ML20034C614
| ML20034C614 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/26/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034C612 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9005040216 | |
| Download: ML20034C614 (6) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
RELATED TO AMENDMENT NO.131 TO PROVISIONAL OPERATING LICENSE NO. DPR-20 CONSUMERS POWER COMPANY PALISADES PLANT i
l DOCKET NO. 50-255
1.0 INTRODUCTION
By letter dated September 12, 1989 (Ref. 1), and supplemented by letters dated l
September 22, 1989 (Ref. 2), September 25, 1989 (Ref. 3), and March 2, 1990 (Ref. 4), Consumers Power Company (CPCc) proposed certain revisions to the
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pressure / temperature (P/T) limits specified in the Palisades Plant i m aical Specifications (TS), Sections 3.1, 3.2, and 3.3.
Theproposedrevisionsglso would change the period of effectiveness of the P/T limits to 1.8E19 n/cm neutron fluence or to about 10 effective full power years (EFPY) (Ref. 2).
l The proposed P/T limits were developed based on Section 1 of Regulatory Guide (RG) 1.99, Revision 2.
The proposed revision provides up-to-date P/T limits for the operation of the primary coolant system (PCS) during heatup, cooldown, criticality, and hydrotest.
On July 12, 1988, the Commission issued Generic Letter 88-11 which advised the addressees that Regulatory Guide 1.99 Revision 2, became effective as of May 3
1988, and that the NRC would be using that revision of the guide in reviewing a
submittals related to pressure / temperature limits and.for analyses that require an estimate (except for pressurized thermal shock) of embrittlement of reactor vessel beltline materials.
Pressurized water reactor licensees also 4
were reminded that low-temperature overpressure protection (LTOP) setpoints also might need revision because of the revised regulatory guide.
Consumers Power Company provided (Ref. I through 4) its updated P/T curves based on the guidance provided in Revision 2 of Regulatory Guio) 1.99 and they are applicable until August 1993.
CPCo also proposed system modMications to install larger PORVs with variable setpoints for LTOP function.
In light of the above. ' nges, CPCo also proposed changes to several other Technical Specifications to assure proper LTOP at the Palisades Plant.
- 2. 0 EVALUATION 2.1 Pressure / Temperature Limit Considerations The staff uses Appendices G and H of 10 CFR Part 50, the ASTM Standards and the ASME Code which are referenced in Appendices G and H, RG 1.99 (Rev. 2),
and Standard Review Plan (SRP) Section 5.3.2 to evaluate the P/T limits.
Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits, and an acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
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j Appendix G of 10 CFR Part 50 specifies fracture toughness and testing i
requirements for reactor vessel materials in accordance with the ASME Code i'
and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn refers to ASTM Standards. These tests define the extent of vessel l
embrIttlementatthetimeofcapsulewithdrawalintermsoftheincreasein reference temperature. Appendix G also requires the prediction of the effects of neutron irradiation on vessel embrittlement by calculating the adjusted referencetem>erature(ART)andCharpyuppershelfenergy(USE).
RG 1.99 defines the Ali as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Palisades reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
Thystaff has determined that the material with the highest ART at 1.8E19 n/cm (about 10 EFPY) was the circumferential weld with 0.20% co er(Cu),0.97% nickel (Ni),andaninitialRT of -56 4 (Ref. 5, 6, 7, 3.
ndt CPCo has removed two surveillance capsules from Palisades. The results from t
i capsule K-290 were published in Westinghouse report WCAP-10637 (Ref. 9). The l
results from capsule A-240 were published in Battelle-Columbus Resort BCL-585-12 (Ref. 10).
Both the surveillance capsules contained Ciarpy impact l
specimens and tensile specimens made from base metal, weld metal, and HAZ metal (Ref. 5,11).
l for the limiting beltline material, the circumferential welds the staff calculatedtheARTtobe233' Tat 1/4T(T=reactorvesselbeltlineghickness) and 177'T at 3/4 and 0.39E19 n/cm). The staff used a neutron fluence of 1.08E19 n/cm at 1/4T at3/4T(Ref.12).
CPCo used the method in RG 1.99 Rev. 2, to calculate an ART of 241'F at 1/4T and177' Fat 3/4Tforthesamelimitingcircumferentialwelds. The staff judges I
that CPCo's ART of 241'F is more conservative than the staff's ART of 233'T and is acceptable. Substituting the ART of 241*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and I
hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
l In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20% of the preservice system lydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference 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 60'F, the staff has determined that the proposed P/T limits satisfy Section IV.A.2 of Appendix G.
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i Section !Y B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb.
