1CAN119608, Forwards Proposed ANO-1 TS Amend Request That Revises Svc Period & Associated Figures IAW TS 3.1.2.7,3.1.2.8 & 10CFR50,App G,For Review & Approval
| ML20135E797 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/26/1996 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20135E801 | List: |
| References | |
| 1CAN119608, NUDOCS 9612120070 | |
| Download: ML20135E797 (15) | |
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ENTERGY inter-Office Correspondence November 26,1996 1CAN119608 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555
Subject:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specification Change To The Reactor Coolant System Pressure And Temperature Curves Gentlemen:
Attached for your review and approval is a proposed Arkansas Nuclear One-Unit 1 (ANO-1)
Technical Specification (TS) amendment request that revises TS 3.1.2. The purpose of this amendment request is to revise the service period and the associated figures in accordance with TS 3.1.2.7,3.1.2.8, and 10 CFR 50 Appendix G. As required by TS 3.1.2.8, this submittal is being made at least 90 days prior to the end of the current service period. The current service period has been evaluated for up to 15 Effective Full Power Years (EFPY).
Based on 100% full power operation, the current service period will end in early March of 1997. If the revised service period limits are not approved by the end of 15 EFPY, ANO will administratively control the operation of the facility to the more restrictive of the current and proposed service period pressure / temperature limits. For the reactor coolant heatup curve the administrative controls will be the more restrictive of the current pressure / temperature limits and the proposed 50 F heatup curve.
The proposed amendment request includes an exemption request pursuant to 10 CFR 50.12 from certain requirements of 10 CFR 50.60. This exemption request will allow an alternate methodology (Code Case N-514) to determine the low temperature overpressure protection (LTOP setpoints for ANO-1. The content of this Code Case has been incorporated and published in the 1993 Addenda to Appendix G of ASME Section XI of Code. However, 10 CFR 50.55a and the applicable Regulatory Guides have not been updated to reflect the acceptability of Code Case N-514 and therefore requires this exemption request for its use.
This submittal also closes out an outstanding issue related to Generic Letter 92-01, 1
Revision 1, Supplement 1.
In the letter dated August 27,1996, (ICNA089601) from goi G. Kalman to J. Yelverton, the Staff required ANO to provide an assessment of the I
application of the ratio procedure to the pressure / temperature limit curves and LTOP limits as described in Regulatory Guide 1.99. The required assessment is i cluded in this submittal.
96 120070 961126
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U. S. NRC November 26,1996 1CANI1%08 Page 2 The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.
Entergy Operations requests the effective date for this change be within 30 days ofissuance.
Although this request is neither exigent nor emergency, your prompt review is requested.
Very truly your,
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NI/ de Attachments To the best of my knowledge and belief, the statements contained in this submittal are true.
SUBSCRIBED AND SWORN TO before me, a Notary Public in and for bdu2m County and the State of Arkansas, this Mlo day of Ittxmh
,1996.9 O /nin M -
Mo or,cu a V0 JUANA M.TAPP Notary PubTay My Commission Empires //-?- SGtb "mIEA*e$m" wcommi..one air -ii.e.nooo l
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i U. S. NRC November 26,1996 1CANI19608 Page 3 i
cc:
Mr. Leonard J. Callan l
Regional Administrator U. S. Nuclear Regulatory Commission Region IV l
611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 l
NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Mr. George Kalman NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 l
l Mr. Bernard Bevill Acting Director, Division of Radiation I
l Control and Emergency Management i
l Arkansas Department ofHealth 4815 West Markham Street Little Rock, AR 72205 l
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ATTACHMENT i
E ICAN119608 PROPOSED TECHNICAL SPECIFICATION AND I
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i RESPECTIVE SAFETY ANALYSES -
IN THE MATTER OF AMENDING LICENSE NO. DPR-51 t
i ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT ONE l
DOCKET NO. 50-313
.1
Attachment to ICAN119608 Page1of11 DESCRIPTION OF PROPOSED CIIANGES Technical Specification (TS) 3.1.2.7 was revised to reflect the new service period of 32 Effective Full Power Years (EFPY).
TS 3.1.2.7 and 3.1.2.8 were revised to remove the requirements associated with e
10 CFR 50 Appendix G, Sections V.B and V.C because these sections have been removed from 10 CFR 50 Appendix G.
TS 3.1.2.10 was revised to reflect the new LTOP enable temperature of 262 F.
