ML20248L876

Safety Evaluation Supporting Amends 124 & 122 to Licenses DPR-80 & DPR-82,respectively

ML20248L876
Person / Time
Site:	Diablo Canyon
Issue date:	03/12/1998
From:	NRC (Affiliation Not Assigned)
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ML20248L873	List: ML20248L871 ML20248L876
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NUDOCS 9803240325
Download: ML20248L876 (10)

Text

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a UNITED STATES j

j NUCLEAR REGULATORY COMMISSION 2

WAsMINGTON, D.C. enana man 1

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1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

RELATED TO AMENDMENT NO.124 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO.122 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND Fl FCTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By application dated February 261997, as supplemented by letters dated December 23,1997, January 30, and February 9,1998, Pacific Gas and Electric Company (or the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License Nos. DPR-80 and DPR-82) for the Diablo Canyon Nuclear Power Plant, Units 1 and 2. The proposed changes revise the combined Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP) Unit Nos.1 and 2 to revise TS 3/4.4.5 and 3.4.6.2 including associated Bases 3/4.4.5 and 3/4.4.6.2 to allow the implementation of steam generator (SG) tube voltage based repair criteria for outside diameter stress corrosion cracking (ODSCC) indications at tube to tube support plant (TSP) intersections. The allowed primary-to secondary operational leakage from any one SG would be reduced from 500 gpd to 150 gpd.

The December 23,1997, January 30,1998, and February 9,1998, supplemental letters provided additional clarifying information and did not change the initial no significant hazards consideration determination published in the Federal Register on April 9,1997 (62 FR 17239).

2.0 BACKGROUND

Steam generator tube flaw acceptance critona (i.e., plugging limits) are specifkxi in the plant technical specifications. The traditional strategy for achieving adequate structural and leakage integrity of the tubes has been to establish a minimum wall thickness requirement in accordance with NRC Regulatory Guide (RG) 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes." Development of minimum wall thickness requirements to satisfy RG 1.121 was govemed by analyses assuming a uniform thinning of the tube wall. This assumed degradation mechanism is inherently conservative for certain forms of steam generator tube degradation. Conservative repair limits may lead to removing degraded tubes from service that may otherwise have adequate structural and leakage integrity for further service.

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I The staff developed generic criteria for voltage-based limits for ODSCC confined within the thickness of the TSPs. The staff published several conclusions regarding voitage-based repair criteria in draft NUREG-1477," Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" and in a draft generic letter titled " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the Federal Reaister on August 12,1994 (59 FR 41520). On August 3,1995, the staff issued GL 95-05, which took into consideration public comments on the draft generic letter cited above, domestic operating experience under the voltage-based repair criteria, and additional data made available from European nuclear power plants.

The guidance of GL 95-05 does not set depth-based limits on predominantly axially oriented ODSCC at TSP locations; rather it relies on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil volage, and tube burst pressure and leak rate. The staff recognizes that although the total tube integrity margins may be reduced following application of a voltage-based repair criteria, the riuidance in GL 95-05 ensures structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR Part 50 and the guidehne values in 10 CFR Part 100. Since the voltage-based repair criteria do not incorporate a minimum tube wall thickness requirement, there is the possibility for tubes with through-wall cracks to remain in-service. Because of the increased likelihood of such flaws, the staff included provisions for augmented steam generator tube inspections and more restrictive operational leakage limits.

The licensee's proposed amendment requests a change to the Diablo Canyon Units 1 and 2 Technical Specifications to incorporate the voltage-based repair criteria in accordance with the guidance of GL 95-05. The guidance specifies, in part, that (1) the repair criteria are only applicable to predominantly axially oriented ODSCC located within the bounds of the TSPs; (2) licensees should perform an evaluation to confirm that the steam generator tubes will retain adequate structural and leakage integrity until the next scheduled inspection; (3) licensees should adhere to specific inspection criteria to ensure consistency in methods between inspections; (f) tubes must be periodically removed from the steam generators to verify the morphology of trio degradation and provide additional data for structural and leakage integrity evaluations; (5) the operational leakage limit should be reduced; (6) licensees should implement an operational leakage monitoring program; and (7) specific reporting requirements should be incorporated into the plant technical specifications.

Each Diablo Canyon unit has four Westinghouse model 51 steam generators, which contain mill-annealed alloy 600 tubing. These steam generators have drilled-hole tube support p!ates and do not have flow distribution baffle plates. The outside diameter of each tube is 7/8 inch.

3.0 EVALUATION 3.1 Materials Evaluation The licensee will comply with the guidance in GL 95-05 when implementing the voltage-based attemate repair criteria. However, as permitted by GL 95-05, the licensee proposed several attematives to the guidance in GL 95-05. They are as follows: (1) use of an altemate i

4 inspection scope for dents greater than or equal to 5 volts in Unit 1, (2) use of an attemate approach for addressing probe wear, and (3) use of an alternate probability of detection (POD).

l These altemative approaches are discussed in Sections 3.1.2 and 3.1.3 of this safety evaluation. In addition, the licensee proposed to use an industry proposed and NRC accepted method to address new bobbin coil probe variability, which is discussed in Section 3.1.2 of this safety evaluation. Lastly, the licensee proposed to incorporate verbatim the model technical specifications in GL 95-05 into the technical specifications for both units, with two exceptions.

