ML20043G852

Application for Amends to Licenses NPF-76 & NPF-80,revising Tech Spec 4.5.6.2.b to Delete Automatic Closure Interlock on RHR Suction Valves

ML20043G852
Person / Time
Site:	South Texas
Issue date:	06/12/1990
From:	Vaughn G HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20043G853	List: ML20043G852 ML20043G857
References
ST-HL-AE-3485, NUDOCS 9006210195
Download: ML20043G852 (12)

Text

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.. .  ; j The Light i co mp any South Temas Project Electric Generating

- Hounton Lig ting & Power Station F.O.Bom249 Wadsworth, Teman 77483  ;

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File No.: 09.06, G20.01 10CFR50.90 l 6

U. S. Nuclear Regulatory Commission I Attention: Document Control Desk Washington, DC 20555 South Texas Project Electric Generating Station  :

Units 1 & 2 Docket Nos. STN 50-498, STN_50 499  ;

Proposed Amendment to the D; tit 1 and Unit 2 Technical Specification 4.5.6.2,J The purpost of this submittal is to request a change to Technical i Specifications deleting the automatic closure irterlock on the Residual Heat Removal suction valves.

Pursuant to 10CFR50.90, Houston Lighting & Power Company (HMP) hereby proposes to amend its Operating Licenses NPF 76 and NPF 80 by incorporating j the attached proposed change to the Technical Specifications for the South  ;

Texas Project Electric Generating Station (STPEGS) Units 1 and 2. l HMP has reviewed the attached proposed amendment pursuant to 10CFR50.92 ,

and determined that it does not involve a significant har.ards consideration. '

The basis for this determination is provided in the attachments. In addition, j based on the informattan contained in this submittal and the NRC Final Environmental Asse:.sment for STPEGS Units 1 and 2, HMP has concluded that, ;I

, pursuant to 10CTR51, there are no significant radiological or non radiological J impacts associated with the proposed action and the proposed license amendment will not have a significant effect on the quality of the environment, i

The STPEGS Nuclear Safety Review Board has reviewed and approved the i proposed changes.

In accordance with 10CFR50.91(b), HMP is providing the State of Texas with a copy of this proposed amendment, j l

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Houston Lighting & Power Company

!- J" 9' South Texas Project Electric Generating Station ST HL $ 3 8S File No.: 09.06, C20.01 Page 2 If you should hava any questions concerning this matter, please contact Mr. M. A. McBurnett at-($12) 972 8530 or myself at (512) 972 7921.

G. E. Vaughn '

Vice President Nuclear Generation  !

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Attachment:

1. Significant Hazards Evaluation for the Residual Heat Removal Automatic Closure Interlock

2. Proposed Technical Specification Change 4.5.6.2.b and Bases 3/4.7.1.7 ,

3. Mark up of the Final Safety Analysis Report '

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Juni 12 1990 St.HL.Ak.3485 Housten Lighting & Power Compa:y . Filo H3.: 09.06, C20.01

. ' South Temas Project Dectric GenerstEg Station p.g. 3 i

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Regional Administrator, Rigion IV Rufus S. Scott Nuclear Regulatory Commission Associate General Counsel 611 Ryan Plaza Drive., Suite 1000 Houston Lighting & Power Company

. Arlington,,TX 76011 P. O. Box 61867 Houston, TX 77208 Coorge Dick, Project Manager U.S. Nuclear Regulatory Commission INPO Washington, DC 20555 Records Center 1100 circle 75 Parkwey J.'I. Tapia . Atlanta,-CA 30339 3064 Senior Resident ~ Inspector i c/o U. ! $'. Nuclear Regulatory Dr. Joseph M. Hendrie l , commission 50 Be11 port Lane P. 0. Box 910- Be11 port, NY 11713 Bay City,1TX 77414 D. K. lacker J. R. Newman,. Esquire Bureau of Radiation Control Newman & Holtzinger, P.Ci Texas Department of Health 1615 L Street, N.W. 1100 West 49th Street l Washington, DC 20036 Austin, TX 78704

