ML20066A004
| ML20066A004 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/18/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20066A003 | List: |
| References | |
| NUDOCS 9101020378 | |
| Download: ML20066A004 (7) | |
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NUCLEAR REGULATORY COMMISSION 5
%*..../gI W ASHING TON, D. C. 20555 SAFETY EVALUATION BY-THE OFFICE OF NUCLEAR-REACTOR-REGULATION RELATED TO AMENDMENT NO_131-TO FACILITY OPERATING LICENSE NO. DPR-6 BALTIMORE GAS AND ELECTR_IC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT.-UNIT 2 DOCKET NO. 50_-318 1.0 I_NTRODUCTION By letter dated October 22, 1990, the Baltimore Gas and Electric Company-(the-licensee) proposed to amend the Technical Specifications of the Calvert Cliffs Nuclear Power Plant, Unit-2L In its submittal, the licensee provided Technical Specification changes to support 10 CFR Part 50,- Appendix G, heatup and cooldown Pressure / Temperature (P/T) limits applicable to the Unit 2-reactor vessel for a period up to' 12 effective full power years (EFPY).
The proposed P/T liinits were developed based on Regulatory Guide -(RG) 1.99, Revision 2.
The proposed revision provides u operation of the reactor coolant system (RCS)p-to-date P/T limits for the during heatup..cooldown, critit.ality, and inservice hydrostatic testing.
revised heatu In addition, the proposed ~ changes. included Valve (PORV) p and cooldown rates, a change in the: Power Operated Relief (LTOP),achanpressure setpoint for Low l Temperature Overpressure-Protection Coolant Pump-(ge in the LTOP enable temperature, a modification to R Pressure Safety Injection (HPSI) pump -controls when in LTOP conditions, and -
changes to the Bases-for the affected Limiting Conditions for 0peration (LCOs) to reflect the proposed changes.
To evaluate the P/T limits and supporting changes, the-staff used the following-
.NRC regulations and guidance:
Appendices-G and H to-10 CFR Part 50; the American Society of Testing Materials (ASTM), Standards. and the American Society of Mechnical Engineers (ASME) Code, which are= referenced in Appendices G and H; 10 CFR 50.36(c)(2); RG 1.99, Revision 2; Standard Review-Plan = (SRP) Sections 5.2.2 and 5.3.2; and Generic Letter:88-11.
- Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for-the operation.of the plant, l
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications, The P/T limits are-among the limiting conditions of operation -in.the Technical Specifications for 9101020370 9012E O PDR ADOCK 05000310 P
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, all comercial nuclear plants in the United States. Appendices G and H of 10 i
CFR Part 50 describe specific requirements for fracture toughness and reactor 1
vessel material surveillance that must be considered in setting P/T limits.
An acceptable method for constructing the P/T limits is described in SRP l
Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and tesing require-ments of reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capales ba tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards.
These tcst; define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature.
Appendix G also' requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and namittees use the methods in DG 1.99, Revision 2, to predict the effect of neutron irradiation on reactnr vessel materials. This guide defines the MT es the sum o' unirradiated reference temperature, the increase in reference temperature resultfng from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires that the licensee establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before initial ~ plant startup and that they-contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
LTOP is provided by the PORVs on the pressurizer.
These PORVs are set at a pressure low enough to prevent violation of the 10 CFR Part 50, Appendix G. P/T limits during heatup and cooldown should a RCS pressure transient occur during low temperature operations. The potential for overpressurization of the RCS can be minimized by a combination of administrative procedures and operator actions.
However, because operator action cannot always.be assumed, and because possible equipment malfunctions must be considered, additional controls must be in place to ensure adequate protection exists for all postulated events.
The two major concerns for LTCP protection are the mass addition-and energy addition transients. The proposed amendment provides restrictions on the use of HPSI pumps to provide protection for mass addition transients. Restrictions are also imposed on the starting and use of the RCPs to provide protection for energy addition transients.
