NRC-97-0111, Forwards Response to 970909 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions,

Forwards Response to 970909 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions,

ML20198J537
Person / Time
Site:	Fermi
Issue date:	10/17/1997
From:	Gipson D DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-97-0111, CON-NRC-97-111 TAC-M96811, NUDOCS 9710210124
Download: ML20198J537 (17)

Text

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October 17,1997 l

NRC-97-0111

! IJ.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D. C. 20555

References:

1) Fenn12 NRC Docket No. 50-341 NRC License No. NPF-43

2) NRC Generic letter 96-06: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions, dated September 30,1996

3) Detroit Edison Letter to NRC," Detroit Edison 30-Day Response to NRC Generic Letter 96-06," NRC 96-0118, dated October 30,1996

4) Detroit Edison Letter to NRC," Detroit Edison 120 Day Response to NRC Generic Letter 96-06," NRC 97-0003, dated January 28,1997

5) NRC Letter to Detroit Edison," Request for Additional Information Related to the Fermi-2 Response to Generic Letter 96-06," dated September 9,1997 (TAC No M96811)

Subject:

Additional Infonnation Related to Detroit Edison Response to NRC Generic I etter 96-06 In Reference 5, the NRC requested Detroit Edison to provide additional 7

ir. formation regarding the 120 day Response (Reference 4) to Generic Letter 96-06.

s

--4 e 9710210124 971017 /L'O /l PDR ADOCK 05000341 ['\

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, USNRC NRC 97 Oll!

Page 2 Detroit Edison was requested to provide summaries of the evaluations performed for the pipes in the six affected penetrations which are susceptible to thermally induced overpressurization. These summaries were to include: (1) fabrication drawings of piping sections evaluated; (2) a description of the analysis, including assumptions used in the analysis and; (3) the results of the analysis. Additionally, the response was to explain how the criteria used for the evaluations met the licensing basis criteria for Fermi 2 and to provide the schedule for the completion of any modifications that may be required. The enclosure to this letter provides the requested additional information as stated above, in Reference 4 Detroit Edison committed to completing necessary modifications by the next refueling outage. Detroit Edison is participating in EPRI analysis and l iesticg c.:'ivities intended to establish a technical basis that can be used to l demc 1 strate aceptability of these penetrations without modification. Because l

compc: tion of t'tese activities may not allow adequate time to properly design and install .nou 5:ations during the next refueling outage, the commitment from l Reference 4 is revised, as follows:

1 Detroit Edison will design and install any modifications which inay be necessary based on the results of these activities. Detroit Edison will provide an update to the NRC by March 31,1998 regarding the progress of this effort, including any proposed modifications identified to date.

if you have any questions, please contact Mr. Norman K. Peterson, Director, Nuclear Licensing at (313) 586-4258.

Sincerely, ,

Enclosure cc: A. B. Beach G. A. Harris M. J. Jordan A. J. Kugler

i i

1

[ USNRC l NRC 97 0111-Page 3 t

I, DOUGLAS R. GIPSON, do hereby afrirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best ofiny knowledge and belief.

b i

DOUGLAS R.%1PSON _.

Senior Vice President On this /7 M day of .O _.are me personally appeared Douglas R. Gipson, being first duly sworn and says that he executed the foregoing as his free act and deed, i

Notary Public

• SHARON K BUCKLEY-g aMonroe cound.Mt WoossiedonhpkesSept 22,20@ q l *

. . . . . , . ~ - . - . - . - . - - . - . - . . . - . . _ - - - , - . , _ - . - - . . . , .

, Enclosure to NRC 97 0111 Page1 Detroit Edison Response to the NRC Reuucst for Addinonalinformallo_ 1 (RAI) Related to Detroit Edison Response to Generie Letter 96-06.

(TAC No. M96811)

The following is the Detroit Edison response to the items requested by the NRC in the enelosure of the NRC letter dated September 9,1997:

e NRC RAI Item Number 1: "The NRC requests that DECO provide summaries of the evaluaties performed for the pipes in the six affected penetrations. The summaries should include fabrication drawings of the piping sections evaluated and describe the method of analysis, including the assumptions used in the analyse,s, and the results "

Detroit Edison Response:

As discussed in Detroit Edison's 120 day response to Generic Letter 96 06 (Reference 4), Fermi 2 has six drywell penetrations that are susceptible to thennally induced overpressurization due to increase in the drywell ambient temperature after a LOCA. As requested in the NRC RAI, enclosed are six l figures describing the piping sections and valves for these six penetrations.

l The figures summarize .he infomiation from the many fabrication drawings that are used to describe the installation. Additionally, three supporting figures are provided describing details of the containment penetrations.