.D. E. Ward /R.,P. Verret Central Power & Light. Company b P. O. Box 2121

Corpus Christi, TX 78403 l

h J. C..Lanier Director of Generation

-City of Austin Electric Utility 721 Barton Springs Road Austin, TX 78704 i R. J. Costello/M. T. Hardt city Public Service Board P. O. Box 1771 San Antonio, TX. 78296 I

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l Revised 12/15/89 L4/NRC/

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UNITED STATES OF AMERICA NUCLEAR REGUIATORY COMMISSION In the Matter )

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Houston Lighting & Power ) Docket Nos. 50 498 Company, et al., ) 50 499

) i South Texas Project ) i Units 1 and 2 ) i AFFIDAVIT C. E. Vaughn being duly s'rorn, hereby deposes and says that he is Vice President, Nuclear Generation of houston Lighting & Power Company; that he is '

duly authorized to sign and file with the Nuclear Regulatory Commission the attached proposed changes to the South Texas Project Electric Generating Station Technical Specifications 4.5.6.2.b and Bases 3/4.5.6 is familiar with the  :

content thereof; cnd that the matters set forth therein are true and correct to the best of his knowledge and belief.

J, G. E. Vaughn '

Vice President, Nuclear Generation Subscribed and sworn to before me, a Notary Public in and for The State of Texas this /fday of hwnt,

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f ATTACHMENT 1 SIGNIFICANT llAZARDS EVALUATION FOR THE RESIDUAL llEAT REMOVAL AUTOMATIC C1hSURE INTER 14CK n

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ATTACHMENT 1 SIGNIFICANT HAZARDS EVALUATION FOR THE RESIDUAL HEAT REMOVAL AUTOMATIC C14SURE INTERIDCK Packtround i

The Automatic Closure Interlock (ACI) feature of the Residual Heat Removal suction valves provides for automatic closure of the valves whenever the Reactor Coolant System (RCS) pressure increases above a level approximating the RHR system design pressure and is intended to ensure that the double barrier (two closed valves in each suction line) is not compromised l by operator error during plant startup. However, during RHR operation, the presence of the ACI offers the potential for inadvertent closure of the valves  ;

and subsequent loss of cooling. Closure of the suction valves also isolates the RHR system pressure relief valves which can help protect the RCS from overpressure transients during water solid conditions. Closure of the suction valves also isolates the low pressure letdown line. These negative aspects of  ;

the ACI have prompted the NRC staff to reconsider the overall need for the suction valve ACI. The NRC indicated that removal of the ACI feature should ,

be based on an evaluation of the potential of an interfacing systems Loss of

Coolant Accident (LOCA) versus the reduced potential for RHR system failures and low temperature overpressurization transients. South Texas Project Electric Cenerating Station (STPEGS) has performed a quantitative assessment of the reduced potential for RHR system failure achieved by deleting the RHR ACI, Additionally, STP2GS has evaluated the potential of an interfacing I systems IDCA and low temperature overpressurization transients, l

By letter dated April 22, 1988 the Westinghouse Owners Group j submitted to the NRC a report, Residual Heat Removal System Autoclosuro l Interlock Deletion Report for the West!sehouse Owners Group, WCAP 11736, i STPEGS is a Group 2 plant as describen ~ "J 11736. The purpose of j WCAP 11736 is to allow various plants to . trence the analysis performed but l i does not take the place of a plant specific ,,alysis. WCAP 11736 recommends- 1 the deletion of the RHR ACI. The results of the interfacing systems IDCA analysis show that the frequency of the event decreases with the removal of the ACI feature. The results of.the RHR unavailability analysis performed in WCAP 11736 show that the removal of the ACI feature increases the RHR availability. By letter dated August, 1989 the NRC staff found that the removal of the ACI for Westinghouse plants covered by WCAP 11736 can produce a net safety benefit provided certain improvements are irplemented. The staff found that VCAP 11736 may be referenced to show compliance with those items that are not plant specific.

Generic Letter 8817 on Reduced RCS Inventory Operation, requested i all operating plants to identify and change Technical Specifications that j restrict or limit the safety benefit of the actions identified in the Generic l Le t te r . The ACI is a feature required by Technical Specifications that limits the safety benefit of Generic Letter 88-17.

