ML20198S086
| ML20198S086 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/31/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198S084 | List: |
| References | |
| NUDOCS 9901110158 | |
| Download: ML20198S086 (8) | |
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curuqk UNITED STATES p
j NUCLEAR REGULATORY COMMISSION y
WASHINGTON, D.C. 20666-0001 i
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.197TO FACILIT/ OPERATING LICENSE NO. NPF-8 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NO. 2 9
DOCKET NO. 50-368
1.0 INTRODUCTION
By letter dated June 29,1998 (2CAN069804), as supplemented by letters dated December 17, 1998 (2CAN129804), and December 22,1998 (2CAN129805), Entergy Operations, Inc. (EOl or the licensee), submitted a request for changes to the Arkansas Nuclear One, Unit No. 2, Technical Specifications (TSs).
The proposed amendment would revise the as-found lift setting tolerance for the ANO-2 main steam safety valves and the pressurizer safety valves, revise the maximum allowable linear power level-hip' + ip setpoint with inoperable steam line safety valves, and relocate part of the specifications for tuam line safety valves to the ANO-2 Safety Analysis Report (SAR).
Administrative and bases changes would also be made.
The information in the December 17,1998 (2CAN129804), and December 22,1998 (2CAN129805), submittals provided clarifying information and did not expand the scope of the original application as initially noticed, or change the staff's proposed no significant hazards determination published in the FederalReaister on October 21,1998 (63 FR 56242).
2.0 BACKGROUND
The reactor coolant system (RCS) has two pressurizer safety valves (PSVs) to provide overpressure protection during normal power operations (other valves provide protection during low temperature operations). They are direct acting, spring-loaded valves meeting ASME Code requirements. As stated in SAR Section 5.5.13," Reactor Coolant System, Component and Subsystem Design, Safety and Relief Valves," these valves are designed to pass sufficient steam to limit the RCS pressure to 110 percent of the 2500 psia design pressure following a complete loss of turbine generator load without a simultaneous reactor trip.
Overpressure protection for the shell side of the steam generators and portions of the main steam lines is provided by 10 direct acting, spring-loaded ASME Code safety valves (main steam safety valves or MSSVs). As stated in SAR Section 10.3.2, " Steam and Power 9901110158 981231 PDR ADOCK 05000368 P
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Conversion System, Main Steam Supply System, System Description," these valves can pass a steam flow equivalent to a power level of 2900 MWt at the nominal set pressure. With one or i
more of these valves inoperable, TSs (see TS 3/4.7.1, " Plant Systems - Turbine Cycle - Safety Valves," and TS Table 3.7-1) require that reactor power be reduced, and that the maximum allowable linear power level-hign trip setpoint be reduced.
1 3.0 EVALUATION Each TS change proposed by the licensee is discussed below.
3.1 TS 3.4.2. " Reactor Coolant System - Safetv Valves - Shutdown" and TS 3.4.3. " Reactor Coolant System - Safetv Valves - Ooeratina"
,e TS 3.4.2 currently reads j
i A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIA +1, -3%*#.
The licensee proposes to change the upper bound, so that TS 3.4.2 would read A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIA +/- 3%*#.
TS 3.4.3 currently reads All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIA
+1, -3%*.
'The licensee proposes to change the upper bound, so that TS 3.4.3 would read All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIA +/-
3%*.
The "*" footnote for each TS reads
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. If found outside of a +/- 1% tolerance band, the setting shall be adjusted to within +/- 1% of the lift setting shown.
and would not be changed.
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TS 4.0.5 (surveillance requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components) specifies that the PSV and MSSV testing must conform to the ASME I.
Boiler and Pressure Vessel Code and applicable addenda. The Code (Section XI,1986 Edition, l
and Addenda through 1987) allows for as-found PSV and MSSV lift setting tolerances of +/- 3%
i when the licensee can demonstrate that the valves will still perform their safety function at these l
limits.
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2 Increasing the upper bound of the lift setting tolerance of the PSVs (during operations and while shutdown) from +1% to +3%, with the caveat that this increase would be for as-found conditions only, will allow normal surveillance testing of the PSVs (with no reporting requirements) to be l
within +3% of the nominallift setpoint of 2500 psia.
