ML18017A162

Draft SER Input Re Mechanical Engineering

ML18017A162
Person / Time
Site:	Susquehanna
Issue date:	04/21/1980
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML18017A161	List: ML18017A160 ML18017A162
References
NUDOCS 8005080373
Download: ML18017A162 (40)

Text

SUS UEHANNA DRAFT SER 3.6 Protection A ainst 0 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 adjacent safety related components or structures have 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 structures, the use of pipe whip restraints, and the use of equipment shields'

Determination of Break Locations and D namic Effects Associated with the Postulated Ru ture of Pi in Our review under Standard Review Plan Section 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.

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

Break opening times should be assumed to be one millisecond.

Break opening times greater than one millisecond must be justified by experimental data or analytical theory.

A table of break opening times or assurance that times of one millisecond or less were used is necessary for completeness.

In order for us to complete our review, the applicant should provide isometric sketches which show the locations of postulated breaks in high energy piping within the drywell.

The applicant has not yet responded to our question 110.26 which requested some of this information.

We consider this an open issue.

Another open issue exists in the applicant's analysis for the effects of pipe breaks in non-seismic Category I piping.

The applicant has assumed that this piping will fail under the effects of the SSE.

Further assurances

must be provided to show that breaks or cracks in these fluid systems have been assumed in the worst case locations.

Pipe whip and jet impingement need only be considered in those high energy lines having reservoirs'ith sufficient capacity to develop a jet stream.

Our review cannot be completed until information showing how reservoirs having sufficient capacity to develop a jet stream are identified and a

list of these systems are provided.

The applicant has stated that a design basis for the Susquehanna SES is that a postulated pipe break inside containment (up to and including a rupture of the recirculation piping), in conjunction with the SSE, plus a

single failure will not prevent the plant from obtaining cold shutdown with no loss of containment integrity and no dose levels above the limits in 10 CFR 100.

The applicant further states that piping not designed to seismic Category I standards is assumed to fail under the effects of the SSE.

In the FSAR Section 3.6 the applicant states that pipe breaks inside and outside of containment are not postulated to occur simultaneously.

This statement appears to be in conflict with the stated design basis.

This inconsistency must be resolved.

The applicant has installed pipe whip restraints to prevent ruptured pipes from whipping into and damaging nearby safety-related equipment.

These pipe whip restraints are designed to withstand the resultant loads and remain intact to assure protection of nearby safety-related structures, systems and components.

The only area requiring further clarification is

the level of plastic strain allowed'in the restraint under loads due to the whipping pipe.

In those piping systems in which breaks are not postulated, an augmented inservice inspection program that commits to'100K volumetric examination of all welds in these systems is required.

The proposed augmented inspection program does not meet these requirements for ASME Class 1

piping 'or for weld-o-lets, half couplings, and socket welds.

Further discussions are required before the augmented inspection program can be accepted.

Additionally, credit for examination of these welds cannot be applied to the general examination of welds in the plant.

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 postu-lated 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.

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

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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 properly in the event of a postulated piping failure.

We 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 Design Criterion 4.

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

This analysis included not only seismic loads but also suppression pool vibratory loads where appropriate.

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.

If different excitations were present at different anchor and support points, the response spectrum analysis was performed using the response spectrum at or above the center of mass of the piping system.

This tends toward conserva-tive results since the response spectrum increases with building height.

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

The relative anchor point displace-ments were then applied to the piping model in a static analysis in order to determine the secondary stresses caused by relative anchor point displacements.

The applicant's procedures for the dynamic analysis of Category I piping 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.

Standard Review Plan Section 3.7.3, "Seismic Subsystem Analysis", requires 5 OBE's with a minimum of 10 cycles each to be utilized in fatigue 'eval-uation.

This requirement has not been met.

The applicant must justify this deviation from Standard Review Plan 3.7.3 or commit to meet our requirements.

Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic

Response

Analysis", outlines the procedures for combining modal responses.

Specifically, modes having frequencies falling within 10K of each other are defined as closely spaced modes and must be combined by the absolute sum method.

Our review of FSAR Section 3.7.3 cannot be completed until assurance is provided that this criteria has been met or that an equivalent level of safety has been achieved.