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. The revised Regulatory Guide 1.99 results in more restrictive P/T limits To meet the revised requirements a new LTOP pressure setpoint and new heatup and cooldown rates are proposed.
These new values are such as to ensure that (1) given a limiting mass or energy input to the RCS during normal operation, anticipated operational occurrences, and hydrostatic testing; the Appendix G pressure-temperature limits are not challenged, and (2) operational flexibility is maintained.
2.0 EVALUATION - APPENDIX G-HEATUp AND C00LDOWN P/T LIMITS The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Calvert Cliffs 2 reactor vessel. The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Revision 2.
The staff has determined that the material with the highest ART at 12 EFPY was the intermediate shell longitudinal welds 2-203A, B, and C with 0.12% copper (Cu),1.01% nickel (Ni), and an initial RT o
-5 W.
ndt The licensee has removed one surveillance capsule from Calvert Cliffs 2.
The results from capsule 263 were published in Southwest Research institute Report SWRI-7524 The surveillance capsule contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline materials, intennediate shell longitudinal weld 2-203A, B, and C, the staff calculated the ART to be 170.8'F at 1/4T (T = reactor vessel beltline thickness) and 1 The staff used a neutron fluence of 1.007E19 n/cm[4.8*F for 3/4T at 12 EFPY 2
at 1/4T and 3.58E18 n/cm at 3/4T. The ART was determined by the Section 1 of RG 1.99, Revision 2 because only one surveillance capsule has been removed from the Calvert Cliffs 2 reactor vessel.
The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 171'F at 12 EFPY at 1/4T for the same limiting weld metal. The staff judges that the licensee's ART 171'F is more conservative than the staff's ART of 170.8'F, and it is acceptable.
Substituting the ART of 170.8'F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Apper. dix G of 10 CFR Part 50.
In additio, ne materials, Appendix G of 10 CFR Part 50 also imposes P/T limits
- he reference temperature for the reactor vessel closure flange ma'Section IV.2 of Appendix G states that when the pressure exceeds 2^'
preservice system hydrostatic test pressure, the temperature closure flange regions highly stressed by the bolt preload must exceed tne iaference temperature of the material in those regions by at least 120"F for normal operation and by 90 F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 30*F. the staff has determined that the proposed P/T limits s6tisfy Section IV.2 of Appendix G.
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4 l i Section IV.8 of Appendix G requires that the predicted Charpy USE at end of
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life be above 50 ft-lb, The intermediate shell plate 08906-1 (Heat No.
C-44E3-1)hasthelowest(limiting)unirradiatedUSEof76.7ft-lbamoggall beltline materials. Using an end of life peak fluence of 2.72E19 n/cm at 1/4T, the staff calculated an USE of 53.7 ft-lb for plate 08906-1 at end of life.
This is above the 50 f t-lb requirement, and is acceptable.
The staff has determined that the proposed P/T limits for the reactor coolant system for heatup, cooldown, inservice hydrostatic test, leak test, and criticality are valid through 12 EFPY because the limits confonn to the requirements of Apper. dices G and H-of 10 CFR Part 50. The licensee's submittal also satisfies Generic letter 88-11 because the licensee used the method in RG 1.99, Revision 2, to calculate the ART.
Hence, the proposed P/T limits may be incorporated into the Calvert Cliffs 2 Technical Specifications.
3.0 EVALUATION - LTOP CONTROLS The PORY lift setpoint is estimated at 430.0 psia to protect the most restrictive pressure of 471.2 psia which corresponds to a rate of 15'F/hr at 70*F in the RCS. The difference in the setting and the protect pressure is due to instrumentation uncertainty and PORV response time allowances. The LTOP enable temperature is 305'F and was estimated using the-Standard Review Plan 5.2.2, Revision 2, for heatup rates to 75'F/hr.
Based on the conservative assumptions and approved methods used, the PORV lift setpoint end the LTOP enable temperature are acceptable.
