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ACCESSION NBR:8902280353 DOC.DATE: 89/02/22 NOTARIZED: NO FACIL:50-305 Kewaunee Nuclear Power Plant, Wisconsin Public Servic AUTH.NAME AUTHOR AFFILIATION

WEBB,T.J. Wisconsin Public Service Corp. BERNHOFT,S.L. Wisconsin Public Service Corp. STEINHARDT,C.R. Wisconsin Public Service Corp. RECIP.NAME RECIPIENT AFFILIATION

SUBJECT: LER 89-001-00:on 890202,three items of generic interest identified by Kewaunee self initiated SSFI & ICSS.

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NOTES:

DOCKET # 05000305

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EXTERNAL:

RECIPIENT ID CODE/NAME

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ACRS MICHELSON ACRS WYLIE AEOD/DSP/TPAB ARM/DCTS/DAB NRR/DEST/ADE 8H NRR/DEST/CEB 8H NRR/DEST/ICSB 7 NRR/DEST/MTB 9H NRR/DEST/RSB 8E NRR/DLPQ/HFB 10 NRR/DOEA/EAB 11 NRR/DREP/RPB 10 NUDOCS-ABSTRACT RES/DSIR/EIB RGN3 FILE 01

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SYSTEM

45 ENCL 44

On February 2, 1989, with the plant at full power, the results of the inspection of the internal containment spray (ICS) were presented to Wisconsin Public Service Corporation (WPSC) Management. The inspection was the first inspection conducted by Kewaunee's internal Safety System Functional Inspection (SSFI) Group. The SSFI identified 38 items of potential concern. These items were reviewed for reportability. None of the items were reportable under the requirements of the Code of Federal Regulations. However, 3 of the findings were determined to be of potential generic interest. These three items are: 1) the single active failure of either ICS pump suction check valve could decrease containment recirculation inventory and increase the refueling water storage tank post accident source term; 2) the large break loss of coolant accident is not the most limiting accident for the ICS system's containment penetrations; 3) the containment analysis for a large break LOCA was re-evaluated because the basis for some of the initial conditions in the original analysis could not be found.

For all three findings, it was determined that the Kewaunee plant was within its design basis and that there were no significant hazards associated with the findings. For cases 1 and 2 above, the ICS system will be modified in order to increase the existing margin of safety. The proposed modifications include installation of additional containment isolation valves and additional isolation capability during containment sump recirculation.

The generic implication of these findings will be evaluated as part of WPSC's continuing SSFI effort.

8902280353 890222 19C3,n PDR ADOCK 05000305

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"C FW. M U.S NUCLEAR REQULATOR, COmilBoDe

APPROVED 0Mm NO 3180-4104

LICENSEE EVENT REPORT (LER) Expois SIgI

FACILITY NAME II1 DOCKET NUARIB 121

Kewaunee Nuclear Power Plant 0 15 o o l o3 0 5 1 OF1 9 TITZEA4, Three Items of Generic Interest Identified by Kewaunee's Self Initiated Safety

System Functional Inspection of the Internal Containment Spray System EVENT DAT (5) LER NUMBER () REPORT DATE 17) OTHER FACILITIES INVOLVED (I

MONTH DAY YEAR YEAR OUNIAL ONY. DAY YEAR OACIIjTY NAMEE DOCKET NUMBERISI

012 0 12 819 819- 10 1- O0 0 12 212 8 NA 0 5000

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LICENSEE CONTACT FOR THIS LE 111 NAME Thomas J. Webb - Plant Contact TELEPHONE NUMBER

Sherry L. Bernhoft - Corporate Contact AREACDD4 3 8 8 - 2 5 6 0 11I 4 3131- 11141116

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRISED IN THIS REPORT (13

CAUSE SYSTEM COMPONENT MANUFAC tEPORT LE CAUSE SYSTEM COMPONENT MAN AC EPORTADSL

SUPPLEMENTAL REPORT EXPECTED (1A, EXPECTED MONTH DAY YAR

SUBMISSION YES Of va, complet EXPEcTED Sustissi0N DAT NO DATE 1151 NA

ABSTRACT firnr te 140 sZa . is, aw e.imately 0ittee wsgle space typew-rte. fnm/ 1181

