ULNRC-04034, Forwards Amerenues Risk Evaluation Summary & Provides Listing of Other Documents Which Have Been Previously Provided to Support Evaluation of Electrosleeves at High Temp Severe Accident Conditions

Forwards Amerenues Risk Evaluation Summary & Provides Listing of Other Documents Which Have Been Previously Provided to Support Evaluation of Electrosleeves at High Temp Severe Accident Conditions

ML20206U574
Person / Time
Site:	Callaway
Issue date:	05/17/1999
From:	Randolph G UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-REGGD-01.174, RTR-REGGD-1.174 TAC-MA3954, ULNRC-04034, ULNRC-4034, NUDOCS 9905250260
Download: ML20206U574 (10)

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Union uxtric PO Box G20 l Callaway Plant Fulton, MO 65251 571 678.8745 Carry L Randolph 5736764056 fax Vice President and Chief Nxlear Officer May 17,1999 U.S. Nuclear Regulatory Commission Attn

Document Control Desk Mail Station PI-137 Washington, DC 20555-0001 Gentlemen: ULNRC-04034 TAC No. MA 3954

$d CALLAWAY PLANT EINN UNION ELECTRIC COMPANY D CKET NUMBER 50-433 UE REVISION TO TECHNICAL SPECIFICATION 3/4.4 REACTOR COOLANT SYSTEM

References:

1) ULNRC-3358 dated April 12,1996

2) ULNRC-3910 dated October 27,1998

3) ULNRC-3920 dated November 13,1998

4) ULNRC-3948 dated January 11,1999

5) ULNRC-3955 dated January 29,10"

6) UL,NRC-3970 dated February 25,1999

7) NRC letter dated April 6,1999, Record ofMarch 18,1999 Telecon

8) ULNRC-4004 dated April 7,1999

9) ULNRC-4005 dated April 7,1999 Reference 1 transmitted an amendment request to revise Callaway Technical Specifications to use Electrosleeves to repair steam generator tubes. Reference 2 transmitted a modified request limited to two operating cycles to address NDE issues. References 3 through 9 modified the request based on Staffinput.

In February,1999, NRC contacted AmerenUE personnel concerning evaluation of Electrosleeve material at high temperature severe accident conditions. Evaluation of Electrosleeves at high temperature severe accident conditions was not anticipated by AmerenUE, since it is beyond the licensing and design basis for Callaway Plant.

A number of telephone conversations and one meeting (April 22,1999) have taken place on this issue. AmerenUE has informally provided risk evaluation information to address the concerns raised by the staff. This letter is a formal transmittal of AmerenUE's risk evaluation summary and provides a listing of other documents which have been previously provided to support the evaluation. Since the referenced amendment request was based on demonstrating that the Electrosleeve steam generator tube repair method met currently approved staff positions and the Callaway Licensing Basis, the risk evaluation was not prepared based on the guidance provided in NRC Regulatory Guide 1.174, An Approach g for Using Probabilistic Risk Assessment In Risk-InformedDecisions On Plant-Specific

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Changes to the Licensing Basis. 0 l 9905250560 990517 i PDR ADOCK 05000483 '

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, U.S. Nuclear Regulatory Commission May 17,1999 Page 2 AmerenUE offers the following items for NRC consideration of the requested amendment:

1) This repair meth.xl meets all regulatory requirements based on approved Staff positions. (This fact was stated by the Staffin the April 22,1999 meeting between AmerenUE and NRC.)

2) A risk evaluation performed by AmerenUE, specific to the Callaway Plant, shows scenarios, under which induced steam generator tube ruptures are postulated to potentially occur, to have an estimated core damage frequency (CDF) of 1 approximately 1.70E-6 yrd .

3) The delta LERF (Large Early Release Frequency), for this repair method, would be less than 1E-6 yr' , since the above CDF would be multiplied by the conditional incremental probability ofinduced tube rupture of Electrosleeved tubes (as compared to non-repaired tubes).

4) Based on evaluations performed by AmerenUE specific to the Callaway design using the MAAP (Modular Accident Analysis Program) Code, the steam generator hot leg tube temperatures do not reach the temperature at which Electrosleeve tube samples failed in testing.

5) The testing associated with Electrosleeve samples is significantly more conservative than what would exist in actual plant conditions. The process of manufacturing a flaw using EDM (Electro Disintegration Machining) produces a flaw that is more susceptible to failure than flaws which develop in tubes in actual steam generator conditions. Framatome Technologies, Inc. (FTI) has recently tested two Electrosleeves with flaws machined to more closely simulate plant flaws, and have obtained results which show failure at 699 C and 689 C.

