ML20216H548
| ML20216H548 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/15/1998 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 98-179, TAC-M97187, TAC-M97188, NUDOCS 9804210187 | |
| Download: ML20216H548 (11) | |
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Vnu;isir Ei.EcTune ann Powsu Cosivasy RICHN10ND, VHuilNI A 232(#1 April 15, 1998 U. S. Nuclear Regulatory Commission Serial No.98-179 Attention: Document Control Desk NL&OS/ETSR1 Washington, D.C. 20555 Docket Nos.
50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CilANGE REVISED LOOP STOP VALVE OPERATION On November 6,1996, Virginia Electric and Power Company (Serial No.96-532) requested amendments, in the form of changes to the Technical Specifications, to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes will modify the requirements for isolated loop startup to permit tilling of a drained isolated loop via backfill from the Reactor Coolant System through partially opened loop stop valves.
j In a March 16,1998 letter, the NRC staff requested additional information to continue review of the proposed Technical Specifications change. T!ie Attachment to this letter provides the information requested by the staff.
Should you have any questions or require additional information, please contact us.
Very truly yours, f
I P
d James P. O'Hanlon Senior Vice President - Nuclear M
Attachment Commitments made in this letter:
1.
None 9804210187 980415 PDR ADOCK 05000338 P
cc:
United States Nuclear Regulatory Commission Region 11 Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident inspector North Annq Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, VA 23219 l
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Response to Request for Additional Information Revised Loop Stop Valve Operation North Anna Power Station, Unite 1 and 2 (TAC Nos. M97187 and M97188)
NRC Question 1 How are the loops verified to be drained prior to opening the loop stop valves?
Response
Loops are verified as being drained by monitoring and recording the rate of Primary Drain Transfer Tank (PDTT) level changes at various points during the draindown evolution. Upon recording the initial PDTT inleakage rate, the Loop Drain Header Isolation, Loop Hot Leg Drain and the Loop Drain are then opened. When the PDTT inleakage approaches the initial rate recorded the loop is verified drained.
This verification eliminates the possibility of unsampled water at an indeterminate boron concentration diluting the water flowing in from the active portion of the Reactor Coolant System (RCS) and subsequently being passed through the core.
NRC Question 2 What is the maxino allowed inleakage into an isolated loop from the RCS and the SGs (i.e., permitted It'a stop valve leakage and allowable leakage from the secondary side through leaking SG tubes) over the two hours between verifying the loops are drained and opening the loop stop valves? What is the effect of this potentially unborated water on the shutdown margin?
Response
There is no specific limit on inleakage from the RCS. This is not needed since the boron concentration of the leaked fluid will equal the active RCS concentration. The PDTT leakage monitoring described in response to question number (1) verifies that backleakage is insignificant since PDTT inleakage must approach the value observed prior to opening the loop drains.
Tube leakage between the primary and secondary sides of the steam generator during and following backfill operations will be negligible. Specification 3.4.6.3 limits primary to secondary leakage during power cperation to 100 gallons per day (gpd). During power l
operation the differential pressure across the steam generator tube is of the order of
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1500 psid. During backfill operations the differential pressure across the tubes will be approximately an order of magnitude lower. Therefore, leakage through the tubes will have a negligible impact on the contents of the loop following com:aletion of the backfill operation. Even if a tube failure during shutdown operations is postulated, the effects of i
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backleakage from the secondary side to the primary side would be readily observed
' duririg the loop draining operations (see the response to Question 1).
NRC Question 3 There are no limitations on the RCS temperature or pressure, other than being in Modes 5 or 6. Please describe the safety impacts of opening the isolation valves at the temperature and pressure extremes allowed by the current Technical Specifications. If opening the valves at the extremes is not permitted or analyzed please clarify proposed specification 3.4.1.6.a to further limit this evolution.
Response
i Existing temperature and pressure limitations and controls as established in the Technical Specifications and Operating Procedures have been determined to be adequate for restoration of an idle loop, including backfill operations. The effects of loop backfill with a temperature differential between the active reactor coolant system and the secondary side of the steam generator are discussed in the response to 4
Question (4).
The maximum theoretical pressure differential across the loop stop valve during backfill operations would occur with a vacuum assist on the loop (i.e. loop pressure below atmospheric) and the active RCS pressure at or just below the residual heat removal suction relief valve setpoint of 467 psig.
Practical considerations dictate that the established differential pressure will be much lower than this. For example, the design differential pressure for operation of the loop stop valves is 200 psid and initial conditions will be established to meet this constraint with considerable margin. The current Surry procedure for performing this operation establishes an initial condition of either (a) reactor cavity flooded (for operation auring refueling shutdown) or (b) pressurizer level on scale and depressurized. The North Anna methodology will be similar.
Backfill to the loop is a slow, gradual process. The loop stop valve position and pressurizer level are carefully monitored and controlled to ensure that the backfill rate (inleakage to the loop) is within the capacity of the source of makeup to the active portion of the RCS. When RCS makeup is no longer required to ensure a stable level in the pressurizer or refueling cavity, the loop is considered full. Pressure equalization between the RCS and the loop occurs gradually as the loop stop valves are stroked open.