These penetrations are:

1. Penetration X 8: Main Steam Drain Line

2. Penetration X 18: Drywell Floor Drain Sump Discharge Line

3. Penetration X-19: Drywell Equipment Drain Sump Discharge Line

4. Penetration X 29Aa: Reactor Recirculation Water Sample Line

5. Penetration X Sla: Reactor Recirculation Pump A Seal Purge Line

6. Penetration X-49a: Reactor Recirculation Pump B Seal Purge Line None of these susceptible penetrations are required to operate after a LOCA.

Five of the six penetration lines, the Main Steam Drain line (Figure 1); Drywell Floor Drain and Equipment Drain Sump Discharge lines (Figures 2 & 3); and Recirculation Pump Seals A and B Purge lines (Figures 5 & 6) are connected to non-safety related piping immediately outside the containment isolation valve.

These systems are not required to operate after accidents. The Recirculation Water Sample line penetration (Figure 4) att*.ched to non-safety related tubing near the sample cooler and would also noi ne used after an accident.

. Enclosure to NRC 97 0111 Page 2 For the isolated drywell penetrations susceptible to thermal overpressurization, operability means that containment integrity is maintained considering the single failure criteria. As discussed in Reference 4, the Fermi 2 operability evaluation for these penetrations relies on engineering judgmentt no detailed stress analysis was performed. The evaluation concluded that operability of containment penetrations was not adversely alTected by the aforementioned phenomeno t This approach is considered satisfactory since there is reasonable assurance that the penetrations will maintain containment integrity.

As discussed in Reference 4, penetration voiding may exist that would prevent overpressurization; or valve seat, bonnet, or packing leakage would be expected to limit any pressure rise to less than the failure pressure. A review of Fermi 2 Local Leak Rate Test data confirmed that all six penetrations exhibit some leakage. Leakage rates of air at 56 psig are not easily correlated to water leakage rates at several thousand psig. Ilowever, for the worst penetration, which is the drywell equipment drain sump discharge line, less than one half gallon is required to leak to completely depressurize the piping.

The main steam drains drywell penetration is isolated aller reaching full reactor pressure and temperature, when drain flow is minimum and the drains are no longer needed. At this condition the penetration piping is expected to be illied with steam, instead of completely full with water. Also, the piping would be above ambient temperature due to interfacing high temperature systems on each side of the isolation valves. Therefore, it is highly unlikely that this piping will be water solid to the extent that it can be subjected to thermal overpressurization.

The containment sump lines would not be expected to be completely full since the penetration is located at the high point of the system. The inboard isolation valves are normally open which allows the piping to leak through the pump discharge check valves. The piping outside containment also includes a vacuum breaker approximately at the same elevation as the penetration piping.

These penetrations are isolated with solid wedge gate valves which would also leak through the packing, bonnet boltface or across the seat before piping failure would be expected.

The Recirculation Pump Seal Purge lines and the Recirculation Sample line are likely to be filled solid with water. However, the containment isolation valves are air-operated spring to-close angle globe valves, and at least one of the isolation valves in each of these penetrations is oriented so that flow is under seat and their expanding volume of water is very small. This type of valve is not leak tight, and would be expected to leak to limit thermal overpressurization before the 3/4" and 1" Schedule 80 or 160 stainless steel

. Enclosure to NRC-97-0111 Page 3 piping would fail. The excess pressure would be expected to lif1 the spring before overpressurizing the piping to failure.

The thermal overpressurization loading is not expected to cause the piping system to fail because the pressure will cause strains in the piping and provide increase in the volume for fluid to expand before causing overpressurization.

The load is self limiting for the temperature increases during postulated LOCA. Self limiting loads are typically not included in emergency or faulted condition stress evaluations since a single occurrence of the load will not cause failure.

In the unlikely event of mechanical failure, containment integrity is expected to be maintained. llecause the penetration assemblies have greater wall thickness, the piping and components inside and outside the containment are more susceptible to overpressure failure than the common piping in the penetration assembly. Simultaneous failures would be required both inside and outside the containment before containment integrity could be affected.

• NRC RAlltem Number _21"The response from DECO should also explain how the criteria used for the evaluations meet the licensing basis criteria for Fenni 2..."