The licensee performed the analyses to justify the proposed changes. EOi stated that a l,.
feedwater line break is the accident which most significantly challenges the PSVs. The
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licensee's current calculated peak pressure for this event is 2742 psia (SAR Table 15.1.14-20),
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which is acceptable because it does not exceed 110% of the design value of 2500 psia (that is, l
2750 psia). This event was reanalyzed for Cycle 13 operation (and bounds the upcoming Cycle
- 14) for several different steam generator tube plugging levels, and accounted for an actual PSV lift setting of +3% of nominal. The assessment was performed using the Combustion
. f Engineering Nuclear Transient Simulation (CENTS), which is an NRC-approved methodology.
The licensee confirmed that all the conditions of the NRC SER approving the methodology were satisfied. Additionally, the licensee utihred appropriate conservative assumptions that were originally reviewed and approved by the NRC staff in the issuance of Amendment Nos.189 and 190. The licensee determined that the peak RCS pressure would be 2730.1 psia.
The licensee stated that the limiting peak RCS pressure anticipated operational occurrence (AOO) is the loss-of-condenser-vacuum (LOCV) event. The licensee's current calculated peak pressure for this event is 2671 psia (SAR Table 15.1.7-2), which is acceptable because it does not exceed 110% of the design value of 2500 psia, which is 2750 psia. This event was also reanalyzed using the CENTS simulation for different tube plugging levels, and included the i
proposed +3% lift setting tolerance. The results of this assessment, which utilized conservative assumptions and for which the licensee confirmed that all conditions of the NRC SER were satisfied, was a predicted peak pressure of 2683 psia, which is lower than the predicted pressure for the feedwater line break accident, and hence this AOO does not result in the limiting pressure for the PSV lift setting tolerance change.
i Since the calculated peak pressure due to the limiting event (feedwater line break accident) does not exceed 110% of the design value of 2500 psia (2750 psia), and since the calculation was performed using an NRC-approved code with appropriate conservative assumptions as determined by the staff in the approval of Amendment Nos.189 and 190, and since the higher lift setting tolerance satisfies the ASME code requirements, and since the as-left setting tolerance will remain +/- 1% of nominal, the increase in the as-found tolerance for the PSV lift pressure to +3% is acceptable for TS 3.4.2 and TS 3.4.3.
3.2 TS Table 3.7-5. " Steam Line Safetv Valves"- Lift Settina Tolerance Chanae TS Table 3.7-5 currently specifies the steam line safety valve lift setting upper bourd as +1% of nominal pressure. The licensee proposes to change the upper bound to +3% of nominal.
Additional information is provided in a footnote which reads:
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. If found outside of a +/- 1% tolerance band, the setting shall be adjusted to within +/- 1% of the lift setting shown.
4-(This is identical to the PSV lift setting footnote) and would not be changed. Increasing the upper bound of the lift setting tolerance of the MSSVs from +1% to +3%, with the caveat that this
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increase would be for as-found conditions only, will allow normal surveillance testing of the MSSVs (with no reporting requirements) to be within +3% of the nominal lift setpoint of 2500 psia.
l The licensee stated that a feedwater line break is also the accident which most significantly challenges the MSSVs. The licensee's current calculated peak pressure for this event is approximately 1100 psia (SAR Figure 15.1.8-3), which is acceptable because it does not exceed l
110% of the design value of 1100 psia (that is,1210 psia). This event was reanalyzed for Cycle l
13 operation (and bounds the upcoming Cycle 14), and accounted for actual MSSV lift settings lI of +3% of nominal. The assessment was performed using the CENTS methodology, and utilized appropriate conservative assumptions. In addition, EOl confirmed that all of the SER conditions were satisfied. The licensee determined that the peak secondary pressure would be 1163.8 l
psia.