It should also be noted that Bechtel topical report BP-TOP-l, Revision 2, has not been accepted.

Revision 3 of this report has been accepted and, therefore, the FSAR should be clarified to reflect the differences between the two versions.

V The use of three components of dynamic motion was not a design basis requirement for the construction permit for this plant.

For a plant of this vintage, an acceptable approach involves calculating responses for the two horizontal and one vertical components of motion.

The largest horizontal response and the vertical response would then be combined using the absolute summation method.

The method described in the FSAR utilizes this method 'and is acceptable.

In FSAR Section 3.7a.2.3.2 the applicant has stated that "the two rotational coordinates about each rode point are excluded because the moment contribu-tion of rotary inertia from surrounding rodes."

Me require clarification of this statement.

Upon resolution of the above open issues, we will report our findings in a supplement to this Safety Evaluation Report.

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

1 through 3.9.6 pertains to the structural 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 components such as control rod drive mechanisms, certain reactor internals, cable trays, 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, seismic qualifica-tion, preoperational

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 combinations of normal operating conditions, system operating transients, postulated pipe breaks, and seismic events'

10 S ecial To ics for Mechanical Com onents In this section we reviewed the proper treatment and inclusion of transients in the fatigue analysis of ASME Code Class 1 components.

Our review also covered the computer programs used in the design of seismic Category I mechanical components.

We also reviewed any experimental or inelastic analytical techniques used by the applicant.

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

This analysis will verify that the sample piping system meets the applicable ASME Code requirements, and will also provide a check on the applicant's ability to correctly model and analyze its piping systems.

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

The design transients specified for use in the fatigue analysis of compo-nents are generally acceptable.

However, we have identified the following open issues in our review.

For the control rod drives and their housing, the list of design transients and their associated

'number of cycles is acceptable with the exception of the number of OBE's assumed.

Standard Review Plan 3.7.3, "Seismic Subsystem Analysis", requires 5 OBE's with a minimum of 10 cycles each to be utilized in fatigue evaluations.

Only 1

OBE has been assumed by the applicant in the analysis of the control rod drives.

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The above criteria also applies to the CRD housings, in-core housings, hydraulic control units, core supports, other reactor internals, reactor

vessel, support skirt, shroud support, shroud plate, MSIV's, SRV's, recircu-lation pumps, and recirculation gate valves.

It is not clear whether the applicant has considered the OBE in the fatigue evaluation of these items.

These apparent exceptions conflict with FSAR Table 3.7a-4 and a statement in Section

3. 7a. 3. 2 that "the OBE is an upset condition" and, therefore, must be included in fatigue evaluations according to ASME Section III.

The applicant must provide clarification of the consideration of OBE loads for NSSS and BOP scope Class 1 components to resolve these apparent con-flicts in the FSAR.

In FSAR Section 3.'9. l. 1.2 it is stated that for the CRD housing a scram with no buffer is considered a normal/upset condition with 1 cycle.

For the CRD the same event has 10 cycles.

The applicant must resolve this inconsistency.

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 calculations of stresses,

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

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13 D namic Testin and Anal sis of S stems Com onents and E ui ment Me have reviewed the criteria, testing procedures, and dynamic analyses employed by the applicant to assure the structural integrity and oper-ability of piping systems, mechanical equipment, reactor internals, and their supports under vibratory loadings.

This review covers four subjects, each of which is described briefly below.

During the Susquehanna plant's preoperational and startup test program, the applicant will test various piping systems for abnormal steady-state or transient vibration and for restraint of thermal growth.

This test program will comply with the ASME Code,Section III, paragraph NB-3622-3 and NC-3622, which requires that the designer be responsible, by observa-tion during start-up or initial operation, for ensuring that the vibration of piping systems is within acceptable levels.

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

This test program will consist of a mixture of instrumented measurements and visual observation by qualified personnel.

Our review has uncovered a few open issues which require resolution.

For NSSS piping, the applicant's acceptance criteria for piping vibration is that the stress due to pressure and the measured steady state or transient vibration should be less than Service Limit B.

