ML20006E809
| ML20006E809 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/15/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20006E807 | List: |
| References | |
| NUDOCS 9002260369 | |
| Download: ML20006E809 (4) | |
Text
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_ UNITED STATES I
I NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO. 70 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO.' 64 T0' FACILITY OPERATING LICENSE NPF DUKE POWER COMPANY, ET AL.
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- CATAWBA NUCLEAR STATION, UNITS 1 AND 2 1.
DOCKETS NOS. 50-413 AND' 50-414
1.0 INTRODUCTION
Sy ' letter dated February 5,21990, Cuhe Power Ccrepany, et al. (the licensee),
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recuested an emergency Technical Specification (TS) amenennt and a temporary waiver of compliance for Catawba Unit 2 only.
The amendment would revise the H
allowable Lift Setting tolerance f rom :1% to 21.5% in TS Table 3.7-2 for the.
Steam Line Safety Vahes of Catawba Unit 2 until the first forced outage, reactor trip, or refueling cutage.
The requested revision resulted from the development of a' new equipment constant by the valve vendor, Dresser Industries,
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b Lift settings for-Catawba Unit I safety valves were established using the new L
equipment constant during the refueling octage that began on January 26,-
1990.
Thus, Unit 1 is in compliance with the current TS requirements and is affected only administratively because it shares a comon TS document with' i:
Unit 2.
l-2.0 E VALUATION Il On Jaisuary 26, 1990, during'.a. Dresser Industries review of the procedures used to adjust the Catawba steam line safety valves, the vendor representative '
mentioned that Dresser Industries had revised the model 1566 hydroset correction factor from 0.339 to.0.352 in August 1989..This model is used at' Catawba to calibrate the. main steam safety valves.
The licensee's recalculation:of the set pressures for all safety valves, using the new correction factor, revealed
. that foer 'out of 20) Unit 2 valves will have set pressures greater than a llowed by Table 3.7-2 of TS 3/4.7.1.
a The licensee stated _that the most limiting scenario for the main steam safety-valves is a turbine trip from full power.
The. reactor will trip due to loss i
of' secondary heat sink, and the time to trip is relatively insensitive' to the safety valve set pressure differences.
The licensee also stated that a design study, MGOS-0176, regarding setpoint '
drift for McGuire safety valves is also applicable to Catawba after allowances are trade for design differences.
The major differences are:
(1) a higher
'T-ave for Cotawba,. and (2) two of Catawba's valves have a setpoint 5 psi greater than corresponding valves at McGuire. While these dif ferences resu lt 9002260369 900215 PDR ADOCK 05000413 7
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1 in a higher steam generater pressure for Catawba, it is still within the 110%
oesign basis.
It should be noted that the above design study utilized a 3%
setpoint drift, whereas the proposed TS change requests 1.51 allowable tolorance for Catawba Unit 2.
4 The most conservative accident for radiological releases is the steam generator-tube rupture.
The licensee stated in the February 5,1989, submittal that the change in safety valve setpoints will not increase the calculated duration of a
atmospheric releases through the valves.
Thus, the dose consequence analysis presented in the licensee's letter dated Cecenber 8,1969, regarding steam generator tube rupture analysis, remains bounding.
i The change of the safety valves' setpoint produces a mininal impact on prirary side temperature 'and pressure and on the capability of the vc htes to perform their design function.
Therefore, the consequences of design basis accidents t
adoressed in the: Final. Safety Analysis Report and the prs 3 ability of valve failures are not adversely affected, t
L Base (on its review, the NRC staff c.grees with the licensee's assessment and h
finds that the revision for Catawba Unit 2 to the allowable value of lift i
setting of main steam safety valves from 1% to 21.5% is acceptable until the l
first forced outage, reactor trip, or refu'eling outage, u
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3.0 EMERGENCY CIRCUMSTANCES o
a The licensee's' application for the TS change has been tinely.
On January 26, 1990, a Dresser Industries review of the procedures used to adjust the Catawba' steam line safety valves revealed that a new equipment constant shoulci be L
used.. On the same day, Catawba Unit.1 entered a refueling outage which enabicd the licensee's personnel to set the pressure for its safety valves -
using the new const6nt.
Therefore, the unit was'in compliance with the TSs.
However, on the same day, Catawba Unit 2 was operating at full pcwer when the licensee's maintenance personnel were informed of.this change by the vendor field service representative. As a result, the responsible maintenance engineer requested an operability evaluation from the licensee's Design
's Engineering personnel.
On February 2,1990,- the evaluation revealed that four valves (out of 20) must be treated as inoperable because they do not meet the
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' TS allowable tolerance. Therefore, TS 3/4.7.1 action statement was entered and power reduction commenced.
Following the operability determination, the 4
licensee promptly contacted the KRC to request enforcement discretion which would allow Unit 2 toTperate at full power until the next forced outage, reactor trip, or refueling outage. The request was supported by the F
-licensee's Design Engineering analysis which demonstrated that the small i
increase in the lift setting of the safety. valves has no nuclear safety signifi cance.
However TS change would be more, after discussion with the NRC, it was determined that a e
appropriate for this situation.
The licensee innedi-ately requested the change by telephone and submitted the TS amendnent on February 5,1990.
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The NRC staff agrees with the -licensee's evaluation that the snell inertase in lift-setting has an insignificant impact on safety.
Furthernore, the NRC staff-finds that f ailure to grant the proposed changes in a tinely manner would result in derating Catawba Unit 2.
We also find that the licensee could not reasor.hbly have avoided this situation, that the licensee has responded in a tinely unner, and has not delayed its application to take novantage of the Emergency License Anendment provisions of 10 CFR 50.91. Accordingly, the NRC staff concludes that the licensee has satisfied the requirenents of 10 CFR 1
50.91(a)(5), and that a ya h d emergency exists.
e 4.0 = FINAL.NO SIGNIFICAhi HAZ ARDS CONSIDERATION DETERMINATION The anEndnent request would revise, on an emergency basis, the existing TS Table 3.7-2 lift setting of the steam line safety valves for Catawba Unit 2.
The revision resulted from the developnent of a new equipment constant _by the valve vendor.
The change in the lift setting is from 11% to 1.5% and would -
remain in effect until the first forced outage, reactor trip, or refueling u
E cu tage.
L 1.ift settings for Catawba-Unit 1 safety valves were established using the new equipnent constant during the refueling outage that began January 26, 1990.
l Thus, Unit 1.is already.in compliana with the current TS requirements and is included aaministratively because it shares a common TS docunent with Unit 2.
-The Consission's regulations in 10 CFR 50.92 state that the Connission may J
n.ake a final determination that a license auendnent involves no significant L
hazards consideration, if operation of the facility, in accordance with the H
anenmnt would not:
i (1)
Involve a significant increase in the probability or consequences of p
arty accident previously evaluated; or-l' (2) Create the possibility of a new or different kind of accident from
. any accident previously evaluated; or (3)
Involve a significant reduction in a nargin of safety.
l' The proposed anendnents would not involve a significant increase in the L
L probability or consequences of:an accident previously evaluated.
The secondary side pressure will be maintained within its design basis limits and the impact on the primary side will be minimal.
The requested change in the valve lift setting tolerance is less than 0.5 of 15.
This small change does not prevent the valves from performing their. design function and will not 5
significantly increase the probability of failure.
The most conservative accident for radiological releases is the steam generator tube rupture.
The change in. safety valve setpoints will not increase the calculated duration of -
atmospheric releases through the valves.
Thus, the dose consequence analysis presented in the licensee's letter dated December 8,1989, regarding steam-u generator tube rupture analysis, remains bounding.
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The proposed amendnents would not create the possibility of a 'new or different 1
~ kind of accident f rom any accident previously evaluated because the design and modes of. operation of the main steam safety valves and Catawba station will not be affected.
As such, no new or different kind of accident would be possible.
The proposed amenaments would not involve a significant reduction in a nargin of safety.
The requested change f rota :1% to 1.5% for the-lift setting tolerance of safety valves is small. A design.stu$ using a 3% setpoint drift concluded that the steam generator pressure reniains within its design limits.
. The. impact on' the~ primary side is rainimal.
As such, no significant reduction in aLnargin of safety would result from the above change.
Accordingly, the Coninission finds that the requested anendnents involve no-p significhnt hazaros consideration.
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5.0 STATE CONSULTATION
-In:occordance with the Coninission's-regulations, the State of South Carolina was contacted on February 6,1989.
The state representative had.no comnents.-
6.0 ENVIRONMENTAL CONSIDERATION
i These amendments involve changes in requirements with respect to the installation or use of facility components locatert within the restricted area as defined in.10 CFR Part 20. The staff has determined that the amendnents t
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 L
~ ignificant increase in individual or cumulative occupational radiation s
exposure.
The NRC staff has made a final determination that the amenonents-L involve no significant hazards consiceration. Accordingly, the anendments
' meet the eligibility criteria for categorical, exclusion set forth in 10 CFR 51.22(c) (9).. Pursuant to 10 CFR 51.22(b), no environnental-impact statement t
p or environmental assessment need be1 prepared in connection with the : issuance of these anendnents.
7.0 CONCLUSION
We have. concluded, based on the consicerations. discussed above, that':
(1) there is, reasonable assurance that the health and safety of the public will L
not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of these amendnents will not' be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
K. Jabbour, PDII-3/DRP-I/II Dated: February 15, 1990 i
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