ML020290177
| ML020290177 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/18/2002 |
| From: | Hartz L Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 01-720 | |
| Download: ML020290177 (12) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 18, 2002 U.S. Nuclear Regulatory Commission Serial No.01-720 Attention: Document Control Desk NL&OS/ETS R2 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 CHANGES ELIMINATION OF SEISMIC EFFECTS FROM CONTROL ROD DROP TIMES REQUEST FOR ADDITIONAL INFORMATION In a June 22, 2000 letter (Serial No.00-307) and a July 26, 2001 letter (Serial No.01-359), Virginia Electric and Power Company (Dominion) requested amendments to the Facility Operating Licenses NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively.
The proposed changes would add a risk-informed license condition.
The license condition will eliminate the consideration of the effects of a concurrent seismic event on the rod control cluster assembly (RCCA) drop time for the non-LOCA accident analyses.
In a November 16, 2001 letter, the NRC requested additional information regarding the analysis used to develop the seismic allowance currently applied to the rod control cluster assembly. The attachment to this letter provides the requested information.
If you have any further questions or require additional information, please contact us.
Very truly yours, Leslie N. Hartz Vice President - Nuclear Engineering Attachment Commitments made in this letter: None
cc:
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060
SN: 01-720 Docket Nos.: 50-338/339
Subject:
Proposed TS Change RAI - Elim. Of Seismic Effects from CRD Times COMMONWEALTH OF VIRGINIA
) )
COUNTY OF HENRICO
)
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.
Acknowledged before me this 18th day of January, 2002.
My Commission Expires: March 31, 2004.
otary Public (SEAL)
Response to Request for Additional Information Elimination of Seismic Effects from Control Rod Drop Times NRC Question 1 Assuming (1) control rods are subject to the postulated seismic-related control rod drop time delay for applicable seismic events, and (2) required reactor trip signals are successful, would reactor coolant system (RCS) pressure open the pressurizer power-operated relief valve(s) and/or the safety relief valves? If so, would the valves be required to function in a steam or water environment?
Provide a discussion on the associated risk of these considerations in terms of Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis" guidance.
Response
The June 22, 2000 (Serial No.00-307) submittal contains a section that discusses defense-in-depth. In this section, the Loss of Load Accident is discussed as the limiting scenario for overpressure transient analyses. The present analysis of record (AOR) makes several conservative assumptions. Using these assumptions, pressure relief is required. However, the reactor actually trips on reactor trip following turbine trip, which occurs sooner than the high pressurizer pressure trip as assumed in the AOR. When the reactor trips on the first signal actually received, pressure relief is not required even if a delay in the rod drop time equivalent to that assumed to result from a seismic event should occur. In actuality, the reactor trip on turbine trip occurs very quickly and limits the system pressure response.
The AOR is but one of many possible scenarios that result from transient operation. As discussed above for a loss of load due to a turbine trip, the PORVs are not expected to open because the turbine trip initiates a reactor trip. For slower transients (such as rod withdrawal or a turbine runback) the transient dynamics may be such that the PORVs open before the reactor trip occurs. When the PORVs open before the reactor trips there is no impact with respect to the rod drop time issue by definition. Similarly, if the trip occurs more than a second or two before the PORV setpoint is reached, the PORVS will not open. Therefore, there is no impact from the rod drop time.
For the rod drop time to have an impact on whether or not the PORV opens, the transient dynamics must be such that the PORVs would open at or just after reactor trip.
In these cases, the seismic effect could contribute to an increased likelihood of PORV demand.
However, the frequency of these "smart" scenarios is small by definition because they require the near simultaneous occurrence of a PORV demand at the time of trip with a concurrent seismic event.
In the conservative AOR for the loss of load scenario only steam is relieved into the pressurizer relief tank.
This result is obtained even though the PORVs are not assumed to be operable and the safety valves operate at the high end of the pressure Page 1 of 9
setpoint range. More generally, it is expected that only the PORVs would be required to operate and that the small impact on pressure would not change the dynamics of the system such that water relief occurs instead of steam relief.
The pressurizer steam volume is large compared to the size of the PORVs so steam relief would be expected to last much longer than a pressure rise due only to the seismic delay on control rod insertion.
From a Regulatory Guide 1.174 perspective, the risk associated with this event would be evaluated as a transient event in which the RCS integrity is potentially lost as a result of the stuck open PORV. A review of the transient event tree from the internal events model shows that the conservative assumptions regarding pressure relief are not included in this event tree. The RCS integrity function has been removed from the tree because the original IPE analysis showed that the sequence frequency for the transient with a loss of RCS integrity was about four orders of magnitude smaller than the initiating event frequency for a small LOCA. The increase in risk for those pressure increases resulting from a delay in rod drop time can be approximated as the product of the transient initiating event frequency, the RCS integrity unavailability, and the conditional probability of core damage from a small LOCA. For the proposed change the CDF increase is 1.01 E-8/yr. The transient initiating event frequency is 1.95E0/yr.