These exceptions are discussed in Section 3.1.3.3 of this safety evaluation. The major issues related to the licensee's implementation of the altemate repair criteria are discussed below.

3.1.1 Tube Repair Limits The proposed criteria will (1) permit tubes having indications confined to within the thickness of the tube support plates with bobbin voltages less than or equal to 2.0 volts to remain in service; (2) permit tubes having indications confined to within the thickness of the tube support plates with bobbin voltages greater than 2.0 volts but less than or equal to the upper voltage limit, to remain in service if a motorized rotating pancake coil probe or acceptable attemative inspection does not detect degradation; and (3) require tubes having indications confined to within the thickness of the tube support plates with bobbin voltages greater than the upper voltage limit be plugged or repaired.

The proposed lower voltage limit of 2.0 volts is consistent with the recommended value specified in GL 95-05 for 7/8-inch steam generator tubing. The upper voltage limit is derived based on the lower 95 percer t prediction interval of the burst pressure versus bobbin voltage correlation, adjusted for lower bound material properties evaluated at the 95 percent confidence level. The upper voltage limit is further reduced to account for uncertainty in the nondestructive examination technique and flaw growth over the next operating cycle. The industry periodically updates the database for burst pressure and bcibbin voltage when the destructive test data from pulled tubes are available; therefore, the upper voltage limit may vary as additional data are incorporated into the database.

3.1.2 Inspection issues insoection of Dent Sianals Section 3.b.3 of Attachment 1 to GL 95-05 specifies that all tube support plate intersections with dent signals greater than or equal to (a) 5 volts should be inspected with a rotating pancake coil (RPC) probe. Any tube with indications found at such intersections by RPC will be repaired if indications are circumferentially oriented or caused by primary water stress corrosion cracking (PWSCC), it may be necessary to expand the RPC sampling plan to include dent signals less than 5 volts.

Unit 1 steam generators contain a large number of dent signals 25 volts at TSP intersections.

RPC inspection of all of these dent signals, as specified in GL 95-05, would not be practical.

The licensee proposed the following attemative to the inspection criteria specified in GL 95-05.

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~ The licensee will RPC inspect 100 percent of hot leg dent signals 25 volts up to and including the highest TSP elevation where PWSCC, circumferential indications, or axial ODSCC not detected by bobbin have been previously detected in that steam generator. in addition,20 percent of dent signals 25 volts at each higher hot leg

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tube support plate elevation will be RPC inspected. If PWSCC, circumferential indications, or axial ODSCC not detected by bobbin are identified by RPC in any 20 percent sample, then in the affected steam generator, the licensee will RPC inspect 100 percent of 25 volt dent signals at that TSP elevation. If PWSCC, circumferential indications, or axial ODSCC not detected by bobbin are ident;fied at the seventh hot leg TSP, then the 20 percent expansion will be applied to the cold leg, and will continue down to the lowest cold leg TSP elevation until a 20 percent sample is obtained that is free from PWSCC, circumferential indications, or axial ODSCC not detected by bobbin. For any 20 percent sample, a minimum of 50 dent signals 25 volts in that steam generator at that TSP elevation will be inspected. If there are less than 50 dent signals 25 volts at that TSP elevation in that steam generator, all the 25 volt dent signals at that TSP elevation in that steam generator will be inspected.

The licensee will follow the guidelines in GL 95-05 to RPC inspect all dent signals 25 volts at TSP intersections in Unit 2 steam generators.

The licensee has also proposed an RPC inspection plan for dent signals less than (<) 5 volts, because axial PWSCC indications have been identified in dent signals <5 volts in Diablo Canyon steam generators. The inspection plan, which applies to Units 1 and 2, is as follows:

The licensee will RPC inspect all dent signals at TSP intersections <5 volts which contain bobbin indications that could remain in service under the voltage-based repair criteria. If this RPC inspection identifies PWSCC or circumferential cracking indications, the tube will be repaired.

In addition to the inspection plan for dent signals <5 volts, described above, the licensee has proposed an augmented inspection plan. The licensee will RPC inspect 100 percent of dent signals at TSP intersections 22 volts and <5 volts up to and including the highest TSP elevation where PWSCC, circumferential indications, or axial ODSCC not detected by bobbin have been detected in that steam generator. In addition, a 20 percent RPC inspection at the next highest TSP elevation will be conducted. If PWSCC, circumferential indications or axial ODSCC not detected by bobbin are identified in the 20 percent sample, then in the affected steam generator, the licensee will RPC inspect 100 percent of the 2 to 5 voit dent signals at that TSP elevation, and they will inspect 20 percent at the next highest TSP elevation. This expansion will continue down to the lowest cold leg TSP elevation until a 20 percent sample is obtained which is free from PWSCC, circumferential indications or axial ODSCC not detected by bobbin.

l The licensee will implement an inspection program for dont signals <2 volts similar to the program for 2 to 5 voit dent signals for both units.

i 5-The proposed attemative inspection plan for dent signals 25 volts in Unit i steam generators and <5 volts in Unit 1 and 2 steam generators is acceptable, because the inspection and expansion criteria would result in inspection of an adequate number of dented intersections to i

identify tubes that should be removed from service. The inspection plan for dent signals 25 volts in Unit 2 steam generators follows GL 95-05 and, therefore, is acceptable. If changes in eddy current technology indicate that the bobbin coil probe can be qualified to cetect degradation in dent signals of a particular voltage range, NRC staff would like to have the opportunity to review the qualification data before PG&E implements the technique.

