LR-N07-0154, Response to Request for Additional Information Request for License Amendment - Extended Power Uprate
| ML071840167 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/22/2007 |
| From: | Barnes G Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LCR H05-01, Rev 1, LR-N07-0154 | |
| Download: ML071840167 (41) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG NuclearLLC 10 CFR 50.90 LR-N07-0154 LCR H05-01, Rev. 1 June 22, 2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354
Subject:
Response to Request for Additional Information Request for License Amendment - Extended Power Uprate
Reference:
- 2) Letter from USNRC to William Levis, PSEG Nuclear LLC, June 7, 2007 In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to increase the maximum authorized power level to 3840 megawatts thermal (MWt).
In Reference 2, the NRC requested additional information concerning PSEG's request. to this letter provides responses to the balance of plant and health physics branch questions. The responses to the remaining questions in Reference 2 are being provided in a separate transmittal.
PSEG is also providing a revised response to request for additional information (RAI) 14.47. The previously submitted response did not reflect the completed re-analysis of the feedwater nozzles. Changes from the previously submitted response are marked.
PSEG has determined that the information contained in this letter and attachment does not alter the conclusions reached in the 10CFR50.92 no significant hazards analysis previously submitted.
95-2168 REV. 7/99
LR-N07-0154 LCR H05-01, Rev. 1 June 22, 2007 Page 2 There are no regulatory commitments contained within this letter.
Should you have any questions regarding this submittal, please contact Mr. Paul Duke at 856-339-1466.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on 6/ &/z70o (date)
Sincerely, George P. Barnes Site Vice President Hope Creek Generating Station Attachment
- 1. Response to Request for Additional Information cc: S. Collins, Regional Administrator - NRC Region I J. Shea, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek K. Tosch, Manager IV, NJBNE
Attachment I LR-N07-0154 LCR H05-01, Rev. I Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Extended Power Uprate Response to Request for Additional Information In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to increase the maximum authorized power level to 3840 megawatts thermal (MWt).
In Reference 2, the NRC requested additional information concerning PSEG's request.
PSEG's responses to the balance of plant and health physics branch questions are provided below. The responses to the remaining questions in Reference 2 will be provided in a separate transmittal.
PSEG is also providing a revised response to request for additional information (RAI) 14.47. The previously submitted response did not reflect the completed re-analysis of the feedwater nozzles. Changes from the previously submitted response are marked.
7 Balance of Plant Branch (SBPB) 7.16 Your response to RAI 7.13 for reactor feed pumps (RFP) stated that the trip of a reactor feed pump at CPPU was analyzed and is documented in attachment 6 to LCR H05-01. Figure 3.2-2 of Attachment 6 shows the reactor vessel level as a function of time for this transient.
Your response also stated that the THOR-BOP model is benchmarked against both plant data and the 110% CPPU thermal kit and that the THOR-BOP results were supplemented by empirical plant data to confirm the adequacy of the THOR-BOP conclusions.
a) Considering the benchmarking of your model and design analysis with actual plant data, explain what the level of uncertainty is and how it was determined for your dynamic analysis (including measurement uncertainties associated with the plant data that is used), particularly in predicting RFP and (secondary condensate pump) SCP suction pressures.
b) If not previously included in Figure 3.2-2 of Attachment 6 to LCR H05-01, resubmit Figure 3.2-2 to include the following additional conditions and information:
Attachment I LR-N07-0154 LCR H05-01, Rev. I
- 1) the effects of the bounding worst-case assumptions (e.g. amount of delay for Reactor Recirculation (RR) runback (loop-logic time delay), delay in RFP speed reduction, reactor vessel level at the start of the transient),
- 2) model and dynamic analysis uncertainties as determined in a) above,
- 3) the reactor pressure drop caused by the RR runback (originally neglected in Vermont Yankees' analysis)
- 4) a graph of RFP suction pressure versus time for the transient, and identify the time and value of the minimum suction pressure.
c) Explain how the results of this analysis as presented in b) above confirm that the loss of an RFP at CPPU power will not cause a loss of feed water and subsequent reactor scram, including a discussion of how much margin is available.
Response
Based on PSEG's discussion with NRC staff during a telephone conference call on June 7, 2007, this response is organized to emphasize design margins and present the results and data in tabular form in order to facilitate comparisons between a total of eight (8) calculations and plant events. Specific responses to the questions above are provided at the end of this response.
The conclusion that a trip of a single RFP at EPU conditions will not cause a loss of all feed-water or a reactor scram is based on two design analyses and evaluation of data from four actual RFP trips (2 single RFP trips and 2 dual RFP trips). A comparison of these, their critical parameters and design margin are summarized in Table 1.
Design Analyses:
GE TransientAnalysis, March 2005 This analysis performs an ELTR-1 plant specific evaluation of a single feed-water pump trip using NRC approved methods and models. This bounding analysis which is conservatively performed at 3952 MWt (120% power) shows that the scram condition (Level 3 - 12.5") is avoided with the lowest level at approximately 19" occurring 12 seconds into the event (margin of 6.5" to scram). The intermediate runback is successful at avoiding any subsequent plant trips. Other assumptions considered include use of an ANSI 5.1 decay heat curve, and consideration of actual speed controller gain and output in modeling the runback. Suction pressures for the remaining two pumps were not determined in this analysis.
PlantRunback Analysis, May 2007, Revision 0 This analysis includes a run of the THOR-BOP simulator model, which is benchmarked to CLTP, under EPU conditions at the target power level of 111.5%. This model is run Attachment I LR-N07-0154 LCR H05-01, Rev. I under nominal conditions to predict actual expected response. The lowest reactor level predicted by this model is 28", a margin of 15.5" to the scram condition. The highest level observed was 44", a margin of 10" to the turbine trip.
This model also demonstrates that the lowest predicted RFP suction pressures occur prior to the trip and that suction pressures increase following the trip of a RFP, thus there is no potential for additional trips on low RFP suction pressures. Since suction pressures go up due to the event, there is no technical need to consider uncertainties, nor would the use of more bounding assumptions be expected to alter this response.
Since the RFP high speed clamp remains the same (5480 rpm), the same level of suction pressure protection is provided as prior to EPU, should the remaining pumps speed up to compensate for the trip of the third pump.
Actual Pump Trips:
Review of plant operating experience at Hope Creek identified four occurrences where either a single or two RFPs tripped or were tripped during power operation. These included the following:
" Single RFP trip - Startup Testing, 'A' RFP was intentionally tripped from 97.4%
OLTP (3293 MWt) thermal power on 12/2/86. Level dropped to 29.8 inches almost avoiding a runback (i.e., the runback requires trip of a RFP and level lowering to 30").
- Single RFP trip - Spurious 'A' RFP Trip from 93% power which occurred on 11/23/02. A runback and scram was avoided by reducing power to 86%
manually.
- Dual RFP trip - A simultaneous trip of two RFPs occurred due to an electrical transient in May of 1993. A scram was avoided in part due to a rapid power reduction from the runback and trip of a reactor recirculation pump which resulted from the electrical transient. This dual pump trip is included in this discussion because it did not result in a total loss of all feed (i.e. the remaining RFP continued running).
- Dual RFP trip - A simultaneous trip of two RFPs resulted from an electrical transient on May 29, 2007. The reactor was manually scrammed due to decreasing level. The remaining RFP increased its speed to its speed-clamp maximum and continued to function through the transient.
In all cases, whether single pump trips or dual pump trips, no total loss of feed-water or consequential pump trips occurred, nor were the trip set-points of the remaining pumps (i.e., 230 psig w/10 second time-delay) challenged. Both single RFP trips and one of the dual RFP trips were successful in avoiding a reactor scram. Available RFP suction pressure trends associated with these trips indicate that the lowest suction pressure occurred prior to the trip (i.e., the decrease in suction pressure due to the remaining pumps speeding up is small in comparison to that gained initially due to the trip of the RFP itself). Refer to Table 1 for a summary of test and analysis results.
Attachment I LR-N07-0154 LCR H05-01, Rev. I Specific Question 7.16 Responses:
Figure 3.2-2 of LCR H05-01 Attachment 6 was generated as a result of the GE bounding design analysis performed and was not generated using the benchmarked THOR-BOP model. As discussed above, this prediction is developed using bounding assumptions (3952 MWt or 120% OLTP) and thus is a very conservative estimate of the minimum level resulting from a RFP trip at EPU conditions.
a) In addition to the bounding GE analysis performed to evaluate the RFP trip at 120% power, the benchmarked THOR-BOP model was run under nominal conditions to predict the actual response at EPU conditions. Because this model and actual plant experience (discussed above) shows that suction pressures improve for the remaining pump(s) following a trip, there was no need to consider additional uncertainties.
b) As discussed above, results are presented in tabular form (Table 1) in lieu of a graph for ease in comparing multiple analyses with actual events. Those analyses which are performed to bound the event and those performed with nominal assumptions used to predict actual responses are annotated in the Table. Reactor pressure drop is considered in both the GE and THOR-BOP models. Graphs of reactor level and RFP suction pressure versus time for a single RFP trip is provided as Figures 1 and 2. These graphs are provided based on the THOR-BOP model run at 115% power and the design maximum allowable combined condensate prefilter/ demineralizer differential pressure (Table 1; Calculation C).
c) Table 1 demonstrates that there is no potential for a total loss of feed-water due to a single RFP trip. Suction pressures for the remaining pumps improve post-trip and a scram condition is avoided.