The RCS integrity unavailability is 1.22E-5. The small LOCA conditional core damage probability is the ratio of the small LOCA contribution to core damage frequency to the initiating event frequency (8.95E-6/yr/2.1E-2/yr).
Thus, the increase in risk for a pressure increase due to a delay in rod drop from a seismic event that only impacts the internal events PRA is conservatively shown to be less than 10E-6/yr.
NRC Question 2 Please provide for review your seismic risk analysis of the failure of a turbine trip-reactor trip scenario mentioned in your July 26, 2001 submittal. Discuss the risk significance of this consideration using the guidance in R.G. 1.174.
Response
The enclosure to the July 26, 2001 RAI response (Serial No.00-359) contains a discussion of the seismic risk due to failure of RCS integrity following a seismically induced loss of offsite power. This discussion can be summarized as follows:
- 1. A seismically induced loss of offsite power (LOOP) was found to contribute to core damage frequency in the Surry Seismic PRA developed in response to GL 88-20, Supplement 4 (IPEEE).
- 2. The impact of overpressure events is explicitly considered in the analysis because the final node of the seismic event tree, CCDP, contains cut sets from the event tree for a LOOP from the internal events model. These cut sets represent the random failures that can occur including failures leading to a loss of RCS integrity.
- 3. Thus, sequence number S10 from the seismic event tree explicitly considers the "seismic risk of the failure of the turbine trip-reactor trip scenario." The sequence frequency is 3.3E-6/yr and the CCDP from random events is 0.12.
Page 2 of 9
- 4. The top cut sets from the CCDP evaluation were provided and PORV failure to close was not among the top contributors.
- 5. Finally, the North Anna event tree that is equivalent to the event tree used to quantify the CCDP for the Surry Seismic PRA was provided to show that the contribution to core damage frequency from a stuck open PORV following a LOOP is negligible.
The precise increase in core damage frequency for the proposed change is difficult to quantify for two reasons. The requisite models are not available and the change is very small. The above summary of the previous submittal illustrates both points. We have used inputs from a combination of the North Anna internal events model and the Surry seismic model as the best available tools. These models indicate that the contribution to core damage from a stuck open PORV following a seismic event with a loss of offsite power is not among the top cut sets.
In fact, it is not even a developed end state because the likelihood is so small. (See sequence T1 P30 or Ti 1 RC from the event tree of the July 26 submittal also attached herein.) The most likely outcome of the seismic event is that the switchyard would fail following the event, but the diesels would start and power would be available to close the PORV. If one or both PORVs would fail-to close, power would be available to close the PORV block valves.
The RCS integrity function, 1RC-11, in the enclosed event tree is the only failure for Sequence TI P30. The top fifty cut sets for this function are also enclosed. The top four cut sets from this list are those discussed in the meeting with the NRC staff in March 2001. As can be seen from the list even if one of the basic events was increased by a factor of 10 the result would still be several orders of magnitude below the dominant cut sets presented in the July 26, 2001 submittal (Attachment 4, page 9 of 10). That is, the dominant cut sets are on the order of 1E-2 while the cut sets for failure of the PORV would be on the order of 1E-5. The seismic sequence frequency is 3.3E-6/year. This number includes the seismic convolution of the hazard and the fragility along with the random failures. The dominant random failures are going to continue to be in the 1 E-2 range so a change in the less dominant random cut sets from 1 E-5 to 1 E-4 is not going to increase the overall sequence frequency.
The guidance in Regulatory Guide 1.174 regarding risk significance recommends that an application for a change to the licensing basis should include an evaluation of the change in core damage frequency and large early release frequency as a result of the proposed change. The proposed change has been evaluated using bounding calculations. Additionally, the Surry seismic PRA model was used to infer the seismic impact assuming that the most likely result of the seismic event would be a loss of the switchyard. The estimated increase in CDF using either method is less than 10E-6/yr.
This small increase in CDF in combination with a baseline CDF in the 10E-5 range means that the proposed change to the license basis would be evaluated against the acceptance criteria for Region Ill.
In this region, small risk increases are permitted without requiring a detailed analysis of the change in total core damage frequency.
Based on these conclusions the small increase in risk associated with the elimination of Page 3 of 9
the seismic penalty from the control rod drop time meets the acceptance criteria in RG 1.174.