For common ferritic steels at 600'F, this stress limit would be about 30 ksi.

Me believe it more appropriate to have a limit on stress due to vibration alone (neglecting pressure) which is related to the material endurance limit in some fashion.

14 The other open issue in this review area is the scope of the test program for BOP piping.

Further detailed discussions will be required to resolve this issue.

Also, we will require the applicant to provide a brief summary of the results of this test program upon its completion.

Subject to resolution of these open issues, our findings are 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 associate'd

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 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 oper-ation.

This test program complies with Standard Review Plan Section

3. 9. 2 and constitutes an acceptable basis for fulfilling, in part, the require-ments of General Design Criteria 14 and 15.

15 During the course of the design and construction of the Susquehanna

plant, the applicant has specified that safety-related mechanical equipment be qualified to be able to perform their safety function during the safe shutdown earthquake and other dynamic events.

In those instances where no specific action of the component is required, only the structural integrity of the component need be assured.

In many cases,

however, the mechanical component must perform a physical movement of some kind to perform its safety function and is necessary for safe shutdown of the plant.

These components are termed "active" components.

Not only their structural integrity, but also their operability, must be demonstrated.

The oper-ability of active pumps and valves is discussed in Section 3.9.3 of this Safety Evaluation Report.

Our review of the dynamic qualification of mechanical equipment is incomplete and will be discussed in a supplement to this Safety Evaluation Report.

The applicant has committed to test the reactor internals in accordance with the provisions of Regulatory Guide 1.20, "Comprehensive Vibration Assessment Program for Reactor Internals During Preparational and Start-Up Testing", Revision 2, for non-prototype Category I plants.

Test procedures will require operation of the recirculation system at rated flow with internals installed (except for fuel).

Test duration will assure that a

minimum of 10 cycles of vibration will be experienced by the critical components during two-loop and single-loop operation of the recirculation system.

At the completion of the flow tests, the vessel head will be removed and the internals will be inspected for evidence of vibration,

wear, and loose parts.

The inspection will cover all components which were examined on the prototype design including the shroud, shroud

head, core support structures, jet pumps, control rod drive, in-core guide

tubes, and lower plenum.

The test results wi 11 be compared with the analytical results.

Our review has resulted in the following open issues.

The applicant has referenced GE topical report NEDE-24057-P, "Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants".

Our review of this report is not yet complete.

We will require the applicant to provide a brief summary of the results of this test program upon its completion.

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

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 analy-

sis, 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 internals in service is essential to assure the proper positioning of reactor fuel assemblies and unimpaired operation of

17 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.

The applicant has analyzed its reactor internals and unbroken loops of the reactor coolant pressure

boundary, including the supports, for the combined loads due to a simultaneous loss-of-coolant accident and safe shutdown earthquake.

We cannot complete our review in this area until the applicant submits the information requested in question 110.32.

We have also requested in question 110.41 that the applicant provide response time histories at one key location for each of the following internal components:

jet pump shroud wall shroud head control rod instrument guide tube core plate This location should be the one having either the maximum stress combina-tion or the most critical deflection combination, whichever governs the design.

Separate responses are required for the loads associated with the

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18 SSE and the most severe pipe break event.

The method of combining the dynamic responses should be outlined and justified.

Our review of this information is incomplete.

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

The dynamic system analysis performed by the applicant provides an acceptable basis for confirming. (he structural design adequacy of the reactor internals and unbroken piping loops to withstand the combined dynamic loads of a postulated loss of coolant accident (LOCA) and the safe shutdown earthquake (SSE).

The analysis provides adequate assurance that the combined stresses and strains in the components of the reactor coolant system and reactor internals do not exceed the allowable stress and strain limits for the materials of construction, and that the resulting deflections or displace-ments at any structural elements of the reactor internals will not distort the reactor internals geometry to the extent that core cooling may be impaired.

The methods used for component analysis have been found to be compatible with those used for the systems analysis.

The proposed com-binations of component and system analyses are, therefore, acceptable.

The assurance of structural integrity of the reactor.internals under combined LOCA and SSE conditions provides added confidence that the design will withstand a spectrum of lesser pipe breaks and seismic events.

