View source for L-PI-14-095, 10 CFR 50.55a Requests 1-RR-5-7 and 2-RR-5-7 Associated with the Fifth Ten-Year Interval for the Lnservice Inspection Program

{{Adams
| number = ML14258A124
| issue date = 09/15/2014
| title = 10 CFR 50.55a Requests 1-RR-5-7 and 2-RR-5-7 Associated with the Fifth Ten-Year Interval for the Lnservice Inspection Program
| author name = Davison K
| author affiliation = Northern States Power Co, Xcel Energy
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000282, 05000306
| license number = DPR-042, DPR-060
| contact person = 
| case reference number = L-PI-14-095
| document type = Inservice/Preservice Inspection and Test Report, Letter type:L
| page count = 14
}}

=Text=
{{#Wiki_filter:Xcel Energy SEP .1 5 2014                                                        L-PI-14-095 10 CFR 50.55a US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 10 CFR 50.55a Requests 1-RR-5-7 and 2-RR-5-7 Associated with the Fifth Ten-Year Interval for the lnservice Inspection Program Pursuant to 10 CFR 50.55a, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), hereby requests NRC approval of 10 CFR 50.55a Requests numbered 1-RR-5-7 and 2-RR-5-7 for the fifth ten-year interval for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, lnservice Inspection (lSI) Program. The details of these 10 CFR 50.55a requests are provided in the enclosure to this letter.
NSPM requests approval of these 10 CFR 50.55a requests by September 15, 2015, to support preparations for the first unit refueling outage in the PINGP fifth ten-year lSI Program interval.
If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-267-1736.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
   /)~~~
K~      Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota Enclosures (1) 1717 Wakonade Drive East
* Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2 cc:   Administrator, Region Ill, USNRC Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC

Enclosure 10 CFR 50.55a Request 1-RR-5-7, Rev. 0 (PINGP Unit 1) 10 CFR 50.55a Request 2-RR-5-7, Rev. 0 (PINGP Unit 2)
Proposed Alternative for Examination of Reactor Vessel Upper Head Closure Head Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)
Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety
: 1. American Society of Mechanical Engineers (ASME) Code Component(s)
Affected Code Class:                      1

==Reference:==
Code Case N-729-1 Examination Category:            Class 1 Pressurized Water Reactor Vessel Upper Head Item Number:                     84.40 1

== Description:==
Nozzles and partial-penetration welds of PWSCC -resistant materials in reactor vessel closure heads Component Number:                157-051 and 257-051

===2. Applicable Code Edition and Addenda===
The applicable Code edition and addenda for the fifth lnservice Inspection (lSI) interval beginning December 21, 2014, is ASME Section XI, "Rules for lnservice Inspection of Nuclear Power Plant components," 2007 Edition through the 2008 Addenda. Note that the upper head of the Unit 2 reactor vessel was replaced in May 2005 and the upper head of the Unit 1 reactor vessel was replaced in May 2006.
: 3. Applicable Code Requirements The Code of Federal Regulations 10 CFR 50.55a(g)(6)(ii)(D)(1 ), requires (in part):
All licensees of pressurized water reactors shall augment their inservice inspection program with ASME Code Case N-729-1 subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section. Licensees of existing operating reactors as of September 10, 2008 shall implement their augmented inservice inspection program by December 31, 2008.
1 Primary water stress corrosion cracking.
Page 1 of 12 0CFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 10 CFR 50.55a(g)(6)(ii)(D)(3) conditions ASME Code Case N-729-1 [1] by stating:
Instead of the specified 'examination method' requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1. A demonstrated volumetric or surface leak path assessment through all J-groove welds shall be performed. If a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.
ASME Code Case N-729-1 specifies that the reactor vessel upper head components shall be examined on a frequency in accordance with Table 1 of this code case. The basic inspection requirements of Case N-729-1 for partial-penetration welded Alloy 690 head penetration nozzles are as follows:
* Volumetric/surface examination of all nozzles every ASME Section XI 10-year lSI interval (provided that flaws attributed to PWSCC have not previously been identified in the head); and
* Direct visual examination (VE) of the outer surface of the head for evidence of leakage every third refueling outage or five calendar years, whichever is less.