Probe Variability With respect to probe variability (Section 3.2.c of Attachment 1 to GL 95-05), the licensee proposed to follow an approach developed through the Nuclear Energy Institute (NEI). The proposed procedures and methodology are described in the January 23,1996, letter from Alex Marion of NEl to Brian Sheron of the NRC and are supplemented in the October 15,1996, letter from Alex Marion of NEl to Brian Sheron of the NRC. Based on a review of data used originally to support the position that only the primary frequency was required for tests on new probes to verify that they met the voltage variability specification of 110 percent of the nominal response, the industry indicated that testing at only the primary frequency was not sufficient. The proposed approach specifies that the voltage responses from the primary frequency and mix frequency channels of new probes be within 110 percent of the nominal voltage responses when voltages are normalized to the 20 percent through-wall flaw values. The nominal voltage responses were established as the average voltages obtained from ASME standard drilled hole flaws for at least 10 production probes, in a letter from Brian Sheron of the NRC to David Modeen of NEl dated July 29,1997, the NRC indicated that this approach to Section 3.c.2 of to GL 95-05 to address probe variability is acceptable. Therefore, the licensee's proposal to follow the industry approach is acceptable.

Probe Wear Section 3.c.3 of Attachment 1 to GL 95-05 specifies guidance for probe wear. The licensee proposed to use an attemative to Section 3.c.3. The attemative approach, developed through NEl, specifies that if the probe does not satisfy the voltage variability criterion for wear of 115 percent limit before its replacement, all tubes which exhibited flaw signals with voltage responses measured at 75 percent or greater of the lower repair limit (i.e., 2 volts) must be reinspected with a bobbin probe satisfying the 115 percent wear standard criterion. The voltages from the reinspection should be used as the basis for tube repair. The NRC staff

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completed a review of the NEl proposed attemative method and concluded that the approach is l

acceptable as discussed in a letter from Brian Sheron of the NRC to Alex Marion of the Nuclear Energy Institute dated March 18,1996. The licensee's proposal to follow the industry approach to address probe wearis acceptable.

Special Consideration Section 1.b.1 of Attachment 1 to GL 95-05 states that the voltage-based repair criteria do not i

apply to TSP intersections where tubes with degradation may potentially collapse or deform as I

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' a result of the combined postulated loss-of-coolant accident and safe shutdown earthquake loadings (e.g., intersections near the wedge supports at the upper TSPs). The licensee stated that the voltage-based repair criteria would not be applied to tubes in this region. The inspection and dispositioning criteria that would be utilized are as follows:

The tubes in this region will initially be inspected with a bobbin probe and further inspected with an RPC probe if degradation is detected with the bobbin probe. All indications, confirmed with RPC to be crack-like, will be repaired. All non-crack-like indications will be repaired unless a qualified sizing technique verifies the indication is less than 40 percent through-wall (the technical specifications repair limit).

3.1.3 Structural and Leakage Integrity Assessments The staff guidance for the implementation of the voltage-based repair criteria focuses on maintaining tube structural integrity during the full range of normal, transient and postulated accident conditions with adequate allowance for eddy current test uncertainty and flaw growth projected to occur during the next operating cycle. Tube structural limits based on RG 1.121 criteria recommend maintaining a margin of safety of 1,43 against tube failure under postulated accident conditions and maintaining a margin of safety of 3 against burst during normal operation. Because GL 95-05 addresses tubes affected with ODSCC confined to within the i

thickness of the tube support plate during normal operation, the staff concluded that the structural constraint provided by the tube support plate ensures all tubes to which the voltage-based criteria applies will retain a margin of 3 with respect to burst under normal operating conditions. For a postulated main steam line break accident, however, the tube support plate may displace axially during steam generator blowdown such that the ODSCC affected portion of the tubing may no longer be fully constrained by the tube support plate. Accordingly, it is appropriate to consider the ODSCC affected regions of the tubes as free standing tubes for the purpose of assessing burst integrity under postulated main steam line break conditions.

In order to confirm the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the conditional burst probability and i

the total primary-to-secondary leak rate from an affected steam generator during a postulated

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main steam line break event. To complete GL 95-05 prescribed assessments, the licensee j

proposes to follow the methodology described in WCAP-14277, Revision 1, "SLB Leak Rate l

and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated December 1996. The staff finds the methodology in WCAP-14277, Revision 1, acceptable.

The licensee requested staff approval to use a voltage dependent probability of detection J

(POD)instead of the constant POD of 0.6 required by GL 95-05. The voltage dependent POD approach calculates the distribution of bobbin indications as a function of voltage at the beginning of the cycle. The staff is currently addressing such an approach generically through NEl. PG&E will be permitted to use a revised POD,in lieu of a constant value of 0.6,if and when a revised POD is approved by the NRC. Until that occurs, PG&E will have to use a constant value of 0.6.

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. GL 95-05 specifies that the structural and leakage integrity assessments should use the latest available data from destructive examinations of tubes removed from Westinghouse-designed steam generators. NRC staff has agreed with NEl on a protocol by which the industry will periodically update the ODSCC database used to perform GL 95-05-specified calculations. The protocol ensures that the latest available data from destructive examination of tubes is considered. The licensee stated that they will follow the protocol.

