ML20238F797
| ML20238F797 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/28/1998 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2554, GL-96-06, GL-96-6, TAC-M96803, NUDOCS 9809040325 | |
| Download: ML20238F797 (17) | |
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- F" a m** Davis-Besse Nuclear Power Station EE 5501 North State Route 2 m Oak Harbor,Chro43449-9760 John K. Wood 419-249-2300 Mce President - Nuclear Fax: 419-321-8337 Docket Number 50-346 License Number NPF-3 Serial Number 2554 August 28,1998 United States Nuclear Regulatory Commission
- Document Control Desk Washington, D. C. 20555-0001
Subject:
Response to NRC Request for Additional Information pertaining to Generic Letter 96-06: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (TAC No. M96803)
Ladies and Gentlemen:
On September 30,1996, the Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 96-06. That letter requested licensees, such as those for the Davis-Besse Nuclear Power Station (DBNPS) Unit Number 1, to address the following generic issues:
(1) Cooling water systems serving the containment air coolers (CACs) may be exposed to the f hydrodynamic effects of water hammer during either a loss-of-coolant accident (LOCA) or a [//
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main steam line break (MSLB). These cooling water systems were not designed to withstand the hydrodynamic effects of water hammer and corrective actions may be needed to satisfy system design and operability requirements. Licensees are to determine if their plant's CACs cooling water systems are susceptible to water hammer during postulated accident conditions. tp (2) Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios were based on single-phase flow conditions. Corrective actions may be needed to satisfy system. design and operability requirements. Licensees are to determine if their plant's CACs are susceptible to two-phase flow conditions during postulated accident conditions.
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(3) Thermally-induced overpressurization of isolated water-filled piping sections in containment could: 1) jeopardize the ability of accident-mitigating systems to perform their safety functions, and 2) could also lead to a breach of containment integrity via bypass leakage.
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Docket Number 50-346 License Number NPF-3 Serial Number 2554 l Page 2 Corrective actions may be needed to satisfy system operability requirements. Licensees are
! to determine if the piping systems which penetrate their plant's containment are susceptible to thermal expansion of fluid so that overpressurization could occur.
l On January 28,1997, Toledo Edison (TE) provided by letter (Serial Number 2439) an interim response to GL 96-06, On February 28,1997, TE provided by letter (Serial Number 2442) a
! summary report describing the actions taken to that date and the results of those actions. On July 28,1997, TE provided by letter (Serial Number 2473) an update on the status of TE's progress in resolving the issues of GL 96-06 and informing the NRC that plans for closure on the issue were delayed due to an unexpected forced outage that occurred because of a plant trip on May 4,1997. On September 30,1997 TE provided by letter (Serial Number 2488) a report on the resolution of post-LOCA thermal overpressurization of containment penetrations, and the water hammer and two phase flow effects on the CAC trains.
On March 18,1998, the NRC issued a Request for Additional Information (RAI) concerning l questions posed by NRC Office of Nuclear Reactor Regulation (NRR) staff regarding the 13 pipe lines penetrating the DBNPS containment that were identified by TE as possibly susceptible to thermally induced overpressurization. On August 17,1998, the backup NRC NRR Project Manager for the DBNPS was notified by telephone that TE's response had been prepared for submittal to the NRC and was undergoing final management review with submission anticipated f
by August 31,1998. Accordingly, the following provides TE response regarding the requested information.
In the submittal of September 30,1997, TE indicated that two pipelines meet the ASME Code pressure limitation under faulted-load combinations (penetrations 14 and 56). The NRC requested the following information for the piping runs associated with these lines:
Question 1: Provide the applicable design criteria for the piping and the valves. Include the ,
t required load combinations.
