ML20141H315
| ML20141H315 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/15/1997 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9707220333 | |
| Download: ML20141H315 (8) | |
Text
IUA Tennessee Valley Authority, Post office Box 2000, Soddy-Daisy, Tennessee 37379-2000 July 15, 1997 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555 Gentlemen:
In the matter of
)
Docket No. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SON) - RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION (TS) CHANGE 96-07
References:
1.
NRC letter to TVA dated March 18, 1997,
" Request for Additional Information -
Technical Specification Change Request TS 96-07 for Sequoyah Nuclear Plant Units 1 and 2 (TAC Nos. M95958 and M96599)"
2.
TVA letter to NRC dated March 27, 1997, on the above subject.
3.
TVA letter to NRC dated April 3, 1997, on the above subject.
In response to NRC questions contained in Reference 1, TVA provided responses to Questions 1, 4,
and 5 in References 2 and 3.
The enclosure to this letter provides responses 4 and 5 that address the balance of the questions (i.e.,
questions 2 and 3).
In addition, TVA is providing responses 1,
2, 3,
and 6 to address informal NRC questions discussed j
during a February 28, 1997 meeting and an April 2, 1997 telephone conference.
These responses are being provided to
{
support NRC review of TS Change 96-07, " Pressurizer Safety Valve (PSV) and Main Steam Safety Valve (MSSV) Setpoint Tolerance Increase."
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U.S. Nuclear Regulatory Commission Page 2 q
r July 15, 1997 l
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- Please direct questions concerning this issue to l.
D. V. Goodin at (423) 843-7734.
H i
Sincerel,
E i
i Pedro Salas
[
-Site Licensing and Industry Affairs Manager l
l L
Enclosure cci(Enclosure):
Mr. R. W. Hernan,-Project Manager Nuclear Regulatory Commission i
One White Flint, North 11555'Rockville Pike Rockville, Maryland' 20852-2739 i
NRC Resident Inspector
-Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624
}
Regional Administrator
[
U.S. Nuclear Regulatory Commission l
Region II Atlanta Federal-Center 61.Forsyth St.,
SW, Suite 23T85 Atlanta, Georgia' 30323-3415 t-I l.
i
's i
i
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i ENCLOSURE TVA RESPONSES TO NRC QUESTIONS FOR SON TS CHANGE 96-07, PRESSURIZER SAFETY VALVE (PSV)
AND MAIN STEAM SAFETY VALVE (MSSV)
SETPOINT TOLERANCE INCREASE i
Question l'.
The environmental consequences analysis for-the loss of AC power transient was evaluated in Section 4.2.8.1 of Framatome Safety Evaluation 77-1257639-01.
The present steam releases assumed in'the environh. ental consequences analysis were identified.as bounding for
- the -3% MSSV setpoint tolerance.
What was the calculated steam release increase when the -l% MSSV setpoint tolerance was relaxed to -3%?
Response
The present environmental consequences analysis for the loss of AC power transient conservatively assumes that the MSSVs actuate 10% below the nominal valve setpoint.
The values in FSAR Section 15.5.1 reflect the -10%
setpoint tolerance.
Since the assumed setpoint tolerance bounds the -3% setpoint tolerance, specific analyses.for -1% and -3% setpoint tolerances;were not performed.
While specific data is not available for steam release differences between -1% and -3% of nominal _ valve setpoints, the_ identified analys is is bounding for the -3% setpoint tolerance.
Question 2.
The environmental consequences analysis for the steam line break transient was evaluated in Section 4.2.8.2 of Framatome Safety Evaluation 77-1257639-01.
The i
presentsteam releases assumed in the environmental l
consequences analysis were identified as bounding for the -3% MSSV setpoint tolerance.
What was the calculated steam release increase when the -1% MSSV setpoint tolerance was relaxed to -3%?
Response
The.present environmental consequences analysis for the
)
steam line break transient conservatively assumes that j
the MSSVs actuate 10% below the nominal valve setpoint.
The values'in FSAR Section 15.5.4 reflect the -10%
setpoint tolerance.
Since the assumed setpoint tolerance bounds the -3% setpoint tolerance, specific analyses for -1% and -3% setpoint tolerances were not performed.
While specific data is not.available for steam release differences between -1% and -3% of nominal valve setpoints, the identified analysis is bounding for the -3% setpoint tolerance.
Question 4
3.
Section 4.2.8.3 of Framatome Safety-Evaluation 77-1257639-01/ indicates that the-environmental consequences analysis of the ejected rod transient _is bounded by the results of the loss-of-coolant accident
' environmental analysis.
No further analysis of the L-ejected rod transient is presented.
Section 15.4.8, l
Appendix A, of-the Standard Review-Plan (NUREG 0800),
establishes the~ acceptance criteria for the ejected. rod 1
L
This
. acceptance criteria is more conservative than the loss-i of-coolant accident acceptance criteria.
Please confirm that the environmental consequences of the ejected rod transient continue to meet the "well within" 10CFR100 limit acceptance criteria with the relaxed MSSV setpoint tolerance.
Response
l As-indic'ated in Framatome Safety Evaluation 77-1257639-01, the present ejected rod transient l
assumes that 81,000 lbs of steam is discharged through L
the MSSVs'during the transient.
The results of the analysis are dominated by releases from the primary system (reactor building leakage) rather than secondary system atmospheric steam releases.
While a specific transient analysis has not been performed to establish the exact steam release for the increased MSSV setpoint tolerance, an evaluation has been j
performed which conservatively assumes that the entire l
secondary system inventory (440,000 lbs) will be released to the atmosphere.
