ML20215M306

Evaluation of South Texas Project Unit 1 Tech Specs, Informal Technical Evaluation Rept

ML20215M306
Person / Time
Site:	South Texas
Issue date:	05/31/1987
From:	Baxter D, Valenti L EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20215M293	List: ML20215M290 ML20215M306
References
CON-FIN-A-6824 EGG-NTA-7702, NUDOCS 8706260346
Download: ML20215M306 (39)

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EVALUATION OF SOUTH TEXAS PROJECT UNIT 1 I

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TECHNICAL SPECIFICATIONS

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DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United j

States Govemrbsnt. Neither the United States Government nor any agency thereof, nor any of thr4r employees, makes any warranty, express or implied, or assumes any legal liabilif'y or responsibihty for the accuracy, c3mpleteness, or usefulness of any inform 3 hon, apparatus, product or process disclosod_ or represents that its use would not infnnge privately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, I

does not necessardy con 30tute or imMy its endorsement, recommendation, or favonng oy the United States Government or any agency thereof. The views and opinions of authors expressed herein 00 not necessarily state or reflect those of the United States Government or any agency thereof.

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EVALUATION OF SOUTH TEXAS PROJECT UNIT 1 TECHNICAL-SPECIFICATIONS i

D. E. BAXTER L. N. VALENTI 1

Published May 1987 Idaho National Engineering Laboratory-~

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Idaho falls, Idaho -83415 4

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j Prepared for the U.S, Nuclear ~ Regulatory Commission Washington, D.C.

20555-i Under DOE Contract No. DE-AC07-76ID01570

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ABSTRACT This document was prepared for'the Nuclear Regulatory Commission (NRC) to assist'them in determining whether the South Texas Project Unit'l

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Technicai. Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the final Safety Analysis Report (FSAR) as' amended, and the requirements of the Safety.

Evaluation Report (SER) as. supplemented.

A comparative audit of the FSAR-as amended..and the SER as supplemented was performed with the South Texas Unit 1 T/S.

Several discrepancies were_ identified and subsequently res'olved by the NRC cognizant reviewer.

Resolutions to.the discrepancies noted in this report were achieved through_ conversations with the NRC Reviewer.

Resolutions were received'for all discrepancies that required resolution.

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1 FORfWORD o

This report is supplied as part of the Power Reactor Technical Specifications Evaluations being conducted for the U.S. Nuclear Regulatory Comission, Of fice of Nuclear Reactor Regulation, Division of Licensing by EG&G Idaho, Inc., NRR and I&E Support Unit.

The U.S. Nuclear Regulatory Comission funded the work under the authorization B&R 20 19 40 41 1, FIN No. A6824 - Power Reactor Technical Specif.ication Evaluation.

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CONTENTS ABSTRACT..............................................................

ii FOREWORD..............................................................

iii 1.

INTR 000CTION.

1 2.

REVIEW CRITERIA...........................

1

l 3.

SUMMARY

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4.

SOUTH TEXAS PROJECT UNIT.1 TECHNICAL SPECIf? CATIONS,

.l FSAR, SER CONSISTENCY COMPARISON.................................

3.

Section I.

Safety. Limits......................................

3-Section II.

Reactor Protection System Setpoints................

13 Section III.

Engineered Safety features Actuation' System Setooints..........................................-

3 Section IV.

Pressure Boundary Isolation Valves.................

4

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Section V.

Containment Isolation Valves......................

4 Section VI.

Containment Depressurization and Cooling' System Limiting Conditions for 0peration (LCO)............

5 Section VII.

Combustible Gas Control System Limiting Conditions for Operation............................

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Section VIII. Technical Specification Requirements Documented q

in the Safety Evaluation Report..................,...

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EVALUATION OF SOUTH TEXAS PROJECT UNIT NO. 1 TECHNICAL SPECIFICATIONS 1.

INTRODUCTION The South Texas Project Unit 1 is a Westinghouse Pressurized Water Reactor (PWR) plant.

It has been selected for an audit to determine if the South Texas Project Technical Specifications (T/S) are consistent with the South Texas Project Final Safety Analysis' Report (FSAR) up to and including.

Amendment 57 and the'. South Texas Project Safety Evaluation Report (SER) up to and including Supplement 2.

The specific sections of the T/S which were audited are listed in Part 2.

Differences between these sections of the T/S and.the FSAR and SER along.with the resolutions are, identified in Part 4 of this report.

2.

REVIEW CRITERIA The following T/S sections were reviewed for this evaluation:

i 1.

Safety Limits 2.

Reactor Protection System (RPS) Setpoints 3.

Engineered Safety Features Actuation System (ESFAS) Setpoints 4.

Pressure Boundary Isolation Valves (PIVs) 5.

Containment Isolation Valves (CIVs) 1 6.

Containment Depressurization and Cooling System Limiting Conditions for Operation (LCO) 7.

Combustible Gas Control System LCOs 8.

Technical Specification Requirements Contained in the Safety Evaluation Report (SER)

The sections of the T/S listed in Part 4 were compared to the FSAR and SER to determine if the T/S are CONSISTENT, CONSERVATIVE'or DIFFERENT than the FEAR and SER.

Setpoints and lists of valves and instruments-in the T/S were checked against tables in the FSAR and SER.

The SER was reviewed to ensure that T/S requirements-in the SER were addressed in the T/S.

-A description of each difference between the T/S and the FSAR and SER is included in this report.

1

3.

SUMMARY

During the performance of this audit, several differences between the

.T/S, SER and FSAR were noted.

Subsequently, discussions were held with the' cognizant'NRC reviewer.

A resolution was provided for each item that required one. The items are listed below and have been assigned a status code which indicates the status of the item.

These items are discussed'in detail in Part 4 of this-report.

All other sections were evaluated and found'to.be consistent or conservative.

Item Title Page Status

• Section VIII Item 5 Crud Deposition 8

6 Section VIII Item 6 Inadequate Core Cooling Detection' 9

5-10 6

Section VIII Item 8 Overpressure Protection 16 5

Section VIII Item 17 Technical Specification ItemsSection VIII Item 18 Leao, Lag, and Rate Time Constants 17 5

Se'ction VIII Item 19

' Design Modification 17 3

Section VIII Item 21 AC Power System 18 5

Section VIII Item 28 EDG Fuel Oil Transfer System 23

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Section VIII Item 32 Nuclear Safety Review Board 25 5

Section VIII Item 33 Accident Analysis.

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- 5 Section VIII Item 34 Uncontrolled RCCA Withdrawal 28 6-Section VIII Item 35 Inadvertent Boron Dilution 28 3

Section VIII Item 36 Technical Specification and-29 3

Emergency Operating RequirementsSection VIII Item 37 Overpressure Protection 30

- 3 Section VIII Item 38 ECCS System Design 30 6

Section VIII Item 39 Generic Letter 83-28 Actions 31 5

Status Code 1.

Unresolved, awaiting NRC/ Utility action 2.

Resolved pending issuance of T/S revision 3.

Resolved pending issuance of SER Supplement 4.

Resolved pending issuance of FSAR Amendment 5.

Resolved, NRC accepts as-is 6.

Resolved, item clarified and accepted 2

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SOUTH TEXAS PROJECT UNIT 1 TECHNICAL SPECIFICATION, FSAR, SER CONSISTENCY COMPARISON Section I.

Safety Limits This section covers the review of the safety limits as defined in Section 2.1 of the Technical Specifications.

It includes reactor core limits and RCS pressure.

FSAR SER 1

Technical Specification Section Section Evaluation 2.1.1 Reactor Core Figure 15.0-1 4.4 CONSISTENT Limits 2.1.2 RCS Pressure 5.2.2 5.2.2 CONSISTENT 3/4.2.5 DNB Parameters 15.0.11 4.4.3.1 CONSISTENT Section II.

Reactor Protection System Setpoints This section covers the review of the Reactor Protection System Setpoints to ensure the T/S values agree with or are conservative to the values assumed in the safety analysis or defined in the SER.

FSAR SER Technical Specification Section Section Evaluation 1

2.2 Reactor Trip System 7.2 & 15 7.2 CONSISTfNT l

Instrumentation j

Setpoints j

Section III.