NRC Question 3 The submittal dated June 22, 2000 indicated that if accident analyses are re-performed, the rod drop time of 2.7 seconds would need to be increased. Such a change would involve concurrent reactor protection system changes (e.g., reductions in high pressurizer pressure and/or low RCS flow reactor trip setpoints).
The submittal indicated this would have the potential to reduce normal operating margin and increase the potential for reactor trip events and associated plant equipment transients. Please discuss the potential magnitude of the decreases in the setpoints.
Response
Our evaluation of the effects of using a full core of advanced fuel products (with their associated higher pressure drop and reduced thimble tube ID) concludes that measured rod drop times could potentially increase by as much as 0.5 seconds at North Anna.
This increase results from the fact that advanced fuel assemblies have a higher hydraulic resistance and core pressure drop than current generation fuel. This higher pressure drop forces more flow up the RCCA guide tubes, creating more resistance to control rod insertion and, therefore, slightly delayed drop times. The seismic effect amplifies the magnitude of the estimated drop time increase.
Current measured drop times provide slightly in excess of 0.5 second of margin to the 2.7 seconds Technical Specification limit (see attached Figure). However, much of this margin is currently allocated for the seismic allowance.
If the seismic penalty is not eliminated, Dominion estimates that the safety analysis rod drop time will have to be increased from 2.7 seconds to 3.2 seconds to ensure adequate margins for BOC startup tests.
Sensitivity studies have been performed with our safety analysis models to estimate the protection setpoint adjustments that would be required to offset a 0.5 second additional delay in rod drop time. The results are summarized below.
Case 1:
Transient - Complete loss of RCS flow Acceptance Criterion - Hot channel DNBR Source of protection-Low RCS Flow Current safety analysis protection setpoint - 87.0% of full flow Required adjusted setpoint to offset 0.5 second rod drop time delay:
88.9% of full flow (+1.9%).
Page 4 of 9
Case 2:
Transient - Loss of external electrical load Acceptance Criterion - Peak RCS pressure Source of protection-High pressurizer pressure reactor trip Current safety analysis protection setpoint - 2381 psig (2360 psig Technical Specification Setpoint + Instrument Uncertainty + Margin)
Required adjusted setpoint to offset 0.5 second rod drop time delay: 2355 psig (-26 psig)
(2334 psig Technical Specification Setpoint + Instrument Uncertainty + Margin)
The second case is of particular concern, since it essentially consumes (by greater than 50%) the currently allocated operating margin between the nominal pressurizer PORV setpoint (2335 psig) and the high pressure reactor trip setpoint. Therefore, the loss of margin has the potential to significantly reduce the effectiveness of the pressure control system to prevent reactor trips.
Given the demonstrably low probability of a significant seismic event, Dominion continues to believe that elimination of the seismic allowance from the control rod drop time requirement will result in an enhancement to overall reactor safety for North Anna following introduction of full cores of advanced design fuel.
Page 5 of 9
Figure 2.2 NORTH ANNA UNIT 1 -
CYCLE 13 STARTUP PHYSICS TESTS ROD DROP TIME -
HOT FULL FLOW CONDITIONS R
P N
M L
K H
G F
E D
C a
A I I I
I I I 1
1.73 1.62 I1.75 2
11.68 1.70 I
1 1.75 1 1.68 1 1
1.79 1 1.7 1 1
4 I
1.781 1-78 5
I.01 1.7Z
- 9.