The bases for specification 3.1.2 was revised to reflect the updated calculation and e
document references and the new service period of 32 EFPY. A sentence was added to the 3.1.2.10 bases to include the use of Code Case N-514 and to specify that instrument error is not included in the LTOP enable temperature of 262*F. The bases section was also modified to reflect that the curves were based on all " allowed" operating reactor coolant pump (RCP) combinations not all combinations as was the case with the 15 EFPY figures. The allowed RCP combinations and their associated temperature limits are listed on the applicable figures.
Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 were revised to reflect the new inservice leak and hydrostatic testing, heatup, and cooldown curves and the restrictions associated with the new service period of 32 EFPY.
DISCUSSION OF CIIANGE The purpose of this amendment request is to revise the service period and the associated pressure and temperature (P/f) limits in accordance with TS 3.1.2.7 and 3.1.2.8. The current service period has been evaluated for up to 15 EFPY. The Pff limitations were evaluated for the 21 and 32 EFPY service periods. The 32 EFPY service period was selected due to the minimal gain in the operating area that would be achieved by selecting the 21 EFPY service period. Therefore, a new set of P/T limits have been developed to assure reactor vessel integrity for up to 32 EFPY and are included in this amendment request. These limits are reflected in the figures for inservice leak and hydrostatic testing, heatups, and cooldowns that are proposed in TS Figures 3.1.2-1,3.1.2-2, and 3.1.2-3 respectively.
The revised P/T curves are not adjusted for instrument error and will therefore not be used for plant operations. However, the P/T curves used for plant operations will be adjusted for instrument error and placed in the plant operating procedures.
l The methodologies used to develop these limitations are consistent with the requirements of 10 CFR 50 Appendix G. The methods and criteria employed to establish operating P/T limits are described in topical report BAW-10046A. Attachment 2 to this submittal contains FTI document 77-1258569-01 "ANO-1 Pressure-Temperature Limits For 32 EFPY" which contains the summary of how these limits were established.
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Attachment to l
ICANI19608 Page 2 of1I Instrument error was not included in the new LTOP enable temperature of 262 F in proposed TS 3.1.2.10 for consistency with the inservice leak and hydrostatic testing, heatup, and cooldown figures.
Flexibility in instrumentation loop component replacements without requiring a TS change is also achieved by use of this methodology. This value will not be l
used for plant operation until instrument error is applied in the operating procedures.
l The bases for specification 3.1.2 was revised to reflect the updated calculation and document references and the new service period of 32 EFPY. FTI document number 77-1258569-01 replaced BAW-2106 and is included with this submittal as Attachment 2.
BAW-2121P replaced BAW-1511P and FTI calculations 32-1245917-00 and 32-1257716-00 replaced BAW-2075. FTI calculations 32-1245917-00 and 32-1257716-00 will be submitted to the staff under a separate cover letter. In addition, a sentence was added to the 3.1.2.10 bases to include the use of Code Case N-514 and to specify that instrument error is not included in the l
LTOP enable temperature of 262 F.
Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 were revised to reflect the new service period and associated limits of 32 EFPY based on FTI document number 77-1258569-01. The notes have been added below the new curves reflecting the applicable RCP operating restrictions.
The notes also include the new heatup and cooldown limitations. The first note on each of the 15 EFPY figures stating "the acceptable pressure and temperature combinations are below and to the right of the limit curves" was removed. This note was replaced by the label on each of the figures showing the " acceptable region." The second sentence of the same note is repeated in the bases and therefore removed to minimize the notes on each of the curves.
ANO-1 REACTOR VESSEL FLUENCE PROGRAM
-b In 1976 several utilities, including Arkansas Power and Light (now Entergy Operations), with reactor vessels designed by Babcock & Wilcox requested exemptions from the 10 CFR 50, Appendix H requirement for an in-vessel material surveillance program. The Staff reviewed and evaluated each licensee's request for an exemption and the plan for an integrated surveillance program. The staff then granted the requested exemption.
A revised 10 CFR 50 Appendix H became effective in July 1983.Section II.C of the revised Appendix H allows an integrated surveillance program provided it was approved by the Director, Office of Nuclear Reactor Regulation. The revised Appendix H provided criteria that was to be used in the evaluation of the surveillance program.
