IR 05000266/1997002

Insp Repts 50-266/97-02 & 50-301/97-02 on 970113-0304.No Violations Noted.Major Areas Inspected:Preliminary Review of Portions of Operability Evaluation of CFCs & Tscr 192

ML20138D185
Person / Time
Site:	Point Beach
Issue date:	04/25/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20138D146	List: IR 05000266/1997002 ML20138D140
References
50-266-97-02, 50-266-97-2, 50-301-97-02, 50-301-97-2, NUDOCS 9705010055
Download: ML20138D185 (10)

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION 111

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Docket Nos.: 50-266, 50-301 '

License Nos.: DPR-24, DPR-27 i

Report No.: 50-266/97002(DRS), 50-301/97002(DRS)

Licensee: Wisconsin Electric Power Company, WEPCo l

Facility: Point Beach Nuclear Plant, Units 1 & 2 Location: WEPCo Corporate Engineering Offices 231 W. Michigan, P.O. Box 2046 Milwaukee, WI 53201-2046 Dates: January 13 through March 4,1997 i

l Inspectors: G. F. O'Dwyer, Reactor Engineer

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Approved by: M. A. Ring, Chief, Lead Engineers Branch l Division of Reactor Safety i

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9705010055 970425 PDR ADOCK 05000266 G PDR

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i EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 & 2 NRC Inspection Report 50 266/97002(DRS), 50-301/97002(DRS)

Tnis inspection report contains the findings and conclusions for a routine inspection conducted from January 13 through March 4,1997, where the inspector performed a preliminary review of portions of the Operability Evaluation of the. Containment Fan Coolers (CFCs) for Point _ Beach Nuclear Plant, Units 1 and 2, and Technical Specification Change Request (TSCR) 192. No violations were identifie i i

I Enaineerina

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The inspector concluded that the containment fan cooler operability evaluation -

referenced a condition report to evaluate the effect of late fan cooler starts. The condition report did not account for the throttled service water outlet valves. This issue is unresolved pending determination of the safety impact of not considering  ;

the effects of the decreased flow to the fan cooler l

The inspecter concluded that the licensee did not have calculations to support the Technical Specification bases allowance of an initial containment pressure of 6 pounds per square inch gauge. Preliminary discussions indicate this may have  ;

been standard practice at the time of Point Beach licensing. This issue is unresolved pending further review of the issue.

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Report Details

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l. Enaineerina El Conduct of Engineering E1.1 _ Containment Fan Cooler (CFC) Ooerability Evaluation (37550) Jnsoection Scooe

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The inspector evaluated the effect of the differing post accident containment temperature and pressure profiles on containment fan coil (CFC) operabilit '

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Documents reviewed included:

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Letter from Wisconsin Electric Power Corporation (WEPCo) to NRC, VPNPD-l 96-065, " Detailed Operability Evaluation of the Service Water System With ,

I Respect to Post-Accident Boiling in Containment Fan Coolers for Point Beach j Nuclear Plant, Units 1 and 2," Revision 0, September 9,1996, Revision 1, ,

) January 17,1997, and Revision 2, January 17,1997;  !

• Technical Specification Change Request (TSCR) 192, " Modifications to TS i

15.3.3 Emergency Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan Coolers and Containment Spray, TS 15.3.7 Auxiliary  ;

Electrical Systems and 15.5.2 Containment," September 30,1996, i Supplement 1, November 26,1996, and Supplement 3, February 13,1997;

WEPCo calculation WE 96-N-165, " Delay Times for ESF Equipment l Actuation for Fuel Upgrade Accident Analyses," Revision 0; i

WEPCo Final Safety Analysis Report (FSAR), updated through June 199 Observations and Findinas l

l NOTE: To facilitate understanding and resolution, the inspector provided written l questions to the licensee, and obtained written responses. These questions and I answers are incorporated as part of this inspection Report, in accordance with the requirements of NRC Inspection Manual Chapter 0620.

1 s The inspector noted that the Operability Evaluation assumed that the containment i temperature and pressure profiles, as presented in the June 1996 revision of the FSAR, were correct. However, in July 1996, the licensee had throttled the service

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water valves on the outlet of the CFCs, which, based upon e May 1996 Westinghouse analysis, extended the post-accident containment adverse environment. Additionally, the inspector noted that, in Supplement 1 of TSCR-192, i the licensee acknowledged that the new steam generators slightly increased the

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post-accident containment temperature and pressure profiles above those provided

, in the FSAR by 0.77 pounds per square inch (psi) and 2 degrees Fahrenheit ( F).