Proposed Chance HL&P proposes to delete the surveillance requirement for the ACI feature from the Technical Specifications. Attachment 2 contains the proposed change.

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Attachment 1  ;

Page 2 Safety Evaluation o Probabilistic Analysis STPECS performed a probabilistic scoping analysis for a loss of RHR during an outage. A baseline analysis with the ACI was performed, followed by i an analysis with the ACI deleted. The net change in system availability was i determined in order to quantify the effect of the ACI on loss of RHR. In this analysis a logic model in the form of a fault tree was employ 6d to delineate i the various pathways in which a loss of all RHR could occur.  ;

With the ACI feature the dominant contributors to loss of RHR are failures which affect both Trains A and B, with Train C either failing or being unavailable due to testing and maintenance. The other important '

contributors involve a failure of a running train (A or B), and an ACI failure ,

in the remaining trains before the initial failure is repaired. These failure combinations comprise over 95% of the total probability of failure.

The failure combinations and the associated probabilities for loss of RHR without the ACI feature were determined. The largest contribution is a common cause failure of all three RHR pumps. The next largest contributor is the independent failure of running trains while C Train is in maintenance or a common cause failure of the running pumps while C Train is in maintenance.

These failure combinations comprise almost 90% of the total probability of the ,

ACI deleted case.

The STPEGS probabilistic analysis determined that the deletion of the ACI results in a decrease in the likelihood of loss of RHR during a seven week mission time from 7.43E 03 to 1.67E 04.

o-Interfacing Systems LOCA Analysis in Support of Removal of RHR Auto Closure Interlock .

The classic interfacing systems Loss-of Coolant Accident (LOCA), the "V" sequence, is caused by a failure of the pressure boundary between the reactor coolant system (RCS) and a connecting low pressure system, such as residual heat removal (RHR). The reason for the interest in this sequence is that the resulting failure has the potential for bypassing the containment building thus leading to a release directly to the environment with no mitigation. This sequence was first identified in WASH 1400. As described in WASH 1400, the following is provided:

1. Gross failure or leakage in the two, series low pressure injection check valves which provide the high pressure to low pressure boundary for the RHR/ Low Head Safety injection (LHSI) system.

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Attachment 1 i Page 3

2. Because the RHR system is located outside containment in the '

i WASH 1400 reference plant, the boundary failure results in pressurization of the RHR system outside containment. Since system design pressure is only 600 psig, rupture was postulated to occur in the RHR piping.

3. The RHR system pumps and piping provide the low pressure safety iajection capability at the reference plant, thus this failure leads to a LOCA and fails the system that is needed to mitigate ,

the IhCA. Core damage was assumed to occur with no mitigation possible.

The RHR suction valve autoclosure interlock presently installed at STPECS was designed to prevent the interfacing systems LOCA through the RHR suction piping by automatically closing the series RCS/RHR suction valves on >

increasing RCS pressure. This interlock is a backup to operator acti)n to i isolate the RHR system after the system has been used for shutdown cooling.

The RHR System installed at STPEGS is somewhat different than the system analyzed in WASH 1400. The major design diff..ences are:

1. The system is completely contained inside the containment building.

2. There are three completely separate trains. Any rupture in a

, single train does not affect the other trains.

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3. The RHR pumps are not used to provide the low pressure safety injection function. Three, physically separate, low head safety injection pumps in the fuel handling building provide this function.

4. Although the IllSI system interfaces with the RHR system on a train hasis, there is an additional check valve inside containment between the IllSI pump and the RHR system tie.

If a failure were to occur that leads to PJIR system overpressurization, one train of IJISI could be affected and a 1DCA could result inside containment, but two trains of IllSI remain available to mitigate the effects of the LOCA. This accident is within the bounds of the large break IDCA analyzed in chapter 15 of the STPEGS FSAR which assumed a single failure that disables one train of safety injection with an additional train

( of safety injection discharging to the affected loop. The likelihood of this L event occurring is less than 1% of the frequency of occurrence of the design l basis large break 1hCA based upon the initiating frequency presented in the South Texas Project Probablistic Safety Assessment (2.0E-04).