Engineered Safety Features Actuation System (ESFAS) Setpoints l

1 This section covers the review of the ESFAS Setpoints to ensure the T/S values agree with or are conservative to the values identified in the f

FSAR sections or as defined in the SER as required values.

FSAR SER Technical Specification Section Section Evaluation 3/4.3.2 6.2, 15.0 7.3 CONSISTENT Table 3.3-4 I

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i Section'IV.

Pressure Boundary Isolation Valves (PIVs)

This review determines if all the PIVs identified in the FSAR and SER are included in the T/S.

FSAR SER Technical. Specification Section Section Evaluation 3/4.4.6 Table 3.4-1 5.2 5.2.5 CONSISTENT Pg. 3/4 4-22 No PIV's were. identified.in the FSAR or SER.

Section V.

Containment Isolation Valves (CIVs)

This review determines if all the CIVs identified in the FSAR and SER are included in the T/S.

FSAR SER Technical Specification Section Section Evaluation 3/4.6 6.2.4

6.2.4 CONSISTENT

Table 6.2.4-1 The complete listing of CIVs has been removed from the T/S and are maintained in the FSAR only.

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'Section VI.

Containment Depressurization and Coolina System (CDCS)

Limitina Conditions for Operation (LCOs)

This section reviews the LCOs for the C0CS to ensure they adequately cover the operation of the CDCT during all required modes of plant j

operation.

FSAR SER j

Technical Specification Section Section Evaluation j

3/4.6.2 6.2.2 6.2.2 CONSISTENT j

LCO'3.6.2.1

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S/R 4.6.2.1 LC0 3.6.2.2 i

S/R 4.6.2.2 LCO 3.6.2.3 S/R 4.6.2.3 The LCOs and Surveillance Requirements (S/R) for these systems are effective during Modes 1, 2, 3, and'4 and require all these systems be i

operable.

4 Section VII.

Combustible Gas Control System (CGCS)

I Limitina Conditions for Operation (LCOs)

This section reviews the LCOs for the CGCS to enture they adequately cover the operation of the CGCS during all required modes of plant operation.

j FSAR SER Technical Specification Section Section Evaluation 3/4.6.4 6.2.5 6.2.5 CONSISTENT LCO 3.6.4.1 S/R 4.6.4.1 LCO 3.6.4.2 S/R 4.6.4.2 The LCOs and Surveillance Requirements (S/R) for.these systems are effective during Modes 1 and 2 and require all these systems be operable.

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Section VIII.

Technical Specification Requirements Documented in the Safety Evaluation Report This section covers the review of all.the items identified in the

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safety evaluation report (SER) and supplements to the safety evaluation report (SSER) as T/S required items and whether they have or have not been adequately addressed in the T/S.

1.

SER Section:

2.4.14 Technical Specifications.and Emergency Operating Requiremer-Pg. 2-28 states:

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As stated in Section 2.4.11.2, plant operation will be permitted only i

when the water level in the ECP is at or above elevation 25.5 feet ms1.

Therefore, a Technical Specification.will be required to define the actions to be taken if the ECP water-level drops below elevation 25.5 feet ms1.

This Technical Specification should'also

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specify the maximum temperature that will be allowed in the ECP during normal operation and-discuss the actions to be taken if this temperature is. exceeded.

j T/S Section:

3/4.7.5 Pg. 3/4 7-13 l

1 T/S 3.7.5 specifies the required water level and intake temperature and specifies the action to be taken if they are not met.

This item is CONSISTENT 2.

SER Section:

3.7.4, Seismic Instrumentation Pg. 3-19' states:

The applicant has met the intent of SRP Section 3.7.4 except that a seismic instrumentation surveillance scheme has not yet been j

provided.

However, tre accordance with stated staff requirements, such a scheme will be incorporated in the Technical Specifications.

The applicant has met 10 CFR 100, Appendix A, by providing the instrumentation that is capable of measuring the effects of an earthquake, which meets GDC 2.

The applicant will meet 10 CFR 50.55a by providing an inservice inspection program that will verify operability by performing channel checks, calibrations, and functional tests at acceptable intervals.

In addition, the installation of the specific seismic instrumentation on the reactor containment structure and at other seismic Category I structures, systems, and components constitutes an acceptable program to record data on seismic ground motion as well as data on the frequency and amplitude relationship of the seismic response of major structures and systems.

A prompt' readout of pertinent data at the control room can be expected to. yield sufficient information to guide the operator on a timely basis for the l

purpose of evaluating the seismic response in the event of an earthquake.

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T/S Section:

3/4.3.3 Pg. 3/4 3-54 S/R 4.3.3.3.1 specifies the required surveillance scheme.

This item is CONSISTENT 3.

SER Section:

3.9.6, Inservice Testing of Pumps and Valves Pg. 3-38 states:

Pressure. isolation valves must be Category A or AC per IWV-2000 and meet IWV_-3420 of'Section-XI of the ASME Code, except as discussed below.

' Limiting conditions for operation (LCOs) must be added to the Technical Specifications to require corrective action (shutdown or-system isolation) when the finally approved leakage. limits are not met.

The Technical Specifications also must include surveillance-requirements that will state the acceptable leak rate testing frequency.

Periodic leak testing of each pressure isolation valve must be performed at least once each refueling outage, after valve maintenance-and before return to service.

Such testing also must be done_for-systems rated as 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 once a year.

Leak testing should also be performed after all-disturbances to the valves are complete, before power operation after.a refueling outage, maintenance, etc.

Leak rates must be equal to or less than 1/2 gpm for each inch of nominal valve size up to a maximum of 5.gpm for each valve.

The requirements of ASME Code Section XI, Paragraph IWV 3427(b), regarding increases in leakage rates are to be strictly observed.

This is required 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 limi_ ting value would be an indication of valve degradation from one test to another.

The Class 1 to Class 2 bound 6ry will-be considered the isolation point that must be protected by redundant isolation valves.

In cases where pressure isolation is provided by two valves,:both will be independently leak tested. When three or_more valves provide isolation, only two of the valves must be leak tested.

T/S Section:

3/4.4.6 Pg. 3/4 4-20 7

T/S 3.4.6.2 specifies the actions required if the approved leakage limits are not' met.

S/R 4.4.6.2.1 specifies the testing frequency.

This item is CONSISTENT 4.

SER Section:

4.3.2, Power Distribution Pg. 4-22 states:

South Texas.has used a design.FoH value of 1.52, slightly less

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than the more usual value of 1.55.

The usual nominal value is 1.44 and an 8% design uncertainty value is added.

South Texas has used a 6% value.

Because the expected uncertainty is about 4%, the use of 6%

is still acceptable. The value of FAH in operation is confirmed by the incore. instrumentation, and the 1.52 value will be required by Technical Specifications.

T/S Section:

3/4.2.3 Pg. 3/4 2-9 i

T/S 3.2.3 specifies that foH shall be less than 1.46 which is less than 1.52.

This item is CONSERVATIVE 5.

SER Section:

4.4.3.2, Crud Deposition and Flow Uncertainty Pg. 4-29 states:

Crud deposition in the core and the associated change in core pressure drop are to be detected by flow measurement as described in the FSAR.

Therefore, South Texas Technical Specifications should require that the reactor coolant system flow be monitored at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The thermal design flow for South Texas is-defined as 91.8% of.the best-estimate flow.

The procedure for verifying this value must be included in the plant Technical. Specifications, and tests of-the primary system before initial criticality must be'made to verify that a conservative primary system coolant flow rate _has.been'used in the design and analyses of the plant.

T/S Section:

3/4.4.1 Pg. 3/4 4-1 S/R 4.4.1.1 specifies that circulating reactor coolant be' verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

However, there was no T/S identified that specified the procedure used to verify what the' thermal design flow is.

This item is DIFFERENT RESOLUTION S/R 4.2.5.3 specifies a precision _ heat balance be run'at.least once per 18 months.

The Staff accepts this as the procedure for verifying the thermal design flow.

This item is CONSISTENT 8

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SER Section:

4.4.6.3, Staff Evaluation, Inadequate Core Cooling Pg. 4-33 states:

" Design and Qualification Criteria for Accident Monitoring j

l Instrumentation." The staff finds that the implementation schedule is i

l acceptable and the ICC detection system is in conformance with these j

i requirements, subject to the staff's ensuring that Technical l

Specifications relating to the final ICC system are submitted and j

l approved before fuel load, l

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T/S Section:

3/4.3.3 Pg. 3/4 3-64 l

T/S 3.3.3.6 and Table 3.3-10 specify the Accident Monitoring Instrumentation.

