ML022540362

Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Vessel Head Penetration Nozzle Inspection Programs

ML022540362
Person / Time
Site:	Farley
Issue date:	09/09/2002
From:	Beasley J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BL-02-002, NEL-02-0186
Download: ML022540362 (12)

Text

Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205 992 7110 Fax 205 992 0403 September 9, 2002 Docket No.

50-364 SOUTHERNNM COMPANY Energy to Serve Your World' NEL-02-0186 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Vessel Head Penetration Nozzle Inspection Programs Ladies and Gentlemen:

In accordance with 10 CFR 50.54(f) and pursuant to the requirements of Nuclear Regulatory Commission (NRC)Bulletin 2002-02, "Reactor Pressure Vessel Head Degradation and Vessel Head Penetration Nozzle Inspection Programs," dated August 9, 2002, Southern Nuclear Operating Company (SNC) hereby submits the enclosed information which serves as the 30-day response required by the bulletin to provide information regarding SNC's planned inspection program for the Farley Nuclear Plant (FNP) Unit 2 reactor vessel head and penetration nozzles.

This letter contains a new SNC commitment to perform penetrant testing of nozzle attachment welds as explained in the enclosure. If you have any questions, please advise.

Mr. J. B. Beasley, Jr. states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, Southern Nuclear Operating Company IqI wroasley co Sworn to and subscribed before mne this ______a Of S u'

,2002.

S"t&)

Notary Public

_ My comimission expires:

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005 J. Barnie Beasley, Jr., P.E.

Vice President

Page 2 U. S. Nuclear Regulatory Commission DWD/sdl: Bulletin 2002-02 30-day U2 response.doc Enclosure cc:

Southern Nuclear Operating Company Mr. D. E. Grissette, Nuclear Plant General Manager - Farley U. S. Nuclear Regulatory Commission, Washington, D. C.

Mr. F. Rinaldi, NRR Project Manager - Farley U. S. Nuclear Regulatory Commission, Region II Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity Provided below for Farley Nuclear Plant (FNP) Unit 2 is the Southern Nuclear Operating Company (SNC) 30-day response to the information request contained in Nuclear Regulatory Commission (NRC)

Bulletin 2002-02, "Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs," dated August 9, 2002. The pertinent portion of the Bulletin is repeated in bold text below.

Note that while a separate 15-day response letter was submitted for Unit 1 on August 23, 2002, some Unit I information is included in this response to aid in explanation of the FNP inspection program.

(1) Within 30 days of the date of this bulletin:

A. PWR addressees who plan to supplement their inspection programs with non-visual NDE methods are requested to provide a summary discussion of the supplemental inspections to be implemented. The summary discussion should include EDY, methods, scope, coverage, frequencies, qualification requirements, and acceptance criteria.

SNC Response to NRC Question L.A SNC has evaluated the current status of FNP Unit 2 with regard to accrued Effective Full Power Years (EFPY) and Effective Degradation Years (EDY), calculated in accordance with Equation 2.2 in Electric Power Research Institute (EPRI) Material Reliability Program (MRP) document MRP-48 (Ref. 12). As of the upcoming refueling outage set to begin September 14, 2002, Unit 2 will be at 15.6 EDY.

The SNC responses to Bulletin 2002-01 (in letters dated March 29, May 16 and June 11, 2002) addressed the adequacy of visual inspection of the reactor pressure vessel (RPV) head and vessel head penetration (VHP) nozzles for compliance with the design and licensing basis of both FNP units. Those responses are still applicable. Additional technical justification for the adequacy of the inspections is provided in this response to Bulletin 2002-02.

SNC previously committed in a letter dated August 31, 2001 to perform a 100 percent bare metal visual (BMV) inspection of Units 1 and 2. The inspection of Unit 1 was completed in November 2001 during the 17 th refueling outage as reported in a letter dated December 3, 2001. The inspection of Unit 2 is scheduled for September 2002 during the 15th refueling outage as reported in a letter dated March 29, 2002.

The planned Unit 2 BMV inspection will be similar to the previous Unit 1 inspection. In the Unit 1 inspection, Level III certified VT-2 visual examination personnel viewed all the examinations performed.