CPCo reported th6t the unirradiated USE for plate i
D 3804-2/C-1308-3 (Ref. 9, 10) in the longitudinal direction was 120 ft-lb.
Based on a reduction factor of 0.65, the USE in the transverse direction was 78 ft-1b.
Using Figure j of RG 1.99, Rev. 2, it was found that the expected USEatEOL(3.35E19n/cm)was43ft-lb.
CPCo has joined the Combustion l
EngineeringOwners'GrouptodeterminetheeffectsoflowUSEvalues(less than50ft-lb)inbeltlinematerials.Section V.E of Appendix G permits the licensee to continue to operate if it submits a proposed program for satisfying the requirements as set forth in V.C at least 3 years prior to the date when the USE value is predicted to fall below 50 ft. Ib. The staff will review the report from the owners' group and monitor CPCo's low USE.
2.2 Low Temperature Overpressure Protection In the current design, LTOP is provided by two pilot-operated relief valves (PORVs)withaliftsettingof310psiawhenthePCScold-legtemperatureis less than 300'F and a lift setting of 575 psia when the PCS cold-leg temperature is less than 430'F. The FORVs must be aligned to the PCS when the PCS is below the specified temperature to provide assurance that the reactor i
vessel will be operated in the ductile region in accordance with 10 CFR Part 50, i
Appendix G, during both normal operation and overpressurization events due to equipment malfunction or operator error. Technical Specifications require alignment of *he PORVs with proper lift settings to the PCS at PCS temperature below 430'F to prevent exceeding Appendix G P/T limits. When the PCS i
temperature is ebove 430'F the pressurizer safety valves will provide overpressurizationprotectIonforthePCS.
l CPCo proposes to replace the existing PORVs and block valves with larger valves. The LTOP systen, will be progranned and calibrated to ensure opening l
cf the PORVs when the combination of reactor coolent pressure and temperature l
is above or to the left of the limits displayed in Figure 3-4 of the proposed Technical Specifications. This limit curve is developed from the more limiting of %e heatup and cooldown limits for the specific PCS temperature I
while heatii t or cooling at the maximum permissible rate for that temperature.
The limit in Figure 3-4 includes an allowance for pressure overshoot-during the interval beginning when a PORY opens to terminate the pressure rise and when the PCS pressure reaches the limit. Either one separately of the two PORVs is sufficient to provide LTOP. The actual PORY lift setting will be lowered from the limit of Figure 3-4 to allow for potentini instrument inaccuracies, measurement error, and instrument drift.
For potential mass addition transients leading to PCS overpressurization, CPCo's analysis was performed with several conservative assumptions which include:
(1) inadvertent start of charging flow coincident with isolation of letdown, maximum pressurizer heating rate, and maximum PCS heating rate; (2) inadvertent start of HPSI coincident with inadvertent start of charging, maximumpressurizerheatingrateandmaximumPCSheating;(3)thePCSis water-solid at beginning of the transient; and (4) flow delivery from two HPSI pumps to the PCS is twice that from one HPSI pump.
The results of the mass addition transients indica 4
- the case of both HPSI 3
pumps starting in conjunction with a charging /letdow;. imbalance, that 10 CFR a
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Part 50, Appendix C limits and the shutdown cooling system maximum allowable pressure will be exceeded when the PCS is less than 260*F. Thus, the proposed Technical Specification 3.3.2g requires that if the reactor head is installed, both HPSI pumps shall be rendered inoperable when the PCs temperature is less than 260F or the shutdown cooling isolation valves are open.
For heat addition transients which lead to PCS pressure overshoot, the current TechnicalSpecificationsprohibitstartingareactorcoolantpump(RCP)ifthe steam generator (SG) secondary water temperature is higher than the PCS temperature and the LTOP system is required to be in service. This is to preclude overpressurizing of the PCS due to reverse heat transfer in the SG when the RCP is started. However, the proposed LTOP modifications provide more flexible plant operation limits. The results of CPCo's analysis show that the operator could initiate a RCP when PCS is less than 430*F with shutdown coolino system isolated and the delta-T between the SG and the PCS not exceeding 20*F when the PCS temperature is between 170*F and 210'F and not exceeding 100'F when the PCS temperature is in the rest of temperature regions. These restrictions are specified in the proposed Technical Specification 3.1.1h.
CPCo's analysis of heat addition transients was performed with conservative assumptions which include:
(1) the PCS is assumed to be water-solid at the initiation of the transient; (2) pressurizer heaters arc assurled to be on at the initiation of the transient; and (3) the maximum-allowable delta-T between the SG and PCS temperatures exist at the initiation
- of the transient.
