ST-HL-AE-3627, Application for Amends to Licenses NPF-76 & NPF-80,changing TS Surveillance 4.4.6.2.2d Re Pressure Isolation Valves

Application for Amends to Licenses NPF-76 & NPF-80,changing TS Surveillance 4.4.6.2.2d Re Pressure Isolation Valves

ML20217A481
Person / Time
Site:	South Texas
Issue date:	11/15/1990
From:	Kinsey W HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217A485	List: ML20217A487 ST-HL-AE-3627, Application for Amends to Licenses NPF-76 & NPF-80,changing TS Surveillance 4.4.6.2.2d Re Pressure Isolation Valves
References
ST-HL-AE-3627, NUDOCS 9011210106
Download: ML20217A481 (9)

Text

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i The Light  !

company l'.O. Box 1700 llouston, Texas 77001 (713) 228 9211

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llouston Lighting k Power .

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November 15, 1990 I ST+1tL AE 3627 )

File No.: 09.06 10CTR50.90 U. S. Nuclear Regulatory Commission -

Attention: Document Control Desk Washington, DC 20555 South Texas Project Electric Generating Station Units 1 and 2 Docket Nos. STN 50 498, STN 50 499 Proposed Amendment to the Unit 1 and 2 Technical Soccification Surveillanec 4.4.6.2.2d The purpose of this submittal is to propose a change to South Texas c Project Electric Generating Station (STPEGS) Technical Specification  :

Surveillance 4.4.6.2.2d. This surveillance verifies pressure isolation valves '

for the Reactor Coolant System are operable following valve actuation due to automatic / manual action or flow through the valves. ,

Houston Lighting & Power Company (illAP) has reviewed the attached proposed amendment pursuant to 10CFR50.92 and determined that it does not involve a significant hazards consideration. The basis for this determination is provided in the attachments. In addition, based on the information contained in this submittal and the NRC Final Environmental AJaessment for STPEGS Units 1 and 2, IllAP has concluded that, pursuant to 10CFR$1, there are no significant radiological or non radiological impacts associated with the proposed action, and the proposed license amendmene will not have a significant effect on the quality of the environment.

The S*dEGS Nuclear Safety Review Board has reviewed and approved the proposed changes.

In accordance with 10CFR50.91(b), IllAP is providing the State of Texas ,

with a copy of this proposed amendment.

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+011210106 901115 PDR ADOCK 03000498 p PDC

- d A1/043.N15 A Subsidiary of Ilouston industries incorporated

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' Houston 1lghting & Power Company -l South Tess Project Dectric Generating $tstion ST H1. AE*3627 f File 30.: 09.06 Page 2 If you should have any questions concerning this matter, please contact Mr. M. A. McBurnett at (512) 972 8530 or myself at (512) 972 7921.

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Warren H. Kinsey, J '

Vice President Nuclear Generation  ;

CCS/sgs- ,

Attachneents: 1. Significant Hazards Evaluation for a  !

Proposed Change to Technical Specification i Surveillance 4.4.6.2.2d ,

2. Proposed Technical Specification  !

- Surveillance'4.4.6.2.2d i

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' Houston l3ghting & Power Company S.HL.AE;3627*06 i

South Tesse Proj,ect Electric Generating Stat,on p,g, 3 cc:

Regional Administrator, Re61on 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 Ceorge Dick. . Proj ect Manager ]

U.S. Nuclear Regulatory Commission INPO j Washington, DC 20555 Records Center 1100 Circle 75 Parkway

- J. 1. Tapia Atlanta, CA 30339 3064  :

Senior Resident Inspector  !

c/o U. S. Nuclear Regulatory Dr. Joseph M. Hendrie  !

Commission 50 Be11 port Lane 1 P. O. Box 910 Be11 port, NY 11713 .

Bay City, TX 77414 D. K. Lacker 1 J. R. Newman, Esquire- Bufl eau of Radiation Control l Newman & Holtzinger, P.C. Texas Department of Health l 1615 L Street, N.W. I't00 West 49th Street ~

- Washington, DC 20036 Austin, TX 78756 3189 R. P. Verret/D. E. Ward

- Central Power & Light Company -F

- P. O. Box 2121- -1 Corpus Christi, TX _78403 ,

J. C. Lanier/M. B. Lee J City of Austin

. Electric Utility Department 1

- P.O. Box 1088 j Austin, TX 78767 ,

R.~J. Costello/M. T. Hardt.

City Public Service Board P. O. Box 1771 -

San Antonio, TX' 78296  !