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Therefore, the inspector questioned the validity of the Operability Evaluation determination,if the plant were to resume power operation prior to approval of

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TSCR 192. The licensee and consultant engineers informed the inspector that the !

Operability Evaluation still provided reasonable assurance that the CFCs would perform their safety function even under the conditions predicted by TSCR 192 (see 1 question 1). During discussions, the licensee indicated that the FSAR profiles

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would be updated in accordance with 10 CFR 50.71(e) once NRC approved i TSCR 19 i Additionally, the ins wctor noted that the licensee, in calculation WE 96-N-165, identified that the B" Containment Spray (CS) pumps and the "B" and "D" CFCs l for both units might not reach full heat removal for 73.1 seconds and 7 seconds, respectively. The licensee prepared condition report CR 96-1486. To resolve the condition report, the licensee used an informal calculation to determine that the actuallate start times were 67 and 62 seconds and concluded that operabi!ity was not affected. This conclusion was based upon the FSAR l

temperature and pressure profiles and did not account for the throttled service water outlet valves. The inspector further identified that Revision 2 of the Operability Evaluation did not address the effect of late CFC fan starts, instead relying upon Action 2 of CR 96-1486 to perform the evaluation. In Supplement 3 1 of TSCR 192, the licensee stated that the containment integrity evaluation for TSCR 192 assumed the same delay times as determined in CR 96-1486. The inspector was concerned that formal calculations might not have been performed to support the TSCR. This item is unresolved pending inspector confirmation that the

, licensee did formal calculations and used correct CFC capacities to determine appropriate CFC and CS pump delay times (URI 50-266/301/97002-01(DRS)). Conclusions The inspector concluded that the containment fan cooler operability evaluation referenced a condition report to evaluate the effect of late fan cooler starts. The condition report did not account for the throttled service water outlet valve E1.2 initial Containment Pressure Assumotion Insoection Scoce During review of the above Operability Evaluation, the inspector identified an additional concern involving the assumed maximum containment pressure prior to a loss of coolant accident for the containment integrity evaluation. The inspector reviewed Technical Specification (TS) 15.3.6.8 and portions of FSAR, June 1996 revision, Section 14.3.4, " Containment Integrity Evaluation." Observations and Findinas The inspector observed that the Containment Integrity Evaluation for both the current FSAR and TSCR 192 assumed that the initial containment pressure was 15 psi absolute (psia). However, TS 15.3.6.B allowed the containment pressure to be as high as 3 pounds per square inch gauge (psig) or 17.7 psia during normal operation. Furthermore, the Safety injection setpoint was approximately 6 psig or 20.7 psia. The TS basis stated that the containment design pressure of 60 psig would not be exceeded even if the pressure in containment was 6 psig prior to a

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l l- loss of coolant accident. The licensee stated that there were no calculations to l' support the TS basis statement and that the statement was probably based on the l peak containment pressure being about 53 psig which allowed an additional 7 psig l before the containment design pressure was reached. Since the containment l pressure profile in TSCR 192 v.as nearly 54 psia, the inspector was con;erned that l the licensee's assumption of a 15 psia initial pressure, rather than 17.7 psia, was  ;

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non-conservative. The licensee stated that TSCR 192 was based on the  !

containment pressure and temperature profiles generated by assuming 15 psia but l

that use of 17.7 psia as an initial containment pressure would be considered if the l licensee requested a power uprate, t

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The inspector discussed this issue with NRR. Preliminary indications were that the particular conditions noted above may have been consistent with NRC practice at the time of initial licensing for Point Beach.

l The issue of the appropriate value for the initial containment pressure assumption l for the containment integrity evaluation will be tracked as an unresolved item l (URI 50-266/301/97002-02(DRS)) pending further inspector review with NRR.

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l The inspector concluded that the licensee did not have calculations to support the  !

original Technical Specification bases allowance of an initial containment pressure

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of 6 psig. This appeared to have been consistent with NRC practice for the time

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perio i 11. Manaaement Meetinas X1 Exit Meeting Summary

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The inspector presented the inspection results to members of licensee management and staff at the conclusion of the onsite inspection on January 23,1997, and during a re-exit interview conducted by conference call on March 4,1997. The persons attending each of those meetings are designated in the list of persons contacted below.- The licensee acknowledged the findings presented, i The inspector asked the licensee whether any materials examined during the inspection

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should be considered proprietary. No proprietary information was identified.

