces caused by differential anchor movement of the pipe considered?

In the discussion delete "conservatively" as a reference to the transients included in the upset conditions.

Based on our revi ew of FSAR 3.9.3.1 and contingent upon the satisfactory resolution o, the open issue, our findings will be as follows:

The soecified 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 load combinations and associated stress and deformation limits specified for ASME Cose Class 1,

2, and 3 components comply with Standard Review Plan Section 3.9 '

and satisfy the applicable portions oF General Design Criteria 1, 2, and 4.

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I'he second area of review in this section concerns the criteria used by the applicant in designing i'ts ASME Class 1, 2, and 3 safety and relief valves, their attached piping, and their supports.

,We have specifically reviewed the applicant's compliance with Regulatory Guide 1.67, "Installation of Overpressure Protection Devices."

Was Regulatory Guide 1.67 used in the design of the relief valves?

How many cycles were used in the fatigue analysis of the pressurizer safety valve inlet piping?

"In lieu of a time-history dynamic analysis, a static analysis has been performed which can be shown to be conservative."

Provide details of where this has been used and show that it is indeed conservative.

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 resultin'g 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.

The third area of our review in this section was the criteria used by the applicant in the design of ASME Class 1, 2, and 3 component supports.

All component supports have been designed in accordance with Subsection NF of the ASME Code,Section III.

We have reviewed the applicant's design criteria pertaining to buckling of component supports and the design of bolts used in component supports.

Pf%7CC Ju.Ms With respect to buckling, we find the applicant's ~~~ acceptable.

With respect to bolt design, 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 15

e r

concrete or steel structures This information has been submitted as a

response to our Office of Inspection and, Enforcement Bulletin 79-02, "Pipe Support Base Plate Designs 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.

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

Based upon our review of FSAR Section 3.9.3.4 and contingent upon resolution of the open items. our findings will be as fo'llows:

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,z< /; <J5 The specified design and service loading combinations used for the design of ASME Code Class 1, 2, and 3 component supports in systems clas'sified 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 structural integrity or supported component operability.

The design and service load combinations and associated stress and deformation limits specified by 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.

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3.9.4 Control Rod Drive S stems Our review under Standard Review Plan Section 3.9.4 covers 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 conditi'ons.

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

The applicant's FSAR references CESSAR for this section.

Our review of this section can be found in the CESSAR SER.

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3.9.5 Reactor Pressure Vessel Internals Our review under Standard Review Plan Section 3.9.5 is concerned with the lead combinations, allowable stress limits, and other criteria used in the design of the Palo Verde'.reactor internals.

The applicant's FSAR references CESSAR for this section.

Our review of this section can be found in the CESSAR.

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3.9.6 Inservice Testin of Pum s and Valves In Sections 3.9.2 and 3.9.3 of this Safety Evaluation Report we discussed the des'ign of safety-related pumps and valves in the Palo Verde 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 plant life.

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

These tests are performed in general accordance with the rules of Section XI of the ASME 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 measurements in order to detect long-term degradation of the pump or valve performance.

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

We 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 ASME Class 1, 2, and 3 components will be in accordance with the revised rules of 10 CRF, Part 50, Section 50.55a, paragraph (g).

The applicant has not yet submitted its program for the preservice and in-service testing of pumps and valves; therefore, 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 safety systems connected to the reactor coolant pressure oun ary a

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that have design pressure below the rated reactor coolant system (RCS) pressure.

There are also some systems which are rated at full 19

reactor pressure 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 an inter-system LOCA.

Pressure isolation valves are required to be category A or AC per INV-2000 and to meet the appropriate requi rements of IMV-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 will require corrective action; i.e.,

shutdown or system isolytion 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 to be approximately one year.

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

The staff's present position on leak rate limiting condi tions 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 valve would be an indication of valve degradation from one test to another 20

Leak rates higher than

.1 GPN will be considered if the leak rate changes are below 1

GPH above the previous test leak rate or system design precludes measuring 1

GPH with sufficient accuracy.