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FACILITY ONAVA (1) DOCKET NUMiwa (2) LIN 1~m si PACE Is LEA R EUMERNTIAL PAVG(ON

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Description of the Safety System Function Inspection (SSFI) Program Findings

On February 2, 1989, with the plant at full power, the inspection results of the internal containment spray (ICS) [BE] system were made available to Wisconsin Public Service Corporation (WPSC) management. The.inspection was conducted by the Safety System Functional Inspection (SSFI) Group. The SSFI program is a new self-initiated endeavor used to assess the operational readiness of the plant's safety systems. The program is tailored after the methodology used by the NRC's augmented inspection teams. The SSFI of the Internal Containment Spray (ICS) system was the first such inspection performed by WPSC. To aid your understanding of this report, a description of the ICS system is provided in attachment 1 of this report.

The SSFI report on the ICS system delineated the inspection teams findings. These findings are documented as Requests for Information (RIs). RIs identify potential areas of concern which require further assessment. After they are initiated, the RIs are assigned to the implementation team for further investigation and resolution. As a result of the ICS system inspection, a total of 38 RIs were generated.

Most of the RIs were originated due to either a lack of retrievable design basis information, or weaknesses in existing plant procedures and drawings. In some cases a lack of retrievable documentation hindered the inspection team's ability to assess the design adequacy of the ICS system. To resolve these RIs, the implementation team is currently pursuing alternate methods to obtain the information. If it is not possible to readily locate the information, supporting calculations and re-analysis are being performed. In the cases of plant procedure and drawing weaknesses, the implementation team is recommending appropriate revisions.

An initial review of the 38 RI's was performed to determine reportability. Based on the preliminary information available, none of the identified areas were considered reportable under the requirements of the Code of Federal Regulations. As the RIs continue to be investigated the implementation team leader will initiate the appropriate actions. If additional information calls into question the system's ability to satisfy its design basis, the finding will be re-evaluated for report-ability. This will ensure proper NRC notification.

The generic implication of the 38 RIs will be evaluated as part of WPSC's continuing SSFI effort.

NAIC FORtM 366A' 1983)

93U NUCLSA. ftEQULATQAY cO LICENSO VENT REPORT (LER) TEXT CONTIN*ION 4PPROVED 0O NO 310-4104

EXPIRES 6131I PACILITY gAILI III DOCKET NUMBER (21 L*N UMIN S$ paG. w

LEAR SOUMSI4IAL PA:: 3

Kewaunee Nuclear Power Plant ""IR Xti2 : NM Wt

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The purpose of this LER is to elaborate on three findings that may be of a generic interest to other licensees. The three findings are:

1. the single active failure of check valves EV] ICS-3A or ICS-3B could result in decreased containment recirculation inventory and increased refueling water storage tank (RWST) [TK] source term,

2. the review of the design of the ICS system's containment penetrations shows that the large break loss of coolant accident (LOCA) may not be the most limiting design basis scenario for these penetrations [PEN], and

3. the containment pressure response for a large break LOCA was re-analyzed because the basis for some of the assumptions used in the original analysis could not be located.

The three findings are described in the remainder of this report. Each finding is addressed separately with sections covering:

1. the Description of the Finding,

2. the Cause of the Event,

3. an Analysis of the Event,

4. Corrective Actions, and

5. Additional Information.

'Equipment Failures

*Similar Events

*Reference

NPC FORM 3O6A 19831

LICENS VENT REPORT (LER) TEXT CONTIN S ION us APEDN EouLAroe cE. ....00

EXPIRES, 8/3146 FACILITY NAME 10 DOCKET NUMaER 121 L INS ~ A IM PA " I

LEM NUM )AL V4A1(

Kewaunee Nuclear Power Plant NEAR . assu.

o 5 o o o 3 05 89 -0 0 1-0 00 4OF 1 9 TIMT (0 -9&~w . ~w adbn I NRC AcWM JS 'a) (17)1F

Active Failure Analysis of Check Valves ICS-3A and ICS-3B

Description of the Finding

While performing an operability assessment of the ICS system, the inspection team engineers questioned the ability of the pump [P] suction check valves, ICS-3A and ICS-3B, to isolate the RWST during containment sump recirculation when the ICS system is operating. Specifically, when recirculation flow is initiated by opening RHR-400A or RHR-400B, the check valves are the only barriers relied upon to prevent the back flow of containment sump water into the RWST. A single active failure of either check valve to prevent reverse flow could deplete sump inventory and increase the RWST source term.

