ML20210P944

Safety Evaluation Accepting Licensee Assessment of Impact on Operation of Plant,Unit 1,with Crack Indications of 2.11, 6.36 & 1.74 Inches in Three Separate Jet Pump Risers

ML20210P944
Person / Time
Site:	Brunswick
Issue date:	08/10/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20210P941	List: ML20210P938 ML20210P944
References
NUDOCS 9908130111
Download: ML20210P944 (5)

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j. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO FACILITY OPERATING LICENSE NO. DPR-71 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 DOCKET NO. 50-325 e

1.0 INTRODUCTION

As requested by the Nuclear Regulatory Commission (NRC) staff, by letter dated May 29,1998 (Reference 1), as supplemented by letters dated February 5 and May 17,1999 (References 2 and 3), Carolina Power and Light Company (CP&L), the licensee, submitted their assessment of the impact on operation of Brunswick Unit 1 with crack indications of 2.11,6.36, and 1.74 inches in three separate jet pump risers. During the Brunswick Unit 1 in-vessel visual inspection (IWI) of the 10 jet pump risers, CP&L identified the crack indications along the heat affected zone of the riser elbow at the RS-1 weld of jet pump numbers 7/8 (riser D),13/14 (riser G), and 19/20 (riser K). CP&L stated that their assessment justified operation with this condition without repair for the next fuel cycle (24-month cycle).

2.0 ENGINEERING EVALUATION 2.1 Summary of Licensee Evaluation The licensee employed the limit load analysis consistent with the latest Appendix C (1996 Addenda) of Section XI of the American Society of Mecnanical Engineers (ASME) Code to perform the flaw evaluation. The limit load analysis assumed that the flaw was through-wall and used a safety factor of 2.77 for ncrmal and upset conditions and 1.39 for emergency and faulted conditions, a Z-factor of 1.0 for non-flux welds, and a flow stress of 3S (S. = 16.9 ksi) to arrive at an allowable flaw size of 18.9 inches for the limiting load condition. The normal load includes dead weight, hydraulic loads, fluid drag, flow-induced vibration (FIV), and thermal loads. The faulted load included the normal load plus the safe shutdown earthquake inertia.

To calculate the predicted flaw size at the end of one additional fuel cycle (16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />), the licensee used a bounding intergranular stress corrosion cracking (IGSCC) crack growth rate of SX10-5 inch / hour. This gives a crack growth of 0.8 inch per crack end per cycle. The licensee also considered the fatigue crack growth due to FlV under normal conditions and determined that the crack growth due to FIV is 0.356 inch. In this calculation, the licensee employed the ENCLOSURE 9908130111 990810 PDR ADOCK 05000325 P PDR

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2 l Appendix C fatigue crack growth rate curves and used the stress ranges and cycles at the cracked weld location ertracted from start =up vibration tasting data for the jst pump 6f l FitzPatrick (the prototype BWR/4 for Brunswick Unit 1). The threshold AK used by the licensee i b- was 5 ksi(in)*, which represents a limit below which the crack will not grow. Adding the IGSCC I crack growth of 1.6 inches, the FIV crack growth of 0.14 inch, and the nondestructive l examination (NDE) uncertainty of 0.61 inch to the limiting detected flaw size of 5.75 inches, the licensee calculated the final crack length to be 8.1 inches. Since the predicted crack length at the end of one additional fuel cycle is less than the allowable crack length, with an adequate margin, the licensee concluded that the observed flaws in the welds of jet pump risers are acceptable as-is until the end of one additional fuel cycle.

In the second submittal, the licensee estimated the leakage to be 47 gallons per minute for the l limiting crack and 89 gallons per minute for all three detected flaws using the Bemoulli equation for incompressible flow- l l

Q = CA)2 GAP lp where Q is the leakage, C is the flow loss coefficient, A is the crack area, p is the fluid density, i and AP is the pressure differential across the pipe. In this application, the licensee conservatively assumed that the flow loss coefficient was 1.0, the crack area was rectangular, and the crack opening displacement that was used in the crack area (A) calculation was 0.01 inch.