===4. Reason for Request===
Code Case N-729-1 with the conditions of 10 CFR 50.55a(g)(6)(ii)(D)(3) requires volumetric and/or surface examination of reactor vessel upper head nozzles and welds. These examinations require accessing the underside of the reactor vessel head which is highly contaminated. The estimated radiation dose for examination is 1 to 2 Rem for each unit. Approximately 1 Rem can be attributed directly to the examination with additional dose the result of supporting activities. Modification of the existing reactor vessel head stands will be required to raise the heads approximately an additional two feet to accommodate examination equipment, and modified shielding will also be required. As such, the code requirement causes a hardship due to significant dose, and some increased risk to industrial safety associated with head stand modifications and handling of heavy loads.
There is no compensating increase in the level of quality and safety. As stated in the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP):
   "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR
[pressurized water reactor] Reactor Vessel Top Head Penetration Nozzles" (MRP-375) the:
       ... current inspection regime was established in 2004 as a conservative approach and was intended to be subject to reassessment upon the availability of additional laboratory data and plant experience on the performance of Alloy Page 2 of 12 OCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 690 and Alloy 52/152. Since that time, plant experience and laboratory testing have continued to demonstrate the much greater resistance of these replacement alloys to PWSCC compared to that for Alloys 600/82/182 for the material conditions relevant to partial-penetration welded nozzles. Although laboratory research is ongoing to investigate and understand the times to crack initiation and the crack growth rates for these materials under various conditions, there are now sufficient data available to develop an improved technical basis for inspection of these components.
The technical basis of MRP-375 demonstrates that the re-examination interval can be extended to a 20 year interval while maintaining an acceptable level of quality and safety. Therefore, NSPM requests approval of this alternative to allow the use of the lSI interval extension for the affected Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2 components.
: 5. Proposed Alternative and Basis for Use Proposed Alternative:
Pursuant to 10 CFR 50.55a(a)(3)(ii), NSPM requests an alternative from performing the required volumetric/surface examinations for the Reactor Vessel Closure Head (RVCH) components identified above at the frequency prescribed in ASME Code, Section XI, Code Case N-729-1. Specifically, NSPM requests extension of the volumetric/surface examination frequency of the RVCH in Table 1, Item 84.40 of ASME Code Case N-729-1 for approximately 10 years beyond the current inspection interval (nominally 10 calendar years) from installation of the Unit 1 and Unit 2 replacement RVCHs. This request would extend the Unit 2 volumetric/surface examinations currently scheduled for the fall of 2015 (baseline exams performed in 2005) for Unit 2 to 2025. This request would also extend Unit 1 volumetric examinations currently scheduled for the fall of 2016 (baseline exams performed in 2006) to 2026. This request applies to the inspection frequencies for volumetric and/or surface examinations only, as the inspection techniques or other requirements may change with later editions of ASME Section XI and 10 CFR 50.55a.
Basis for Use:
The basis for the inspection frequency for ASME Code Case N-729-1 comes, in part, from the analysis of laboratory and plant data presented in report MRP-111 [2], which was summarized in the safety assessment for RVCHs in MRP-11 0 [3]. The material improvement factor for PWSCC of Alloy 690/52/152 materials over that of mill-annealed Alloys 600 and 182 was shown by this report to be on the order of 26 or greater.
Further evaluations were performed to demonstrate the resistance of Alloys 690/52/152 to PWSCC under a recent EPRI MRP initiative provided in MRP-375 [4].
This report combines an assessment of the test data and operating experience Page 3 of 12

Enclosure- I OCFR50.55a Requests 1-RR-5-7 and2-RR-5-7 developed since the technical basis for the 10-year interval of Case N-729-1 was developed in 2004 with deterministic and probabilistic evaluations to assess the improved PWSCC resistance of Alloys 690/52/152 relative to Alloys 600/82/182.