Staff approval of the ICC portion is needed.

This item is NOT EVALUATED RESOLUTION The Staff has reviewed and approves the ICC Detection system This item is CONSISTENT 7.

SER Section:

5.2.2.2, Overpressure Protection During Low Temperature Operation Pg. 5-5 states:

Low-temperature overpressure protection is provided by the cold overpressure mitigation system (COMS), which includes the two i

l pressurizer PORVs and the interlocks for RCS pressure control during low-temperature operation (SER Section 7.6).

The PORVs will have their opening setpoints automatically adjusted as a function of RCS temperature. Two independent protection sets (i.e., interlocks) are provided.

The system will be manually armed by the operator below 350*F.

The RCS wide-range temperature measurements.(hot-leg temperature for one set, cold-leg temperature for the other) will be auctioneered to obtain the lowest value.

This temperature signal will then be sent to a function generator (in the same protection set)

I which has a PORV setpoint curve program.

This function generator will produce a calculated maximum allowable (reference) pressure for the j

prevailing temperature.

The reference pressure will then be compared l

with the indicated RCS pressure from a wide-range pressure channel.

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If the measured reactor coolant pressure approaches the maximum allowable pressure within a certain limit, an alarm will be sounded on the main control board indicating a pressurization transient.

If the reactor coolant pressure continues to increase, the PORVs will be automatically opened to mitigate the pressure transient.

Thus, the system pressure will always be below the maximum allowable pressure.

This PORV setpoint curve shall be periodically updated, as shall be specified in the Bases for the Technical Specifications, to ensure that the stress intensity factors for the reactor vessel at any time in life are lower than the reference stress intensity factors as specified in 10 CFR 50, Appendix G.

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T/S Section:

B3/4.9 Pg. 83/4 4-?5 The bases specify that the PORV setpoint will be updateo based on the examinations of the vessel material irradiation surveillance.

This item is CONSISTENT 8.

SER Section:

5.2.2.2, ')verpressure Protection During Low Temperature Operation Pg. 5-5 states:

A.

The design mass input transient in mode 4 assumes that, with failure of one PORV to open, a safety injection signal will start one high head safety injection (HHSI) pump.

The normal charging and letdown flow paths would be isolated by a containment isolation "A" signal, but a reactor coolant pump (RCP) seal flow rate of 100 gpm would be maintained (normal seal flow is 20 gpm).

The capacity of each PORV is sufficient to discharge the combined HHSI and RCP seal flow rate at RCS pressure below the present maximum allowable PORV setpoint pressure for 200*F (590 psig).

In mode 5, the mass input transient assumes the operation of one centrifugal charging pump (CPP) with letdown isolated.

The plant Technical Specifications will require that two HHSI pumps, one CCP, and the positive displacement charging pump be locked out before reaching COMs activation conditions, and that the remaining HHSI pump would be locked out at 200*F.

T/S Section:

3/4.5.3, 3/4.1.2 Pg. 3/4 5-8, 3/4 1-11 T/S 3.5.3.2 requires all HHSI pumps be inoperable for mode 5 and 6.

S/R 4.1.2.3.2 specifies that the CCP shall be inoperable during modes 4.5 and 6.

This item is CONSISTENT B.

The heat input analysis was performed for an inadvertent RCP start Msuming that the RCS was water solid at the initiation of the et at and W;t a 50*F mismatch existed between the RCS and seconJ6ry side Of the steam generators.

The heat input analysis took into account the single-failure criteria.

The applicant states that the allowable limits will not be exceeded.

The staff 5

requires Technical Specifications on the maximum permissible primary-secondary temperature mismatch before starting an RCP if the pressurizer is water solid.

The applicant must also provide j

PORV setpoint values in the Technical Specifications.

T/S Section:

3/4.4.9 Pg. 3/4 4-36 l

T/S 3.4.9.6 and Figure 3.4-4 specify the required PORV setpoints, however, no T/S was identified that specified the maximum permissible i

primary to secondary temperature mismatch before starting an RCP with l

l the pressurizer solid.

l This item is DIFFERENT 10 l

RESOLUTION T/S 3.4.1.3 a thru d end T/S 3.4.1.4.1 all have a footnote that specifies the water temperature difference must be less than 50*F for

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Modes 4 and 5.

This item is CONSISTENT c

9.

SER Section:

5.2.4.3, Evaluation of Compliance with 10 CFR 50.55a(g)

Pg. 5-10 states:

In accordance with 10 CFR 50.55a(g), examination requirements of Subsection IWE of Section XI will not be included in the PSI program.

Eddy current PSI examinations of steam generator tubing will be j

conducted at the site in accordance with the South Texas Technical j

Specifications and Section XI.

T/S Section: ~3/4.4.5 Pg. 3/4 4-12 S/R 4.4.5.2 and 4.4.5.3 specify the required inspections and frequencies.

This item is CONSISTENT 10.

SER Section:

5.2.5, Reactor Coolant Pressure Boundary Leakage Detection Pg. 5-13 states:

For monitoring of intersystem leakage, radiation monitors are used to detect reactor coolant leakage into the component cooling water system, which supplies the r.esidual heat removal heat exchangers, letdown heat exchangers, seal water, and thermal barrior heat exchangers.

Leakage through steam generator tubes is detected by condensate vacuum pump radiation monitors in the condenser offgas line and by using the sampling system.

Accumulator leakage is detected by level and pressure indications and alarms provided for each accumulator.

Thus, RG 1.45, Position C.4, regarding intersystem leakage is satisfied.

The applicant has provided indication and alarm for the leak detection system in the control room as well as providing for testing and calibration during plant operation.

Thus, RG 1.45, Positions C.7 and C.8, regarding instruments and alarms and provisions for testing and calibration is satisfied.

The applicant has stated that the plant Technical Specifications will provide limiting conditions for identified and unidentified leakage, thus satisfying RG 1.45, Position C.9.

T/S Section:

3/4.4.6 Pg. 3/4 4-20 T/S 3.4.6.2 specifies limiting conditions for identified and

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unidentified leakage.

This item is CONSISTENT 11

11.

SER Section:

5.4.2.2.2, Evaluation of the Inspection Program Pg. 5-23 states:

The applicant has committed to perform a preservice examination of the steam generator tubing at the site in accordance with RG 1.83, Revision 1, and the requirements of the Standard Technical Specifications for Westinghouse PWRs and Section XI of the ASHE Code.

I NUREG-0452, Revision 4, requires a preservice examination of_the full length of each tube in each steam generator using eddy-current techniques to establish the baseline condition of the tubing.

The examination is to be performed before initial power operation using the equipment and techniques expected to be used during subsequent inservice examinations.

The staff considers the PSI examination of steam generator tubing at Unit 1 a confirmatory issue, contingent on the applicant conforming i

i with the requirements of the applicable Standard Technical Specifications.

T/S Section:

3/4.4.5 Pg. 3/4 4-15 S/R 4.4.5.4.a.9 specifies the preservice inspection as described in the Standard Technical Specifications.

This item is CONSISTENT 12.

SER Section:

5.4.7.2, Cold Shutdown Capability Pg. 5-26 states:

The safety-grade auxiliary feedwater storage tank (AFST) has a net usable volume of 525,000 gallons.

The present design, as described in the FSAR, does not provide safety-grade backup to the AFST.

The applicant has identified the most limiting failure as the loss of A train ac power, which results in the loss of two atmospheric dump valves and auxiliary feedwater trains.

In response to a staff inquiry, the applicant indicated that, with the most limiting failure, RHR cut-in conditions can be achieved 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> after reactor trip, j

based on maintaining hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> natural circulation coolda n and an 8-hour soak period.

Approximately j

445,000 gallons of water would be added to the steam generators during this period.

Allowance is made for unusable volumes from instrument error, vortexing, and miscellaneous losses.

The staff finds this acceptable.

The plant Technical Specification for minimum AFST volume will require 518,000 gallons BTP RSB 5-1 also specifies the need for surveillance of boron concentration.

