ML20028A386

Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station,Unit 3.Docket No. 50-302. (Louisiana Power and Light Company)

ML20028A386
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Site:	Waterford
Issue date:	10/31/1982
From:	Office of Nuclear Reactor Regulation
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References
NUREG-0787, NUREG-0787-S04, NUREG-787, NUREG-787-S4, NUDOCS 8211190358
Download: ML20028A386 (31)

Text

NUREG-0787 Supplement No. 4 Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1982 g>nnrog kk....h.

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GPO Pnnted copy price: _$4.50

NUREG4787 Supplement No. 4

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Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power & Light Company U.S. Nuclear Regulatory t

Commission Office of Nuclear Reactor Regulation October 1982 p.s..w s

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TABLE OF CONTENTS Page 1.

INTRODUCTION AND GENERAL DISCUSSION..............................

1-1 1.1 Introduction................................................

1-1

1. 7 Summary of Outstanding Issues...............................

1-1

1. 8 Confirmatory Issues........................,................

1-2' 2.

SITE CHARACTERISTICS.............................................

2-1 2.2 Nearby Industrial, Transportation, and Military Activities..

2-1 2.4 Hydrologic Engineering......................................

2-1 2.4.2 Flood Potentia 1......................................

2-1 3.

DESIGN CRITERIA-STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS...

3-1 3.5 Missile Protection..........................................

3-1 4.

REACT0R..........................................................

4-1 4.4 Thermal-Hydraulic Design....................................

4-1 5.

REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS.....................

5-1 5.4 Component and Subsystem Design..............................

5-1 5.4.3 Shutdown Cooling System..............................

5-1 6.

ENGINEERED SAFETY FEATURES.......................................

6-1 6.3 Emergency Core Cooling System...............................

6-1 11.

RADI0 ACTIVE WASTE SYSTEM.........................................

11-1 11.3 Process and Effluent Radiological Monitors.................

11-1 15.

ACCIDENT ANALYSIS................................................

15-1 15.3 Limiting Accidents.........................................

15-1 15.3.3 Loss-of-Coolant Accident............................

15-1 22.

TMI-2 REQUIREMENTS...............................................

22-1 22.2 Discussion of Requirements.................................

22-1 Waterford SSER 4 111

ff TABLE OF CONTENTS (Continued)

Page APPENDICES A.

Continuation of Chronology of Radiological Review................

A-1 B.

Continuation of Bibliography.....................................

B-1 C.

NRC Unresolved Safety Issues.....................................

C-1 D.

List of Principal Contributors to SSER No.

4.....................

D-1 E.

Additional Errata to the Safety Evaluation Report................

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1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction On July 9, 1981, the Nuclear Regulatory Commission (NRC) issued a Safety Evalua-tion Report (SER) related to the operation of Waterford Steam Electric Station, Unit No. 3.

Subsequently, three supplements to the SER have been issued by the staff.

This fourth supplement s7 dates the SER by providing the staff's evalua-tion of information submitted by the applicant since the SER and its three sup-plements were issued.

Each of the following sections of this supplement is numbered the same as the section of the SER that is being updated and the discussions are supplementary to and not in lieu of the discussion in the SER.

Appendix A is a continuation of the chronology of the safety review.

Appendix B is an updated bibliography.

Appendix C addresses unresolved safety issues.

Appendix D is a list of princi-pal contributors to this SSER.

Appendix E contains additional errata. The Proj-ect Manager is Suzanne Black; she may be reached on (301) 492-7702.

1. 7 Summary of Outstanding Issues Section 1.7 of the SER and its supplements contained a list of outstanding issues.

This supplement addresses the resolution of a number of issues previously iden-tified as open.

These issues are listed below, along with the section of this report wherein their resolution is discussed.

(1) Turbine missiles (3.5.1.3, 3.5.3)

(2) Thermal-hydraulic design (4.4)

(3) Site hazards (explosions) (2.2)

At this time there remain a number of safety issues that have not yet been resolved. These will be addressed in a future supplement to the SER.

The fol-lowing is a list of these items.

(1) Fire protection (7.4, 7.5, 7.7, 9.5.2)

(2) PSI /ISI (3.9.6, 5.2.4, 6.6)

(3) Environmental qualification (3.11)

(4) Seismic qualification (3.10)

(5) Reactor coolant pump shaft break analysis (7.1, 7.2, 7.3, and 7.5)

(6) Therfral-hydraulic design (CPC's) (4.4)

(7) Indemnity requirements (21)

(8) Licensee qualification (13.1, 13.2)

(9) TMI issues Operating procedures (I.C tasks-long term)

Control room review (I.D.1)

Containment system design (II.E.4.2)

ICC instrumentation (II.F.2)

Waterford SSER 4 1-1

1.8 Confirmatory Issues Confirmatory issues are those which were essentially resolved to the staff's satisfaction but for which certain confirmatory information had not yet been provided by the applicant.

For the following issues, the staff has received that information and has confirmed the preliminary conclusion.

(1) Shutdown initiation using safety grade equipment (5.4.3)

(2) Containment sump vortex text (6.3.3)

At this time several issues remain for which the staff has not yet received the necessary confirmatory information. These issues, which are listed below, will be addressed in a future supplement to the SER.

(1) Testing the ultimate heat sink (2.4)

(2) Piping analyses (3.9.2)

(3) Containment isolation actuation signal (7.3)

(4) Performance of PWR relief and safety valves (22)

(5) Boron dilution events (5.4.3, 15.2.4.4)

(6) Emergency feedwater control (7.3)

(7) Feedwater line break analysis (15.3.2)

(8) Clarification of transient analyses with potential for fuel damage (15.3.1) 1.9 McenseConditions In addition to those issues listed in the SER as requiring a license condition to ensure that NRC requirements are met during plant operation, the staff has identified the following license condition:

(1) The low pressure turbine dircs must be inspected during the first refueling outage, and thereafter in accordance with the schedule established in the technical specifications.

