ML20210S679
| ML20210S679 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/12/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20210S676 | List: |
| References | |
| NUDOCS 9908180168 | |
| Download: ML20210S679 (8) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF N'JCLEAR REACTOR REGULATION 1
RELATED TO AMENDMENT NOf183 To FACILITY OPERATING LICENSE NO. DPR-72 REVISED PRESSURE / TEMPERATURE LIMITS REPORT 6_N.D N
L LOW TEMPERATURE OVER-PRESSURE PROTECTION LIMITS
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- FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 l-DOCKET NO. 50-302 l
1.0 INTRODUCTION
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i By letter dated October 30,- 1998, as supplemented by letters dated December 31,1998, and L
. May 12,1999, Florida Power Corporation (the licensee), proposed changes to the improved Technical Specifications (ITS) for Crystal River Unit 3 (CR-3) to reflect new low temperature overpressure protection (LTOP) limits based on revised fluence values, and to include j
references to American Society of Mechanical Engineers (ASME) Code Case N-514, " Low L
Temperature Overpressure Protection" and Topical Report BAW-1543A, "Inte0 rated Reactor i
Vessel Surveillance Program." In addition, the licensee proposed to revise the CR-3 Pressure / Temperature Limits Report (PTLR) (1) to reflect the use of the fluence methodology l
in Topical Report BAW-2241P, " Fluence and Uncertainty Methodologies" for developing j
pressure / temperature (P/T) limit curves, and (2) to place the LTOP curve, developed using l
ASME Code Case N-514, in the PTLR. The current PTLR, which contains P/T limit curves for l
15 effective full power years (EFPY), was approved by the U.S. Nuclear Regulatory l
Commission (NRC or Commission) on December 20,1993. The proposed P/T limit curves are L
for 32 EFPY In addition to the revised fluence values, new information regarding material l
data and heatup and cooldown rates are also reflected in the proposed P/T limit curves. The l
proposed changes will affect Technical Specification (TS) Sections 5.6.2.19 and 3.4.11.
Likewise, the Bases for Sections 3.4.11 and 3.4.3 are to be revised accordingly. The j
December 31,1998 and May 12,1999, supplements did not affect the original proposed no sigrwficant hazards determination, or expand the scope of the request as noticed in the Federal Reaister.
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2.0 BACKGROUND
The LTOP system is designed to protect the pressure vessel boundary from overpressurization during low temperature operation. At CR-3, overpressure mitigation is accomplished using a combination of a pressurizer power operated relief valve (PORV) and a restricted water level in the pressurizer and/or a reactor coolant system (RCS) vent to depressurize the reactor. The i
system is manually enabled by the operator and uses a single setpoint as the lift pressure for the PORV. The design basis for the Crystal River LTOP system considers mass as well as i
heat addition transients.- The analysis of the mass addition transient accounts for the injection from one makeup pump to the RCS with the control valve failed to the fully open position. The analysis for the heat addition transient accounts for the heat input from the secondary sides of
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.. the once-through steam generators (OTSG)into the RCS upon starting a single reactor coolant pump.'. In the heat addition transient the analysis assumes the OTSGs are filled to 95 percent with 420 degree F feedwater and with RCS water at 150 degrees F. The results of the transient analyses indicate that the mass addition transient is limiting while the heat addition transient is self-limiting below the pressure-temperature limits.
2,1 Reaulatory Reauirements The staff evaluates pressure / temperature limits using the guidance of Generic Letters (GL) 88-11, "NRC Position on Radiation Embrittlement of Vessel Materials and its impact on Plant Operations", and GL 92-01, " Reactor Vessel Structural Integrity," and their revisions and supplements; Regulatory Guide (RG) 1.99, Rev. 2, " Radiation Embrittlement of Reactor Vessel Materials," Standard Review Plan (SRP) Sections 5.2.2 and 5.3.2, and Branch Technical Position RSB 5-2. SRP 5.2.2, " Overpressure Protection" provides review criteria for the evaluation of the adequacy of overpressure protection for the reactor coolant pressure boundary to meet the requirements of General Design Criterion (GDC) 31. Branch Technical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures,"is part of SRP 5.2.2. For the protection of the RCS boundary GDC 14 and 31 are applicable. GDC 14 requires that the RCS boundary be designed, fabricated, erected and tested so as to have an extremely low probability of j
abnormalleakage, rapidly propagating failure or gross rupture. GDC 31 requires that sufficient margin be provided to assure that the reactor coolant pressure boundary behaves in a non
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brittle manner under the stresses of normal operation, maintenance, test and accident I
conditions, with a low probability of rapidly propagating fracture.
