ML062120472
| ML062120472 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 07/03/2006 |
| From: | Pace P L Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Lu S, NRR/DSS/SSIB, 415-2869 | |
| Shared Package | |
| ML062120461 | List: |
| References | |
| GL-04-002, TAC MC4730 | |
| Download: ML062120472 (47) | |
Text
PROPRIETARY INFORMATION ENCLOSED UNDER 10 CFR 2.390(b)(4)
July 3, 2006
U. S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, D.C. 20555-0001
Gentlemen: In the Matter of ) Docket No. 50-390 Tennessee Valley Authority )
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - GENERIC LETTER 2004 REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NUCLEAR
REGULATORY COMMISSION STAFF AUDIT ON THE CONTAINMENT SUMP
MODIFICATIONS (TAC NO. MC4730)
The purpose of this letter is to respond to NRC's request for additional information (RAI) dated May 10, 2006 concerning the
subject Staff audit of the containment sump modifications.
TVA coordinated an extension of this response with NRC Project
Manager to July 5, 2006.
TVA's responses to NRC's questions are provided in Enclosure
- 1. The documents requested by NRC's RAI are provided on the
enclosed Compact Discs (CD) two per set. A list of documents
on the CDs is provided in the Attachment to Enclosure 1.
Calculation FSDA-C-597, "RHR Pump NPSH," in response to
Question 1 under Net Positive Suction Head/Loss-of-Coolant Accident and the "WBN ECCS Analysis Report" in response to Question 1 under Downstream Effects (Components) on the CD contain information proprietary to Westinghouse Electric
Corporation for which withholding is being requested.
Westinghouse is providing these documents for use by the NRC
staff in its audit activities and requests that these
documents be considered proprietary in their entirety. As
such, a non-proprietary version will not be issued.
PROPRIETARY INFORMATION ENCLOSED UNDER 10 CFR 2.390(b)(4)
U.S. Nuclear Regulatory Commission Page 2
July 3, 2006
The proprietary information for which withholding is being
requested is further identified in Affidavit CAW-06-2163 and
CAW-06-2169 signed by the owner of the proprietary
information, Westinghouse Electric Company LLC. The affidavit
sets forth the basis on which the information may be withheld
from public disclosure by the Commission and addresses with
specificity the considerations listed in 10 CFR 2.390(b)(4).
contains Westinghouse authorization letters,
CAW-06-2163 and CAW-06-2169, accompanying affidavits, Proprietary Information Notices, and Copyright Notices.
Correspondence with respect to the proprietary aspects of the
application for withholding or the Westinghouse affidavits
should reference CAW-06-2163 or CAW-06-2169 and should be
addressed to B.F. Maurer, Acting Manager, Regulatory
Compliance and Plant Licensing, or J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse
Electric Company LLC, P. O. Box 355, Pittsburgh, Pennsylvania
15230-0355.
The remaining open items to respond to NRC's audit request for
additional information are being tracked as part of the
previous commitment to provide a supplemental response. If
you have any questions concerning this matter, please call
P. L. Pace at (423) 365-1824.
I declare under penalty of perjury that the foregoing is true
and correct. Executed on this 30th day of June 2006.
Sincerely,
P. L. Pace
Manager, Site Licensing
and Industrial Affairs
Enclosures
cc See page 3 PROPRIETARY INFORMATION ENCLOSED UNDER 10 CFR 2.390(b)(4)
U.S. Nuclear Regulatory Commission Page 3
July 3, 2006
Enclosures
cc (Enclosures): NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381
Mr. D. V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-1 Break Selection and Zone of Influence Analysis QUESTION 1 Tennessee Valley Authority (TVA, the licensee) stated that
because the quantity of reflective metallic insulation is not a
significant contributor to head loss, and the quantity of fibrous
material, Min-K, would remain relatively unchanged for each
break, the bounding case for each loop is the reactor coolant
system break which would destroy the most coatings. The licensee
indicated that a thorough analysis showed that a break in each of
the crossover legs near the steam generator nozzle yielded the
most coating debris due to the size of the zone of influence (ZOI) applied in the analyses. The Nuclear Regulatory Commission (NRC) staff (the staff) determined that such an analysis was not
clearly documented in the calculations and information provided
for the staff's audit. Please provide the referenced analysis to
verify that the limiting break is at the base of the steam
generator.