FSAR Table 5.4.A-1 states that boron sampling is not required during cooldown to cold shutdown.

The staff's position is j

that periodic boron measurements are necessary, particularly if the plant is in natural circulation.

In response to a staff inquiry, the applicant has indicated that boron concentrations can be measured periodically by use of the postaccident sampling system.

The staff will require that performance of periodic boron measurements during cooldown be specified in the plant Technical Specifications.

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T/S 3.7.1.3 specifies the required'AFST minimum volume.

This item is CONSISTENT:

1 T/S Section:

3/4.1.1 Pg. 3/4 1-1

-S/R 4.1.1.1.1.d afd 4.1.1.2.b.1 specify boron concentration verification once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This item is CONSISTENT 1

13.

SER Section:. 5.4.12, Reactor Coolant System High Point Vents Pg. 5-33. states:

The FSAR states that a break in the RVHVS line would result in a-LOCA

.]

no greater than 1 inch in diameter.

The-applicant was asked to.

1 provide-additional'information to demonstrate that the isolation l

valves are capable'of closing'against the dynamic forces associated with a broken vent line or provide a-flow restriction that wouldL result in leakage flows lower than the LOCA definitions in the event

-i I

of a pipe break.

The applicant responded that the RVHVS isolation valves are.normally closed, fail-closed valves.. The valve is designed so that valve flow tends,to close the valve disk; thus, the greater the' flow rate, the' more rapidly the valve would tend to close~. Also, as.noted above, the applicant has provided redundancy so that the-system can be isolated..

'l with a single failure. The staff concludes.that the RVHVS design j

satisfies the requirements in NUREG-0737..The Technical Specifications for South Texas must include. operability requirements for the RVHVS.

T/S Section:

3/4.4.11

'Pg. 3/4'4-10 i

T/S 3.4.11 specifies operability requirements for the RVHVS' This item is CONSISTENT 14.

SER'Section:

6.1.1, Engineered Safety Features Materials Pg'. 6-1 states:

The applicant will use borated water with 'a concentr'ation of 2500-2600 ppm boron (as boric acid).from the refueling water storage tank during the. initial injection phase.of containment spray.

The borated water will be mixed with a 34% to 36% by weight' sodium-hydroxide solution from the chemical addition tank.

]

.The'resulting solution will have a.pH greater than 7.and will-drain.to the containment sump. Mixing is achieved as the solution-is-continuously retirculated from the sump ~to the containment spray nozzles during the' recirculation phase of containment. spray.

4 13

.l

l l

{

l l

The staff evaluated the pH of the water (mixture of refueling water l

storage tank and sodium hydroxide solution) in the containment sump.

i It verified by independent calculations that sufficient sodium hydroxide is available to raise the containment sump water pH above the minimum 7.0 as required by Branch Technical Position (BTP)

MTEB 6-1 to reduce the probability of stress corrosion cracking of austenitic stainless steel components.

The effectiveness of the chemical additive in removing fission products in containment is j

reviewed in Section 6.5.2 of this SER.

The staff will include surveillance requirements in the plant Technical Specifications to verify that sufficient sodium hydroxide is maintained in the containment spray additive tanks.

j SSER 1, Appendix L (SER ERRATA) revised the solution concentration to f

30-32%.

I i'

T/S Section:

3/4.6.2 Pg. 3/4 6-15 T/S 3.6.2.2.a specifies a 30 to 32% by weight NaOH solution and S/R 4.6.6.2.b.2 requires verifying NaOH concentration.

This item is CONSISTENT 15.

SER Section:

6.2.4, Containment Isolation System Pg. 6-12 states:

The 48-inch normal containment purge lines are sealed closed during operating conditions other than cold shutdown and refueling.

Valve position indication lights are provided to permit verifying that the valves are closed.

The supplementary containment purge system may e used during nornal plant operation (operating modes 1 through 4).

The plant Technical Specifications, which will be reviewed separately, will prescribe limitations on system usage.

Normal and supplementary purge system isolation valves are designed to close on receipt of a containment ventilation isolation signal.

This signal is initiated by the following:

safety injection signal, containment phase A isolation manual actuation, containment spray manual actuation, and high containment purge radiation.

The staff has reviewed information provided by the applicant to demonstrate compliance with NU3EG-0737 Item II.E.4.2, " Containment l

Isolation Dependability." As previously described, the applicant has complied with the provisions regarding diversity in parameters sensed for initiation of containment isolation, identification of essential and nonessential systems, automatic isolation of nonessential systems, l

and closure of containment purge and vent isolation valves on a high radiation signal.

In addition, the FSAR states that the reopening of the containment isolation valves requires deliberate operator action and can only be performed on a valve-by-valve basis.

The containment v

setpoint pressure that initiates containment isolation should be

~

reduced to the minimum value compatible with normal operating conditions.

The containment setpoint pressure and the justification l

l for it should be provided by the applicant; this information will be 1

14

reviewed by the staff in conjunction with the development of the plant Technical Specifications.

In the response to Q480.17N, the applicant stated that debris screens are provided on the containment supplemental purge system, are located on the containment side of the inboard isolation valves, and are attached to seismic Category I piping.

The debris screens are also designed to withstand the LOCA differential pressure.

Finally, the applicant has committed to keep the 48-inch containment purge valves closed during the operational conditions of power operation, startup, hot standby, and hot shutdown and verify that the valves are closed as specified in the Technical Specifications.

T/S Section:

3/4.6.1 Pg. 3/4 6-12 T/S 3.6.1.7 specifies the allowable system usage during Modes 1 through 4 and specifies maintaining the 48-inch valves closed during operating Modes 1 through 4.

This item is CONSISTENT 16.

SER Section:

6.3.6, ECCS Conclusions p. 6-25 states:

1 g

)

The ECCS includes the piping, valves, pumps, heat exchangers, l

instrumentation, and controls used to transfer heat from the core after a LOCA. The scope of review of the ECCS included piping and i

instrumentation diagrams, equipment layout, failure modes and effects I

analyses, and design specifications for essential components.

The review included the applicant's proposed design criteria and design bases for the ECCS and the manner in which the design conforms to 1

these criteria and bases. The staff review of the failure modes and effects analyses found these analyses acceptable.

The staff concludes that the design of the ECCS is acceptable and meets GDC 2, 5, 17, 27, 35, 36, and 37, except as noted.

(1) The applicant has met GDC 2 with regard to the seismic design of the ECCS and the structures nousing it and also of non-safety systems or portions thereof that could have an adverse effect on the ECCS by meeting Position C.2 of RG 1.29.

This is further discussed in SER Section 3.2.

(2) The applicant has met GDC 5 with respect to the sharing of structures, systems, and components by demonstrating that the ECCS and ancillary systems are not shared between Units 1 and 2.

(3) The applicant has met GDC 17 with respect to providing sufficient capacity and capability to ensure that the core is cooled and vital functions are maintained in the event of postulated accidents, with and without the availability of offsite power.

c 15

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(4) THe applicant has met GDC 27 with regard to providing conbined reactivity control system capability in conjunction with poison added by the ECCS to ensure that, under postulated accident conditions and with appropriate margin for stuck rods, the capaoility to cool the core is maintained.

(5) The applicant has met GDC 35 with regard to abundant core cooling capability for the ECCS by providing redundant safety-grade systems to transfer heat from the core at a rate so that fuel and cladding damage that could interfere with ef fective core cooling is prevented, and cladding metal-water reaction is limited to negligible amounts, providing the applicarit supplies the additional open and confirmatcry information asked for in item 8 below.

(6)

The applicant has met GDC 36 with respect to the design of the ECCS to permit appropriate periodic inspection of important components of the system.

(7) The applicant has met GDC 37 with respect to designing the ECCS to permit testing of the operability of the system throughout the life of the plant, including the full operational sequence that brings the system into operation.

The plant Technical Specifications will need to be reviewed to confirm compliance with the criteria.

T/S Section:

3/4.5.2; 3/4.8.1 Pg. 3/4 5-5, 3/4 8-6 S/R 4.5.2.and 4.8.1.1.2 specify the testing necessary to comply with the. requirements of GDC 37.

l This item is CONSISTENT l

17. SER Section:

7.1.4.3, Technical Specification Items Pg. 7-3 states:

(1) The staff requested detailed information on the methodology used to establish the Technical Specification trip setpoints and allowable values for the reactor protection system (including reactor trip and engineered safety feature channels), including i

the lead, lag, and rate time constant setpoints assumed to operate in the FSAR accident and transient analyses.