Title 10; Code of Federal Regulations (10 CFR) Sections 50.60 and 50.61 require that licensees demonstrate that the effects of progressive embrittlement by neutron irradiation do not compromise the integrity of the reactor pressure vessel. To this end, two analyses are required: one to determine the P/T limits for normal heatup and cooldown operations and one
- to assess the vessel's ability to maintain its integrity during an emergency shutdown with cold water injection (i.e., pressurized thermal shock (PTS)).10 CFR 50.60 invokes Appendices G and H to 10 CFR Part 50, while 10 CFR 50.61 is the PTS rule which requires a PTS assessment.
Appendix G to 10 CFR Part 50 specifies fracture toughness requirements for.forritic materials of the reactor coolant boundary, it requires that the pressure-temperature limits for the reactor coolant system be at least as conservative as those obtained by the methodology in the 1989 edition of Appendix G to Section XI of the ASME Code. Attematives to Appendix G may be used when an exemption is granted by the NRC. An exemption approving the use of ASME Code Case N-514 was granted on July 3,1997. Appendix H to 10 CFR Part 50 requires a reactor vessel materials surveillance program to monitor changes in the fracture toughness' properties of ferritic materials in the reactor vessel beltline region. These changes result from exposure of these materials to neution irradiation and changes of the thermal environment., Material specimens exposed in the surveillance capsules are removed and tested at specified time intervals to monitor changes in the fracture toughness of the material.
. r 3.0 EVALUATION-Crystal River is currently operating at about 13 EFPY and is projected to reach the current license limit of 15 EFPY during the forthcoming cycle 12. The proposed revised operating limits are for 32 EFPY and are based on: (1) revised fluence values derived from the recently approved methodology described in BAW-2241, (2) more conservative chemistry data for the limiting beltline material, (3) more accurate instrument uncertainty calculations for the i4CS temperature, pressure and level instruments (4) the most limiting transient, i.e., mass addition and (5) an inadvertent discharge of the core flood tanks. Code Case N-514 was used in the determination of the current limits and is also used in the determination of the pioposed limits.
The methodology used in the current limits evaluation was approved in December 1997.
All components of the RCS are designed to withstand the effects of cyclic loads resulting from system pressure and temperature changes. These loads are introduced by heatup and cooldown operations, power transients, and reactor trips. Appendix G to 10 CFR Part 50 defines P/T limit curves for heatup, cooldown, LTOP, and inservice leak and hydrostatic testing. Each curve defines an acceptable region for normal operation. The curves are used for operational guidance during heatup and cooldown maneuvering, when P/T indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
3.1 The LTOP Curve As was pointed out previously in Section 2.0, for Crystal River, the limiting transient is mass addition with a failed open makeup valve. A modification of the analytical pressure-temperature limits allows a 10-minute time interval for operator action to mitigate the mass addition which is the limiting LTOP transient. The curve is a composite of the heatup, cooldown, hydrostatic pressurization and the 10-minute delay curves. Inadvertent discharge of the core flood tanks is taken into account. The curve has been added to the PTLR because it has the potential to be more limiting than either the heatup or the cooldown curves. The LTOP curve was developed using a methodology previously approved by the NRC and therefore including the curve in the PTLR is acceptable. The basis for TS 3.4.3 has also been revised to reflect the new methodology for the 32 EFPY curves.