The licensee stated that the limiting peak RCS pressure AOO is the LOCV event. The licensee's current calculated peak secondary pressure for this event is 1144 psia (SAR Table l
15.1,7-2), which is acceptable because it does not exceed 110% of the design value of 1100 i
psia (that is,1210 psia). This event was also reanalyzed using the CENTS simulation, including the proposed +3% lift setting tolerance. The results of this assessment, which utilized conservative assumptions as determined by the staff with the approval of Amendments Nos.189 l
and 190 and for which the licensee confirmed that all of the SER conditions were satisfied, was a predicted peak secondary pressure of 1195 psia, which is higher than the predicted pressure for the feedwater line break accident, and hence this AOO leads to the limiting secondary pressure for the MSSV lift setting tolerance change.
l The increase in the MSSV as-found lift pressure impacts the peak clad temperature (PCT), and l
consequently the cladding oxidation, during a small-break loss-of-coolant accident (SBLOCA).
The licensee determined that the increase in MSSV lift pressure results in a higher steam generator (SG) pressure, which results in a higher RCS pressure during a SBLOCA. The higher RCS pressure leads to decreased safety injection flow and increases the break flow, resulting in a higher PCT.
l To ensure that the criteria of 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling l
Systems [ECCSs] for Light-Water Nuclear Power Plants," continue to be satisfied by the ANO-2 l.
ECCSs with higher lift-setting MSSV tolerances, the SBLOCA was reanalyzed using the NRC-approved methodology described in CENPD-137, Supplement 2-P-A, Calculative Methods for the ABB CE SmallBreak Evaluation Model(S2M). The licensee confirmed that there are no conditions in the NRC SER approving the methodology that need to be specifically ' addressed.
The licensee provided a comparison between the S1M and S2M models in their November 24, 1996 (2CAN119610) submittal for Amendment No.179. This submittal contained an analysis which demonstrated the sensitivity of these codes to break size and steam generator pressure.
This information was used in determining the limitinc brcd size of 0.05 ft as described in SAR 2
Section No. 6.3.3.2.3.5 and demonstrated that the 0.05 ft break size remained bounding for both the S1M and S2M models. Since this sensitivity study was performed using the +3% main steam safety valve lift setting tolerance as stated in the licensee's letter dated December 22,
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'\\ 1998 (2CAN129805), the results of this analysis are appropriate for the identification of the limiting break size. The licensee also indicated in their December 22,1998 (2CAN129805) letter that the same inputs and assumptions associated with this previous analysis were conservative and remain applicable for the proposed change.
The results of the calculated parameters used to satisfy criteria 1,2, and 3 of 10 CFR 50.46 are summarized below:
Current New Parameter GuinDQD Value Value Peak Clad Temperature
< 2200 2011 1798 I
(degrees Fahrenheit)
Maximum Cladding Oxidation (percent)
< 17 5.47 4.8 Core-Wide Cladding Oxidation (percent)
<1
< 0.835
< 0.36 Though the increase in MSSV lift pressure results in a higher steam generator (SG) pressure, which results in a higher RCS pressure during a SBLOCA, the new calcelated PCT is actually
. lower than the current value due to the differences in the calculation methodologies. Based on their analysis as summarized by the information provided above, the licensee has concluded that the proposed change satisfies criteria 4 and 5 of 10 CFR 50.46. The NRC staff has reviewed the information provided and concluded that the proposed change meets the requirements of 10 CFR 50.46.
Since the calculated peak pressure due to the limiting event (LOCV AOO) does not exceed 110% of the design value of 1100 psia (1210 psia), and_since the calculation was performed. _.
using an NRC-approved code with appropriate conservative assumptions, and since the higher lift setting tolerance satisfies the ASME code requirements, and since the criteria of 10 CFR 50.46 continue to be satisfied, and since the as-left setting tolerance will remain +/- 1%
of nominal, the increase in the as-found tolerance for the MSSV lift pressure to +3% is acceptable. Future changes to input parameters which affect this analysis will be handled
- through the core operating limits report (COLR) process which ensures the appropriate configuration controls and regulatory processes are utilized.
3.3 TS Table 3.7-5. " Steam Line Safety Valves"- Removal of Orifice Size TS Table 3.7-5 currently specifies the orifice size of each MSSV. Tne licensee noted that this information is already contained in the SAR, and proposes to remove this information from the TS table..
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( Criteria for inclusion of requirements in the TSs is provided in 10 CFR 50.36, as follows:
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Criterion 1:
l l-Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2:
A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
lV Criterion 3:
'A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
l Criterion 4:
A structure; system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.
Requirements that are in the existing TSs, but do not meet the criteria set forth in 10 CFR 50.36 for inclusion in TS, can be relocated to appropriate licensee-controlled documents.
The MSSVs are part of the primary success path to mitigate accidents and transients as described in Criterion 3. The lift settings of the MSSVs are critical to the proper functioning of the safety valve, and will be maintained in the TSs. Verification of the safety valve lift setting provides assurance that the valve will function es appropriate to mitigate a pressure transient.
The orifice size is a sub-component design detail that does not meet the criteria established in l
10 CFR 50.36 for the inclusion of requirements in the Technical Specifications. The orifice is a passive sub-component which controls the maximum relief capacity of the safety valve after the valve has fully lifted and therefore does not satisfy Criterion 3. Criterion 1,2, and 4 do not apply l
to the MSSVs as the safety valve does not provide indication, does not establish the initial L
conditions for a design basis accident or transient, and has not shown to be significant to public j
health and safety. The orifice size is listed in the SAR along with other design information l
associated with the MSSVs. The SAR is the appropriate licensing document to contain the
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P design detail describing the technical attributes of a component that support the execution of its i
design function. The staff has determined that, since the orifice size does not meet the criteria described above, removal of the orifice size from TS Table 3.7-5 is consistent with the requirements of 10 CFR 50.36. The description of the orifice size in the SAR ensures that the process specified in 10 CFR 50.59 will provide the appropriate controls for future changes.
Therefore, the proposed removalis acceptable.
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. 3.4 TS Table 3.7-1. " Maximum Allowable Linear Power Level-Hiah Trio Setooint With inocerable Steam Line Safetv Valves Durina Ooeration With Both Steam Generators" The maximum allowable linear power level-high trip setpoint values provided in this table are currently determined using a calculation based on MSSV relieving capacity, as stated in TS Bases 3/4.7.1.1. The licensee is proposing to include a second calculation method based on the analysis of the LOCV event (assuming +/- 3% lift setting tolerance). This AOO leads to the limiting secondary pressure. In addition, the licensee proposes to utilize the more conservative trip setpoint from the two methods.
kb The LOCV event was analyzed using the NRC-approved CENTS methodology which is appropriate for use at ANO. The licensee confirmed that the SER conditions associated with the approval of this methodology have been satisfied. Since an approved code was used and the SER conditions were satisfied, the use of this second methodology is acceptable. In addition, the continued use of the current methodology together with the new methodoiogy is acceptable because the most conservative setpoint is being utilized. Therefore, the proposed change is acceptable.
3.5 TS 6 9.5.1. " Administrative Control-Core Ooeratina Limits Reoort." Analvtical Methods This TS provides a list of the NRC-approved analytical methods used to determine core operating limits. The licensee proposes to add a reference to " Calculative Methods for the CE Small Break LOCA Evaluation Model"(Topical Report CENPD-137, Supplement 2-P-A, dated April 1998) and to renumber the other references as appropriate.
Since this methodology has been approved by the NRC, is appropriate for use at ANO, and will ensure that values for cycle specific parameters will be determined such that applicable limits of the plant safety analysis are met, the staff finds it acceptable to reference in the TSs for use in the COLR process.
3.6 TS Bases 3/4.7.1.1. " Plant Svstems - Bases - Turbine Cvele - Safety Cveles" The licensee has proposed changes to this bases section related to the MSSV lift settings, reactor trip settings and setpoint methodology. Since these changes are consistent with the TSs, they are acceptable. In addition, the licensee is removing relief valve capacity information from the TS Bases. The staff has concluded that this change is acceptable since it is consistent with NUREG-1432 and this information is already captured in the SAR.
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4.0 STATE CONSULTATION
1 In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility l
component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 56242). This amendment also changes reporting or record i
p keeping requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by l
operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: A. Hansen C. Nolan Date: December 31, 1998
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