RESPONSE As a result of questions raised during the audit, ALION has
revised and expanded the debris generation calculation. The
revision to the debris generation calculation (Revision 2) no
longer makes reference to undocumented analyses for the paint
calculations. Since the ZOIs used in the debris generation
analysis are large, moving locations along the primary loop
piping would not have a significant impact on the debris
quantities generated. The break selection considered all debris
sources. The silicon coatings protecting the carbon steel shell
of the steam generators were the reason to select the steam
generator nozzle as well as the large amount of reflective metal
insulation (RMI) on the steam generators. The new steam
generators that are being installed in the Fall 2006, will not
have a coating on the shell. The steam generators continue to be
the largest source of RMI debris for large breaks. Also, since
the crossover leg is larger than the hot and cold legs, selecting
a break on the crossover leg piping is conservative because the
ZOI is larger.
The revised analysis results in revised debris quantities
projected for WBN. Some of the fiber quantities due to min-K and
3M fire wrap have increased with respect to that tested in WBN's
strainer test. WBN is looking at several options to reduce these
quantities to within the tested configuration. These include:
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-2 credit for additional jet shielding due to robust barriers and large structures, material testing under jet impingement loading
to reduce the ZOI for encapsulated fiber, removal of material, and/or sump strainer re-testing. The total fiber quantities
still remain low and with WBN's large strainer area, TVA is
confident the WBN strainer design will have a low head loss. The
final debris calculation will be provided as part of an update of
the remaining open items.
QUESTION 2 As discussed in Sections 3.1 - 3.4 of Watts Bar calculation
ALION-CAL-TVA-2739-03, the licensee credits the reactor annulus
and refueling canal as robust barriers in the analysis. As
stated, the licensee's analysis showing that a break in each of
the crossover legs near the steam generator nozzle yielded the
most coating debris was not clearly documented in the
calculations and information provided for the staff's audit.
Therefore, Watts Bar calculation ALION-CAL-TVA-2739-03 does not
clearly show the extent to which the licensee credited truncation
due to robust barriers. Using the response to question 1 above, please show the extent to which truncation is credited.
RESPONSE The revised debris generation calculation (Revision 2) now shows the shielding that is currently credited and includes appendices
to clearly document which of the line items from the insulation
spreadsheet that were included as debris for each break location.
The revised analysis will be provided to NRC in the supplemental
report. Follow-up work may be required as described in the
response to Question 1 above.
QUESTION 3 Steam line breaks in the debris generation calculation are ruled
out because recirculation is not required for cooling the core
following a steam line break. However, recirculation using spray
flow for environmental qualification of equipment is required.
Please explain why this scenario was not analyzed.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-3 RESPONSE A main steam line break (MSLB) in the lower compartment would result in a smaller ZOI volume compared to the reactor coolant
system (RCS) line break since the main steam pressure is less
than half of the RCS pressure. A loss-of-coolant accident (LOCA)
was considered to be bounding to a MSLB since ECCS recirculation
is not required for decay heat removal following a postulated
MLSB. Recirculation using spray flow for environmental
qualification of equipment is required long term following a
MSLB. The ice condenser ice melt depletion is bounded by the
LOCA and occurs later in time due to less energy release for the
MSLB. Eventually the ice is depleted, even for the MSLB, and
containment spray in conjunction with air flow from the lower
compartment coolers is used to maintain the containment
temperature in the long term. However, operators are not
required to restart the lower compartment cooler fans to
recirculate air throughout the lower compartment and the dead-
ended spaces to prevent hot spots from developing for at least
1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s
after the event. In addition, the containment spray is
only required to remove ambient heat loss from the RCS. Periodic
use of one train of spray is needed. Therefore, there would be
less flow to transport debris, less debris to transport, and
intermittent flow to move the debris. Thus, it was determined
that a MSLB was bounded by a Large Break LOCA and was not
required to be analyzed.
