ML20137V516
| ML20137V516 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/15/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137V473 | List: |
| References | |
| NUDOCS 9704170325 | |
| Download: ML20137V516 (9) | |
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7 Cro UNITED STATES i
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NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 20066-4001
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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
RELATED TO AMENDMENT NO. 87 TO FACILITY OPERATING LICENSE NO. NPF-37, j
AMENDMENT NO. 87 TD FACILITY OPERATlgn_1] CENSE N0 _MEE-51, i
jliEHQHEHL.HQ. 79 TO FACILIrY OPERATIMG LI(.gNSE NQ4 HEE-22,
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j AHQ.8H[tglgML NO. 79_.JO FACILITY QEEB&HHQlICENSE NO. NEE-ll j
i IOMONWEALTH EDISON COMP 8H1 i
B,YJJQH SIAUQth,_ UNIT N01. I AND 2 i
BBAIDWOOO_, STATION. UNIT HQ). 1 AND 2
-DOCKET N05. STN 50-454. STN 50-455. STH,,50-456 AND STN 50-457 i
1.0 1HI80 DUCTION By letter dated April 29, 1996, as supplemented on January 21 and March 25, 1997, Commonwealth' Edison. Company (Comed, the licensee) requested an amendment to-Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, which j
would change the Technical Specification (TS) Section 3/4.7.1, " Turbine Cycle i
Safety Valves" and the associated Bases. Additional information was provided in a letter from Comed to the NRC dated March 25, 1997, and in a Westinghouse letter to the NRC dated January 28, 1994. The March 25, 1997, submittal provided additional information that did not change the initial proposed no i
significant hazards consideration determination.
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The proposed TS would:
j 1.
Revise TS 3.7.1.1, Action a., to require the unit to be in hot shutdown, i
.rather than cold shutdown, for consistency with Revision 1 of MUREG-1431, " Standard Technical Specifications for Westinghouse Plants,"
and adding a new Action b to clarify the shutdown requirements when there are more than three inoperable main steam line American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) l safety valves on any one steam generator.
2.-
- Revise TS Surveillance Requirement (SR) 4.7.1.1 to clarify that i
_ Specification 4.0.4 does not apply for entry into Mode 3 for Byron and i
Braidwood and for Braidwood only, deleting the one-time requirements for
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Unit. 1, Cycle 5 and Unit 2 after outage A2F27 (during Cycle 5).
3.
Revise the maximum allowable power range neutron flux high trip setpoints in Table 3.7-1.
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4.
Revise Table 3.7-2 to increase the as-found main steam safety valve (MSSV) lift setpoint to 13 percent, provide as-left setpoint tolerance of il percent, and change a table notation.
5.
Delete the orifice size column from Table 3.7-2.
6.
Revise the Bases for TS 3.7.1.1 to be consistent with the proposed changes to TS 3.7.1.1.
2.0 EVALUATION The evaluations of each of the six proposed changes listed above in Section 1.0 are given below.
2.1 Ig hnical See _i.fic.Ation 3.7.1.1 a.
Description Comed proposes to revise TS 3.7.1.1, Action a., to delete the phrase "four reactor coolant loops and associated steam generators in operation and with."
l The final mode requirements are changed to require the unit to be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of reaching hot standby, rather than cold shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of reaching hot standby.
The action statement adds words to clarify that the action ~ applies when there are up to three inoperable main steam line Code safety valves on any one steam generator.
Comed proposes to add a new Action b. to clarify the shutdown requirements when there are more than three inoperable main steam line Code safety valves on any one steam generator; i.e., there is no provision to restore the.
inoperable valves to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock to het standby begins.
The current Action b. becomes Action c.
Comed also proposes to revise the title to Table 3.7-1 to delete "during four loop operation." The Table of Contents is also revised to reflect the title change.
b.
Imoact of the Chanoes The proposed change by Comed to require the final mode to be hot shutdown rather than cold shutdown is consistent with the applicability section of the specification, which does not require the main steam safety valves (MSSVs) to be operable in hot shutdown.
There are no credible transients requiring the MSSVs in Modes 4 and 5.
The steam generators are not normally used for heat removal in Modes 5 and 6 and, thus, can not be overpressurized.
NUREG-1431 does not include requirements for the MSSVs to be operable in these modes.
The change will also eliminate the unnecessary transient that had been imposed on the unit by forcing entry into cold shutdown. Therefore, these proposed changes have no significant negative impact on any system or operating mode, and provide a benefit of eliminating unnecessary plant transients.
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The new Action b.-provides clarification to the current Action a. and is consistent with the requirements in NUREG-1431.