3.2 LTOP Limits For the limiting transient above, (technical specification 3.4.11) the set-points reflect the projected material properties at 32 EFPY and the revised instrument uncertainties. The PORV i
lift pressure limit is set at less or equal to 454 psig (the old value was 457 psig). The lowest i
pressure (derived from Code Case N-514) in the LTOP curve is 458 psig which is adjusted for instrument uncertainty to 454 psig. The proposed LTOP enable temperature is 264 degrees F, including instrument uncertainty (the old value was 259 degrees F). The calculation for tne enable temperature was also estimated in accordance with Code Case N-514. The maximum pressurizer level setpoint is 155 inches (including instrument uncertainty) which allows for 10 minutes of potential operator mitigative action to limit the LTOP transient. Finally the temperature at which inadvertent discharge of the core flood tanks would overpressurize the RCS with respect to the LTOP curve has been raised from 197 degrees F to 208 degrees F.
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. L The LTOP enable temperature defines the limit below which the LTOP system is required to be l.
operable. The methodology used in this determination is the same as that used in the 1997 evaluation. The methodology complies with the ASME Code using an enable RCS liquid temperature corresponding to the reactor vessel thickness of 1/4T (where T is the vessel wall thickness) metal temperature of RT, + 50 degrees F or 200 degrees F, whichever is higher.
Instrument error of 5 degrees F is also applied. A very small temperature difference (about 1 degree F) was estimated between the RCS coolant temperature and 1/4T when heatup is suspended for 90 minutes. The Technical Specification Bases 83.4.11 reflect a heatup holding period of at leasi 90 minutes after the enable temperature has been exceeded and prior to exiting this LTOP limiting condition for operation. The licensee proposed an enable temperature of 268 degrees F consisting of RT, of 213 degrees F, the required 50 degrees F marcin and instrument error of 5 degrees F. This temperature is acceptable because it is conservative with respect to the minimum enable temperature ullowed by the ASME code.
The proposed PORV lift pressure limit is s 454 psig. This value is set to mitigate low temperature overpressure transients and to prevent violation of the Appendix G Section IV.2 (to 10 CFR Part 50) " Pressure-Temperature Limits and Minimum Temperature Requirements."
In the proposed application, Code Case N-514 is applied which allows peak pressures of 110 percent of Appendix G Section IV.2.b. As stated above, the NRC has approved the use of Code Case N-514 at CR-3. The analyticallimit is 458 psig with an allowance of 4 psig for instrument error. The licensee's analysis methodology is the same as the methodology applied in the estimation of the currently approved limits. The licensee performed an analysis of the mass addition transient (limiting transient) assuming water injection from one makeup pump through a failed fully open control valve with two reactor coolant pumps operating. The j
peak pressure would remain below 458 psig at 85 degrees F. The calculation was based on a j
pressurizer level of 155 inches, thus providing sufficient gas volume to allow at least 10 minutes for operator action. In these 10 minutes it is anticipated that the operator would take mitigating action to stop mass addition and the overpressurization. However, should the operator fail to take the appropriate action, the PORV flow capacity is sufficient to protect the vessel. The licensee estimated an analytical result of 458 psig and proposed 454 psig.
Because this value protects the Appendix G Section IV.2 required limits we find the proposed PORV lift limit value acceptable, j
Associated with the determination of the PORV lift limit is the maximum pressurizer water level.
The licensee assumed a 160-inch water level and demonstrated that the peak pressure will
, remain below the pressure temperature limit at 85 degrees F. Allowing for instrument error the proposed maximum allowable pressurizer level was set at 155 inches This value is acceptable because it supports the pressure-temperature limitt.
l A make-up tank mass addition overpressurization was analyzed. With RCS pressure equal or less than 464 psig, pressurizer level at or lower than 160 inches and the make-up tank level less than 88 inches, the make-up tank will deplete in less than 10 minutes without resulting in overpressurization. Therefore, the 88 inch make-up tank level remains the same for 32 EFPY as for 15 EFPY.