In a letter dated March 14, 1984, the B&W Owners Group submitted an updated integrated surveillance program for Staff review and approval. The program was documented in BAW-1543, Revision 2. In the SER for BAW-1543A, the Staff concluded that the topical report meets the evaluation criteria of Section II.C of Appendix H. The following statements are from the conclusions section of the SER.
In-cavity dosimetry testing should continue in order to reduce uncertainties in neutron fluence for vessels that do not contain in-vessel dosimetry. If these test results provide an effective method of monitoring vessel neutron fluence, the in-cavity dosimetry should be incorporated in plants.
Attachment to l
l 1CANI19608 l
Page 3 of11
'I In a letter dated September 16,1985, the B&W Owners Group requested an evaluation of a l
" Cavity Dosimetry Program" that was under development for use in B&W plants. This program is described in topical report BAW-1875. This report was approved by the Staffin l
June 1986. It was noted in BAW-1875 that the material surveillance program will have l
l provided all the required empirical information for the fluence-toughness relationship by the i
early 1990's when most or all of the surveillance capsules would have been removed. The l
cavity dosimetry program will then continue to provide vessel irradition data beyond the end I
of the capsule dosimetry program in an accurate and convenient manner.
The calculational methodology for predicting the fluence using the cavity dosimetry was
.l validated in the benchmark phase of the cavity dosimetry program. The benchmark consisted of both capsule and cavity comparisons of calculations to dosimetry measurements. The results of these benchmarks are documented in an FTI topical report BAW-2205-00. The l
results demonstrate that the conclusions in the ANO-1 fluence data included in Attachment I of this submittal for the fluence analysis are sufficient for safety and licensing evaluations of reactor vessel embrittlement.
I Pressurized Thermal Shock (PTS) 1 As required by the NRC Safety Evaluation for the original ANO-1 PTS evaluation, ANO-1 i
has re-evaluated RTvrs as part of generating the new Appendix G limits. This evaluation was performed in accordance with 10 CFR 50.61. It should be noted that all the referenced BAW reports listed in this section have been previously submitted to the Staff.
The welds and materials ofinterest are listed in Table 1. Listed as well are the extrapolated fluences at 32 EFPY for the inside surface. The discussion of the determination of the inside surface fluence is provided in Attachment I to this submittal. Weld SA-1788 (Lower Shell to Dutchman weld), was removed from the evaluation given its location and subsequently lower fluence value (the fluence is less than 10" n/cm ),
2 Table 1 L
RV Beltline Materials and Fluence Values at 32 EFPY 1
Beltline Materials Material Fluence,Inside l
Identity Surface (n/cm )
j 2
Nozzle Belt Forging AYN 131 7.90E+18 Upper Shell Plate C5120-2 8.71 E+18 Upper Shell Plate C5114-2 8.71E+18 Lower Shell Plate C5120-1 8.58E+18 j
l Lower Shell Plate CSI14-1 8.58E+18 i
NB to US Cire. Weld (100%)
WF-182-1 7.90E+18 US Longit. Weld (Both 100%)
WF-18 6.07E+18 US to LS Cire. Weld (100%)
WF-112 8.35E+18 LS Longit. Weld (Both 100 %)
WF-18 6.06E+18
Attachm:nt to i
ICANI19608 j
Page 4 of 1I
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The PTS reference temperature is determined by the following correlation from 10 CFR 50.61:
RTns = I + M + ARTns j
where I is the initial reference temperature (RTer) for an unirradiated material and M is a margin term to cover uncertainties in the values of the initial RTmr, copper and nickel contents, fluence and the calculation procedures. ARTns is the mean value of the adjustment in reference temperature caused by irradiation. The following correlation is used to calculste the ARTns.
ARTm = (CF)f* "' ' 8 0 j
l In this expression, CF is the chemistry factor and is a function of the copp.er and nickel content for base metals and welds, as tabulated in 10 CFR 50.61. The term fe.2 tioios o is the fluence factor,ff, where fis the best estimate neutron fluence, in units of 10" n/cm, at the 2
clad-base-metal interface on the inside surface Sthe vessel at the location where the material receives the highest fluence.