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

The Class 1 to Class 2 boundary wi 11 be considered the isolation point which must be protected by'edundant isolation valves.

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

When three or more valves provide'solation, 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.

21

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APPENDIX A QUESTIONS FOR PALO VERDE SER

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~ 'HB/NRC Palo Verde 3/30/81 3.2 Classification of Structures, Com onents, and S stems 3.2.1 Seismic Classification Page 3.2-10 Mhy is the spent fuel pool liner non-seismic?

Page 3.2-]6 Why are valves on "III-1" systems given a equality Group Classification of 8?

Page 3.2-17 Explain the non-seismic classification for radioactive pumps and valves.

Page 3.2-17 h'hy are the isolation valves for the HYAC given a non-seismic classification?

3.6.2 Oetermination of Break Locations and D namic Effects Associated with the Postulated Ru ture of Pi in 3.6.F 1.2.2, Page 3.6-32 This pipe break criteria is not in compliance with SRP 3.6.2.

Page 3.6-33 Loading condi tions for determining pipe break are normal and upset conditions plus the OBE.

Pa e 3.6-34 All references to 10K of the Code allowable for fatigue should be changed to a cumulative usage factor of.l.

Page 3.6-34 Are they any Class 1 lines that penetrate containment?

Any breaks exclusion areas must be shown on the drawings and a list of the systems provided.

Page 3.6-35 Identify any piping where pipe restraints or supports are welded to the surface.

Provide an example of the analysis done on these attachments.

Page 3.6-36 Are piping penetrations designed to the ASNE Code?

Page 3.6-36 "Pressure-temperature

analysis, assuming a single area non-mechanistic break of the main steam lines, was performed to establish design parameters for the main steam support structure."

This needs to be expanded to give details of what actually was done.

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"A main steam line pipe rupture is not postu')ated to occur between the containment penetrations and the HSIVs nor between the double wal1 (designed as a pipe whip restraint) downstream from the MISYs."

Is this part of the break exclusion area?

Page 3.6-36 Are there any Class 1 and/or high energy lines outside of containment?

Justify why breaks at terminal ends of piping are postulated only if the piping is "located within the protective structure".

3.6-.2.1.2.3, Page 3.6-38 Oetails must be provided of the location and the analysis used to limit the circumferential separation of a ruptured pipe.

What break opening times are used in the analysis of pipe breaks?

Times greater than one millisecond must be justified.

3.6.2.1.4, Page 3.6-42 How is the energy associated with a whipping pipe determined?

How i s it dete~ined that a ruptured pipe will not damage an impacted pipe to an unacceptable level?

3.6.2.1.5, Page 3.6-43

~z.~Id Upset conditions>include the OBE.

3.6.2.2.2, Page 3.6-43 What is an "appropriate ductility factor"?

How is it used/determined?

3.6.2.3.2, Page 3.6-44 Loading conditions for a plant, prior to rupture must include normal and upset loads plus the OBE.

How is a fluid reservoir having sufficient capacity to develop a jet stream determined?

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3.6.2 '.2, Page 3.6-45 On what systems has plastic analysis been used?

Provide an example of this analysis.

The reference to ".5 ultimate strain" should read "505 of ultimate strain".

3.6.2.3.3.1, Page 3.6-45 Provide assurances that break location and pipe restraint locations'or the reactor coolant system are the same as those in CENPO-168A.

Notes on CENPO-168-A 1.

Palo Verde must provide assurances that their plant is designed within the parameters used in this report.

2.

Are the pipe stops described in 4.2.1 the same as those at Palo Verde?

Are the component supports used in Palo Verde the same as discussed in 2.0?

3.

Palo Verde must justify the break opening times used and any displace-ment limited break areas.

4.

Palo Verde must justify the criteria used for determining intermediate breaks.

5.

The forcing functions used in the reactor coolant system analysis for Palo Verde must be validated.