Evaluation of Existing Alloy 690/52/152 Data and Experience by MRP-375 Operating experience to date for replacement and repaired components using Alloys 690/52/152 has shown a proven record of resistance to PWSCC during numerous examinations in the approximate 25 years of its application. This includes steam generators, pressurizers, and RVCHs. In particular, Alloy 690/52/152 operating experience includes inservice volumetric/surface examinations performed in accordance with ASME Code Case N-729-1 on nine (expected to be 13 at the end of the spring 2014 outage season) of the 40 replacement RVCHs currently operating in the United States of America (USA) as of February 2014. Some of these examined heads had continuous full power operating temperatures that may approach 613&deg;F.
None of these examinations revealed PWSCC cracking and these examination results further support the low likelihood of the potential for the RVCH to experience PWSCC during the extension period.
The evaluation performed in MRP-375 considers a simple Factor of Improvement (FOI) approach applied in a conservative manner to model the increased resistance of Alloys 690/52/152 compared to Alloys 600 and 182 at equivalent temperature and stress conditions 2 . FOis were estimated for the material improvements of Alloys 690/52/152 materials using an extensive database of test data. Results for both crack initiation and crack growth conclude a substantially higher resistance to PWSCC for Alloy 690 base material and Alloy 52/152 weld materials. Figures 3-2, 3-4, and 3-6 of MRP-375, provide crack growth rate data for Alloy 690/52/152 materials and heat affected zones with curves plotting FOis of 1, 5, 10, and 20 on a statistical basis reflecting the material variability exhibited in MRP-55 [5] for Alloy 600 material and in MRP-115 [6] for Alloy 82/182/132 weld material. An FOI of 20 bounds most of the data plotted, and an FOI 3
of 10 or less bounds all of the crack growth rate data . Table 3-6 of MRP-375 provides a summary of FOis determined on the basis of crack growth rate and crack initiation data. For crack initiation, FOis reported, although significant, are conservatively small because crack initiation of Alloys 690/52/152 was not observed during testing; instead, the initiation time was assumed to be equivalent to the test duration.
2 Alloy 600 wrought material is the appropriate reference for defining the FOI for Alloy 690 wrought material. As discussed in Section 3.1 of MRP-375, Alloy 182 weld metal is chosen as the reference for defining the FOI for Alloys 52 and 152 weld metals because Alloy 182 is more susceptible on average to PWSCC initiation and growth than Alloy 82 (due to the higher Cr content of Alloy 82).
3 As discussed in Section 3.3 of MRP-375, the laboratory crack growth rate data compiled in MRP-375 represent the values reported by individual researchers, without any adjustment by the authors of MRP-375 other than for temperature and stress intensity factor. The data presented in Figures 3-2, 3-4, and 3-6 of MRP-375 represent essentially the entire set of data points reported by the various laboratories. No screening process was applied to the data on the basis of test characteristics such as minimum required crack extension or minimum required engagement to intergranular cracking. Instead, an inclusive process was applied to conservatively assess the factors of improvement apparent in the data for specimens with less than 10% added cold work.
Page 4 of 12 OCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 Additional Evaluations Performed under MRP-375 MRP-375 applied the FOI results to perform a combination of deterministic and probabilistic evaluations to establish an appropriately conservative inspection interval for Alloy 690 RVCHs. The deterministic technical basis applies industry-standard crack growth calculation procedures to predict time to certain adverse conditions under various conservative assumptions. A probabilistic evaluation is then applied to make predictions for leakage and ejection risk, generally using best-estimate inputs and assumptions, with uncertainties treated using statistical distributions.
The deterministic crack growth evaluation provides a precursor to the probabilistic evaluation to directly illustrate the relationship between the improved PWSCC growth resistance of Alloys 690/52/152 and the time to certain adverse conditions. These evaluations apply conservative crack growth rate predictions and the assumption of an existing flaw (which is replaced with a PWSCC initiation model for probabilistic evaluation). The evaluations provide a reasonable lower bound on the time to adverse conditions, from which a conservative inspection interval may be recommended. This evaluation draws from various EPRI MRP and industry documents that evaluate, for Alloys 600/82/182, the time from a detectable flaw being created to leakage occurring and from a leaking flaw to the time that net section collapse (nozzle ejection) would be predicted to occur. Applying a conservative crack growth FOI of 20 to circumferential and inside diameter (I D) axial cracking and of 10 to outside diameter (OD) axial cracking for Alloy 690 versus Alloy 600, the results show that more 20 years is required for leakage to occur and that more than 120 years would be required to reach the critical crack size subsequent to leakage.