The shutdown requirements are clarified based on the number of inoperable valves. There are no technical changes to these times.
l The title change to TS 3.7-1 is an editorial change.
i The staff has found the above changes to be acceptable as they have no significant negative impact and they are consistent with NUREG-1431.
i 2.2 Surveillance Reauirement 4.7.1.1
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i a.
Descriotion 1
l Comed proposes to revise TS SR 4.7.1.1 to clarify that TS 4.0.4 does not apply for entry into Mode 3 for Byron and Braidwood; i.e., entry into Mode 3 would be allowed prior to completing surveillance testing on the MSSVs. The l
proposed revision also deletes the one-time requirements for Braidwood, Unit 1, Cycle 5 and Braidwood, Unit 2, after outage A2F27 (during Cycle 5) l that were added by Amendment Nos. 49 and 51.
The current practice for testing the MSSVs is to enter Action a. for TS 3.7.1.1 while testing the valves, since the ASME Code,1983 Edition through summer 1983 Addenda requires the plant to be at normal operating temperature i
and pressure (Mode 3) and TS SR 4.0.4 would require the testing to be completed prior to entering Mode 3, which places severe time restriction on
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the valve testing.
b.
Imoact of the Chanaes 4
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Changing TS SR 4.7.1.1 to delete the one-time requirements impos @ by previous Braidwood amendments and allow entry into Mode 3 for MSSV testit ; for Byron j
and Braidwood will permit testing of the MSSVs at normal operating pressures and temperatures in accordance with the applicable codes while allowing a j
reasonable amount of time for completion of the surveillance.
i The staff finds these changes to be acceptable as they have no significant i
negative impact and they are consistent with NUREG-1431.
l The one-time requirements for Braidwood -are no longer applicable and, j
therefore, deletion of them is acceptable.
2.3 Chances to Maximum Allowable Power Ranae Neutron Flux Hiah Setooints l
a.
Descrintion
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Comed proposes to revise the power range neutron flux high setpoints in the event of MSSVs inoperability based on a revision to the method for calculating i
the setpoints that was provided in a Westinghouse Nuclear Safety Advisory Letter, (NSAL-94-001), which was sent to the NRC on January 28, 1994. The 6
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existing equation for the Reactor Trip Setpoint reductions for four-loop operation in TS Bases 3/4 7-1 has been replaced by:
i Highe 100 ( Qif)
O K
l Where:
Highe -
Safety Analysis power range high neutron flux setpoint, in i
percent.
Q Nominal NSSS power rating of the plant (including reactor i
coolant pump heat), in Mwt (-3427.6 MWt).
l K
I Conversion factor - 947.82 (BTU /sec.)/MWt.
l w,
minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as i
appropriate, in lbm/sec.
l h,,
Heat of vaporization for steam at the highest MSSV opening i
pressure including tolerance and accumulation, as
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appropriate, in BTU /lbm.
l N
Number of loops in the plant (=4).
The TS Bases clarify that the values from the above algorithm must be adjusted lower for use in TS 7.1.1.1 to account for instrument and channel i
i uncertainties (typically 9 percent power) as specified in NSAL-94-001. This reduction has been included in the proposed values in the TS Table 3.7-1.
4 The proposed changes in TS Table 3.7-1 to the maximum allowable power range j
neutron flux high setpoint are as follows:
{
Maximum Number of Maximum allowable power range neutron Inoperable Safety flux high setpoint Valves on any (percent of rated thermal power)
Steam Generator Current TS Value Proposed Value i
1 87 60 j.
2 65' 43
{
3 43 -
25 -
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b.
Imoact of the Chanaes Westinghouse has determined (NSAL-94-001) that the method used to calculate t
the current setpoints may not be valid under certain conditions.
That method assumes a linear relationship between the maximum allowable initial power level and available MSSV relief capacity.
In particular, a loss of load /
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l turbine trip (LOL/TT) transient from a reduced power condition may result in overpressurization of the main steam system.
With fully operational MSSVs, it can be demonstrated that overpressure protection is provided at all initial power levels.
However, TS 3.7.1.1 allows operation with a reduced number of operable MSSVs at a reduced power level.
In certain LOL/TT scenarios from low initial power levels with. pressure control available, the reactor trip may be delayed until low-low steam generator level is reached.
By this time, using the allowed reactor power levels of the current setpoints in TS 3.7.1.1 with one or more safety valves inoperable, the secondary side pressure may exceed the acceptance criterion of 110 percent pressure. As a result, Westinghouse recommended that the maximum allowable power range neutron flux high trip setpoints of TS 3.7.1.1 be lowered.
The staff has found these proposed changes to be acceptable as they are more conservative than the current TS requirements and are based on calculations performed by Westinghouse.
There is no operational impact on Braidwood or Byron since both stations have already adiainistratively incorporated these reduced power range neutron flux high trip setpoints as a result of NSAL-94-001.