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Attachment 1 Page 4 To summarize, in the case of S'iPEGS, an RHR interfacing systems LOCA, should it occur, would result in a release of coolant to containment r ather than bypassing containment. The break will be isolated by the containment preventing the direct release of radioactive material to the environment in the remote event of core uncovery. The availability of LHSI will still be assured by two, independent, unaffected trains, o RCS Pressure Control During Low Temperature Operation and the Effect of Removal of the RHR Auto Closure Interlocks Cold overpressure protection at STPEGS is provided by the pressurizer power operated relief valves (PORVs). The STP PORVs are safety related and Class 1E powered. They are designed in accordance with ASME code; are qualified via the Westinghoute pump and valve operability program described in Section 3.9.3.2.1 of the UFSAR; are seismically qualified as described in Section 3.10N; and are environmentally qualified as described in Section 3.11N.

l The residual heat removal (RHR) discharge relief valves are not required to operate to mitigate the consequences of an overpressurization event at low reactor coolant system (RCS) temperature. If the RHR discharge relief valves are available (e.g. not isolated by inadvertent autoclosure operation), plant risk is decreased because of the additional redundancy j available to mitigate an overpressure transient occurring at low RCS l temperatures.

Each RHR train is equipped with a pressure relief valve that is designed to relieve the combined flow of all the charging pumps at the relief valve set pressure of 600 psig. No credit is taken in the cold overpressure analysis for these relief valves in protecting the RCS from overpressurization

-(UFSAR Appendix 5.4A). These valves are installed downstream of the RHR suction isolation valves and would be automatically isolated by any autoclosure signal.

While it is true that the ACI initiates an automatic closure of the RHR suction valves on high RCS pressure, overpressure protection of the RHR  ;

system is provided by the RHR system relief valves and not by the slow acting suction valves that isolate the RHR system from the RCS, A o RHR Valvo Position Alarm An alarm will be added to alert the operator to a RHR suction valve being open with the RCS pressurized. The intent of this alarm is to alert the operator that a RHR suction valve is not fully closed, and that double isolation from the RCS to the RHR is not maintained. Position indication will be provided from the spare Limitorque Limit switch contacts on the RHR suction ,

valves. The alarm will actuate if the valve is open and the pressure is  !

greater than the open permissive setpoint and less than the RHR design pressure minus the RHR pump head pressure. This alarm will not be affected by j power lockout of the RHR valve.

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e Attachment 1 Page 5 The alarm response procedure will be revised to direct the operator to close the open RHR suction valves. If this is not possible the operator will be instructed to depressurize the RCS to below the alarm setpoint. ~ A procedure will be revised to include testing of the RHR suction valve alara.

In addition the procedures listed in WCAP 11736 have been reviewed for the impact of removing the ACI and installing a control rooi alarm.

o Impact of Hardware Changes The open permissive circuit is not removed or affected and the valve position indication will be available in the control room with power removed from the RHR suction valves. In addition, the RHR motor operator for the suction valves are sized to prevent the valves from opening against full RCS pressure, o RHR Suction Valve Testing Technical Specification Surveillance 4.4.6.2.2 requires leak testing of the RHR suction valves at least once per 18 months, prior to entering Mode 2 whenever the plant has been in cold shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, and if leakage testing has not been performed in the previous 9 months. The valves are required to'be leak tested prior to returning the valves to service following maintenance, repair or replacement work. This leak testing is

. performed as outlined by ASME Code,Section XI. paragraph IW.3427(b). Since this testing is conducted at low temperature and pressure conditions for personnel safety considerations, the train is likely to be restored to service following the test. Since the test must ce conducted on all three trains, at a minimum, at least one train of RHR will be in service following the testing as the plant temperature and pressure are raised to the point where the RHRs is secured to a standby condition. Because this requirement will not allow the testing configuration to be maintained for the suction valves, H1AP will not leak test the valves with the power removed.

Determination of Significant Hazards Pursuant to 10CFR50.91 this analysis provides a determination that the proposed change to Technical Specifications does not involve any l significant hazards consideration as defined in 10CFR50.92.

(1.) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The accidents previously evaluated are the WASH-1400 interfacing LOCA, RCS cold overpressure mitigation, )

and a loss of RHR capability. I l

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Determination of Sinn!ficant Hazards (Cont'd.)

The probabilistic analysis performed by STPEGS used a baseline case of the ACI feature in the RHR system. This was followed by an analysis without the ACI feature. The analysis results concluded that the deletion of the ACI results in a decrease in the likelihood of loss of RHR during a seven week mission time [

from 7.43E 03 to 1.67E 04.

The evaluation of an RHR interfacing systems LOCA at STPEGS resulted in the conclusion that the LOCA would not bypass i containment. Therefore, the containment would prevent the .

direct release of radioactive material to the environment in -

the remote event of core uncovery and two trains of LHSI would '

be available for mitigation due to the design of STPEGS.

Cold overpressure mitigation at STPEGS is provided by the pressurizer power operated relief valves. The residual heat  ;

removal discharge relief ralves are not required to operate to -

mitigate the consequences of an overpressurization event at low reactor coolant syster temperature. With the RHR discharge relief valves availa',1e (e.g. , ACI deleted), plant risk is  ;

decreased because of the additional redundancy available to mitigate an overpressure transient occurring at low RCS  :

temperatures.

The deletion of the ACI was analyzed in WCAP-11736 in terms of (1) the frequency of an interfacing LOCA (2) the availability I of the RHR system, and (3) the effect on overpressure transients with the net result of a safety benefit.

With the above considerations it is concluded that the deletion of the RHR ACI does not involve a significant increase in the ,

probability or consequences of an accident previously (

evaluated. ,

(2.) The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. The q PORVs at STPEGS mitigate an overpressure transient and the RHR discharge relief valves are not utilized in the analysis for mitigation of an overpressurization transient. While it is true that the ACI initiates an automatic closure of the RHR .

suction valves on high RCS pressure, overpressure protection of the RHR system is provided by the RHR system relief valves and i not by the slow acting suction valves that isolate the RHR 1 system from the RCS. The deletion of the ACI will decrease plant risk because without the RHR suction valves closed due to the ACI the RHR relief valves provide additional redundancy, t

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i Attachment I l Page 7 Determination of Significant Harards (Cont'd. )

The STPECS design prevents bypassing containment which is part of the classic interft ,ing systems LOCA, the 'V" sequence._ The l RHR is completely contained in containment and this design  ;

feature prevents bypassing containment. The purpose of ACI l feature is to ensure there is a double barrier between the RHR l system and the RCS when the plant is at normal operating conditions, (i.e., pressurized and not in the RHR cooling )

mode). There are several methods to ensure that there is a j double barrier between the RHR system and the RCS when the  ;

plant is in the normal operating mode. First, plant operating j procedures instruct the operators to isolate the RHR system 1 during plant heatup. l l

Second, an alarm will be installed as part of this change that )

will annunciate in the control room when a RRR suction valve is  ;

open at.d the RCS is at pressure. In conjunction with this alarm, the alarm response procedure will be evised to ensure l that the RHR suction valves are closed upon alarm indication. I Third, the open permissive interlock, which is not being ,

removed, will prevent the opening of the RHR suction valves J unless the RCS pressure is substantially below the RHR design pressure.

Since the PORVs prevent overpressurization of the RHR system during shutdown conditions and several other methods are available to ensure that the RHR systs as isolated from the RCS during normal plant operating conditions, removal of the ACI does not create the possibility of a new or different kind of accident from any accident previously evaluated, s

(3.) The proposed change does not involve a significant reduction in a margin of safety. The probabilistic analysis performed for the deletion of the ACI at STPEGS indicates there is a decrease ,

in the likelihood of a loss of RHR. The potential for an interfacing systems LOCA and low temperature overpressurization transients have been evaluated with the ACI deleted. Therefore the overall safety has been increased.

Conclusion The proposed change is similar to changes approved by the NRC for the s Salem and Diablo Canyon plants.

L Based upon the results of the WCAP-11736 and the STPEGS evaluation l the deletion of the RRR ACI feature does not involve a significant hazards l consideration.

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