IR 05000498/1996025

Insp Repts 50-498/96-25 & 50-499/96-25 on 961031-1206.No Violations Noted.Major Areas Inspected:Operations & Engineering

ML20149M476
Person / Time
Site:	South Texas
Issue date:	12/11/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20149M472	List: IR 05000498/1996025 ML20149M471
References
50-498-96-25, 50-499-96-25, NUDOCS 9612170112
Download: ML20149M476 (9)

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U.S. NUCLEAR REGULATORY COMMISSION I

REGION IV

l Docket Nos: 50-498;50-499 License Nos: NPF-76; NPF-80 Report No: 50-498/96-25:50-499/96-25 l

Licensee: Houston Lighting & Power (HL&P)

Facility: South Texas Project Electric Generating Station, Units 1 and 2  ;

i Location: 8 Miles West of Wadsworth on FM 521 l Wadsworth, Texas 77483 l

l Dates: October 31 through December 6,1996 '

Inspectors: D. P. Loveless, Senior Resident inspector i

J. M. Keeton, Resident inspector i W. C. Sifre, Resident inspector  ;

l Approved by: J. l. Tapia, Chief, Project Branch A l Division of Reactor Projects

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. EXECUTIVE SUMMARY South Texas Project, Units 1 & 2 NRC Inspection Report 50-498/96-25;50-499/96-25 This specialinspection included a review the circumstances surrounding the failure of plant personnel to identify that leakage from the emergency core cooling system (ECCS) into the fuel handling building (FHB) was significan Operations

• Licensed operators identified leakage from the ECCS into the FdB pump room sump and wrote a condition report describing the degraded conditio a, however, the condition was not adequately classified and as a result did net receive an adequate evaluation. (Section E2.1).

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• The system engineer was aware that the emergency core cooling system was degraded, however, no evaluation of the leakage was undertaken until the issue was raised by the resident inspectors during routine inspection (Section E2.1).

• The licensee's no significant hazards determination utilized a nonconservative containment leak rate assumption that was below the maximum allowable Leak Rate, L, defined in Technical Specifications (Section E2.1).

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! Reoort Details l s- i l

Summary of Plant Status a

Both units operated at essentially 100 percent reactor power throughout this inspection period.

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i E2 Engineering Support of Facilities and Equipment E Excess ECCS Leakaae Contribution to Loss of Coo' ant Accident (LOCA) Dose j Inspection Scope (37551)

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The inspectors reviewed the sequence of events leading to the identification and ,

repair of a leaking safety system in Unit 2. This review also included )

documentation and procedures related to the design basis of the ECCS and the  !

impact of the leakage on the radiation doses expected during a design basis LOC This issue was previously discussed as an unresolved item in NRC Inspection Report 50-489/96-07:50-499/96-0 Observations and Findinas On February 22,1996, operators identified that the ECCS was leaking into the FHB pump room sump. Condition Report 96-2271 was written to document the deficiency. The report stated that the leakage was suspected to be seat leakage through Pump Suction Relief Valve 2-PSV-3941. A water sample taken from the ECCS sump contained boron, indicating that the leakage was from the refueling water storage tank (RWST). Based on the historical pumping rates from the sump I and associated decreases in the RWST water level, the leakage had been estimated to be 300 gpd. This condition report was initially classified as only documenting a material deficiency and entered the licensee's work process program. No evaluation of the leakage impact on the ECCS or other systems was undertaken as a result of that classificatio During a routine inspection of the ECCS in the FHB on September 11,1996, inspectors noted leakage into the pump room sumps. Through discussions with the system engineer, the inspectors were made aware of the outstanding condition report. The system engineer informed the inspector that the relief valve was the suspected source of the leak and that it had been scheduled for repairs in June 199 The inspector noted that the estimated leakage from the valve was greater than the leak rate documented in Table 15.6-12 of the Updated Final Safety Analysis Report (UFSAR). UFSAR Section 15.6.5.3.2 describes the leakage from ECCS components located in the FHB as a potential source of fission product leakage

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following a LOCA. This leakage is postulated to occur during the recirculation phase for long-term core cooling and containment cooling by sprays. The water i contained in the containment sumps is used after the injection phase and is l recirculated by the ECCS pumps and the containment spray pumps located in the FHB. The maximum potential recirculation loop leakage external to containment and into the FHB which was utilized in the dose assessment for a postulated large-break LOCA was tabulated in UFSAR Table 15.6-12 as 4,140 cubic centimeters per hour, )

or about 26 gpd. The leakage rate assumed for dose calculation purposes was conservatively selected as twice the leakage rate given in the table. The estimated leak rate of 300 gallons per day (gpd) was therefore in excess of the leak rate assumed in the offsite dose assessments for all ECCS components in the FHB.