1 a
I I
I I
I I
I I
1.68 1
0 7B 1 1.70 1 1 1-74 1
7 1.76 1
7.79
.0 1 D I
I1.I 1.7 1.70Z77 1
I 8
1.1 2
1.1.70771.97 I
9
.15 I.X 1 -
O DRO TIM 10 DS I E TY ISEC)
I I
NE13 NI1 Stru Py ic Tet Reor Pag 21 of 57 I1.00 I l.7 T
I I.70 I 1.74 1.76 I 1.77 I
.00 I 1
I 1.151 I
I 1.761 I
II1 I.69 I
I 0.81 I I
1 71 I I 1.73 I 2
I I I
I I
I I
I I
I I
1.7T 11.721 I
Il 13 I
I I I
I I
I I 1.97 I 1.75 I I 1.76 I I
1 I
I I
I15 I 1.90 I--> ROD DROP TIME 10 IAS)IPOT EN002 01EC)
NE-1133 NICI3 Startup Physics Tests Report Page 21 of 57 Page 6 of 9
NUPRA 2.33 FILE
- lRC-1I.FTP NURELMCS Solution VIRGINIA Minimum Cut Set Solution for fault tree RC100 Serial no.=
7 Performed :
21:19 13 Feb 1998 Cut Set Equation produced is
- 1RC-II.EQN RCS PORVs Fail To Reclose -TR NAPS Unit 1 At-Power PSA, N7B Top event: GRC1112 Top event unavailability (r.ev. appr)-
1.224E-005 Cutoff value used 1.00E-009 Number of Boolean Indicated Cut Sets 5.107597E+00 Number of MCS in equation file 65 MINIMAL CUT SETS SORTED BY UNAVAILABILITY 3.126E-006 3.126E-006 1.812E-006 1.812E-006 2.489E-007 2.489E-007 1.873E-007 1.873E-007 1.742E-007 1.742E-007 1.475E-007 1.475E-007 7.613E-008 7.613E-008 5.729E-008 5.729E-008 5.329E-008 5.329E-008 3.885E-008 3.885E-008 3.506E-008 3.506E-008 2.924E-008 IRCPORV-T3 1RCRV--FO-1455C 1RCPORV-T3 IRCMOV-FO-1536 1RCMOV-FO-1535 IEEEDG-TM-EEEGIJ HEP-0OP6:3 lEEEDG-TM-EEEGIH HEP-0OP6:3 IEEEDG-FS-IJ HEP-0OP6:3 1EEEDG-FS-1H HEP-0OP6:3 1EEEDG-FR-IH HEP-0OP6:3 IEEEDG-FR-lJ HEP-0OP6:3 lEGEDG-CC-ALL IEGEDG-CC-ALL 1EEEDG-TM-EEEG1H AACEDG-FS-DGOM IEEEDG-TM-EEEGIJ AACEDG-FS-DGOM IEEEDG-FS-lJ AACEDG-FS-DGOM 1EEEDG-FS-1H AACEDG-FS-DGOM 1EEEDG-FR-IH AACEDG-FS-DGOM IEEEDG-FR-IJ AACEDG-FS-DGOM 1EEEDG-TM-EEEGlJ AACEDG-TM-DGOM IEEEDG-TM-EEEGlH AACEDG-TM-DGOM IEEEDG-TM-EEEGIH AACEDG-FR-DGOM IEEEDG-TM-EEEGOJ AACEDG-FR-DGOM IEEEDG-FS-lJ AACEDG-TM-DGOM IRCRV--FO-1456 IRCPORV-T3 IRCPORV-T3 1RCPORV-T3 IRCPORV-T3 IRCPORV-T3 1RCPORV-T3 IRCPORV-T3 1RCPORV-T3 1RCPORV-T3 IRCPORV-T3 IRCPORV-T3 IRCPORV-T3 IRCPORV-T3 IRCPORV-T3 IRCPORV-T3 1RCPORV-T3 1RCPORV-T3 IRCPORV-T3 1RCPORV-T3 1RCPORV-T3 IRCPORV-T3 HEP-IE0-22 HEP-IEO-22 1RCRV--FO-1455C IRCRV--FO-1456 1RCRV--FO-1456 IRCRV--FO-1455C IRCRV--FO-1456 1RCRV--FO-1455C 1RCRV--FO-1455C lRCRV--FO-1456 1RCRV--FO-1455C 1RCRV--FO-1456 IRCRV--FO-1455C IRCRV--FO-1456 IRCRV--FO-1456 IRCRV--FO-1455C 1RCRV--FO-1455C IRCRV--FO-1456 IRCRV--FO-1456 IRCRV--FO-1455C 1RCRV--FO-1455C IRCRV--FO-1456 IRCRV--FO-1456 Page 7 of 9 1.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
- 8.
- 9.
- 10.
- 11.
- 12.
- 13.
- 14.
- 15.
- 16.
- 17.
- 18.
- 19.
- 20.
- 21.
- 22.
23.
- 24.
2.924E-008
- 25.
2.720E-008
- 26.
2.720E-008
- 27.
2.639E-008
- 28.
2.639E-008
- 29.
2.454E-008
- 30.
2.454E-008
- 31.
5.781E-009
- 32.
5.781E-009
- 33.
5.586E-009
- 34.
5.586E-009
- 35.
5.586E-009
- 36.
5.586E-009
- 37.
5.586E-009
- 38.
5.586E-009
- 39.
4.351E-009
- 40.
4.351E-009
- 41.
4.046E-009
- 42.
4.046E-009
- 43.
3.590E-009
- 44.
3.590E-009
- 45.
3.485E-009
- 46.
3.485E-009
- 47.
- 48.
- 49.
50.