In addition, the Joseph M. Farley Nuclear Plant (FNP) recently identified pulled tube data which was determined to be "significant" per item 5 of the protocol discussed above. As a result, these data have to be incorporated into the database on an expedited schedule. The licensee indicated that leak and burst correlations, developed using data from the ODSCC database, have been updated to incorporate the FNP data, and these updated correlations will be used when the licensee implements the voltage-based repair criteria.

NRC Information Notice (lN) 97-79," Potential inconsistency in the Assessment of the Radiological Consequences of a Main Steam Line Break Associated with the implementation of Steam Generator Tube Voltage-Based Repair Criteria," states that a licensee implementing the voltage-based repair criteria had used two different temperature conditions when comparing the projected end-of-cycle tube leakage with the maximum allowable tube leakage. The same temperature conditions should have been used in the calculations. The IN also states that other licensees may have made similar mistakes. PG&E indicated that they were aware of this issue and that they intend to use identical temperature conditions when comparing the projected end-of-cycle tube leakage with the maximum allowable tube leakage.

3.1.3.1 Conditional Probability of Burst The licensee will use the methodology described in Revision 1 of WCAP-14277 for performing a probabilistic analysis to quantify the potential for steam generator tube ruptures given a main steam line break event. The results of the probabilistic analysis will be compared to a threshold value of 1x102per cycle in accordance with GL 95-05. This threshold value provides assurance that the probability of burst is acceptable considering the assumptions of the calculation and the results of the staffs generic risk assessment for steam generators contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety issues A-3 A-4 and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value indicates ODSCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation assumed and evaluated as acceptable in NUREG-0844. NRC staff concludes that the i

licensee's proposed methodology for calculating the conditional burst probability is consistent l

with the guidance in GL 95-05 and is acceptable.

l 3.1.3.2 Accident Leakage The licensee will use the methodology described in Revision 1 of WCAP-14277 for calculating the steam generator tube leakage from the faulted steam generator during a postulated main steam line break event. The model consists of two major components: (1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of

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leakage model); and (2) a model predicting leak rate as a function of voltage, given that leakage occurs (i.e., the conditional leak rate model). The staff concludes that the licensee's proposed methodology for calculating the tube leakage is consistent with the guldance in GL 95-05 and is acceptable.

3.1.3.3 Primary-to-Secondary Leakage During Normal Operation Because the voltage-based repair criteria would allow degraded tubes to remain in service, the degraded tubes may develop through-wall cracks during an operational cycle, thus creating the l

l potential for primary-to-secondary leakage during normal operation, transients, or postulated I

accidenta. Therefore, as a defense-in-depth measure, GL 95-05 specifies that the operational leakage limits of the plant technical specifications be limited to 150 gallons per day from any l

one steam generator. The staff concludes that adequate leakage integrity during normal l

operation is reasonably assured by the technical specification limits on allowable primary-to-secondary leakage. The proposed amendment includes a modification to TS 3.4.6.2.c.

Technical Specification 3.4.6.2.c. specifies an operational primary-to-secondary system leakage l

limit through the steam generators. Currently, this leakage limit applies when operating in l

Modes,1,2,3, and 4. The amendment proposes to limit the allowable primary-to-secondary l

leakage through each steam generator to 150 gallons-per-day (gpd), and changes the modes of applicability for the proposed steam generator leakage limit. The proposed modification states that, when operating in Modes 3 and 4, the licensee can assume that the leakage requirement of 150 gpd of primary-to-secondary leakage is met if the steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/ml for principal gamma emitters. The licensee justifies this proposed change in the Bases (Section 3/4.6.2 Operational Leakage) by stating that in Modes 3 and 4 the primary system radioactivity level may be very low, making it difficult to measure a primary-to-secondary leakage of 150 gpd. Therefore, if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/mi for principal gamma emitters, then the licensee will consider the 150 gpd leakage limit to be met. The licensee will still have to measure for primary-to-secondary leakage when in Modes 1 and 2. The staff is in agreement with this proposed change since the leakage limit is consistent with the guidance in GL 95-05, and because it will be very difficult to accurately determine the primary-to secondary leakage through measurements of the primary j

system radioactivity during Modes 3 and 4.

l The licensee proposes to add an additional surveillance requirement to Section 4.4.6.2.1 of the Technical Specifications to measure reactor coolant leakages. The proposed modification states that the licensee intends to make a determination of the primary-to-secondary leakage at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while operating in Modes 1,2,3, and 4. During normal operations, the licensee will use a correlation of the steam jet air ejector radiation monitors to determine if l

the leak rate is below the 150 gpd limit at the 72-hour frequency requirement. During periods of I

significant bource term and mass flow rate changes or when the primary system radioactivity levels may be low, errors in the primary-to-secondary leak rate calculations can be large.

During these conditions, the licensee may have to use engineering judgement to aid in determining leak rates. The staff finds the licensee's proposed surveillance requirement to measure the primary-to-secondary leakage at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to be acceptable.

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Additionally, GL 95-05 states that licensees should review their leakage monitoring measures to ensure that should a significant leak occur in service, the leak will be detected and the plant will be shut down in a timely manner to reduce the likelihood of a potential tube rupture. As stated in its February 26,1997 submittal, the licensee will provide an effective leakage monitoring program by following the guidance of EPRI TR-104788,"PWR Primary-to-Secondary Leak Guidelines," Research Project S 550, Final Report, May 1995. The staff finds the licensee's leakage monitoring program acceptable.