LR-N07-0154 LCR H05-01, Rev. 1 Table 1 - Reactor Feed Pump Trips (Actual vs. Design)
PARAMETER Trip Calc Trip B Trip Trip Calc Calc Calc Min A A (Note 3) C D B C D Margin InitialRxPower 96 100 93 97 100 112 115 120 NA
(% CLTP)
InitialRxPress. 1015 1020 NA 1010 1018 1020 1020 1020 NA (psia)
Min. RPV Level 29.8 29.5 >30 16 (-)41 28 28 19 6.5 (Scram-12.5") 28Note 1 Max. Level <54 (Turbine trip- 38.5 43 43 35 44 44 40 10 54")
Lowest RFP Suction Sure See Noe not not not 115 available available avail Note 4 (Trip-230 psig 2 for 10 seconds)
Time of lowest See Prior to not not Prior to Prior to Prior not RFP Suction Note not to Note 4 Pressure 2 Trip available available Trip Trip Trip avail Final Rx Power 68 72 86 50 0 70 70 70 NA
(% of Initial)
Final985 983 NA 961 934 968 968 960 NA (psia) Note 5 Legend:
Trip A Single 'A' RFP trip from 97.4% reactor power on 12/2/86 (OLTP) performed during startup testing. Data obtained from original GETAR traces contained in TE-SU.AE-233(Q).
Trip B Single 'A' RFP trip which occurred on 11/23/02 from 93% reactor power.
Trip C Dual RFP trip from 100% reactor power in May 1993 due to an electrical transient. Note that the simultaneous tripping of a recirculation pump caused a rapid reduction in plant power avoiding a scram (provided for information, not a design transient).
Trip D Dual RFP trip from 100% reactor power in May 2007 due to an electrical transient. Reactor was manually scrammed due to decreasing level (provided for information, not a design transient).
Calc A THOR-BOP analysis of a RFP trip occurring at current power (Benchmark Case - representative of Trip A).
Calc B THOR-BOP analysis of the 'A' RFP trip occurring at the approximate EPU Target Power Level (111.5%).
The model is run using nominal conditions and predicts the actual post-EPU response.
Calc C THOR-BOP analysis of a RFP trip occurring at the EPU License Power Level (115%). The model is run using maximum design allowable highest d/p across the condensate prefilters/demineralizers and predicts the actual post-EPU response.
Calc D GE analysis of an RFP trip. This analysis is performed with worst-case bounding assumptions including power level at 3952 MWth (120% reactor power).
Notes:
- 1) In both the actual trips and calculations, initial RPV levels were at the normal water level maintained by the digital feed-water system of 35".
- 2) Suction pressures were not recorded during the original startup test. However, GETAR traces show that total feed water flow never exceeded its pre-trip value as the two remaining RFPs increased speed to restore level. Thus, since the RFPs share a common suction header, suction pressures would have increased from their pre-trip values following the trip.
- 3) Plant historian data was not available during or after this trip. However, a runback was avoided by reducing reactor power by operator action. Thus, level was maintained above 30 inches following the transient since the runback set-point was not reached (information obtained from the corrective action program).
- 4) Trip A and D, and Calc A, B and C demonstrate that RFP suction pressures increase following the trip.
Thus, the 230 psig w/10 second time delay trip set-point is not challenged. Margin at EPU conditions would Attachment I LR-N07-0154 LCR H05-01, Rev. 1 be a minimum of 115 psig. Since RFP suction pressures increase, uncertainties or bounding assumptions are not relevant to this analysis.
- 5) Pressure due to the scram lowered to 887 psig then increased and was controlled by the bypass valves (920 psig pressure set).
Attachment I LR-N07-0154 LCR H05-01, Rev. I 0.)
0I-,
0) 0 U
0) 0)
z CO 4)
- 0) 0 0
Co
)------
4------
0 0
0----
0.°
..- I- I -- - - -
0 .l M
C4
~~-- ---I- - -- ---- 0 0
0 0 0 0 o0 to 0 0 0 0 ino 0 0 M 0 0 0 . 0 0 0
(SaqOUI) I.IAO-1 Attachment I LR-N07-0154 LCR H05-01, Rev. I
- p
° ,,,,,i CL 0 0
0 LL 0 0 0
Cq I -_I
-- 2,- - - - - - -
0 0
!0 0
0 0 0 0 0! 0 0. 0 C 0 0 0 0D 0 0 0 0 0 0 0 0
'a 0 CD VISd Attachment I LR-N07-0154 LCR H05-01, Rev. 1 7.17 Your response to RAI 7.13 for the loss of an SCP stated that with a more realistic delay in RFP speed reduction and runback effectiveness, RFP suction pressure may momentarily drop below the 230 psig RFP trip setpoint but will recover well within the existing 10-second time delay.
a) For a trip of an SCP at CPPU, show the time dependent reactor vessel level and RFP suction pressure curves (identify the time and value of the minimum suction pressure) and compare them to the 230 psig RFP suction pressure trip and the 10 sec time delay, including in your analysis the following if not already considered:
- 1) the effects of the bounding worst-case assumptions (e.g. amount of delay for Reactor Recirculation (RR) runback (loop-logic time delay), delay in RFP speed reduction, reactor vessel level at the start of the transient),
- 2) model and dynamic analysis uncertainties (as referred to in 7.16),
- 3) the reactor pressure drop caused by the RR runback (originally neglected in Vermont Yankees' analysis) b) Do the results confirm that a trip of an SCP will not result in a loss of feedwater and subsequent reactor scram? Explain, include a discussion of how much margin is available.
Response
Similar to the response to Question 7.16, this response is organized to emphasize design margins and present the results and data in tabular form. Specific responses to the questions above are provided at the end of the response.
The conclusion that a trip of a single SCP at EPU conditions will not cause a loss of all feed-water or a reactor scram is based on the results of two design dynamic models and evaluation of data from an actual SCP trip which occurred on 6/22/02. A comparison of these, their critical parameters and design margin are summarized in Table 2.
Design Analyses:
PlantRunback Analysis, May 2007, Revision 0 This analysis includes runs of the THOR-BOP simulator model, which is benchmarked to CLTP, under various EPU conditions. Runs were made assuming both nominal conditions (target power level, 111.5%, and clean demineralizers) and bounding conditions at 115% (see Table 2). Bounding conditions included evaluation at the maximum allowable combined condensate prefilter/demineralizer differential pressures, and use of a 4-second delay in establishing the reactor recirculation pump runback (see below for discussion of this conservatism). This model was used due to its strengths in 9-
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 evaluating plant integrated effects including the reduction of reactor pressure during the runback.
This analysis also included runs of the transient module of PROTO-FLO to conservatively predict bounding lowest potential suction pressures during and following the trip, to ensure bulk boiling does not occur in the system, and to evaluate the existing suction pressure trip time-delays. For these analyses, no credit is taken for the runback, the highest procedurally allowable condensate pre-filter and demineralizer differential pressures are used, and conservative pump moment of inertias are used for pump coast-down.
Actual Pump Trip:
The plant experienced a trip of the 'B' SCP from 100% power on 6/22/02. No significant level change was seen immediately following this transient. Running RFP suction pressures decreased from approximately 460 psig to a low of 383 psig and then turned and increased due to the effect of the runback. A margin of 153 psig was maintained to the 230 psig RFP low suction pressure trip set-point during this event. It should be noted also that at the completion of the runback, the plant scrammed due to a turbine trip due to high moisture separator (MS) level caused by the decrease in reactor pressure. This was ultimately attributed to slow MS dump valve operation which was corrected via the corrective action process.
Specific Question 7.17 Responses:
a) As shown in Table 2, assuming little or no RR runback delay, the THOR-BOP model accurately predicts the RFP suction pressure response at CLTP.
Therefore, little to no delay is expected. Based on startup test data involving a RFP trip sampled using GETARS every millisecond, the time from the trip to an established core flow change is less than 2 seconds. Data taken from the SCP trip of 6/22/02 taken at a once/second sample rate showed that the time between the trip and RFP suction pressure recovery (indicative of RFP speed reducing due to the RR runback) was approximately 3.11 seconds; thus a 4-second time delay in the THOR-BOP model is chosen to conservatively bound any realistic delay. In addition, RFP runback response is expected to actually be faster under EPU conditions since RFP clamp set-down speeds which currently limit the RFP speed increase should a SCP trip occur will serve to immediately reduce RFP speed at the time of the trip since the new EPU RFP running speeds will be above these clamps. Using this 4-second bounding assumption, assuming condensate pre-filters and demineralizers are at their maximum allowable differential pressure, and starting with a normal water level of 35" and pressure of 1020 psia, the THOR-BOP model is run to predict a conservative expected response.
- 1) Using the bounding assumptions discussed above, the time-dependent CPPU vessel level and RFP suction pressure curves from THOR-BOP for a SCP trip are presented as Figures 3 and 4 (Table 2; Calculation D).
Attachment I LR-N07-0154 LCR H05-01, Rev. 1
- 2) In lieu of calculating the summation of all uncertainties from the non-safety related instruments used to benchmark the THOR-BOP model (information not readily available), the approach taken was to ensure each model accurately predicts plant results, and using multiple models and bounding assumptions, to conservatively predict worst-case results and if appropriate, take actions to improve margins if necessary. CLTP benchmarking uncertainty and instrument TLA (total loop allowance) is considered as shown in Note 2 to Table 2. When considering these uncertainties, there is a minimum margin of 18 psid to the 230 psig trip setpoint at design conditions. In addition to this pressure margin, there exists 10 seconds of time-delay to avoid additional pump trips.
- 3) Reactor pressure drop was considered in the THOR-BOP analysis. The time dependent pressure for the trip discussed in the response to Paragraph la, 1 above, is presented as Figure 5 (Table 2; Calculation D).
b) Table 2 demonstrates that no total loss of feed-water or reactor scram will occur due to a single SCP trip at EPU conditions. Margins are shown in Table 2.