3.6.2.3 '.2, Page 3.6-48 What level of strain hardening has been used in the analysis and design of restraints?

" In other fluid system piping the energy absorbed by the pipe is taken into account only where it is significant."

Provide an explanation of what is done-here.

Provide a reference and a justification of the ductility ratios used.

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Page 3.6-49 Provide justification for the DIF used.

Justify the procedure used to increase the yield stress for yielding members.

Explain in detail and provide an example of the canti levered compression restraint.

3.6.2.4, Page 3.6-50 If no guard pipe are used, why is whipping prevented by restraints?

3.6.2.5.l, Page 3.6-5l "A summary of the dynamic analyses applicable to the RCS main loop piping and component supports that determine the loadings resulting from the postulated RCS pipe breaks is covered in CESSAR Section 3.6.2.5."

The summary is not there.

General uestions What do jet impingement restraints look like?

How are they designed/

analyzed?

Provide examples.

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3.7.3 Seismic Subs stem Anal sis 3.7.3.2, Page 3.7-23 How many cycles of eyrthquake were used in the BOP scope of analysis' 3.7.3.3.2, Page 3.7-26 How are modal frequencies combined before using the Component Factor Method?

Provide justification for the use of this method.

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E 3.9 Mechanical S stems and Com onents 3.9.1.1, Page 3.9-1 Are there any Class I piping in the BOP scope? lf so, provide a list of design transients.

3.9.1.2, Page 3.9-1 All computer codes used in the design and analysis must be included in this list.

Verification is required for all codes.

3,9.1.3, Page 3.9-7 A description of the experimental stress analysis performed on the essential ducting is required.

3.9.1.4.2, Page 3.9-9 "A component is assumed to be in the elastic range if yielding across a section does not occur."

How much yielding across the section is allowed?

Provide examples of where some net section yielding has occurred.

Paragraph F-1223 of Appendix F of the ASNE Code doesn't exist.

What was used?

Provide a list of those cases where component stresses exceed yield.

Provide a detailed example of the simplified methods of analysis used in those cases.

What computer was used for the elastic-inelastic time history analysis.

3.9.2.1, page 3.9-12 Nore detail is needed for the pre-op testing program.

What, systems will be examined'?

What are the acceptance levels?

We will require details of the program an'd the results.

3.9.2',

Page 3.9-17 Wi 11 the preoperational testing program for the Palo Verde reactor internals be in accordance with Reg.

Guide 1.20 for prototype or non-prototype?

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3.9.3.1, Page 3.9-20 In the desia of supports, how are reaction forces caused by differential anchor movement of the pipe. considered?

h 3.9.3.1, Page 3.9-20 What transients are used in the design of Class 1 piping?

3.9 '.1.1.2, Page 3.9-20 Delete "conservatively" as a reference to the transients included in the upset conditions.

3.9.3.3, Page 3.9-70 Was Reg.

Guide 1.67 used in the design of the relief valves?

3.9.3.3.B.7, Page 3.9-73 How many stress cycles were used in the fatigue analysis of the pressurizer safety valve inlet piping?

3.9.3.3, Page 3.9-74

" In lieu of a time-history dynamic analysis,'

static analysis has been performed, which can be shown to be conservative."

Provide details of where this has been used and show that it is indeed, conservative.

3.9.6, Page 3.9-75 Provide a commitment that your inservice testing of ASi~1E Class 1,

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

Provide a schedule for the submittal of the 120 month program for in-service testing of pumps and valves.

Any request for relief from ASHE Section XI should be submitted as soon as possible.

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3.9.6 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor cooland 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 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 an inter-system 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 ASNE Code except as discussed below.

Limiting Conditions for Operation (LCO) are required to be added to the technical specifications 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 to be 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 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

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redundant pressure isolation function and give an indication of valve degradation over a finite period of time.

Significant increases over this limiting valve 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 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 independentlv 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.

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