3.1.4 Degradation Monitoring To confirm the nature of the degradation occurring at the tube support plate elevations, tubes are periodically removed from the steam generators for destructive analysis. Tube pulls can confirm that the nature of the degradation being observed at these locations is predominantly axially oriented ODSCC, provide data for assessing the reliability of the inspection methods, and supplement the existing databases (e.g., burst pressure, probability of leakage, and leak rate). GL 95-05 contains guidance that states licensees should remove at least two pulled tube specimens with the objective of retrieving as many intersections as practical (a minimum of four intersections) during the plant steam generator inspection outage that implements the voltage-based repair criteria or during an inspection outage preceding initial application of the voltage-based criteria. On an ongoing basis, additional tube specimen removals (minimum of two intersections) should be obtained at the first refueling outage following 34 effective full power months of operation or at the maximum interval of three refueling outages after the previous tube pull. Altematively, the licensee may participate in an industry-sponsored tube pull program endorsed by the NRC as described in GL 95-05.

The licensee removed five intersections (four tubes) during the Cycle 7 outage from Unit 1 steam generators that contained axial ODSCC. None of the OD indications were detected by field bobbin or RPC inspection. But, destructive examination confirmed that the ODSCC indications were predominantly axially oriented and confined within the tube support plates as required by GL 95-05.

The licensee plans to remove at least two pulled tube specimens (minimum of four intersections) during the plant steam generator inspection outage that implements the voltage-based repair criteria for both Units 1 and 2. The licensee also stated that it will comply with GL 95-05 for future tube removals.. The staff concludes that the licensee satisfies the tube removal guidance of GL 95-05, and therefore the tube removal program is acceptable.

3.2 Radiological Consequences Evaluation The licensee proposes to change their TS to implement a voltage-based attemate SG tube plugging repair criteria per the requirements of NRC Generic Letter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." In their limnse amendment submittal dated February 26,1997, the licensee requested that the specific activity limits of dose equivalent *l in the primary coolant be established at 1.0 pCl/g for the 48-hour limit and at 60 pCi/g for the maximum instantaneous limit (in accordance with Generic Letter 95-05). The allowable activity level of dose equivalent l

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'8'l in the secondary coolant was assumed to be equal to the TS limit of 0.1 pCi/g. This license amendment also requestvf that Diablo Canyon be approved to operate based upon a 12.8 gpm primary to secondary leak initiated by an accident in the faulted steam generator and the TS allowable value for primary to secondary leakage from each of the three intact steam generators of 150 gpd per steam generator. As part of this amendment request, the licensee performed an assessment of the radiological dose consequences of a main steam line break accident. The licensee found the radiological dose consequences of incorporating these proposed changes to be acceptable based on the NRC acceptance criteria for doses at the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ), and the control room.

The staff reviewed the licensee's calculations and performed confirmatory calculations to check the acceptability of the licensee's methodology and resulting doses. As part of the stafs review, the staff calculated the doses resulting from a main steam line break accident using the methodology associated with Standard Review Plan (SRP) 15.1.5, Appendix A (NUREG-0800).

The staff performed two separate assessments. One was based upon a pre-existing iodine spike activity level of 60 pCi/g of dose equivalent '8'l in the primary coolant and the other was based upon an accident initiated iodine spike. For the accident initiated spike case, the staff assumed that the primary coolant activity level was 1.0 pCi/g of dose equivalent '8'l. The accident initiated an increase in the release rate of iodine from the fuel by a factor of 500 over the normal release rate to maintain an activity level of 1.0 pCi/g of dose equivalent '8'l in the primary coolant. For these two cases, the staff calculated the thyroid doses for individuals located at the EAB and at the LPZ. The staff also calculated the thyroid dose to the control room operator. The parameters which were utilized in the stafs assessment are presented in Table 1. For the control room makeup and recirculation flow rates, the staff used the licensee's flow rate value less ten percent, as allowed by the TS. The values used for the efficiencies of the control room ventilation filters were slightly reduced (below the licensee's numbers) to account for one percent bypass flow. The EAB, LPZ, and control room doses calculated by the staff are presented in Table 2.

i The stafs calculations confirmed that the doses from a postulated main steam line break i

accident meet the acceptance criteria and that the licensee's calculations are acceptable. The l

results of both the licensee's and stars calculations showed that the thyroid doses at the EAB and LPZ would be less than the guidelines established by SRP 15.1.5, Appendix A (acceptance criterion of 300 rem thyroid dose at the EAB and LPZ for the pre-existing spike case and 30 rem thyroid dose at the EAB and LPZ for the accident initiated spike case). The control room operator thyroid dose would be less than the guidelines of SRP 6.4 (acceptance criterion of 30 rem thyroid to the control room operator). On this basis, the staff approves the' licensee's request to implement a voltage-based repair criteria for the steam generator tube support plate intersections at Diablo Canyon. Use of this voltage-based repair criteria will permit the licensee to maintain specific activity limits of dose equivalent '8'l in the primary coolant of 1.0 pCi/g for the 48-hour limit and 60 pCi/g for the maximum instantaneous limit, as well as restricting the allowable maximum primary to secondary coolant leakage to 12.8 gpm.