The licensee's calculations indicate that the PORV and RCS vents provide adequate flow l
(withtheir respective 1.049 inch and 0.75 inch diameter openings) to avoid RCS L
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, overpressurization due to an inadvertent make-up flow transient at an RCS pressure of 446 i
. psia. Therefore adequate flow will exist at 458 psig (473 psia) to ensure that the LTOP limit is not exceeded up to 32 EFPY.-
A spectrum of seven overpressurization scenarios was investigated. Review of these scenarios indicate that the licensee has covered the potential causes of overpressurization.
' Thus, we find that the mass addition is indeed the critical transient.
3.3 Review of the 32 EFPY P/T Limit Curves Licensees' evaluation The licensee determined that the limiting beltline material in the CR-3 vessel is the lower nozzle beltline to the upper shelf circumferential weld. Forty percent of this weld from the reactor pressure vessel inside the wall was fabricated from weld wire heat number 71249. The remaining weld was fabricated from weld wire heat number 8T1554. They will be referred to as Wold 71249 and Weld 8T1554. The licensee evaluated Weld 71249 using both Position 1.1 (surveillance data not available) and Position 2.1 (surveillance data available) of RG 1.99, Rev. 2 to calculate the adjusted reference temperature (ART) of the limiting weld at the l
1/4 thickness (1/4T) vessel location. In the first approach, the licensee used chemistry data of i
0.26 percent copper (Cu) and 0.61 percent nickel (Ni) to obtain a chemistry factor (CF) of j
181.6 degrees F for Weld 71249. The product of the CF and a fluence factor is the mean value of the adjustment in reference temperature, delta rte, For Weld 71249, the delta RT, at 32 EFPY is 138.6 degrees F based on the neutron fluence of 4.27 x 10" n/cm at the
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2 1/4T vessellocation. The ART is the sum of the initial RT,(10 degrees F), delta RT, (138.6 degrees F), and a margin term (56 degrees F). This resulted in a calculated ART for i
. this material of 204.6 degrees F.
The licensee also performed an evaluation using three surveillance data points of the same heat from Turkey Point Unit 3 and 4. CR-3 does not have Weld 71249 in its surveillance capsules. The CF, based on the surveillance data is 192.6 degrees F, Since one of the three surveillance data points is not credible, the licensee used a full margin of 56 degrees F, and calculated the ART to be 213 degrees F. This larger ART was used in the subsequent P/T limits calculaten for 32 EFPY.
To calculate the ART at 3/4T of the vessel wall for the lower nozzle beltline to the upper shelf circumferential weld, the chemistry data for Weld 8T1554 was used. Since surveillance data fx Wold 8T1554 is not available, Position 1.1 of RG 1.99, Rev. 2 was used to calculate ART
' for this material. Based on the chemistry data of 0.18 percent Cu and 0.63 percent Ni, an initial RTa f-5 degrees F, a margin of 68 degrees F, and a neutron fluence of 1.55 x 10" o
2 n/cm at the 3/4T vessel location, the licensee determined that the ART at 3/4T for 32 EFPY is
- 144.5 degrees F. This ART was used in the subsequent P/T limits calculation.
L NRC Staff's evaluation The staff compared the information supplied by the licensee in this submittal to the previously l
docketed information from the licensee's response to GL 92-01, Rev.1, Supplement 1, dated June 30,1998. This comparison indicated that the initial RTa value for Weld 71249 has
a been revised from 'a generic value of -5 degrees F to a material-specific value of 10 degrees F.
This change was based on data from Electric Power Research Institute Report NP-373 for welds manufactured from weld wire heat number 71249. The initial RT, value of +10 degrees F is acceptable because using the sum of this value and its corresponding margin
' term of 56 degrees F is more conservative than using the generic initiel RT, of -5 degrees F
'and its' margin term of 68.5 degrees F.-
The comparison also indicated that the licensee has used outdated chemistry data of 0.26 percent Cu and 0.61 percent Ni for Weld 71249 and 0.18 percent Cu and 0.63 percent Ni l
for Weld 8T1554 in its evaluation._ The chemistry data should be 0.23 percent Cu and 0.59 percent Ni for Wold 71249 and 0.16 percent Cu and 0.57 percent Ni for Weld 8T1554 as reported in BAW-2325, which was referenced in the submittal dated June 30,1998. Using outdated chemistry data for both welds did not impact the proposed P/T limits because (1) the licensee's chemistry data gives larger CF, which, in tum, gives more restrictive P/T limits, and (2) the more conservative CF based on surveillance data for Weld 71249 was eventually used in the proposed P/T limits calculation. The RVID includes Weld 8T1554 in the summary report l
for CR-3 because, although this weld is not reviewed in accordance with 10 CFR 50.61, the l
PTS rule, it is reviewed in accordance with Appendix G of 10 CFR Part 50 for P/T limits.