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 Table 2- Secondary Pump Trip (Actual vs. Design)
PARAMETER Trip Calc Calc Calc Calc Cale Calc Min A A B C D E F Margin InitialRxPower 100 100 111.5 115 115 111.5 115 NA
(% CLTP)
InitialRx Press. 1019 1020 1020 1020 1020 1020 1020 NA (psia)
Min. Level 35 not Min. Note 1 34.8 33 32 31 not avail avi (Scram-12.5") Note 1 avail 18.5 Max. Level not urbine 40.3 38.5 41 43 44 not avail avi (Turbine trip-54") avail 10 Lowest RFP 330 258 260 190 100 Suction Pressure nominal (Trip-230 psig for 383 375 nominal 323 bounding no no 28 10 seconds) Note 2 case case runback runback bounding Time of lowest REP 3.1 sec 7 sec 6 sec 7 sec 7 sec 6-8 sec 6-8 sec NA Suction Pressure Note 3 Note 4 Final Rx FnlRPoe Power 72 72 72 not 74 72 72 not avail nt NA
(% of Initial) Note 1 avail Final Rx Press.
FialR Pes.not N 935 983 968 968 968 not avail avi NA (psia) avail Legend:
Trip A - Single 'B' SCP trip from 100% reactor power on 6/22/02. At the conclusion of the runback, the plant scrammed due to a turbine trip associated with high moisture separator level. The root cause of this event was slow MS dump valve response which was corrected.
Calc A - THOR-BOP analysis of a SCP trip occurring at current 100% power level (Benchmark Case - similar to 6/22/02 trip). No runback time-delay is assumed.
Calc B - THOR-BOP analysis of a SCP trip occurring at the target power level of 111.5%. This case is performed using nominal parameters to predict the actual suction pressure response.
Calc C - THOR-BOP analysis of a SCP trip occurring at the Licensed EPU Power Level (115%). This case is performed with the maximum design allowable highest d/p across the condensate prefilters/demineralizers.
Calc D - THOR-BOP analysis of a SCP trip (same as Calc C) occurring with a bounding 4-second RR runback delay at 115% and with the maximum design allowable highest d/p across the condensate prefilters/demineralizers. This is the bounding case for purposes of RFP suction pressure margin.
Calc E - PROTO-FLO analysis of a SCP trip at target power level of 111.5% (no runback assumed). This case is provided for information only since it was used to determine the magnitude and timing of the minimum RFP suction pressure in the absence of any RR runback.
Calc F - PROTO-FLO analysis of a SCP trip at the Licensed EPU power level of 115% (no runback assumed). This case is provided for information only since it was used to determine the magnitude and timing of the minimum RFP suction pressure in the absence of any RR runback.
Notes:
- 1) Conditions which existed after the runback but prior to the scram on high moisture separator level. Final pressure shown was post-scram with pressure being maintained by the bypass valves.
- 2) Lowest expected (bounding) EPU margin assuming a bounding 4-second RR runback delay (see discussion later). PROTO-FLO results (Cases E & F) assume runback does not occur. Existing 10-second trip time-delay provides margin should the delay be slightly longer than 4 seconds. The margin shown is conservative since the CLTP benchmarked model under predicts suction pressure by 8 psid. When considering instrument total loop accuracy, the bounding margin would be reduced to 18 psid (instrument TLA is 9.6 psig). The nominal margin (from TPU power) is expected to be 100 psid.
- 3) Exact time of trip not available on plant historian. Duration shown is from first noticeable parameter change to lowest RFP suction pressure; therefore, pump coast-down may not be included.
- 4) A second low pressure was reached at approximately 75 seconds into the event due to a slight speeding up the RFPs to restore level.
Attachment I LR-N07-0154 LCR H05-01. II Rev. I I-
4-- -------------- -
a) -i---I------j-j- -jijij--------------
z
---------~I---------I- -- ---
A)
-- -- -- V-K-------
Ln N0 0ý 0i L
-3I 0.
0 U - 0 C'4 U
-- - - - - - - - - - - - - - - - - - - - - - - - - - - - - -_ - - - -- U LI 0 0 U,
0 0
0 IL, 0
o 0 0D 0 0) 0 0 0 ) 0 0)
LO o 0 0> 0t00 2 0 0) 0 M) C1 N -
'NI Attachment I LR-N07-0154 LCR H05-01, Rev. 1 In ci.)
A-0 0
LL ci) fn I
a----
0
U---------
0
---- ----- ---- j-il-- - - - - - - - - - - 0 CD
J -------------------------
M LO 0
o CD 0 0 0* 0 0* M 0 0 0 0 C.
0 0 0'
ID 0 In CC)
ViSd Attachment I - LR-N07-0154 LCR Hg-131- Rpy- I 0
- =)
a)
CM 0
LO v
Cu
- ------- ------ C3 LO M
U, I
w LLI C>
oN 0* I-0*
0r CD CD 0 0 0 0D 0 R CD C.-
04 0 0 0 0z q)
C., 0r- 0, a3) CO 0I) Cn 0 0 VISd Attachment I LR-N07-0154 LCR H05-01, Rev. I 7.18 Your response to RAI 7.13 for primary condensate pumps (PCP) stated that with a 4-second delay imposed on RR runback, SCP suction pressure dropped to 39 psig (only 9 psi above the 30 psig trip setpoint) following a PCP trip. Therefore, the SCP trip setpoint time delay will be increased to less than or equal to 15 seconds for CPPU.
a) For a trip of a PCP at CPPU, show the time dependent reactor vessel level and the RFP suction pressure and SCP suction pressure curves (identify the time and value of the minimum suction pressure) and compare them to their respective pump suction pressure trip setpoints and time delays. Include in your analysis the following if not already considered:
- 1) the effects of the bounding worst-case assumptions (e.g. amount of delay for Reactor Recirculation (RR) runback (loop-logic time delay), delay in RFP speed reduction, reactor vessel level at the start of the transient),
- 2) model and dynamic analysis uncertainties (as referred to in 7.16),
- 3) the reactor pressure drop caused by the RR runback (originally neglected in Vermont Yankees' analysis) b) Do the results confirm that a trip of a PCP will not result in a loss of feedwater and subsequent reactor scram? Explain, include a discussion of how much margin is available.
c) Based on the analysis that was completed, including the cases described in 7.16 and 7.17 above, explain what the minimum allowable SCP suction pressure trip delay is for CPPU, including the basis for this determination.
Response
Similar to the response to Question 7.16, this response is organized to emphasize design margins and present the results and data in tabular form. Specific responses to the questions above are provided at the end of the response.
The conclusion that a trip of a single PCP at EPU conditions will not cause a loss of all feed-water or a reactor scram is based on the results of two design transient models and evaluation of data from an actual PCP trip which occurred on 2/04/06. A comparison of these, their critical parameters and design margin are summarized in Table 3.
Attachment I LR-N07-0154 LCR H05-01, Rev. I Design Analyses:
PlantRunback Analysis, May 2007, Revision 0 This analysis includes runs of the THOR-BOP simulator model, which is benchmarked to CLTP, under various EPU conditions. Runs were made assuming both nominal conditions (target power level, 111.5%, and clean demineralizers) and bounding conditions at 115% (see Table 3). Bounding conditions included evaluation at the maximum allowable combined condensate prefilter/demineralizer differential pressures, and use of a 4-second delay in establishing the reactor recirculation pump runback (see below for discussion of this conservatism). This model was used due to its strengths in evaluating plant integrated effects including the reduction of reactor pressure during the runback.
This analysis also included runs of the transient module of PROTO-FLO to conservatively predict bounding lowest potential suction pressures during and following the trip, to ensure bulk boiling does not occur in the system, and to evaluate the existing suction pressure trip time delays. For these analyses, no credit is taken for the runback, the highest procedurally allowable condensate pre-filter and demineralizer differential pressures are used, and conservative pump moment of inertias are used for pump coast-down.
Actual Pump Trip:
The plant experienced a trip of the 'A' PCP from 100% power on 2/04/06. Reactor level response was in the upward direction post-trip due to the runback. Running SCP suction pressures decreased from approximately 123 psig to a low of 95 psig and then turned and increased due to the effect of the runback. A margin of 65 psig was maintained to the 30 psig SCP low suction pressure trip set-point during this event.
Specific Question 7.18 Responses:
a) As shown in Table 3, assuming little or no RR runback delay, the THOR-BOP model accurately predicts the SCP suction pressure response at CLTP.
Therefore, little to no delay is expected. Based on startup test data involving a RFP trip sampled using GETARS every millisecond, the time from the trip (runback logic) to an established core flow change is less than 2 seconds. Data taken from the PCP trip of 2/04/06 taken at a once/second sample rate showed that the time between the trip and SCP suction pressure recovery (indicative of RFP speed reducing due to the RR runback) was 2.94 seconds; thus a 4-second time delay in the THOR-BOP model is chosen to conservatively bound any realistic delay. In addition, RFP runback response is expected to actually be faster under EPU conditions since RFP clamp set-down speeds which currently limit the RFP speed increase should a PCP trip occur will serve to immediately reduce RFP speed at the time of the trip since the new EPU RFP running speeds will be above these clamps. Using this 4-second bounding assumption, assuming condensate pre-filters and demineralizers are at their design maximum allowable differential pressure, and starting with a normal water level of 35" and Attachment I LR-N07-0154 LCR H05-01, Rev. 1 pressure of 1020 psia, the THOR-BOP model is run to predict a conservative expected response.
- 1) Using the bounding assumptions discussed above, the time-dependent CPPU vessel level and SCP suction pressure curves from THOR-BOP for a PCP trip are presented as Figures 6 and 7 (Table 3, Calculation D).
RFP suction pressure is presented as Figure 8 (Table 3, Calculation D).
- 2) In lieu of calculating the summation of all uncertainties from the non-safety related instruments used to benchmark the THOR-BOP model (information not readily available), the approach taken was to ensure each model accurately predicts plant results, and using multiple models and bounding assumptions, to conservatively predict worst-case results and if appropriate, take actions to improve margins if necessary. CLTP benchmarking uncertainty and instrument TLA is considered as shown in Note 2 to Table 3. When considering these uncertainties, there is a minimum margin of 4 psig to the 30 psig trip set point at design conditions.