The examiners were cognizant of boric acid corrosion problems experienced at other utilities and reviewed examples of boric acid corrosion provided in EPRI Report 1006296, "Visual Examination for Leakage of PWR Reactor Head Penetrations," issued in August 2001 and since updated. The examiners specifically looked for boric acid residue in the area around the annulus between the penetration nozzle and the penetration hole in the RPV closure head. Boric acid in this area could be indicative of a through wall leak in the Inconel penetration nozzle or the attachment weld on the underside of the RPV closure head. Any boron on the RPV head was recorded for evaluation.

I

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity The general condition of the Unit I RPV head as noted in the November 2001 inspection was very good, with some slight debris and/or foreign objects noted. There was no apparent evidence of boric acid residue from active leakage in the vicinity of any of the head penetrations at the interface between the RPV closure head and the penetration nozzles. Several locations had small white spots on the penetration nozzle above the nozzle-to-RPV-head interface, a condition that is not unusual for locations that have been vented or disassembled in the past. After performance of the Unit 2 BMV inspection the RPV head and penetration nozzles will be cleaned to remove boric acid residue to the extent practicable.

A full 3600 visual examination of all 69 control rod drive mechanism (CRDM) nozzles and the head vent nozzle (70 VHP nozzles total) was achieved during the Unit 1 BMV inspection by means of remotely operated video equipment. The Unit 2 BMV inspection will use similar equipment; therefore, complete visual coverage of all 70 VHP nozzles is expected. The acceptance criterion for the BMV inspection is no evidence of leakage originating from the annulus between the RPV head and each VHP nozzle.

The MRP Inspection Plan has been developed, reviewed, and approved by the PWR utilities (Refs 13 and 14). It presents a technically credible inspection regimen that assures to a high degree of certainty that leaks will be detected at an early stage long before wastage or circumferential cracking can challenge the structural integrity of the reactor coolant system (RCS) pressure boundary. Furthermore, implementation of the MRP Inspection Plan will assure continued compliance with the Regulatory Requirements cited within NRC Bulletin 2002-02.

Therefore, SNC will implement the MRP Inspection Plan and will comply with its requirements beginning with the 100 percent BMV inspection of Unit 2 in September 2002. The MRP Inspection Plan is consistent with the inspection commitments SNC made in its responses to Bulletin 2001-01 and Bulletin 2002-01, and is more rigorous.

Summary Discussion of Supplemental Inspections Due to the seriousness of the RPV head integrity issue, SNC has elected to perform supplemental ultrasonic testing (UT) of the Unit 2 VHP nozzles in addition to the BMV inspection. SNC committed to this supplemental inspection in a March 29, 2002 letter responding to Bulletin 2002-01.

The Unit 2 UT examination will be performed on all 69 of the control rod drive mechanism (CRDM) nozzles and the head vent nozzle. The UT techniques used will cover the entire wall thickness of the penetration nozzles from the ID surface for detection of pressurized water stress corrosion cracking (PWSCC). Inspection coverage will extend 360' around each nozzle and approximately 11 inches from the bottom of each nozzle to the upper travel limit of the UT probe. At all nozzles the UT probe will reach up past the attachment weld (J-groove weld) and the heat affected zone above it. However, due to the particular geometry of each nozzle with respect to the RPV head, the highest level the probe can reach is expected to vary from approximately 0 to 5 inches below the highest level of the upper head surface.

Also, the bottom 1.5 inches of each CRDM nozzle is threaded on the OD, which will create undesirable geometric reflectors in the UT data. An additional potential inspection limitation is created by three tabs which center the thermal sleeve suspended within each CRDM nozzle. In some nozzles the centering tabs may restrict the height the UT probe can reach. None of these anticipated inspection limitations are significant since the region of highest stress (and thus of highest susceptibility to PWSCC) is closely associated with the attachment weld, but any limitations in the final inspection coverage achieved will be 2

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity noted when the inspection results are reported.