Debris Generation
QUESTION 1 Please provide the complete walk-down report, "Report on Watts
Bar Unit 1 Containment Building Walkdowns for Emergency Sump
Strainer Issues," TVAW001-RPT-001, Revision 0.
RESPONSE A compact disc (CD) is enclosed with the requested information in
electronic format.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-4 Chemical Effects QUESTION 1 Please provide the amounts of various Watts Bar containment materials (I) submerged and (ii) in the containment spray zone
for the following materials: aluminum, zinc (from galvanized
steel and inorganic zinc (IOZ) coatings), copper, carbon steel, and uncoated concrete. These amounts should include any
scaffolding material or metallic-based paints (e.g., aluminum-
based paints used on pressure vessels).
RESPONSE The material amounts requested were provided in TVA's response dated April 11, 2006 in response to NRC's Request for Additional
Information dated February 10, 2006 under Plant Materials, Question 2. The quantities provided included scaffolding stored
inside the crane wall that would be subject to spray or
submergence. WBN controls this material to minimize quantities.
There is no other metallic based paint other than those listed in
the April 11, 2006 response.
QUESTION 2 Provide a discussion concerning the post loss-of-coolant accident (LOCA) containment pool pH, including the range of pH values
possible. The values discussed by the licensee at the audit
meeting were more refined than the licensee's response to the NRC
Generic Letter (GL) 2004-02. Please clarify.
RESPONSE The expected sump pH is 7.8 to 8.2 for a LOCA at any time during
the fuel cycle. The sump pH range includes conditions for the
beginning and end of core life, the minimum and maximum
quantities of boron and buffering agent in the RCS, the
accumulators, the refueling water storage tank (RWST), and in the
ice condenser. The range also includes the maximum and minimum
water and ice volumes. The temperature variation of the RWST and
accumulators was included in developing this range.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-5 QUESTION 3 If possible, provide the containment pool temperatures as a
function of time during the emergency core cooling system (ECCS)
mission time for the limiting combination of conditions that
would produce (i) the highest pool temperatures with time, and (ii) the lowest pool temperatures with time.
RESPONSE
SUMP WATER TEMPERATURE
The figure above shows the sump temperature for the limiting
large break LOCA. The analysis is based on one train of ECCS and
containment spray which minimizes containment heat removal. The
analysis also assumed an ultimate heat sink temperature of 88
degrees Fahrenheit (F) which is higher than the current technical
specification limit of 85 degrees F. It also assumes that river
stays at this temperature for the entire 30 day period. This is
a very conservative assumption. It should be noted that TVA has
submitted a Proposed License Amendment Request (WBN-TS-06-09)
dated May 8, 2006, to increase the design basis ultimate heat
sink temperature to 88 degrees F. The RWST temperature was 60 80 100 120 140 160 180 2000.005.0010.0015.0020.0025.0030.00DaysTemperature (F)
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-6 assumed at the technical specification maximum of 105 degrees F.
The amount of ice in the ice condenser was assumed to be at the
minimum safety limit value. It should also be noted that the
maximum pH used to evaluate chemical effects was based on the
maximum amount of ice in the ice bed. Using this ice mass in the
containment analysis would have resulted in a lower sump
temperature and a higher water level for net positive suction
head (NPSH).
A similar analysis for minimum sump temperature has not been
performed. A sensitivity study on the amounts of chemical
precipitants was performed assuming that sump temperature was
lowered considerably. This sensitivity study showed that the
amount of corrosion products produced was lower than in the high
temperature case. As such, the high temperature case is limiting
and there is not a need for a detailed formal analysis of minimum
sump temperature.
