ML081930274
| ML081930274 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/10/2008 |
| From: | Gerald Bichof Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 08-0369 | |
| Download: ML081930274 (54) | |
Text
Dominion Nuclear Connecticut, Inc.
'loon Dominion Boulevard, Glen Allen, Virginia", i()60 1X"e!, i"ldress: www.dom.com July 10, 2008 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2378 Serial No.:
08-0369 NLOS/GAW: RO Docket No.:
50-423 License No.: NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST DNC COMMENTS ON DRAFT SAFETY EVALUATION - STRETCH POWER UPRATE Dominion Nuclear Connecticut, Inc. (DNC) submitted a stretch power uprate license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3) in letters dated July 13, 2007 (Serial Nos.
07-0450 and 07-0450A),
and supplemented the submittal by letters dated September 12, 2007 (Serial No. 07-04508), December 13, 2007 (Serial No. 07-0450C), March 5, 2008 (Serial No.
07-0450D), March 27, 2008 (Serial No. 07-0450E), April 24,2008 (Serial No. 07-0450F), May 20, 2008 (Serial No. 07-0450H) and May 21, 2008 (Serial No. 07-04501).
The NRC staff forwarded requests for additional information (RAls) in October 29,2007, November 26,2007, December 14,2007, December 20,2007 and April 23, 2008 letters.
DNC responded to the RAls in letters dated November 19, 2007 (Serial No. 07-0751), December 17, 2007 (Serial No. 07-0799), January 10, 2008 (Serial Nos. 07-0834, 07-0834A, 07-0834C, and 07-0834F), January 11, 2008 (Serial Nos. 07-08348, 07-0834E, 07-0834G, and 07-0834H), January 14, 2008 (Serial No. 07-0834D), January 18, 2008 (Serial Nos.
07-0846, 07-0846A, 07-08468, 07-0846C, and 07-0846D), January 31, 2008 (Serial No. 07-08341), February 25, 2008 (Serial Nos. 07-0799A and 07-0834J),
March 10, 2008 (Serial Nos. 07-0846E and 07-0846F), March 25, 2008 (Serial No. 07-0834K), April 4, 2008 (Serial No. 07-0834L), April 29, 2008 (Serial No. 08-0248), and May 15, 2008 (Serial No. 08-0248A).
On June 12, 2008, the NRC issued for comment a draft safety evaluation report regarding the stretch power uprate license amendment request. The attachment contains DNC comments on the draft safety evaluation report.
The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the December 13, 2007 DNC letter (Serial No. 07-0450C).
Serial No. 08-0369 Docket No. 50-423 Comments on Draft Safety Evaluation Page 2 of 3 Should you have any questions in regard to this submittal, please contact Mr.
Geoffrey Wertz at 804-273-3572.
Sincerely, Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T.
Bischof, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc.
He has affirmed before me that he 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 his knowledge and belief.
Acknowledged before me this I D"'h,day Of~,2008.
My Commission Expires: ~UQ 1; 3/, 2008.
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Serial No. 08-0369 Docket No. 50-423 Comments on Draft Safety Evaluation Page 3 of 3 Commitments made in this letter: None Attachment cc:
U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. G. Lamb Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-8B1A Rockville, MD 20852-2738 Ms. C. J. Sanders Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-8B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Serial No. 08-0369 Docket No. 50-423 ATTACHMENT LICENSE AMENDMENT REQUEST STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST DNC COMMENTS ON DRAFT SAFETY EVALUATION REPORT MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.
Serial No. 08-0369 Docket No. 50-423 Table of Contents DNC Comments on Draft Safety Evaluation Report for the Stretch Power Uprate License Amendment Request 1
Section 1.4, Plant Modifications 1
Section 2.1, Materials and Chemical Engineering 3
Section 2.2, Mechanical and Civil Engineering 5
Section 2.3, Electrical Engineering 7
Section 2.5, Plant Systems 8
Section 2.6, Containment Review Considerations 16 Section 2.7, Habitability, Filtration, and Ventilation 18 Section 2.8, Reactor Systems 19 Section 2.8, Reactor Systems (continue) 22 Section 2.9, Source Terms and Radiological Consequences Analyses 32 Section 2.10, Human Performance 47 Section 3.2, Technical Specifications 48
Serial No. 08-0369 Docket No. 50-423 Page 1 of 49 DNC Comments on Draft Safety Evaluation Report for the Stretch Power Uprate License Amendment Request Section 1.4, Plant Modifications Page Location Sentence Comment 3
Plant Modifications, "4. For the turbine generator, provide the Please delete "(a) new operating point for Item 4 following: (a) new operating point for generator excitation", "(d) throttle pressure generator excitation; (b) control valve and excess throttle pressure circuit position demand against lift settings for recalibrations" and "(g) main control board the valve position cards; (c) changes to and panel meter replacements." Please refer load imbalance circuits; (d) throttle to DNe letter dated May 20, 2008 (Serial No.
pressure and excess throttle pressure 07-0450H), Attachment, Updated Table 1.0-1 circuit recalibrations; (e) sensor rescaling MPS3 Power Uprate Planned Modifications, for steam pressure changes; (f)
Page 2, "Turbine Generator" which revised instrument scaling; and (g) main control the list of planned modifications.
board and panel meter replacements."
4 Plant Modifications, "7. Pipe support modifications for the Please replace "extraction steam" with Item 7 condensate system, feedwater system, "containment recirculation." Please refer to CCW system, and extraction steam DNC letter dated May 20, 2008 (Serial No.
system."
07-0450H), Attachment, Updated Table 1.0-1 MPS3 Power Uprate Planned Modifications, Page 3, "Pipe Support Modifications, Condensate, Feedwater, Component Cooling water, and Containment Recirculation,"
which revised the list of planned modifications.
Serial No. 08-0369 Docket No. 50-423 Page 2 of 49 Page Location Sentence Comment 4
Plant Modifications, "9. Provide instrument loop rescaling for Please delete "(a) isophase bus duct cooler Item 9 the following: (a) isophase bus duct flow" and "(b) moisture separator reheater cooler flow; (b) moisture separator steam flow." Please refer to DNC letter reheater steam flow; and (c) first stage dated May 20,2008 (Serial No. 07-0450H),
turbine pressure."
Attachment, Updated Table 1.0-1 MPS3 Power Uprate Planned Modifications, Page 3, "Instrument Loop Rescaling" which revised the list of planned modifications.
Section 2.1, Materials and Chemical Engineering Serial No. 08-0369 Docket No. 50-423 Page 3 of 49 Page Location Sentence Comment 8
2na paragraph, last "The methodology used to project the As stated in DNC letter dated November 19, sentence neutron fluence values for surveillance 2007 (Serial No. 07-0751), in response to capsules X (WCAP-15045) and W RAI CVIB-07-002, the fluence analyses per (WCAP-16629-Reference 2) adhered to WCAP-15405 were compliant with the the guidance in Regulatory Guide (RG) requirements of Draft Regulatory Guide DG-1.190 "Calculational and Dosimetry 1053, which was the precursor to Regulatory Methods for Determining Pressure Vessel Guide 1.190. A more accurate statement Neutron Fluence."
would be: The methodology used to project the neutron fluence values for surveillance capsules X (WCAP-15405) and W (WCAP-16629-Reference 2) adhered to the guidance in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," or its predecessor."
8 3ra paragraph, first "The projected peak neutron fluence As stated in DNC letter dated November 19, sentence values on the inside surface of the RV for 2007 (Serial No. 07-0751) in response to RAI 32 and 54 EFPYs are 1.97 x 1019 n/cm2 CVIB 07-003, the fluences given in these (E > 1 MeV) and 3.31 x 1019 n/cm2 (E > 1 sentences correspond to earlier projections MeV) respectively. In addition, the based on capsule X with the assumption that calculation assumed that the equilibrium the flux values from cycle 4 through 6 would cycle 15 core loading will be implemented continue throughout plant life. While these through the 54 EFPYs."
values are conservative, the values more appropriate to SPU are 1.63 x 1019 n/cm2 and 2.70 x 1019 n/cm2 for 32 EFPY and 54 EFPY respectively.