2.4 Main Steam Safety Valve Setooint Tolerance a.
Description 4
Comed proposes to revise Table 3.7-2 to allow MSSV as-found setpoints to be within f3 percent of the lift settings.
The obsolete note is deleted and replaced with a note that requires all tested valves to be set to 1 percent tolerance. The new note maintains the current as-left setpoint tolerance requirement and removes the reference to the outdated provision.
The change is consistent with the ANSI /ASME OM-1-1981 Code, Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices" which is a more recent code endorsed by the NRC.
Comed has determined that, over an operating cycle, the setpoint of the MSSVs often change by more than 1 percent from the original set-pressure. As a result, when valves are tested and one or more are found outside the current il percent criterion, the plant is placed in an action statement.
Changing the as-found setpoint to f3 percent of the lift settings is expected to preclude frequently entering the action statement. Also, the conditions requiring the one-time note in Amendment No. 49 for Braidwood Station and Amendment No. 51 for Byron Station to address calculational errors in a vendor MSSV calibration procedure have been corrected.
b.
Imoact of Chanaes The effects of increasing the as-found lift setpoint tolerance on the MSSVs were examined by the licensee.
The various evaluations were made by Westinghouse (letter from Comed to NRC dated March 25,1997) and included non-loss-of-coolant accident and loss-of-coolant (LOCA) related evaluations.
The non-LOCA evaluations included AT protection and departure from nuclear
' e boiling (DNB) events.
The AT protection events included the affect of the increase in MSSV lift tolerance on the overtemperature AT and overpower AT setpoints.
The DNB events reviewed included the following areas (Updated Final Safety Analysis Report (UFSAR) sections given):
Feedwater system malfunction-reduction in temperature (15.1.1); Feedwater system malfunction-increase in feedwater flow (15.1.2); Excessive increase in secondary steam flow (15.1.3);
Inadvertent opening of a steam generator relief or safety valve (15.1.4);
Steam system piping failure (15.1.5); Partial loss of forced reactor coolant flow (15.3.1); Complete loss of forced reactor coolant flew (15.3.2); Reactor coolant pump shaft seizure (15.3.3); Reactor coolant pump shaft break (15.3.4); Uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal from a subcritical condition (15.4.1); Uncontrolled RCCA bank withdrawal at power (15.4.2); RCCA misalignment (15.4.3); Inadvertent operation of the Emergency Core Cooling System (ECCS) (15.5.1); Inadvertent opening of a pressurizer safety or relief valve (15.4.4); Startup of an inactive reactor coolant pump (15.4.4); and Chemical Volume Control System (CVCS) malfunction (boron dilution) (15.4.6).
It was found that the above non-LOCA DNB transients are not adversely impacted by the proposed change, and the results and conclusions presented in the UFSAR remain valid.
The following long-term heat removal events were reviewed:
Loss of non-emergency AC power to plant auxiliaries (15.2.6); Loss of normal feedwater (15.2.7); and Feedwater system pipe break (15.2.8).
It was found that the proposed change does not impact the long-term cooling overpressurization requirements.
For the -3 percer,t tolerance, the secondary steam releases generated for the offsite dose calculations for the following non-LOCA transients were examined:
steam system piping failure (Table 15.1-3), the loss of external load (Table 15.2-4), and the RCP shaft seizure (locked rotor - Table 15.3-3).
It was found that the current releases remain valid.
The loss of external load / turbine trip is the limiting non-LOCA event for overpressurization.
It was determined that this is the only UFSAR transient that is impacted such that a new analysis must be performed in order to address effects of the MSSV lift setpoint increase from 11 to f3 percent. The loss of external load / turbine trip event was analyzed by Westinghouse in order to quantify the impact of the setpoint tolerance relaxation. The evaluation by Westinghouse (letter from Comed to USNRC, dated April 29, 1996) concluded that all applicable acceptance criteria for this event remain satisfied and demonstrated that the conclusion presented in the FSAR remains valid.
Regarding the LOCA and LOCA-related evaluations, the following accidents were reviewed for the licensee by Westinghouse (letter from Comed to NRC dated March 25, 1997) for the effects of increasing the safety valve setpoint tolerance from 11 to f3 percent:
Large Break LOCA (15.6.5); Small Break LOCA (15.6.5); LOCA blowdown reactor vessel and RCS loop forces (3.9); LOCA mass and energv released for containment integrity analyses (6.2); Steam generator
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tube rupture (15.6.3); Hot leg switchover of the ECCS to prevent potential 4
boron precipitation (6.3.2.5); and Post-LOCA long-term core cooling (15.6.5).
The effect of an increase in the MSSV lift setpoint increase from 11 to i3 percent on the FSAR LOCA was evaluated.
In each case, the applicable regulatory or design limit was satisfied.