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10 CFR Part 50, Appendix A, Criterion 19, requires that control rooms be provided ,

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adequate radiation protection to permit access and occupancy under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole

body, or its equivalent to any part of the body, for the duration of the acciden The equivalent is 30 rem thyroid or beta-skin dose. As described in Section 6.4. of the UFSAR, the postulated control room dose analysis uses the analysis for

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offsite environmental consequences described in UFSAR Section 16.6.5. The UFSAR analysis assumes that the airborne release from the FHB is transported to the intake of the control room heating, ventilation, and air conditioning system by atmospheric conditions. The release pathway from the FHB is through the main exhaust. The release and control room pathways are shown in UFSAR

Figures 15.B-2 and 15.B-3. Based on the postulated UFSAR leakages, control room operator dose was calculated at 23.29 rem to the thyroid.

. On September 24, licensee engineers concluded that the relief valve leakage i

required additional assessment. On September 25, rough calculations indicated that relief valve leakage would not greatly affect offsite doses following a LOC ,

However, mechanical maintenance personnel determined that the relief valve was

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not leaking. A plan of action was then developed to identify the source and i quantify the leak. On October 2, engineers determined that the average leak rate over the previous 2 months had been 255 gpd. It was determined that if the leak l was coming from the high pressure side of the system, offsite LOCA doses could be a unacceptable. On October 3, engineers verified that the leak into the FHB pump

, room sump was coming from the high pressure side of the ECCS pump On October 4,1996, engineers determined that Safety injection System Flush Valve 2-SI-0120C was leaking past its seat and into the ECCS Train C pump room sump. Manual Valve 2-SI-0120Cis a single isolation valve on the discharge side of High Head Safety injection Pump 2C. The 2-inch valve isolates a flush line which discharges to the open ECCS sump in the FHB basement. With the pump shut down, the valve leak rate was found to be 172 gallons per day. This leakage was due to the head of water in the RWST leaking through the pump. The leak rate was determined to be 1641 gpd with the high head safety injection pump ' >erating in testing conditions. Engineers calculated that following a LOCA, the a ' ate would

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have been approximately 460 gpd. This was based on adding the maximum design system leakage stated in the UFSAR to the identified leak rate through Valve 2-SI-120C. Using the design assumptions listed in the UFSAR, engineers l calculated that, following a postulated LOCA, control room operators would have j received from 32 to 37 Rem to the thyroid, depending upon when the high head 4 safety injection pump was shut down. This exceeded the 10 CFR Part 50, Appendix A, General Design Criterion 19, limit. The leakage that was allowed to exist would have contributed an additional 14 Rem to control room operator dos A 1-hour, nonemergency notification was made in accordance with 10 CFR Section 50.72 (b)(1)(ii)(B), because the condition had been determined to be outside

the design basis of the plant. Licensed operators declared the train inoperable and l closed the associated containment sump isolation valve. Technical Specification Action Statement 3.5.2.a was entered, Valve 2-SI-0120C was cut out, replaced, and tested, and the train was returned to servic On October 10,1996, licensee engineers presented the inspectors with a report which concluded that no significant safety hazard existed. This revised analysis l concluded that the actual estimated dose to control room operators during a postulated LOCA was 21.4 Rem to the thyroid, within General Design Criterion 19 limits. However, changes to the initial evaluation included significantly decreasing the containment design leak rate. The subsequent analysis utilized a previously measured primary containment leakage rate and not the maximum allowable rate utilized in the accident analysis and referenced in the Technical Specifications. The ,

analysis was therefore considered nonconservative by the inspectors. The !

engineers' report entitled "ECCS Inleakage into FHB Sump Reportability" included the use of a maximum allowable primary containment leakage rate (L) of 0.0767 weight percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the leakage was reduced to half. These values were based on an integrated primary ;

containment leak rate test performed on September 22 and 23,1991. The test value was adjusted using current local leak rate test data. The maximum allowable leakage rate is delineated in Technical Specification 6.8.3.j as 0.3 percent of primary containment air weight per day. UFSAR Section 6.2.1.1.1.5 also documents 0.3 percent per day as the design basis for containment. The inspectors noted that the use of the measured value was not conservative for the following reasons:

(1) The leak rate was significantly below the design basis containment leak rate in Technical Specification (2) The assumption utilizes a containment leak rate that was measured 5 years ago. An evaluation of the local leak rate test data was not performed to evaluate the potential impact of test equipment inaccuracies and statistical error . . _ _ _ - _ _ _

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(3) If the measured leak rate was accurate and found acceptable for use as essentially a new L,, correct use of Appendix J would require that

! containment as-left leak rate be .75 L,. This would require that the analysis I be performed at a 33 percent greater leak rate than the measured valu The ECCS was designed to be, essentially, leak tight to provide its required safety function. The leakage that was allowed to exist for 7 months represented a significant degradation of the system safety function in that the leakage in excess of UFSAR values would have increased radioactive material bypassed to the

atmosphere. The system was degraded to the extent that detailed evaluation was

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required to determine the system operability. The condition was determined to be

adverse to quality from February 22,1996, when the system degradation was identified, until the condition was corrected on October 4,1996. During that time, several opportunities for recognition of significance and corrective action implementation were missed. Licensee personnel were not aware that operational leakage of the ECCS could affect Technical Specification system operability. The I failure to correct the excessive leakage represented a condition adverse to quality, which was not promptly identified and corrected. This represents an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI. In addition to failing to take prompt corrective action, the licensee did not evaluate the decision to delay repairing the leak until June 1997 in accordance with 10 CFR 50.59.

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The licensee's Event Review Team investigation report determined that the root cause of the failure to identify the significance of this condition was a " General knowledge deficiency to the sensitivity of the Control Room dose limits as they i relate to ECCS leakege outside containment." At least six reviewers of the  !

condition report, including licensed operators and the system engineer, failed to recognize the potential impact of system leakage. In addition, monthly system health reviews, as well as several work planning sessions, also failed to identify the significance of the deficiency. The inspectors concluded that this issue would not !

have been adequately addressed by the licensee had it not been identified during routine NRC inspection, c. Conclusions  ;

The Unit 2 ECCS was allowed to have excess leakage for a period of 7 month This represented a significant degradation of a safety system, because leakage in excess of UFSAR values would have increased radioactive material bypassed to the atmosphere during a LOCA and would have resulted in exceeding General Design Criterion 19 dose rate limits for the control room. The degraded system was initially evaluated by licensed operators and the system engineer; however, no one identified the significance of ECCS leakage outside the primary containment. The system was degraded to the extent that a detailed evaluation was required to determine system operability. The condition was determined to be adverse to quality from February 22,1996, when the system degradation was identified, until the condition was corrected on October 4,1996. During that time, several

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opportunities sar recognition of significance and corrective action implementation were missed. The licensee's subsequent no significant hazards determination utilized nonconservative containment leak rate assumptions and a leakage rate that was below the maximum allowable leakage rate defined in Technical Specification j d

The inspectors concluded that this issue would not have been adequately addressed

by the licensee had it not been identified during routine NRC inspection and that it represented apparent violations of 10 CFR Part 50, Appendix B, Criterion XVI, and 10 CFR 50.5 !

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ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee

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A. Aldridge, Supervisor, Engineering Specialist H. Butterworth, Operations Manager, Unit 2

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T. Cloninger, Vice President, Nuclear Engineering K. Coates, Manager, Maintenance, Unit 2 W. Cottle, Executive Vice-President D. Daniels, Manager, Operating Experience

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B. Dowdy, Assistant to Group Vice President J. Groth, Vice President, Nuclear Generation E. Halpin, Manager, Design Engineering Department W. Harrison, Supervising Licensing Engineer S. Head, Licensing Supervisor T. Jordan, Manager, Systems Engineering M. Kanavos, Manager, Mechanical Fluid Systems D. Leazar, Director, Nuclear Fuels and Analysis J. Lovell, Plant Operations Manager, Unit 1 B. Masse, Plant Manager, Unit 2 M. McBurnett, Licensing Manager G. Parkey, Plant Manager, Unit 1 D. Rencurrel, Manager, Electrical / Instrumentation and Controls J. Shepperd, Assistant to Executive Vice-President D. Schulker, Compliance Engineer S. Thomas, Manager, Design Engineering Department W. Waddell, Manager, Maintenance 1 G. Weldon, Manager, Nuclear Training Department INSPECTION PROCEDURES USED IP 37551: Onsite Engineering ITEMS OPENED. CLOSED. AND DISCUSSED Opened 498 499/96025-01 eel Failure to take corrective action to reduce ECCS leakage for a period of 7 months (Section E2.1)

Closed 499/96007-01 URI Review the circumstances surrounding the f ailure of plant personnel to identify that leakage from the emergency core cooling system into the fuel handling building was significant

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-2-LIST OF ACRONYMS USED

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ECCS emergency core cooling system eel escalated enforcement item (apparent violation)

FHB fuel handling building gpd gallons per day LOCA loss of coolant accident RWST refueling water storage tank UFSAR Updated Final Safety Analysis Report URI unresolved item i VIO violation -

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