I Response 1: Refer to the following matrix:
Design Criteria Pen.# Component / Item Design Code Design Design ANSI Press. Temp. Class Piping Line CCB-21 ASME Section III, Class 2500 600 T M/A 2 1971 Edition / No psi Addenda 14 Valve MU2A (Inside Ctmt ASME Code for Nuclear 2500 600*F 1500#
l Isolation Valve) Pumps and Valves, Class psi j 1 (Draft 1968)
Valve MU3 (Outside Ctmt ASME Code for Nuclear 1400 100 T 1500#
isolation Valve) Pumps and Valves, Class psi I (Draft 1968)
Continued on Next Page
.. 4 Docket Number 50-346 License Number NPF-3 Serial Number 2554 l Page 3 Design Criteria Pen.# Component / Item Design Code Design Design ANSI Press. Temp. Class Piping Line CCB-20 ASME Section III, Class 2500 300 T N/A 2 1971 Edition / No psi Addenda 56 Valves MU59A/B/C/D ASME Code for Nuclear 2500 300 T 1500#
(Inside Ctmt Isolation Pumps and Valves, Class psi Valves) 1 (Draft 1968)
Valve MU38 (Outside Ctmt ASME Code for Nuclear 2500 300 T 1500#
Isolation Valve) Pumps and Valves, Class psi I (Draft 1968)
Load combinations for both penetrations are Pressure + Dead Weight + SSE Stresses (Reference USAR Table 3.9-3). The combination of these stresses are
. to be less than or equal to a factor of 1.0 of the material yield stress, at temperature, as represented by the following equation:
General Membrane Stress ( o ) + Bending Stress ( o 6 ) s 1.0
- Yield Stress (Sy).
Question 2: Provide a drawing of the piping run between the isolation valves. Include the lengths and thicknesses of the piping segments and the type and thickness of the insulation.
Response 2: Attachments A and B provide the piping geometry located between the containment isolation valves for penetrations 14 and 56, respectively. Also note that neither piping segment is insulated.
Question 3: Provide the maximum calculated temperature and pressure for the pipe run.
Describe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model used in the analysis and the basis for the heat transfer coefficients used in the analysis.
Response 3: The maximum calculated temperatures and pressures for each of the referenced penetrations are as follows:
P14 P56 Maximum Calculated Fluid Pressure 5135 psig 5235 psig Initial Fluid Temperature 120 T 100 P Maximum Calculated Fluid Temperature 240 T 250 P I
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t Docket Number 50-346 License Number NPF-3 Serial Number 2554 Page 4 Maximum Fluid Pressure Calculation Method The method used to calculate the resultant internal piping pressure of the isolated section of containment piping at the elevated accident temperature is based on the mass of the entrapped water remaining constant. Specifically, the initial mass of the entrapped water was first determined by calculating the internal pipe volume at initial ambient conditions and dividing by the specific volume of water at the same initial ambient condition. Similarly, the pipe volume was calculated at the post accident temperature for the affected pipe section which included metal expansion due to both temperature and pressure.
The accident pressure was then determined by applying an iterative process of selecting a fluid pressure that resulted in a water mass, at accident temperature, that closely approximated the water mass of the entrapped fluid at initial conditions.
Maximum Fluid Temperature Calculation Method The DBNPS has a very large (approximately 2.8x106 ft' net volume) free standing steel containment vessel. A 4.5 foot annular gap is provided between the steel containment liner and the 2.5 foot thick concrete shield building. The annular space is not exposed to direct steam heating following a LOCA. Due to the large containment size, the peak vapor temperature for design basis LOCA (approximately 260oF) is relatively low compared to similar plants with smaller containments. The large volume also increases the air to steam mass ratio, which tends to reduce condensing heat transfer. The design basis Main Steam Line Break (MSLB) has a more rapid blowdown and smaller mass release than the design basis LOCA. Therefore, the duration of elevated i temperatures and magnitude of heat transfer coefficients are smaller for MSLB than for LOCA. Thus, design basis LOCA was taken as the limiting event for penetration heating.
During the initial screening for potentially affected piping, peak piping temperatures following a design basis LOCA were conservatively chosen by reviewing the calculated thermal response of other heat sinks in the contain-ment model. Piping associated with penetrations 14 and 56 are both non-insulated inside containment. This resulted in estimates of 240 F for 2-1/2" piping associated with penetration 14 (letdown) and 250 F for the smaller 1" piping associated with penetration 56 (RCP seal return). The initial operating temperature of penetration 14 was conservatively assumed to be 120$F, while the initial temperature of penetration 56 was assumed to be 100 F. A simplified approach was used to provide temperature distribution along the ,
length of penetration piping. This approach assumes that the piping located in l
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Docket Number 50-346 License Number NPF-3 Serial Number 2554 Page 5 the annular space outside containment would be heated to the same temperature as an "infm' ite" length of piping located inside containment. Outside the 4-1/2 foot annular space, the piping temperature was assumed to be constant. This assumption is considered reasonable since heat transfer coefficients would be low in the auxiliary building and the initial temperature was chosen conservatively low.