QUESTION 4 Provide the Watts Bar plant-specific chemical effects analysis.
Indicate if any more chemical effects related testing is planned.
RESPONSE Chemical effects were evaluated using a correlation developed by
Westinghouse from separate effects precipitation test data (WCAP-16530-NP, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191) and considering the results of the integrated chemical effects tests (ICET). The
evaluation using the WCAP correlation showed a total of 10 milli-
grams per liter (mg/l) for the precipitants based on the total
weight of the precipitants. The total weight of precipitants for
the base case was less than 45 pounds. The precipitants
predicted by the Westinghouse correlations were composed
principally of NaAlSi 3 O 8 (aluminum silicate) with a small amount of AlOOH (aluminum oxide hydroxide). This result was obtained
using the sump temperature profile discussed in response to
Question 3 above and the maximum sump pH was reached at about 30
minutes into the event. The maximum pH will not occur until ice
bed melt out at just over an hour into the event.
Temperature and pH sensitivities were run using the Westinghouse
correlations. Lower temperatures and lower pH result in lower
concentrations and total quantities. A case was run with a ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-7 maximum sump pH of 7.8 corresponding to a minimum ice case with somewhat lower long term sump temperatures. The amount of
precipitant was just over 23 pounds. Using the same temperature
profile with a maximum sump pH of 8.2, results in a total
precipitant weight of less than 28 pounds. Aluminum silicate and
aluminum oxide hydroxide were the only precipitants in all cases.
ICET 5 is the test most representative of the WBN environment of
the ICET series of tests. The boron concentration in the test is
2800 parts per million (ppm) versus a maximum WBN concentration
of 3300 ppm. The buffer is sodium tetraborate contained in the
ice of the ice condenser. A concentration for the sodium
tetraborate is not calculated. The solution used to form the ice
is sampled and has to have a boron concentration of 1800 to 2000
ppm and the pH is required to be between 9.0 and 9.5. The ICET 5
test pH range is 8.0 to 8.5 and the WBN sump pH is between 7.8
and 8.2 as discussed in the response to Question 2 above. The
amount of aluminum evaluated in ICET 5 is much higher than is
present in the plant. Since aluminum is the predominant
precipitant, this difference is significant. The other
significant difference is the ICET temperature is much higher
than is present in the plant. ICET 5 showed concentrations of
dissolved aluminum of 55 milligrams per liter (mg/l) and calcium
of 35 mg/l.
Given the very low quantities of chemical precipitants, TVA does
not plan further chemical testing.
QUESTION 5 During the integrated chemical effects testing (ICET), in certain
chemical environments such as sodium tetraborate, precipitates
formed as the solution cooled from the 140 o F test temperature.
These products could interact with other downstream debris to
cause clogging in narrow passages of downstream components such
as valves and pump internals, or affect internal surfaces of heat
exchangers or the reactor vessel. Describe your evaluation of
potential downstream effects related to interaction with chemical
products and the criteria used to determine that performance of
downstream components is acceptable for your plant-specific
chemical products and debris combination.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-8 RESPONSE The chemical analyses showed that the quantity of precipitants formed would be less than 45 pounds. These formed over the
course of the 30 day mission time not instantaneously. The
precipitants would initially form as small particles. If the
precipitants were to agglomerate, it would be more likely to
occur in the general sump pool as opposed to in piping where the
flow rates are much higher. Larger particles would be more
likely to settle out and be removed as a potential problem for
down stream effects. TVA has included chemical precipitants in
the evaluated particulate load for downstream wear and plugging.
The total quantity of chemical precipitants is so small that
there would be no noticeable effect on heat transfer in the RHR
and containment spray heat exchangers. The chemical load is less
than two percent of the total particulate load and as such does
not appreciably affect wear. The strainer hole size was selected
to be the smallest opening in the ECCS flow path when fuel bottom
nozzle changes are complete. The size of the strainer holes was
chosen to preclude plugging.