In addition to this pressure margin, there will be a minimum of 10 seconds of time-delay to avoid additional pump trips (the time-delay set points are being changed from 1-second to between 10-15 seconds in a staggered manner). By staggering the three set points, defense-in-depth is provided ensuring a single PCP trip cannot result in a total loss of feed water.
- 3) Reactor pressure drop was considered in the THOR-BOP analysis. The time dependent pressure for the trip discussed in the response to Paragraph la, 1 above, is presented as Figure 9 (Table 3, Calculation D).
b) Table 3 demonstrates that no total loss of feed-water or reactor scram will occur due to a single PCP trip at EPU conditions. Margin is discussed above and presented in Table 3.
c) The minimum allowable SCP suction pressure trip delay for CPPU has been chosen to be 10 seconds. As discussed in the response to la above, a 4-second RR runback delay bounds any previous runbacks experienced and under EPU conditions, this response, although not quantified, would be faster. As shown in Table 3, Calculation D, the trip set point is not expected to be reached under EPU design conditions; therefore, 10-seconds of time-delay will exist as margin.
However due to the low pressure margin, the results of the PROTO-FLO runs which assume no runback occur, and the possibility although not expected for delays beyond 4 seconds, the time-delays are being extended from their current 1-second. By providing a staggered time-delay over the range of 10-15 seconds additional defense-in-depth is provided to ensure a single pump trip does not result in the loss of all feed water. These time-delays are also chosen to be consistent with those already in place for the RFPs and are acceptable per discussions with the pump vendor.
18-
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 Table 3 - PrimaryPump Trip (Actual vs. Design)
Trip Calc Calc Calc Calc Calc Calc Min PARAMETER A A B C D E F Margin InitialRxPower 100 100 111.5 115 115 111.5 115 NA
(% CLTP)
Initial Rx Press. 1019 1020 1020 1020 1020 1020 1020 NA (psia)
Min.
Min.Level 35 35 35 35 35 not avail not avail 22.5 (Scram-1 2.5")1 Max. Level 40 41.2 40 42 42 not avail not avail 12 (Turbine trip-54")
Lowest SCP 81 39 9 3 51 Suction Pressure 95 97 nominal 59 bounding no no [nominal]
(Trip-30 psig for 1 9al 59 b d no 9 second) Note 2 case case runback runback [bounding]
Time of lowest SCP Suction 3.0 sec 5 sec 8 sec 8 sec 8 sec 2-3 sec 2-3 sec NA Pressure Final Rx Power 58% 59% 72 70 71 not avail not avail NA
(% of Initial) Note 1 Final Rx Press. NA 971 968 968 968 not avail not avail NA (psia)
Legend:
Trip A - Single 'A' PCP trip from 100% reactor power on 2/04/06 due to electrical fault.
Calc A - THOR-BOP analysis of a PCP trip occurring at current 100% power level (Benchmark Case - similar to 2/04/06 trip).
Calc B - THOR-BOP analysis of a PCP trip occurring at the target power level of 111.5%. This case is performed using nominal parameters to predict the actual suction pressure response.
Calc C - THOR-BOP analysis of a PCP trip occurring at the Licensed EPU Power Level (115%). This case is performed with the maximum design allowable highest d/p across the condensate prefilters/demineralizers.
Calc D - THOR-BOP analysis of a PCP trip (same as Calc C) occurring with a bounding 4-second RR runback delay at 115%. This is the bounding case for purposes of SCP suction pressure margin.
Calc E- PROTOFLO analysis of a PCP trip at target power level of 111.5% (no runback assumed). This conservative case is provided for information only since it was used to determine the magnitude and timing of minimum SCP suction pressures in the absence of any RR runback.
Calc F - PROTOFLO analysis of a PCP trip at the Licensed EPU power level of 115% (no runback assumed). This case is provided for information only since it was used to determine the magnitude and timing of minimum SCP suction pressure in the absence of any RR runback.
Attachment I LR-N07-0154 LCR H05-01, Rev. I Notes:
- 1) Power lowered to 58% following the full runback. Rods were inserted to 53% power to get out of the OPRM exit region. Calculation data (Calc B-F) reflects the change of the full to the intermediate runback for a PCP trip.
- 2) Lowest expected EPU margin assuming a bounding 4-second RR runback delay. Considering that the CLTP benchmarked model under-predicts the pressure drop by 2 psid and the trip instrument has a total loop accuracy of 3.3 psig, the margin could be as low as 4 psid. Considering this low (bounding) margin, and since the conservative PROTO-FLO results indicate the potential for downward pressure spikes to be below the trip set point quickly (2-3 seconds), the existing SCP suction pressure trip time delays are being increased and staggered between 10-15 seconds to provide additional margin beyond the more realistic 9 psig bounding margin stated. The expected nominal margin (from TPU power) is expected to be 51 psig.
Attachment I LR-N07-0154 LCR H05-01, Rev. I I
-Series 1 - NR Reactor level 45.00 -~-n~-~ -- r--r rT-I ll t i1 l 1 J I l l I I t 1 I I i i i i 40.00 i 2 1 21 )II 2 I 2I I 2i l I i i I I i i i i i i ii2 4 ,I 0 2,
, 2 2 2 2 2, t I I,1 2 I I 2 2 I I I I I I I II2 I I 2I I I l2 2l I 2I 2I I I I 30.00 222 _ _ _ _ __ _ _ _- _ _ _ _
2 2 2 2 2 I 2 I 2I I2I2-- 2 2 2 I 2 2 2 2 i i i 2 2II 22 i I I 2i 2II i i i i i i 2 2 2 2 2 I I I2 I2 2 2 2 i i t i 2i2i i i 1 1 I 21 I I 2II 2I 2 2 2 I I I I2 I II I2 I 2I I l I I I l l 2l 2 2 I i i i 2i i i i 2
i 2
i 2
i 2i 2
I J 2i 2
I ~ i i 2l l 2
~ 2 I I2 Lo Level I i (Leve 2i i
23)1 Tip 2i 1
2l 1
I I
J i
ii i
l l i
I i
I i
2J I 2I I
JI I
i I
ll I
jIF 25.00 . 21 1 2 I 2 I I 2I 2 I I I I2 I21 I I I I I I I I I I I I 2 2I I I I I 2 I2 I I I I I I I I l i l I 2I 2 2 I I 2l 2 2 2 2 2 2 I 1 2 2I l2l 211 2 2 I 2I 2I2 2 I 2I 2I II2I I I I I I I I I I 2I 2 I I I2
. . . . . . . . . . . . . . . . . 2. ' ' '2, 2. . . 2 2 l i l 2t 2I l 2 2I I I I I l I I I I I I 2I I 2 I I 2 I I I I2 2I I 2II 2 2 I I I I I I I I I I I I I I I 7;
35.00 . . . . . . .. .. .. .
- il I I ILo Levl (eve 3) ITri 2 2 2 2 ' l l t I 1 J I 1 1 1 1 t 1 0.00 .i,. l , 2 2f I 2 21 I-2 2 2 I 2 2 l 2l 2 2 2 2 2 2 t l l 2 2l 2 2i 2t t 1 2 2 1 2t 2 t i l t 2I t 2I l I 2I I I 2t t 2l t i l l2 I I I I2 I i i i I I I I 2I 2 1 2 2II I 21 I I I 2I I I 2 2 2 J i l2l2 2 212112 1 2 2I2 2 I 2I 2 I I I i i i i2 I I I2 I I I2 2I 2I I l 2I I I 2t I 2I I I I I I I I 2I2 I I 2I 2l2 I2 I2l 2 12I I 2I 2I r 2 2 1 2 2 I I Il2 I2I I I I2 I I I2I2I2 I I l 2 21 I I 2I I 2 1 12 2 I2 2 2 2I 2I I t l l 2l i l 2i l 2I I 2I I I I I I I l2 2 2 2l 2l l 1 1 2i 2lI 2I 2 I I I i 21 2-02 2 2 2l ' l2 I I I 2' 2i 24+ I I 2I I 2 I2 2II I I I I 2 2I 2I 2 2I I I I 2I I I I 2 2 I I I 2 2 t2 l i l 2t2 2 2 2 2 2 l22 2 2 2 l 2 i 2 l 2 21* 2 2 2 i i i i 2I 2I 2I 2 2I 1 1 2 21 21 2 21 1 1 i i 211 2 i i i i I I I I 2I 2i 2i I 2I i i 2 2i 2i 2 2I2I I I I I 2 2 I2 I I I
i i I
i I2 i
2I2I I2I 2
2 2
2 2
2 2I 2 2 2 I I I
2t I
2l I
2 2I 2I i
2I 2 i
I2 l2 i 2I i I I
~
2I I
2I 2I i
i i
i 2I 2 2 2l 2 t2 2 2I 2l 2 l I 2lI i i i 2 J i l l2 I I I I2 05.00.
I 2 I I I I 2I I I l l2 2 2 l2 l I 2 l2 L
2 1 2 f
1 l
I2 2
' 1 2l
' ~l 2 l 2
_______...______,___I l
0 s0 100 150 200 250 300 350 401 TIME (SEC)
Figure 6 Attachment I LR-N07-0154 LCR H05-01, Rev. I
- Series 1 - SCP suction press 140.00 : I I I 2'~2 2~~~
. I 2 2 2 I 2 2 I 2 2 2 2 2 I 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 I
- 2 I 2 2 2222, 222; "'2 2 2 2 2 2 2 - I 2 2 2 I 2 2 2 I 2 2 2 2 2 I 2 2 2 2 I I 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2. 2 I 2 2 I 2 2 2 2 2 2 2 2 2 2 2 2i I 1I I 1 l l i i i I l l l 121 1 i 2
-l 2 i2 i i I i 1 l I I 2 t 2Ii l _ _f f__I _ _ _ _ ( t' l i t t 2i 2 2 l I I I 2l 2 I 2i 1 21 1 I I 2 2 2ii 2 2 I l2l2 I2 l I I I 2II I 2I I Fl l2 1 2} I I I I l2II 2I 2I 2 - 2 2 I I I I 2 I 2I I 2 2I l 2 I 2 I 2 I I I 2 2 2I I I I 2 I t i l I 2 2 2 2t 2 rl i I I I I I t i ll2 I 21 2 00.0 I
2II 2I 2 ~ 2 2 2iII i21 i2l i i 2 2 i I 2 2l l 2
2 2
1 2
-I 2
2 l
2i 2
l2.