The copper and nickel content for the plates and forging can be found in BAW-1820. This report also contains the measured initial RTwor alues for the plates. The initial RTer value v
for the forging is located in BAW-10046P and is an estimated value. The generic value of
-5 F for the initial RTuor from BAW-1803, Revision 1, is used in this analysis for the welds.
The copper and nickel content of the welds can be found in BAW-2121P.
The margin calculation is done to obtain conservative, upper bound values for the RTns calculation. It is calculated as the square root of the sum of the squares of the standard deviations for the initial RTer (oi) and the ARTns (ca). If a measured value ofinitial RTer for the material in question is available, ci is to be estimated from the precision of the test method. If not, and generic mean values for that class of material are used, or is the standard deviation obtained from the set of data used to establish the mean. The standard deviation ca for ARTns, is 28 F for welds and 17 F for base metal. The margin term is then calculated according to the equation below and the results are listed in Table 2.
M = Margin = 2]o' + a 2
However, in accordance with 10 CFR 50.61, ca need not exceed 0.50 times the mean value of ARTns. This alternate approach is applied only to the Nozzle Belt Forging, AYN 131, where the value of oa can be taken as the value of ARTns/2. The margin terms prescribed for each l
material are listed in Table 2.
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Attachment to i
1CAN119608 Page 5 of11 Table 2 RTrrs Margin Values Beltline Materials Material ai ca Margin Identity (F)
( F)
(F)
Nozzle Belt Forging AYN 131 31 9.5 65 Upper Shell Plate C5120-2 0
17 34 Upper Shell Plate C5114-2 0
17 34 Lower Shell Plate C5120-1 0
17 34 Lower Shell Plate C5114-1 0
17 34 l
l NB to US Cire. Weld (100%)
WF-182-1 20 28 69 UT Longit. Weld (Both 100%)
WF-18 20 28 69 US to LS Cire. Weld (100%)
WF-112 20 28 69 i
WF-18 20 28 69 i
The values for initial RTmyr, ARTvrs, and Margin terms are listed in Table 3 with the resultant values for RTns. Table 3 indicates weld WF-112 is the limiting weld for ANO-1 and is significantly lower than the screening criteria of 300 F.
Table 3 l
RTrrs Calculation Beltline Materials Matl.
I Margin ARTvrs
-RTvrs Identity (F)
("F)
(*F)
( F)
Nozzle Belt Forging AYN 131 3
65 19 87 Upper Shell Plate C5120-2
-10 34 118 142 Upper Shell Plate C5114-2
-10 34 102 126 Lower Shell Plate C5120-1
-10 34 117 141 Lower Shell Plate C5114-1 0
34 101 135 NB to US Cire. Weld (100%)
WF-182-1
-5 69 166 230 US Longit. Weld (Both 100%)
WF-18
-5 69 131 195 US to LS Cire. Weld (100%)
WF-112
-5 69 187 251 LS Longit. Weld (Both 100 %)
WF-18
-5 69 131 195 l
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Attachment to 1CANI19608 Page 6 of11 Close Out Of Generic Letter 92-01. Revision 1. Sunnlement i In letter dated August 27,1996 from G. Kalman to J. Yelverton addressing the close out of Generic Letter 92-01, Revision 1, Supplement 1, ANO was required to demonstrate that the application of the ratio procedure described in Position 2.1 of Regulatory Guide 1.99, Revision 2 would not cause the RTm values for the limiting material to exceed the screening i
criteria. The Stafr required ANO to provide an assessment of the application of the ratio l
procedure to the P/T limit curves and LTOP limits. This section provides the required i
assessment.
l Three of the materials listed in Table 4 have surveillance data to apply the ratio procedure for l
the ANO-l' vessel. These materials are the base metal plate heat number CSI14-1 and weld metals WF-182-1 and WF-112. When applying the credibility criteria as defined in 10 CFR 50.61 to the data sets for these materials, the scatter of the measured ARTmyr values is greater than 17 F for some of the base metal plate data, and 28*F for weld wires data. However, the l
scatter for these data points are less than twice the values listed above. For these materials a i
two standard deviation (aa) from the generic database (i.e.,34 F for base metals and 56 F for weld metals) were used in the margin term to ensure that all the surveillance data is bounded.
When there is clear evidence the copper or nickel content of the survdll2nce weld differs from i
that of the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and surveillance weld, the measured values of ARTm should be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld to that for the l
surveillance weld (the ratio procedure). A material specific chemistry factor is determined for j
the disparate materials to determine the appropriate ratio by employing the below listed
. formula.
CF = [(ARTmyr
- ff)
Z(frq Base metal plate heat number C5114-1: The ANO-1 plant specific reactor vessel surveillance program (BAW-2075, Revision 1) provides data for predicting the reference temperature shift for the base metal plate heat number C5114-1. The chemistry factor used to calculate the RTm in Table 3 was 105.6. A revised chemistry factor of 51.2 was determined for plate C5114-1 by applying the ratio procedure. This revised chemistry factor was used in the calculation of RTm nd is listed in the 32 EFPY column in Table 4. This EOL value is a
significantly below the screer.ing criterion listed in 10 CFR 50.61 of 270*F for plates.
Weld metal WF-182-1:
The Master Integrated Reactor Vessel Surveillance Program l
(BAW-1543, Revision 4), provides the surveillance data for weld metal WF-182-1 on l
predicting the reference temperature shift. BAW-2125 and BAW-2190 provides relevant surveillance data for this weld. The chemistry factor used to calculate the RTm in Table 3 was 177.95. A revised chemistry factor of 172.7 was determined for weld metal WF-182-1 by applying the ratio procedure. This revised chemistry factor was used in the calculation of RTm and is listed in the 32 EFPY column in Table 4. This EOL value is significantly below the screening criterion listed in 10 CFR 50.61 of 300 F for circumferential weld materials.
Attachment to ICANI19608 Page 7 of11 l
Weld metal WF-112: BAW-1543, Revision 4 also provides the surveillance data for weld metal WF-ll2 on predicting the reference temperature shift. The relevant surveillance data for weld WF-112 can be found in BAW-2075, Revision 1, BAW-2050, BAW-2074, BAW-1920P, and BAW-2140. The chemistry factor used to calculate the RTns in Table 3 was 196.7. A revised chemistry factor of 185.6 was determined for weld metal WF-ll2 by applying the ratio procedure. This revised chemistry factor was used in the calculation of RTne and is listed in the 32 EFPY column in Table 4. This EOL value is significantly below the screening criterion listed in 10 CFR 50.61 of 300 F for circumferential weld materials.
Table 4 RTrrs Comparison with 10 CFR 50.61 Screening Criterion RTns ( F)
Beltline Materials Material 32 Screening Identity EFPY Criterion t
Nozzle Belt Forging AYN 131 87 270 Upper Shell Plate C5120-2 142 270 Upper Shell Plate C5114-2 126 270 Lower Shell Plate C5120-1 141 270 Lower Shell Plate C5114-1 83 270 NB to US Circ. Weld (100%)
WF-182-1 225 300 US Longit. Weld (Both 100%)
WF-18 195 270 US to LS Cire. Weld (100%)
WF-112 240 300 LS Longit. Weld (Both 100 %)
WF-18 195 270 It should be noted that weld WF-112 remains the limiting material for the ANO-1 vessel and there is significant margin to the screening criterion. Also note that the reference temperature was reduced from 251*F to 240 F when the ratio procedure _was used for this material. The reduction in the reference temperature values is allowed by 10 CFR 50.61. The 32 EFPY proposed P/T limit curves and LTOP limits were developed using the material specific chemistry factors determined above. Therefore, the use of the ratio procedure in this submittal will allow closure of Generic Letter 92-01, Revision 1, Supplement I for ANO-1.
LTOP Pft' Limits The low temperature transient P/T limits provide pressure restrictions for the protection l
against non-ductile failure of the RCS under transient conditions. The LTOP system protects the LTOP P/T limits. The design basis event for the LTOP system is a failure of the makeup system control valve to the full open position.
Attachment to ICANI19608 Page 8 of 11 The LTOP transient P/T limits have been calculated in accordance with the methodology described in AShE Section XI Code Case N-514. ASME Code Case N-514 was prepared to provide an alternate approach to AShE Section XI Appendix G for determining loads and temperature conditions during reactor startup aad shutdown. NRC Regulatory Guides 1.84, 1.85, and 1.147 list the AShE Code Cases that have been approved by the NRC. Code Case N-514 has not been added to these regulatory guides to date, although it has been previously l
approved for use at other facilities including ANO-2. However, the content of Code Case N-l 514 has been incorporated into Appendix G of Section XI of the AShE Code and published i
in the 1993 Addenda to Section XI.10 CFR 50.60 allows licensees to use proposed l
alternatives to the requirements of 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.12. Therefore, pursuant to 10 CFR 50.12, an exemption to 10 CFR 50.60 is requested to use Code Case N-514 for ANO-1. Justification for use of the code case is provided in the next section of this submittal.
The code case limits the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G, paragraph G-2215 of ASME Section XI, Division 1 as a design limit.
Code Case N-514 defines the enable temperature as the greater of the coolant temperature corresponding to a metal temperature of RTer +50 F or a minimum of 200 F. As discussed i
in Attachment 2, RTer has been determined to be 212 F. Therefore, this resulted in a LTOP enable temperature of 262 F (212 F + 50 F) for the limiting weld material (WF-112). The new LTOP enable temperature is included in this amendment request in specification 3.1.2.10.
The isothermal conditions were used to develop the LTOP setpoint. This is consistent with the Westinghouse standard methodology (WCAP-14040, Revision 1) that endorses the use of l
110% of the isothermal curve as the LTOP design limit. This methodology has been previously reviewed and approved by the staff.
The value of 262 F plus instrument uncertainty will be added the plant operating procedures.
l JUSTIFICATION FOR ASME CODE CASE N-514 EXEMPTION REOUEST The following provides the basis for the exemption request to 10 CFR 50.60 for use of AShE i
Section XI Code Case N-514, " Low Temperature Overpressure Protection Section XI, Division 1"in lieu of10 CFR 50, Appendix G.
10 CFR 50.12 Reauirements: The requested exemption to allow the use of AShE Code Case N-514 for determining the LTOP enable temperature meets this criteria as discussed below.
10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that:
i l.
The reauested exemption is authorized by law: No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendices G and H when an exemption is granted by the Commission under 10 CFR 50.12.
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Attachment to 1CANI19608 Page 9 of11
- 2. The reauested exemption does not oresent an undue risk to the oublic health and safety:
l A revised LTOP relief valve enable temperature is being proposed for ANO-1. The enable j
temperature has been developed to provide bounding low temperature reactor vessel integrity j
protection during the LTOP design basis transient. The LTOP setpoint will utilize 110% of j
the pressure determined to satisfy Appendix G, paragraph G-2215 of ASME Section XI, Division - 1 as a design limit.
The approach is justified by consideration of the l
ovegressurization design bads events and the resulting margin to reactor vessel failure.
Restrictions on allow < ' operating conditions and equipment operability requirements have l
been established to ensure that operating conditions are consistent with the assumptions of the accident analysis. Specifically, RCS pressure and temperature must be maintained within the heatup and cooldown rate dependent pressure / temperature limits specified in TS 3.1.2.
Therefore, this exemption does not present an undue risk to the public health and safety.
3.
The reauested exemption will not endanner the common defense and security: The common defense and security are not endangered by this exemption request.
- 4. Special circumstances are present which necessitate the reauest for an exemotion to the renulations of 10 CFR 50.60: Pursuant to 10 CFR 50.12(aX2), the NRC will consider granting an exemption to the regulations if special circumstances are present. This exemption meets the special circumstances of paragraphs:
l (a)(2Xii).- demonstrates that the underlying purpose of the regulation will continue to be achieved; j
(aX2Xiii) - would result in undue hardship or other cost that are significant if the regulation is l
enforced and; i
(aX2Xv) - will provide only temporary relief from the applicable regulation and the licensee has made good faith efforts to comply with the regulations.
i 10 CFR 50.12(aV2Viih ASME Code Case N-514 recognizes the conservatism of the ASME j
Appendix G curves and allows setting the LTOP setpoint such that the ASME Section XI, Appendix G limits are not exceeded by more than 10%. The code case permits use of an LTOP enable temperature equal to an RLr + 50 F or 200 F whichever is greater for the limiting material. This allows the implementation of a LTOP setpoint that preserves an j
acceptable margin of safety while maintaining operational margins for reactor coolant pump l
operation at low temperatures and pressures. The LTOP setpoint established in accordance i
with ASME Code Case N-514 will also minimize the unnecessary actuation of protection j
system pressure relieving devices. Therefore, establishing the LTOP setpoint in accordance with ASME Code Case N-514 criteria satisfies the underlying purpose of the ASME Code and the NRC regulations to ensure an acceptable level of safety.
I l
Attachment to ICANI19608 Page 10 of 11 l
10 CFR 50.12(a)(2)(iii): The reactor coolant system pressure / temperature operating window at low temperatures is defined by the LTOP setpoint. Implementation of a LTOP setpoint l
without the additional margin allowed by AShE Code Case N-514 would restrict the pressure / temperature operating window and would potentially result in undesired actuation of the LTOP system. This constitutes an unnecessary burden that can be alleviated by the application of AShE Code Case N-514. Implementation of an LTOP setpoint as allowed by AShE Code Case N-514 does not significantly reduce the margin of safety associated with normal operational heatup and cooldown limits. Further, the LTOP guidelines will reduce the potential for an undesired lift of the LTOP valve.
10 CFR 50.12(a)(2)(v): The exemption provides only temporary relief from the applicable regulation and ANO has made a good faith effort to comply with the regulation. We request that the exemption be granted until such time that the NRC generically approves AShE Code Case N-514 for use by the nuclear industry.
However, to retain suflicient pressure / temperature operating margin to the end of the current ANO-1 TS P/r limits, we require the exemption to use Code Case N-514.
Code Case N-514. Conclusion for Exemption Acceptability: Compliance with the specified requirements of 10 CFR 50.60 would result in hardship or unusual difliculty without a compensating increase in the level of quality and safety. AShE Code Case N-514 allows setting the LTOP actuation setpoint and enable temperature such that the AShE Section XI Appendix G limits are not exceeded by more than 10%. This proposed alternative is acceptable because the Code Case recognizes the conservatism of the ASME Appendix G curves and allows establishing a LTOP setpoint which retains an acceptable margin of safety while maintaining operational margins for reactor coolant pump operation at low temperatures and pressures. As discussed above, the Code Case provides an acceptable margin of safety against reactor vessel failure, and reduces the potential for an undesired LTOP actuation.
Therefore, application of Code Case N-514 for ANO-1 will ensure an acceptable level of safety.
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:
Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.
I The proposed change revises the pressure / temperature limits in accordance with the 10 CFR 50.60 requirements or in accordance with Code Case N-514. This approach utilizes the latest NRC guidelines relative to estimating neutron irradiation damage of the reactor vessel, as well as maintaining conservative limits with respect to the low temperature overpressure protection (LTOP) system. Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.
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Attachment to ICANI19608 Page 11 of11 Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident l
from any Previously Evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any previously evaluated since it does not introduce new systems, failure modes or plant l
perturbations. Therefore, this change does n_ol create the possibility of a new or different kind q
of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.
The proposed change will not involve a significant reduction in the margin of safety since the proposed pressure / temperature limitations have been developed consistent with the req /.ccments of 10 CFR 50.60. The operational limits have been developed to maintain the n cessary margins of safety through 32 effective full power years using methodologies prevively reviewed and approved by the NRC. The objective of these limits is to prevent non-ductile failure during any normal operating condition, including anticipated operational occurrences and system hydrostatic tests.
The LTOP safety factors are based on reanalyzed conditions for 32 effective full power years of operation utilizing methodology contained in ASME Code Case N-514. The LTOP evaluation under Code Case N-514 for low temperature transients is considered more I
appropriate than the ASME Section XI. The code case establishes a factor of 110% of the pressure determined to satisfy Appendix G, paragraph G-2215 of ASME Section XI, Division 1 as a design limit, instead of 100% required by Section XI. This proposed altemative is acceptable because the Code Case recognizes the conservatism of the ASME Appendix G curves and allows establi:hing a LTOP setpoint which retains an acceptable margin of safety while maintaining operational margins for reactor coolant pump operation at low temperatures and pressures. The Code Case provides an acceptable margin of safety against flaw initiation and reactor vessel failure, and reduces the potential for an undesired LTOP actuation. The application of Code Case N-514 for ANO-1 will ensure an acceptable level of safety. Therefore, this change does po_t involve a significant reduction in the margin ofsafety.
Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does not involve a significant hazards consideration.
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