Waterford SSER 4 1-2

2 SITE CHARACTERISTICS 2.2 Nearby Industrial, Transportation, and Military Activities In Supplement 2 to the Waterford 3 SER, the staff had reported its continuing review of two remaining offsite hazards; namely, the overpressure hazard due to potential explosions of LPG trucks on nearby Highway 18 and the potential for multiple pipeline failures near the exclusion boundary.

In response to our concerns, the applicant has submitted additional information regarding the above two hazards.

With respect to potential LPG truck explosions on Highway 18, the staff requested the Naval Surface Weapons Center to review the applicant's blast capacity evaluation of the Waterford 3 seismic Category I structures.

The findings of the review indicate that all elements of the seismic Category I structures exposed to the blast have sufficient capacity to withstand the overpressures due to an LPG truck explosion on Highway 18.

Hence, we conclude that this particular hazard does not pose a significant risk to the safe operation of Waterford 3.

The applicant has analyzed the question of multiple pipeline failures near the exclusion boundary.

Two failure mechanisms were identified:

thermal effects and mechanical forces.

The analyses indicate that although historical data on multiple line failures are very limited, it is theoretically possible to cause multiple line failures by thermal or mechanical means.

However, the conse-quences of multiple line breaks are limited to introducing additional sources of fire.

This implies the possibility of more extensive fires than in the case of a sirple line failure.

Since the pipelines are at or beyond the exclusion bounda'y, the potential fires, even when multiple failures are assumed, are suf 'iciently far away such that the thermal effects at the plant would be negligiale.

The staff thus concludes that potential multiple pipeline failures in the vicinity of Waterford 3 do not pose a significant hazard to the plant.

2.4 Hydrologic Engineering 2.4.2 Flood Potential 2.4.2.3 Local Intense Precipitation Waterford 3 has wet and dry cooling towers which are open at the top.

There are two open cooling tower areas A and B.

Local intense precipitation which falls directly over these open areas plus runoff from adjacent roofs will accumulate and pond on the floors of the dry cooling tower areas.

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of floor drains and a network of drainage pioing will convey this water to two sumps where a set of duplex pumps in each sump will remove water from the cooling tower areas.

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l Waterford SSER 4 2-1 L

Design Basis Rainfall Event In Section 2.4.2.3 of the SER, the staff concluded that the applicant's analysis of potential flooding in cooling tower areas A and B did not meet the design cri-teria suggested in Regulatory Guides 1.59 and 1.102 nor the requirements of GDC-2 i

because certain safety-related transformers and motor control centers located in the cooling tower areas could be flooded during a design basis rainfall event.

The staff' stated that safety-related components in coaling tower areas A and B should be flood protected to 4.2 ft and 3.6 ft, respectively.

The staff also stated that lower flood protection depths would be acceptable if additional or larger pumps were used to reduce ponding levels or if the applicant could provide assurances that roof drains would not become clogged.

The applicant subsequently presented an evaluation of the potential for blockage of roof drains.

This evaluation showed that clogging of roof drains would be highly unlikely.

However, as described below, the applicant conservatively assumed that 33 percent of the roof drains would be clogged during a design basis rainfall event.

Based on the information presented by the applicant, the staff agrees that it is highly unlikely that all of the roof drains would become clogged.

The staff further agrees that a 33 percent blockage of roof drains is a conservative assumption.

In Amendment 21 to the FSAR, the applicant presented a revised analysis of potential flooding in cooling tower areas A and B.

In this analysis roof drains were assumed to be 33 percent blocked.

The applicant also assumed that one of the sump pumps in each cooling tower area would be inoperable during a probable maximum precipitation (PMP) event.

Amendment 21 also stated that the sump pumps in the cooling tower areas each have a capacity of 325 gal per minute (gpm).

Initially, the FSAR had shown these pumps as having a capacity of 140 gpm.

This revised analysis by the applicant resulted in lower ponding levels in the cooling tower areas.

These levels, however, were not low enough to prevent flood-ing of the motor control centers which are located on the floor of the dry cool-ing towers.

To further reduce ponding levels in the cooling tower areas, the applicant proposed to allow water to flow into and pond in the Fuel Handling Building.

Openings will be provided between the cooling tower areas and the Fuel Handling Building.

The applicant has determined that three 4-in. diameter openings have to be installed in the sills beneath the exit door located on each side of the Fuel Handling Building.

However, to allow for some clogging of pipes, a total of eight 4-in. diameter pipes will be installed.

The appli-cant has estimated that by allowing water to pond in the Fuel Handling Building, a maximum of 1.6 ft of water will pond in the cooling tower areas and in the Fuel Handling Building.

The maximum height to which water can pond in the cool-ing tower areas before flooding of essential portions of the transformers occurs is 3.0 ft, and for the motor control centers it is 1.71 ft.

The staff has reviewed the material presented by the applicant and has performed an independent analysis.

The staff, therefore, concludes that, with the eight 4-in. diameter opening installed as indicated by the applicant, water depths in the cooling tower areas will remain below 1.6 ft following a PMP event and will thus not affect the safe operation of Waterford 3.

Combination of Events Regulatory Guide 1.59 suggests that a sufficient number of combinations of flood causing events be tested or discussed to assure that the highest flood level has Waterford SSER 4 2-2 j

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been determined.

An alternative combination which should be considered is an operating basis earthquake (OBE), which fails the sump pumps, coincident with a rainfall event less severe than the PMP.

This combination is considered appro-priate since the pumps are not seismically qualified 1, and thus cannot be shown to be operable following a seismic event.

The staff, therefore, requested that the applicant provide an analysis of the effects of a standard project storm (SPS)2 assuming all four sump pumps in the cooling tower areas are inoperable.

The applicant's analysis of this combination of events showed that there would be some flooding of r:otor control centers about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after ponding began in the cooling towers.

The applicant stated that reactivation of even one of the sump pumps would, at an increasingly rapid rate, reduce the total accumulated water level.

There was no discussion or description of how the sump pumps would be reactivated nor how long it would take to do so.

Thus the staff was unable to conclude that a rainfall event coincident with an OBE would not result in flood-ing of motor control centers and transformers in the cooling tower areas.