2 2 2I 2l 2
2l 2
I 2l 2
I I t
1 22 TII Fi i
21 l
l 2I l
I I
2 2I TT 2
iT
~tI -
rT I
I2 2 l-- 2 1 1
1 2
2 1 2 2 1 1 2 2 2. 2 I2i2 2l 2 *l i l 2 2I I I 2 2 2 2l2l2I I 2I I I I I I I 2I2I2 2 I 22 I I2I 2 I I I 2 2I2 2 I I I I 2 2 2 2I I 2I I I l 2l 2 I 2 2 2 2I I I I 2 2 2l2I 2I 2 2 2 I I2 2l I 2 I 2I 2 I
, 2 2 , 2 2 1 1 2 2 I I2 2 I 2 2I2I2 2 2 t i l 2 l 2l 2 2 2 20* 2 2 2 2 i 2 l 22 j , 4 211 2 2'2 2o i 2 2 2 2 L I I I II222 2 2 2 l 1 2[ 2 2i 2Il i i i i ii2 2t2 I 2l2 I I I 2I2 I2 2 2 2 I 2II 2t 2I I 2 i l t2 2 2 2l 2 I I I I2 r2 f l l I I 2I 1 1 1 2 2 l2 0<
222 .2 2 .2 2I I
2 . . . . . . . . . . . . . . . . . . . . . . . . . . .2 2I2 2t2 2
2 I
I I
I2 2 I2 I l
I I l
I l
I2 II l
I I l t I
l I I l
I l
I l 2 I 2 2 2 I
l2 2I It LowS resr Trip (0) sz l I *l,:I I_____l__I I _
2 2 2 2 2 21 I I 2I I I I l2 2 2I2 i i i 2I 2I2 I2I 1 2 2I I 2 I I2I 2t 2t i l l2 I2 21 2t 2I 2t 2
20 0 2 2,
, t
.. . .. .. 2 2
.. 2 2 2.
2* 2 !
. . 2. .
. . 2. .
2 2 2 2 i l 2 2l 2J 2*J 2 2l2 2 2 2 2 2 I I I I l i2tI i i i2I2 l f i 2i l l I I I 2 2 I I i l l I I 21I I 1 2II 12 I2i2t 1 1 1 20.0 2 2' ' Low~2 Sucton ressre -i (30pig) T " . . . . . . . -
2 2 I 2I I 2 2 2 2 2 2 I f i l i i i I t2
- 40. i 0 l 2 2 2 2 2 2 I t I I2 I 2I 2 2 1 2 2 l2 l 2 f i l 2l2 2 2 2 ` 1 1 I I 2 2I l l l 2 2 l l l l 2l 2 l2 2 2 2l2 2 l l 2 2 l2l 2l I I I 2 1 1 12 i -
2I I i I i i 2i2 2i i i 2 2 2 l I2i l 2 l 2l 2 i i i2 2 2 I 2J 2 2 2 2J f l 2ll I 2l2 i l2 r I I i 2 I 1 2 1 1 I I I 2I2 2I I 1 2 2 2 I2I I2 2ii I I I 2 I2 I2 2 2 21 12 2 2 I2 I2 I 2I22 l I I I i i 2 I 2I 2 1 1 12 2 2 2 2 2 t i 2i l 2 2 2l l2 l 2 I 2i 2I I I I2 i
2t l t 2l i 2 2l 2I i 2I 2 I I I 2l I 2I 21 I l 0.0 I _____I __1 1_ ___ l______ I II I I I I _ _11 I I I I I I I_____I__III I__I I .I I I - 1. ___ 1 1 1 1__
0 so 100 150 200 250 300 350 40(
TIME (SEC)
Figure 7
- 22 -
Attachment I LR-N07-0154 LCR H05-01, Rev. 1
-Series 1 - RFP suction press 600.00 I t i I m
illi J i i r t IIII 2 2 2 2 2 2 r I 2 2 2 2 2 2 2 2 2 2 I r 2 2 I 2 2 2 2 I 2 2 2 2 2 2 2 I 2 2 2 2 2 2 I 2 2 2 2 2 2 2 2 2 2 2 I I 2 2 2 2 2 2 2 I 2 2 2 2 I . 2 2 2 2 2 2 2 2 2 I 2 I I I 2 2 2 I 2 2 2 2 r 2 2 2 I 2 1 .1 2 2 I 2 I 2 I I 2 2 2 2 2 I 2 2 2 2 2 2 2 I 2 2 2 2 2 2 2 I I 2 I I I 2 r 2 2 2 2 2 2 I 2 2 2 2 I I 2 I I t 2 2 2 2 2 2 I I 2 2 I 2 2 2 2 I 2 2 2 I 2 2 I 2 2 2 2 2 I 2 .2 2 2 2 2 2 2 2 2 2 I 2 2 2 2 2 I 2 2 2 2 2 I 2 2 I I 2 I 2 2 2 2 2 2 2 2 2 2 2 I 2 2 2 I 2 2 I 2 2 2 2 I 2 2 2 2 I 2 2 2 2 2 2 2 2 2 2 2 2 2 I 2 2 I 2 I 2 I 2 2 2 2 2 2 2 2 2 I 2 2 I 2 2 2 2 2 2 2 I 2 2 2 2 2 2 2 I 2 2 I 2 2 2 I 2 2 2 I 2 2 I 2 2 2 2 I 2 2 2 2 2 2 2 2 2 2 2 2 I 2 2 2 2 2 2 2 2 I 2 I 2 2 2 2 2 2 2 2 2 2 2 I 2 2 2 2 2 2 2 2 2 2 2 2 I 2 2 2 2 4 I 2 2 400.00- 2 2 2 ______________
2 2 I I -
2 2 2 2 2 2 I 2 I I 2
1--r--r- F -r r-- -F-' r-r-ri--- T
< 300.00 i i
i i
i L F .1--
Low Suction Pressure Trip (230 psig)
I r I 2 i i 2 i i 2l i l 2 I I I I2 I I I2 i
- - + . . F. .
100.00 .4------------------------------+ .
0.00 +----------------------------------------------..2...-..................~ 401 50 100 150 200 250 300 350 0
TIME (SEC)
Figure 8 Attachment I LR-N07-0154 LCR H05-01, Rev. 1
-Series 1 - Reactor Wide Range Press 1030.00 I 2 I 2 I 2 2 I 2 2 2 2 I 1 1t 2 .0 i I I I lI - I l i2Il 2 I l l I I I I I I2 I 2l2 I I 2 I L I I I I I I I r 2 2 I I I I I I l I 2 I 2 I I II I 2 2 IIII 12 I I 21 i i l l I I I i i i iII I i l IIIl I Il2 ii Ii IlI i I 2 t l Il i i i 2 i r l 2 l I I I I2 2I 2 I2 i I I I I 2 2 i i 2 i2 i2 1 I 1 I 2- I 2 i2 2 l l l I I I I I I I I 1 1 1 1 l I l2 l2 l J i I I i2 2 i i I I I I l 2l I I I I I I I I I I2 I I 2 II 1 1 1 1 12 I I I2 I 2 t i l l2 l l 2 i l l 2l l l I 2 I I ] l l II I l lI 2 II l 1 1 21 1 2 I I I I I 2 I I l f i l I I 2 I I22 I I I f i l f i l 2 f i l 1010.00 2I2l 2l I I
I I l I2 2
I I I l
I I l ~
II l
I I I I
II 2I I f Ir l
I 2 I I i
I I l
I f
I l l
2 I I 2 2 2 2 I I) I I 1 l i l t2 I I2 2 I 2 2 l Ii f i l lI 2l f i ll222 I III Il I I l 2f i l l2 f i l I I2 I f i l lI 2 I l 2 I 2 l2 I I I I l f i l l2 I I I2 I I I I2 I 2 I I I I 2 i I l I I I I I 2 I i i l I I Il 2 2 I 2 f2 i I I l lI 2 2 I 2 I f i I l II l 2 I 2 2 2 2 22I I I JI l2 l2It 2 2 l 2 l 2 I I 2) J i 2 1 2 I I I212 1 l l I I I l I l L 2l l 2l l r I l l l f2 i l l I I I 2 2 2 2 I 2 2 I I2 2 I I*. I I I I I 2 2 i 2 I I 2 2 I2 2 I2 2 22I 1 1 I l l l2 I 2 I I I I I I 2 2 2 f2 i l l l lI 2 I I2 I I I I I I I I I I I 2 I I 2 I I I I I 2 l I I 2 I I 2 I 2 2 I2I 2 2 II 2
lIl 2 III l l l2 I I l l l lI I I I I 2 2 I I I 2 2 t i t i 2 I l I 2 2 2II I 22I 2I 2 I 2 I II i 2 2 2 t 2I I 2 I.2 I I I I 2 I2li 2 2I l t i2I l I I I I I I III22 2 2 I I I I I2 2 I. I 2 l2 i l l I I I I I I .2 I I2 2I 2 2 I 2I l l I I 2 I I 2 i i I II22 I I I I I2 I 2 I2
- i l 2I I l i i I I I I 1 2 l l U)
CL 2 i 2 2 I2 I2 2 I2 l2 i22l i l i l2l i i i2 J i l l i l 2 l 2 2 2 . i2i,=.,
2 2 2 2 2 1 I 1 2 1 I 2 I*
+ 1 2 l I I 2 I I II2 2 I 2 I I I I 2 I I I I I I I II 2 2 I I2 I I I I 2I 2 I 2i l l i l i l l2 I I I I 2 2 i l i l2" l i22 l i l I I 2 I I I I 2 I 2I 2 I I 2 II 2 I 2 2 I 2 IIII I i i i 2 I l 2 l I I I l I I I I 1 1 12 I I I I l2 l2 I 2 2 2fi I Il l 2 i I 2 I It' 2 f i l l I , I 2 I 2l i l l I 2 I I I 2 2 I 2 II 1 I 2 2 2 I I I I l2 I l I 2 l l 2 i~ l I II2 2 I I I I 2I I 2 I 2 I 2I II I I I 2 2I 2I i l J i l 2 l l I II2 2 I I 2 l i 2 2 I I 2 I 2 2 l Ii I .2 II 2 I 2 2 I I I I I I2 o2Io2I 2 I 2 I I
I I I 2
I l
"1I IIII1 lI I I I I I
I I I2 Il l I lI 2 2l 2I I I I 2 I
I f i I I l
2 lI JII I
i I
l I I2 I
I l
I i 2
- I2 I 2 2 I I I I I I I II II I I l II 2 1 1 1I I I I I I I I 9* .I 0I I 2 2 I 2I2I1 1 i l l I I I I 2I i iIIIi i I I I I2I '
i I I 2 2 I 2 22l I I i i i l 2l I I I 2 i 2I l I i i I I 2 2 I 2I2II II i I2ii Il 2I l J i l I I 2 I i IJ l l II I
. . .. 2. I . . 2
. I ' I 2I . . . . . . . . . . . . . 2 2 . . . . .