(2) The staff requested detailed information on response time testing as specified in the plant Technical Specifications, j

The detailed review of trip setpoints and response time testing will be performed as part of the staff's review of the plant Technical Specifications.

T/S Section:

Pg.

}

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16 l

1

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s og!f 4

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This item requires staff review and approval.

lfl This item is NOT EVALUATED RESOLUTION The Staff has reviewed the trip setpoints and response time testing and found them acceptable.

This item is CONSISTENT t

18.

SER Section:

7.2.2.1, Lead, Lag, and Rate Time Constant Setpoints Used in Safety System Charinels Pg. 7-8 states:

Several safety system channels make use of lead, lag, or rate signal compensation to. provide single time responses consistent with assumptions in the FSAR Chapter 15 analyses.

The time constants for l

these single compensations are adjustable setpoints within the' analog portion of the safety system.

The staff position is that the time constant setpoint be incorporated into the plant Technical Specifications.

This issue will be reviewed during the review of the plant Technical Specifications.

Additional information and formal documentation is required.

This is a Technical Specification item.

T/S Section:

2.2 Pg. 2-7 T/S Table 2.2-1 Notations specifies the constant values to be used.

The staff needs to review these values and determine if they are acceptable setpoints as required.

This item is NOT EVALUATED RESOLUTION 1

The Staff has reviewed the time constants and found them acceptable.

l This item is CONSISTENT 19.

SER Section:

7.2.2,2, Design Modification for Automatic Reactor Trip Using Shunt Coil Trip Attachment Pg. 7-8 states:

The Westinghouse Owners Group (WOG) has submitted a generic design modification to provide automatic reactor trip system (RTS) actuation i,

of the breaker shunt trip attachments in response to Salem anticipated transient without scram events.

The staff.has reviewed and accepted the generic design modification and has identified additional information required on a plant-specific basis.

By letter dated October 14, 1985, the applicant provided a response to the.

plant-specific questions identified by the staff in its SER on the generic Westinghouse design (Eisenhut, August 10, 1983).

The staff has reviewed the applicant's proposed design for the automatic 17 i

i J

actuation of the reactor trip breaker shunt trip attachments and finds it acceptable except for the breaker response time testing, which l

should be included in the Technical Specifications.

This is a Technical Specification item.

T/S Section:

3/4.3.1 Pg. 3/4 3-10 T/S Table 3.3-2 does not specify any required response time testing of these breakers as required.

This item is DIFFERENT RESOLUTION SER Supplement 4 will revise Section 7.2.2.2 to delete the requirement to response time test these breakers.

This item is CONSISTENT 20.

SER Section:

7.3.2.6, Testing of Engineered Safeguard P-4 Interlock Pg. 7-22 states:

In a letter dated November 7, 1979 (from T. M. Anderson), Westinghouse j

notified the Commission of an undetectable failure that could exist in the engineered safeguards P-4 interlocks.

In a letter dated July 15, 1985, the applicant addressed the design on P-4 interlocks.

The South Texas design is different from that of the' earlier Westinghouse plants in that the P-4 contacts for each train's trip and

~

bypass breakers are wired individually to the solid state protection system.

The status of the trip and bypass breakers is indicated on he control board.

Verification of P-4 contact status will be administratively controlled following any condition that requires opening of the reactor trip breakers and following reclosure of the trip breakers.

Operability of the P-4 contacts will also be checked as part of the Technical Specification surveillance testing program.

T/S Section:

3/4.3.2 Pg. 3/4 3-46 T/S Table 4.3-2 specifies an operational test of the P-4 contacts as required.

This item is CONSISTENT 21.

SER Section:

8.3.1, AC Power System Pg. 8-7 states:

Each of the three trains in each unit is provided with two levels of

{

undervoltage schemes at the 4160-V bus.

The first-level scheme called q

loss of voltage uses four undervoltage relays in a two-out-of-four logic and a time delay Of 0.5 to 5.0 seconds to isolate the safety-related buses from the offsite power system, disconnect selected loads, and start and load the associated diesel generator through its sequencer.

THe second-level scheme called degraded grid l

18 j

I

I voltage relays are set to alarm in the control room on a tolerable degraded bus voltage.

Should an SI signal be present coincident with degraded voltage, the sequencer enters the mode III state, which is a loss of offsite power coincident with an SI signal.

This causes a trip of the offsite power source breakers and also causes the starting and loading of the diesel generator with the accident loads.

However, if degraded voltage persists beyond the tima settings, or further degrades for nonaccident conditions, the degraded voltage relays or loss of voltage relays will cause the sequencer to enter mode II, which is a loss of offsite power.

In this mode, the offsite ac power breakers are tripped and the diesel generator is started and loaded with tha shutdown loads. The voltage and time delay setpoints for both loss of voltage and degraded voltage scheme relays will be selected when the applicant completes the analysis of the onsite distribution system voltages in compliance with BTP PSB-1 and includes them in the Technical Specifications.

The staff will provide its evaluation of the analysis and the setpoints in a supplement to this SER.

T/S Section:

3/4.3.2 Pg. 3/4 3-32 T/S Table 3.3-4 specifies the setpoint values for Loss of Voltage.

Tolerable Degraded Voltage and Sustained Degraded Voltage.

The staff needs to supply its evaluation of these values.

This item is NOT EVALUATED RESOLUTION The Staff has reviewed these values and found them acceptable.

This item is CONSISTENT 22.

SER Section:

8.3.1, AC Power System Pg. 8-7 states:

BTP SB-1 further requires that the Class IE bus load snsdding scheme i

automatically prevent load shedding during sequencing of the emergency loads to the bur.

The load shedding feature should, however, be reinstated upon completion of the load sequencing action.

The Technical Specifications must include a test requirement to demonstrate the operability of he automatic bypass and reinstatement features at least once per 18 months during shutdown.

Regarding this requirement, the applicant indicated that the South Texas design automatically prevents load shedding durina sequencing of the emergency loads on the diesel generator.

Upon completion of sequencing, the load shedding feature is reinstated by the manual resetting of the reset button in the main control room or at the sequencer panel.

The Technical Specifications will include testing of the automatically bypassed load shedding feature during load sequencing on the diesel generator and the manual operation for its reinstatement.

l 19

)

1

I 1

T/S Section:

3/4.8.1 Pg. 3/4 8-5

-l S/R 4.8.1.1.2 provides the requires system surveillance testing.

This item is CONSISTENT 23.

SER Section:

8.3.1, AC Power Systems Pg. 8-9 states:

Section C.2.a.3 of the guide recommends that the preoperational and i

18-month tests include the full-load capability test for an interval of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of which 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> should be at the load equivalent to the continuous rating of the diesel generator and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the load equivalent to the 2-hour rating of he diesel generator.

South Texas Technical Specifications call for 22-hour operation at the continuous rating as required by the guide but 2-hour operation at the 2000-hour rating instead of the 2-hour rating of the diesel generator.

The 2-hour rating test has been performed on each diesel generator.

The FSAR states that.the type qualification test performed on a South Texas diesel generator provided that it can operate at the 2-hour rating for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of any 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation with no reduction in annual maintenance interval.

The 2-hour rating test, if performed on an 18-month basis, will impose unnecessary stresses on the machine because the maximum load required for the design-basis accident shall not exceed the i

I 2000-hour rating of the diesel generator.

The applicant was requested to add in the FSAR that no transient effect or future addition of

~

loads on the emergency buses will exceed the 2000-hour rating of the diesel generator.

Subsequently, the applicant revised FSAR Section 8.3.1.2.10 to add the above statenent.

The staff concludes that the periodic testing of diesel generators (every 18 months) at the 2000-hour rating rather than 2-hour rating is acceptable.

However, should the future load conditions exceed the 2000-hour rating of the diesel generator, the diesel generator will have to be tested at the 2-hour rating thereafter.

This condition will be incorporated into the Technical Specifications.

T/S Section:

3/4.8.1 Pg. 3/4 8-6 The footnote to S/R 4.8.1.1.2.e.7 specifies the requirement to return to testing at the 2-hour rating if the 2000-hour rating is ever exceeded.