3.159E 3.159E 2. 012E 2.- 012E-
-009
-009
-009
-009 1EFEDO FS-lH AACEDG-TM-DGOM 1EEEDG-FR 1J AACEDG-TM-DOOM 1EEEDG-FR 1H AACEDO TM-DOOM 1EEEDG-ES 1J AACEDG-FR-DGOM 1EEEDG-FS-1H AACEDG-FR-DOOM 1EEEDG-FR 1H AACEDG-FR-DGOM 1EEEDG-FR 1J AACEDG-FR-DGOM 1EEEDO--TM-EEEO1H 1RCRV-ED 14550 1EEEDG-TM-EEEG1J 1RCRV---FO 1456 1EEBKR-SO-15J8 1EEBKR-SO-15H8 1EEBKR-SO-14H1 7 1EEBKR-SD-14HI-1 1EEBKR-SD 14J1 1EEBKR-SD 14J5 lEEEDO ES 1H 1RCRV--ED 14550 1EEEDO-ES 1J 1RCRV-EO-1456 IEEEDO-ER 1H 1RCRV-ED 1455C 1EEEDO FR-1J 1RCRV-ED 1456 1EEBKR-ED 15J2 HEP-00P6:3 1EEBKR-ED 15H2 HEP-00P6 :3 1EOEDO CC 1H-1J HEP-ODP6 :3 1EOEDO-CC 1H 1J HEP--0P6 :3 1EETEM-LP-IHI IEETFM-LP-1J 1EEBUS-LU 1H 1EEBUS-LU IJI 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1EPBKR-FC-iSFi IEPBKR-FC-15DI 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1EPBKR-EC-15E1 IEPBKR-FC-15D1 1EPBKR-FC-15Fi IEPBKR-EC-15D1 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 1RCPORV-T3 IRCRV -ED 1455C 1RCRV-ED 1456 1RCRV -ED 14550 1RCRV---FO 1456 1RCRV-ED 1455C 1RCRV-ED 1455C 1RCRV-ED 1456 1RCPORV-T3 1RCPDRV-T3 1RCRV-ED 1456 1RCRV-ED 1455C 1RCRV-ED 14550 1RCRV-ED 1455C 1RCRV--FO 1456 1RCRV-ED 1456 1RCPDRV-T3 1RCPDRV-T3 1RCPDRV-T3 1RCPDRV-T3 1RCRV-ED 1456 1RCRV-FD 14550 1RCRV-ED 1455C 1RCRV-ED 1456 1RCRV--F-ED14550 1RCRV-- ED 1456 1RCRV -ED 1455C 1RCRV-ED-1456 Page 8 of 9
POI TI OK P02 T11PW OK P03 TIIFWIBLD O0 PII TIIFWIOIL01SW O0 iLRCF.
P05 TIIFWIBLDISWLERF I
S.800.1 0
P00 TlIFW18LDOISRS OK<
LR02 PO7 TIIFWlBLOIOSRSLERF 2
1.42E.10 P08 TIIFWiCHR OK JLR21 PIS TI1ITWIC1RLEOF 21 1.200.00 H
lW4P10 T11FWICHOlSW OK ICR2LR-2 P11 T11FWICHR18WLERF 22 0 00E-00 P12 TlIFWICHRIOSRS OK LR23 P13 Ti1FWICHRIOSRSLERF 23 0.00E-00 P14 TI1FWICH OK L RF2 PIS TIIFWICHLERF 20 1.67E.06 P16 TIIFWICHISIR OK LIRF-1 P17 TlIFWICJI1SIRLERF 21 5.82E.09 11H-
.0-P18 TIIPWICK1SW 0K Pig TIIPWICHISWLERF 22 2.056-09 P20 T11FWICHIOSRS 00 LERF.23 P21 TI11FWIC1 Q100SLERF 23 09065E08 P22 TIIFWIRC 00 LR08 P23 T11FWIRCLERF 0
8.94E.07 ISR7P24 T11FWIRCISIR 0K LR-9 P25 T11FW1RC1SIRLERF 9
4,86E.09 P26 T1OFWIRCiSW OK IR-21S4ý-0 P27 T1IFW1OCISWLERF 10 0000.+00 P28 TI1FW1RC105RS 0OK OSS3LOOP
=1 III T111IR1011OSRSLERF 11 1.376.07 IOC-11 P30 TiIOC TO 1.40E.06 TTR lE-TI IVP31 T11HV TO 708,.04000 ZEE-EG
-P32 T12EE T
3.48E-06 2TR 166.00 P33 T11EE TO 1,36E.04 TTR TI: LOSS OF OFF-SITE POWER (LOOP)