In analyzing the PMP, both the applicant and the staf f determined that ponded water in the cooling tower areas would peak at about the 5th or 6th hour.

After this, levels would decrease because the capacity of the sump pumps would exceed the amount of water coming in.

Thus, consideration of a 6-hour PMP as a design basis event was adequate, For the coincident SPS and OBE event, however, storm dura-tion is a much more critical parameter because the sump pumps are assumed to be inoperable, allowing water to accumulate for the entire duration of the storm.

In Section 2.4.2.3 of the SER, a 48-hour PMP is estimated to be 43.5 in. The 48-hour SPS rainfall would be about 21.8 in. assuming that the SPS is equal to 50 percent of the PMP.

Since ponding depth in the cooling tower areas is depend-ent on the duration of the rainfall event, the staff considered a SPS of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> duration.

This event would produce a total rainfall of about 23 in, and would result in a ponding depth of about 1.9 ft in the cooling tower areas assuming that all four sump pumps are inoperable.

Since this is higher than the maximum allowable ponding depth of 1.71 ft, the applicant has proposed to provide a portable pump with a pumping capacity of 100 gpm and sufficient head to pump over the cooling tower wall.

This pump will be stored on pallets placed away from any nonseismic Category I equipment which could fall and damage the pump.

In addition, the pump will be included in the surveillance testing program which will include a demonstration at least once per refueling that the, pump will cir-culate water.

As part of the station's emergency procedures, a provision will be included for emplacing the portable pump within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of a seismic event if the installed pumps fail.

The staff has determined that a 100 gpm pump capable of lifting water 75 ft vertically is adequate to prevent flooding of safety-related equipment in the I

tThe applicant has described the sump pumps as designed to seismic Category 1 requirements but not classified as seismic Category 1 (FSAR Table 3.2-1).

2Tne SPS is a storm used for design of flood control structures by the U.S. Army Corps of Engineers.

Rainfall resulting from a SPS is generally equal to about 40 to 60 percent of the PMP.

Waterford SSER 4 2-3

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cooling tower ares during a combined SPS-0BE event provided the pump is placed,,'

in operation within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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i Conclusion The staff'now concludes that with respect to potential flooding of the cooling tower areas, the station meets the requirements of GDC-2 and the criteria of Regulatory Guides 1.59 and 1.102.

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1 Waterford SSER 4 2-4

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3 DESNN CRITERIA STRUCTURE, COMPONENTS, EQUIPMENT AND SYSTEMS 3.5 Missile Protection f

s 3.5.1.3 Turbine Missiles According to General Desig'n Criterion 4 of Appendix A to 10 CFR Part 50, nuclear power plant structure, systems, ard components important to safety shall be appropriately protected against dynamic effects including'the effects of tur-bine missiles. Of those systems important to safety, this section is primarily concerned with safety-related systems; i.e., those structures, systems, and components necessary to perform iequired safety functions and to ensure:

1.

The integrity of the reactor coolant pressure boundary.

2.

'.The ca'pability to shut down the reactor and maintain it in a cold shutdown condition, or 3.

The capability to prevent accidents that could result in potential offsite exposures that are a significant fraction of the guideline exposures of 10 CFR Part 100, " Reactor Site Criteria."

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- TheWaterford3turbinegeneratorplacementandorientatioi$isunfavorablewith respect to the station reactor buildings.

This configuration places the reactor auxiliary building, control room, battery room, primary { water condensate storage tanks, main steam lines, and intake cooling water struc ure, as we.ll as the containment building, within,the low trajectory missile (LTM) strike zone (see Regulatory Guide 1.115).

Applicant's Analysis and Conclusion j

The applicant has performed an analysis to evaluate the probability of damage from postulated turbine missiles to safety-related systems.

Using the NRC recommended (Regulatory Guide 1.115) Pi values of approximately 10 4, and the Modified National Defense Research Council (NDRC) perforation formula recom-mended in the *cgulatory guide, the applicant has calculated the total damage probability due to low trajectory missiles (LTM) to be 3.4 x 10 8, and the same probability for high trajectory missiles (HTM) to be 6.0 x 10 8 per year.

The applicant calculated that the LTM design and destructive overspeed strike

- probabilities are 0.0 and 8.4 x 10 4 per turbine failure, respectively.

Based or/the historical missile producing turbine failure rate, 6 x 10 5 per year for design overspeed failure, and 4 x 10 s per year for destructive overspeed failure, their analysis yields a total probability of unacceptable damage to safety-related systems due to LTMs of 3.4 x 10 8 per year.

4 With regard to high trajectory missiles (HTMs), the applicant has performed an analysis and calculated that for safety-related systems the total HTM strike probabilities for postulated missiles from design and destructiv'e overspeed Waterford SSER 4 3-1 s

failures are 4.4 x 10 4 and 8.5 x 10 4 per turbine failure, respectively.

This yields a total HTM risk rate of 6.0 x 10 8 per year.

Staff Evaluation The staff has reviewed applicant's LTM risk assessment and performed an indepen-dent analysis based on the applicant's postulated missile spectrum which con-sidered the geometric relationship of the turbine to safety-related systems and which, assuming straight line trajectories, estimated the solid angle subtended by these systems to arrive at a probability of a LTM perforating barriers, and striking and damaging the systems.

The staff used the NORC missile perforation formula for concrete barriers.

Our analysis concentrated on the reactor coolant pressure boundary, and the interior of the battery and control rooms.

The intake structure was not explicitly included in the analysis since it lies in the shadow of the containment building; i.e., LTMs would have to pass through the containment building to strike the intake structure.

The main steam lines on the turbine side of the containment building wall were not considered safety-related targets since they would be effectively ruptured when the missile leaves the turbine casing and regardless of where such a rupture occurs, the MSIV is designed to close fast enough to protect the steam generators.

The staff assumed that postulated missiles leaving the turbine casing have uniform velocity distributions from zero to the maximum values stated in the FSAR.