Even if a gross procedural violation occurred and the loop stop valve (s) were rapidly opened prior to filling the loop, the initial pressurizer level conditions are established to prevent loss of suction to the Residual Heat Removal pumps. The analysis to establish this level conservatively assumed simultaneous opening of all the stop valves with all Page 2 of 9
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loops drained and one of the loops under vacuum.
Because of the low initial temp'eratures associated with cold and refueling shutdown, even in this extreme situation the RCS depressurization would not be expected to result in any steam production in the core.
We therefore conclude that no additional Technical Specification constraints on temperature or pressure are needed to support the proposed method of operation.
l NRC Question 4
'li the SG secondary side is at a much higher or lower temperature than the RCS, how much will the overall RCS temperature change when the stop valves are opened (i.e.,
can the SGs act as a heat source or a heat sink)? Please be sure to consider if one or more of the other loops are in service with an RCP running. What is the effect of this potential temperature change on the shutdown margin?
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Response
As indicated in the response to question number 3, the loop stop valves are designed to open with less than 200 psid. With the loop at atmospheric pressure or a slight vacuum, this requires the active volume of the RCS to be at 200 psi or less to open the l
valves. Reactor Coolant pumps require greater than 200 psid to establish proper seal operation, which equates to approximately 300-325 psi in the RCS. Thus, a RCP will l
not be operating when loops are being backfilled.
The two scenarios are discussed below.
(a)
Steam Generator secondary side at a lower temperature For the case of the steam generator secondary at a lower temperature the maximum theoretical temperature difference would result from (a) the maximum allowable temperature in MODE 5, i.e. 200 F and the minimum physically achievable temperature i
in the steam generator secondary, i.e. 32 F. So the maximum theoretical temperature
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difference is 168*F. This is not a credible scenario, but it does represent a theoretical upper limit.
For operation at temperatures in the active portion of the RCS at < 200 F, the required boron concentration (the so-called minimum shutdown boron) ranges from about 1000 ppm at end of core life to in excess of 2200 ppm at beginning of core life. At these high boron concentrations, the absolute value of the isothermal temperature coefficient of reactivity is low, so the potential reactivity defect associated with introduction of colder water from a recently backfilled loop into the core is also very low.
For a limiting case of a backfill occurring at end of core life followed by introduction of Page 3 of 9
32 F water from the backfilled loop into a core operating at 200 F with no mixing
' betw'een the cold loop and other loops, we estimate a total reactivity addition of about i
800 pcm (1 pcm = 10'* % Ak/k), or less than one-half of the minimum shutdown margin required by Technical Specifications.
This is an extremely conservative assessment. It considers the following:
32 F water in the steam generator secondary side. More realistic values during.
shutdown range from 80 -180 F.
Maximum temperature of 200 F in the core. A more realistic value for this type of operation would be 140 F.
The most limiting time in life (end of life).
No credit for the heat capacity effects of the water in the SG being backfilled. In reality, as the backfill occurred, the secondary water would heat up as the i
primary water cooled down, resulting in convergence of the two temperatures to somewhere between the two initial temperatures.
No credit for mixing between loops after the RCP stert. The reactivity estimate above is for an instantaneous reduction of core fuel and water temperature from 200 F to 32 F.
Initial core shutdown margin equal to the minimum Technical Specification value of 1.77%. The available shutdown margin at cold shutdown is normally well in excess of this value.
In summary, the reactivity effects of backfilling into a cold SG are acceptable.
(b)
Steam Generator secondary side at a higher temperature.
It is highly unlikely for the secondary side of a steam generator to be at a higher temperature than the active portion of the reactor coolant system, since the normal heat source is decay heat on the primary side. As a limiting case, we considered backfill of the RCS at 100 F into a steam generator with secondary side temperature of 200 F.
We modeled free convection heat transfer on both sides of the tubes. Secondary inventory was set at 2500 cubic feet of liquid (more than enough to cover the tubes).
Perfect instantaneous fluid mixing on both sides of the tubes was assumed.
In any real backfill process, the reverse heat transfer would begin as soon as primary liquid is introduced into the SG tubes and the unfilled tube volume would be available to accommodate fluid thermal expansion. As a conservative simplification, we considered an instantaneous backfill, i.e. at Time 0 the primary sides of the tubes are filled with 100 F water and the secondary side is covered with 200 F water.
The graphs on the following pages illustrate the behavior of the key parameters during the loop fill evoludon.
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i Figure 1 Prima'ry/ Secondary Temperatures Following " Instantaneous' Steam Generator Backfill n r m i. s.cwm.d s me n,v.=,
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Figure 2 The surge late out of the affected SG tubes (into the pressurizer) due to thermal expansion.
Reverse NT tv, B.ckfleed SG 200F Seconda,y.100 F Primary Surge Rate kito t'rsesertner 360 p. -.
m ano m.-
25a
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m.-. --
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12s 4
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m The. Seconds Note that the calculated peak surge rate above is well within the capacity of a single pressurizer PORV when actuated at the Low Temperature Overpressure Protection System (LTOPS) setpoint. It is also less than the half of the rated capacity of one of the two residual heat removal RHR suction relief valves (900 gpm).