CFCo proposed changes of Technical Specifications 3.1.1h, 3.1.8, 3.3.29, 4.1.1c, 4.6.1b, Figure 3-4 and their associated bases sections reflect the above discussed LTOP system modifications and the heatup and cooldown rates identified by the updated Figures 3-1, 3-2, and 3-3.
The staff finds thet they are reascnably conservative ar.d acceptable.
300 CONCLUSION The staff concludes that the prcposed P/T limits for the reactor coolant system for peatup, cooldown, leck test, and criticality are valid through 1.8E19 n/cm neutron fluence or about 10 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
CPCo's submittal also satisfies Generic Letter 88-11 because CPCo used the method in RG 1.99, Rev. 2 to calculate the ART.
Hence, the proposed P/T limits may be incorporated into the Palisades Technical Specifications.
The staff will review the owners' group' report on the effect of having low USE values before making a decision on CPCo s low predicted USE.
4.0 Elv,lRONMEllTAL CONSIDERATION This amendment involves changes to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or a change to a surveillance requirement. The staff has determined that the amendment involves no significant increase in l
l the amounts, and no significant change in the types, of any effluents that may
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be relcastd offsite and that there is no significant increase in individual cr cunulativt occupational radiatien exposure. The Cervissicn has previcurly I
issued a proposed finding thet this an.endroent involves no-significant hazards censideration and there has been no public coment on such finding.
i Accordingly, this amendment meets the eligibility criteria for categorical l
excluriersetforthin10CFR51.22(c)(9).
Pursuantto10CFR51.PP(b),no environt..tntel intact stattsent nor environmental assessnient need be prepared in contcction with the issuance of this amendtrent.
5,0 C0llCLUSION The staff has concluded, based en the considerations discussed above, that:
(1) there is reascr.able assurance that the health and safety of the will ret be endangered by operation in the proposed manner, and (2) public such activitics will be conducted in compliance uiti the Comission's regulations, and (3) the issuance of this antndment will ret be inimical to the comon defense and security or to the health and safety of the public.
Principal Cor.tributor: Chu Liang, NP.R/SRXD John Tsao, ilPR/ECl1B Dated: Aptil 26, 1990
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REFERENCES
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1.
Letter from K. Berry of Consumers Power Company to USNRC, " Technical Specification Change Request - LTOP/ Variable Setpoint," dated September 12, 1989.
J 2.
Letter from K. Berry of Consumers Power Company to USNRC, "Technica.
Specification Change Request - LTOP/ Variable Setpoint - Revision 1,"'
y,p dated September 22, 1989.
A 3.
Letter from R. Smedley of Consumers Power Company to USNRC, " Technical-Specification Change Request - LTOP/ Variable Setpoint - Revision 1 -
Additional Information," dated September 25, 1989.
4.
Letter from R. Smedley of-Consumers Power Company to USNRC, " Technical Specification Change Request - LTOP/ Variable Setpoint - Revision 2 3
Additional Information," dated March 2,1990.
5.
May 23, 1978, Letter from D. P. Hoffman (CP) to D. L. Ziemann (USNRC),
yj subject:
Palisades Plant e
6.
October 12, 1984, Letter from T. U. Marston (EPRI) to J. Tosky (CP),
S subject:
Reactor Vessel and Surveillance Welds for Palisades Plant 7.
Attachment III:
Summary of Findings Relative to Palisades Plant Seactoi, e q' Vessel Materials, June 14, 1985
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8.
October 9, 1984, Letter from N. J. Ocrt m (C-E) to J. B. Tosky (CP),
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subject:
Palisades Vessel Weld Documentation
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9.
October 31, 1984, Letter from B. D. Johnson d'P)' to H. R. Denton (USNRT.),
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subject:
Palisades Plant--Test Report of Reactor Vessel Specimen Capsules T-330 and W-290 Removed during the 1983 Refueling Cycle (WCAP-10637) s 10.
BCL-585-12, Final Report on Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program:
Capsule A-240, March 13, 1979, s
Battelle-Columbus Laboratories
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11.
February 5,1985, Letter from B. D. Johnson (CP) to Director, USNRC
>!f Nuclear Reactor Regulation, subject:
Palisades Plant--Reactor Pressure
~ t Vessel Surveillance Program Additional Information i
12.
November 30, 1988, Letter from R. W. Smedley (CP) to USNRC Docue nt Control Desk, subject:
Palisades Plant--Compliance with Pressurized Thermal Shock Rule 10CFR50.61 and Regulatory Guide 1.99, Revision 4'
2--Fluence Reduction Status
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