Serial No. 08-0369 Docket No. 50-423 Page 4 of 49 Page Location Sentence Comment 11 Last paragraph, third "The licensee used position 2.1 of the RG It should be noted that information is sentence 1.99 Revision 2, where credible provided in Table 2.1.3-4 of the License surveillance capsule test data was used Report for both Position 1.1 and Position 2.1.
for establishing RTPTS value for the As stated on page 2.1-35 of the License intermediate shell plate."
Report, "The limiting material is Intermediate Shell Plate B9805-1, with the more limiting RTPTS value occurring for calculations using the RG 1.99, Rev. 2 Position 1.1 Chemistry as opposed to the Position 2.1 Chemistry Factor calculated from credible surveillance data."
12 Second paragraph, "Specific review criteria are contained in Please replace "WCAP-14277,Ll" with last sentence SRP Section 4.5.2, WCAP-14277,21 and "WCAP-14577,21".
BAW-2248.22" WCAP-14577 is the correct reference for the License Renewal Evaluations: Aging Management for Reactor Internals.
22 Flow-Accelerated "The licensee's FAC program is based on The statement should read "The licensee's Corrosion, First NUREG-1344, "Embrittlement Criteria for FAC program is based on NRC Bulletin 87-paragraph, sixth Zircaloy Fuel Cladding Applicable to 01, "Thinning Pipe Walls in Nuclear Power sentence Accident Situations in Light-Water Plants",..." This is stated in the licensing Reactors: Summary Report,...."
report, page 2.1-76 under MPS3 Current Licensing Bases.
Section 2.2, Mechanical and Civil Engineering Serial No. 08-0369 Docket No. 50-423 Page 5 of 49 Page Location Sentence Comment 34 Second paragraph, "The licensee performed its evaluations The wording "and its supports" should be First line under for the MPS3 RPVand its supports at removed since the code of record is different "Reactor Vessel and SPU conditions in accordance with the for the RPV Supports. The RPV supports are Supports" current plant code of record, ASME Code, ASME III 1974 Edition including Summer 74 Section III, Division 1, 1971 Edition addenda. Note that in this paragraph, the through summer 1973."
reference is made to FSAR Table 5.4-18 which provides the correct information.
Also please add the word "addenda" after "summer 1973."
35 First paragraph, third "In its response to the NRC staff's RAI, Please use "simplified elastic-plastic sentence the licensee provided a summary of the evaluation" instead of "elastic-plastic results of the elastic-plastic evaluation evaluation." This is stated in DNe letter which shows that the special rules for dated April 4, 2008, (Serial No. 07-0834L) exceeding 3Sm as provided by (a) page 1 of the Attachment.
through (f) of Subparagraph NB-3228.3 have been met."
36 Last paragraph, "The Code of Record for the SG supports The word: "summer" should be "winter." The seventh line.
is the ASME Boiler and Pressure Vessel correct reference is provided in FSAR Table Code,Section III, Subsection NF 1974 5.4-18, "Equipment Supports, Loading Edition including 1974 summer Addenda."
Combinations, and Design Allowable Stresses."
39 End of First
"...Section III, Division 1, 1974 Edition The word: "summer" should be "winter."
The Paragraph at top of through summer 1974 Addenda."
correct reference is provided in FSAR Table page.
5.4-18.
44 Safety-Related "The licensee has reviewed system level Please insert "Cat 1 AOVs in the" between Valves and Pumps, design basis calculations for main steam "for" and "main steam." The following is a second paragraph, system (MSS), service water system, and quotation from the license amendment report
Serial No. 08-0369 Docket No. 50-423 Page 6 of 49 Page Location Sentence Comment second sentence CVCS system."
section 2.2.4, page 2.2-141 :
"The system level design basis review calculations for the Category 1 AOVs in the following systems were reviewed:
- Main steam system (main steam pressure relieving valves, turbine-driven auxiliary feedwater pump steam supply valves)
- Service water system (service water diesel generator heat exchanger outlet valves)
- Chemical and volume control system (reactor coolant letdown inside/outside Containment isolation valves)"
44 Safety-Related "The Code of record for MPS3 is the 1980 Please insert "original" between "The" and Valves and Pumps, Edition through winter 1980 Addenda of "Code". As stated in the last sentence of the fourth paragraph, the ASME Code,Section XI."
paragraph, "The 1ST program must be second sentence periodically updated to meet applicable ASME O&M Code Requirements specified in 10 CFR 50.55a," which means that the code years and addenda are updated as required.
The "original" code was 1980 Edition through winter 1980 addenda - This would be a clarification.
Section 2.3, Electrical Engineering Serial No. 08-0369 Docket No. 50-423 Page 7 of 49 Page Location Sentence Comment 49 Main Steam Valve "The valves fail closed, and the failure of Please replace with "Once the valves move Building, third the limit switches would not cause the to their fail safe position, closed, the failure of paragraph, fourth reopening of the valve."
the limit switches would not cause the sentence from the reopening of the valve." Please refer to the bottom response to question EEEB-07-0052 Follow-Up (EEEB-08-0108) in DNC letter dated May 15, 2008 (Serial No. 08-0248A).
50 Turbine Building, "The pressure for the turbine building Please add "from a MSLB" after "turbine third line of the last remains bounding for SPU conditions."
building." The phrase "from a MSLB" is paragraph missing.
51 Turbine Building, "The primary function of the transmitters is The second sentence in the second second sentence in to provide input to the rod control system, paragraph does not read the same as the second where a transmitter failure could initiate a intended in the RAI response to questions paragraph rod withdrawal demand signal coincident EEEB-07-0056 and EEEB-07-0057 in DNC with a steam line break."
letter dated January 10, 2008 (Serial No. 07-0834C). Please consider replacing the sentence with:
"One function of the transmitters is to provide input to the rod control system, where a transmitter failure could initiate a rod withdrawal demand signal coincident with a steam line break."
Section 2.5, Plant Systems Serial No 08-0369 Docket No, 50-423 Page 8 of 49 Page Location Sentence Comment 61 Section 2.5.1.1.1
"..... and vessels was based on Please delete the word "instantaneous" since Flooding Protection, instantaneous release of fluid from the there is no basis for the use of Technical Evaluation, tanks."
instantaneous. The license amendment first Paragraph, third report does not discuss "instantaneous line release."
63 Technical Evaluation, "As discussed in UFSAR Section 3.8.4, UFSAR Section 3.8.4 does not discuss First line of the first the CWS provides a continuous supply of circulating water system heat removal.
paragraph cooling water to the MC to remove Therefore please delete the phrase "As excess heat from the steam turbine discussed in UFSAR Section 3.8.4."
exhaust cycle and auxiliary systems."
63 Conclusion, third line "the increased volumes of fluid leakage This section reads as though DNC is that could potentially result from these performing modifications to the CWS and modifications would not result in the that there is an increase in fluid leakage failure of safety-related SSCs following volume. Please consider revising the implementation of the proposed SPU."
conclusion as suggested below:
"Since the CWS flow and operating pressures will remain unchanged for the SPU, and no modifications to the CWS or turbine building are performed that would affect the analyses associated with flooding, the proposed SPU is acceptable with respect to flooding of CWS and would not result in the failure of safety-related SSCs following implementation of the proposed SPU."
66 Second paragraph "With respect to a postulated MSLB that This paragraph has mixed up the results for results in an increase in the turbine the turbine building and the auxiliary bUilding building peak temperature from 500 OF to with the results for the main steam valve 565.5 OF, the previously performed building.
Serial No. 08-0369 Docket No. 50-423 Page 9 of 49 Page Location Sentence Comment temperature analysis is no longer bounding for SPU conditions. The impact The LAR section for a TB HELB is located on on the qualification of equipment as a LAR page 2.5-28 (bottom).
This LAR says result of an MSLB, is discussed in that DNC is not evaluating the TB HELB Section 2.3 of this report. The increase in temperature increase because ONC will peak temperature may also impact change the TB HELB safe shutdown analysis specific valves that are essential for to eliminate crediting any equipment located shutdown in the event of an HELB in the in the TB.
auxiliary building. MPS3 UFSAR Table 3.6-5 lists the two specific valves. The There are two isolation valves located in the EQ of these two valves to remain TB that are credited for an AB HELB (Le.,
effective under increased temperature 3ASS-AOV102NB). The subject SER condition is also discussed in Section 2.3 paragraph discusses these valves. SPU has of this report."
no impact upon the AB temperature/pressure profile associated with an AB HELB.
Because the AB temperature/pressure is unaffected by SPU and the AB is not adjacent to the TB, an AB HELB has no impact upon TB temperature/pressure.
Thus, SPU has no impact upon the AB HELB credited steam isolation valves (Le., 3ASS-AOV102NB) that are located in the TB.
The "500 of to 565.5 of temperature increase is associated with the Main Steam Valve Building (MSVB). Specifically, this temperature increase is for a MSVB MSLB event.
DNC suggests simplifying/clarifying the
Serial No. 08-0369 Docket No. 50-423 Page 10 of 49 Paqe Location Sentence Comment subject paragraph as follows:
"For a Main Steam Valve Building (MSVB)
HELB, MSVB temperature qualification profiles are impacted (i.e., peak temperature increases from 500 of to 565.5 OF).
Section 2.3 documents assessment of MSVB harsh environmental changes with respect to credited SSC's.
For a Turbine Building (TB) HELB, SPU will increase the TB HELB temperature.
However, to address this consequence, the licensee proposes to revise the TB HELB safe shutdown analysis to eliminate crediting equipment located in the TB. In the proposed design, a TB HELB has no adverse impact on credited SSC's because no credited equipment is located in the TB and there is no impact to adjacent areas due to building layout/design.
Review of pipe failure assessments for all other buildings concluded that SPU has no adverse impact upon existing moderate or hiqh enerqy pipe failure analysis."
Serial No. 08-0369 Docket No. 50-423 Page 11 of 49 Page Location Sentence Comment 69 First paragraph, Last "The licensee indicated that for the Consistent with RAI response to Question sentence proposed SPU condition, there is no AFPB-07-0006, please consider revising increase in the potential for a radiological from "there is no increase in the.." to "there is release resulting from a fire."
no significant increase in the... "
69 Sections after the first "By letter dated December 17, 2007."
A more up to date response to RAI AFPB two paragraphs 0007 is provided in letter dated February 25, 2008. The response in the letter dated December 17 2007 was superseded by the response in the letter dated February 25, 2008. The primary difference between the two responses is the clarification of the MSIV closure time in the later response. Therefore please use the response from February 25, 2008 letter.
71 Second paragraph, "The licensee also stated that in order to Please consider changing from "fire third line provide margin, the AFW initiation time shutdown and design compliance report" to will be reduced to less than 21.5 minutes "fire shutdown procedure and design prior to implementation of the SPU by compliance report."
making changes to the fire shutdown and design compliance report."
Serial No. 08-0369 Docket No. 50-423 Page 12 of 49 Page Location Sentence Comment 78 Technical Evaluation, "No modifications are being made to the DNC proposes to replace this sentence with:
second sentence TGSS and non-condensable gases will "No physical changes to system components continue to be monitored for radiation."
or changes in system operation are required due to the slight increase in sealing flow and gland steam condenser cooling flow. There is no radiation monitoring at the gland seal condenser vent, as radioactive gaseous releases fall within the total unmonitored steam release specifications from the Turbine Building as defined in NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Pressurized Water Reactors, April 1976, Section 2.2.6."
The proposed change reflects the results of DNC evaluation as provided on pages 2.5-64 and 2.5-66 of the licensing report.
81 First sentence, last "The operating band of the CCW surge Since the operating band of the CCW surge paragraph tank is sufficient to accept a 100 gallon tank can accept more than 100 gallons, increase in system thermal expansion please revise the sentence to read as "The volume relative to the current system operating band of the CCW surge tank is volume."
sufficient to accept the increase in system thermal expansion volume relative to the current volume." The LAR section 2.5.4.3.2.3 describes an exact value of 100 gallons which is an estimated value.
Serial No. 08-0369 Docket No. 50-423 Page 13 of 49 Page Location Sentence Comment 84 Fourth paragraph, "Upon a loss of normal feedwater The proposed 7/6 DWST sizing criteria is 7-second sentence (LONF), the licensee's proposed 7-hours hours at hot standby. Therefore, DNe at hot shutdown with higher decay heat suggests changing "7-hours at hot shutdown" as a result of SPU does not impact the to "7-hours at hot standby."
available water volume in DWST to provide cooldown."
85 Third paragraph, first "The DWST inventory requirement is ONC suggests using: "The DWST inventory sentence also relative to fire shutdown."
is also credited within the fire shutdown analysis."
Serial No. 08-0369 Docket No. 50-423 Page 14 of 49 Page Location Sentence Comment 86 Technical Evaluation, "The proposed SPU would increase DNC proposes replacing the sentence with:
first sentence steam temperature to 533 of from 500 "The proposed SPU would increase steam of.
temperature of the reheat lines to 533 of from 500 of.
The temperature increase to 533 of is associated only with the hot reheat lines as discussed in section 2.5.5 of the license report, page 2.5-111, under System/Component Design PressurelTemperature Versus SPU.
89 Technical Evaluation, "The licensee's analysis of a TT without Please replace with: "The licensee's analysis first paragraph.
reactor trip showed the pressurizer of a IT without reactor trip showed the PORVs could be challenged at SPU pressurizer PORVs could be challenged at conditions. A proposed modification of SPU conditions. A proposed modification to the steam dump load rejection controller the steam dump load rejection controller will "Hi 2" trip open setpoint from 12.0 of to include a change to both the Hi 1 and Hi 2 7.0 of, prior to the start of the coast setpoints. Hi 1 will be changed from 6.6 of down maneuver, and reset the controller to 5.3 of. Hi 2 will be changed from 15.8 of back to 12.0 of prior to the subsequent to 12.0 of. Should the plant choose to use startup for the new fuel cycle at SPU the alternate coastdown method identified as conditions would ensure that a TT would Option 2 in license amendment report Table not result in a challenge to the 2.4.1-4, then the Hi 2 setpoint will be pressurizer relief valve lift setpoint.
changed to 7.0 of for the coasdown and then During a load runback transient, the MSS reset to the new at power Hi 2 setpoint of pressure would remain less than the 12.0 of prior to subsequent startup for the MSSV setpoint."
new fuel cycle. Otherwise the Hi 2 setpoint will remain at 12.0 of for the Option 1 coastdown. The modified setpoints would ensure that a IT would not result in a challenge to the pressurizer relief valve lift
Serial No. 08-0369 Docket No. 50-423 Page 15 of 49 setpoint. During a load runback transient, the MSS pressure would remain less than the MSSV setpoint."
Please refer to table 2.4.1-4 (page 2.4-24) of the license amendment report.
Section 2.6, Containment Review Considerations Serial No. 08-0369 Docket No. 50-423 Page 16 of 49 Page Location Sentence Comment 97 Regulatory liThe containment maximum internal design MPS3 does not apply a maximum Evaluation 1st pressure is 45 pounds per square inch containment vapor design temperature limit paragraph, last gauge (psig), the minimum internal design of 296°F. The containment structural sentence pressure is 8.00 psia, and the design integrity is assured by applying a temperature is 296°F."
containment liner temperature limit of 280°F.
For equipment inside containment, a temperature qualification profile is used for qualification of equipment inside containment. As can seen from LR section 2.6.1.2.3.3.2, page 2.6-14, the predicted short term peak from a steam line break is 343°F which exceeds 296°F. This is also the case for the current licensing basis results documented in FSAR Section 6.2.1.1.3.7, page 6.2-16 where the predicted short term peak from a steam line break is 335.9°F.
Please consider revising or deleting this sentence.
98 LOCA Short-Term liThe LOCA short-term peak pressure and MPS3 does not apply a maximum Containment temperature are less than the containment containment vapor design temperature limit Response last design pressure and temperature 45 psig of 296°F. See discussion above. Please sentence and 296°F respectively."
consider revising or deleting this sentence.
98 LOCA Long-Term liThe initial conditions of maximum This should read liThe initial conditions of Containment pressure, maximum temperature and maximum pressure, maximum temperature Response 2nd minimum relative humidity along with a and maximum relative humidity... " This is sentence.
single failure or loss of an EDG produces stated on page 2.6-11 of the LR.
the slowest containment depressurization for this break."
Serial No. 08-0369 Docket No. 50-423 Page 17 of 49 Page Location Sentence Comment 99 Main Steam Line "The peak pressure occurs at 12.6 seconds The 12.6 seconds should be 194.3 seconds.
Break Containment from the time of the instantaneous break."
This is stated in Table 2.6.1.2.3-2 of the LR.
Response 1st paragraph, third sentence
Section 2.7, Habitability, Filtration, and Ventilation Serial No. 08-0369 Docket No. 50-423 Page 18 of 49 Page Location Sentence Comment 108 Technical Evaluation,
"(2) for a fuel handling accident (FHA),
Please consider changing "pressurized First paragraph, item the CR emergency ventilation system filtration" to "filtered recirculation." This (2)
(CREVS) is required to be in the change will make this item consistent with pressurized filtration MODE within 30 other sections of the SER. For example, the minutes of the accident, whereas in the term 'filtered recirculation' is used for the current licensing basis analysis it is condition of the control room 30 minutes post required to be in the pressurized filtration FHA. See section 2.9.2.2, page 199, 2nd, MODE within 1.685 hour0.00793 days <br />0.19 hours <br />0.00113 weeks <br />2.606425e-4 months <br />s97 of the 3rd and last paragraph; Control Room accident."
Habitability, page 211, 1st and 4th paragraph; Changes in Emergency and Abnormal Operating procedures, page 229, 1st bullet.
Serial No. 08-0369 Oocket No. 50-423 Page 19 of 49 Section 2.8, Reactor Systems As part of ONC review of the NRC's draft SPU safety evaluation report (SER),
errors were identified in Table 2.8.2-1, "Range of Key Safety Parameters Safety Parameter," page 2.8-18 of attachment 5 of the license amendment request dated July 13, 2007 (Serial Nos. 07-0450 and 07-0450A). The value given as the current vessel average temperature is the core average temperature rather than the vessel average temperature. Footnote c was incorrectly assigned to "Most Positive MDC (Klg/cm3)." The following are the proposed corrections:
The current vessel average temperature is 587.1 of.
Foot note c "Linear ramp from 70% to 100% power" applies to the MTC
>70%.
Footnote d applies to the second entry for the Most Positive MOC.
Footnote d "MOC assumed for feedline break analysis (HFP ARO)"
applies to the Most Positive MOC of 0.45.
The values used in Table Line item 6 and 7, on Page 120 of the SER, came from license report Table 2.8.2-1. Therefore, ONC proposes changes to the following two lines in order to incorporate the above mentioned corrections.
Replace the following three lines:
Current Design Uprate Analysis Values Vessel Average Coolant Temp. HFP (OF) 591.6 581.5 to 589.5 a,b Most Positive MTC, >70% (pcmJOF) 0.0 0.0 Most Positive MCC ***
0.50 0.50, linear ramp from 70% to 100% power to 0.45 With:
Current Design Uprate Analysis Values Vessel Average Coolant Temp. HFP (OF) 587.1 581.5 to 589.5 a,b Most Positive MTC, >70% (pcmJOF) 0.0, linear ramp from 0.0, linear ramp from 70% to 100% power 70% to 100% power Most Positive MDC 0.50 0.50 0.45 for FWLB at HFP ARO
Serial No. 08-0369 Docket No. 50-423 Page 20 of 49 The following is the corrected license report Table 2.8.2-1:
Range of Key Safety Parameters Safety Parameter Current Design Analysis Values Values Reactor Core Power (MWt) 3411 3650 587.1 581.5 to 589.5 a,D Vessel Average Coolant Temp. HFP CF) 2250 2250 Coolant System Pressure (psia) 5.45 5.83 Core Average Linear Heat Rate (kW/ft)
Most Positive MTC (pcmrF)
Power < 70%
+ 5.0
+ 5.0 O.Oe O.Oe Power ~ 70%
Most Positive MDC (Klg/cm3) 0.50 0.50 0.45d Doppler Temperature Coefficient
-3.20 to -0.91
-3.20 to -0.90 (pcmrF)
Doppler Only Power Coefficient (pcm/%Power)
-9.55 to -5.42 Least Negative, 118%RTP to HlP
-19.40 to -11.36 Most Negative, 118%RTP to HZP
-11.55 to -7.02 Least Negative, 121%RTP to HlP
-19.40 to -11.17 Most Negative, 121%RTP to HlP
-9.55 to -5.00 Least Negative, 130%RTP to HlP
-19.40 to -10.56 Most Negative, 130%RTP to HlP Beta-Effective 0.0040 to 0.0070 0.0040 to 0.0075
Serial No. 08-0369 Docket No. 50-423 P
21 f 49 age 0
Current Design Analysis Values Values Normal Operation FNH 1.70 1.65 Normal Operation FQ(Z) 2.60 2.60 Shutdown Margin (%)
1.30 1.30
- a. The vessel average coolant temperature can decrease to 571.5 0 F during a coast down.
b.
Constant temperature program assumed during nominal depletion.
c.
Linear ramp from 70% to 100% power.
d.
MDC assumed for feedline break analysis (HFP ARO)
Section 2.8, Reactor Systems (continue)
Serial No. 08-0369 Docket No. 50-423 Page 22 of 49 Page Location Sentence Comment 116 Technical Evaluation "The fuel system in use at MPS3 While these fuel types have been used in the 1st paragraph includes Westinghouse 17X17fuel past, for the SPU cycle only RFA-2 assemblies matrices of the Standard (STD),
will be in use. A more accurate sentence would VANTAGE 5 Hybrid (V5H), Robust be The fuel systems that have been placed in Fuel Assembly (RFA) and RFA-2 service at MPS3 include Westinghouse designs."
17x17... "
117 5th paragraph "Similar to evaluation at the current For accuracy, this should read "... reference fuel license conditions, fuel rod system comprised entirely of RFAlRFA-2 fuel... "
performance for the uprated core In fact, the first SPU uprate cycle will consist of was evaluated using a reference fuel entirely RFA-2 assemblies.
system comprised entirely of RFA fuel... "
Serial No. 08-0369 Docket No. 50-423 Page 23 of 49 Page Location Sentence Comment 129 Last sentence "The licensee responded by letter DNC suggests changing the part: "proposed dated January 11, 2008, stating that CCW return flow design temperature of 145°F" the peak shell side design to: "proposed CCW return line 145°F operating temperature of the RHRS heat temperature."
exchanger is 200 of, which is well The LAR Table 2.8.4.4-2, "Proposed Cooldown over the newly proposed CCW Related Design Changes" explains that 145°F is return flow design temperature of a maximum operating temperature. The design 145 OF."
temperature (and stress analyzed temperature) for the subject CCW lines is increased from 150°F to 160°F.
130 Second paragraph, "The licensee stated that, for normal DNC suggests changing the sentence to read:
second sentence.
cooldown, this evolution will take "The licensee stated that for the normal place within 4-hours."
cooldown analysis, RHR entry occurs at 4-hours after a reactor trip."
The proposed change reflects the information provided in LAR Section 2.8.4.4.2.2.1.
130 Sixth paragraph, first "Because there was no change in DNC suggests changing the part: "Because sentence cooldown time for normal RHRS there is no change in cooldown time for normal cooldown, the NRC staff requested RHRS cooldown..." to: "Because there is no additional information regarding increase in cooldown time for normal RHRS differences in the normal and safety-cooldown..."
grade cold shutdown analyses from Per LAR Table 2.8.4.4-1, there is a decrease in pre-to post-uprate implementation."
normal cooldown times due to SPU and the associated SPU design changes.
Serial No. 08-0369 Docket No. 50-423 Page 24 of 49 Page Location Sentence Comment 133 1sr paragraph, fourth "However, the NRC staff does not From License Report page 2.8-125, Table sentence consider the reactor trip to be the 2.8.5.1.1.2.2-2, the sequence of events shows direct mitigation for this increase in the following:
feedwater flow event, especially Reactor Trip on Turbine Trip - rod motion since the minimum DNBR occurs initiated: 32.4 seconds only 0.3 seconds after rod motion is Minimum DNBR (1.88) reached: 32.5 seconds.
begun."
The difference is 0.1 seconds rather than 0.3 seconds.
133 Second paragraph, last "Therefore, the NRC staff finds these ONC fully concurs with the NRC that credit sentence results to be acceptable, despite the cannot be taken for the reactor trip from the crediting of reactor trip on IT in the Turbine Trip signal. As correctly discussed by analysis."
the NRC, once feedwater isolation occurs, the reactivity addition and power rise will be terminated and the RCS will begin to pressurize.
This will result in an increase in ONBR. The reactor trip following turbine trip was modeled to terminate the transient analysis at a point beyond the time of minimum DNBR. As such the reactor trip does not represent a credit in determining the minimum ONBR. A more accurate wording is : "Therefore, the NRC staff finds these results to be acceptable, despite the modeling of reactor trip on IT in the analysis."
Serial No. 08-0369 Docket No. 50-423 Page 25 of 49 Page Location Sentence Comment 135 4th paragraph in "However, the sequence of events Although the low steam line pressure setpoint is Technical Evaluation, (LR Table 2.8.5.1.2.2.1-1) lists low reached first, and it generates an SI signal second sentence steam line pressure as the first (which initiates feedwater isolation), the analysis safety injection signal."
does not credit the initiation of SI flow until after the SI signal from low pressurizer pressure is generated at 25.8 seconds. This timing is reflected in the following comment.
135 5In paragraph in "Although the safety injection system Per LR Table 2.8.5.1.2.2.1-1, the SI flow starts Technical Evaluation, is actuated early in the transient (0.5 at 72.8 seconds.
first sentence seconds) the minimum... "
136 5In paragraph "The limiting case... which is less Westinghouse considers the peak linear heat than the peak linear heat rate SAL of rate SAL as proprietary. The sentence can be re-worded as follows to eliminate the proprietary information. "The limiting case...which is less than the peak linear heat rate SAL given in LR Table 2.8.3-1.
Page 140 140 141 Location 1st paragraph and items 1 through 3 Conclusion Technical Evaluation 1st paragraph, last sentence Sentence "Analyses for the loss-of-non-emergency ac power event are not reported in the LR, since this event is bounded by the complete loss-of-flow event:
- 2. LR Section 2.8.5.2.1, with respect to RCS pressure and MSS pressure SALs, and by the LONF event with loss-of-non-emergency ac power;
- 3. LR Section 2.8.5.2.3, with respect to the capabilities of RCS natural circulation and the AFWS to remove stored and residual heat."
"The NRC staff has reviewed the licensee's analyses of the loss-of-non-emergency ac power to station auxiliaries event...."
"The AFW system starts automatically on SG low-low water level, following a safety injection signal, on LOOP, or on trip of all main feedwater pumps."
Serial No. 08-0369 Docket No. 50-423 Page 26 of 49 Comment The following rewording is recommended:
"Analyses for the loss-of-non-emergency ac power event are not reported in the LR, since this event is bounded by:
- 3. the LONF event with loss of non-emergency ac power, LR Section 2.8.5.2.3, with respect to the capabilities of RCS natural circulation and the AFWS to remove stored and residual heat."
This sentence is inconsistent with the first sentence on this page which states "Analyses for the loss-of-non-emergency ac power event are not reported in the LR.." The following rewording is recommended: "The NRC staff has reviewed the licensee's discussion of the loss-of-non-emergency... "
While this is an accurate statement, it is noted that only the two motor driven AFW pumps will start automatically following a safety injection signal, loss of offsite power or trip of the Main feedwater pumps. Only the SG low-low level signal will automatically start all pumps, the
Serial No. 08-0369 Docket No. 50-423 Page 27 of49 Page Location Sentence Comment turbine driven AFW pump as well as the two motor driven AFW pumps. It is recommended that this sentence be clarified to avoid confusion in the future.
140 Last paragraph, third "RCP heats of 20 MWt and 16 MWt A more accurate statement would be "RCP sentence were added to the cases with and heats of 20 MWt and 16 MWt were modeled in without offsite power."
the cases with and without offsite power" as is stated in License Report 2.8-183. Post-trip, for the offsite power cases, 20 MWt was included as heat source from the RCPs. Post-trip for the loss of offsite power cases, the Rep pumps are not a heat source since they will coastdown.
The initial power level was assumed to be 1.02 times 3666 MWt, which includes 16 MWt for RCP pump heat.
144 4tn full paragraph "There are FLB analyses results, This would be clearer if it is re-written as follows:
available for MPS3, for both "A comparison of the current licensing basis FLB assumptions. The licensing basis analysis and the SPU FLB analysis can be used FLB analyses, in which the to evaluate this assumption. The current pressurizer PORVs are assumed to licensing basis FLB analyses,..."
be available, predict that the Note:
minimum RCS subcooling would be While not documented in the Licensing Report, it 22°F."
should be noted that the SPU FLB analyses included a number of sensitivity studies including the impact of PORV availability. With credit taken for the PORVs, the margin to hot leg saturation was 15°F, demonstrating that the assumption of the PORVs being unavailable is conservative at SPU conditions.
Serial No. 08-0369 Docket No. 50-423 Page 28 of 49 Page Location Sentence Comment 149 2na full paragraph.
The licensee analyzed a postulated The RETRAN model for this event took into locked rotor, and noted that the account both the impact of the locked rotor and consequences of the locked rotor the sheared shaft. A locked rotor will result in accident are very similar to those of fastest reduction in forward flow through the an RCP shaft break."
loop. A sheared shaft will result in the largest reverse flow in the affected loop, consequently, resulting in the maximum reduction in the core flow once the flow in the affected loop reverses.
The RCP homologous pump curve in the forward flow direction was modified to take into account the increased resistance due to the locked rotor. The RCP homologous pump curve in the reverse flow direction was modified to take into account the reduced resistance due to a sheared shaft. This information provides additional basis for the statement in the SER.
154 Last sentence "Therefore, the licensee concluded, A more accurate wording for this sentence is as the generic analysis was not follows: "Therefore, the licensee concluded, applicable to MPS3, and the results reference to the generic analysis was not of the full power analysis presented required for MPS3, and the results of the full in the SPULR adequately power analysis presented in the SPULR demonstrated that the plant would adequately demonstrated that the plant would not overpressurize during an RWAP not overpressurize during an RWAP transient."
As stated in DNC letter 07-0450F dated April 24, 2008, DNC believes that it was unnecessary to reference the generic analysis. The generic analysis while bounding of MP3, was overly conservative.
Page 157 158 Location Conclusion 1st sentence Conclusion 1Sf sentence.
Sentence "The NRC staff has reviewed the licensee's analyses of control..."
"The NRC staff has reviewed the licensee's analyses of the inactive loop startup event... "
Serial No. 08-0369 Docket No. 50-423 Page 29 of 49 Comment This sentence is inconsistent with the discussion at the top of this page. As stated at the top of the page, generic dropped RCCA statepoints are used rather than licensee analyses. The following re-wording is recommended: "The NRC staff has reviewed the licensee's discussion of control rod misoperation... "
This is inconsistent with the last sentence in the Technical Evaluation where it states "The staff finds that this event need not be analyzed to implement the proposed power uprate." The following re-wording for this paragraph is recommended: The staff has reviewed the licensee's discussion of the inactive loop startup event and concludes that this event need not be analyzed to implement the proposed power uprate. Therefore, the NRC staff finds the proposed SPU acceptable with respect to the increase in core flow event."
Serial No. 08-0369 Docket No. 50-423 Page 30 of 49 Page Location Sentence Gomment 164 3ra full paragraph "The LR indicates that, "EGGS DNG concurs that this is an inaccurate injection is not credited for mitigating statement. However, DNG could not identify steam line or feedwater line breaks."
where this statement is made. The EGGS Actually, EGGS injection is credited modeling for the SLB and FLB SPU analyses is for mitigating steam line and FLBs."
as follows:
For Steam Line Break DNBR analyses, no credit is taken for boron addition from the charging pumps.
For Steam Line Break Mass and Energy releases, EGGS charging flow is modeled.
However, EGGS flow is assumed to be initiated on low pressurizer pressure and this is unaffected by the implementation of the P-19 permissive.
For FLB, charging flow EGGS is credited for assisting in maintaining hot leg saturation.
However, EGGS flow is assumed to be initiated on low pressurizer pressure and thus is unaffected by the implementation of the P-19 permissive.
164 Last sentence "The P-19 permissive essentially A more accurate statement is as follows:
eliminates the ability of the EGGS to "The P-19 permissive essentially eliminates the deliver flow to the RGS cold legs at ability of the EGGS to automatically deliver flow pressures above the low pressurizer to the RGS cold legs at pressures above the low pressure reactor setpoint."
pressurizer pressure reactor trip setpoint." The sentence as written could be interpreted to mean that the operators would not be able to perform feed-and-bleed. The P-19 permissive has no impact on the operator to manually initiate EGGS charging flow to the colds legs
Serial No. 08-0369 Docket No. 50-423 Page 31 of 49 Page Location Sentence Comment when required for beyond design basis events such as feed-and-bleed.
169 Top paragraph, last "Preventing SG overfill is necessary The most significant reason to demonstrate that sentence in order to prevent the failure of the the SG will not overfill is to assure that credit can steam lines."
be taken for iodine partitioning in the SG. If water is released from the SG directly to the environment, a significant increase in iodine releases will be predicted since a DF of 1 will be assumed. In fact, the MPS3 steam lines are designed for the loads of a water solid condition.
This information is contained in NU letter B12787 to the NRC dated January 22, 1988, "Millstone Nuclear Power Station, Unit NO.3 Steam Generator Tube Rupture (SGTR) - Plant Specific Information." The following re-wording is recommended: "Preventing SG overfill is necessary in order to prevent radioactive liquid releases to the environment and to prevent the failure of the steam lines."
170 4tn paragraph, fifth "These actions are defined in the A more accurate statement would be "These sentence Westinghouse Owner's Group actions are defined in the Westinghouse Emergency Response E-3 Owner's Group Emergency Response E-3 Guidelines."
Series of Guidelines."
The E-3 Series consists of E-3 Steam Generator Tube Rupture; ES-3.1 Post SGTR Cooldown Using Backfill; ES-3.2 Post-SGTR Cooldown Using Blowdown and ES-3.3 Post-SGTR Cooldown Using Steam Dump.
Section 2.9, Source Terms and Radiological Consequences Analyses Serial No. 08-0369 Docket No. 50-423 Page 32 of 49 Page Location Sentence Comment 190 Section 2.9.2, "The inventory, consisting of 66 isotopes Please change 66 isotopes to 72 isotopes. The second to the last at end of fuel cycle curie levels, formed basis for the number 72 is contained in SPU line the input for the RADTRAD-NAI dose Licensing Report, Table 2.9.2-1, (note that the evaluation code."
RADTRAD-NAI inventory file contains 72 isotopes, but Co-58 and Co-60 are not listed in the Table due to 0.0 curies). Also, RAI Response to question AADB-07-0107 in the letter dated 1/18/08 (Serial No. 07-0846),, page 4 of 106.
192 Section 2.9.2.1, "For conservatism the licensee The values given for sprayed/unsprayed region second paragraph, maintained the partitioning of the fission applies only to the period when quench spray second sentence products source between the alone is in operation. Different values are sprayed/unsprayed regions constant used for when quench spray and recirculation (0.4963/0.5037) during this period."
spray are in operation together and when recirculation spray is in operation alone.
Please change to: "The licensee maintained the partitioning of the fission product source between the sprayed/unsprayed regions consistent with the percentage of containment sprayed/unsprayed during these periods." The basis for this change is contained in DNC response to RAI question AADP-07-0107, DNC letter dated January 18, 2008 (Serial No.
07-0846), Attachment 3, page 4 of 106.
Serial No. 08-0369 Docket No. 50-423 Page 33 of 49 Page Location Sentence Comment 194 Section 2.9.2.1, last "As a result of these changes, the The EAB and LPZ doses dropped, only the CR paragraph before licensee calculated larger EAB, LPZ, and dose increased. Suggested change to "The Containment CR doses and therefore the changes are licensee calculated a larger CR dose due to Leakage, the 1st shown to be conservative."
the changes shown and, although the EAB and sentence LPZ doses decreased slightly due to the new source term, the changes are shown to be conservative." The basis for this proposed change is contained in the license amendment request, attachment 5, Licensing Report, Table 2.9.2-5.
194 Section 2.9.2.1, "The containment leak rate La, and the Please delete "and the bypass leak rate."
Containment bypass leak rate are reduced by one-half Bypass leak rate is included in La. The basis Leakage, fifth at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for offsite calculations, and at for this proposed change is contained in the sentence 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for CR dose calculations."
license amendment request, attachment 5, Licensing Report, Section 2.9.2.1.2.8.
196 Section 2.9.2.1, "The licensee calculated a DF of 450 for Please change "450" to "330". The basis for Releases from the the release of iodines from the RWST as this proposed change is contained in DNC RWSTdueto a result of back leakage."
response to RAI question AADP-07-0107, ECCS Back DNC letter dated January 18, 2008 (Serial No.
Leakage, fourth 07-0846), Attachment 2, page 40 of 95.
paragraph, fourth sentence
Serial No. 08-0369 Docket No. 50-423 Page 34 of 49 Page Location Sentence Comment 201 Section 2.9.2.3, "The 35.75 hour8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> time period includes 24 Please replace with: "The 35.75 hour8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> time third paragraph, hours for the primary system to cool period includes 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the primary third and fourth sufficiently to allow an alignment to the system to cool sufficiently to allow an sentences RHRS. As a result of the assumption of a alignment to the RHRS, an additional 11.75 concurrent LOOP, the 35.75 hour8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> period hours of steaming required to reduce the includes an additional 11.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> of system heat load to the point where the RHRS steaming required to reduce the system can remove all the decay heat using only heat load to the point where the RHRS safety grade equipment, and the impact of the can remove all the decay heat using only assumption of a concurrent LOOP."
safety grade equipment."
The basis for the proposed change is as follows: The discussion from ONe letter dated January 18, 2008 (Serial No. 07-0846),, Page 53 of 95 - 2nd paragraph -
The 11.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> is not the result of the concurrent LOOP assumption, but the 35.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> includes the impact of the concurrent LOOP assumption 203 Section 2.9.2.3, "The licensee assumed the same CR Please replace with: "The licensee assumed second paragraph, ventilation timing sequence as was used the CR ventilation timing sequence indicated in last sentence for the LOCA, which does not credit Table 7, which does not credit automatic automatic initiation of the CREVS."
initiation of the CREVS."
The basis for the proposed change is as follows: The CR ventilation timing sequences for the LOCA and the SGTR are very similar.
However, isolation of the Control Room for the SGTR is the result of a radiation monitor (RM) responding to the plume and automatically initiating the CBI signal. The RM and CBI siQnals each take about 5 seconds. As a
Serial No. 08-0369 Docket No. 50-423 Page 35 of 49 Page Location Sentence Comment result of relying on the RM signal there is a 10 second period after the plume reaches the control room intake before the control room is isolated. During the LOCA, the CBI signal is generated and the control room is isolated in 5 seconds which is before the plume reaches the control room intake. Therefore, as a result of the RM signal delay the manual initiation of CREVS occurs 5 seconds later during the SGTR than it does in the LOCA.
In Table 3.3-4 of DNC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Page 62 of 95 - Dominion indicated that the Control Room Ventilation Timing was equal to the current licensing basis FHA due to isolation based on RM not the LOCA and isolation based on SI.
205 Section 2.9.2.4, 1st "In order to ensure proper accounting of Dominion did not use a TS limit of 100/Ebar.
paragraph, 2nd gross gamma, iodine and noble gas The gross gamma activity is based on the sentence releases from the unaffected SGs, the equivalent level of fuel damage as 1 JlCi/gm licensee evaluated all the significant DE 1-131 (0.29% FF).
nuclide transport models for the MSLB The basis for the proposed change is accident. The licensee evaluated the contained in RAI Response to question AADB-release of the gross gamma activity from 07-0107, DNC letter dated January 18,2008 the primary coolant, at the TS limit of (Serial No. 07-0846), Attachment 2, page 54 of 100/E-Bar, leaking into the unaffected SG 95.
volume at a primary-to-secondary leak Please consider replacing the sentence with:
rate of 0.65 gpm. Radionuclides initially in "In order to ensure proper accounting of gross the SG liquid and those entering the SG gamma, iodine and noble gas releases from
Serial No, 08-0369 Docket No. 50-423 Page 36 of 49 Sentence
~-
~-~~
Comment Page Location from the primary-to-secondary leakage the unaffected SGs, the licensee evaluated all flow are released as a result of secondary the significant nuclide transport models for the liquid steaming. An assumed PC of 100 MSLB accident. The licensee evaluated the results in 1 percent of the particulates and release of the gross gamma activity from the iodines in the SG bulk liquid being primary coolant, at a level of fuel failure released to the environment at the consistent with 1 uCi/gm DE 1-131, leaking into steaming rate. Radionuclides initially in the unaffected SG volume at a primary-to-the steam space do not provide any secondary leak rate of 0.65 gpm.
significant dose contribution. The Radionuclides initially in the SG liquid and transport to the environment of noble those entering the SG from the primary-to-gases from the primary coolant is secondary leakage flow are released as a assumed to occur without any mitigation result of secondary liquid steaming. An or holdup."
assumed PC of 100 results in 1 percent of the particulates and iodines in the SG bulk liquid being released to the environment at the steaming rate. Radionuclides initially in the steam space do not provide any significant dose contribution. The transport to the environment of noble gases from the primary coolant is assumed to occur without any mitigation or holdup."
206 Section 2.9.2.5, "The licensee assumed the same CR Please consider replacing with: "The licensee third paragraph, ventilation timing sequence as was used assumed the same CR ventilation timing last sentence for the LOCA, which does not credit sequence as was used for the SGTR as shown automatic initiation of the CREVS."
in Table 7, which does not credit automatic initiation of the CREVS."
As with the SGTR, the CR ventilation timing sequence for the LOCA and the LRA are very similar, but the SGTR and LRA CR ventilation timing sequences are identical. Isolation of the
Serial No. 08-0369 Docket No. 50-423 Page 37 of 49
~~~-~~
Page Location Sentence Comment Control Room for the LRA is the result of a radiation monitor (RM) responding to the plume and automatically initiating the CBI signal. The RM and CBI signals each take about 5 seconds. As a result of relying on the RM signal there is a 10 second period after the plume reaches the control room intake before the control room is isolated. During the LOCA, the CBI signal is generated and the control room is isolated in 5 seconds which is before the plume reaches the control room intake.
Therefore, as a result of the RM signal delay the manual initiation of CREVS occurs 5 seconds later during the LRA than it does in the LOCA.
The basis for the proposed change is contained in RAI Response to question AADB-07-0107, ONe letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Table 3.3-4, Page 81 of 95. Just like the SGTR, Dominion indicated that the Control Room Ventilation Timing was equal to the current licensing basis FHA due to isolation based on RM not the LOCA and isolation based on SI.
Serial No, 08-0369 Docket No. 50-423 Page 38 of 49 Page Location Sentence Comment 207 Section 2.9.2.6,
The licensee has determined that Please change "determined' to "conservatively second paragraph, containment sprays will not initiate due to assumed." This assumption did not change first sentence an REA and, as a result, the licensee did from the current licensing bases and is listed not evaluate dose contributions from as such in the license amendment request, ECCS leakage and RWST back leakage attachment 5, licensing report, page 2.9-16, as in the LOCA analysis."
section 2.9.2.1.6.3 and RAI Response to question MDB-07-0107, ONC letter dated January 18, 2008 (Serial No. 07-0846),, page 85 of 95, No.2.
Further, it should be noted that ECCS leakage and RWST back leakage are dose contributors when containment recirculation is initiated.
Containment recirculation is not necessary for the rod ejection event. This is unrelated to spray operation.
211 Control Room "The period when the CREVS is Please replace with: "The period when the Habitability, fifth operational is referred to as the CR CREVS is providing pressurization flow is paragraph, second positive pressure period."
referred to as the CR positive pressure period."
sentence As stated in the first sentence CREVS is pressurization and recirculation. After 30 minutes following a FHA, only recirculation flow is operating and the control room is neutral until 101 minutes.
The basis for the proposed change is contained in RAI Response to question MOB-07-0107, ONC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Table 3.2-1, page 48 of 95
Serial No. 08-0369 Docket No. 50-423 Page 39 of 49 Page Location Sentence Comment 215 Table 2, (pg 1 of 2)
Millstone Stack\\l)
The Millstone Stack X/Q are listed for 0 - 2 0-2 hrs and 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. They should be 0 - 4 and 4 -
2-8 hours.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The basis for the proposed change is contained in RAI Response to question AADB-07-0107, ONC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, page 11 of 95.
223 Table 8, The third Primary-to-secondary leak rate TS limit 1
Please replace with:
item in the table gpm (to unaffected SGs)
Primary-to-secondary leak rate TS limit 150 gpd to any 1 steam generator The basis for the proposed change is contained in RAI Response to question AAOB-07-0107, ONC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, page 68 of 95.
Serial No 08-0369 Docket No. 50-423 Page 40 of 49
"~-_ *...__._-------
Page Location Sentence Comment 223 Table 8, last line "CR plume and CR filter shine dose The MSLB does not use LOCA values, please conservatively set at values from the replace with: "CR plume and CR filter shine LOCA analyses" dose are 4.37E-03 rem and 0.26 rem, respectively."
The basis for the proposed change is contained in RAI Response to question AAOB-07-0107, ONC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 3, pg 67 of 106.
219 Table 5, (page 2 of "CR ventilation timing for the LOCA, Please delete: "for the LOCA, SGTR, MSLB, 2)
SGTR, MSLB, LRA and the REA" LRA and the REA."
Each table has its own ventilation timing assumptions.
T= 1 minute, 5 seconds T=1 hour, 41 min, 5 sec (1.685 hours0.00793 days <br />0.19 hours <br />0.00113 weeks <br />2.606425e-4 months <br />)
On Page 219, Table 5, (page2 of 2):
Please replace the following Section:
CR ventilation timing for the LOCA, SGTR, MSLB, LRA and the REA:
T= 0 seconds Normal CR unfiltered intake flow: 1595 cfm T= 5 seconds CBI signal generated T= 10 seconds CR isolates on radiation monitor signal Intake flow: 0 cfm; neutral condition Assumed unfiltered inleakage: 350 cfm delay for CREPS response (Not credited)
Assumed unfiltered inleakage: 350 cfm CREVS filtered intake flow: 230 cfm Assumed unfiltered inleakage: 100 cfm CREVS filtered recirculation flow: 666 cfm With:
CR ventilation timing for the LOCA, SGTR, MSLB, LRA and the REA:
T=O seconds CR isolated on SI signal Intake flow: 0 cfm; neutral condition Assumed unfiltered inleakage: 350 cfm Serial No. 08-0369 Docket No. 50-423 Page 41 of 49 T=1 minute T=1 hour, 41 min (1.683 hours0.00791 days <br />0.19 hours <br />0.00113 weeks <br />2.598815e-4 months <br />) delay for CREPS response (Not credited)
Assumed unfiltered inleakage: 350 cfm CREVS filtered intake flow: 230 cfm Assumed unfiltered inleakage: 100 cfm CREVS filtered recirculation flow: 666 cfm
Serial No. 08-0369 Docket No. 50-423 Page 42 of49 The basis for the proposed change is contained in RAI Response to question AADB-07-0107, DNC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 3, Page 5 of 106; Also stated correctly on page 212, first full paragraph of the draft SER.
On Page 221, Table 7 (Page 1 of 2)
Please replace the following Section:
Table 7 (Page 1 of 2)
MPS3 SPU Data and Assumptions for the SGTR Accident Primary-to-secondary leak rate TS limit LOOP RCS TS iodine limit for normal operation Gross gamma Iodine With:
1 gpm (to unaffected SGs)
Coincident with release 100/ E-Bar 1.0 J,JCi/gm DEI Table 7 (Page 1 of 2)
MPS3 SPU Data and Assumptions for the SGTR Accident Primary-to-secondary leak rate TS limit LOOP RCS TS iodine limit for normal operation RCS Gross Gamma activity 1 gpm (to unaffected SGs)
Coincident with release 1.0 J,JCi/gm DEI Equivalent to fuel failure associated with DEI limit Basis:
The SGTR does not use the 100/E-bar LCO. Instead, as indicated in RAI Response to question AADB-07-0107, DNC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Page 54 of 95, the gross gamma activity was based
Serial No. 08-0369 Docket No. 50-423 Page 43 of49 upon the more operationally limiting Dose Equivalent 1-131 LCO. The discussion from Letter Serial No. 07-0846, Page 54 of 95 is repeated below.