AmerenUE believes Electrosleeves represent a steam generator tube repair method superior to any other tube repair method available to date. Based on the attached information and the referenced letters, we request that the Staffissue the license amendment requested in References 2 through 9. Please respond to this request by May 21,1999, ifyou determine that the License Amendment cannot be approved as requested, please identify the basis for that determination by May 21,1999.

Sincerely, Garry L. Randolph  !

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DS/ACP/jdg Attachments: 1) List of PRA-Related transmittals to NRC on Electrosleeves

2) Issues Related Quantification of LERF Due to Installation of Electrosleeves i

i

e e cc: M. H. Fletcher Professional Nuclear Consulting, Inc.

19041 Raines Drive Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-8064  !

Senior Resident Inspector '

Callaway Resident Office U.S. Nuclear Regulatory Commission-8201 NRC Road Steedman, MO 65077 Mr. Mel Gray (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Denny Buschbaum TU Electric P.O. Box 1002 Glen Rose, TX 76043 Pat Nugent Pacific Gas & Electric Regulatory Services P.O. Box 56 Avila Beach, CA 93424

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ATTACHMENT 1 1

List of PRA-Related Transmittals to NRC on Electrosiceve

/ 3/18/99 - A preliminary AmerenUE probabilistic model was faxed to NRC to support a telecon that same day.

/ 3/24/99 - Following another telecon on AmerenUEs PRA arguments, we faxed the updated Callaway Level 1 event trees, with core damage sequence frequencies printed thereon, to NRC.

/ 3/25/99 -We ovemighted hard copies of the following documents to NRC:

/ Updated Callaway Level 1 event trees, with sequence frequencies.

/ Severe accident management guideline SAG-1, Rev. O.

/ The OCL file (T1TC.OCL) that takes sequence T(1)S06 (LOSP-initiated loss of RCP seal cooling), and generates the resultant core damage sequence cutsets.

/ AmerenUE calculation BB-97, Rev. O, which determines the probabilities of core uncovery, given a loss of RCP seal cooling under various scenarios.

/ Prior to the 4/22/99 meeting, between AmerenUE, FTl and NRC, AmerenUE provided a white paper documenting our PRA assessment of the induced SGTR issue related to the Electrosleeve amendment.

/ 4/27/99 - A white paper entitled " Issues Related to Quantification of ALERF Due to Installation of Electrosleeves" was transmitted to NRC. (This document is provided as Attachment 2.) This transmittalI document should be considered AmerenUEs PRAjustification for approval of the Electrosleeve license amendment, as it was developed based upon a more complete understanding of NRCs LERF concerns, and represents our refined position with respect to ALERF due to installation of Electrosleeves.

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ATTACHMENT 2 AmerenUE Callaway Plant ELECTROSLEEVE LICENSE AMENDMENT Issues Related to Ouantification of ALERF Due to Installation of Electrosleeves April,1999 Page 1 of 4

. . . . . .. _ . . . . . . . . . _ . .,-:.- ,,..,n.m .

.. ATTACHMENT 2 ELECTROSLEEVE LICENSE AMENDMENT Issues Related to Ouantification of ALERF Due to Installation of Electrosleeves

1) Approximately 25% (1.63E-6 y(') of the total CDF (6.61E-6 y(') of the candidate sequences for induced SGTR are non-station blackout core damage sequences. These sequences could be mitigated by injection of firewater into one (1) steam generator, pursuant to FR-H.1. Core damage would be precluded if firewater were successfully injected into a steam generator. This success path is not credited in the CDF estimate noted above. Using an estimated probability of 0.13 for failure to inject firewater into a steam generator, the non-SBO contribution is reduced to 2.12E-7 y('.

2) The highest frequency (2.75E-6 y(') SBO core damage sequence, which could contribute to the frequency of induced SGTR, is T(1S)S26. This is a station blackout, followed by failure of the turbin64 riven auxiliary feed pump, and failure to recover AC power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In actuality, core damage will begin at approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 1-hour timeframe was chosen, for CDF quantification, to provide approximately.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the operators to start and align mitigating systems, such as ECCS, to prevent core damage. For quantification of the potential for induced SGTR, use of the full 2-hour timeframe is more appropriate. At 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, core uncovery begins. However, the SGs (secondary side) are still pressurized. If AC power is recovered '

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, operators can inject into the RCS with the ECCS, and begin a controlled injection with the motor-driven auxiliary feed water pumps into the SGs.

NUREG-1032 is used in the Callaway PRA to determine probabilities of AC power recovery following station blackout. The probability of failure to recover AC power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, used in the quantification of CDF for T(1S)S26, is 0.389. The probability for failure to recover within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is 0.222. If the 2-hour timeframe is credited, the frequency of T(1S)S26 is reduced from 2.75E-6 y(' to 1.57E-6 y('.