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Attachment: As stated

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PARTIAL LIST OF PERSONS CONTACTED Licensee

- G. Adams, Senior Licensing Enginaw J. Anciaux, Senior Engineer K. Castell, Senior Engineer S. Dawson, Fauske engineer

1. H. Hanneman, former supervisor of Nuclear Safety Analysis Group R. Henry, Senior Vice President of Fauske and Associates, In P. Hubbard, Senior Engineer

1. ' J. McNamara, Manager, Design Engineering

1. • T. Malanowsky, Senior Project Engineer

S. Pellinen, Senior Engineer T. Zaki, Sargent and Lundy engineer U. S. Nuclear Reaulatorv Commission

1. M. Kunowski, Project Engineer i r A. McMurtray, Senior Resident inspector

1. M. Ring, Chief, Lead Engineers Branch

• Denotes individuals that attended the exit interview on January 23,199 # Denotes individuals that participated in the conference call re-exit on March 4,1997, l l l

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. i INSPECTION PROCEDURES USED

!P 37550: . Engineering ITEMS OPENED

'50 266/301/97002-01(DRS) URI CFC and CS pump Late Start Calculation 50-266/301/97002-02(DRS) URI Pre-LOCA Containment Pressure TIA LIST OF ACRONYMS USED l

CFC Containment Fan Cooler CR Condition Report

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CS Containment Spray l EDG Emergency Diesel Generator FAI Fauske and Associates, In FSAR Final Safety Analysis Report  :

IFl inspection Followup item 1 LOCA Loss of Coolant Accident i MOV Motor Operated Valve

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NRC United States Nuclear Regulatory Commission l

NRR Office of Nuclear Reactor Regulation i PBNP Point Beach Nuclear Plant psi pounds per square inch psia pounds per square inch absolute psig pounds per square inch gage S: Safety injection SW Service Water TS Technical Specifications TSCR Technical Specification Change Request l

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WEPCo Wisconsin Electric Power Company F degrees Fahrenheit i

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Attachment Q 1. For the startup of Point Beach Nuclear Plant (PBNP) Unit 2,is Wisconsin Electric Power Company (WEPCo) planning on relying on the containment fan cooler (CFC)

ovaluation sent to the NRC by letter VPNPD-96-065 dated September 9,1996, to provide reasonable assurance of the operability of the Unit 2 CFCs during the transient identified in that letter?

A 1. Yes. Any clarification or supplement to the 9/9/96 operability discussion will be submitted in the 120-day response to Generic Letter 96-0 Q For each curve included with Technical Specification Change Request (TSCR) 192 Supplement 1, please identify the analysis that generated each curv A All the curves were generated frorn data based on the results of analyses performed by Westinghouse. We have not obtained any information other than the results .!

presented at this time. It is expected that if the results of these new analyses are accepted by the NRC, appropriate Final Safety Analysis Report (FSAR) changes will l be generated. FSAR updating will be completed in accordance with 10 CFR 50.71(e) as stated in the answer to question 1 in TSCR 192 supplement Q Please specify the predicted length of time from the beginning of a. double-ended cold line break to when the following motor operated valves (MOVs) would begin to open: 1-MOV 2907,1-MOV-2908, 2-MOV-2907, and 2-MOV-290 l A Assuming a loss of offsite power:

1/2 second to safety injection (SI) signal initiated by containment Hi Pressure 2 seconds for SI Signal processing 10 seconds for emergency diesel generator (EDG) startup 12-1/2 seconds The time for the No Loss of Offsite Power case will not include the time delay for EDG startu The above is an informal estimate based on the following: 1/2 seconds is based on the time to a containment pressure of 9.5 psig; 2 seconds is a conservative estimate of signal processing times; and 10 seconds is the EDG start time acceptance lirni Q Please specify for both units the normal service water flow through each CFC during normal full power operation with the following valves shut: 1 -MOV-2907, 1-MOV 2908, 2-MOV-2907, 2-MOV-290 A The bypass throttle valves for the fan cooler MOVs (i.e., SW-144 (U1) and SW-273 (U2)) are set by Operations Checklists CL-10J U1&U2. The valves are throttled open to obtain 500-600 gpm with SW-2907/2908 shut and a service water (SW)

supply header pressure of approximately 60 psig. This checklist is run prior to leaving shutdown after refuelings. Normal flow will then vary depending on SW header pressure. SW header pressure is maintained at a pressure > 50 psig during