Accomplishment of the dynamic system analysis constitutes an acceptable basis for complying with Standard Review Plan Section 3.9.2 and for satisfying the applicable requirements of General Design Criteria 2 and 4.

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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 wi'th the structural integrity and operability of pressure-retaining components, their supports, and core support structures which are designed in accor-dance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, or earlier industry standards.

This review is divided into four parts, each of which is discussed briefly below.

The first area of rev'iew is the subject of load combinations and allowable.

stresses.

Our required loading combinations for the Susquehanna plant have been included in question 110.42.

Mith one exception 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.

This one exception is related to our position that for load cases 1

and 2, as identified in question 110.42, all ASME Code Service Level B requirements are to be met, including fatigue usage factor requirements, and should take into account all SRV discharge load effects (initial actuation and continuous suppression pool vibratory) taken for the number of cycles consistent with the 40 year design life of the plant.

On the other hand, the applicant states that its ASME Class 1 piping fatigue calculations have considered OBE and SRV loads separately, not in a combined fashion as we require.

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l 20 Me consider the subject of seismic sloshing loads in the Susquehanna suppression pool to be an open issue because the applicant has not responded to our question 110.4.

Another open issue related to load combinations is the applicant's method for combining peak responses to multiple dynamic loads.

The applicant has used the "square root of the sum of the squares" method (SRSS) for all dynamic loads.

Our position, as outlined in NUREG-0484, "Methodology for Combining Dynamic Responses",

is that the SRSS method is acceptable for combining peak dynamic responses due to LOCA and SSE.

For other dynamic loads we are currently preparing a generic position which should be available in the near future.

The applicant has not yet responded to question 110.53 concerning the use of ASME Code Cases.

This is an open issue.

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

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

The second area of our review under Standard Review Plan Section 3.9.3 is the manner in which the applicant has assured the operability of active pumps and valves.

Active pumps and valves are those which must perform a mechanical motion in order to shut down the plant or mitigate the conse-quences of an accident.

For instance, under accident conditions, certain valves must open or close, and certain pumps are required to start.

On the other hand, inactive pumps and valves are only required to maintain their position during an accident.

Our review has resulted in two open issues concerning the applicant's operability assurance program for active valves.

The applicant's response to question 110.22 stated that the 20 inch Lunkenheimer recirculation loop gate valves were qualified by similarity to a 4 inch Anchor/Darling gate valve.

Our review has not concluded that these valves are, indeed, similar.

We require the applicant to either justify this similarity or qualify the 20 inch valve prototypically.

The applicant has stated in the FSAR that as long as a valve and its internal parts remain elastic, valve deformations will remain so small

22 that valve operability will not be impaired.

We do not necessarily agree.

It is our position that even elastic deformations must be checked to provide assurance of the valves'ot binding.

Similarly, elastic deforma-tions in active pumps should also be checked.

Additionally, our Seismic qualification Review Team will review the seismic qualification of the electrical motors,

switches, and other appurtenances attached to these active pumps and valves.

Subject to resolution of the above open issues, our findihgs are as follows.

The component operability assurance program for ASME Code Class 1, 2, and 3 active valves and pumps provides adequate assurance of the capability of such active components (a) to withstand the imposed design and service loads without loss of structural integrity, and (b) to perform necessary "active" functions (e. g., valve closure or opening, pump operation) during postulated events and conditions expected when plant shutdown is required.

The specified component operability assurance test program complies witb Standard Review Plan Section 3.9.3 and satisfies the applicable portions of General Design Criteria 1, 2, and 4.

The third area of review in this section concerns the criteria used by the applicant in designing its 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."

As noted above, we have not completed

23 our review of the seismic qualification of active valves in general, or of safety and relief valves in particular.

Also, as noted above, open issues remain in the areas of load combinations and methods for combining dynamic responses.

Subject to resolution of these open issues, our findings are 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'ressure 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 fourth 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.

Some supports for ASME Code components have been designed in accordance within Subsection NF of the ASME Code,Section III.

The remaining supports for ASME Code components were designed to remain elastic under maximum

24 loads.