The containment building walls were approximated as plane surfaces paral-1el to the turbine axis with the outer wall being reinforced concrete 3 ft thick and the inner wall being steel 1.9 in. thick.

The west walls of the battery and control rooms are reinforced concrete 2 ft thick.

Though the wall of the turbine building was considered a negligible barrier, credit was taken for the effect of the moisture separators on the turbine floor.

By our analysis, reasonable estimates of design and destructive overspeed strike probabilities are 1 x 10 8 and 5 x 10 3 per turbine failure, respectively.

Assumingdesignanddestructiveoverspeedturbinefailurerates6x10s and 4 x 10- per year, respectively, we obtained a total probability of unacceptable damage to systems important to safety due to LTMs of 3 x 10 7 per year.

The sum of this value and the one for HTMs of 1.6 x 10 7 per year, is within the 10 8 to 10 7 range and is acceptable to the staff.

The staff has also reviewed other factors that have a bearing on the probability of missiles generation.

With regard to destructive overspeed, the staff has reviewed FSAR Section 10.2 " Turbine-Generators" (see SER Section 10.2.1)

Overspeed protection is accomplished by three independent systems; i.e., normal speed governor, mechanical overspeed, and electrical backup overspeed control systems.

The normal speed governor modulates the turbine control valves to maintain desired speed load characteristics and it will close the intercept valves and control valves at 103 percent of rated speed.

The mechanical over-speed sensor trips the turbine stop, control, and combined intermediate valves by deenergizing the hydraulic fluid systems when 111 percent of rated speed is reached.

The turbine steam stop valves, control valves, reheat stop valves, and intercept valves are designed to fail closed on loss of hydraulic system pressure.

The electrical backup overspeed sensor will trip these same valves when 111.5 percent of rated speed is reached by independently deenergizing the hydraulic fluid system.

Both of these actions independently trip the energizing Waterford SSER 4 3-2

trip fluid system. The overspeed trip systems can be tested while the unit is on-line.

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

Similarly, for the design overspeed failure case, the staff has reviewed FSAR Sec-tion 10.2.3 " Turbine-Disc-Integrity" (see SER Section 10.2.2). The turbine was manufactured by Westinghouse.

The turbine discs and rotors are forged from vacuum degassed steel by processes that minimize flaws and provide adequate fracture toughness. These materials have the lowest fracture appearance trans-ition temperatures and highest Charpy V-notch energies obtainable on a consist-ent basis.

The preservice inspection program calls for 100 percent ulrasonic test (UT) of each rotor and disc forging before finish machining and magnetic particle test (MT) after finish machining. No MT flaw indications are permis-sible in bores, holes, keyways and other highly stressed regions.

Since 1979 the staff has known of the stress corrosion cracking problems in low pressure rotor discs in Westinghouse turbines. Appropriately conservative inspection intervals have been effective in monitoring crack growth to permit repair or replacement of discs well in advance of failure.

The applicant has submitted to the staff the materials properties of the low pressure turbine discs as well as the calculations of critical crack sizes. The method used to predict crack growth rates is based on evaluating all of the cracks found to date in Westing-house turbines, past history of similar turbine disc cracking and results of laboratory tests.

This prediction method takes into account two main parameters:

the yield strength of the disc, and the temperature of the disc at the bore area where the cracks of concern are occurring. The higher the yield strength of the material and the higher the temperature, the faster the crack growth rate will be.

The staff has evaluated the data submitted by the applicant, and in addition, performed independent calculations for crack growth and critical crack size.

The staff concludes that Waterford 3 may be safely operated until the first refueling outage, at which time the LP turbine discs should be inspected.

Inservice inspection will include UT of the bore and keyway areas of each disc and MT and visual inspection of all accessible areas.

The inspection interval has been selected using the criterion that any postulated crack must not be allowed to grow to a size greater than one-half of the critical crack size, assuming a conservative crack growth rate. The staff concludes that these provisions provide reasonable assurance that the probability of disc failure with missile generation is low during normal operation, including transients up to design overspeed.

Sunaary The staff concludes that the total turbine missile risk from high and low trajectory missiles for the Waterford 3 design is acceptably low so that the plant structure, systems, and components, important to safety are adequately protected against potential turbine missiles.

Waterford SSER 4 3-3

4 REACTOR 4.4 Thermal-Hydraulic Design In Supplements No. I and 2 to the SER, all open issues were resolved except the effects of the HID-1 spacer grid design on the minimum departure from nucleate boiling ratio (DNBR).

This SER supplement addresses the remaining open issue.

Since the Waterford 3 thermal-hydraulic design calculations were performed for fuel assemblies having one Inconel and eleven Zircaloy grids per fuel assembly, the staff asked the applicant how the new fuel design, one Inconel and ten Zircaloy grids, affected diversion crossflow in the thermal margin calculations.

The applicant responded that the HID-1 grid would increase crossflow and tur-bulent mixing; therefore, using the standard grid loss coefficients in the DNB analysis was conservative.

The applicant supplied additional information (CE Report Un-numbered, January, 1982; CE Report Un-numbered, February, 1982) which did not support the original contention but did demonstrate that the effects of spacer grid design on DNBR were negligible.

Based on our review of the above information, the staff concludes that the use of the standard grid design in the Waterford 3 thermal-hydraulic design analyses is acceptable. The staff will ensure that the actual fuel design is used in determining the DNBR trip set points and the CPC algorithms.

h 4.4.1 Summary This SER supplement addresses the effects of the HID-1 grid design on the mini-mum DNBR.

Based on our review of tha information provided by the applicant, l

the staff considers this issue resolved with the restrictions given above. We also concluded that the tnermal-hydraulic design presented in Section 4.4 of l

l the FSAR is acceptable for issuance of an operating license.

4.4.2 References Un-numbered, " Response to Questions on Effect of Spacer Grid Design on DNBR Prediction," January 1982.

Un-numbered, " Response to Supplementary Question on Effect of Spacer Grid Design on DNBR Prediction," February, 1982.