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1 Revised 10/08/90 L4/NRC/

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter ) )

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

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South Texas Project )

Units 1 and 2 )  ;

AFFIDAVIT Warren 11. Kinsey, Jr. being duly sworn, he.eby deposes and says that he ,

is Vice President, Nuclear Generation, of Hous'.on Lighting & Power Company; that he is duly authorized to si n 6and file witn the Nucicar Regulatory Commission the attached proposed changas to the South Texas Project Electric Generating Station Technical Specification Surveillance 4.4.6.2.2d; is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge and belief. ,

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Warren H. Kinsey, Jr. [ )'

Vice President, Nuclear GeYeration Subscribed and sworn to before ine, a Notary Public in and for The State of Texas this 15'6day of Noven11Ar .1990-l ,

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I ATTACHMENT 1 SIGNIFICANT HAZARDS EVALUATION FOR A PROPOSED CHANGE TO TECHNICAL SPECIFICATION SURVEILLANCE 4.4.6.2.2d I

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Attachment 1 ST.}{L. AE 3627 Page 1 of 4 SIGNIFICANT llAZARDS EVALUATION FOR A PROPOSED CilANGE TO TECilNICAL SPECIFICATION SURVEILLANCE 4.4.6.2.2d Backcround As stated in the bases for Technical Specification 3/4.4.6, the purpose of leak testing the RCS pressure isolation valves (PIV) is to prevent overpressurization and rupture of the Emergency Core Cooling System low pressure piping which could result in a 10CA that bypasses containment. This sequence of events is identified and discussed in WASil.1400 and is referred to as the "V" sequence, which is discussed as background information used in the development of Standard Technical Specifications It should be noted that the "V" sequence is not a credible event for STPEGS.

Because of the STPEGS design, lilAP is 9roposing to delete the requirement to leak check the PIV within 24 hoars after there has been flow through them. This test requirement imposes s.n unnecessary burden on the operators during restarts of the plant at a time when there are other, more c.ritical evolutions. PIV operability is adequately assured by the other PIV serveillance requirements.

Proposed Change Delete Technical Specification Surveillance 4.4.6.2.2d  ;

Saferv Evaluation The interfacing systems Loss.of Coolant Accident (LOCA) as originally identified and discussed in WASil 1400, the "V"_ sequence, is caused by a failure of the pressure boundary between the reactor coolant system (RCS) aad a connecting low pressure system outside of containment, such as the residual heat removal (RilR) system. The reason for the interest in this sequence is s that the resulting failure has the potential for bypassing the containment building thus leading to a release directly to the environment with little or i no mitigation. This sequence was first identified in'WASit-1400 and was based upon a two train plant which is different from the design of STPEGS. The following is a discussion of WASil 1400:

1. Cross failure or leakage in the two in series, low pressure injection check valves which provide che high pressure to low pressure boundary between the Ri!R/ Low llead Safety injection (LilSI) system and the primary system.

2. Because the RllR system is located outside containment in the WASil-1400 refere..co plant, the boundary failure results in *

pressurization of the RilR system outside containment. Since l system design pressure is only 600 psig, rupture was postulated' j to occur in the RllR piping.

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Attachment 1 ST.HL AE 3627 Page 2 of 4 )

l Safety Evaluation (Cont'd) l

3. The RilR system pumps and piping provide the low pressure safety l

injection capability at the reference plant, thus this failure 1eads to a 1DCA and fails the system that is needed to mitigate j the lhCA. Core damage was assumed to occur with no mitigation possible. J The most likely interfacing systems thCA identified in other utility ,

PRAs and in the WASil 1400 PRA is a lhCA that bypasses containment via ,

the RllR system. The RHR system at STPEGS is located entirely within ,

containment. Failure in the low pressure RHR lines connected to the )

RCS is a LOCA which is bounded by the STPEGS accident analyses.

Interfacing systems lhCA in an RHR train may effectively disable one

lowhead safety injection train but will not result in a lhCA outside containment which is the concern for an interfacing systems LOCA.

o Comparing the STPEGS Design to the System Analyzed in WASil 1400 shows the following significant differences and advantages:

l- 1. The Residual lleat Removal (RilR) system at STPEGS is completely contained inside the containment building. The WAS}l 1400 system was outside of the containment building.

2. There are three completely separate ECCS trains at STPEGS. Any failure in a single train does not affect the other two trains.