This item is CONSISTENT 24.

SER Section:

8.3.2, DC Power Systems Pg. 8-11 states:

Minimum battery voltage at the 2-hour-duty cycle is 106 V, and the minimum voltage required by the control equipment is 100-V..The i

battery float voltage is 130 V dc, and the equalizing charging of the

~

battery requires a maximum of 141 V dc (during shutdown when all

. equipment is disconnected).

During operation, the maximum equalizing charge voltage is 135 V dc.

All the 125-V dc Class lE equipment will-20 L

i i

be qualified for operation over a voltage range of 105 to 135 V.

In addition, all the 125-V dc Class lE loads will be tested to verify i

correct operation at the maximum and minimum system voltages.

The i

response also indicated that no non-Class 1E loads are supplied from the Class lE 125-V de system.

All equipment of the Class lE de power systems is located in a ventilated, centrolled environment outside the reactor containment building (RCB).

Cables or supporting structures penetrating into the 1

RC8 are designed to operate in the postaccident environment for the j

period required to maintain the plant in a safe shutdown condition i

following a design-basis accident.

Independence of the four battery systems is secured by separation of cables and equipment and by prohibiting cross-ties between load groups in different trains.

Periodic testing of the Class 1E dc power syst.em equipment is performed in accordance with RG 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants," to verify the de system's ability to perform its safety function.

The batteries and chargers are inspected and tested in accordance with the Technical Specifications.

T/S Section:

3/4.8.2 Pg. 3/4 8-10 S/R 4.8.2.1 specifies the testing required for the batteries and chargers.

This item is CONSISTENT

25. SER Section:

9.2.1, Service Water Systems Pg. 9-11 states:

The ECWS is designed to meet the single-failure criteria.

Power is supplied to each train from a separate emergency bus backed by a diesel generator so that the failure of one diesel generator only affects one ECW train. Any two of the three ECW cooling trains can supply the minimum cooling requirements during a design-basis accident, including a loss-of-coolant accident, and during nornal cold shutdown with or without offsite power.

Thus, GDC 44 is satisfied.

The ECWS design incorporates provisions for functional testing and inspection during normal plant operation according to the plant Technical Specifications.

The ECWS trains will be operated alternately.

Temperature, pressure, and differential pressure indications are provided to monitor the parameters of the ECWS.

T/S Section:

3/4.7.4 Pg. 3/4 7-12 S/R 4.7.4 specifies the required surveillance to ensure system operability as required.

This item is CONSISTENT f

21

26.

SER Section:

9.2.2, Component Cooling Water Systems Pg. 9-12 states:

In response to a staff concern (SRP Section 9.2.2) regarding luss of cooling water flow to the RCPs as a result of a single failure in the common supply line that might result in the occurrence of a multiple locked rotor condition, the applicant indicated that testing performed by Westinghouse has shown that the RCPs will incur no damage as a result of cooling flow interruption for 10 minutes.

This 10-minute test with no damage indicates that the pumps could potentially run longer with loss of cooling water without the need for operator action.

Safety-grade instrumentation and alarms have been provided that annunciate in the control room on the detection of low CCW flow to the.RCP motor and pump lube eel bearing coolers.

Therefore, sufficient time and adequate indication are available to ensure operator action to prevent ar unacceptable locked-rotor etent.

Seal water flow to the RCP seale will be maintained from the chemical and volume control system.

rurther, the CCW pumps are powered from redundant onsite sources, thus preventing loss of RCP motor and seal cooling in the event of a loss of offsite power with a concurrent single failure.

Therefore, the design of the CCWS meets i

Item II.K.3.25 of NUREG-0737 and satisfies GDC 44.

During normal plant operation, one train of the CCWS is in continuous operation.

The operating train may be varied to ensure equal operating times for the pumps. Availability of pumps not running will be ensured by periodic tests and inspections according to the plant Technical Specifications.

The system components are located in accessible areas to permit periodic inservice ir.spections as required.

T/S Section:

3/4.7.3 Pg. 3/4 7-11 S/R 4.7.3 specifies the required surveillance to ensure system operability as required.

This item is CONSISTENT 27.

SER Section:

9.4.1.3, Essential Chilled Water System Pg. 9-27 states:

(

l The essential chilled water system is designed to meet the

(

single-faiiure criterion with three 50% capacity trains fed by separate Class lE buses.

A single failure will result in loss of only one train and not prevent delivery of 100% of system requirements.

The essential chilled water system is designed to provide cooling water to various safety-related components under all operating conditions with or without offsite power.

Thus, it meets G0C 44.

The essential chilled water system incorporates provisions for functional testing and inspection during nornal plant operation.

Temperature, pressure, and differential pressure indications are provided to monitor parameters of the essential chilled water system.

The system will receive periodic testing according to plant Technical Specifications.

22 i

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T/S Section:

3/4.7.14 Pg. 3/4 7-44 S/R 4.7.14 specifies the required surveillance to ensure system

+

operability as required.

This item is CONSISTENT 28.

SER Section:

9.5.4.2, Emergency Diesel Engine fuel Oil Storage and Transfer System Pg. 9-58 states:

In addition, the diesel fuel oil storage system conforms with RG 1.137, " Fuel-011 Systems for Standby Diesel Generators,"

Positions C.2.a through C.2.g, with the following exceptions.

The fuel oil quality tests will conform to the McGuire Technical Specification for fuel oil quality, as modified by the applicant's letter of September 19, 1985.

T/S Section:

3/4.8.1 Pg. 3/4 8-4 S/R 4.8.1.1.2.c.1 specifies the fuel oil quality tests to be

{

performed.

The staff needs to t.rify that these are in conformance with the McGuire Technical Specifications.

)

This item is NOT EVALUATED j

i RESOLUTION

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The Staff has reviewed the fuel oil quality tests identified by South Texas and, although different than McGuires, found them acceptable.

This item is CONSISTENT 29.

SER Section:

10.2, Turbine Generator Pg. 10-3 states:

An inservice inspection program for the main steam throttle, governor, reheat stop, and interceptor valves is provided and includes (1) dismantling and inspection of one of each type of turbine steam valves at approximately 3-1/3-year intervals during refueling or maintenance shutdowns coinciding with the inservice inspection schedule and (2) 1 exercising and observing at least once a week the main steam stop and control, reheat stop, and interceptor valves.

This will be included in the plant Technical Specifications.

The applicant is also providing a monthly inservice inspection program for the extraction steam valves.

The inspection will check that the extraction check valve closing mechanism travels in the closing direction in a free and positive manner.

T/S Section:

3/4.3.4 Pg. 3/4 3-83 S/R 4.3.4.2 specifies all of the above stated requirements.

This item is CONSISTENT 23

30.

SER Section:

10.3.5, Secondary Water Chemistry Pg. 1016 states:

~

In late 1975, the staff incorporated provisions into the Standard

./

Technical Specifications that required limiting conditions for operation and surveillance requirements for secondary water chemistry parameters.

The Technical Specifications for all PWRs that were issued an operating license from 1974 until 1979 contain either these provisions or a requirement to establish these provisions af ter baseline chemistr y conditions have been determined.

The intent of the provisions was to provide added assurance that the operators of newly licensed plants would properly monitor and control secondary water chemistry to limit corrosion of steam generator components such as tubes and tube support plates.

In a number of instances, the plant Technical Specifications have significantly restricted the operational flexibility of the plant with little or no benefit with regard to limiting degradation of the steam generator tube and the tube support plates.

On the basis of this experience and the knowledge gained in recent years, the staff has concluded that Technical Specification limits are not the most effective way of ensuring that steam generatcr degradation will be minimized.

Because of the complexity of the corrosion phenomena and the state of the art as it exists today, the staff considers that it is more effective to specify a Technical Specification that requires the implementation of a secondary water chemistry monitoring and control program containing appropriate procedures and administrative controls.

This has been the approach usec for control of secondary water programs since 1979.

T/S Section:

6.8 Pg. 6.15 T/S 6.8.3.c requires the implementation of a secondary water chemistry program as required.

This item is CONSISTENT 31.

SER Section:

10.4.9, Auxilin y feedwater System Pg. 10-19 states:

The AFW system has been oesiryied to permit periodic testing, including full-flow pump testing.