Cause of Event

A review of the design documentation was performed to locate the justification for this configuration. This involved a review of early correspondence between WPSC and the ICS system designer (Pioneer Service & Engineer Co., now Fluor Daniel), a review of the original Final Safety Analysis Report (FSAR), and discussions with personnel involved with the plant design.

The relevant design basis criterion is Atomic Energy Commission (AEC) criterion 41 "Engineered Safety Features Performance Capability. This criterion states:

Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.

WPSC's response provided in the original FSAR states in part:

"Sufficient redundancy and duplication is incorporated into the design of the engineered safety features to insure that they may perform their function adequately even with the loss of a single active component."

The Updated Safety Analysis Report (USAR) does provide a single failure analysis of the ICS system. However, the failure of the check valves ICS-3A and ICS-3B to prevent reverse flow was not postulated. Based on a review of the available documentation and from conversations with personnel involved with the plant's original design, it appears that during the design phase of the ICS system, the suction check valves may not have been considered active components. Hence, it would not have been necessary to postulate a single active failure of either check valve to close or fully seat in order to prevent reverse flow.

IN"- FORM 366A 19 93 ,

NR. FC 3A19431 U.S1. MUCLEAR REQULATORY11 COMaleSI1400 LICENS* VENT REPORT (LER) TEXT CONTINS ION APPOVEDO POO. 3180-04o

EXPIRES, 8/31ig FACWLIT NAM 1 (1; DOCKET NUMBift 12) L1f1 NI lft le PAGE cm

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It appears the original designers may have assumed that the ability to accommodate a single active failure during recirculation is provided by the two redundant trains of ICS and the two redundant trains of containment fan coils units.

Analysis of Event

WPSC is reporting this event because of its potential for generic interest. This report describes how the active failure of a single check valve could decrease containment sump inventory and increase RWST source term post accident.

Although the single active failure of either ICS suction check valve could result in decreased containment sump recirculation inventory and increased RWST source term, the system is within its design basis. It appears that at the time Kewaunee was licensed, these check valves may not have been considered active components. Hence, they were not included in the active failure analysis.

The safety implications of this finding are negligible. During the recirculation phase of a large break LOCA, 2 of the 4 containment fan coil units supply sufficient cooling to maintain containment pressure below 46 psig, the containment's design basis pressure, with post accident decay heat removal operating. Therefore, ICS is not required during recirculation to satisfy design basis assumptions. If the operators elect to initiate containment spray during containment sump recirculation and one of the check valves fails to seat, the RWST liquid level would increase.. The operators would be able to diagnose the problem and isolate the effected train. Since all water would be contained in the RWST, it could be re-injected into containment in order to supply additional recirculation inventory and to decrease RWST source term.

Corrective Actions

Although Kewaunee is within its design basis for single active failure analysis, the industry's experience with check valve failures and current engineering practices suggest that additional isolation capability should be added to the system. Therefore, the SSFI implementation team has recommended that motor operators be installed on the manual isolation valves, ICS-2A and ICS-2B (please see Figure 2). They have also recommended that the emergency operating procedure for long term recirculation be revised to require closing valves ICS-2A/2B prior to opening RHR-400A/400B. These recommendations are presently under management review and are subject to change based on equipment availability and further analysis.

Additional Information

Equipment Failures: None

Similar Events: None

NMC FOftM 36A 19 831

NItC For 3S6A t9431 LICENSEI ENT REPORT (LER) TEXT CONTNUISON

A.MNvEA Oe ono Co~oio, EXPIRES: 8/3t1tt

FACILITY NAME (1) DOCKT NUMBER 12) LE M 11 PA

Kewaunee Nuclear Power Plant "^" EDI

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Design Review of the ICS System's Containment Penetrations

Description of the Event

As part of the SSFI of the ICS system, the ICS system's containment isolation capabilities were investigated. The investigation involved comparing actual ICS containment isolation capabilities to the requirements and guidance provided by the following documents.