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normal operation. The flow is recorded each shift on PBF-2036, Safeguards Shift Log. The logsheet for January 16,1997 indicates flows of 690, 690, 680, and -

680 gpm for Unit 1 A, B, C, and D coolers respectively, and zero flow through the Unit 2 coolers (isolated). SW header pressure was 78-79 psig. The log spec is for flow > 500 gp . Please specify the latest recorded opening stroke times for the following MOVs:

l 1 MOV-2907,1-MOV-2908, 2-MOV-2907, 2-MOV-2908.

A5. Inservice Test (IT-72) results of December 14,1996:

l l Ooen Shut l 1-2907 28.36 s 27.62 s 2-2907 27.71 s 27.21 s 1-2908 28.08 s 26.25 s l 2-2908 27.61 s 26.14 s 06. What are the revisions to the S&L analysis that supports VPNPD-96-065 Attachment B? l

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A6. Description of those revisions is attached. (DIT-PB-EXT-222-01 and -02). '

l 07. Will the new Unit 2 containment temperature / pressure profile affect the operability i conclusion related to the restoration of single-phase flow to the CFCs and i satisfaction of the design basis CFC response requirement of 60 seconds?

l A7. Analysis of the new Unit 2 steam generators provides a new containment peak pressure that is 3/4 psi higher than the FSAR profile and 2 F higher than the FSAR profile. Based on the FSAR profile, the S&L report (Attachment B to VPNPD-96-065) concluded that adequate flow would be restored within 36 seconds. Based on the sensitivity of the S&L analysis,it'is our judgment that the i restoration of in the containment P/T profile, and in any case, the 60-second FSAR l

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response requirement would not be exceede :

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08. VPNPD-96-065 considered the large break loss of coolant accident (LOCA) to create the limiting conditions of the transient. Could there be a smaller break LOCA 1 l that creates more limiting results?

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A8. Fauske conducted scale model experiments of the PBNP configuration. To simulate i l a range of LOCAs, the experiments varied the steam injection rates and concluded I

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that the large break LOCA bounded other smaller LOCAs.

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09. Will the new Unit 2 containment temperature / pressure profile affect the operability conclusion related to the peak water hammer loads?

l A9. No. Our existing analysis shows that the peak pressures associated with

condensation-induced water hammer are bounded by the peak pressures associated

with the " refill" water hammer. The refill water hammer event provides the l maximum loads that we use in the limiting stress analysis. We believe that a more

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adverse containment environment (e.g. a 2 Fahrenheit and 3/4 psi increase) will l

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I even further reduce the probability for condensation-induced water hamme Therefore, the assumption that the refill water hammer event is bounding is still a valid conclusion with the higher containment profil With iespect to condensation induced water hammer, a highet pressure / temperature profile would result in a higher steam generation rate and, accordingly, a higher Froude number in the 8" return lines Based on the principles described in 1 FAl/96 75, the potential for condensation-induced water hammer is further reduced ;

as the Froude number increases. Therefore, a higher containment pressure / temperature profile would reduce condensation-induced water hammer load With respect to refill water hammer loads, a higher pressure / temperature profile would result in a larger void in the return lines. If the PBNP configuration were not already postulated to be subject to the maximum refill velocity, it could be postulated that the larger void could result in a higher refill velocity and a larger water hammer load. However, this scenario has been bounded by the conservative ,

assumptions that form the basis for our refill water hammer analysis. For  :

conservatism, the refill water hammer loads are already based on the steady-state l terminal velocity with 6 SW pumps running. This is the largest refill rate that is ;

physically possible in the PBNP SW System, regardless of void size. Also, for l l conservatism, the refill water hammer is assumed to occur at the throttle valve l

rather than at the middle of the return line. Therefore, no credit is taken for i compliant surface column rejoining. Based on the above, the magnitude of the i

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calculated refill water hammer loads will bound the loads that may be generated with the higher containment temperature / pressure profil Therefore, the results of VPNPD-96-065 are still valid for the revised containment I pressure / temperature profile and continue to provide the basis for Unit 1 and Unit 2 operability, i l l  !

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