Removing three additional conservatisms that exist in the calculation can credibly reduce the 1.57E-6 y(' value further. These three conservatisms are:

a) The diesel-generator (DGN) fault trees model a direct dependency of the DGNs on diesel building HVAC. However, AmerenUE has previously determined that this dependency does not actually exist, unless the outside ambient temperature is greater than 65*F. Using data in the Callaway FSAR Site Addendum, the probability that the outside ambient temperature is greater than 65"F is, conservatively, 0.583.

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ATTACHMENT 2 b) Test and maintenance unavailability point estimates used in the updated  ;

Callaway PRA are based on data from the IPE (1/87 - 5/90) and Cycles 7, 8 and 9 (11/93 - 5/98). A review of this data shows that trains of l equipment were out of service for test / maintenance less time in Cycles 7, l 8 and 9 than during the IPE data collection period. Accordingly, more representative test / maintenance point estimates, for the diesel-generators, turbine-driven auxiliary feed water pump and essential service water pumps, can be derived using the Cycles 7,8 and 9 data only.

c) A review of cutsets for sequence T(1S)S26 reveals cutsets with both the turbine-driven auxiliary feed water pump and a diesel-generator in test /

maintenance. In fact, the plant configuration risk matrix (contained in plant procedures), developed pursuant to (a)(3) of the Maintenance Rule, )

identifies simultaneous outages of the turbine-driven auxiliary feed water pump and a diesel-generator as being undesirable from a plant risk perspective. Accordingly, concurrent planned outages of this equipment would not be undertaken.

Factoring the above three conservatisms into the frequenc T(1S)S26 further reduces the frequency, fromto1.57E-6 y('y calculation for 9.58E-7 y('.

In addition to the above discussion, MAAP runs for this sequence show that the steam generator tubes will not reach the requisite temperature for induced SGTR. (Refer to item 5, below.)

3) The second highest frequency SBO CD sequence, under evaluation for induced SGTR potential, is T(1S)S10. This sequence is a SBO, successful operation of the turbine-driven auxiliary feed water pump, successful cooldown and depressurization, but AC power is not restored within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The CDF for this sequence is 1.93E-6 y('.

MAAP runs for Callaway show that, given this sequence, the time at which core uncovery begins is actually 20.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The failure to recover probability for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is 4.44E-3 (based on NUREG-1032). NUREG-1032 only provides data for times up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, if 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> is used as a surrogate for the actual 20.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> timeframe, the failure to recover AC power probability is

< 1.22E-3. Therefore, the frequency for this sequence can be reduced to

< 5.31E-7 y('.

Also, per item 5, below, MAAP runs show that the SG tubes will not reach the temperatures required for induced SGTR.

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ATTACHMENT 2 ,

4) The RCP seal LOCA model, used in the Callaway PRA for sequences in which seal injection and cooling is lost, assigns a 90% probability to a o

leakage rate of 21.4 gpm per RCP. This leakage rate is insufficient in most scenarios to depressurize the RCS and thereby prevent induced SGTR from occurring. However, larger postulated RCP seal leakage rates would depressurize the RCS such that induced SGTR would not occur. It is important to note that Callaway CDFs are calculated assuming either (1) a 21.4 gpm per RCP LOCA or (2) a 90% probability of a seal LOCA of this size.

5) The discussion of increases in ALERF for the purposes of the Electrosleeve amendment should be limited to proposed scenarios where the steam generator tube temperatures fall between the failure temperature of an Alloy-600 tube (~1400 F with no tube degradation) and the failure temperature of an Electrosleeved tube (~1100 F with a 2 inch axial flaw). At temperatures above 1400 F, no increase in ALERF occurs since any tube (including the original Alloy-600 tube) is predicted to fail. At temperatures below 1100 F, no increase in ALERF occurs since no tubes are predicted to fail. This narrow

- temperature band limits the scenarios which would lead to an increase in ALERF.

The MAAP code was used to estimate the steam generator tube temperatures for the sequences identified in items 2 and 3 above. Figure 1 1 provides the steam generator hot leg tube temperatures for sequence T(1S)S26. Figure 2 provides the steam generator hot leg tube temperatures for sequence T(1S)S10. Both figures show that the steam generator tube temperatures remain below 1100 F, and thus result in no increase in ALERF.

Conclusions Based on analysis of issues related to quantification of ALERF due to installation of Electrosleeves, the following conclusions are drawn.

The frequency of Callaway core damage sequences that may precede induced SGTR is very small (approximately 1.70E-6 yr). In addition, this frequency would have to be multiplied by the conditional incremental probability of induced tube rupture of Electrosleeved tubes. MAAP runs for Callaway show that, for the

. dominant relevant two station blackout core damage sequences, the steam generator tubes would not reach the required temperature for induced rupture of either the Electrosleeved tubes, or Alloy-600 tubes. Therefore, the conditional probability of induced tube rupture can be taken to be zero.

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