"Primary side iodine and gross gamma source concentrations are based on the more limiting Technical Specification limit of 1.0 ~Ci/gm Dose Equivalent (DEQ) 1-131. The failed fuel equivalent associated with the iodine Technical Specification limit is 0.29%. The Technical Specification concentrations of the non-iodine isotopes at the 1OO/E-bar limit result in greater than 1 % failed fuel. Therefore the more limiting failed fuel percentage associated with the specific activity limit for the iodines is used for all primary coolant isotopes."
On Page 223, Table 8, 4th item on the table:
Please replace the following Section:
Table 8 - MPS3 SPU Data and Assumptions for the MSLB Accident RCS volume RCS mass Primary-to-secondary leak rate TS limit RCS TS limit for normal operation Gross gamma Iodine With:
11,750 ft3 5.216E+05 Ibm 1 gpm (to unaffected SGs) 1001 E-Bar 1.0 ~Ci/gm DEI Table 8 - MPS3 SPU Data and Assumptions for the MSLB Accident RCS volume RCS mass Primary-to-secondary leak rate TS limit RCS TS iodine limit for normal operation RCS Gross Gamma activity 11,750 ft3 5.216E+05Ibm 1 gpm (to unaffected SGs) 1.0 ~Ci/gm DEI Equivalent to fuel failure associated with DEI limit
Serial No. 08-0369 Docket No. 50-423 Page 44 of 49 Basis:
Dominion did not use a technical specifications limit of 100/Ebar. The gross gamma activity is based on the equivalent level of fuel damage as 1 /lCi/gm DE 1-131 (0.29% FF). The basis for the proposed change is contained in RAI Response to question AADB-07-0107, DNC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Page 54 of 95 (See comment 28)
Serial No. 08-0369 Docket No. 50-423 Page 45 of 49 Page Location Sentence Comment 222 Table 7 (Page 2 of "CR ventilation timing for the SGTR:
Please delete "Same as the LOCA". The CR 2)
Same as the LOCA" ventilation timing sequences for the LOCA and the SGTR are very similar. However, isolation of the Control Room for the SGTR is the result of a radiation monitor (RM) responding to the plume and automatically initiating the CBI signal. The RM and CBI signals each take about 5 seconds. As a result of relying on the RM signal there is a 10 second period after the plume reaches the control room intake before the control room is isolated. During the LOCA, the CBI signal is generated and the control room is isolated in 5 seconds which is before the plume reaches the control room intake.
Therefore, as a result of the RM signal delay the manual initiation of CREVS occurs 5 seconds later during the SGTR than it does in the LOCA. The basis for the proposed change is contained in RAI Response to question AADB-07-0107, DNC letter dated January 18, 2008 (Serial No. 07-0846), Attachment 2, Page 62 of 95. DNC indicated that the Control Room Ventilation Timing was equal to the current licensing basis FHA due to isolation based on RM not the LOCA and isolation based on SI.
223 Table 8 (Page 1 of "CR ventilation timing for the MSLB:
Please delete "Same as the LOCA". The basis 1)
Same as the LOCA" for the proposed change is contained in RAI Response to question AADB-07-0107, DNC letter dated January 18, 2008 (Serial No. 07-
Serial No. 08-0369 Docket No. 50-423 Page 46 of49 0846), Attachment 2, Page 72 of 95. DNC indicated that the Control Room Ventilation Timing was equal to the current licensing basis FHA.
224 Table 9 (Page 1 of "CR ventilation timing Same as the Please replace with: "CR ventilation timing 1)
The CR ventilation timing sequences for the LOCA and the SGTR are very similar.
However, isolation of the Control Room for the SGTR is the result of a radiation monitor (RM) responding to the plume and automatically initiating the CBI signal. The RM and CBI signals each take about 5 seconds. As a result of relying on the RM signal there is a 10 second period after the plume reaches the control room intake before the control room is isolated. During the LOCA, the CSI signal is generated and the control room is isolated in 5 seconds which is before the plume reaches the control room intake. Therefore, as a result of the RM signal delay the manual initiation of CREVS occurs 5 seconds later during the SGTR than it does in the LOCA.
Section 2.10, Human Performance Serial No. 08-0369 Docket No. 50-423 Page 47 of 49 Page Location Sentence Comment 227 Changes in Emergency "There will be an additional manual action in Please replace "psig" with "psia". The unit and Abnormal the EOPs to verify ECCS flow when RCS "psig" was incorrectly used on page 2.11-3 of Operating Procedures, pressure is less than 1900 psig."
the SPU licensing report, section 2.11.1, second bullet, First Human Factors. However, the correct units sentence.
were used in reference to same parameter on pages 2.4-29, 2.4-30, 2.4-39 and 2.4-40 of the SPU licensing report.
Section 3.2, Technical Specifications Serial No. 08-0369 Docket No. 50-423 Page 48 of 49 Page Location Sentence Comment 234 Last sentence "This surveillance needs to be Please replace with: "This surveillance needs performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after to be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching reaching 90 percent of the RTP."
90 percent of the RTP following each fuel loading." This change will make the sentence consistent with the revised surveillance requirement.
235 First full sentence
'Thus, to comply with the uncertainty Please replace with: "Thus, to comply with the analysis assumptions and to meet the uncertainty analysis assumptions and to meet heat balance requirements, it is the heat balance requirements, it is necessary necessary to require the RCS flow to require the RCS flow rate to be determined rate to be determined at not less than at not less than 90 percent of the RTP and 90 percent of the RTP and within 24 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 90 percent of hours after reaching 90 percent of the the RTP following each fuel loading." This RTP."
change will make the sentence consistent with the revised surveillance requirement.
235 Second paragraph, "As described above, the Please replace with: "As described above, the third sentence instrumentation calibrations required instrumentation calibrations required for heat for heat balance are calibrated once balance are calibrated once per 18 months per 18 months and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and within 7 days prior to performance of the after reaching 90 percent of the RTP."
calorimetric flow measurement." This sentence is describing the current surveillance requirement.
236 TS 3/4.3.2 Engineered "Pressurizer Water Level-Low" Please replace the word "Water Level" with Safety Features "Pressure" to read: "Pressurizer Pressure-Actuation System Low". Please refer to Technical Specifications Instrumentation, Table 2.2-1, Functional Unit 9.
second paragraph, third line.
Serial No. 08-0369 Docket No. 50-423 Page 49 of 49 PaQe Location Sentence Comment 239 Additional Changes, liThe licensee's application dated July Please delete this paragraph. The licensing Item 1, Safety Grade 13, 2007, provided revised TS Bases bases change associated with the proposed Cold Shutdown, pages to be implemented with the change is contained in FSAR Section second paragraph.
associated TS changes. The NRC 5.4.7.2.3.5 as addressed in MPS3 SPU license staff reviewed this change in Section amendment request, attachment 1, section 2.5 of the SE and did not have any 3.1, page 10, last paragraph.
concerns. The NRC staff notes that the MPS3 TS Bases Control Program is the appropriate process for updating the TS Bases."
240 Additional Changes, liThe licensee's application dated July Please delete this paragraph. The licensing Item 2, BTP CMEB 13,2007, provided revised TS Bases bases change associated with the proposed 9.5.1 Sections 5.c.3 pages to be implemented with the change is contained in a DNC document and 5.c.5-Fire associated TS changes. The NRC entitled: "MP3 Branch Technical Position 9.5-1 shutdown strategy for staff reviewed this change in Section Compliance Report." This document will be long-term steam 2.5 of the SE and did not have any updated upon NRC approval of this item to generator inventory concerns. The NRC staff notes that credit the use of domestic water, make-up, second the MPS3 TS Bases Control Program demineralized water or fire water to make-up paragraph.
is the appropriate process for updating the DWST and CST.
the TS Bases."