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. 3.3 Loading Evaluation During a combined postulated loss-of-coolant accident (LOCA) plus safe shutdown earthquake (SSE) loading condition, the potential exists for yielding of the tube support plates (TSP) in the vicinity of the wedges at the periphery of the TSPs, followed by collapse of deformed tubes and subsequent loss of flow area. In addition to tubes that may collapse following a combined SSE plus LOCA event, there may be a number of tubes that will undergo a limited amount of permanent deformation. This deformation may also lead to loss of flow area and consequent increase in peak clad temperature.

The licensee will not apply voltage-based repair criteria to tube-to-tube support-plate intersections where the steam generator tubes with degradation may potentially collapse or deform as a result of the combined postulated LOCA and SSE loadings. The specific tubes that are susceptible to collapse and in-leakage during a postulated LOCA plus SSE event have been identified to the licensee by Westinghouse. A maximum of 254 tubes per SG could be effected and these tubes are located near the wedge supports at each TSP.

A discussion of the most critical loads, load combinations and Westinghouse's analytical methodology to determine tubes susceptible to deformation / collapse during a postulated LOCA plus SSE event is provided in the following paragraphs.

LOCA Rarefaction Wave loads The principal tube loading during a postulated LOCA is caused by the rarefaction wave in the primary fluid. This wave initiates due to sudden rupture at the postulated break location and travels around the tube U-bends. A differential pressure is created across the two legs of the tube which causes an in-plane horizontal motion of the U-bend. This differential pressure, in tum, induces significant lateral loads on the tubes.

The pressure-time histories which are part of the input data for the structural analysis were obtained from transient thermal-hydraulic (T/H) analyses using the MULTIFLEX computer code.

A break opening time of 1.0 msee to full flow area (that is, instantaneous double-ended rupture) is assumed to obtain conservative hydraulic loads. Pressure time histories were determined for three tube radil, identified as the minimum, medium, and maximum radius tubes. For the structural evaluation, the pressures of concem occur at the hot and cold leg U-bend tangent points. As the tube is loaded, it moves laterally and rotates within the TSP. After a finite amount of rotation, the tube becomes wedged within the TSP and will no longer be able to l

rotate. The second set of boundary conditions, therefore, considers the tube to be fixed at the top TSP location, and is referred to as the " fixed" case. Continued tube loading causes the tube to yield in bending at the top TSP and eventually a plastic hinge develops. This represents the third set of boundary conditions, and is referred to as the " pinned" case. The licensee has considered all three sets of boundary conditions in the analysis for the Diablo Canyon SGs.

The hot to cold leg AP resulting from a LOCA pipe break is strongly dependent on the representation of the divider plate located in the inlet chamber of the steam generator that separates the primary inlet and outlet flows. When LOCA rarefaction analyses were first

4 performed, the divider plate was modeled as a rigid structural member. Later analyses accounted for the divider plate flexibility and a significant reduction in the resulting hot to cold leg AP. The divider plate flexibility is included in the present analysis. Based on its review as discussed above, the staff finds the licensee's methodology for determining the loading due to the LOCA rarefaction wave acceptable.

LOCA Shakina Loads Concurrent with the rarefaction wave loading during a LOCA, the tube bundle is subjected to additional bending loads due to the shaking of the SG caused by the break hydraulics and reactor coolant loop motion. However, the resulting tube stresses from this motion have been determined to be small compared to those due to the rarefaction wave induced motion.

To obtain the LOCA induced hydraulic forcing functions, a dynamic blowdown analysis was performed to obtain the system hydraulic forcing functions assuming an instantaneous (1.0 msec break opening time) double-ended guillotine break. The hydraulic forcing functions were applied, along with the displacement time-history of the reactor pressure vessel, to a system structural model, which includes the SG, the reactor coolant pump and the primary piping. This analysis yields the time history displacements of the SG at its upper lateral and lower support nodes. These time-history displacements formulate the forcing functions for obtaining the tube stresses due to LOCA shaking of the SG.

To avaluate the SG response to LOCA shaking loads, the computer code WECAN was used.

Input to the WECAN model was in the form of acceleration time histories at the tube /tubesheet interface. These accelerations were obtained by differentiation of the system model displacement time histories at this location. Accolaration time histories for all six degrees of freedom were used. Based on its review, as discussed above, the staff finds the licensee's methodology for determining LOCA shaking loads acceptable.

Seismic Analysis The SSE loads are developed as a result of the motion of the ground during a design basis earthquake. A nonlinear time-history analysis was used to account for the effects of radial gaps between the secondary shell and the TSPs, and between the wrapper and shell. The seismic l

excitation defined for the SGs was in the form of acceleration response spectra at the SG l

supports, in order to perform the nonlinear time history analysis, the response spectrum input was converted into acceleration time history input. Acceleration time-histories for the nonlinear analysis were synthesized from El Centro Earthquake motions, using a frequency l

suppression / raising technique, such that the resulting time history spectra closely enveloped l

the corresponding specified spectra. The three orthogonal components of the earthquake were applied simultaneously to perform the analysis.

l The mathematical model for the seismic analysis consisted of three dimensional lumped mass beam and pipe elements as well as a general matrix input to provide a plant-specific representation of the SG and reactor coolant pipe stiffness. In the nonlinear analysis, performed using the WECAN computer program, the TSP /shell, and wrapper /shell interactions

e

- were represented by a concentric spring-gap dynamic element, using impact damping to account for energy dissipation at these locations.