The UT examination has been demonstrated to detect both axial and circumferential indications initiating from the ID or OD surface of the nozzle tube material. In addition, UT examination of the interference fit region above the nozzle attachment weld will be performed to determine if a potential reactor coolant leak has occurred, either through the nozzle tube material or through the attachment weld. Such a leak can corrode the interference fit region and create a recognizable "leak path" UT indication. The leak path UT detection technique has been developed by Framatome ANP, the vendor contracted for the FNP Unit 2 inspection, based on experience with approximately 270 nozzle UT examinations to date. In March 2001 Framatome ANP began consistently scanning the nozzle interference fit region during UT examinations.

In all subsequent UT examinations performed on CRDM nozzles with known leakage, indication of a leak path through the interference fit region was observed which corresponded to the known leakage. Since the UT exam will detect circumferential cracks in the tube, the concern for penetration ejection from crack propagation in the nozzle tube material is effectively addressed. The UT leak path exam provides an additional confirmation of the visual results and also addresses the concern of potential wastage resulting from a leak. A complete UT examination for detection of axial and circumferential flaws combined with UT leak path detection therefore addresses both the concern for wastage resulting from leakage and the potential for a nozzle ejection due to a circumferential crack above the weld.

Personnel performing the UT examination will be required to have current Level II or III NDE certification and additional training in VHP nozzle flaw evaluation. EPRI has conducted MRP mockup blind sample examination demonstrations with Framatome ANP, and the UT techniques that will be used at FNP have been evaluated and accepted by means of these demonstrations. Only procedure techniques have been demonstrated at this time; personnel qualification testing similar to ASME Section XI, Appendix VIII is not yet available.

Dye penetrant (PT) examinations will be performed on the VHP nozzle attachment welds as required to help resolve discrepancies between UT cracking indications and leakage indications. Thus a PT of the attachment weld will be done if there is visual or LIT indication of suspected leakage but little or no corresponding UT indication of cracking in the nozzle tube material. A PT examination will also be performed on the attachment weld of any nozzle for which a UT leak path exam and BMV exam cannot be performed, although no access restrictions resulting in such cases are anticipated as explained above.

The acceptance criteria for the UT examination will be determined based on the length, depth and location of each identified indication. It is anticipated that flaws will be removed unless evaluated to be acceptable. For flaws that are evaluated, the approach will be to locate and size the flaw, apply the growth rate identified in MRP-55 (Ref 5) to the next inspection interval and evaluate using ASME flaw tolerance methods and acceptance criteria as modified by the NRC recommendation letter of November 21, 2001 (Ref 11).

SNC plans to work with the EPRI MRP and NRC to determine the frequency for future supplemental inspections.

3

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity Response to the Six Concerns of Bulletin 2002-02 With regard to the planned 100 percent BMV inspection, SNC offers the following responses as justification for continued reliance on visual examinations as the primary method. Included in these responses is a discussion on the reliability and effectiveness of visual examinations as it relates to the six concerns expressed in Bulletin 2002-02 and the basis for concluding that unacceptable wastage will not occur between refueling outages.

Concern 1:

Circumferential cracking of CRDM nozzles was identified by the presence of relatively small amounts of boric acid deposits. This finding increases the need for more effective visual and non-visual NDE inspection methods to detect the presence of degradation in CRDM nozzles before nozzle integrity is compromised.

Response 1:

Since the initial discovery of circumferential cracks above the J-groove weld in 2001, visual inspection techniques and approaches employed have been dramatically improved and a heightened sense of awareness exists for the range in size and appearance of visual indications that must be further investigated. Non-visual techniques similarly have and continue to evolve to more effectively examine the penetration tube and associated welds for evidence of cracks. Nothing in the recent events at Davis Besse has altered the fundamental inspection capability requirements previously established as necessary to identify the presence of PWSCC and subsequent associated wastage. The effectiveness of inspection techniques continues to be evaluated and improved.

The EPRI MRP has published detailed guidance for performing visual examinations of reactor pressure vessel (RPV) heads (Ref 1). A utility workshop was recently conducted to discuss this guidance and lessons learned from recent field experience (including Davis-Besse). RPV head bare metal visual inspections at FNP Units 1 and 2 are/will be performed and documented in accordance with written procedures and acceptance criteria that comply with the guidance of the MRP Inspection Plan.

Evaluations and corrective actions will be rigorous and thoroughly documented.