The probabilistic model in MRP-375 was developed to predict PWSCC degradation and its associated risks in RVCHs. The model utilized in this probabilistic evaluation is modified from the model presented in Appendix B of MRP-335, Rev. 1 [7] that evaluated surface stress improvement of RVCHs with Alloy 600 nozzles. The integrated probabilistic model in MRP-375 includes submodels for simulating component and crack stress conditions, PWSCC initiation, PWSCC growth, and flaw examination. The submodels for crack initiation and growth prediction for Alloy 600 reactor pressure vessel head penetration nozzles (RPVHPNs) in MRP-335, Rev. 1 were adapted for RVCHs with Alloy 690 nozzles by applying FOis to account for the superior PWSCC resistance of Alloys 690/52/152. The average leakage frequency and average nozzle ejection frequency were determined using the Monte Carlo simulation model with conservative FOI assumptions. The results show that, using only modest FOis for Alloys 690/52/152, the potential for developing a safety significant flaw (risk of nozzle ejection) is acceptably small for a volumetric/surface examination period up to 40 years.
The evaluations performed in MRP-375 were prepared to bound all PWR replacement RVCH designs that are manufactured using Alloy 690 base material and Alloy 52/152 weld materials. The evaluations assume a bounding continuously operating RVCH Page 5 of 12

Enclosure- I OCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 temperature of 613&deg;F and a relatively large number of RVCH penetrations (89 penetrations).
RVCH Design and Operation The analysis presented in MRP-375 was intended to cover all replacement heads in USA PWRs, including the PINGP RVCHs. The MRP-375 analyses assume a reactor vessel head operating temperature of 613&deg;F to bound the known reactor vessel head temperatures of all USA PWRs currently operating. The average RVCH operating temperature for PINGP Units 1 and 2 over the operating period from installation of the replacement head until the end of the requested volumetric/surface inspection period is conservatively no more than 595 oF based on hot leg temperatures recorded over a year of operation [8]. Actual temperatures based on thermocouple readings since the heads were replaced have ranged between 574 oF and 580 &deg;F, based on Unit 2 thermocouples which are indicative of performance of both units. Based on this, the Unit 1 and 2 RVCH average operating temperatures (which is the measure of temperature relevant to potential PWSCC degradation) are bounded by the MRP-375 evaluation, which assumes 613&deg;F for its main deterministic and probabilistic calculations.
As stated in MRP-375, "... to further allow consistent interpretation, all results are adjusted to an operating temperature of 613&deg;F (323&deg;C) using the Arrhenius relationship with an activation energy of 130 kJ/mol. This operating temperature is believed to be an upper bound for operating Alloy 690 top heads in service today." (Footnote omitted)
Reduced operating temperature results in a significant reduction in both crack initiation and crack propagation. As stated in MRP-375 Case M2- Reduced Operating Temperature:
Reducing the head temperature from 613&deg;F to 600&deg;F (323oC to 316&deg;C) reflects that most Alloy 690 hot heads operate below 613&deg;F (323&deg;C), with a majority operating between 590&deg;F and 600&deg;F (31 ooc to 316&deg;C). The reduced temperature decreases the thermally activated PWSCC flaw initiation and growth processes (i.e., through the Arrhenius relation in the model).
Reducing the head temperature leads to a more than tenfold reduction in AEF
[Average Ejection Frequency]. Similarly, the frequency of leakage is decreased to less than half its base case value.
The PINGP RPVCHs were designed and fabricated using materials and techniques to reduce susceptibility to PWSCC and facilitate prompt detection of potential leakage by visual examination. The resistance of the PINGP heads to PWSCC and detectability of leakage are well within the scope of MRP-375.