Detroit Edison Response:

The licensing basis of the primary containment is to provide the capabili:y of limiting the release of fission products within the values specified in 10 CFR 100, in the event of a postuimed LOCA. The licensing basis for the piping systems penetrating primary containment is to maintain the integrity of the containment with double barrier protection, such that no single active failure will result in the loss of containment integrity. To maintain the containment integrity, containment penetrations have the following design characteristics: (a) capability to withstand peak transient pressures expected during a LOCA: (b) capability to withstand without failure the forces caused by impingement of the fluid from the rupture of the largest local pipe or connection;(c) capability to accommodate without failure the thermal and mechanical stresses that may be encountered during all modes of operation.

The UFSAR Sections 3.1,3.2,3.8,3.9, and 6.2 describe Fermi 2's compliance with the above mentioned licensing basis of containment penetrations, isolation valves and piping between the isolation valves.

Enclosure to NRC 97 0111 Page 4 Generic Letter 96-06 identifies conditions where thermal expansion of fluid trapped in isolated piping could lead to overpressurization of the piping syste:s The Generic Letter also states that piping systems u hich have the potential to experience pressurization due to trapped fluid expansion shall either be designed to withstand the increased pressure or shall have provisions for relieving the excess pressure. This requirement is contained in AShiE Section Ill. NC 3621.2 " Fluid Expansion Effect". Ilowever, per AShiE Code Interpretations N197 008 and N197 011, this requirement is only related to nonnal design and operating conditions, and is not applicable to cases where i

thennal expansion of fluid trapped in isolated piping occurs only during one-l time accident conditions. The stress caused by the volume expansion of the l trapped fluid due to temperature increase after LOCA exhibits the self limiting characteristics of a secondary stress as defined in AShiE Section 111, Nis-3213.9 " Secondary Stress", and failure from such stress is not expected.

For Fenni 2, the design basis for the components subjected to thermal overpressurization loads is AShiE Section 111 (1971), and the components are classified as Class I or Class 2. As these components are not subject to thermal overpressure during their normal operating conditions, their design basis does not include the thermal overpressure load case and the loading is not defined in the AShiE Design Specifications for these components. Also, the component stress analysis reports do not include an evaluation of the thermal overpressure load case. The practice of not including the thermal overpressure load case is based upon Detroit Edison's conclusion that evaluation of this load case is not required by AShiE Section 111, as supported by Code Interpretations N197-008 and N197 011.

Nevertheless, during the period ofinitial licensing of Fermi 2, Detroit Edison has addressed thermal expansion of fluid trapped in isolated piping where it wa:: required by the operating conditions of the piping system. In response to an NRC question (FSAR Question 212.169) specific to the RilR suction line between isolation valves El150F008 and El150F009, which is used for shu*down cooling (penetration X-12), provisions were made to accommodate post-LOCA thennal expansion of water trapped between these valves.

Although other penetrations were also reviewed by the NRC, the context of this request did not provide a discussion or reason for being specific to this penetration. As described in Fermi 2 UFSAR Section 6.3.2.16, a 3/4 inch thennal reliefline was added which allowed the heated water trapped between the isolation valves to be relieved back to the reactor. This provision was found acceptable by the NRC in Section 5.4.2 of the Fermi 2 Safety Evaluation Report (NUREG-0798). This 4 unique case related to a specific penetration where the operat: g conditions may have warranted consideration 1

. Enclosure to NRC 97 0111 Page 5 4

of thermal expansion of fluid trapped in isolated piping during accident conditions.

In summary, Detroit Edison concludes that the Fermi 2 licensing basis does not generally contain requirements to design for pressurization due to thermal expansion of trapped fluid during a one time accident occurrence. The licensing basis requires that containment iritegrity be maintained following an accident. Detroit Edison's evaluation criteria provides reasonable assurance that containment integrity would be maintained following postulated accidents.

• NRC RAI Item Number 3: "The response from DECO should also... provide the schedule for the completion of any modifications that may be required."

Detroit Edison Response:

Detroit Edison is participating in the EPRI GL 96 06 Utility Group to perform testing and analysis to establish a technical basis that can be used to demonstrate acceptability of these penetrations without modification. Detroit Edison will design and install any modifications which are necessary based on the results of these activities. Detroit Edison will provide an update to the NRC by March 31,1998 regarding the progress of this effort, including any proposed modifica' ions identified to date.

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