1. Kewaunee's Updated Safety Analysis Report (USAR),

2. Kewaunee's 10 CFR 50 Appendix J commitments,

3. NUREG 0800, Standard Review Plant (SRP), and

4. 10 CFR 50 App. A, General Design Criteria (GDC).

Presently, containment isolation of the ICS system is provided by the following:

1. check valves ICS-SA and ICS-8B,

2. the leak tightness of the ICS system outside of containment, and

3. the discharge pressure of the ICS pumps which is greater than containment design pressure.

These barriers are consistent with KNPP's USAR's classification of the ICS system's penetrations as class 6 penetrations. The USAR defines class 6 penetrations as "penetrations for systems required to operate in a postaccident condition. The design and operation of isolation valves for this class of penetration is governed by the functional requirements of the system." During a large break LOCA, all three barriers would be inservice.

However, in the event of a small or intermediate break LOCA, containment pressure may not reach the ICS system's actuation setpoint of 23 psig. Therefore during these accidents, the third barrier (ICS pump discharge pressure) will not actuate automatically.

Cause of Event

This event does not describe equipment or personnel failures. It is being provided to the NRC as an item of potential generic interest. The present containment isolation capabilities of the ICS system are adequate and satisfy Kewaunee's design basis. Therefore this section of the Licensee Event Reporting system does not apply to this incident.

NMC FORM 36SA 19-83 )

k9C43 U.S. NUL AM RE ULATORY COP' hSg LICENSEWENT REPORT (LER) TEXT CONTINLON APPROVED OMG 000 31C04104 EXPIRES'lS/2'NS

FACILITY NIAMI II DOCKET NUMBER () LIR NUER S PAM IS

Kewaunee Nuclear Power Plant "L Mg"

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Analysis of the Event

Figure 1, attached, shows the ICS system's existing design. Containment isolation of the system is provided by the following:

1. Check valves ICS-8A and ICS-8B,

2. The leak tightness of the ICS system outside of containment, and

3. The discharge pressure of the ICS pumps.

The as built configuration of the ICS system's containment penetrations match the description of the penetrations contained in Kewaunee's USAR. Therefore, the penetrations meet their design basis. NRC approval of Kewaunee's method of testing these penetrations is provided by the reference identified at the end of this section.

The operability of check valves ICS-8A and ICS-8B is verified by surveillance procedure 23-201 (SP 23-201), "Inspection of Check Valves ICS-8A and ICS-8B". The inspection, which is conducted once every five years, requires that the valves be removed from the pipe and visually inspected for the following:

1. Internal wear,

2. Pin wear,

3. Sprinq condition,

4. Seat leakage,

5. Freedom of disc movement.

All problems are recorded and corrected.

The leak tightness of the system outside of containment is verified on an annual basis by SP 23-193, "Containment Spray System Leakage Test". In this test, valves ICS-7A and ICS-7B, see Figure 1, are closed and the pumps are started. The pump pressurizes the system to approximately 270 psig. The flow from the pumps is discharged into the RWST through the pump miniflow recirculation line. Once flow has been established and the system has been pressurized, the piping and the valves are visually inspected for excessive leakage. System leakage would then be detected and corrective actions would be taken to reestablish leak tightness. To date, no excessive leakage has been noted.

The leak tightness of the system outside of containment is also verified on a ten year basis by SP 23-228, "ICS Suction Piping Hydrostatic Test", and SP 23-229, ICS "Discharge Piping Hydrostatic test". These SPs pressurize the ICS system to ASME code requirements of 125% of design pressure. For both tests the piping and valves are inspected for signs of system degradation. Abnormal indications are recorded and resolved in accordance with the ASME code.

P~C FORM 308'-PPM is 83,

LICENSE VENT REPORT (LER) TEXT CONTINS ION A.S MJCELIA 000. onM-0 9EPVIES: 8/31 86

FACILITY A E fl DOCKET NUMBER (21 11 NUM 1s PA E I

Kewaunee Nuclear Power Plant "E" "MOU*N*L :v..o

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Although the third barrier, the discharge pressure of the ICS pumps, would be available during small and intermediate break LOCAs, it is not assumed to perform a containment isolation function since it is not expected to automatically actuate. The report from the Franklin Research Center attached to the reference identified the potential for losing this barrier.

The safety significance of this finding is minimal. Two of the three barriers would be available for all design basis accidents. Furthermore, the source term associated with small and intermediate break LOCAs is significantly less than those for a large break LOCA. Additionally, if the ICS pumps are not started, the discharge valves, ICS-6A/6B and ICS-5A/5B, will remain closed and act as a third barrier.

This event is being reported because of its generic interest. It provides an example where the large break LOCA design considerations are not as limiting as those for small and intermediate break LOCAs.

Corrective Actions

Although the penetrations meet their design basis, the SSFI Implementation Team has recommended that they be reclassified as class 3 penetrations This will increase the margin of safety associated with small and intermediate break LOCAs. A class 3 penetration is defined by Kewaunee's USAR as "Incoming lines connected to open systems outside the Reactor Containment Vessel are provided with two check valves in series, one located inside and one outside the Reactor Containment Vessel. The internal check valve is located near the Reactor Containment Vessel Shell..."

In order to satisfy the containment isolation requirements specified for a class 3 penetration, the SSFI team has recommended that the system be modified. Figure 2, attached, shows the proposed change to the ICS system. A check valve and a manual isolation valve will be added to the discharge piping of each ICS train. In addition, vent lines will be added between the new check valves and the new manual isolation valves. This will allow pneumatic leak testing of all 4 check valves. Presently the modifications are undergoing WPSC Management review and are subject to change based on equipment availability and further analysis.

Additional Information

Equipment Failure: None

Similar Events: None

Reference: Letter from D.G. Eisenhut (NRC) to C.W. Giesler (WPSC), dated September 30, 1982.

NRC FOAM 3e8.A 19 82

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FACILITY NA*t III DOCKET NUMBER (2) LIN M AN IJ PA" t SIUNIL MV4I0N

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Re-Analysis of Containment Pressure Response

Description of the Event

The SSFI of the ICS system compared the actual design of the ICS system to the description of the system contained in the USAR. The comparison identified differences between the system's as built configuration and some of the assumptions used in the containment pressure response analysis for a large break LOCA, Sections 14.3 and 14-C of Kewaunee's USAR. As a result of the differences, the containment was re-analyzed for the most limiting RCS large break LOCA; i.e., a 3ft 2 RCS pump suction break.

The original USAR analysis was conducted by Westinghouse using the COCO code. The re-analysis was conducted by Kewaunee's architect and engineer, Fluor Daniel (formerly Pioneer Service and Engineering) using the CONTEMPT code. The reanalysis resulted in increases in the calculated peak containment pressures following a large break LOCA. However, the calculated peak pressures did not exceed the containments design pressure of 46 psig.

Cause of the Event

The initial conditions assumed in the original analysis may be justified by other conservatisms associated with the analysis. However, the basis for these assumptions could not be located. Therefore, the performance of the containment cooling system was re-analyzed using more conservative initial conditions.

Analysis of the Event

This event is being reported because of its potential for generic interest. It identifies possible non-conservative assumptions used in containment analyses for a large break LOCA.

The USAR analysis and the re-analysis determined that the most limiting LOCA was a 3ft 2 RCS pump suction break with a concurrent loss of offsite power. The reanalysis resulted in higher peak containment pressures. However, the calculated pressure did not exceed the containment design pressure of 46 psig. Therefore, this event does not represent a significant safety hazard.

Table 1 identifies the initial conditions used in the original analysis contained in the USAR and those used in the re-analysis. In all cases the values used for the ICS system in the re-analysis identified in Table 1 are either identical to or more conservative than those used in the original analysis.

In the re-analysis, the initial containment pressure was increased from 15 psia to 16.85 psia. The increase assumes an initial containment pressure of 2 psig as allowed by Kewaunee's Technical Specifications and .15 psi associated with instrument error.

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The original analysis assumed that the temperature of the cooling water to the containment fan coil units (FCUs) would be 660F. This is approximately the expected high temperature of Lake Michigan water; i.e., the cooling water, during the summer months. As part of the SSFI, the lake temperatures for a 30 month period (April 1986 to October 1988) were reviewed. The highest temperature recorded was 75.9*F. To be conservative, the re-analysis assumed a cooling water temperature of 850 F.

The actuation time for the ICS system is the time it takes the spray nozzles to reach full flow after accident initiation. The re-analysis determined that ICS system's actuation time could be conservatively estimated at 135 seconds. This assumption is based on the equipment start times and valve cycle times delineated in Table 2. The actuation time for the containment FCUs is the time it takes to get full flow through the FCUs after accident initiation. The reanalysis determined that the containment FCUs actuation time could be conservatively estimated at 84 seconds. This assumption is based on the equipment start times and cycle times delineated in Table 3. These actuation times are conservative because:

1. they do not assume any flow through the pumps until the pump reaches full speed, and

2. flow through the valves is not assumed until full flow through the valves is reached.

The basis for the original 60 second actuation time assumed in the USAR's analysis could not be found.

In addition to the structural heat sinks assumed in the USAR, the SSFI verified the existence of additional structural heat sinks. The structural heat sinks used in the original analysis and the additional structural heat sinks used in the new analysis are described in Tables 4 and 5 respectively.

-0A F0 - 36

LICENSE*ENT REPORT (LER) TEXT CONTINU N

FACILITY NIAM: M

Kewaunee Nuclear Power Plant

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The results of the new analysis are provided in the following table.

ACTIVE HEAT REMOVAL 1 SPRAY (SPRAY & FAN COILS) 2 FAN COILS 2 SPRAYS 4 FAN COILS Maximum peak 45.0 45.2 44.7 pressure, psig at 168 sec. at 168 sec. at 183 sec.

The original USAR analysis resulted in a calculated 43.6 psig.

peak containment pressure of

Although the re-analysis shows an increase in peak containment pressures, the pressures do not exceed the containment's design pressure. Due to the conservatisms associated with the re-analysis, the peak pressures identified in the table are considered upper bound values.

Three of the conservatisms associated with the re-analysis are:

1. based on inservice testing results, the actuation time for the ICS system could be reduced to 120 seconds,

2. based on containment pressure history, the initial containment pressure could be reduced to 16.35 psia,

3. based on Lake Michigan temperature history, the temperature of the cooling water could be reduced to 75'F.

If these conservatisms are removed from the analysis, it is estimated that approximately 2 psi can be subtracted from the peak pressures.

Corrective Actions

Since the re-analysis calculated that the peak containment pressure for the most limiting large break LOCA would not exceed the containment's design pressure of 46 psig, no corrective actions are planned. However, the USAR will be revised to incorporate this new information.

Additional Information

Equipment Failure: None.

Similar Events: None.

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Attachment 1

The internal containment spray (ICS) system sprays cool borated water into containment during and following a design basis accident. The design basis of the ICS system is to provide sufficient heat removal capability post-accident to maintain the containment pressure below the containment's design value of 46 psig at 268 0 F. The following equipment combinations satisfy the design basis heat removal requirements.

1. Both containment spray pumps,

2. All four containment Fan Coil Units .(FCUs) [FCU], or

3. One containment spray pump and two FCU's.

The containment spray system automatically actuates on a hi-hi containment pressure signal (23 psig). This signal starts the pumps, opens the discharge valves to the spray headers, and opens the caustic standpipe isolation valves. During the initial injection phase, the ICS pumps take suction from the refueling water storage tank (RWST) and caustic standpipe, and discharge into containment via the spray ring header.

When injection from the RWST has been terminated, water recirculated from the containment sump by the residual heat removal (RHR) [BP] pumps [PI can be provided to the suction of the ICS pumps. After the injection phase is completed, it is expected that spray flow will no longer be required for containment heat removal. In the recirculation mode two FCUs are sufficient to maintain containment pressure below 46 psig.

However, if containment pressure is observed to increase, recirculation spray flow may be initiated by the control room operators. Discharge flow from the RHR heat exchanger [HX] can be diverted to the suction of the ICS pumps by opening motor [MO] operated valves RHR-400A or RHR-400B. Spray flow can be terminated at any time by the operators from the control room.

NAC FOORM 36A 19831

11109143"A U.. MUCLEA RECULATQAV COM"sSSION LICENSEWENT REPORT (LER) TEXT CONTIN ON APPROVED OMA M0 310-0104 ExPIRES, $131 l5

FACILITY NA1E IlI DOCKET NOUMBER 121 LE 1UMBE1 li PAQ IM

AaI IEUENTIALT MV.ON

Kewaunee Nuclear Power Plant -mant[..I7 n lus oo in o o310 5 89 0 0 0 1 3o 19

TEXT (F M ow"c a nepus, un ecubo NRC Fom =14 a) (17

TABLE 1

INITIAL CONDITIONS

Original Analysis

Containment Pressure

Air Partial Pressure

Initial Containment Pressure

Instrument Error

Containment Temperature

Containment Free Volume

Concrete and Structural Steel Initial Temperature

RWST Water for Spray during Injection

Cooling Water Temperature for the Fan Coil Units

Actuation Time for the ICS System

Actuation Time for the Containment Fan Coil Units

15.00 psia

14.70 psia

.3 psi

0 psi

120OF

1.32 X 106 ft3

120OF

100aF, 1300 gpm/train

660F, 900 gpm/unit

60 seconds

60 seconds

Re-Analysis

16.85 psia

14.70 psia

2.0 psi

0.15 psia

120OF

1.32 X 106 ft 3

120OF

100*F, 1300 gpm/train

85-F, 900 gpm/unit

135 seconds

85 seconds

NOC FORM 366A 1983 ,

194cF 3 0 U. NUCLEAR MIEQULATDM kAY ~ o 43) LICENSE#ENT REPORT (LER) TEXT CONTINUoN

APPROVED OMSM NO. 3180-MO4

EXPIRES7 8/31as FACILITY NAMYE (1 DOCKET WeUMER 2 LIN WUMM 10 PA(2 13

vEAP StautNTIAL OgVeoCN Kewaunee Nuclear Power Plant

05 01003 015 8 9 001 0 10 1 4 oF1 9 TE (0 n a pawc d. &. toe NW Fwm 14 u 17)

TABLE 2

ICS SYSTEM INITIATION TIMES

FOR A LARGE BREAK LOCA CONCURRENT WITH A BLACKOUT

Events in Sequence Duration, sec. Accumulation, sec.

3ft 2 pump suction break concurrent with loss of offsite power. 0 0

SI signal on containment pressure at > 4 psig (Tech Spec Table TS 3.5-1). <1 0

Two emergency diesel generators at full speed 10 seconds after start on SI signal, and beginning to accept loads. 10 10

ICS pumps loaded or started after the diesel at full speed. 20 30

ICS pumps to full speed. 4 34

Full flow achieved through ICS valves 5A/6A & 5B/6B. Full flow is achieved 40 seconds after the ICS pumps reach full speed. Full flow occurs when the valve is half open. 40 74

Full sprays, RWST water to reach the last nozzle. 61 135

N04C FORM 364A IS9e31

hAC Fo. MSA

LICENSE ENT REPORT (LER) TEXT CONTINU N1943

US. SAUCLEAR RSOULATORY COMmthfthON

xPOVIE owl 30-mod EW*I14t3, -- I -

Kewaunee Nuclear Power Plant

TEXT fK wow a m 'esWe. - edfto NRC F gm B =4s) (171

TABLE 3

FAN COIL INITIATION TIMES

FOR A LARGE BREAK LOCA CONCURRENT WITH A BLACKOUT

Events in Sequence

3ft 2 pump suction break concurrent with loss of offsite power.

SI signal on containment pressure at > 4 psig.

Two emergency diesel generators beginning to accept load 10 seconds after start on SI signals.

1st Service Water Pump starts 25 seconds after the diesel at full speed (USAR T8.2-1B)

Containment Fan Coil Fan starts 30 seconds after the diesel at full speed. Also, return MOV starts opening (USAR T8.2-1B)

The return MOV stroke time (from fully close to fully open)

Duration, sec.

0

<1

10

25

30

45

Accumulation, sec.

0

0

10

35

40

85

N14C FOOAM 366A 19-_a3

FA IIY NA f fIll e e *snae .6

NRC Form 38"A ... WL. MGUA01 OM 943 U.S. NCEMRGLTR LICENSEW ENT REPORT (LER) TEXT CONTINU3ON APPROVIED OMs 00 31W-0106

ExPIRIS 8 /3146

FACILITY NAM I1l DOCKET NUMSI (2)

Kewaunee Nuclear Power Plant

TABLE 4

STRUCTURAL HEAT SINKS

Linings Materi alExposed Area

(Ft2)

Containment Cylinder

Containment Dome

Reactor Vessel Liner

Refueling Canal

Carbon Steel

Carbon Steel

Carbon Steel Concrete Backup

Stainless Steel Concrete Backup

Stainless Steel Concrete Backup

Steel Structures

The following items have been grouped according to the indicated thickness:

41,300

17,300

1,260 1,260

1,100 1,100

5,500 5,500

Supports Pressurizer Support ) Reactor Coolant Pump)

Supports Crane Crane Rail Seismic Restraints Hangers )

Handrails Grating Exposed Pipe Exposed Conduit and Cable Trays

Ductwork Accumulators

Carbon Steel

Carbon Steel Carbon Steel (None assumed for calculations)

Carbon Steel Carbon Steel Carbon Steel

IVI IS QUINT IAL F WEVSOOON

o 15soo 3 0 oIoIo13C 5 8 9 0 1 0i0 1 6 o 19

Thickness (In.)

1.5

0.75

0.25 12.00

0.1875 12.0

0.25 12.0

( 4,055 (16,925 (28,500

2,000 ( 500

1,695 12,400

2,000 18,000 2,200

0.356 0.5 0.75 1.5 2.0

0.145 0.09

0.1 0.07 1.44

N14C FOR 366A

- - P." a*--If-9.,,& ... ww N*mn~ MW17

04PC o.m 366A1943

FACILITY MAMI fl)

Kewaunee Nuclear Power Plant

'TrT 1 - - - -.- .. 'W~ - -w ~ ~ W*ai Ill)

U.S NUCLEAM REGULATORY COM~,ao..

APPROVED OMS NO 3100-4i0 ExPIURS Salit.

Concrete Floors

Total Thickness, Inches

4

6

8

10

12

15

>15

One Side Area, Ft2

(one side exposure)

1785

2156

464

2405

6592

880

6512

Concrete Walls

Total Thickness, Inches

1.0

1.167

1.667

1.75

2.0

>2.0

One Side Area, Ft2

(one side exposure)

238

846

175

144

1,842

Paint

Total Thickness, Inches

21,900

One Side Area, Ft2 (one side exposure)

195, 334

LICENSEE NT REPORT (LER) TEXT CONTNUON

.012

Ac 100Mu 366A

EXIE, 5~DOCKET 1UMER (21t

1510 01"10" 819 I10 O O o jslo g go 0 8|9 -0 0 1 - 0 0 1 7 oF

TABLE 5 STRUCTURAL HEAT SINKS

TUC (9 ma

%%C ]"* 2naA US NUCLIAR atGULAT0"COM"Smok

LICENSEO ENT REPORT ILER) TEXT CONTINU N **EOSO 0 D'D*

$ACiLITY NAMt i DOCatT .Uelf 1 LIa %uMBER Ill PAGI 3

4%~gtR J&.MgE*

un' Nut car Pow r Plntl 0 90 9 0 0 1 _ 0 0 1 8 or 1

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04021

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c 3US NUCLAm S uLATO ' COMM'SSOne

" UCENSEE ENT REPORT ILER) TEXT CONTINUAN a"mo 1 os .o a0e.

SACILITY NAME III DOCast enusU o III Las NUMBE lI PAGE 13.

,-.. SGtOut .T.&L qV Io: vet

1 *0 ~

Kewaunce Nuc ear Pow r PlanL o o o | o 5 8 9 - 0 0 1 - 0 0 1 9 F 1 9

TEXT IN awn aw"

I I

2 it ma3

i~1

,v A t

WPSC (414) 433-1598 TELEC~gg a43O 27

WISCONSIN PUBLIC SERVICE CORPORATION

600 No-th Adams * P.O. Box 19002 * Green Bay, WI 54307 9002

NRC-89-19

EASYLINK 62891993

February 22, 1989 10 CFR 50.73

U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Gentlemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Reportable Occurrence 89-001-00

The attached Licensee Event Report for reportable occurrence 89-001-00 is being submitted in accordance with the requirements of 10 CFR 50.73, "Licensee Event Report System."

Sincerely,

1#UCAAijJ " o-&km C. R. Steinhardt Manager - Nuclear Power

PIS/jms

Attach.

cc - INPO Records Center Mr. Robert Nelson US NRC, Region II

I