The tube bundle straight leg region on both the hot-leg side and cold-leg side was modeled as two equivalent beams. The U-bend region, however, was modeled as five equivalent tubes of different bend radil, each equivalent tube representing a group of SG tubes. In addition, a single tube representing the outermost tube row was also included in the model. Continuity between the straight leg and U-bend tubes, as well as between the U-bend tubes themselves, was accomplished through appropriate nodal couplings. Typically, the tubes were coupled to the TSP for transnational degree of freedom. Packed intersections, which exist at Diablo Canyon, are not expected to significantly affect the resulting TSP loads. Based on its review, as discussed above, the staff finds the analytical methodology for the seismic analysis ecceptable.

Combined Tube Sunoort Plate Loads

)

The combined TSP loads consist of appropriate combinations of LOCA rarefaction, shaking and l

seismic loads. The LOCA rarefaction and shaking loads were combined directly to obtain overall LOCA loads. These LOCA loads were combined with the SSE loads using the square l

root of the sum of the squares. The LOCA plus SSE loads were transferred from the TSP to the SG shell through wedges (also referred to as wedge groups) located at discrete locations l

around the plate circumference. For the Series 51 SGs, there are six wedge groups located every 60* around the plate circumference. Except for the bottom TSP, the wedge groups for each of the TSPs are located at the same angular location as for the top TSP. Thus, if TSP deformation occurs at the lower plates, the same tubes are affected as for the top TSP. For the top TSP, however, the wedge groups have a 10-inch width, compared to a 6-inch width for the other plates. This larger wedge group width distributes the load over a larger portion of the TSP, resulting in less plate and tube deformation for a given load level. For the bottom TSP, the wedge group width is 6 inches, and the wedge groups were rotated 36' relative to the other TSPs. The distribution of the total TSP load among the various wedge groups results in a maximum wedge load of 0.634 of the total TSP load. For seismic loads, which can have a random orientation, the maximum wedge load was determined to be 0.667 of the maximum TSP load. For this analysis, however, it is conservatively assumed that the factor of 0.667 of the total TSP load applies to both the LOCA and seismic loads. The staff finds the licensee's load combination methodology acceptable.

Determination of Flow Area Reduction in estimating the flow area reduction, one of the key parameters is the force / deflection characteristic of the TSP. The licensee used data for the Model D plates as a basis, and extrapolated those values for use in the current analysis.

The Model D crush test force / deflection results represent inelastic behavior of the TSPs and tubes. In order to make use of this data, an approximation was made between the elastic analyses that determine the TSP loads and the inelastic crush test. This approximation was based on the area under the force / deflection curve for the crush test versus the area j

4

- corresponding to the elastic response. Comparing the areas from the two sets of calculations showed that the elastic areas are greater than the inelastic areas for both the seismic and the combined LOCA plus SSE loads for the Diablo Canyon SGs.

Since it is estimated that the LOCA plus SSE plate loads exceed the Model D crush test plate loads, a factor of 3.0 was applied to the number of collapsed tubes at each wedge location.

This resulted in a total of 42 tubes that were considered collapsed at each of the six wedge locations. Therefore, the total number of affected tubes was determined to be 252 and, out of a total of 3388 tubes, the reduction in flow area was 7.5 percent. The staff finds the methodology for determining the affected tubes and flow area reduction acceptable.

l j

Identification of Potentially Susceptible Tubes l

Identification of the potentially susceptible tubes is based on crush test results performed for Series 51 SGs subsequent to the calculations to establish the flow area reduction. Wedge group orientations typical of Series 51 SGs were considered in the tests. In performing each of l

the crush tests, the plate samples were loaded to the point of load instability. Evaluation of the tube deformations experienced in each of these tests showed that the level of tube deformation I

(diameter reduction) does not exceed the diameter change that would result in tube collapse l

under the post-LOCA secondary to primary AP. Thus, the test results show general tube i

deformation trends, but do not provide specific tubes that will potentially collapse at any given load level. As such, it was not possible to identify exactly the 42 tubes that might be limiting at l

each wedge group.

l l

To account for the uncertainty in selecting the susceptible tube locations, an enveloping group i

of tubes were selected at each of the six wedge locations, resulting in more than 42 tubes identified at each wedge group as being limiting. As a result, a total of 468 tubes per SG have been conservatively included in the wedge region exclusion zone and will be excluded from application of voltage-based repair criteria, even though a maximum of 252 of these tubes can potentially collapse and cause in-leakage following a LOCA plus SSE. Tabular summaries of the potentially susceptible tubes, showing tube row and column numbers and wedge locations have been provided.

I Based on its review, as discussed above, the staff finds the licensee's analytical evaluation, and methodology for the identification of the SG tubes to be excluded from the application of voltage-based repair criteria due to pstential tube collapse during a postulated combined LOCA plus SSE event acceptable.

3.4 Proposed TS Changes

in order to incorporate a voltage-based steam generator repair criteria, the licensee has proposed the following changes to the TS.

O

=.

. 1.

Proposed New TS 4.4.5.2b.4).

" Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages."

l 2.

Proposed New TS 4.4.5.2d.

" Implementation of the steam generator tube / tube support plate repair criteria requires a 100% bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their length."

l 3.

Proposed Changes to TS 4.4.Sa.6).

I An exemption is added for the voltage-based repair criteria. "This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4a.10) for the repair limit applicable to these intersections."

4.

