ML18338A238
| ML18338A238 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 12/04/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-1218-63693 | |
| Download: ML18338A238 (11) | |
Text
RAIO-1218-63693 December 04, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Supplemental Response to NRC Request for Additional Information No. 52 (eRAI No. 8855) on the NuScale Design Certification Application
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 52 (eRAI No. 8855)," dated June 02, 2017
- 2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 52 (eRAI No.8855)," dated April 03, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).
The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 8855:
- 03.06.02-13 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.
Sincerely, v ~
Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A Enclosure 1: NuScale Supplemental Response to NRC Request for Additional Information eRAI No.8855 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com
RAI0-1218-63693 :
NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 8855 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com
- ~! NUSCALE '"
Powe, f oe o ll humonhnd Response to Request for Additional Information Docket No.52-048 eRAI No.: 8855 Date of RAI Issue: 06/02/2017 NRC Question No.: 03.06.02-13 NuScale FSAR Tier 2, Section 3.6.1.2, "Types of Breaks," states that high-energy lines are evaluated for both line breaks and through-wall leakage cracks. Line breaks include both circumferential (i.e., complete rupture around the circumference of the pipe) and longitudinal breaks (i.e., rupture of the pipe along its axis). In addition, it states that through-wall leakage cracks are as defined in BTP 3-4, Revision 2. Moreover, in FSAR Section 3.6.2.1, "Criteria Used to Define Break and Crack Location and Configuration," the applicant refers to BTP 3-4 guidance that the types of breaks postulated in high-energy lines include circumferential breaks in fluid system piping greater than 1 inch nominal diameter; longitudinal breaks in fluid system piping that is 4-inch nominal diameter and greater, and through-wall leakage cracks in fluid system piping greater than 1 inch nominal diameter. Furthermore, in FSAR Section 3.6.5, "Integral Jet Impingement Shield and Pipe Whip Restraint," the applicant states that per the criteria of BTP 3-4, longitudinal pipe breaks need not be postulated at terminal ends.
To ensure the compliance with GDC 4 requirements that SSCs important to safety be designed to accommodate the effects of postulated pipe ruptures, the NRC staff in BTP 3-4, Part B, Item C provides its guidance for postulating the type of breaks and leakage cracks in fluid system piping which includes the associated postulated rupture location and configuration. The NRC staff reviewed the above FSAR information against the NRC staff's guidance as delineated in BTP 3-4, Part B, Item C(i), C(ii), and C(iii) for circumferential pipe breaks, longitudinal pipe breaks, and through-wall leakage cracks respectively. The NRC staff found either insufficient information or no information (or pointer) which addresses certain NRC staff's guidelines delineated in BTP 3-4, Part B, Items C(i)(1 ), C(i)(3), C(i)(5), C(ii)(1 ), C(ii)(3), C(ii)(5), C(iii)(2),
C(iii)(3) and C(iii)(4). Explain how the referenced NRC staff's guidelines are considered in determining the break and crack location and configuration for the NuScale design.
NuScale Nonproprietary
NuScale Response:
The response to RAI 8855 Question 03.06.02-13, submitted by NuScale letter RAIO-0418-59380, dated April 3, 2018, was discussed in a follow-up public telecon with NRC on May 5, 2018. NRC noted that, although the submitted response properly addressed the original RAI question, the accompanying FSAR revision resulted in two sections in the FSAR Section 3.6 with the identical title ("Types of Breaks") but with slightly different information (3.6.1.2 and the RAI response-added 3.6.2.1.7). This section title duplication could lead to potential confusion.
NuScale agreed to supplement its RAI 8855 response to correct this condition. This issue was corrected in the recently submitted FSAR Revision 2, with the content of the original Section 3.6.1.2 rolled into FSAR Section 3.6.1, titled "Plant Design for Protection against Postulated Piping Ruptures in Fluid Systems," and the FSAR Section 3.6.2.1.7 retaining the title "Types of Breaks" and its correct content.
Additionally, NuScale has incorporated the following information regarding non-mechanistic breaks into the FSAR Section 3.6.
Nonmechanistic breaks in containment penetration area BTP 3-3 B.1.a.(1) states:
Even though portions of the main steam and feedwater lines meet the break exclusion requirements of item 2.A(ii) of BTP 3-4, they should be separated from essential equipment. Designers are cautioned to avoid concentrating essential equipment in the break exclusion zone. Essential equipment must be protected from the environmental effects of an assumed nonmechanistic longitudinal break of the main steam and feedwater lines. Each assumed nonmechanistic longitudinal break should have a cross sectional area of at least one square foot and should be postulated to occur at a location that has the greatest effect on essential equipment.
NuScale has identified the containment penetration area for the FWS and MSS as the portions outside containment from the FWS CNV safe end / FWS containment isolation valve (CIV) weld to the FWS check valve/ FWS piping weld and the MSS CNV safe end / MSS piping tee weld to the MSS CIV / MSS piping weld.
The NuScale design has the following characteristics that make nonmechanistic ruptures low risk:
NuScale Nonproprietary
- 1) The essential SSC in the vicinity of the MSS and FWS piping to which break exclusion criteria apply are CIVs, DHRS actuation valves, and instrumentation cables and sensors.
- 2) Unlike safety-related valves in other plant designs that use motor-operators, the NuScale CIVs fail shut upon a loss of electric power or failure of their hydraulic operator lines. The DHRS actuation valves similarly fail open.
- 3) Failure of the NuScale MSS and FWS piping is unlikely because a) Piping in the containment penetration area is made of stainless steel.
b) The physical length of MSS and FWS piping in the containment penetration area is zero (i.e., there are only valves and fittings).
c) The MSS and FWS have a design pressure and temperature of 2100 psi a and 650 degrees Fahrenheit, respectively, in the containment penetration area, similar to the RCS.
The flow area for the nonmechanistic longitudinal break (1 ft 2 ) specified in BTP 3-3 is disproportionately large for application to smaller pipe sizes. NuScale MSS piping is NPS 12 Schedule 120 and the FWS piping is NPS 4 and NPS 5 Schedule 120 in the containment penetration area. For those piping sizes, a 1 ft 2 flow area would be about 159 percent for MSS and 1396 percent for FWS of the area for a full circumferential rupture, which is unrealistic. The NuScale implementation for nonmechanistic breaks of MSS and FWS piping in the containment penetration area considers these design differences from the larger LWR plants. Comparing existing LWR designs to that of NuScale, the NuScale piping flow area is one-eighth to one-twelfth. On this basis, NuScale analyzes for the environmental effects of a MSS break with an area of 12 in 2 , versus 144 in 2 (1 ft 2 ). A FWS break size of 5.87 in 2 is used.
Additionally, BTP 3.4 B.3.(iii) specifies postulating leakage cracks with a flow area of one-half of a pipe diameter by one half pipe wall thickness in piping in the vicinity of essential SSC, regardless of system. This guidance yields an equivalent flow area of 2.7 in 2 for MSS piping, 0.093 in 2 for FWS piping, and 0.199 in 2 for eves piping. Therefore, the effects of leakage cracks are bounded by those of an MSS nonmechanistic break.
The volume under the bioshield is small (roughly a cube 20 ft on a side). Therefore, even outside the containment penetration area, where only leakage cracks are to be considered, the temperature and pressure conditions caused by the nonmechanistic break are used for environmental qualification.
NuScale Nonproprietary
Impact on DCA:
The FSAR Tier 2, Section 3.6 has been revised as described in the response above and as shown in the markup provided in this response .
NuScale Nonproprietary
Protection against Dynamic Effects Associated with Postulated Rupture NuScale Final Safety Analysis Report of Piping permanent piping. The spool piece and subsequent piping are also ASME Class 3 to the junction of an additional valve (or check valve) in each line, and subsequently become ASME B31.1 after that last valve. At the first spool piece breakaway flange, the four lines become part of the eves. Breaks in these lines are postulated in accordance with BTP 3-4 Section B.A.(iii)(2) at intermediate locations where stresses calculated by the sum of equations (9) and (1 O) in NC/ND-3653 of Section Ill of the ASME Boiler and Pressure Vessel Code exceed 0.8 times the sum of the stress limits given in NC/ ND-3653.
RAI 03.06.02-6 Final stress analysis is performed concurrent with fabrication of the first NPM. There are no postulated break locations based upon the current analysis.
RAI 03.06.02-6 Due to the unique nature of the DHRS piping, these lines are specifically discussed in Section 3.6.2.7.
RAI 03.06.02-6, RAI 03.06.02-15 COL Item 3.6-2: A COL applicant that references the NuScale Power Plant design certification will verify that the pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the CNV (under the bioshield) is applicable. If changes are required, the COL applicant will update the pipe rupture hazards analysis, design additional protection features as necessary, and update Table 3.6-2.
RAI 03.06.02-13S1 3.6.2.1.2.1 Non-mechanistic Secondary Line Breaks in Containment Penetration Area RAI 03.06.02-13S1 BTP 3-3 B.1 {a){l) specifies:
RAI 03.06.02-13S1 "Even though portions of the main steam and feedwater lines meet the break exclusion requirements of item 2.A{ii) of BTP 3-4. they should be separated from essential equipment. Designers are cautioned to avoid concentrating essential equipment in the break exclusion zone. Essential equipment must be protected from the environmental effects of an assumed non-mechanistic longitudinal break of the main steam and feedwater lines. Each assumed non-mechanistic longitudinal break should have a cross sectional area of at least one square foot and should be postulated to occur at a location that has the greatest effect on essential equipment."
RAI 03.06.02-13S1 For the NuScale design, the following considerations apply:
- MSS and FWS piping is the largest. high energy piping near containment boundary Tier 2 3.6-14 Draft Revision 3
Protection against Dynamic Effects Associated with Postulated Rupture NuScale Final Safety Analysis Report of Piping
- The lines have a single CIV outside containment in accordance with GDC 57 for lines closed inside containment
- MSS and FWS piping is usually made of less corrosion resistant material than used for the NuScale design. MSS and FWS piping in many pressurized water reactors is carbon or low alloy steel. which has greater susceptibility to degradation than stainless steel.
RAI 03.06.02-13S1 Analyzing for non-mechanistic ruptures provides assurance that multiple essential sscs are capable of withstanding the effects of a limited piping failure should one occur. In the NuScale plant. the dual CIVs are located outside the containment and exposed to the same environmental conditions. which makes protection against unexpected ruptures particularly important. However. the NuScale design has the following characteristics that make non-mechanistic ruptures low risk:
RAI 03.06.02-13S1
- The essential SSCs in vicinity of MSS and FWS piping in the containment penetration area are CIVs. DHRS valves. and instrumentation cables and sensors.
- Unlike some safety-related valves in other plant designs that use motor-operators. the NuScale CIVs are hydraulically held open against pneumatic pressure from an accumulator and shut upon a loss of power or a failure of the hydraulic line. The DHRS actuation valves similarly fail open.
- Failure of MSS and FWS piping is unlikely because:
- Piping in the containment penetration area is made of stainless steel.
- The physical length of MSS and FWS piping in the containment penetration area is zero {i.e .. there are only valves and fittings).
- MSS and FWS piping has a design pressure and temperature of 2100 psi a and 625°F, respectively, eguivalentto the RCS piping.
RAI 03.06.02-13S1 The flow area of 1 ftl specified in BTP 3-3 for a non-mechanistic. longitudinal break is disproportionately large for a small modular reactor with small pipe sizes. NuScale MSS piping is NPS 12 Schedule 120 and FWS piping is NPS 4 and NPS 5 Schedule 120 in the containment penetration area. For those piping sizes. a 1 ftl flow area exceeds the area for a full circumferential rupture. which is physically unrealistic.
RAI 03.06.02-13S1 For the NuScale design. non-mechanistic breaks of MSS and FWS piping in the containment penetration area are evaluated. after consideration of the design differences from larger LWR plants. Comparing the typical PWR pipe MSS flow area to that of NuScale {NPS 30 to 38 vs NPS 12) yields a ratio of one-eighth to one twelfth. On this basis. NuScale analyzes for environmental effects of an MSS non-mechanistic break with an area of 12 inl, versus 1 ftl {144 inl). The Tier 2 3.6-15 Draft Revision 3
Protection against Dynamic Effects Associated with Postulated Rupture NuScale Final Safety Analysis Report of Piping non-mechanistic FWS break size applied for the NuScale design
{NPS 4 and NPS 5) is 5.87 inl~
RAI 03.06.02-13S1 The volume under the bioshield is small: roughly a cube 20 feet on a side.
Therefore. even though only leakage cracks are required to be considered outside the containment penetration area. analysis is performed for a 12 inl_
MSS break at the highest point of the pipe run. resulting in a conservative pressure and temperature profile over time for environmental qualification and bounding breaks occurring in any section of the piping under the bioshield.
RAI 03.06.02-13S1 3.6.2.1.2.2 Break Exclusion RAI 03.06.02-13S1 BTP 3-4 B.A.{iii) identifies specific criteria for which ruptures need not be considered from the containment wall to and including the inboard or outboard isolation valves {usually referred to as the containment penetration area "break exclusion zone"). The concept was necessary due to constraints on ability to cope with breaks between the CIVs. Should a break occur between the CIVs followed by a single failure of a CIV, then containment bypass could occur. To preclude bypass. criteria were developed to ensure that the probability of a piping failure was sufficiently low to make it implausible.
RAI 03.06.02-13S1 The NuScale plant has both CIVs in a single valve body. There are no break locations between the valves. However. the weld between the valve body and the CNV safe end is equivalent to those to which break exclusion applies.
Therefore, NuScale has extended this boundary outside the CNV to include:
RAI 03.06.02-13S1
- The outboard check or excess flow check valve nozzle weld in pressurizer spray. injection. and discharge lines
- DHRS piping welds outside the CNV RAI 03.06.02-13S1 Accordingly. the guidance of BTP 3-4 B.A.{ii) is used in piping design to ensure that breaks and leakage cracks can be excluded in the containment penetration area. BTP 3-3 non-mechanistic breaks of MSS and FWS piping are also addressed. The remaining high energy piping under the bioshield applies BTP 3-4 B.A.{iii) for ruptures and {v) for leakage cracks. Figure 3.6-33 is a representation {not all lines shown) of application of the BTP 3-4 guidance on break location and size. as applied in the NPM bay and the RXB.
RAI 03.06.02-13S1 Tier 2 3.6-16 Draft Revision 3
Protection against Dynamic Effects Associated with Postulated Rupture NuScale Final Safety Analysis Report of Piping The length of piping and number of welds inside the Nu Scale CNV is limited.
For the NuScale design. no primary or secondary piping other than about 160 feet of DHRS piping is within the break exclusion zone outside containment. The design pressure and temperature of MSS. FWS. and DHRS piping in the break exclusion zone is the same as for the RCS.
RAI 03.06.02-13S1 Break exclusion is not applied to any of the piping in the RXB.
RAI 03.06.02-13S1 3.6.2.1.2.3 Leakage Cracks RAI 03.06.02-13S1 Leakage cracks are excluded in containment penetration areas where the criteria of BTP 3-4 B.A.{ii) are satisfied.
RAI 03.06.02-6 3.6.2.1.3 Pipe Breaks in the Reactor Building (outside the Bioshield)
RAI 03.06.02-6 Within the NPM, there are a large number of essential SSC that require protection and relatively small amounts of piping . Therefore, postulated pipe break locations within the NPM or in close proximity to the NPM (i.e., under the bioshield) are specifically addressed by analysis, as discussed in Section 3.6.1.3.
RAI 03.06.02-6 Beyond the NPM, there are fewer SSC that require protection and a large amount of high- and moderate-energy piping (See Table 3.6-1 ). The SSC that require protection are evaluated for effects of line breaks or are separated within compartments of the RXB from areas that contain piping. In addition, the building structure necessary to support the modules and to maintain the integrity of the pool (i.e., the ultimate heat sink) is evaluated.
RAI 03.06.02-6 Piping arrangements in the RXB have not been finalized yet. It is appropriate, therefore, for evaluation of potential rupture locations beyond the reactor pool bay wall, to identify the bounding dynamic effects of postulated breaks and then to determine if protection is required. The approach is to evaluate:
RAI 03.06.02-6
- blast, unconstrained pipe whip, and jet impingement caused by rupture of a main steam pipe.
- subcompartment pressurization, spray wetting, flooding , and other adverse environmental effects caused by main steam or eves breaks that are potentially limiting where they might occur in the building.
- multi-module impacts in common pipe galleries.
Tier 2 3.6-17 Draft Revision 3
RAI 03.06.02-6, RAI 03.06.02-13S1 Figure 3.6-33: Application of BTP 3-4 Break Location Guidance in the NPM bay and RXBNat Used Pipe NPM Bay under bioshield Gallery Any break BTP 3-4 B.A.(iii) and Other Areas Breaks at terminal ends but can leakage exclude elsewhere if acceptance crack criteria met, but subject to leakage cracks unless more stringent criteria are met per BTP 3-4 B.A.(v) w Non-mechanistic Breaks anywhere; analyze for all
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