l2 l l I I I2 I 2 I I I I 2 2 2 III I2 I I 2 I I I I 2 I I I 2 2 2 I I 2 I I 2l i 2 2 " Ii l2 2 Il 2 2l l i l2 i 2l l I 2 2 I l l i 2 2 2 I I I i Ii2I 2 2 2 2! I 2 I I I 21 I I I IIII IIII IIII I l l l2 IIII U i I I * - I I I 2 2 2 2 I I 2 I I 2 2 I I I I I I 2 900.00-50 100 150 200 250 300 350 40 TIME (SEC)
Figure 9
-24 -
Attachment I LR-N07-0154 LCR H05-01, Rev. 1
- 11) Health Physics Branch (IHPB) 11.10 In response to RAI 11.7, explain the basis for the dose rates (Column A and B),
mission times (first column), and the resultant dose for each vital mission function (Column C) given in Table 11.7.a-1 of your response. Clarify how the mission times are calculated for each required vital area to include time to perform each activity and access to and egress from a vital area.
Response
Mission Times (first column): The mission times to perform each of the vital functions are extracted from the Hope Creek original plant-licensing basis UFSAR Table 12.3-3, Rev 0. The last column of Table 11.7.a-1 shows the reference drawings that illustrate the access and egress routes between the control room and the vital function locations (please see the note underneath Table 11.7.a-1). The access and egress routes shown in the reference drawings are not affected by EPU implementation.
Dose Rates (Columns A and B): The Hope Creek operating license has been amended for the Alternate Source Term (AST) on October 3, 2001 via License Amendment No. 134, that replaced the whole body and thyroid dose criteria by TEDE (total effective dose equivalent) criteria, therefore, the whole body and TEDE dose rates in Column A and B are calculated using the AST and 20%
uprated core inventory.
Resultant Dose for Each Vital Mission Function (Column C): Column C provides the whole body dose to perform the given vital function in the estimated mission time. The Column C whole body dose is the product of the post-EPU whole body dose rate in Column A and the mission time in the first column. Compliance to NUREG-0737, Section ll.B.2, assures the shielding adequacy necessary to reduce the whole body dose (external dose) to an operator performing a vital function in a given mission time to less than the allowable dose limit of 5 rem.
The whole body doses calculated to demonstrate compliance with NUREG-0737, Section ll.B.2 do not include inhaled dose. Plant procedures control radiological respirator use to protect against airborne radiological hazards.
a) Include in your discussion what source term assumptions were used, any factors that were used, an explanation of any conservative assumptions used, for what power level these calculations are based on, (include in your discussion where these radiation sources (i.e., RCS piping) are coming from for each vital area).
Attachment I LR-N07-0154 LCR H05-01, Rev. 1
Response
The following assumptions were used:
- 1. The whole body dose results from the airborne activities from the post-LOCA containment, ESF, and MSIV leakages.
- 2. Core inventory based on assumed thermal power level of 4031 MWt (120% of original licensed thermal power times 1.02 uncertainty).
- 3. The use of the atmospheric dispersion factors (X/Qs) for the various post-accident vital function location specific air intakes.
- 4. The direct dose contribution from the containment shine was negligible based on the presence of the concrete shielding of the containment wall and dome, and the concrete roof, floor, and wall shielding of the building where the various post-accident vital functions are to be performed.
- 5. The remaining assumptions were as stated in Regulatory Guide 1.183, Appendix A for the LOCA analysis.
The Hope Creek Emergency Core Cooling System (ECCS) and components that contain the post-LOCA suppression pool water are located in the reactor building. All areas to perform the post-accident vital functions are located outside the reactor building. Therefore, there is no unshielded piping carrying the post-LOCA fluid in proximity of the ingress/egress routes and areas where the post-accident vital functions are to be performed.
b) Include in your discussion a detailed explanation of any unshielded systems carrying reactor water in any of the vital areas and if there are any unshielded systems carrying reactor water that the operators will have to come into proximity for access and egress to each vital area.
Response
Please see response RAI 11.10.a, in second paragraph.
c) Include in your discussion an explanation of why the mission time for the filtration, recirculation, and ventilation system (FRVS) radiation monitoring system (RMS) skid changed from 0.9 hrs identified in the Hope Creek Power Uprate Safety Analysis Report (PUSAR) Table 8-1, to 1.5 hrs.
Response
The mission time of 0.9 hrs for the FRVS skid in the PUSAR Table 8-1 was calculated based on the TEDE dose rate that includes the inhaled dose rate. The mission time of 1.5 hrs in Table 11.7.a-1 is the estimated mission time abstracted from the Hope Creek original plant-licensing basis UFSAR Table 12.3-3, Rev 0, and included in the first column of Table Attachment I LR-N07-0154 LCR H05-01, Rev. 1 11.7.a-1. The estimated mission time of 1.5 hrs is the sum of the UFSAR, Rev. 0 specified times required for taking, transporting, and analyzing the sample.
11.11 Describe in detail the basis for calculations used for the percent increase in N-16 activity in the steam turbine equipment and condenser components as a result of this EPU (increase in N-16 concentration in reactor vessel, change of N-16 concentration in steam piping based on steam mass-flow-rate, decrease in decay time of N-16 concentration due to transit time, and increase in N-16 concentration at the steam turbine equipment and condenser components).
Response
The N-16 transit times for the steam components in the turbine building (TB) complex were analyzed in various design calculations for the steam mass flow rate corresponding to the original licensed thermal power level of 3,293 MWt.
Instead of revising a large number of design calculations, a simplified approach was adopted based on the change in the parameters that determine the N-16 transit time. For the given main steam piping configuration (i.e., pipe diameter, length, fittings including the elbows, tees, valves, and reactor dome pressure, all of which are assumed to remain constant), the N-16 transit time is dependent on the steam mass flow rate and the downstream pressure which is assumed to be the pressure at the turbine stop valve (TSV). The post-EPU reactor pressure vessel (RPV) dome pressure and TSV pressure are 1,020 psia and 937, psia respectively. The following two cases were analyzed to determine the decrease in the N-16 transit time:
Case 1: Determine the N-1 6 transit time for the original licensing basis:
Two N-16 transit times were determined between the RPV nozzle and TSV using the steam mass flow rate corresponding the original licensed core thermal power level of 3,293 MWt. The pressure between the RPV nozzle and TSV was assumed constant at either 1,020 psia or 937 psia, and two different transit times were determined for each constant pressure value.
Case 2: Determine the N-16 transit time for EPU licensing basis:
Two N-16 transit times were determined between the RPV nozzle and TSV using the steam mass flow rate corresponding to the post-EPU core thermal power level of 3,952 MWt. The pressure between the RPV nozzle and TSV was assumed constant at either 1,020 psia or 937 psia, and two different transit times were determined for each constant pressure value.
From the N-16 transit times for the original and EPU design basis conditions at the given pressure conditions, the percentage reductions in the N-16 transit times were calculated and averaged. The resulting reduction in N-16 transit time due to the EPU was determined to be 15.1%, which reduces the N-16 transit time by a factor of 0.849 (1.0- 0.151 = 0.849). The N-16 transit time reduction factor of Attachment I LR-N07-0154 LCR H05-01, Rev. 1 0.849 is applied to all originally calculated transit times for the steam components located downstream of the TSV. The newly calculated N-16 transit times were used to calculate the N-16 activity in the various steam components.
a) Include in your discussion the basis for the change in N-16 source strength at the reactor vessel, in the steam piping, and the resultant change in the turbine equipment at EPU conditions and an explanation of any assumptions or factors used.
Response
As stated in the response to RAI 11.1.a, while the magnitude of the source production increases in proportion to power, the concentration in the steam remains nearly constant. This is because the increase in activation production is balanced by the increase in steam flow. Therefore, the N-16 concentration of 50 jtCi/g at the reactor pressure vessel (RPV) nozzle remains bounding for the EPU.
The N-16 activity concentration in various steam components was calculated using the newly calculated post-EPU transit times as follows:
A = Ao e-xt Where; Ao = Initial N-16 Activity @ RPV Nozzle A = Post-EPU N-16 Activity In Given Steam Component X.= N-16 Decay Constant = 0.0972 sec-1 t = Newly Calculated Post-EPU N-16 Transit Time (sec)
Using the preceding equation, the N-16 activity concentrations in the various steam components were calculated. The N-16 gamma radiation exposure is directly proportional to the N-16 activity concentration.
Therefore, the relative increase in N-16 radiation exposure can be calculated by comparing the original and EPU activity concentrations in the component.
The N-16 transit time of interest is the first 10 seconds, because during this period the main steam has already traveled through the major steam components including the steam headers, HP turbine and outlet piping, cross-over and cross-under piping, moisture separators, and condenser, which contribute to the major in-plant (direct dose) and offsite skyshine doses. The N-16 source strength increases by approximately 16% during the first 10 seconds. Therefore, a multiplier of 1.16 can be used to calculate the post-EPU N-16 radiation exposure in all steam components.
The steam components having the larger transit times are less significant dose contributors.
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 b) Include in your discussion the pre-EPU radiation dose rates and post-EPU radiation dose rates near the steam turbine equipment and condenser components based on the percent change of N-16 concentration determined above.
Response
The post-EPU radiation exposures in various areas of TB complex are expected to increase by 16% above the pre-EPU radiation level and the resulting post-EPU radiation exposures are expected to remain within the applicable radiation zone allowable dose rate limits as shown in the following table:
Current Maximum Radiation Design Item Turbine Building Steam Zone Dose Rate Pre-EPU Post-EPU No Component Location Designation Limit Dose Rate Dose Rate 1 137' Turbine Building Moisture VI 10 R/hr 6.7 R/hr 7.7 R/hr Separators Rooms 1512 and 1514 2 137' Turbine Building Enclosure Room 1513 Turbine VII 100 R/hr 6.7 R/hr 7.7 R/hr 3 120' Turbine Building Room 1405 VII 100 R/hr 12.9 R/hr 15 R/hr 102' Condenser Bay Rooms - 1310, 4 1311, 1312, 1313, 1319, 1320, VII 100 R/hr 18.4 R/hr 21.4 R/hr 1321, and 1324 5 137' Turbine Building Reactor Feed VI 10 R/hr 3.3 R/hr 3.9 R/hr Pumps Rooms 1509, 1510, & 1511 137' Turbine Building Where 6 Generator and Turbine Enclosure 111 10 mR/hr 5 mR/hr 5.8 mR/hr for "C" LP Turbine Meet, Room 1501 11.12 Describe in detail the basis for the estimated doses to members of the public based on the calculated increase of N-16 concentration in the turbine equipment and condenser components and the resulting skyshine.
Response
The expected post-EPU increase in the in-plant radiation exposure in the TB complex has a negligible effect on the estimated doses to members of the public.
The TB concrete shielding and distance between the TB and offsite boundary are such that the post-EPU direct dose contribution from the steam components in the TB is negligible. The post-EPU N-16 skyshine dose rate at the nearest west site boundary is expected to be near the background radiation level. Therefore, it does not add into the total estimated doses to members of the public. Although the post-EPU direct dose and skyshine dose contribution to a member of the public (MOP) are negligibly small, a 16% increase in the combined direct dose and skyshine dose contributions was applied to the MOP doses.
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 a) Include in your discussion the dose rate at CLTP to members of the public onsite and offsite, the basis for calculating the expected post-EPU dose rates for members of the public onsite and offsite, the actual value of the post-EPU dose rate to the member of the public onsite and offsite, and a description of any occupancy factors used for your calculations.
Response
The dose limits for individual members of the public (MOP) are given in 10 CFR 20.1301 and compliance to these dose limits for the maximum exposed MOP and continuously present MOP are given in 10 CFR 20.1302(b)(1) and 1302(b)(2)(ii) respectively.
Compliance with 10 CFR 20.1302(b)(1) - maximum exposed MOP Due to installation of the security wall, the MOP is not allowed beyond the security post, which is located a considerable distance from the protected area fence. A conservative estimate was done postulating a hypothetical condition using the historical information when the food vendors were allowed to provide food catering service to onsite worker during the lunch time. The critical group for evaluating compliance with § 10 CFR 20.1302(a)(1) is assumed to be the food vendors that used to cater in front of the Security Center. The catering operation was restricted to one hour per day. The area that they utilized was partially shielded by the Security Center. For conservatism, the dose to this MOP was assumed to be at the location with highest exposure rate. The post-EPU annual dose to this MOP was calculated to be 7.31 mrem/year as follows due to combined hydrogen water chemistry (HWC) and EPU, which is much less than the allowable limit of 100 mrem/year.
Maximum Dose Rate 20 gtrem/hr Background Dose Rate -6 1.rem/hr Area Dose Rate 14 prem/hr Occupancy Time 1.5 hours/day (conservatively assumed = 1.5 x 1 hr)
Vending Days/year x 250 days/year Annual Exposure Time 375 hours/year Annual Dose to MOP
= 14. ptrem/hr x 375 hour0.00434 days <br />0.104 hours <br />6.200397e-4 weeks <br />1.426875e-4 months <br />s/year = 5,250 ýtrem/year x 1/1000 mrem/ptrem = 5.3 mrem/year The annual dose contribution to the MOP from gaseous and liquid releases dose contribution is much less than 1 mrem. Conservatively, this annual dose contribution was assumed to be 1.0 mrem/year.
- 30 -
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 CLTP The total annual dose to the maximum exposed MOP In CLTP = (5.1 +
1.0) mrem/year = 6.3 mrem/year. This is much less than the allowable limit of 100 mrem/year.
EPU The total annual dose to the maximum exposed MOP due to EPU = 1.16 x 6.3 mrem/year = 7.31 mrem/year, This is much less than the allowable limit of 100 mrem/year.
The maximum dose point was selected for this evaluation such that the estimated annual dose to the maximum exposed MOP remains bounding for other MOPs in the owner controlled area.
Compliance with 10 CFR 20.1302(b)(2)(ii) - continuously present MOP The site boundary locations were reviewed on the basis of continuous occupancy. The south and west site boundaries are adjacent to the Delaware River, where personnel occupancy will be very low. Therefore, only north and east site boundaries are considered for continuous occupancy at unrestricted area. The dose survey results indicated that the dose rate at the east site boundary (DP # 6) was higher than the north site boundary (DP # 5). Therefore, the annual dose to the continuously present MOP at the east site boundary is calculated to be 9.3 mrem/year as follows due to EPU. This annual dose is much less than the allowable limit of 50 mrem/year.
East Boundary Dose Rate 6.8 grem/hr Background Dose Rate -6. grem/hr Area Dose Rate 0.8 jtrem/hr Occupancy Time 24 hours/day Days/year x 365.25 days/year Annual Exposure Time 8,766 hours/year Annual Dose to MOP
= 0.8 prem/hr x 8,766 hour0.00887 days <br />0.213 hours <br />0.00127 weeks <br />2.91463e-4 months <br />s/year = 7,013 ptrem/year x 1/1000 mrem/[trem = 7 mrem/year The annual dose contribution to the MOP from gaseous and liquid releases dose contribution is much less than 1 mrem. Conservatively, this annual dose contribution is assumed to be 1.0 mrem/year.
Attachment I LR-N07-0154 LCR H05-01, Rev. I CLTP The total annual dose to the continuously present MOP at CLTP = (7.0 +
1.0) mrem/year = 8.0 mrem/year. This is much less than the allowable limit of 50 mrem/year.
EPU The total annual dose to the continuously present MOP due to EPU = 1.16 x 8 mrem/year = 9.3 mrem/year. This is much less than the allowable limit of 50 mrem/year.
Compliance with the 10 CFR 20.1301 dose limits is documented in the following table:
Onsite Offsite Hypothetical Continuously Maximum Exposed Present MOP MOP Compliance 10 CFR 10 CFR Compliance ___ 20.1302(b)(1) 20.1302(b)(2)(ii)
AnnualDose 6.3 mrem 8 mrem (CLTP)
Dose Rate Basis 16% Increase 16% Increase (EPU)
Occupancy Requirement 375 hour0.00434 days <br />0.104 hours <br />6.200397e-4 weeks <br />1.426875e-4 months <br />s/year 8,766 hour0.00887 days <br />0.213 hours <br />0.00127 weeks <br />2.91463e-4 months <br />s/year Annual Dose (EPU) 7.31 mrem 9.3 mrem llowable Annual 100 mrem 50 mrem Dose Limit b) Include in your discussion the dose contributions from skyshine and other radiation sources to members of the public at offsite locations, where these locations are located, and demonstrate that the maximum exposed member of the public offsite meets the requirements of Title 40 of the Code of FederalRegulations, Part 190 (40 CFR Part 190).
Response
As stated in Response to RAI Question 11.12, the post-EPU skyshine dose rate contribution to the closest site boundary location is expected to be near background radiation level and it does not add into the estimated doses to members of the public.
The direct dose contributions from other radioactive sources such as the condensate storage tank, the radwaste transport casks, and the spent fuel shipping cask at the site boundaries become negligible based on the Attachment I LR-N07-0154 LCR H05-01, Rev. 1 distances between the radioactive sources and closest site boundary location. Therefore, the assessment done to comply with 10 CFR 1302(b)(2)(ii) for the continuously present MOP at the site boundary and the resulting post-EPU offsite dose of 9.3 mrem/year are expected to be bounding for the maximum exposed MOP at the offsite to comply with 40 CFR 190 requirements. This resulting annual dose of 9.3 mrem/year is much less than the allowable limit of 25 mrem/year. The inhaled and ingestion doses from the gaseous and liquid effluent releases are included in the annual site boundary dose.
c) Include in your discussion the dose contributions from skyshine and other radiation sources to members of the public at onsite locations, where these locations are located, and demonstrate that the maximum exposed member of the public onsite meets the requirements of 10 CFR Part 20.
Response
Please see the responses to RAI Questions 11.12, 11.12.a, and 11.2.b.
d) Include in your discussion whether these changes continue to meet the requirements of the offsite dose calculation manual 6.9.1.8 and 6.9.1.7.
Response
The ODCM Technical Specifications 6.9.1.7 (Hope Creek) and 6.9.1.8 (Salem) require producing the Annual Radioactive Effluent Release Report (ARERR) to comply with the requirements outlined in Regulatory Guide 1.21 including the maximum annual doses to the MOP from the liquid and gaseous releases to comply with the 10 CFR 20.1301 and 40 CFR 190 requirements.
The annual doses to the MOP calculated in design calculation (as summarized in the response to RAI Questions 11.12 through 11.12.d) are based on conservative assumptions so that they remain bounding for the remaining design life of the plant to eliminate the need for future revisions to the design calculations that document this information. Unlike the design calculations, the ODCM calculates the annual doses based on the actually measured effluent release parameters, which are much less severe than those modeled in the design calculations. Therefore, the annual doses to the MOP calculated in the design calculations will continue to remain bounding for the information reported in the ARERR.
11.13 Demonstrate based on the values described above, that a member of the public which includes a member of the New Jersey National Guard working a 40-hour week inside the site boundary for 12 months, meets the requirements of 10 CFR Part 20.
Attachment I LR-N07-0154 LCR H05-01, Rev. 1 a) Include in your discussion explanations of any assumptions or factors used.
Response
The ARERR, Part E, classifies the New Jersey National Guard (NJNG) as a MOP performing activities inside site boundary. The definition changed on September 11, 2001. The various food vendors that have previously comprised the maximally exposed group are no longer allowed on site. The definition of MOP is the members of NJNG to augment the security force at the site. Their typical patrol spans the site including the Hope Creek barge slip (TLD 16S1);
dredge spoils (TLD CA8); and baseball field (TLD CA15). The annual doses at these locations are averaged to estimate their doses, which are subject to change annually based on the radiation exposure data collected every year from the TLDs located on the NJNG patrol route. A clarification is required to note that the conservatively calculated annual dose of 9.3 mrem for the continuously present MOP to comply with 10 CFR 1302(b)(2)(ii) requirement (as summarized in the response to RAIs 11.12 through 11.12.d) will constitute the licensing basis for the EPU and it will bound the corresponding NJNG annual dose information in the ARERRs for remaining design life of plant. Although, the NJNG dose in the following section is calculated based on the annual occupancy, typically, individual NJNG members are not assigned to the site for an entire year.
The annual dose to the NJNG for 2005 was reported to be 0.243 mrem TEDE, which can be calculated as follows:
NJNG annual occupancy = 40 hrs/week x 52 weeks/year = 2080 hrs Annual TLD dose reading = 365.25 days/year x 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day = 8766 hrs NJNG Occupancy Factor = 2080 hrs / 8766 hrs = 0.237 = 0.25 2005 annual TLD total body dose for full annual occupancy = 0.969 mrem (from table below) 2005 annual TLD total body dose corrected for occupancy = 0.969 x 0.25
= 0.242 mrem Inhaled annual organ dose due to effluent releases = 4.84E-4 mrem/year (ARERR-2005)
Attachment I LR-N07-0154 LCR H05-01, Rev. I New Jersey National Guard Annual Offsite Dose - 2005 Average Monthly Offsite Average Nat Guard Dose Rate (mremlmonth) Quarterly Quarterly Quarter Including Above Offsite Offsite ID Background Background Background Dose Rate Dose (mrem/Quarter) (mrem)
A B C=A-B D=Cx3molqtr E=DxO.25 First Quarter 3.76 3.94 0.00 0.00 0.00 Second Quarter 4.38 4.27 0.10 0.31 0.08 Third Quarter 4.47 4.25 0.22 0.65 0.16 Fourth Quarter 4.33 4.51 0.00 0.00 0.00 Total Annual NJ National Guard Dose (mrem) 0.969 0.242 CLTP 2005 total annual dose to NJNG = 0.242 mrem total body + 4.84E-4 mrem inhaled = 0.243 mrem TEDE, which is a small fraction of annual allowable dose limit of 100 mrem and the same as that reported in the 2005 ARERR.
EPU Corresponding post-EPU total annual dose to NJNG would be = 0.243 mrem x 1.16 = 0.282 mrem, which is a small fraction of the annual allowable dose limit of 100 mrem/year.
Dose to members of the public, including members of the New Jersey National Guard working a 40-hour week inside the site boundary for 12 months, will continue to meet the requirements of 10 CFR Part 20.
- 14) Mechanical & Civil Engineering Branch (EMCB) (revised response)
The response to RAI 14.47 previously submitted did not reflect the completed re-analysis of the feedwater nozzles described in PUSAR section 3.2.2. Changes from the previously submitted response are marked.
14.47 In the PUSAR Section 3.2.2 through 3.2.2.3, "Reactor Vessel Structural Evaluation," states that "The effect of CPPU was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME B&PV Code. For the components under consideration, the 1968 code with addenda to and including winter 1969, which is the code of construction, is used as the governing code. However, if a Attachment I LR-N07-0154 LCR H05-01, Rev. 1 component's design has been modified, the governing code for that component is the code used in the stress analysis of the modified component. The Hope Creek CPPU utilizes the original code of construction as the governing code for all components for CPPU conditions. New stresses are determined by scaling the "original" stresses based on the CPPU conditions (temperature and flow).
The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions."
a) Provide a summary of the analyses stated above which justify that loading changes due to CPPU in the analysis of the components affected at the normal, upset, emergency and faulted conditions do not affect the structural integrity of these components.
Response
As defined in the LTR for CPPU, the only components that require evaluation are: (a) those with a pre-existing (OLTP) CUF >0.5 and (b) those that experience an increase in flow, temperature, Reactor Internals Pressure Differences (RIPDs), and other mechanical loads due to CPPU. Therefore, the only components in the HCGS vessel that require evaluation for CPPU operating conditions are the main closure studs, shroud support, and core spray nozzle.
For the main closure studs and the core spray nozzle, there is a small (0.2%)
change in temperature, but no change in pressure or flow. Hence, this small change in temperature has an insignificant effect on primary plus secondary stresses (P+Q) and fatigue Cumulative Usage Factor (CUF).
For the shroud support, there is no increase to the thermal stresses due to CPPU. The dynamic loads and recirculation LOCA loads do not change for CPPU conditions. The only changes are due to an RIPD increase of 11.1%
across the shroud support. This causes an increase in the primary stresses (P) of the same percentage. As a result, (P+Q) increases approximately 3%. This is an insignificant increase; it is judged that the CUF remains unchanged.
Table 14.47-1 provides the Original Licensed Thermal Power (OLTP) CUF compared to the Constant Pressure Power Uprate (CPPU) results.
Similarly, (P+Q) values are presented in the Table 14.47-2.
b) Provide a list of components that have their design modified along with a description of the design modifications, governing code, and maximum stress summary versus allowable stress limits at critical locations.
36-
Attachment I LR-N07-0154 LCR H05-01, Rev. I
Response
Table 14.47-3 provides a summary of the applicable Code, stresses, and allowable stresses for all components that have been modified since the original design.
In addition to the component evaluations for modifications summarized in Table 14.47-3, welded overlays on the core spray nozzle to safe end weld (N5B) and the reactor vessel recirculation inlet nozzle to safe end weld (N2K) were satisfactorily evaluated for EPU conditions as described in response to RAI 5.3 (PSEG letter LR-N07-0056 dated March 22, 2007).
c) Provide the procedural tensioning method of Main Closure Studs that includes loads due to CPPU conditions. Show that the tensioning stud load including tolerances is within code allowables when including the higher CPPU conditions.
Response
As provided above in response to RAI 14.47a, the effect of CPPU on main closure stud CUF and (P+Q) is not significant. Therefore, no change to the stud tensioning method or elongation tolerances is required.
Table 14.47-1 Cominonent OLTP ýC.U F.CPPU QUF . :Conclusion Main Closure Studs 0.755 0.755 Component Acceptable Core Spray Nozzle 0.796 0.796 Component Acceptable Shroud Support 0.672 0.672 Component Acceptable Attachment I LR-N07-0154 LCR H05-01, Rev. 1 Table 14.47-2 Component OLTP P+ýQ CPPU P+Q Code Conclusion (ksi)
Main Closure 108.9 109.11 110.4 Component Studs Acceptable Core Spray 13.76 13.79 43.11 Component Nozzle Acceptable (See Note 1)
Shroud 22.03 24.41 69.9 Component Support Acceptable (See Note 1)
Note 1:
Thermal Bending has been removed as allowed by the ASME Code.
Table 14.47-3 Co'm'pon~ent: Modificationý Code Modification Maximum Allowable Recirculation Safe End and ASME Boiler & 26.4 30.2 41.17 Inlet Nozzle Thermal Pressure Vessel Sleeve Code, 1974 Replacement Edition and Addenda up to Summer 1976 CRD Nozzle ASME Boiler & 39.48 39.56 80.1 Hydraulic Capped Pressure Vessel System Code, 1974 Return Edition and Nozzle Addenda up to Winter 1975 Feedwater Safe End ASME Boiler & 41.813 41.352 50.850 Nozzle Safe Replacement Pressure Vessel End Code, 1998 Edition including the 2000 Addenda References
- 1. PSEG letter LR-N06-0286, Request for License Amendment: Extended Power Uprate, September 18, 2006 Attachment I LR-N07-0154 LCR H05-01, Rev. I
- 2. NRC letter, Hope Creek Generating Station - Request for Additional Information Regarding Request for Extended Power Uprate (TAC NO. MD3002), June 7, 2007