Proposed New TS 4.4.5a.10).

This new section provides the detailed requirements for the voltage-based repair criteria.

5.

Proposed Changes to TS 4.4.5.4be The change removes the statement "and all tubes containing through-wall cracks."

6.

Proposed New TS 4.4.5.5d.

This new section provides the detailed reporting requirements for the voltage-based repair criteria.

7.

Proposed Changes to Bases The Bases for TS 3/4.4.5 and 3/4.4.6.2 are revised to incorporate the voltage-based repair criteria.

The staff has reviewed the TS changes discussed above and finds that they consistently incorporate the voltage-based repair criteria in accordance with the requirements of NRC GL 95-05 as previously discussed in this safety evaluation. Therefore, the proposed changes are acceptable.

l 3.5 Summary The licensee submitted an application for a license amendment to permit the use of the voltage-based repair criteria for steam generator tubes at Diablo Canyon Units 1 and 2. The staff has

s e

' reviewed the proposed amendment and concludes that the proposed altemate repair criteria are consistent with GL 95-05 and are acceptable. Conceming the use of a voltage dependent probability of detection for application in steam generator voltage-based attemate repair criteria, the staff is currently addressing such an approach generically through the Nuclear Energy Institute. PG&E will be permitted to use a revised POD,in lieu of a constant value of 0.6, if and when a revised POD is approved by the NRC. Until that occurs, PG&E will have to use a constant value of 0.6. The staff also concludes that adequate structural and leakage. integrity can be assured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied. The staff approves the proposed voltage-based repair criteria based, in part, on the licensee being able to successfully demonstrate after each inspection outage that the conditional probability of burst and the primary-to-secondary leakage during a postulated main steam line break will be acceptable per the guidance in GL 95-05.

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4.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Califomia State official was notified of the l

proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 17239). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(g). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Attachments: 1. Table 1

2. Table 2 Principal Contributor: C. Beardslee l

C. Hinson

(

J. Rajan Date:

March 12, 1998

e TABLE 1 INPUT PARAMETERS FOR DIABLO CANYON UNITS 1 AND 2 EVALUATION OF MAIN STEAMLINE BREAK ACCIDENT 1.

Primary Coolant Concentration of 60 pCi/g of Dose Equivalent "'I Pre-existina Soike Value (uCi/a)

• l = 46.77 n2l = $ $,93

• i = 65.63

• l = 8.18 "51 = 34.77 2.

Data on Primary Coolant and Secondary Coolant Primary Coolant Volume (ft')

12,566 Primary Coolant Volume (gal) 94,000 Primary Coolant Temperature (*F) 576 Secondary Coolant Liquid Mass (pounds /SG) 81,500 Secondary Coolant Steam Mass (pounds /SG) 7,200 Secondary Coolant Operating Temperature (*F) 519 3.

TS Limits for DE "'l in the Primary and Secondary Coolant Maximum Instantaneous DE "'I Concentration (pCi/g) 60.0 Primary Coolant DE *l Concentration (pCi/g) 1.0 Secondary Coolant DE *1 Concentration (pCi/g) 0.1 4.

TS Value for the Primary to Secondary Leak Rate Primary to secondary leak rate, maximum any SG (gpd) 150 Primary to secondary leak rate, total all 4 SGs (gpd) 600 5.

Maximum Primary to Secondary Leak Rate to the Faulted and Intact SGs Faulted SG (gpm) 12.8 Intact SGs (gpm/SG) 0.1 6.

lodine Partition Factor j

Faulted SG 1.0 Intact SG 1.0 l

7.

Steam Released to the Environment (lbs) i Faulted SG (0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 162,784 l

Faulted SG (2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) 0

e intact SGs (0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 393,464 intact SGs (2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) 860,461 8.

Letdown Flow Rate (gpm) 75 9.

Release Rate for 1.0 pCi/g of Dose Equivalent '8'l Release Rate (Ci/hri 500X Release Rate (Ci/hr) l

'8'l =

10.3 5,160 1321 =

17.8 8,900

'8 i =

22.9 11,500

• l =

29.33 14,700

'851 =

22.8 11,400

10. Atr,.ospheric Dispersion Factors sec/m8 EAB (0-2 hours)

• 5.3 x 10" LPZ (0-8 hours) 2.2 x 104 Control Room (0-8 hours)(pressurized) 7.05 x 104 (0-8 hours)(infiltration) 1.96 x 10d

11. Control Room Parameters Filter Efficiency (%)

Air recirculation filter 95 (Value used by staff) 94.05 8

Volume (ft )

170,000 Makeup flow (cfm) 1,890 Recirculation Flow (cfm) 1,890 Unfiltered Inleakage (cfm) 10 Occupancy Factors 0-1 day 1.0 1-4 days 0.6 4-30 days 0.4

• NRC staff calculated values

=

i 9

t l

Table 2 - THYROID DOSES FROM DIABLO CANYON UNITS 1 AND 2 MAIN STEAM LINE BREAK ACCIDENT (REM)

(VALUES CALCULATED BY NRC STAFF)

DOSE LOCATION Pre-Existing Spike Accident Initiated Spike **

EAB 50.3*

27.2 LPZ 7.8*

15.6 Control Room ~

7 x 10-2 1.45 x 10 4

• Acceptance Criterion = 300 rem thyroid

• • Acceptance Criterion = 30 rem thyroid i

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