This evaluation concluded j
that thc. expected radiation exposure for the two-hour exclusion area boundary (0.5 Rem Whole Body, 35 Rem l
Thyroid) and the 30-day low population zone (0.1 Rem l
Whole Body, 10 Rem Thyroid) continue to be "well l
within" the 10CFR100 acceptance criteria of 25 Rem-Whole Body, 300 Rem Thyroid, i
i i
l
l j
~ Question l
- 4..
'In the August-28, 1996 submittal, you indicated that
+3% and +5% tolerances were used-for the MSSVs and PSVs, respectively.
However, the Mark-BW Fuel Assembly Topical-Report indicates that 6% and 5%
tolerances were used for the MSSVs and PSVs,
{
-respectively.
Provide a discussion of the actual tolerances used for'the peak pressure analysis.
- Also,
' discuss and justify your modeling of.(or assumptions L
for) valve lift characteristics and accumulation.
i
. Response Tolerances of +6% and +5% for the.MSSVs and PSVs, respectively, were assumed in the analysis of the limiting pressurization event for SQN (i.e.,
loss of electric load).
In all other transients analyzed to support the fuel reload, the modeling of MSSVs and PSVs is not as critical.
Tolerances of +3% and'+5%
for the MSSVs and PSVs, respectively, were assumed for these events.
Safety valves are modeled to open abruptly at valve setpoint plus tolerance.
This modeling closely simulates actual valve characteristics observed in EPRI safety relief valve tests (EPRI Document NP-2770-LD, Volume 6, Research Project V102-2,"EPRI/C-E 1
PWR Safety Valve Test Report", March 1983) in that there is little or no accumulation associated with the operation of these valves..With no accumulation, the valve tolerances indicated above are adequate to bound actual valve performance.
Question 5.
Current TSs require the PSV lift setpoint to be set at.2,485 psig (2500 psia).
According to Table 6.1-3, l
of the Mark-BW Fuel Assembly Topical Report, the high pressurizer pressure reactor trip setpoint can be as high~as 2,425 psig including instrumentation channel-error and setpoint error.
For the proposed negative i
l tolerance change of -3% show that the: change does not L
result in lifting the PSVs at or before the reactor trip.
Provide a similar discussion for the effects of the -3% tolerance for the PSVs on the operation of the pressurizer PORVs and -3% tolerance for the MSSVs l
on the steam generator relief valves.
4 4
1 l
Response
SQN TS Table 2.2-1, indicates a high pressurizer pressure setpoint of 2,385 psig. Section 3.4.1.1, of the TSs indicates the requirement for a pressurizer safety valve setpoint of 2,485 psig.
This givesLa l
difference in setpoints values of 100. psi.
The_ loop uncertainty for the high pressurizer pressure trip setpoint instrumentation is 5.6% of instrument span.
With a span from 1,700 psig to 2,500 psig, the loop uncertainty is equivalent to 44.8 psi. An.as-found tolerance allowance of -3% on the pressurizer safety valve setpoint (74.6 psi) is proposed.
Since the J
reactor trip setpoint uncertainty and safety valve j
tolerance have no correlation with each other, they may be combined in a square-root, sum of the squares relation:
uncertainty = (44.8 + 74.6)u2 = 87.0 psi 2
2 The combined uncertainty associated with the RPS instrument error and the safety-valve tolerance is less than the 100 psi difference between the low pressure setpoint and the pressurizer safety valve j
setpoint.
Thus, there is no overlap of functions.
The main steam atmospheric relief. valve setpoint is 1,025 psig (Section 10.3.2.1 of the SON FSAR).
This is'approximately 3.7% below the lowest MSSV setpoint,. greater than the proposed 3% tolerance.
The pressurizer PORV setpoint is 2,335 psig, approximately 6.6% below the pressurizer safety valve setpoint, greater than the 3% tolerance.
Question 6.
FSAR Section 5.2.2.3, Page 5.2-38, third from last paragraph, states, "The upper limit of overpressure protection is based upon the positive surge of the reactor coolant produced as a result of turbine trip under full load, assuming the core continues l
to produce full power.
The self-actuated safety valves are sized on the basis of steam flow from j
the pressurizer to accommodate this surge at a j
setpoint.of 2,500 psia and a total accumulation of 3 percent.
Note that no credit is taken for the relief capability.provided by the power operated relief valves during this surge."
i
,e-In addition, Reference 3 to that FSAR section (WCAP-7789) which describes the sizing of the
. pressure relief devices, also states that for
~
sizing purposes, no credit is taken for the reactor protection system.
1 Please justify the acceptability of the changes to the tolerance with this regard (i.e.,.with regard to sizing of the PSVs).
Also, provide a discussion j
and a similar justification of the requested changes with respect to the sizing of the MSSVs..
Response
The steps described above pertain to the initial j
design and rizing of the pressurizer and i
pressurizer safety valves.- This should not be l
interpreted as a requirement'for evaluating the change to these components, particularly in the instance in which a re-design or replacement is not being sought.
In this case, the rated valve flow and pressure is unchanged.
The design basis of the plant was reviewed with respect to the relaxed' safety valve setpoint tolerance.
The evaluation was performed in accordance with the requirements of Regulatory Guide 1.70 and the"ASME
'f code.
\\
The loss of electric load event it the limiting pressurization event for Sequoyah.
The event was analyzed from 102% power, in a manner that i
resulted in a maximum insurge to the pressurizer.
A single failure was taken, associated with the failure of the reactor to trip on. turbine trip.
The' safety valve setpoint tolerances were modeled in accordance with the relaxed setpoint tolerances.
The analysis shows that, regardless of the method chosen to size the safety valves, the applicable Regulatory Guide 1.70 and ASME acceptance criteria for this event are met with MSSV and PSV setpoint tolerances of +6% and +5%,
respectively.
.,.