3,1 _HPSI pump Controls Overpressurization events due to mass addition, in their most limiting case include:
HPSI pump flow, charging pump flow and the coolant expansicn due to loss of decay heat removal.
The only controllable component in this case is the HPSI flow. Thus, the maximum PORY flow determines the HPSI flow af ter the charging pump and the expansion equivalent have been subtracted and the instrumentation uncertainty been accounted.
In this manner a total flow limit of 380 gpm yields an HPSI indicated flow of 210 gpm to ensure an Appendix G pressure limit of 471.2 psia in the pressurizer.
The HPSI flow rate was then compared to the requirements of other design basis events.
The most limiting l
such event is the loss of shutdown cooling which requires an actual flow rate of 175 gpm to prevent core uncovering.
The proposed flow of 210 gpm meets this limiting design requirement and is, therefore, acceptable.
3.2 RCP Controls RCP start is the primary concern for the limiting energy addition LTOP transient.
In this case we assume RCP start with: letdown isolation, energy addition from two RCPs, energy addition from the pressurizer heates, and loss of decay heat removal. Mitigation of such transients :s provided by the initial pressurizer pressure, pressurizer level, and the steam generator primary
~to-secondary change in temperature (delta-T).
For two RCPs starting assuming; an initial pressurizer level of 170 inches, a steam generator delta-T of 30*F,
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. initial pressurizer pressure of 3.0 psia, decay heat at a' level of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after shutdown, and no operator action, the pressurizer insurge peak pressure will be below the PORY setting, thus within the Appendix G acceptance criteria and is acceptable.
3.4 PORV Response Time The PORV response time is part of the estimated assumptions for the Appendix G limits.
For Unit 2 this response time has not been directly measured but assumed to be the same as for Unit 1.
The justification for this assumption is that the designs are identical.
For Unit 1 the maximum total response time is 0.49 seconds based on confinnatory analysis and testing. The licensee's i
confirmatory test results are consistent with results of similar tests performed I
by other utilities and-with bench tests perfonned by EPRI.
Based on the-results of the Unit 1 tests and the other industry tests, the response time assumed for Unit 2 PORVs is acceptable.
4.0 TECHNICAL SP_ECIFICATION CHANGES The licensee provided updated P/T curves in the proposed technical Specification
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Figures 3.4-2b and 3.4-2c for heatup and cooldown, respectively. Technical Specification 3.4.9.1.a provides-a new heatup rate of 75'F/hr for all temperatures.
Technical Specificetion 3.4.9.1.b provides new cooldown rates as follows:
100'f/hr for T3y, greater than N0*F 40*F/hr for T,y, between 180*F and 140*F I
15'F/hr for,T less than 140 F ave The Action Statements for Technical Specifications 3.4.9.1 and 3.4.9.2 are changed to reflect the proposed cooldown rates.
The Bases for Technical Specification 3/4.4.9 have been changed to reflect the rev'sions in the heatup and cooldown rates.
Technical Specification 3.4.9.3 proposes to lower the PORY lift setting to less than 430 psia, and require system vents equivalent to the PORVs for RCS temperatures less than 305'F (for system testing).
In-addition, two of the three HPSI pumps will be disabled and the HPSI loop motor operated valves be prevented from aligning pump flow to the RCS for RCS temperatures less than 305'F.
For one HPSI pump operable, the total flow will be throttled to 210 gpm. The above are not applicable if a system vent
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greater than 8 square inches exist.
The current action times are 7-da restore a PORV to " Operable" status or must be vented (depressurized) ys to within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
This is changed to 5 days to restore the PORV to operable and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to vent.
The tot; time is increased for ease of operation.t.ime is still the same (7 days) but the venting Surveillance requirements are added to verify system operability conditions. The Technical Specification 3.4.9.3 Bases -have also been changed to reflect the proposed set of conditions,
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' veral % hnical Specificationr have been changed tu reflect the new etquirements on the HPSI pumpa A footnote is added to Technical Specifications 3.1.2.1 and 3.1.2,3 which def.nes an " operable" HPSI pump.