In addition, the applicant will perform periodic monthly tests in ccnformance with the Standard Technical Specifications for Westinghouse Pressurized Water Reactors (NUREG.0452).

This meets GDC 46.

TheAFWsystemhasbdendesignedtopermitinserviceinspectionand periodic inspection of valves and pumps, thus meeting GDC 45.

e 24

The AFW system has diverse power sources that consist of offsite or onsite (Class lE) ac power for the motor-driven pumps and steam for the-turbine-driven pump.

There are no auxiliaries in the train for the turbine-driven pump that require ac power to maintain operation of the train.

This meets BTP ASB 10-1.

The AFW system is designed to supply water to the steam generators without throttling, thus avoiding throttling as a potential source of waterhammer. Waterhammer is also prevented b

• maintaining the lines full of water and preventing the formation of steam voids in the inlet piping by incorporation of (1) a separate nozzle for the introduction of AFW to the. steam generators, (2) minimizing the horizontal length of piping immediately upstream of the steam generators, (3) self-venting inlet piping within the steam generators, and (4) the outlet of the AFW nozzle below the normal steam generator water level.

See Section 10.4.7 of this SER for a further discussion of waterhammer prevention.

The staff has evaluated the AFW system against the generic short-.(GS) and generic long- (GL) term recommendations of NUREG-0611

" Generic Evaluation of feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants." The results of the staff's review of these recommendations as applicable to the South Texas design are discussed below.

GS-1--Technical Specification Time Limits:

The applicant has indicated that outage time limit and the subsequent action time will be as specified in the Westinghouse Standard Technical Specifications or equivalent.

This commitment is acceptable.

T/S Section: 3/4.7.1 Pg. 3/4 7-4 T/S 3.7.1.2 specifies outage times and corrective actions consistent with th* Standard Technical Specifications.

This item is CONSISTENT

32. SER Section:

13.4.2.1, Nuclear Safety Review Board Pg. 13-15 states:

NSRB members are to be qualified in accordance with ANS 3.1-1981 and RG 1.8.

The NSRB will undertake independent reviews and audits of potentially significant activities related to nuclear safety and will advise the Group Vice President--Nuclear of the results of their activities.

NSRB reviews will include station operations, modification of. station design, procurement of safety-related equipment and services, and unusual or. unanticipated events.

The applicant plans to have the NSRB in place about 6 months before fuel is loaded into Unit 1.

Details about the NSRB will be described in the plant Technical Specifications.

25

1 l

The staff has evaluated the applicant's general description of the NSRB and concludes that it is an acceptable appr oach to satisfying the requirement for independent review and audit'.

However, because of the lack of detail in the FSAR, the staff cannot reach a conclusion about the adequacy of the NSR8. When the Technical Specificatione are being prepared for Unit 1, the staff shall make sure that the app.lcant fully meets the staff requirements regarding independent review and audit.

T/S Section:

6.5.2 Pg. 6-9

.T/S 6.5.2.1 through 6.5.2.9 describe in detail the operation of the NSRB.

This item is CONSISTENT

/

The staff needs to assure itself that the NSRB as detailed in-I T/S 6.5.2 fully meets its requirements regarding independent review

and audit.

This item is NOT EVALUATED I~

RESOLUTION The Staff has reviewed the NSRB and found it to Le acceptable.

l This item is CONSISTENT O.

33.

SER Section:

15.0, Accident Analysis Pg. 15-3 states:

The applicant has stated that dose analyses performed for the Chapter 15 A00s and pas were performed assuming Standard Technical Specification steam geaerator tube leakage.

The applicant was requested to review the A00 and PA analyses'to provide assurance that all equipmentand systems relied on for A00 or PA mitigation whose availability and operability are assumed by the Technical Specifications in modes 1 and 2 can also be relied on to provide mitigation in other modes, If this assurance could not be-provided, the applicant was requested-to provide a detailed accounting 1

of what systems, equipment, and protective functions were assumed for.

these modes, a justification why the modes 1 and 2 analyses were bounding, and a confirmation'that the Technical Specifications applicable in modes 3,: 4, and 5 will'be consistent with and provide the same level of intended protection as'the Technical Specifications l,

a9plicable in modes 1 and 2.

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The applicant responded that each A00 and PA was revieweo with attention to modes that are not identified in the FSAR and tc' the protection operability requirements in these modes, including consideration of the applicable Technical Specifications, to determine i

whether the protection and operability required by the analysis were ensured by the Technical Specifications.

The applicant provided the results of the A00 and PA review with the exception of the steam generator tube rupture (SGTR) and boron dilution event (BDE), for i

which the analyses have not been completed. The analysis for a LOCA i

after shutdown is discussed in Section 6.3.

The occurrence of certain A00s is impossible or highly unlikely in modes other than 1 and 2 because of plant operating characteristics, j

for example, feedwater system malfunction resulting in increased core heat removal, loss of electrical load, and turbine trip.

Occurrence of pas such as the LOCA, steamline break (SLB), feedwater line break (FLB), and SGTR would be less likely because of reduced temperatures and pressure during modes 3, 4, and 5.

For certain A00s and pas the consequences of occurrence in modes 3, 4, and 5 would be less severe than in modes 1 and 2, for example, spurious main steam isolation valve closure, loss of ac power, loss of main feedwater, FLB, loss of flow, locked rotor, and startup of an inactive loop.

In the revised response to staff Q211.21, the applicant stated that, i

in the event of an SLB during shutdown, safety injection would be actuated and the steamlines isolated by the high-1 containment

~

pressure signal.

Steamline isolation would also be effected by the high negative steamline pressure rate signal.

SGTR during shutdown will be reviewed as part of the SGTR safety evaluations.

The applicant concluded that, in general, appropriate protection for i

all applicable A00s and pas is available and ensured by the Technical Specifications below modes 1 and 2 is consistent with the reduction in the severity and potential consequences of each A00 and PA in the other modes.

The staff will provide its evaluation after review of the South Texas Technical Specifications, subject to provision of the additional information requested for a small-break LOCA during shutdown, and the completed BDE analysis.

T/S Section:

Pg.

This item requires staff review and approval.

This item is NOT EVALUATED RESOLUTION The Staff has reviewed the appropriate Technical Specifications and found them acceptable.

W This item is CONSISTENT 21

34.

SER Section:

15.4.1, Uncontrolled Rod Cluster Control Assembly (Rod)

Bank Withdrawal from Zero Power Conditions Pg. 15-14 states:

The analysis also assumed that two reactor coolant pumps are in operation. The staff has been reviewing this aspect of the assumptions because Technical Specifications have not been requiring two pumps to be in operation in modes 3 and 4.

Several Technical Specifications have required that two pumps be in mode 3 or that the control rods not be operational.

The same problem for mode 4 is under generic review, but may require the same specifications.

The Technical Specifications will be examined later with this question in mind.

T/S Section:

3/4.1.3 Pg. 3/4 1-16 T/S 3.1.3.1 through 3.1.3.6 do not specify any conditions where two pumps must be in operation during modes 3 or 4.

This item is DIFFERENT RESOLUTION T/S 3.4.1.2 addresses Mode 3 operation.

Mode 4 operation with 2 pumps is a Generic Issue and is not addressed in plant specific Technical Specifications.

This item is CONSISTENT I

35. SER Section:

15.4.6, Inadvertent Boron Dilution Pg. 15-17 states:

]

The staff requested that the applicant provide additional information on this event, including a list of alarms and indications that would alert the operators to the occurrence of a boron dilution event (BDE),

including verification of their redundancy, and a description of any automatic mitigation systems; BDE analyses for modes 4, 5, and 6 that demonstrate that the minimum time available for operator action following receipt of an alarm before shutdown margin is lost will conform with SRP Section 15.4.6, assuming maximum charging and reacter makeup flow capability; demonstration that all possible dilution flow paths have locked-closed valves during mode 6; confirnation that Technical Specifications require operability of BDE alarms during all operational modes and surveillance of locked-closed valves.

The staff also requested a description of the analytical model used in the BDE calculations.

In response, the applicant provided a list of alarms and indications available to alert the operators to the occurrence of a BDE.

During I

mode 1, if the reactor is in automatic control, the rod insertion limit alarms would alert the operators to a dilution event, and, in manual control, power range neutron flux alarms would alert the operators.