QUESTION 6 If all the coatings are assumed to fail, justify why this large
additional debris loading would not increase the analyzed amount
of chemical effects, or add another unanalyzed chemical product.
RESPONSE The principal coating materials in the containment are inorganic
zinc and phenolic topcoat. The chemical testing has established
that there are no noteworthy precipitants associated with the
zinc. Amounts of zinc in excess of the amount present in WBN
were considered in the ICET tests. The cured phenolic is not
chemically active in alkali and acid solutions per manufacturer's
data. The silicon coatings on the steam generator do not need
further consideration as the replacement steam generators do not
have a coating. WBN has not removed the contribution of this
coating from head loss used to design the new strainer. This
becomes margin, therefore, chemical considerations from these
coatings do not need further consideration. The other coating
type in containment is alkyds. This paint does not have a high
resistance to acidic or alkali solutions. While this is the
case, the sump pH is moderately acidic at the start of an
accident but the pH rapidly rises to a mild alkali. This would ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-9 limit chemical effects as would the low sump temperature. The total amount of alkyd paint in containment is 44 pounds. This
small amount in conjunction with low quantity of fiber and the
large strainer area is sized to prevent the formation of a
uniform fiber bed. There will be no measurable effect on head
loss due to alkyd based chemical effects. The alkyd coatings are
already assumed to be a debris source as are the other coatings.
As such, these coatings are accounted for in both head loss and
downstream effects considerations. If the coatings were assumed
to stay on the equipment or structure which the coatings were
applied and were not a debris source, the chemical material could
add to the debris loading as is the case with aluminum, where
absent the chemical consideration there would not be an aluminum
debris term. Given the low quantity and the fact all of the
alkyd coatings are considered as debris, further considerations
from a chemical effects standpoint are not needed.
Net Positive Suction Head / Loss-of-Coolant Accident QUESTION 1 Section 2.3 of ALION-REP-TVA-2739-02, Revision 0, notes that the
maximum containment sump temperature used to establish the
available net positive suction head (NPSH) for the containment
spray pumps during the recirculation phase was 190 o F. Please provide the temperature used to establish the available NPSH for
the residual heat removal (RHR) pumps during the recirculation
phase, and justify if it is different from that used for the
spray pumps during recirculation.
RESPONSE The Westinghouse calculation FSDA-C-597, "RHR Pump NPSH," dated
November 16, 1994 is provided on the enclosed CD. Note that 190
degrees F was used in the analysis. Please note this document is considered "Proprietary Information" as classified by Westinghouse Electric company, LLC for which withholding is being
requested under 10 CFR 2.390(b)(4). Enclosure 2 to this letter
provides the required Affidavit (CAW-06-2163).
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-10 QUESTION 2 Please summarize the methodology and assumptions used to
determine the maximum sump pool water temperature at the
initiation of sump recirculation. Please justify if there is a
deviation of this temperature from the calculated maximum
containment temperature following a LOCA. If such calculation
assumptions were used to maximize containment pressure, please
explain the effect of such assumptions on containment
temperature.
RESPONSE The maximum sump temperature at recirculation was determined from
the containment analysis performed for the replacement steam
generator project. This analysis maximizes initial conditions to
determine the worst case containment pressure for the ultimate
heat sink worst case summer conditions. Some of the assumptions
which maximize containment and sump temperature include: Single train containment spray, Conservative core decay heat, Minimum technical specification ice mass, Single train RHR core cooling, Large break LOCA maximizing initial energy release, Single air return fan operation, Minimum RWST level and Maximum RWST temperature, Conservatively low heat sink area and mass, Maximum steam generator water inventory (includes uncertainty), Conservative RHR and Containment Spray heat exchanger coeffient, [UA], assumptions.