The following table summarizes the design attributes of the PINGP RPVCHs.
Page 6 of 12 0CFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 Feature                                     Description Reactor vessel (RV) head           SA-508, Grade 3, Class 1 with a targeted chromium material                           content of 0.15% or less, and sulfur content limited to 0.015% maximum. Cladding is generally 309L or 308L, 0.125" minimum and 0.2" nominal thickness RV head assembly upgrade           Enhanced access doors for inspection of RV head package                            penetrations.
Coatings                           The replacement RVCH (RRVCH) is not painted to ensure full compliance with the NRC bare metal inspection criteria for RV head inspections.
RV head 4" OD penetration          SB-167 UNS N06690 (lnconel 690) material Penetration to head weld           Alloy 52 ERNiCrFe-7 (UNS N06052) and Alloy 152 material, type, interference fit   ERNiCrFe-7 (UNS N86152), Narrow J-Groove, 0.001-0.002" interference fit Full-length control rod drive      29 mechanism penetrations Spare penetrations                 4 Thermocouple penetrations          3 core exit thermocouple nozzle assemblies Reactor coolant gas vent           1" Schedule 160 SB-167 UNS N06690 (lnconel690) system penetration pipe nozzle with Alloy 52/152 ERNiCrFe-7 narrow J-groove weld Reactor vessel level               1" Schedule 160 SB-167 UNS N06690 ( lnconel690) instrumentation system             with Alloy 52/152 ERNiCrFe-7 narrow J-groove weld penetration pipe nozzle Note that the probabilistic analysis in MRP-375 was performed assuming a head with 89 partial-penetration welded nozzles which bounds the number of penetrations in the PINGP replacement heads. The number of penetrations included in the probabilistic model is not a key variable, and the assumed number of penetrations results in a small change in results relative to other sensitivity cases. Thus, the probabilistic calculations of MRP-375 cover all USA replacement RVCHs regardless of the precise number of penetrations.
The RRVCHs were buttered with Alloy 690 in the area of the penetration. A partial penetration J-groove weld using Alloy 52/52M (ERNiCrFe-7 UNS N06052) filler metal was used between the Alloy 690 penetration and the head on the inside of the RRVCH. Two modifications were introduced in the weld to reduce residual stress: a narrow gap J-groove weld edge preparation was used to reduce the volume of weld metal deposited; and automatic welding processes were used. Manual welding of the first and reinforcement layer was permitted; all remaining layers were deposited using an automatic welding process.
Page 7 of 12 0CFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 Based on the above, the PINGP RVCHs are expected to have resistance to PWSCC as good, if not better, then the MRP-375 base case.
A bare metal visual examination (VE) was performed on the Unit 1 RRVCH in 2011 and Unit 2 RRVCH in 2013, in accordance with ASME Code Case N-729-1, Table 1, Item 84.30. These visual examinations were performed by VT-2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. These examinations did not reveal any surface or nozzle penetration boric acid that would be indicative of nozzle leakage. Examination will be performed again for Unit 1 in the upcoming 1 R29 refueling outage scheduled to commence in October 2014. Examination will be performed again for Unit 2 in the upcoming 2R30 refueling outage scheduled to commence in September 2017. Continued visual inspection in accordance with Code Case N-729-1 will provide prompt detection of any potential leakage.
As an alternative to the requirements of Code Case N-729 -1 with conditions of 10 CFR 50.55a for volumetric examination of the reactor vessel closure heads on a nominal ten year interval, NSPM requests approval to perform volumetric examination of the reactor vessel closure heads to the requirements of Code Case N-729-1 with conditions of 10 CFR 50.55a on a nominal twenty year interval.
Minimum FOI Implied by Requested Inspection Period ASME Code Case N-729-1 is based upon conclusions in MRP-117 [9] that a reexamination interval between volumetric/surface examinations of one 24-month operating cycle is acceptable for a head with Alloy 600 nozzles operating at a temperature of 605&deg;F. The inspection period for heads with Alloy 690 nozzles in Code Case N-729-1 is a nominal10 years, which represents a minimum implied factor of improvement (FOI) of five over Alloy 600.