Following deterrnination that physical modi 5 cations might not be needed for penetration 14 and 56, scoping calculations were performed to affirm that the chosen temperatures were conservative for design basis LOCA and MSLB.
These scoping calculations utilized a PC based version of the licensed computer code (COPATTA) used for US AR Chapter 6 containment analysis.
A simple " infinite" length penetration model was added to the containment model to simulate a non-insulated 2-1/2 inch diameter stainless steel cylinder, filled with water. (The 2-1/2 inch pipe was chosen for this review because the predicted temperature of the 2-1/2 inch pipe was farther from the peak containment temperature. Time above 250oF is less than 100 seconds, while temperatures above 240 F continue for nearly 500 seconds. Therefore the 250 F peak temperature used for the 1"line was considered conservative by inspection.) The temperature difference between the internal pipe wall and the liquid center was minimized by conservatively maximizing the overall heat transfer within the penetration model. The " modified Tagami" condensing heat transfer coefficient was chosen for the penetration. This coef5cient is described in the Bechtel Topical Report for COPATTA, BN-TOP-03. In this particular case, use of the modified Tagami heat transfer coefficient resulted in 2
a maximum heat transfer coefficient of over 245 BTU /hr-ft *F early in the 2
transient, decreasing to a little over 70 BTU /hr-ft *F at the time of the peak penetration temperature. A peak penetration temperature of approximately 236
- F was found for the LOCA case, while the MSLB case was considerably lower. Therefore the 240*F temperature which was used for the stress analysis was deemed to be appropriate.
In the submittal of September 30,1997, TE indicated that the air-operated isolation valves associated with penetrations 21 and 32 pipe lines provide inherent relief to prevent overpressure.
The NRC requested the following information for each pipe line.
Question 4: Describe the applicable design criteria for the piping and valves. Include the required load combinations.
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Docket Number 50-346 l License Number NPF-3 Serial Number 2554 Page 6 l Response 4: Refer to the following matrix:
l Design Criteria Pen.# Component / Item Design Code Design Design ANSI Press. Temp. Class Piping Line HCB-23 ASME Section Ill, Class 2 150 psi 150"F N/A l 1971 Edition / No Addenda 21 Valve DW6831 A (Inside ASME Section Ill, Class 2 160 psi 150"P 150#
Ctmt Isolation Valve) 1971 Edition / W'71 Addenda Valve DW6831B ASME Section III, Class 2 160 psi 150"F 150#
(Outside Cimt Isolation 1971 Edition / W'71 Valve) Addenda Piping Line HCB-32 ASME Section III, Class 2 165 psi 300"F N/A 1971 Edition / No Addenda 32 Valve RC1773A (Inside ASME Section III, Class 2 255 psi 150"F 150#
Ctmt isolation Valves) 1971 Edition / W'71 Addenda Valve RC1773B (Outside ASME Section Ill, Class 2 255 psi 150 *F 150#
Ctmt Isolation Valve) 1971 Edition / W'7i Addenda Load combinations for both penetrations are Pressure + Dead Weight + SSE Stresses (Reference USAR Table 3.9-3). The combination of these stresses are to be less than or equal to a factor of 1.0 of the material yield stress at the maximum accident operating temperature as represented by the following equation:
General Membrane Stress ( o m ) + Bending Stress ( o b ) s 1.0
- Yield Stress (Sy ).
Question 5: Provide a drawing of the valve. Provide the pressure at which the valve was determined to lift off its seat or leak and describe the method used to estimate this pressure. Discuss any sources of uncertainty associated with the estimated liftoff orleakage pressure.
Response 5: Penetration 32 is isolated with a 3" air-operated diaphragm valve manufactured by ITT Grinnell, which is designated as RC1773B. This valve utilizes a single acting, spring loaded," fail closed" actuator. A drawing of this valve is attached (Attachment C). The calculated lift-off pressure for this valve is 238 psig. The calculational method utilized a simple force balance. Inputs to this force balance included effective valve diaphragm area, actuator diaphragm area, and the as-measured air pressure (on the actuator) that is required to begin opening the valve, without system pressure assisting the valve. Dimensions of the internal valve parts were obtained from measurements of spare parts. The effective valve diaphragm area was taken as one half of the total diaphragm area, minus the area of the weir. Uncertainty in the lift-off pressure would be i
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Docket Number 50-346 License Number NPF-3 Serial Number 2554 Page 7 comprised primarily of changes in mechanical friction of the parts, or the presence of system pressure in the downstream port. The latter would supplement the opening force and would be conservative. Another uncertainty would involve potential future changes in'the spring pre-load following valve maintenance. However, considering the very low lift-off pressure, it is not considered credible for any accumulation of changes to threaten the efficacy of the inherent over-pressure protection.
Penetration 21 is isolated with a 4" air-operated unbalance globe valve with
! full area trim manufactured by Hammel Dahl, which is designated as DW6831 A. This penetration provides demineralized water to containment.
This valve utilizes a single acting, reverse spring loaded, " fail closed" actuator.
g A drawing of this valve is attached (Attachment D). The calculated lift-off pressure for this valve is 152 psig. The calculational method also utilized a simple force balance. Inputs to this force balance included plug area, actuator diaphragm area, and the as-measured air pressure (on the actuator) that is required to begin opening the valve, without system pressure assisting the valve. Dimensions of the internal valve parts were obtained from measurements of spare parts. Uncertainty in the lift-off pressure would be L comprised pr%arily of changes in mechanical friction of the packing and the presence of any system pressure in the downstream port. The latter would be limited to Containment back pressure, and could increase the lift pressure by up to 38 psig (maximum Containment pressure). Another uncertainty would involve potential future changes in the actuator spring pre-load following valve maintenance. However, again considering the wry low lift-off pressure, it is -
not considered credible for any accumulation of changes to threaten the efficacy of the inherent over-pressure protection.
Question 6: Provide the maximum calculated stress in the piping run based on the estimated liftoff or leakage pressure.
Response 6: As identified in the matrix provided in Response 4 above, the ANSI B16.5 pressure class for the containment isolation valves associated with penetrations g 21 and 32 is 150#. As such, ASME III, Table NC-3521-1 as referenced by Article NC-3521(b), allows the maximum internal pressure resulting from a
- Level D event to be as high as 1.5 times the allowable working pressure (i.e.
225 psig) at the applicable service temperature. This maximum pressure criteria equates to 338 psig for these penetrations. Since the liftoff or leakage pressure for these air operated penetration valves is less than this value, actual calculation of stresses for this piping was not required and was not performed.
Docket Number 50-346 License Number NPF-3 Serial Number 2554 Page 8 l
On April 15,1998 the NRC issued another RAI containing additional questions posed by the I NRC NRR staff. These questions relate to the Containment Air Coolers and possible water hammer and two phase flow in piping systems servicing the Containment Air Coolers.
Toledo Edison enlisted the consulting services of Fauske and Associates,Inc. to determine potential water hammer loads on the containment air cooler system piping. This original work was performed as promptly as could be accomplished, attempting to meet the schedule requirements of the original GL 96-06. Since that time, the NRC has refined the objectives of the effort and better defined the methods that are considered acceptable. The RAI reflects this more ,
focused definition and requires that the DBNPS obtain the services of the original consultant to describe how the previously completed work meets the April 15,1998 revised guidelines of the NRC. However, the consultant cannot complete the full scope of work within the requested response time of the RAI. Therefore, while TE has full confidence in the quality and acceptability of the consultant's methodology, TE requests that it be allowed until January 29, 1999, to provide its response.
In the previously referenced letter of September 30,1997 (Serial Number 2488), TE provided a report on the resolution of post-LOCA thermal overpressurization of containment penetrations, and the water hammer and two phase flow effects on the CAC trains. Toledo Edison committed to provide a summary of the actions taken to ensure compliance with the requirements of Generic Letter 96-06 after the modifications were incorporated during the Eleventh Refueling Outage in the Spring of 1998 (11RFO). The following table provides the final resolution for the thirteen containment penetrations previously identified as potentially susceptible to post-LOCA thermal overpressurization.
Penetration Name Final Status 14 RCS Letdown Meets ASME Code Faulted Stress Allowables and ASME Code pressure limitations under faulted conditions. No further action.
56 Reactor Coolant Pump Seal Meets ASME Code Faulted Stress Allowables and Return ASME Code pressure limitations under faulted conditions. No further action.
12 Component Cooling Water Installed Bypass Check Valve during May 4,1997 to the Control Rod Drive outage. No further action.
Mechanisms 21 Demineralized Water to AOV Globe valve-Provides inherent relief to Containment prevent overpressurization. No further action.
32- Reactor Coolant Drain to AOV Diaphragm Valve-Provides inherent relief to Reactor Coolant Drain Tank prevent overpressurization. No further action.
49 Refueling Canal Fill Procedure changes assure piping remains partially drained to prevent potential overpressurization.
Continued on Next Page I !
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Docket Number 50-346 License Number NPF-3 Serial Number 2554 Page 9 Penetration Name Final Status 74C Pressurizer Auxiliary Spray Procedure changes assure piping remains partially drained during plant operation to prevent potential overpressurization.
1 Pressurizer Sample Line Meets ASME Code Faulted Stress Allowables, however, procedure changes assure weekly sampling is coordinated with valve stroke testing to assure the penetration is isolated with the fluid at a higher temperature than possible after a design bases accident.
13 Containment Normal Sump Relief Valve installed during iIRFO to achieve full ASME Code compliance.
3 Component Cooling Water Meets ASME Code Faulted Stress Allowables. Bypass to Containment Check Valve Installed during 11RFO to achieve full ASME Code compliance.
4 Component Cooling Water Bypass Check Valve Installed during iIRFO to Return from Containment achieve full ASME Code compliance.
47a Core Flood Tank Sample Bypass Check Valve installed during iIRFO to Line achieve fu'l ASME Code compliance.
48 Pressurizer Quench Tank Bypass Check Valve installed during iiRFO to Outlet achieve full ASME Code compliance.
With respect to the issue pertaining to potential water hammer in Containment Air Cooler piping, Toledo Edison's letter, Serial 2488, identified that Anchor A160 had one weld that potentially could become over-stressed and that further analysis was required. Calculations are now complete and show that the stresses for Anchor A160 are within allowable values.
With submission of the information contained in this letter in response to the NRC RAI dated March 18,1998, and the modifications completed during 11RFO for the subject containment penetrations TE considers the issue of post-LOCA thermal overpressurization as applicable to the i DBNPS to be closed, i Should you have any questions or require additional information, please contact i Mr. James L. Freels, Manager - Regulatory Affairs, at (419) 321-8466.
l V, ly yours, l
/
FWK/laj cc: A. B. Beach, Regional Administrator, NRC Region III S. J. Campbell, DB-1 NRC Senior Resident Inspector A. G. Hansen, DB-1 NRC/NRR Project Manager Utility Radiological Safety Board
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Docket Number 50-346 License Number NPF-3 Serial Number 2554
- Page 10 I
REQUEST FOR ADDITIONALINFORMATION TO NRC GENERIC LETTER 96-06 FOR THE DAVIS-BESSE NUCLEAR POWER STATION -
UNIT NUMBER I
- I, John K. Wood, state that (1) I am Vice President - Nuclear of the Centerior Service Company,
. (2) I am duly authorized to execute and file this certification on behalf of the Toledo Edison
- Company and The Cleveland Electric Illuminating Company, and (3) the statements set forth .
herein are true and correct to the best of my knowledge, information and belief.
By: u John I/ Wood, iceMesident - Nuclear Affirmed and subscribed before me this 28th day of August,1998.
" /+6f AJ w Notary Public, Sta'le of Ohio Nora Lynn Flood.- My Commission expires September 4, 2002.
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