2.2 NRC Staff Evaluation 2.2.1 Flaw Evaluation The staff evaluated the licensee's flaw evaluation and determined that the limit load analysis meets the rules of the ASME Code (1996 Addenda), ard therefore is appropriate. The predicted crack length at the end of one cycle is also appropriate because the licensee used the bounding IGSCC crack growth rate and considered the FIV crack growth and NDE uncertainty. This FlV crack growth calculation methodology by General Electric Nuclear Energy has been reviewed and accepted by the staff as indicated in the Safety Evaluations (SEs) for Vermont Yankee, Dresden Unit 2, and Hatch Unit 1. The threshold aK of 5 ksi(in)" has also been accepted by the staff in these SEs.

2.2.2 Leakaae Evaluation Although the Bernoulli equation is a simple model for flow through a crack, the licensee used the equation very conservatively. First, using a flow loss coefficient of 1.0 is equivalent to assuming that there is no pressure loss due to phase change, area change, and friction loss associated with surface roughness. This would overestimate the leakage as indicated by the Bemoulli equation. Second, using a rectangular crack area is more conservative than using the elliptic crack area based on linear elastic fracture mechanics (LEFM). Since the predicted crack length is much smaller than the critical crack length based on limit load analysis, it is appropriate to use LEFM to predict the crack opening displacement. Using the larger rectangular crack area also

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i overestimated the leakage. Third, instead of using the crack tip displacement based on fracture mechanics anaiysis, the licensee coscrvatidy used an assumed value of 0.01 inch. This value is 20 times the crack tip displacement used by anotner hewnee fut M bcilbg WW reactor (BWR) plant based on LEFM analysis from EPRI Report NP 2472, Volume 2, " Growth ~ {

and Stability of Stress Corrosion Cracks in Large Diameter BWR Piping." Considering these l conservatisms, the staff has accepted the licensee's leakage calculation methodology and has venfied the calculated leakage rate.

3.0 SYSTEM EVALUATION Jet pump assemblies are not designed to meet the ASME code. However, they are classified as safety-related components since the structuralintegrity of the jet pump assembly is relied upon for assuring the abmty to reflood the core, up to two-thirds core height, following a design basis accident (DBA) los.s-of-coolant accident (LOCA) for BWR/3s through BWR/6s. An additional safety function of the jet pump assemblies at Brunswick Unit 1 is to provide a path for low pressure coolant injection (LPCI) flow into the core.

CP&L evaluated the effect of potential leakage through the crack indications at the end of the next fuel cycle. The evaluation assumed a through-wall crack at the maximum predicted length at the end of the operating cycle (EOC) for each crack indication. Additionally, the licensee i assumed a crack opening displacement of 0.01 inches for each indication. Based on these assumptions, the following leakage rates were calculated for each indication.

Riser Current Length (inches) Predicted Length at EOC Leakage (gallons per (inches) minute)

D 2.11 3.81 22 G 6.36 8.2 47 K 1.74 3.44 20 Using the predicted flaw size at the end of the next fuel cycle for each indication, the licensee calculated the total leakage from the indications to be 89 gallons per minute. Although this leakage is not significant with' regard to total recirculation flow, a reduction of core cooling capability due to the leakage must be considered.

CP&L considered the decrease in LPCI flow during the most limiting DBA with respect to peak cladding temperature (PCT). The bounding case for Brunswick Unit 1 is a recirculation suction line break with a failure of DC power. In this scenario, two LPCI pumps would inject into one recirculation loop. According to Table 6.3.1-1 of the Brunswick Unit 1 Updated Final Safety Analysis Report (UFSAR, Reference 4), the design LPCI injection rate for one pump operating is 9000 gallons per minute. The UFSAR also states that a flow rate of 14000 gallons per minute is assumed for two LPCI pumps injecting into one recirculation loop (UFSAR Table 6.3.3-5).