In order for outside diameter (OD) circumferential cracks above the J-groove weld to initiate and grow, a leak path must first be established to the control rod drive mechanism (CRDM) annulus region from the inner wetted surface of the RPV head. If primary water does not leak to the annulus, the environment does not exist to cause circumferential OD cracking. Axial cracks in the CRDM nozzles or cracks in J groove welds must first initiate and grow through-wall. Experience has shown that through-wall axial cracks will result in observable leakage at the base of the penetration on the outer surface of the vessel, even with interference fits. Alloy 600 steam generator drain pipes at Shearon Harris (1988) and pressurizer instrument nozzles at Nogent 1 and Cattenom 2 (1989) were all roll expanded but still developed leaks during operation (Ref 2). Plant specific top head gap analyses have been performed for a large number of plants, with nozzle initial interference fits ranging from 0 to 0.0034". These analyses have confirmed the presence of a physical leak path in essentially all nozzles under normal operating pressure and temperature conditions (Ref 2).

The probability of detecting small CRDM leaks by visual inspections alone is high. Visual inspections of the reactor coolant system pressure boundary have been proven to be an effective method for identifying 4

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Inte2rit leakage from primary water stress corrosion cracking (PWSCC) cracks in Alloy 600 base metal and Alloy 82/182 weld metal. Specifically, visual inspections have detected leaks in RPV head CRDM nozzles, RPV head thermocouple nozzles, pressurizer heater sleeves, pressurizer instrument nozzles, hot leg instrument nozzles, steam generator drain lines, a RPV hot leg nozzle weld, a power operated relief valve (PORV) safe end and a pressurizer manway diaphragm plate (Ref 3). To date, no leaking CRDM nozzles have been discovered by non-visual NDE examinations except for the three nozzles at Davis-Besse where leakage would have been detected visually had there been good access for visual inspections and the head cleaned of pre-existing boric acid deposits from other sources (Ref 2).

Finally, as described under Concern 3 below, detailed probabilistic fracture mechanics (PFM) analyses have been performed to demonstrate the effectiveness of visual inspections in protecting the CRDM nozzles against failure due to circumferential cracking (Ref 4). Even though the above discussion illustrates that visual inspections performed in accordance with MRP recommendations have a high probability of detecting through-wall leakage, a very low probability of detection was assumed in the PFM analyses. The PFM analyses assume only a 60% probability that leakage will be detected if a CRDM nozzle is leaking at the time a visual inspection is performed. Furthermore, if a nozzle has been inspected previously, and leakage was missed, subsequent visual inspections are assumed to have only a 12% probability of detecting the leak. Even with these conservative probabilities of detection assumptions, the PFM analyses show that visual inspection every outage reduces the probability of a nozzle ejection to an acceptable level for plants with 18 or more EDY. Visual inspections of plants with fewer than 18 EDY in accordance with the MRP Inspection Plan will maintain the probability of nozzle ejection for these plants more than an order of magnitude lower than that for the greater than 18 EDY plants.

In summary, the industry has responded to the need to detect small amounts of leakage by increased visual inspection sensitivity, increased inspection frequencies, and improved inspection capabilities.

Small amounts of leakage can be detected visually and it has been shown that timely detection by visual examination will ensure the structural integrity of the RPV head penetrations with respect to circumferential cracking.

Concern 2:

Cracking of 82/182 weld metal has been identified in CRDM nozzle J-groove welds for the first time and can precede cracking of the base metal. This finding raises concerns because examination of weld metal material is more difficult than base metal.

Response 2:

Cracks in the J-groove weld do not pose an increased risk regarding nozzle ejection as compared to penetration base metal cracks. J-groove weld cracks that initiate and grow through-wall will leak the same as cracks in the penetration base metal. Therefore, weld cracks pose a similar risk as cracks in the base material and are equally detectable by visual examination. Although higher crack growth rates have been observed in laboratory testing of weld metal, the industry model of time-to-leakage includes plants that have had weld metal cracking as well as base metal cracking. The visual examination frequencies from the MRP Inspection Plan have been conservatively established based on the risk informed analyses considering leakage due to both weld metal and base metal cracking.

5

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity Through-wall circumferential cracking from the outside diameter of the CRDM nozzle has been identified for the first time. This raises concerns about the potential for failure of CRDM nozzles and control rod ejection, causing a LOCA.