FOI Approach Per the technical basis documents for ASME Code Case N-729-1 for heads with Alloy 600 nozzles ([3], [9], and [1 0]), the effect of differences in operating temperature on the required volumetric/surface reexamination interval for heads with Alloy 600 nozzles can be easily addressed on the basis of the Re-lnspection Years (RIY) parameter. The RIY parameter adjusts the effective full power years (EFPYs) of operation between inspections for the effect of head operating temperature using the thermal activation energy appropriate to PWSCC crack growth. For heads with Alloy 600 nozzles, ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D)(2) limits the interval
                                                                     =
between subsequent volumetric/surface inspections to RIY 2.25. The RIY parameter, which is referenced to a head temperature of 600&deg;F, limits the time available for potential crack growth between inspections.
Page 8 of 12 0CFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 The RIY parameter for heads with Alloy 600 nozzles is adjusted to the reference head temperature using an activation energy of 130 kJ/mol (31 kcal/mol) [1]. Based on the available laboratory data, the same activation energy is applicable to model the temperature sensitivity of growth of a hypothetical PWSCC flaw in the Alloy 690/52/152 material of the replacement RVCH. Key laboratory crack growth rate testing data for Alloy 690 wrought material investigating the effect of temperature are as follows:
(1) Results from Argonne National Laboratory (ANL) indicate that Alloy 690 with 0-26% cold work has an activation energy between 100 and 165 kJ/mol (24-39 kcal/mol) [11]. NUREG/CR-7137 [11] concludes that the activation energy for Alloy 690 is comparable to the standard value for Alloy 600 (130 kJ/mol).
(2) Testing at Pacific Northwest National Laboratory (PNNL) found an activation energy of about 120 kJ/mol (28. 7 kcal/mol) for Alloy 690 materials with 17-31%
cold work [12].
These data show that it is reasonable to assume the same crack growth thermal activation energy as was determined for Alloys 600/82/182 (namely 130 kJ/mol (31 kcal/mol)) for modeling growth of hypothetical PWSCC flaws in Alloy 690/52/152 PWR plant components.
As discussed in the MRP-117 [9] technical basis document for heads with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate reexamination interval to detect any PWSCC in a timely fashion. USA PWR inspection
                                                                                =
experience for heads with Alloy 600 nozzles has confirmed that the RIY 2.25 interval results in a suitably conservative inspection program. There have been no reports of nozzle leakage or of safety-significant circumferential cracking for times subsequent to the time that the Alloy 600 nozzles in a head were first examined by non-visual inservice non-destructive examination [13].
Minimum FOI Implied by Requested Inspection Period NSPM has assessed the minimum Alloy 690/52/152 FOI that supports the requested Unit 1 and Unit 2 extension period for comparison with the laboratory crack growth rate data presented in MRP-375. An extension of the examination interval to 20 years would imply a factor of 20/2 or 10 for Alloys 690/52/152 relative to Alloys 600 and 182 for the proposed period between volumetric/surface examinations for a head operated at a temperature of 605&deg;F. To calculate the minimum implied FOI for the Unit 1 and Unit 2 RVCHs operating temperature of 595&deg;F, the RIY parameter for the requested examination interval is compared with the N-729-1 interval for Alloy 600 nozzles of RIY
   = 2.25.
The representative Unit 1 and Unit 2 RVCH operating temperatures of 595&deg;F corresponds to an RIY temperature adjustment factor of 0.882 (versus the reference temperature of 600&deg;F) using the activation energy of 31 kcal/mol (130 kJ/mol) for crack growth of ASME Code Case N-729-1. As discussed previously, this standard activation energy is appropriate to apply for modeling crack growth of Alloy 690/52/152 Page 9 of 12

Enclosure- IOCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7 plant components. Conservatively assuming that the EFPYs of operation accumulated for Unit 1 and Unit 2 since RVCH replacement is equal to the calendar years since replacement, the RIY for the requested extended period would be (0.882 factor for growth rate)(20 total calendar years for extended interval)= 17.63 RIY690. The FOI implied by this RIY value for Unit 1 and Unit 2 is (17.63 RIY690)/(2.25) = 7.8 FOI.