Specific analyses were performed for small break LOCA assuming the current MSSV TS set pressures plus the proposed additional 3 percent uncertainty.
The calculated peak cladding temperatures were well below the 10 CFR 50.46 2200 degrees Fahrenheit limit.
In summary, the departure from nucleate boiling design basis, primary and secondary pressure limits and dose release limits continue to be met.
For i
the LOCA, the peak cladding temperatures remain well below the limits specified in 10 CFR 50.46. Neither the mass and energy release to the containment following a postulated LOCA, nor the analysis of containment response following a LOCA, credit the MSSV in mitigating the consequences of an accident.
Therefore, changing the MSSV lift setpoint tolerances would have no impact on the containment integrity analysis.
In addition, based on the conclusion of the transient analysis, the change to the MSSV tolerance will not affect the calculated steam line break mass and energy releases inside
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containment.
The conclusions for the current accident analyses for LOCAs and non-LOCAs as described in the UFSAR remain valid.
Therefore, the staff finds the change to allow the M3SV as-found setpoints to be within i3 percent of the lift settings to be acceptable.
2.5 Chanae to Delete the Orifice Size Column from Table 3.7-2 a.
Descriotion i
Comed proposes to delete the orifice size column from Table 3.7-2.
This information is not used by the operator and the operators have no control of HSSV orifice size. MSSV capacity is discussed in UFSAR Section 10.3, Main Steam Supply System, and in UFSAR Chapter 15, Accident Analyses.
The information is not appropriate for inclusion in the TSs and is, therefore, proposed to be deleted.
b.
Imoact of Chanaes The staff finds this change to be acceptable as the proposed change does not introduce any new equipment, equipment modifications, or any new or different modes of plant operation. This change will not affect the operational characteristics of any equipment or systems.
The MSSVs are described in detail in the UFSAR and the information is not required in the TSs.
2.6 Technical Specification Bases 3/4.7.1.1 a.
Description l
The Bases for TS 3/4.7.1.1 contain information on secondary system pressure i
limits and how operability of the MSSV maintains design pressure during transients. The applicable ASME Code (1971 Edition) is also referenced when i
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discussing MSSV design flow rates and setpoints.
The bases also contain a discussion on thermal power limitations based on the total. number of safety valves inoperable.
j Comed proposed to add new information to the bases section. One addition states that the requirement to set the steam line safety valves to within il percent of the appropriate setpoint is consistent with Section III of the ASME Code.
The change also states that the allowed operating tolerance of i
f3 percent is supported by the Commonwealth Edison Company, Byron /Braidwood Unit 1 & 2 Overpressure Protection Report.
The equation that provides the j
bases for the reactor trip setpoint reductions is revised. The changes to the Bases are proposed to address and support the changes in TS 3.7.1.1 and Tables 3.7-1 and 3.7-2.
b.
Imoact of Chanaes No component or system operating characteristics will be affected by these revisions.
The proposed setpoints in Table 3.7-1 are more limiting than those currently allowed in Specification 3.7.1.1.
Westinghouse has determined that the current setpoints are non-conservative for some combinations of reduced 4
MSSV availability and reactor power levels.
Tho reactor trip settings were calculated using a revised methodology to account for the non-linear 4
relationship of reactor trip setpoints and reduced MSSV availability.
The revised equation, as proposed in the Bases, is used to calculate the reduced reactor trip setpoints.
By reducing the setpoints, the original design margin of safety is maintained.
The staff has found the revisions to Bases 3/4.7.1.1 to be acceptable.
Increasing the as-found valvo setpoint tolerance from il percent to 13 Dercent does not have a significant impcct on any accident.
The peak primary and secondary pressures remain below 110 percent of design at all times. The MSSVs are actuated after accident initiation to protect the secondary systems from overpressuriztion.
Increasing the as-found setpoint tolerance will not result in any hardware mcdification to the MSSVs.
Therefore, there is not an increase in the probability of the spurious opening of an MSSV.
Sufficient margin exists between the normal steam system operating pressure and the valve setpoint with the increased tolerance to preclude an increase in the probability of inadvertently actuating the valves.
3.0
SUMMARY
The impacts of the proposed TS changes have been revised by the staff as discussed in Section 2.
The proposed changes are acceptable.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments.
The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
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The amendments change 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 and change surveillance requirements. The NRC staff has 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 i
offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public connent on such finding (62 FR 11486). Accordingly, the amendments meet the eligibility criteria for j
categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR i
51.22(b), no environmental impact statement or environmental assessment need i
be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
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The Commission has concluded, based on the considerations discussed above, j
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, 4
i and (3) the issuance of the amendments will not be inimical to the common defense and secerity or to the health and safety of the public.
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Principal Contributors:
H. Balukjian 4
G. Hammer j
.0 ate: April 15, 1997 l
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