Response 3:

Probability Fracture Mechanics (PFM) analyses using a Monte-Carlo simulation algorithm were performed to estimate the probability of nozzle failure and control rod ejection due to through-wall circumferential cracking (Ref 4). The PFM analyses conservatively assume that, once a leak path has extended to the annulus region, an OD circumferential crack develops instantaneously, with a length encompassing 30' of the nozzle circumference. Fracture mechanics crack growth calculations are then performed for this initially assumed crack, using material crack growth rate data from EPRI Report MRP 55 (Ref 5). The parameters used in the PFM model were benchmarked against the most severe cracking found to date in the industry (B&W Plants) and produced results that are in agreement with experience to date. The analyses were used to determine probability of nozzle failure versus EFPY for various head operating temperatures. Analyses were then performed to estimate the effect of visual and non-visual NDE inspections of the plants in the most critical inspection category, using the conservative assumption discussed above (see Concern #1 response) for probability of leakage detection by visual inspection.

These analyses demonstrate that performing visual inspections significantly reduces the probability of nozzle ejection, and that performing such examinations on a regular basis (in accordance with the inspection schedule prescribed in the MRP Inspection Plan) effectively maintains the probability of nozzle ejection at an acceptably low level indefinitely.

In the extremely unlikely event that nozzle failure and rod ejection were to occur due to an undetected circumferential crack, an acceptable margin of safety to the public would still be maintained (Ref 6). The consequences of such an event are similar to that of a small-break LOCA, which is a design-basis event.

The probability of core damage given a nozzle failure (assuming that failure leads to ejection of the nozzle from the head) has been estimated to be 1 x 103. The PFM analyses demonstrate that periodic visual inspections are capable of maintaining the probability of nozzle failure due to circumferential cracking well below 1 x 103. Therefore, the PFM analyses demonstrate that the resulting incremental change in core damage frequency due to CRDM nozzle cracking can be maintained at less than 1 x 10-6 (i.e., 1 x 103 times 1 x 10-3 equals 1 x 10.6) per plant year, through a program of periodic visual examinations performed in accordance with the MRP inspection plan. This result is consistent with NRC Regulatory Guide 1.174 that defines an acceptable change in core damage frequency (1 x 10.6 per plant year) for changes in plant design parameters, technical specifications, etc.

Concern 4:

The environment in the CRDM housing/RPV head annulus will likely be more aggressive after any through-wall leakage because potentially highly concentrated borated primary water may become oxygenated. This raises concerns about the technical basis for current crack growth rate models.

Response 4:

The MRP panel of international experts on stress corrosion cracking (SCC) (including representatives from ANL/NRC Research), prior to the Davis-Besse incident, gave extensive consideration to the likely environment in the annulus between a leaking CRDM nozzle and the RPV head and revisited this issue subsequently (Ref 5). When revisited, the relevant arguments remain valid for leak rates that are less than 6

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity 1 liter/h or 0.004 gpm, which plant experience has shown to be the usual case.

The conclusions were:

1. An oxygenated crevice environment is highly unlikely because:

(a) Back diffusion of oxygen is too low compared to counterflow of escaping steam (two independent assessments based on molecular diffusion models were examined).

(b) Oxygen consumption by the metal walls would further reduce its concentration.

(c) Presence of hydrogen from leaking water and diffusion through the upper head results in a reducing environment.

(d) Even if the concentration of hydrogen was depleted by local boiling, coupling between low alloy steel and Alloy 600 would keep the electrochemical potential low.

(e) Corrosion potential will be close to the Ni/NiO equilibrium, resulting in primary water stress corrosion cracking (PWSCC) susceptibility similar to normal primary water.

2. The most likely crevice environments are either hydrogenated steam or PWR primary water within normal specifications and both would result in similar, i.e. non-accelerated, susceptibility of the Alloy 600 penetration material to PWSCC.

3. If the boiling interface happens to be close to the topside of the J-weld, itself a low probability occurrence, concentration of PWR primary water solutes, lithium hydroxide and boric acid can in principle occur. Of most concern here would be the accelerating effect of elevated pH on SCC, but calculations and experiments show that any changes are expected to be small, in part because of the buffering effects of precipitates. A factor of 2x on the crack growth rate (CGR) should conservatively cover possible acceleration of PWSCC, even up to a high-temperature pH of around 9.

For larger leakage rates which could lead to local cooling of the head, concentration of boric acid and development of a sizeable wastage cavity adjacent to the penetration, the above arguments no longer directly apply. However, limited data (Berge et al., 1997) on SCC in concentrated boric acid solutions indicate that:

(a) Alloy 600 is very resistant to transgranular SCC (material design basis).

(b) High levels of oxygen and chloride are necessary for intergranular cracking to occur at all.

(c) The effects are then worse at intermediate temperatures, suggesting that the mechanism is different from PWSCC.

The above considerations show that there is no basis for assuming that a post-leakage crevice environment in the CRDM housing/RPV head annulus would be significantly more aggressive with regard to SCC of the Alloy 600 penetration material than normal PWR primary water, irrespective of the assumed leakage rate and/or annulus geometry. The current industry model (Ref 5), which includes a factor of 2x on CGR to cover residual uncertainty in the composition of the annulus environment, remains valid.

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ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity Concern 5:

The presence of boron deposits or residue on the RPV head, due to leakage from mechanical joints, could mask pressure boundary leakage. This raises concerns that a through-wall crack may go undetected for years.

Response 5:

The experience at Davis-Besse has clearly demonstrated that effective visual inspection for leakage from CRDM nozzle and weld PWSCC requires unobstructed inspection access and that the head surface be free of pre-existing boric acid deposits. Accumulations of debris and boric acid deposits from other sources can interfere with a determination as to the presence or absence of boric acid deposits extruding from the tube-to-head annulus. Therefore, to effectively perform a visual examination of the RPV head outer surface for penetration leakage, such deposits and debris accumulations must be carefully inspected, removed, and the area re-inspected. Evaluation may show that it is necessary to perform a non-visual examination to establish the source of the leakage.

Accordingly, each inspection at FNP Units 1 and 2 will be conducted with a questioning attitude, and any boric acid deposit on the vessel head will be evaluated to determine its source in accordance with existing industry guidance supplemented by the most recent industry experience at the time of the inspection.

These requirements are incorporated in the visual inspection guidance contained in the MRP Inspection Plan. Implementation of these requirements will preclude the cited condition of a through-wall crack remaining undetected for years.

Concern 6:

The causative conditions surrounding the degradation of the RPV head at Davis-Besse have not been definitively determined. The staff is unaware of any data applicable to the geometries of interest that support accurate predictions of corrosion mechanisms and rates.

Response 6:

The causes of the Davis-Besse degradation are sufficiently well known to avoid significant wastage. The root cause evaluation performed by the utility (Ref 7) clearly identifies the root cause as PWSCC of CRDM nozzles followed by boric acid corrosion. The large extent of degradation has been attributed to failure of the utility to address evidence that had been accumulating over a five-year period of time (Figure 26 of Ref 7).

The industry has provided utilities with guidance for inspecting the top of the RPV head to ensure that conditions approaching that which existed at Davis-Besse will not occur. Visual inspection guidelines have been provided (Ref 1), and a workshop was conducted to thoroughly review industry experience, regulatory requirements, leakage detection, and analytical work performed to understand the causes of high wastage rates (Ref 8).

Subsequent to significant wastage being discovered on the Davis-Besse RPV head, the industry has performed analytical work to determine how a small leak such as seen at several plants can progress to the significant amounts of wastage discovered at Davis-Besse. This work is referenced in the basis for the MRP Inspection Plan (Ref 9) and was presented to the NRC (Ref 10).

8

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity The analytical work shows that the corrosion rate is a strong function of the leakage rate. Finite element thermal analyses show that leak rates must reach approximately 0.1 gpm for there to be sufficient cooling of the RPV top head surface to support concentrated liquid boric acid that will produce high corrosion rates. The leak rate is in turn a strong function of the crack length. The effect of crack length above the J-groove weld on crack opening displacement and area has been confirmed by finite element modeling of nozzles including the effects of welding residual stresses and axial cracks. Leak rates have been calculated using crack opening displacements and areas determined by the finite element analyses and leak rate models based on PWSCC cracks in steam generator tubes.

Cracks that just reach the annulus through the base metal or weld metal will result in small leaks such as those that produced small volumes of boric acid deposits on several vessel heads at locations where the CRDM nozzles penetrate the RPV head outside surface. These leaks are typically on the order of 10-6 to 104 gpm. There is no report of any of these leaks resulting in significant corrosion. A leak rate of 10-3 gpm will result in a release of about 500 in3 of boric acid deposits in an 18-month operating cycle, which will be detectable by visual inspections.

The time for a crack to grow from a length that will produce a leak rate of 10-3 gpm to a leak rate of 0.1 gpm has been determined by deterministic analyses based on the MRP crack growth models to be 1.7 years for plants with 602'F head temperatures. Probabilistic analyses show that there is less than lxi0-3 probability that corrosion will proceed to the point that the inside surface cladding of the head would be uncovered over a significant area before the wastage would be detected by supplemental visual inspections as required under the MRP Inspection Plan. During the transition from leak rates of 10-3 gpm to 0.1 gpm, loss of material will be by relatively slow processes (Ref 9).

The ability to detect leakage prior to the risk of structural failure is illustrated by Figure 26 of the Davis Besse root cause analysis report. There was visual evidence of boric acid deposits on the vessel head for five years prior to the degradation being detected. Guidance provided in the MRP Inspection Plan would not permit these conditions to exist without determining the source of the leak, including nondestructive examinations if necessary.

Therefore, while the exact timing of the event progression at Davis-Besse cannot be definitively established, the probable duration can be predicted with sufficient certainty to conclude that a visual inspection regimen can ensure continued structural integrity of the RCS pressure boundary.

9

ENCLOSURE Joseph M. Farley Nuclear Plant - Unit 2 30-Day Response to NRC Bulletin 2002-02 Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity REFERENCES

1. EPRI Technical Report 1006899, Visual Examination for Leakage of PWR Reactor Head Penetrations on Top of the RPV Head: Revision 1, Report 1006899, March 2002.

2. Appendix B of EPRI Document MRP-75, "Probability of Detecting Leaks in RPV Top Head Nozzles," September 2002.

3.

EPRI TR-103696, PWSCC of Alloy 600 Materials in PWR Primary System Penetrations

4. Appendix A of EPRI Document MRP-75, Technical Report 1007337, "Technical Basis for CRDM Top Head Penetration Inspection Plan," September 2002.

5. MRP-55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," July 18, 2002.

6.

Walton Jensen, NRC, Reactor Systems Branch, Division of Systems Safety and Analysis (DSSA), Sensitivity Study of PWR Reactor Vessel Breaks, memo to Gary Holahan, NRC, DSSA, May 10, 2002.

7.

Davis-Besse Nuclear Power Station report CR2002-0891, "Root Cause Analysis Report Significant Degradation of the Reactor Pressure Vessel Head," April 2002.

8. EPRI Technical Report 1007336, "Proceedings of the EPRI Boric Acid Corrosion Workshop, July 25-26, 2002 (MRP-77)", September 2002.

9. Appendix C of EPRI Document MRP-75, Technical Report 1007337, "Supplemental Visual Inspection Intervals to Ensure RPV Closure Head Structural Integrity," September 2002.

10. Glenn White, Chuck Marks and Steve Hunt, Technical Assessment of Davis-Besse Degradation, Presentation to NRC Technical Staff, May 22, 2002.

11. NRC letter, "Flaw Evaluation Criteria," Jack Strosnider, NRC, to Alex Marion, NEI, November 21, 2001.

12. MRP-48, "PWR Materials Reliability Program Response to NRC Bulletin 2001-01," August 2001.

13. EPRI Letter MRP 2002-086. "Transmittal of PWR Reactor Vessel (RPV) Upper Head Penetrations Inspection Plan, Revision 1, August 6, 2002", from Leslie Hartz, MRP Senior Representative Committee Chairman, August 15, 2002.

14. EPRI Document MRP-75, Technical Report 1007337, "PWR Reactor Pressure Vessel (RPV)

Upper Head Penetrations Inspection Plan (MRP-75), Revision 1", September 2002 10