Following a LOCA, the pressure difference (AP) between the jet pump riser and the reactor downcomer region would be much smaller than the AP during normal operations. According to 1 UFSAR Table 6.3.1-1, APm would be 20 psid. Since the leakage rate is proportional to the square root of the pressure, CP&L calculated the potential leakage through the three indications during a LOCA using the following equation:

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Quri = (Leukagem.a)NkPuxV / APwim.a

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The resultant look tste from tus thrco indlc;tisns wa; 30 gs!'cas per minute. The Brunswick

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UFSAR states that the jet pumps were designed for a maximum potentialleakage of 807 gallons

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per minute. The staff has determined that the calculated leakage due to the three indications

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during a LOCA is fairly insignificant in comparison to the design leakage of the jet pumps and i the total flow of the two LPCI pumps. Additior. ally, the Brunswick SAFER /GESTR LOCA Analysis (Reference 5) concluded that the PCT is well below the regulation limit of 2200 degrees l Fahrenheit. The current results of the licensing basis PCT analysis are 1533 degrees Fahrenheit for Unit 1 and 1537 degrees Fahrenheit for Unit 2 and are presented on page 6.3.3-9 of the Brunswick UFSAR. Based on this information, the staff has determined that the resultant leak rate of 30 gallons per minute will have little effect, if any, on the calculated PCT in the l Brunswick Unit 1 LOCA analysis. '

4.0 CONCLUSION

1 The staff has determined that the flaw evaluation meets the rules of the ASME Code and the f assumed IGSCC and FIV crack growth rates are adequate for this application. Since the I predicted final crack length at the end of one additional cycle (8.1 inches) is far less than the allowable crack length (18.9 inches) from the limit load analysis, the staff determined that, from the standpoint of flaw evaluation, Brunswick 1 can be operated without repair for one additional l fuel cycle. The staff has also verified the licensee's calculated leakage rate and supporting documentation.

The calculated PCT of 1533 degrees Fahrenheit for Brunswick Unit 1 is well below the regulation limit of 2200 degrees Fahrenheit based on the estimated length of the crack indications at the end of the next fuel cycle. Based on the PCT value, the staff has determined that the calculated jet pump leakage will not impact the LPCI flow into the core during a DBA LOCA, so that operation in the proposed manner for the next fuel cycle meets the requirements of 10 CFR 50.46 with the calculated jet pump leakage.

The staff concludes that operating Brunswick Unit 1 without repair of the jet pump riser cracks until the next refueling outage is acceptable. It is the staff's expectation that the jet pump will be reexamined in the next refueling outage in accordance with the provisions of the latest version of BWRVIP-41.

5.0 REFERENCES

(1) Jury, K.R., CP&L, to USNRC, " Brunswick Steam Electric Plant, Unit No.1 - Docket No.

50-325/ License No. DPR-/1 - Jet Pump Riser Weld Inspection Results," May 29,1998.

(2) Jury, K.R., CP&L, to USNRC, " Brunswick Steam Electric Plant, Unit No.1 - Docket No.

50-325/ License No. DPR Jet Pump Riser Detailed Flaw Evaluation Technical Report (NRC TAC No. MA2436)," February 5,1999.

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(3) Jury, K.R., CP&L, to USNRC, " Brunswick Steam Electric Plant, Unit No.1 - Docket No.

50-325/ License No. DPR Additional information Regarding Jet Pumo Riser Weld f

inspection Results (NRC TAC No. MA2436),* Mty 17,1999.

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(4) Keenan, J.S., CP&L, to USNRC, " Brunswick Steam Electric Plant, Units 1 and 2, Revision Number 16 to the Brunswick Updated Final Safety Analysis Report,"

September 14,1998.

(5) GE Nuclear Energy, " Brunswick Steam Electric Plant Units 1 and 2 - SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis: Application to GE13 Fuel," NEDC-31623P Supplement 3 Revision 0, January 1996. ,

Principal Contributors: S.Sheng i K.Kavanagh Date: August 10, 1999 l

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