Page 5 of 9
Figure 3 The ihtegrated volume surge into the pressurizer Reverse HT to BackAlled 80 2007 Secondary.100 F Primary 20
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g.
3.._
2 0
0 20 40 60 80 100 120 140 180 180 200 Time. Seconds -
The total surge represents a change in pressurizer level of about 1.3% of indication span. As shown, the total insurge is less than the volume above the upper level tap in l
the pressurizer. What this says is that as long as the cold calibrated level channel is l
indicating on span, there is more than enough available vapor voluma in the pressurizer to accommodate the insurge.
l Once the backfill operation is complete, loop stop valves are completely opened and one or more reactor coolant pumps may be started. At this point some additional expansion of the primary coolant will take place as the comer (100 F in the example above) water moves through the hot steam generator, f
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To obtain a bounding assessment of the effects of opening the loop stop valves after j
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' completing the backfill evolution, we repeated the assessment above with the following
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l modifications:
The heat transfer coefficient on the pnmary side of the steam generator tubes was l
increased to a conservatively high forced circulation value of 8,000 btu /hr-ft _op,
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2 The active volume of the primary side was increased to the active RCS volume of l
8284 FT.
Instantaneous fluid mixing was assumed on both sides as before.
j Conservative initial temperature assumptions were: RCS; 100 F; steam generator l
secondary side: 200 F.
Free convection heat transfer was again assumed on the secondary side.
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Figure 4 l
Reverse HT to Backfilled SG 200 F Secondary -100 F Primary i
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_ 1 rore.d conv.ction on Primary Side i
o n Loop stop. ca.
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150 y
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135 m
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3 120 115 Primary 110 105 0
50 100 150 200 250 300 Time Seconde The total primary side temperature increase in this case is about 22.5 F.
Page 7 of 9
Figure 5 The ' volume surge rate Reverse HT to RCS kRet openine Loop stop vawes) 1 200 F Secondary.100 F Priinary
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Surge Rate into Prosaurizer 1=
7 seo e60 0
=
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l fn j
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=
y 300 250 200 l
rr
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=
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.0 The maximum surge rate remains below the relief capacity of a single RHR suction relief valve or a pressurizer power operated relief valve (PORV).
j The total volume surge (for the initial temperature differential of 100*F) is estimated to be about 45 cubic feet, or about 3.5% in pressurizer level span shown below in Figure 6. Again, this is l
less than the pressurizer volume available above the top pressurizer level tap.
Figure 6 Reverse HT to RCS
[Upon Opening Loop Stop Valves]
j 200 F Secondary.100 F Primary i
80 44 1
a 3
1 5n ja J,,0 1$
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10
,3.y 7
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2$
50 78 100 12S 16G 175 200 225 2$8 278 300 325 350 376
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%.m In summary, the effects of reverse heat transfer through the steain generator tubes during backfill and subsequent pump start are well understood and do not pose an unacceptable challenge to RCS integrity.
Page 8 of 9
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' NRC Question 5 Is there any potential that the gas volume in the loop will affect RHR flow when the isolation valves are opened? Describe how this gas volume is vented.
Response
The initial pressurizer level requirement has been established such that even if the loop stop valve (s) were rapidly opened prior to filling the loop, the initial pressurizer level conditions are established to prevent loss of suction to the Residual Heat Removal pumps. The analysis to establish this level conservatively assumed simultaneous opening of all the stop valves with all loops drained and one of the loops under vacuum.
The Reactor Coolant system water level will not drop below the middle of the reactor vessel nozzles. This ensures continued adequate suction conditions for the residual heat removal pumps. The loop gas volume will not affect RHR flow.
NRC Question 6 Proposed TS 3.4.1.6 requires 450 ft' of water in the pressurizer when filling the loop from the RCS volume. How many pressurizer level instruments are required to be operable in these modes?
Discuss if _ these level instruments are temperature compensated and any possible effects? What is the instrument uncertainty associated with pressurizer level measurement and was it included in the determination of the 450 ft' limit?
Response
The 450 ft' limit was developed from an extremely conservative analysis.
It was assumed that procedural violations result in simultaneous opening and ba'ckfilling of three drained loops, while in reality this operation will only be performed one loop at a time, as discussed in the BASES. As such, there was no need to specifically address pressurizer level instrument errors in the development of the limit.
During the backfill evolution, the level will be monitored via pressurizer level indicator RC-Ll-1462/2462. This is a level channel that has been specifically calibmted for cold conditions in the pressurizer.
In addition, the normal pressurizer level protection channels, which are calibrated for hot conditions (saturated liquid at 2235 psig), will be i
available for trending purposes.
No specific channel accuracy calculation has been done for RC-LI-1462/2462, but typical indication errors for channels of this type are of the order of 3% level span, which corresponds to a pressurizer water volume of about 40 cubic feet. This is accommodated by the conservatism discussed above.
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