Subsection NF could not be used because it had not yet been published at the time these supports were being designed.

This is an acceptable approach.

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

With respect to buckling, the applicant has not yet responded to our question 110.43.

With respect to bolt design, the'pplicant has supplied k

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.

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 by our Office of Inspection and Enforcement.

We do not intend to further pursue this issue in this Safety Evaluation Report.

As noted previously, several issues related to load combinations are still open.

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

The specified design and service loading combinations used for the design of ASME 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 dug to postulated events or system operating

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25 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 pro-vides a conservative basis for the design of support components to with-stand 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 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.

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26 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 1

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 life.

The only open issue concerning the applicant's design criteria is the statement in the FSAR that "Deformations are not a limiting factor in the analysis of the CRD's components since stresses are in the elastic region."

Further information must be provided to show that elastic deformations cannot effect the performance of the CRD system.

Also, as noted in Section 3.9.3 of this report, there are several open issues related to load combinations.

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

The design criteria 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.

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

We have reviewed the physical configuration of the Susquehanna reactor internals and would note that they are essentially identical to the reactor internals at several operating plants which use the General Electric BWR/4 nuclear steam supply system.

As we stated in Section 3.9.2 of this Safety Evaluation Report, the proto-type of the Susquehanna reactor internals were tested for steady state and transient vibration primarily at Brown's Ferry. Unit l.

These tests and analyses plus the successful operating experience at Brown's Ferry and other similar plants indicate that the Susquehanna reactor internals will remain structurally sound during their design life of 40 years of normal operation.

Our review is also concerned with the structural integrity of the Susquehanna reactor internals under the combination of loads that would be experienced t

during postulated events such as the safe shutdown earthquake and loss-of-coolant accident.

As we noted in Section 3.9.2 of this Safety Evaluation Report, the applicant 1s currently reanalyzing the reactor internals for the combined loads due to a simultaneous safe shutdown earthquake and loss-of-coolant accident, including any asymmetric effects.

28 Also, as noted in Section 3.9.3 of this Safety Evaluation Report, there are several open issues related to load combinations.

Therefore, subject to resolution of these

issues, our findings are as follows.

The specified transients, design and service loadings, and combination of loadings as applied to the design of the Susquehanna reactor internals provide reasonable assurance that in the event of an earthquake or of a system transient 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 of structural integrity or impairment of function.

The design procedures and criteria used by the applicant in the design of the Susquehanna reactor internals comply with Standard Review Plan Section 3.9.5 and constitute an acceptable basis for satisfying the applicable requirements of General Design Criteria 1, 2, 4, and 10.

29 Inservice Testin of Pum s and Valves In Sections 3.9.2 and 3.9.3 of this Safety Evaluation Report we discussed the design and seismic qualification of safety-related pumps and valves in the Susquehanna

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 applicant 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 has not yet submitted its program for the preservice and inservice testing of pumps and valves, as requested by question 110.47, therefore we have not yet completed our review.

We will report the resolution of this issue in a supplement to this Safety Evaluation Report.

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30 Seismic uglification of Cate or I Instrumentation and Electrical

~Eui ment 1

IOur review under Standard Review Plan Section

3. 10 is concerned with the tests and analyses performed by the applicant to assure the operability of its safety-related instrumentation and electrical equipment in the event of an earthquake or other dynamic event at the Susquehanna site.

In instances where components have been qualified by testing or analysis to other than current standards (IEEE Std. 344-1975 and Regulatory Guides 1.92 and l. 100), the components will require reevaluation and possible requal-ification.

Our Seismic qualification Review Team is scheduled to review and inspect the nuclear steam supply system and balance-of-plant equipment of the Susquehanna plant.

This review will reevaluate the qualification testing and analysis already performed to determine that the effects of multi-axis seismic input and multi-mode equipment response have been properly accounted for.

Our review will also determine that the effects 1

of suppression pool hydrodynamic loads were properly considered.

On the basis of the review and site visit, the Seismic qualification Review Team will ascertain whether any nuclear steam supply system or balance-of-plant equipment have to be requalified.

Resolution of this issue will be pre-sented in a supplement to this Safety Evaluation Report.

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