Considering the statistical compilation of data provided in Figures 3-2, 3-4, and 3-6 of MRP-375, this factor of improvement is conservatively less than the FOI of 10 that bounds the crack growth rate data presented. Furthermore, as discussed in Sections 2 and 3 of MRP-375, PWR plant experience and laboratory testing have demonstrated a large improvement in resistance to PWSCC initiation of Alloys 690/52/152 in comparison to that for Alloys 600/82/182. Hence, the demonstrated improvements in PWSCC initiation and growth confirm on a conservative basis the acceptability of the requested period of extension.
Conclusions The proposed alternative provides reasonable assurance of structural integrity in that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the Unit 1 and Unit 2 replacement RVCHs provide a superior reactor coolant system pressure boundary where the potential for PWSCC has been shown by analysis and by years of positive industry experience to be minute. The minimum FOI implied by the requested extension period represents a level of reduction in PWSCC crack growth rate versus that for Alloys 600/82/182 that is completely bounded on a statistical basis by the laboratory data compiled in MRP-375. Given the lack of PWSCC detected to date in any PWR plant applications of Alloys 690/52/152, the simple FOI assessment clearly supports the requested period of extension.
NSPM proposes to perform volumetric/surface examinations of the Unit 1 and Unit 2 reactor vessel heads in accordance with Code Case N-729-1 with conditions of 10 CFR 50.55a on a nominal20 year interval as an alternative to the nominal10 year interval required by the Code Case N-729-1.
: 6. Duration of Proposed Alternative Relief is requested for the fifth 10-year inspection interval of the lnservice Inspection Program for PINGP Units 1 and 2. The fifth interval is effective for Units 1 and 2 from December 21, 2014 through December 20, 2024.
: 7. Precedents Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1, Arkansas Nuclear One, Unit 1 -
Currently under NRC review (NRC ADAMS Accession No. ML14118A477 [14]).
Page 10 of 12

Enclosure- IOCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7
: 8. References
: 1. ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1,"approved March 28, 2006.
: 2. Materials Reliability Program (MRP): Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI, Palo Alto, CA, U.S. Department of Energy, Washington, DC:
2004. [ML041680546]
: 3. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-110NP), EPRI, Palo Alto, CA.
[ML041680506]
: 4. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375), EPRI, Palo Alto, CA: 2014. [freely available at www.epri.com]
: 5. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA. [freely available at www.epri.com]
: 6. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA.
: 7. Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Revision 1),
EPRI, Palo Alto, CA. [freely available at www.epri.com]
: 8. Prairie Island Nuclear Generating Plant, Emergency Response Computer System points 1T0419A, 1T0439A, 2T0419A and 2T0439A from July 28, 2013 to July 28, 2014.
: 9. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP-117), EPRI, Palo Alto, CA.
[ML043570129]
: 10. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis of PWR Reactor Pressure Vessel Top Head Nozzle Cracking (MRP-105 NP), EPRI, Palo Alto, CA. [ML041680489]
: 11. U.S. NRC, Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment- 2009, NUREG/CR-7137, ANL-10/36, published June 2012. [ML12199A415]
: 12. Materials Reliability Program: Resistance of Alloys 690, 152, and 52 to Primary Water Stress Corrosion Cracking (MRP-237, Rev.2): Summary of Findings Between 2008 and 2012 from Completed and Ongoing Test Programs, EPRI, Palo Alto, CA: 2013.3002000190. [freely available at www.epri.com]
: 13. EPRI MRP Letter 2011-034, ''Tcold RV Closure Head Nozzle Inspection Impact Assessment," dated December 21, 2011. [ML12009A042]
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Enclosure- IOCFR50.55a Requests 1-RR-5-7 and 2-RR-5-7
: 14. Letter from S. L. Pyle (Entergy) to U.S. NRC, Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51, 1CAN041402, dated April28, 2014, including Request for Alternative AN01-ISI-024. [ML14118A477]
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