These assumptions are described in TVA's letter to NRC dated
June 7, 2006, "Watts Bar Nuclear Plant (WBN) - Unit 1 - Technical
Specification (TS) Change No. WBN-TS-05 Ice Condenser Ice
Weight Increase Due to Replacement Steam Generators -
Supplemental Information - (TAC No. MC 9270)"
The containment sump liquid temperature is not in equilibrium
with the containment atmosphere temperature at the time of
switchover to sump recirculation. This is due in part to the ice
condenser design where ice melt water mixes with the break
discharge and the containment spray drainage in the lower
compartment and active sump region. Maximum containment ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-11 atmosphere temperatures are approximately 240 degrees F reducing to 200 degrees F at approximately 800 seconds whereas maximum
containment sump liquid temperature is approximately 190 degrees
F reducing to approximately 165 degrees F at 800 seconds. The
cold ice melt water exits the ice condenser and falls through the
lower compartment picking up energy and cooling the air/steam
mixture but does not attain full equilibrium prior to reaching
the sump. The efficacy of this process is based on NRC approved
Westinghouse ice condenser containment analysis models including
scale test results.
For WBN recirculation occurs around 1600 seconds. The
assumptions for NPSH are based on sump temperatures at the
beginning of the event (190 degrees F) which is nearly identical
to the lower compartment temperature at 1600 seconds (191 degrees
F). At the same time the predicted sump temperature is less than
165 degrees F. Therefore, it can be concluded that a
conservative temperature is used when comparing to actual sump
temperature at recirculation or it can be concluded that a
consistent temperature is used when comparing to lower
compartment temperature at sump recirculation.
In summary, the analysis is based on one train of ECCS and
containment spray which minimizes containment heat removal. The
analysis assumes that the ultimate heat sink (river) stays at the
technical specification maximum temperature for the entire 30 day
period. This is a very conservative assumption because it also
minimizes containment heat removal. The RWST temperature was at
the technical specification maximum of 105 degrees F. The amount
of ice in the ice condenser was assumed to be at the minimum
safety limit value. The assumptions were set to produce the
maximum containment pressure. Other assumptions made were that
ECCS spill water is at the injection temperature and there is
limited heating of the ice melt water. Changing these last two
assumptions would increase sump temperature. Because of the ice
condenser containment design these assumptions have a minimal
impact on sump temperature at the initiation of sump
recirculation. The containment spray in the upper compartment is
not condensing steam. The spray flow does not provide heat
removal from the containment as long as ice is present. This
water enters the sump at the RWST temperature. The ice condenser
melt water even with different assumptions enters the sump pool
at a temperature below 100 degrees F. These are the two largest
sources of water prior to sump recirculation. The NPSH
calculations were performed assuming a sump temperature of 190
degrees F. The sump temperature at the time of switchover is ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-12 less than 190 degrees F and is decreasing. Containment temperature at this time is approximately 190 degrees F.
QUESTION 3 Please provide copies of the following calculation reports
referenced in Section 2.5 of ALION-REP-TVA-2739-02, Revision 0:
- N2-72-4001, R Containment Spray System
- N3-74-4001, R12 - RHR System
- Watts Bar calculation EPM-RCP-120291 Revision 2, Containment Spray Pump Net Positive Head (NPSH) Calculation.
- Westinghouse calculation FSDA-C-597 dated 11/6/94 - RHR Pump NPSH. RESPONSE The requested information is provided on the enclosed CD. Note that the containment spray system description (N3-72-4001) is at
a higher revision level than requested. The changes are noted in
the revision log.
Debris Transport QUESTION 1 Please provide ALION's FLOW-3D Version 9 executable and the corresponding input deck for the Watts Bar analysis.