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Waterford SSER 4 4-1

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5. 4 Component and Subsystem Design 5.4.3 Shutdown Cooling System In SSER No. 2, the staff reported that the applicant had demonstrated compliance with BTP RSB 5-1.

The applicant had stated that the plant has the capability to be placed into shutdown cooling within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> even though a single failure results in a loss of one of the two atmospheric relief valves and the nonsafety grade steam cross tie is not utilized in the analysis.

However, because the cooldown method described by the applicant requires relatively complex operator action, the staff required LP&L to submit the operating procedures and details of the training program for staff review.

The staff has since reviewed this material, provided comments to LP&L, and been satisfied that the comments have been incorporated into the procedures.

Therefore, the staff considers this issue to be resolved.

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6 ENGINEERED SAFETY FEATURES 6.3 Emergency Core Cooling System Section 6.3.3 of the SER stated that the staff will require experimental verification that the Waterford 3 plant can operate in the recirculation mode without cavitation or air entrainment problems.

The applicant has committed to full-scale model tests of the SIS sump for vortex formation and NPSH.

The NPSH calculations will be verified by using the pump performance curves developed during preoperational testing and the data from the SIS sump model tests.

The staff required the test model to include the entire containment emergency sump compartment with gratings above the sump screen, the flow path entering the sump compartment and the proposed screen partition in between the two pump suctions. Also, the staff required that the amount of air entrainment be measured during tests.

The applicant has submitted a r$oort of the model testing of the SIS sump by a letter dated June 28, 1982.

The applicant has performed tests of containment emergency sump hydraulic behavior to study intake head losses and vortex control using a full scale simulation.

The test model includes a grating cage encapsulating the end of the intake pipe and a screen cage around the top of the sump opening with a grating floor above the screen cage, simulating the emergency containment sump features used at Waterford.

During the tests, heated water was circulated through the sump system at flow rates greater than the maximum value postulated for the worst recirculation case, and at water depth equal to the minimum postulated water level after a LOCA.

In the test, the model screen cage was partially blocked with various geometry blockages to simulate the effects of an accumulation of debris.

The tests demonstrated that the Waterford 3 sump arrangement, including the grating cage, screen cage and grating floor above the sump, prevents vortexing during post-LOCA recirculation.

The staff has reviewed the test report and observed a part of the tests, and concludes that for up to 50 percent blockage of the sump screen cage, the sump performance is not degraded.

Significant (i.e., more than 50 percent) blockage i

of the sump screen is precluded by insulation design (see Section 6.3.2 of the SER).

Since the measured sump inlet losses are smaller than the assumed value for analyses and flow blockage exceeding 50 percent is precluded by insulation design, and the pump runout flow was conservatively used for the calculation of NPSH requirements, we conclude that the Waterford 3 emergency containment sump design is acceptable.

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Waterford SSER 4 6-1

11 RADI0 ACTIVE WASTE SYSTEM 11.3 Process and Effluent Radiological Monitors The staff in its July 1981 SER reached the following conclusions with respect to the normal and potential release pathways:

(1) The containment purge line does not include either an in-line process monitor or an automatic control feature which would terminate the release upon a high radiation signal.

It is our position that an in-line monitor is required for the containment purge line and that isolation of the containment purge on a high radiation signal from either the process monitor or the stack monitor be an automatic control feature and not dependent upon operator action.

(2) The applicant has indicated that upon a high radiation signal from the normal exhaust monitor of the fuel handling building the plant operator will be alerted to the fact that additional radiation sur-veys and sampling are required to determine the source of the radio-active leakage.

It is our position that, upon a high radiation signal from the normal fuel handling building exhaust monitor, the normal fuel handling building exhaust should be automatically iso-lated and release routed through the fuel handling building ESF filter system.

(3) The spent fuel treatment system does not contain any process monitors which alert the plant operator of the buildup of activity in the spent fuel pool.

It is our position that such a monitor is required.

(4) The applicant has indicated that the contents of the regenerative waste tank are pumped to the regenerative w&ste transfer sump.

From there the wastes are pumped to the waste collection basin #2 (Unit 1 and 2's metals waste pond).

There is no radioactivity monitor on this line.

It is our position that since this is an unmonitored release point, a radiation monitor will be required to be installed.

The applicant has provided the staff a letter dated April 13, 1982 which dis-cusses the analysis conducted and the design and procedural changes made.

In accordance with the staff's request, the applicant has agreed to provide automatic isolation of the containment purge line upon receipt of a high radia-tion signal from the plant stack radiation monitors.

The applicant provided an analysis of two release scenarios as evidence that their design of the fuel handling building exhaust was adequate.

The staff's concern, with respect to the operation of the fuel handling building exhaust, was with the design of the system and, in particular, with the operation pro-cedures that were to be the mode of operation for the plant operator.

The applicant has changed its procedures so that the operator will isolate the fuel Waterford SSER 4 11-1

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1 handling building on an alarm from the normal exhaust.

This change satisfies the staff's concerr.s.

The revision to SRP 11.5, NUREG-0800, no longer requires that the spent fuel treatment syster,i contain process monitors if the spent fuel pool is not dis-charged as an effluent.

Therefore, the staff will not require a monitor as originally indicated in the SER.

The applicant and the staff have discussed the pumping of the contents of the regenerative waste tank to the regenerative waste transfer sump and the dis-charge of the sump to waste collection basin #2 (Unit 1 and 2's metals waste pond). The applicant agreed to direct steam generator blowdown treatment system filter flush water and demineralizer regenerative waste to the radio-active waste management system when confirmed primary to secondary leakage exists. This was documented in Amendment 23 to the FSAR.

The staff has agreed to this method of handling the waste in lieu of installing a radiation monitor on line from the regenerative waste transfer sump.

With the above changes, the process and effluent monitoring system is in accor-dance with requirements of GDC 60, 63, and 64 of Appendix A to 10 CFR Part 50 and the guidelines of Regulatory Guide 1.21.

Waterford SSER 4 11-2

15 ACCIDENT ANALYSIS 15.3 Limiting Accidents 15.3.3 Loss-J '-Coolant Accident Waterford 3 t r ~5 analysis does not assume any steam generator tubes are plugged.

tube plugging is treated on an as needed basis for C-E operating The effect<.

plants.

A v,sitivity analysis assuming 6 percent plugging showed minimal cha,'aas,.

ECCS performance and no change in the allowable peak linear heat gener.icn d e.

Based on this, the current ECCS performance analysis, which does not

.,4 der steam generator tube plugging, is acceptable and no new analye'

• equired unless the plugging exceeds 6 percent.

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Waterford SSER 4 15-1

22 TMI-2 REQUIREMENTS 22.2 Discussion of Requirements II.K.2.13 Thermal Mechanical Report - Effect of High Pressure Injection Vessel Integrity for Small Break Loss of Coolant Accident With No Auxiliary Feedwater In Amendment 27 to the FSAR, the applicant has referenced a report "CEN-189" prepared by CE for CE Owners Group to address this subject for NRC Staff review.

Staff review of this item will be covered in NRC Unresolved Safety Issue A-49 " Pressurized Thermal Shock."

II.K.3.30 Revised Small Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K In Amendment 28 to the FSAR, the applicant has referenced a report "CEN-203, Revision 1" which addresses the justification of CE small break LOCA methods.

This report is being reviewed by the staff.

We will report our evaluation of this CE topic report in a later supplement to this SER.

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t t

1 i

Waterford SSER 4 22-1

APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW March 18, 1982 Letter from applicant transmitting ASME/B&PV Code,Section XI program for repair to reactor vessel closure head flange and control element drive mechanism welds March 19, 1982 Letter from applicant transmitting ASME/B&PV Code,Section XI repair program for relocation of steam generator nameplates March 25, 1982 Letter from applicant forwarding " Response to Supple-mentary Question on Effect of Spacer Grid Design on DNBR Production" (proprietary)

March 27, 1982 Letter to applicant concerning depressurization and decay heat removal March 29, 1982 Letter from applicant transmitting "CPC/CEAC Protection Algorithm Test Plan" "CPC/CEAC Software Modifications for Waterford Unit No. 3," CEL-197(c)-P, and " Safety Evaluation of the Reactor Power Cutback System,"

CEN-200-P (proprietary and nonproprietary versions)

March 30, 1982 Letter from applicant concerning control of heavy loads April 9,1982 Board Notification 82 ACRS Letter on Reliability of the Shutdown Heat Removal System on the System 80

(

Design April 13, 1982 Letter from applicant addressing Effluent Treatment j

Systems Branch concerns April 14, 1982 Letter to applicant discussing upcoming May 10-13, control room visit April 15, 1982 Generic Letter 82 Transmittal of NUREG-0909 relative to the Ginna tut'e rupture l

April 16, 1982 Letter to applicant concerning adequacy of operating j

procedures needed to place plant into shutdown cooling l

assuming loss of an atmospheric valve April 16, 1982 Letter to applicant forwarding evaluation of Fermi's post-accident chemistry procedures i

l Waterford SSER 4 A-1

l April 20, 1982 Letter to applicant forwarding list of human engineering discrepancies identified in the Preliminary Control Room Assessment report April 20, 1982 Generic Letter 82 Environmental Qualification of Safety-Related Electric Equipment April 21, 1982 Mcating with applicant to discuss its plans for addressing human engineering discrepancies in control room and to discuss proposed agenda for control room

~

audit April 22, 1982 Letter to applicant advising of relaxed schedule for completion of emergency operating procedures April 26, 1982 Submittal of Amendment No. 26 April 30, 1982 Letter from applicant forwarding information regarding control room April 30, 1982 Issuance of Supplement No. 3 to Safety Evaluation Report May 3, 1982 Letter to applicant forwarding schedule for control l

room design audit to be conducted May 10-14 l

May 6, 1982 Letter from applicant concerning depressurization and decay heat removal May 10-13, 1982 Staff audit of control room May 12, 1982 Letter from applicant forwarding " Turbine Missile Report-R,esults of Probability Analyses of Disc Rupture l

and Missile Generation," CT-24836 Revision 0, October 1980 (proprietary)

May 14, 1982 Letter from opplicant confirming changes in design basis pipe breaks May 17, 1982 Letter from applicant transmitting monthly staffing report May 21, 1982 Letter from applicant forwarding Annual Report May 26, 1982 Letter from applicant regarding schedule for SQRT audit May 26, 1982 Letter from applicant transmitting comments on the Safety Evaluation Report and Supplements 1 and 2 to the Safety Evaluation May 31, 1982 Letter from applicant regarding proposed changes to technical specification requirements Waterford SSER 4 A-2

June 7, 1982 Letter to applicant forwarding request for additional information June 9, 1982 Generic Letter 82 Transmittal of NUREG-0916 relative to the restart of R. E. Ginna Nuclear Power Plant June 14, 1982 Letter to applicant forwarding control room design review / audit report and advising of July 20 meeting to discuss corrective actions to be taken June 15, 1982 Meeting with applicant to discuss its quality assurance program June 17, 1982 Generic Letter 82 Reactor Operator and Senior Reactor Operator Examinations June 17, 1982 Letter to applicant requesting summary report on security matter June 18, 1982 Letter from applicant transmitting monthly report on staffing June 22, 1982 Board Notification 82 Steam Generator Tube Rupture June 22, 1982 Letter to applicant transmitting comments on Emergency Action Levels of Radiological Emergency Plan.

June 23, 1982 Letter from applicant forwarding control wiring diagrams June 25, 1982 Letter from applicant providing responses to questions on operability of containment purge valves June 25, 1982 Letter from applicant transmitting "CPC/CEAC Phase 1 Software Verification Test Report," CEN-209(c)-P, "Waterford 3 Cycle 1 CPC/CEAC Data Base Document,"

CEN-207(c)-P, and "Waterford 3 Cycle 1 CPC/CEAC Phase II Software Verification Test Report," CEN-208(c)-P-proprietary and nonproprietary versions l

June 28, 1982 Letter from applicant transmitting report of model test-ing of the safety injection system sump July 1, 1982 Letter from applicant concerning pressurizer safety valve functionability July 2, 1982 Letter from applicant in response to staff comments on Natural Circulation and Cooldown Procedure July 2, 1982 Letter from applicant transmitting information on containment purging Waterford SSER 4 A-3

July 2, 1982 Generic Letter - Commission Approved Charter for the Committee to Review Generic Requirements (CRGR).

July 7, 1982 Letter from applicant providing confirmation on several items July 13, 1982 Letter from applicant providing confirmation regarding reactor coolant pump sheared shaft analysis July 15, 1982 Letter to applicant transmitting comments on revision to site security plan July 16, 1982 Letter from applicant regarding schedule for conducting the seismic qualification review team audit July 20, 1982 Letter to applicant forwarding draft technical evaluation report on ccitrol of heavy loads July 20, 1982 Meeting with applicant to discuss control room audit results July 26, 1982 Letter from applicant forwarding comments on Supplement No. 3 to the Safety Evaluation Report July 27, 1982 Meeting with Combustion Engineering to discuss its responses to staff questions on CESEC model July 26, 1982 Letter from applicant transmitting monthly staffing report July 28, 1982 Submittal of Amendment 28 consisting of information on emergency shutdown from outside control room, small break LOCA methods, design review of plant shielding, and other information August 6, 1982 Letter from applicant confirming schedule for upcoming EQ audit August 9, 1982 Board Notification 82 Accident Sequence Precursor Program Report August 9, 1982 Generic Letter 82 Submittal of Documents to the NRC August 11, 1982 Meeting with applicant to discuss its proposed design verification program August 19, 1982 Letter from applicant advising of transmittal of updated SQRT Summary Sheets August 19, 1982 Letter from applicant providing responses to questions at June 15 meeting on its quality assurance program August 23, 1982 Letter from applicant forwarding monthly staffing report 1

Waterford SSER 4 A-4

August 25, 1982 Letter from applicant advising that responses to questions on depressurization and decay heat removal to be provided by August 15, 1983 August 26, 1982 Meeting with applicant to discuss its proposal for a design verification program August 30, 1982 Letter from applicant concerning compliance with 10 CFR 73.21 Aug 31 - Sep 3, 1982 Seismic qualification audit September 1, 1982 Letter from applicant transmitting revised Program Plan for Independent Design Review of the Emergency Feedwater System September 3, 1982 Letter from applicant regarding Core Protection Calculator software functional inconsistency September 8, 1982 Letter from applicant advising of review of heavy load handling September 9, 1982 Letter from applicant concerning depressurization and decay heat removal September 13, 1982 Letter from applicant forwarding pump and valve inservice test plan September 15, 1982 Letter from applicant transmitting press release announcing delay in commercial operation until January 1984 September 15, 1982 Letter from applicant transmitting monthly staffing report September 21, 1982 Letter from applicant transmitting interim procedure l

for post-accident estimate of core damage September 23, 1982 Letter to applicant forwarding open items on purge valve operability l

Septerrber 24, 1982 Board Notification 82 Semiscale Test Results September 27, 1982 Letter from applicant providing justification for test exception to delete the stuck control rod physics test September 27, 1982 Letter from applicant transmittal of record of Seismic Qualification Review Team audit September 27, 1982 Letter from applicant regarding recent meeting to discuss a proposed revision to Chapter 14 concerning transferring a system from Construction to Startup to Plant Operations Waterford SSER 4 A-5

September 28, 1982 Letter from applicant transmitting test reports on fire resistance rating of penetration seals September 28, 1982 Letter to applicant advising that revised program plan for independent quality assurance evaluation is acceptable September 29, 1982 Letter to applicant transmitting copy of ANSI N 14.6 evaluation October 1, 1982 Board Notification 82 Welds in Main Control Panels October 3-5, 1982 Audit of applicant's training program October 4, 1982 Letter from applicant forwarding " Program Plan -

Independent Design Review of Waterfcrd SES Unit No. 3 Emergency Feedwater System," Revision B, September 13, 1982.

Waterford SSER 4 A-6

Appendix B

\\

Bibliography

• USNRC REGULATORY GUIDES l

1.115 Protection Against Low-Trajectory Turbine Missiles.

COMBUSTION ENGINEERING REPORTS i

" Response to Questions on Effect of Spacer Grid Design on DNBR Prediction,"

January 1982.

" Response to Supplementary Question on Effect of Spacer Grid Design in DNBR Prediction," February 1982.

CEN-189, Evaluation of Pressurized Thermal Shock Effects Due to Small Break t

LOCA's With Loss of Feedwater."

CEN-203, Rev. 1, " Response to NRC Action Plan Item II.K.3.30 Justification of j

Small Break LOCA Methods."

l

• Available for inspection and copying for a fee in the NRC Public Document Room, 1717 H St., NW, Washington D.C.

Waterford SSER 4 B-1

APPENDIX C NRC UNRESOLVED SAFETY ISSUES C.5 DISCUSSION OF TASKS AS THEY RELATE TO WATERFORD A-49 Pressurized Thermal Shock Severe reactor-system overcooling events in a pressurized water reactor (PWR) which could be followed by repressurization of the reactor vessel can result from a variety of causes.

These include instrumentation and control system mcifunctions and postulated accidents such as small break loss-of-coolant ace.idents (LOCAs), main steamline breaks, or feedwater pipe breaks.

Rapid cocling of the reactor vessel internal surface causes a temperature gradient across the reactor vessel wall.

This temperature gradient results in thermal st' ess, with a maximum tensile stress at the inside surface of the vessel.

Th magnitude of the thermal stress depends on the temperature differences a.ross the reactor vessel wall.

Effects of this thermal stress are compounded f.y the hoop stress if the vessel is repressurized.

As lang as the fracture resistance of the reactor vessel material remains high, such transients will not cause failure.

After the fracture toughness of the vessel is reduced by neutron irradiation, severe thermal transients could cause fairly small flaws near the inner surface to initiate and result in significant cracking.

The vessels of most concern are those with high radia-tion exposure, which are made of material that has a relatively high sensitivity to radiation damage (such as those made with welds of high copper content).

For failure of the RPV to occur, a number of contributing factors must be present.

These factors are: (1) a reactor vessel flaw of sufficient size to initiate and propagate; (2) a level of irradiation (fluence) and material properties and composition sufficient to cause significant embrittlement (the exact fluence is dependent upon materials present, i.e., high copper content causes embrittlement to occur more rapidly); (3) a severe overcooling transient with repressurization; and (4) the crack resulting from the propagation of initial cracks must be of such size and location that the vessel fails.

The staff's preliminary review of overcooling events and their probabilities included a study on overcooling events at Babcock and Wilcox (B&k) plants; a survey of operating experience on Westinghouse and Combustion Engineering plants; a review of available accident analysis in Final Safety Analysis Reports and in vendor topical reports; and a preliminary probabilistic analysis.

The preliminary results of these evaluations indicate that there is a probability of about 10 3 per reactor year that a B&W-designed plant will experience a severe overcooling transient similar to or worse than that experienced at Rancho Seco on March 20, 1978.

The Rancho Seco transient is the most severe overcooling tran-sient experienced by any PWR in the United States.

The staff estimates that l

the probability of such an overcooling event in CE-or W-designed reactors is l

lower, perhaps by an order of magnitude, than for B&W-designed reactors.

This difference is based on design differences and on operating experience.

Waterford SSER C-1

In the 1978 Rancho Seco transient, reactor pressuie was maintained at a fairly high level (1500 psig to 2100 psig) throughout the cooldown.

The minimum temperature of the reactor coolant (280*F) during the transient was high enough so that material toughness of the reactor vessel was not affected.

This evaluation leads the staff to believe that if this transient were to be repeated at Ranch Seco or any other B&W-designed facility within the next few years, the reactor vessel failure would still be unlikely.

Nonetheless, the.

possibility of vessel failure as a result of an overcooling event cannot be completely ruled out.

If an overcooling event such as that at Ranch Seco were to occur, even for the vessel with the most limiting material properties in existence today, the staff would not expect a failure.

The staff conclusion is supported by ORNL analyses of the Rancho Seco event which indicate that the threshold irradiation level (neutron fluence) for crack initiation (that is, small cracks growing to larger ones assuming conservative initial material properties such as RTNDT=40 F and copper content of 0.35 percent) would be in the range of 1019 neutrons /cm.

The highest neutron fluence to date in a B&W-designed facility is less than half the minimum value listed above.

It would, therefore, be several years before any B&W-designed facility reached its threshold irradiation level.

The reactor vessels in CE and W facilities have somewhat higher fluences; however, other mitigating factors--such as lower values of initial RT

- provide NDT a significant margin to failure should an overcooling event similar to that at Rancho Seco occur.

As a result of its evaluations to date, the staff has concluded that the probability of a severe overcooling transient (similar in magnitude to the Rancho Seco event) is relatively low.

For B&W-designed reactors this probability is estimated to be about 10 3 per reactor per year, and for W-and CE-designed reactors, it is lower, perhaps by an order of magnitude.

Furthermore, the staff anticipates that this issue will be resolved before the irradiation history at Waterford is large enough to cause a significant pressurized thermal shock concern.

Therefore, based on the foregoing, the staff concludes that Waterford can be operated before resolution of this issue, without undue risk to the public.

Watertord SSER C-2

APPENDIX D PRINCIPAL CONTRIBUTORS TO SSER NO. 4 K. Campe Siting Analysis R. Gonzales Hydrology J. Schiffgens Materials Engineering C. Liang Reactor Systems J. Holonich Core Performance J. Hayes Effluent Treatment Systems C. Anderson Generic Issues Branch i

l l

l Waterford SSER 4 D-1

i APPENDIX E ADDITIONAL ERRATA TO THE SAFETY EVALUATION REPORT Section 11.2.1.5, Page 11-12 The last line of Section 11.2.1.5 should read "... discharged through the oil separator to the Arpent Canal."

Section 11.2.2.5, Page 11-16 Eighth line should read "2.0 x 10 4 uCi/sec."

Sectiqn 6.3.2, Page 6-22 Second paragraph, item (1), delete 2SI-V1556 and 2SI-V1559 Waterford SSER 4 E-1

" ' "" 335

1. mm mBm weem m er 7n u.s. NucLEan nEcuLATony commission NUREG-0787 BIBLIOGRAPHIC DATA SHEET Supplement No. 4

4. TITLE AND SUBTITLE (Add Volume No., ereargres*1

2. (Leave b/m&J Safety Evaluation Report Related to the Operation of W2terford Steam Electric Station, Unit 3

3. RECIPIENT 3 ACCESSION NO.

7. AUTHOR (S)

5. DATE REPORT COMPLETED MON TH l YEAR October 1982

9. PERFORMING ORGANIZATION N AME AND MAILING ADDRESS (Inclu* I,a Co*/

DATE REPORT ISSUED "T"

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U.S. Nuclear Regulatory Commission to 1W Office of Nuclear Reactor Regulation ft,,,,,,,,

W:shington, D.C.

20555

8. (Leave Nank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS //nclude lip Codel p

Same as 9. aboVO

11. CONTRACT NO.

13. TYPE OF REPOR T PE RIOD COV E RE D (/nclus,re defes/

15. SUPPLEMENTARY NOTES

14. (Leave WmA)

Pertains to Docket No. 50-382

16. ABSTR ACT (200 words or less)

Supplement No. 4 to the Safety Evaluation Report for the application filed by Louisiana Pbwer & Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Iouisiana has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of infomation submitted by the applicant since the Safety Evaluation Report and its three supplements were issued.

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17. KEY WORDS AND DOCUMENT AN ALYSIS 17a DESCRIPTORS l

17b. IDENTIFIE RS!OPEN ENDE D TE RMS 18 AV AILABIL4TY ST ATEMENT 19 SE CURITY CLASS IThis reportl

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