A footnote is added to Technical Specification 3.5.3 which states that a maximut of one HPSI pump shall be operable when the RCS temperature is less or equal to 305'F. A footnote is added to the surveillance requirements of Technical Specification 4.5.2 allowing full fl0w testing of a HPS! pump. A footnote is also added to Table 3.3-3 providing infonntion on HPSI pump operation.
The Technical Spt.cification 3.5.3 Bases have also been changed to reflect the new requirements on the operation of the HPSI pumps.
The following changes reflect the new restrictions on the RCP operation.
A footnote to fechnical Specification 3.4.1.3 is changed to require that an RCP not be started if the RCS temperature is less (or equal) to 305'F unlessi (1) the pressuriter level is less or equal to 170 inches, (2) the primar secondary steam generator delta-T is less than or equal to 33'F, and (3)y to the pressurizer pressure is less than or equal to 320 psia. A footnote is added to Technical Specification 3.4.1.2 to provide RCP ctart control consistent with that of 3.4.1.3.
Finally, the bases of 3/4.4.1 and 3/4.4.9 are changed to reflect the new requirements.
5.0
SUMMARY
The proposed P/T limits for the RCS for heatup, cooldown, inservice hydrostatic testing, and criticality are valid through 12 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
(he P/T limits 1
also satisfy Generic Letter 88-11 because the licensee used the methods in RG 1.99, Revision 2 to calculate the ART.
methodology were,used for the 1. TOP analyses. Conservative assumptions and approved The analyses defined the HPSI pump and RCP control limitations.
We found these proposed revisions acceptable.
In addition, Technical Specification changes were defined which correctly i
reflect the new limitations and restrictions.
The staff has concluded, based on the above snd details provided in Sections 2, 3, and 4, that the proposed Technical Specifications and Bases supporting the new 12 EFPY P/T limits and LTOP controls are acceptable.
6.0 ENVIRONMENTAL CONSIDERA_ TION This amendment involves a change to a requirement with respect to the installation or use of the facilities' components located within the restricted areas as defined in 10 CFR Part 20 and to a surveillance requirement. The staff has determined that this amendment involves no significant increase in the amounts and no significant change in the types, 1
of any effluents that may be release,d offsite and that there is no significant j
increase in individual or cumulative occupational radiation exposure.
The Coninission has areviously issued a proposed finding that this amendment involves no significant lazards consideration and there has been no public comment on such finding. Accordingly, this amendment m?ets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
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e 7-assessment need be prepar(ed in connection with the issuance of this
7.0 CONCLUSION
We have concluded, based on the considerations discussed above. thatt (1) there is reasonable assurance-that the health and safety of the public will not Le endangered by operation in the proposed mannert and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimir.a1 to the common defense and l
security or to the health and safety of the public.
Dated: December 18, 1990
.PRINCIPAt CONTRIBUTORSr J. Tsao L. Lois S. Sanders D. Mcdonald n
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i Mr. G. C. Creel December 18,.1990 A copy of the related Safety Evaluation is enclosed, Notice of Issuance will be included in the Comission's next regular bi-weekly Federal Register notice.
Sincerely, i
glolNAL slGNED BYi.
Daniel G. Mcdonald, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II I
Office of Nuclear Retetor Regulation
Enclosures:
1
-1.
Amendment No.131 to DPR-69 2.
Safety Evaluation cc: w/ enclosures See next page
)1STR BijTION:
mcKet m e.,
'J. Calvo NRC/ Local =PDRs' ACRS(10)
PDI-1 Reading GPA/PA r
SVarga OC/LFMB EGreenman Plant File CVogan Wanda Jones DMcDonald SSanders OGC DHagen RACapra Edordan GHill(4)
JLinville JTsao LLois SSanders
- See previous concurrence PDI-1:LA PDI-1:PM OGC 1:D' CVogan DMcDonald:rse EHoller*
pRACapre 3
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/90 yttrgd DOCUMENT NAME:
CC AMEND 77840
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