1 Substituting the ART of 213 degrees F at 1/4T and 144.5 degrees F at 3/4T into equations in SRP 5.3.2, the staff verified that the proposed P/T limit curves (at 32 EFPY) for heatup, cooldown, and inservice leak hydrostatic tests meet the beltline material,equirements in Appendix G of 10 CFR Part 50. This verification was made after the staff included the difference between the RCS coolant temperature and the wall metal temperature in these curves. Therefore, the staff finds the proposed P/T limit curves using the new fluence methodology and the inclusion of this fluence methodology in the PTLR acceptable.
3.4 References of Code Case N-514 und Topical Report BAW-1543A The licensee proposed to reference ASME Code Case N-514 and Topical Report BAW-1543A, L
in ITS 5.6.2 and in the ITS Bases. Since both the Code Case and the Topical Report have l
been approved for use at CR-3, the addition of these references is an editorial change and is l
acceptable.
4.0 STATE CONSULTATION
Based upon a letter dated March 8,1991, from Mary E. Clark of the State of Florida, Department of Health and Rehabilitative Services, to Deborah A. Miller, Licensing Assistant, U.S. NRC, the State of Fluida does not desire notification of issuance of license amendments.
5.0 ENVIRONMENTAL CONSIPERATIONS The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has
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determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The I
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^ ' mmission har previously issued a proposed finding that this amendment involves no Co L
significant hazards consideration and there has been no public comment on such finding (63 FR 71965). Accordingly, the amendment meett, the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). 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
S Based on its review of the licensee's proposal, the staff has determined that the proposed LTOP limits and P/T curves using the fluence methodology described in Topical Report BAW-2241P and the changes to the CR-3 ITS and PTLR to reflect their use, provide adequate assurance that the reactor coolant system will be adequately protected from the effects of pressure and temperature fluctuations. The staff concludes that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: S. Sheng, EMCB L. Lois, SRXB
- Date:
August 12, 1999-I i
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Mr. John Paul Cowan CRYSTAL RIVER UNIT NO. 3 Florida Power Corporation l
cc:
l Mr. R. Alexander Glenn Chairman Corporate Counsel (MAC-BT15A)
Board of County Commissioners Florida Power Corporation Citrus County P.O. Box 14042 110 North Apopka Avenue St. Petersburg, Florida 33733-4042 Inverness, Florida 34450-4245 Mr. Charles G. Pardee, Director Ms. Sherry L. Bernhoft, Director Nuclear Plant Operations (PA4A)
Nuclear Regu'atory Affairs (NA2H)
Florida Power Corporation Florida Power Corporation Crystal River Energy Complex Crystal River Energy Complex 15760 W. Power Line Street 15700 W. Power Line Street l
Crystal River, Florida 34428-6708 Crystal River, Florida 34428-6708 Mr. Michael A. Schoppman Senior Resident inspector Framatome Technologies Inc.
Crystal River Unit 3 1700 Rockville Pike, Suite 525 U.S. Nuclear Regulatory Commission Rockville, Maryland 20852 6745 N. Tallahassee Road Crystal River, Florida 34428 Mr. William A. Passetti, Chief Department of Health Mr. Gregory H. Halnon Bureau of Radiation Control Director, Quality Programs (SA2C) 2020 Capital Circle, SE, Bin #C21 Florida Power Corporation Tallahassee, Florida 32399-1741 Crystal River Energy Complex 15760 W. Power Line Street Attomey General Crystal River, Florida 34428-6708 Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 l
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