RESPONSE This executable was separately provided for NRC use by ALION. It is TVA's understanding that ALION considers this information to
be proprietary.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-13 Downstream Effects (Core)
These questions refer to the Watts Bar downstream effects calculations found in calculation CN-CSA-05-36, Fuel Evaluation:
QUESTION 1 Page 5 states that a fiber bed of less than 0.125 inch at the core inlet is acceptable. Page 40 states that a 7-foot head loss
is predicted for a 1/8-inch fiber bed. What head loss would be
produced at the core inlet following a large cold leg break?
Please explain and justify whether adequate flow to the core
would be provided with this head loss.
RESPONSE The 1/8-inch fiber bed is a pass/fail criterion based, in part, on NRC calculations of head loss in the Safety Evaluation of Nuclear Energy Institute (NEI) 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology. The 7-foot head loss approximates the maximum water head in the downcomer for a cold
leg break. No attempt was made to calculate the head loss
associated with the 1/8-inch fiber bed.
QUESTION 2 Page 7 states that 95 percent of fibrous material would be trapped in the bottom fuel nozzle and that the remaining 5
percent is assumed to be returned to the sump. This assumption
is stated to be based on the similarity of the dimensions of the
flow path through the sump screen and the dimensions through the
screen at the bottom of the fuel.
- a. Please provide drawings of the fuel element inlet screens showing the dimensions of the flow path into the fuel.
- b. Provide comparisons of the dimensions of the sump screen holes to the debris screen at the inlet at the fuel
elements.
RESPONSE Drawings of the fuel design are shown in the Updated Final Safety Analysis Report (UFSAR), Section 4.2. Detailed drawings of the ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-14 fuel are considered proprietary to the fuel vendor and have been separately discussed with NRC staff on behalf of WBN on
June 8-9, 2006. It is TVA's understanding that information was
sufficient to disposition this RAI question.
QUESTION 3 Page 10 lists the volume concentration for 3M fiberglass passing through the sump screens as 2.351e-3 and the total fibrous
concentration to be 2.559e-3. Page 5 of calculation CN-CSA-05-14
lists the fibrous concentration passing through the sump screens
as 5 parts per million. Please relate these quantities.
RESPONSE The volume concentration on page 10 is an initial concentration value used in the fuel evaluation, and is the ratio of fiber
volume to sump pool volume assuming no sump screen filtering has
occurred.
The mass concentration on page 5 of CN-CSA-05-14 is used in the
wear and abrasion calculations, and is the ratio of fibrous mass
to sump pool water mass assuming that the entire fiber load has
passed through the sump screen one time. Therefore, the fiber
mass is multiplied by 0.05, and then divided by the sump pool
water mass.
QUESTION 4 Page 10 states that decay heat is based on American Nuclear Society (ANS) Standards 79 with 2 Since this is a LOCA calculation, please explain why the decay heat was not calculated
using ANS Standard 71 + 20 percent to be consistent with Appendix
K to Title 10 Code of Federal Regulations Part 50.
RESPONSE The evaluation method employed is based on the decay heat rate at the time of ECCS switchover from RWST injection to containment
sump recirculation and maintains the decay heat as a constant for
the time period considered in the calculation. While the
American Nuclear Standard (ANS) 1979 + 2 sigma decay heat curve
may slightly under-predict the head load for a cold leg break at
the time of switchover from RWST injection to sump recirculation, ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-15 the use of constant decay heat at time of switchover is conservative for the overall calculation.
QUESTION 5 Page 17 shows that following a hot leg break, the fiber bed at the core inlet will exceed the 1/8-inch acceptance criterion
within the first hour of recirculation. Please explain the
effect of this condition on the core. Describe alternate flow
paths for water to reach the core. Describe the transport and
deposition of debris through these alternate flow paths.
RESPONSE Alternate flow paths are currently being considered generically under a PWR Owners Group program. Calculations such as those
suggested by NRC have not been performed at this time.
QUESTION 6 The staff plans to perform audit calculations using the TRACE code to evaluate flow of water to the core through alternate flow
paths in the event that the core inlet becomes blocked. