ML23289A152

Attachment 1 - Monticello Trp 76 Suppl Audit Questions

ML23289A152
Person / Time
Site:	Monticello
Issue date:	10/18/2023
From:	Mary Johnson NRC/NRR/DNRL/NLRP
To:	Hafen S Northern States Power Co
References
EPID L-2022-SLR-0000
Download: ML23289A152 (6)

Text

1 ATTACHMENT 1 MONTICELLO SLRA: SUPPLEMENTAL AUDIT QUESTIONS TRP 76: IRRADIATION EFFECTS ON BIOLOGICAL SHIELD CONCRETE AND STRUCTURAL STEEL Audit Question C13

Background

SLRA Section 3.5.2.2.2.6, as amended by Supplement 5 dated August 28, 2023 (ML23240A695), Enclosure 03c, states in page 3 of 4 of the applicants evaluation of the effect gamma radiation heating and its cumulative impact on concrete temperature, that:

"NUREG/CR-7171 (ML13325B077) and RIL 2021-07 (ML21238A064) contain equations backed by test results that show the cumulative effect of heating due to irradiation. The heating effect from gamma ray irradiation has been determined to be limited to approximately 1.12o F of temperature change due to gamma irradiation. The basis assumptions for the gamma irradiation dose (integrated energy absorption of the bioshield) evaluation at MNGP aligns with the irradiation demonstrated and discussed in RIL 2021-07.

The expansion of the concrete from this temperature increase results in a maximum of 0.9 ksi additional tensile stress during operation. Even in conservative and bounding scenarios, this level of additional stress will not affect conclusions of the basis bioshield or piping analysis or result in any additional age-related degradation mechanisms that must be addressed, monitored, or assessed."

SLRA Supplement 5, Enclosure 03c, also states, in part:

"Plant areas that bound high temperature considerations are the drywell general area and the biological shield wall piping penetration local area, which experience temperatures of 135oF and 179oF, respectively. Insulation is credited with maintaining..

In addition, at mid-core height, where peak radiation is expected, thermal concrete expansion might induce additional stress on the steel liner and a finite element model was used to evaluate the radial temperature profile (of the bioshield wall) from heat loads resulting from the reactor vessel temperature under operating conditions.. Based on the heat transfer analysis results, the maximum expected temperature on the concrete surface of the MNGP bioshield is 140.69oF. Temperature increase due to gamma-heating can be approximated to 1.12oF. which brings the maximum expected concrete surface temperature to approximately 142oF; which is below the ACI limit of 150oF."

Further SLRA Section 3.5.2.2.1.2 in Supplement 5, Enclosure 04 (page 3 of 3), states, in part:

The calculation of local concrete temperatures conservatively assumed a 25oF increase in temperature due to nuclear heating effects. This bounds the anticipated gamma heating effects for concrete discussed in Section 3.5.2.2.2.6.

2 Issue It is not clear how the applicant determined the heating effect of gamma to be 1.12o F. The reference to NUREG/CR-7171 and RIL 2021-07 appears superficial with no specifics or context, and the staff is unclear what equations, test results, and assumptions from these referenced documents or other were used to arrive at 1.12oF and where the location of this temperature is in the bioshield wall concrete. Also, the staff is unclear how the additional tensile stress of 0.9 ksi from concrete expansion from this temperature increase was determined, and whether the stress is in the concrete or in the steel components of the bioshield wall. This information was not provided during the regulatory audit.

The magnitude of the above temperature estimate of gamma heating effect appears to be significantly inconsistent with similar estimates found in the literature with a lower level of gamma radiation than that estimated for MNGP; i.e., typical total temperature profiles including the effects of gamma radiation for varying air gap temperatures and airflow shown in Fig. 7 of the reference P.M. Bruck et al Structural assessment of radiation damage in light water power reactor concrete biological shield walls, Nuclear Engineering and Design 350 (2019) 9-20, provided on the ePortal. Figure 7 in the above reference also indicates that the peak temperature occurs at a distance inside the bioshield concrete wall inner surface, and not on its surface. Another literature example in the ePortal of such temperature variation in concrete is Figure 4-13 of ORNL/TM-2018/769, Revision 0, Expected Condition of Concrete Exposed to Radiation at Age 80 Years of Reactor Operation, US Department of Energy, January 2018.

From review on the ePortal of calculation CA-97-007, Revision 0 (1/2/97), Calculation of Local Temperature in Biological Shield Wall, assumption 5 on page 2 of 7 states: the temperature at the inside surface of the biological shield wall was estimated to be approximately 125oF. For this analysis, the bulk temperature of biological shield is assumed to be 150oF to account nuclear heating effects. The staff noted that this calculation is based on bulk temperature of air cavity being 125oF which is less than the air cavity temperature of 135oF or 140.69oF as stated in the Background section above; therefore, considering the above, the increase in temperature to account for nuclear heating effects appears to be inconsistent and significantly lower than the 25oF claimed in Enclosure 04 of SLRA Supplement 5.

Discussion Question / Request

1. Describe in sufficient technical detail with supporting calculation(s), assumptions, and justification and applicability of basis how the applicant determined the temperature increase in concrete due to heating effect of gamma radiation to be 1.12o F (or other) and the corresponding increase in tensile stress of 0.9 ksi. State whether the increase in stress is in the bioshield concrete or in its encapsulating steel. State the location of the maximum cumulative temperature effects, including those due to gamma heating, in the bioshield wall section.

2. Reconcile the inconsistencies (identified above in the Issue section) for the net temperature increase considered to account for gamma heating in the evaluation of total temperature effects on concrete.

3. If the 0.9 ksi stress increase is in concrete, clarify how the increase in stress due to expansion was considered into the strength reduction evaluation of the structural reinforced concrete section of the bioshield wall and acceptance criteria.

3 Audit Question C14

Background

SLRA Section 3.5.2.2.2.6, as amended by Supplement 5 dated August 28. 2023 (ML23240A695), Enclosure 03c, summarizes the separate plant-specific analysis that was performed to address the impact of potential reduction of strength and mechanical properties of the bioshield wall reinforced concrete due to gamma dose exceeding the SRP-SLR threshold limit for the subsequent period of extended operation. For the plant specific analysis the enclosure references an industry methodology and states (on Enclosure 03c page 2 of 4, third paragraph) that this analysis considered attenuation through the concrete, and the potential for radiation induced volumetric expansion (RIVE) of the biological shield concrete thickness that is above the damage threshold, as well as impact to gamma heating considerations.

Issue The summary description of the plant-specific analysis performed and results is generally qualitative and lacks specific clarity. It is not clear what is the applicants referenced methodology, whether it is applicable in the gamma dose bioshield concrete irradiation evaluation, and how it is used in the plant-specific analysis and what the sources of unspecified industry literature are. The staff needs additional information to make its regulatory finding. Also, SLRA Supplement 5 (Enclosure 03a) Figure 3.5.2.2.2.6-1 is not legible and the numbers indicated are not clear.

Discussion Question / Request

1. Confirm that the generic curve from industry literature for the variation of gamma flux/dose normalized to flux at core-midplane along the height of the bioshield wall used in the SLRA analysis (and supporting audited calculation 2200285.301 in ePortal) for attenuation of gamma dose is in publicly available sources as follows: In Figure 3-7 Axial Gamma Flux Variation Relative to Core Fuel Mid-Height of open source Report ORNL/TM-2018/769, Revision 0, Expected Condition of Concrete Exposed to Radiation at Age 80 Years of Reactor Operation, US Department of Energy, January 2018 (https://info.ornl.gov/sites/publications/Files/Pub107682.pdf), and similarly in Figure 2-7 in EPRI Report 3002011710, Irradiation Damage of the Concrete Biological Shield - Basis for Evaluation of Concrete Biological Wall for Aging Management, May 2018.

2. State the specific publicly available source that was used to quantify the reduction in concrete strength and material properties subjected to gamma dose exposure (e.g., Hilsdorf et al paper).

3. State the specified minimum design compressive strength of concrete (fc) for the original design of bioshield wall and the degraded (reduced) strength that was used for the evaluation of gamma radiation.

4. State the controlling loading combination that was used for the evaluation of reduction in strength of the bioshield wall reinforced concrete, the code-of-record, and the maximum demand to capacity ratio(s) (D/C) for the degraded case for the controlling loading combination.

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5. Clarify how the plant specific analysis evaluated the potential for radiation induced volumetric expansion (RIVE) of the biological shield wall concrete thickness, that is above the damage threshold, given the exceedance is for gamma dose only.

6. If possible, provide a legible SLRA Figure 3.5.2.2.2.6-1.

Audit Question S12

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Background===

SLRA Section 3.5.2.2.2.6, as amended by Supplement 5 dated August 28, 2023 (ML23240A695) in Enclosures 01 and 02 provide AMR line items and corresponding program enhancements to monitor and manage loss of fracture toughness as an aging effect that requires management for the biological shield wall steel components and or the RV steel supports, respectively, for the subsequent period of extended operation (SPEO). SRP-SLR Appendix A, Section A.1.2.3.3 states that the selected AMP should identify the aging effect(s) and should provide a link between the[se and] parameter or parameters to be monitored and how the monitoring of these parameters will ensure adequate management of the aging effect.

Issue As indicated in the GALL-SLR, loss of fracture toughness of a steel component is an aging effect that is indirectly managed using visual or volumetric examination techniques to monitor for symptoms (e.g., cracking). While the proposed SLRA AMR items identify the indirect aging effect requiring management, they do not appear to also specify the direct aging effect and associated parameter(s) monitored or inspected to ensure adequate management of the loss of fracture toughness aging effect.

Discussion Question / Request

1. Clarify and update the proposed AMR items in SLRA Supplement 5 (Enclosures 01 and 02) to also include direct aging effect(s) associated with managing the indirect aging effect of loss of fracture toughness for the MNGP biological shield wall structural steel components and the RV support steel components.

Audit Question S13

Background

SLRA Section 3.5.2.2.2.6, as amended by Supplement 5 dated August 28, 2023 (ML23240A695) in Enclosure 03e related to evaluation of the biological shield structural steel, states in part that the initial nil-ductility transition (NDT) temperature of the ASTM A36 material is

-28°F per the guidance in NUREG-1509. In the unamended SLRA Section 3.5.2.2.2.6, the applicant stated that the initial NDT temperature value of the ASTM A36 material is 39°F, which was consistent with the Table 4-1 of NUREG-1509 guidance for a hot rolled, carbon-manganese (C-Mn) steel, i.e., with the 1.3 standard deviation (1.3) uncertainty margin. The staff notes that the initial NDT temperature value of -28°F is without the uncertainty margin of 1.3 and is for the C-Mn steel, normalized case.

5 Issue The applicant did not explain the selection of the initial NDT temperature value of -28°F for the C-Mn steel, normalized case for the MNGP biological shield wall structural steel. The guidance in NUREG-1509 (Table 4-1) lists generic initial NDT temperature values for common steels used in reactor vessel supports. NUREG-1509 states that the initial NDT temperature values in the column labeled NDT + 1.3 of Table 4-1 should be used. The staff notes that unless the use of mean initial NDT temperature values (column NDT in Table 4-1 of NUREG-1509) is justified with plant-specific data, the initial NDT temperature values in the NDT + 1.3 column should be used consistent with the NUREG-1509 guidance. The staff also notes that it is not clear whether the maximum delta NDT temperature value stated on page 5 of Enclosure 03e of SLRA Section 3.5.2.2.2.6 Supplement 5 should be 129.6°F per page 4 of Enclosure 03e, instead of 121.04°F.

Discussion Question / Request

1. Justify the use of an initial NDT temperature value of -28°F for the ASTM A36 material in Supplement 5 for the MNGP biological shield wall structural steel, without consideration of the 1.3 uncertainty margin. This justification should include plant-specific data and an explanation for using the normalized value instead of the as-hot rolled value.

2. Clarify whether the maximum delta NDT temperature on page 5 of Enclosure 03e of SLRA Section 3.5.2.2.2.6 Supplement 5 should be 129.6°F instead of 121.04°F.

Audit Question S14

Background

SLRA Section 3.5.2.2.2.6, as amended by Supplement 5 dated August 28, 2023 (ML23240A695) in Enclosure 03e summarized the fracture mechanics evaluation of the biological shield structural steel. The applicant stated in part that the fracture toughness, KIC, is 58.7 ksi-in1/2 based on plant-specific data from certified material test reports (CMTR), and the limiting stress intensity factor (SIF) is 19.4 ksi-in1/2.

Issue The applicant did not provide the details of the fracture mechanics evaluation of the biological shield wall structural steel. For example, the applicant did not describe how KIC and the limiting SIF were determined. It is not clear whether this fracture mechanics evaluation performed for the biological shield wall structural steel was at the location of maximum dpa and/or maximum (or limiting) stress.

Discussion Question / Request Provide the following clarification and descriptions of the fracture mechanics evaluation of the biological shield wall structural steel:

1. Clarify that the location evaluated for fracture mechanics of the biological shield structural steel was at the location of maximum dpa of 2.07 x 10-3 and/or maximum (or limiting) stress.

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2. Describe the calculation of the KIC value of 58.7 ksi-in1/2 (strain rate effects and the embrittlement due to the 2.07 x 10-3 dpa radiation exposure level should be included in this description).

3. Describe the calculation of the limiting SIF value of 19.4 ksi-in1/2, including:

i. the applied loads used, including weld residual stress; ii. the contribution of the additional stress acting on the biological shield wall structural steel due to the 0.9 ksi tensile stress (see RAI 3.5.2.2.2.6-1 and SLRA Section 3.5.2.2.2.6, as amended by Supplement 5, Enclosure 03c); and iii. the SIF model (or models) used.

4. Describe conservatisms in the calculation of KIC and for the limiting SIF in parts (2) and (3).