During modes 1 and 2, reactor trip would occur on either overtemperature AT or power range high neutron flux (low or high 2B

l i

setting).

During modes 3, 4, 5, and 6, the Class 1E, redundant extended range neutron flux shutdown monitor alarms would alert the 1

operators to any reduction in shutdown margin.

A source range neutron j

flux high flux level alarm and a source range audib'le counter are also J

~

provided for modes 3, 4, 5, and 6.

Thus, both redundance and i

diversity are provided.

Automatic mitigation systems are not used.

The applicant has indicated that all valves that can result in boron dilution during mode 6 will be locked closed, and these will be specified in the South Texas Technical Specifications.

T/S Section:

3/4.1.2; 3/4.9.1 Pg. 3/4 1-9; 3/4 9-1 j

i S/R 4.9.1.3 specifies which valves must be closed and locked during mode 6 operation.

This item is CONSISTENT No T/S was identified that specified that the BOE alarms be operational during ALL modes of operation.

This item is DIFFERENT RESOLUTION With the issuance of SSER 3 this item will be clarified and acceptable as is.

This item is CONSISTENT 36.

SSER #1 Section:

2.4.14, Technical Specifications and Emergency Operating Requirements Pg. 2-2 states:

In Section 2.4.14 of the SER, the staff stated that the plant will be permitted to operate only when the water level in the ECP is at or above elevation 25.5 feet MSL.

The staff also stated that operation will be permitted only when the temperature in the ECP is less than a neximum value.

The staff, however, did not specify what this maximum temperature would be.

In performing its independent analysis of the thermal performance nf the ECP, the staf f assumed that the ECP water temperature on the intake side of the pond would be at 95'F.

Thus an ultimate heat sink technical specification should define the actions to be taken in the event that the ECP water level drops below elevation 25.5 feet MSL or the water temperature at.the intake side of the pond rises above 95'F.

T/S Section:

3/4.7.5 Pg. 3/4 7-13 T/S 3.7.5 specifies an intake temperature of 99*F.

e 29

This-item'is DIFFERENT RESOLUTION SSER 4 will revise the allowable intake temperature to 99*F.

i l

This item is CONSISTENT

37. SSER #1 Section:

5.2.2.1, Overpressure Protection During Power Operation Pg. 5-1 states:

In: response to the requirement of NUREG-0737 Action Item II.K.3.3, j

" Reporting SV and PORV Failures-and Challenges," the applicant stated

)

that the South Texas Project Technical Specifications will include a j

requirement to promptly report to NRC a' failure of a power-operated-

]

relief. valve (PORV) or a safety valve to close and will also include a-requirement.to document challenges to the PORVs or safety valves.in-s the monthly operating report.

The staff will review the South Texas Project Technical Specifications for compliance with this commitment.

T/S Section:

6.9 Pg. 6-20 T/S 6.9.1.5 requires monthly reporting o'f all challenges to the PORV's or Safeties, however, no T/S was identified that requires. prompt reporting to-the NRC any failures of a PCRV or.. Safety valve to close.

This item is DIFFERENT RESOLUTION The SER will be revised thru SSER 4 to only requir'e repor. ting of challenges.

This item is CONSISTENT 38.

SSER #1 Section:

6.3.1, ECCS System Design Pg. 6-1 states:

In response to the requirement of'NUREG-0737 Action Item II.K.3.17,

" Report on Outages of Emergency Core Cooling Systems-(ECCS),"Lthe applicant-in FSAR Appendix 7A committed to report ECCS outage data to-the NRC. The staff finds this acceptable and will review the South Texas Technical Specifications for compliance with this. commitment.

T/S Section:

Pg.

No T/S was identified that requires reporting ECCS ' outage data to the' NRC.

.j f

I l

l l

This item is DIFfERENT RESOLUTION

~

ECCS outage data reporting is covered by 10CFR50.73.

T/S 6.6.1 requires reports pursuant to 10CFR50.73.

This item is CONSISTENT

39. SSER #1 Section:

15.8.2, Generic Letter 83-28--Actions Pg. 15-23 states:

(10) Verify that each bypass breaker will be tested to demonstrate its operability before placing it into service for reactor trip breaker testing.

The applicant stated that the Technical Specifications were revised to include testing of.the bypass breaker before placing it into service for reactor trip breaker testing.

The staff finds this acceptable.

(11) Verify that the test procedure used to determine reactor trip breaker operability will also demonstrate proper operation of the associated control room indication / annunciation.

The applicant noted that the revised test procedures used to determine reactor trip breaker operability will demonstrate proper operation of the associated control room indication / annunciation.

The staff finds l

this acceptable.

{

(12) Verify that the response time of the automatic shunt trip feature will be tested periodically and shown to be less than or. equal to that assumed in the FSAR analyses or that specified in the-4 Technical Specifications.

The applicant stated that Westinghouse has prepared a report of the reactor trip breaker UVTA and shunt trip attachment (STA) life cycle test which concludes that periodic testing for STA can be limited to verifying that it can trip the breaker with.70 V dc (minimum design i

voltage).

Therefore, periodic testing of the automatic shunt trip feature response time is not required.

The staff finds this acceptable.

(13) Propose Technical Specification changes to require periodic testing of the undervoltage and shunt trip functions and the manual reactor trip switch contacts and wiring.

~

The applicant submitted the proposed Technical Specification changes to require pericalc testing of the undervoltage and shunt trip functions and the manual reactor trip switch _ contacts and wiring.

The staff finds this in accordance with GL 85-09 and, therefore, acceptable.

However, the applicant added ACTION 11 to Table 3.3-1, but failed to reference it in Item No. 19.

Therefore, the staff 31

f:

required that the applicant' add ACTION 11 to Item No. 19lof

, Table 3.3-1.

The applicant in a submittal dated May-30, 1986, informed the staff-that ACTION 11 is referenced in Item 19 of; Table 3.3-1.

T/S Section:

3/4.3.1 Pg. 3/4 3-4 T/S Table 3.3-1 Item 19 does not reference ACTION 11 as stated.

Thisittem is DIFFERENT RESOLUTION

-T/S Table 3.'3-1 has been renumbered and Item 19 is now Item 20'and Action 11 is now Action 12.

This has been clarified and accepted.

This item is CONSISTENT 4'

e 32-

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U S. DeuCLS AA 6 81U'. ATE 1Y COMhelgssose i m EPOAY NVM88 A fasespoes ty TsOC. deur Vair Wm. nf er,ys 82 Sei

$723' BIBUOGRAPHIC DATA SHEET EGG-NTA-7702 Sit #NSTRUCTIONS om f at alvtR$t 2 fif tt AN0 liuellTLE J 4t AV8 8k A4'E EVALUATION OF SOUTH TEXAS PROJECT UNIT 1 TECHNICAL SPECIFICATIONS e oaf t Rt*OM f COWpLtito asose T H vtAm May 1987

. Auf oais.

D. E. Baxter, L. N. Valenti

.oartauoariuuto g

vtAa oNTN May 1987 i

> nasoawimo oaGANizatioN NAus ANo manuNo Acoatsa <=,m e c.ms a PacatCTitasamons umir NvMetm NRR and I&E Support

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EG&G Idaho, Inc.

P. O. Box 1625 Idaho Falls, ID 83415 A6824 10 $PONSOmsNG ORG AN#2 AfiON NAME Amo MaiLineG ADOm458 dierneme te cesirs its TYPto#asPoni Division of Licensing Office of Nuclear Reactor Regulation

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U.S. Nuclear Regulatory Commission Final Technical Washington, DC 20555 Evaluation Report 13 SUPPLlutNT Anv NOTES tj AGSTR ACT (Jap wegs or ' eses Final technical evaluation report on the audit of the South Texas Project j

Unit 1 Technical Specifications performed for the NRC in connection with the issuance of Low Power and Full Power License for the applicant.

All identified discrepancies have been resolved.

to DOCUMENT AN AL es*5 e et t 's*Q8'DS '00 5c a,P roms in avaikAt utv STAftMENT Unlimited-

'S SECumsTv CLA$$1FICATION ir

..otNTi.it s,onN E~oso fi Unclassified ern rrs Unclassified 17 NuMetm o* *sGts is snict