3. The RilR pumps are niet used at STPEGS to provide the low pressure safety injection ft.netion. Three, physically separate, low head safety injection p. imps in the fuel handling building provide this function.

l 4. Although the Low I ead Safety Injection (UISI) system interfaces with the RHR systes on a train basis, there is an additional check. valve inside containment between the UlSI pump and the RilR system tie. Consequently, any intersystem LOCA would require not two, but three series failures.

l l 5. An interfacing system LOCA was screened from the STPEGS l'

Probabilistic Safety Assessment (PSA) for the following reasons:

UlSI Cold Legs 3 normally closed,. leak tested, check valves l

inside containment. The RHR heat exchanger (the most likely l failure location in the Seabrook interfacing systems IECA analysis) is located inside containment downstream of thr. first L check valve inside containment.

ll L UlSI ilot Legs - 3 normally closed, leak tested, check valves inside containment and one normally closed MOV inside containment. The RHR heat exchanger is located inside

, containment downstream of the first check valve inside l containment.

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Attachment 1 ST HL AE 3627 Page 3 of 4 Safety Evaluation (Cont'd)

HHSI Cold Legs 3 normally closed, leak tested check valves and '

pump discharre piping that is designed for high pressure.

HHSI Hot Legs 3 normally closed, Icak tested, check valves, one i normally closed MOV, and pump discharge piping that is designed for high pressure, o Probability of a RHR/LHSI/HHSI Overpresa arization and Mitigation of ,

this event.

For STPEGS if a failure were to occur that leads to -

overpressurization of a singlo tralr. of RHR, one train of LilSI could ,

be affected and a LOCA could re., ult inside containment, but two trains of LHSI and RHR remain available to mitigate the effects of the LOCA. This accident is within the bounds of the large break LOC /.  !

analyzed in Chapter 15 of the STPEGS UFSAR. The likelihood of this event occurring is less than 1/2 of. It of the frequency o', occurren o of the design basis large break LOCA based upon the initiating frequency presented in the South Texas Project Probabili1 tic Safe',y Assessment (2.03E.04). This result does not take credi', for the i third. check valve which would further irsprove the above compari,on. -

To summarize, in the case of STPEGS, an RER interfacing systems LOCA, _

should it occur, could result in a release of coolant to containment  :

but not bypassing containment, thus preventing the direct release of i radioactive material to the environment. It is for this reason that the "V" sequence is not a credible event at LiPEGS. The availability of UlSI will still be assured by two-indep;. dent unaffected trains.

o Summary The STPEGS design includes thre6 check valves. If one check valve were to fail, two check valves would still be available to prevent ,

the bypassing of containment.

Since the "V" sequence was used in the development of Standard

• Technical Specifications and an interfacing system LOCA that bypasses '

containment is not s credible event for the STPECS, it is requested that the subject survefilance be deleted. .

The Beaver Valley design also has the RHR system completely inside .

containment and the required surveillances on P1V at Beaver Valley are similar to the proposed change.

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Attachment 1 ST HL AE 3627 Page 4 of 4 Determination of Sienificant Hazards Pursuant to 10CFR$0.91 this analysis provides a determination that the proposed change to Technical Specifications does not involve a significant hazards consideration as defined in 10CFR$0.92,

1. The proposed change does not involve a significant increase in the probability of an accident previously evaluated. The previously evaluated accident is a large break LOCA and this proposal does not change the probability of the dominant large break LOCA. The likelihood of an intersystem LOCA is less than 1/2 of it of the frequency of occurrence of the design basis large break LOCA without taking credit for the third check valve.

The probability of a large break LOCA as stated in the STPEGS PSA is 2.03E 04.

The consequences of the accident previously evaluated do not significantly increase. The consequences of a large break LOCA remain unchanged. A LOCA inside containment is bounded by the Updated Final Safety Analysis Report Chapter 15 analysis.

2. The proposed change does not create the possibility of a new or different accident from any previously evaluated. There are no

. changes to the design or operation of STPECS. Therefore, a new or t.ifferent accident from the previously evaluated LOCA is not created.

3. Tie proposed change does not involve a significant reduction in t'io margin of safety. The margin of safety is defined by the design of the systems connected to the Reactor Coolant. System through the subject pressure isolation valves. Since the design of these systems is unchanged, there is no reduction in tie

'aargin of safety. The Table 3.4 1 pressure isolation valves will be leak tested (1) once every 18 months, (2) prior to Mode 2 if l 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 /> and if leakage testing has not been performed in the previous 9 months, (3) following maintenance, repair or repiccoment, (4) as outlined in ASME Code,Section XI, paragraph IWV 3427(b). This testing ensures there is not a significant reduction in the margin of i safety.

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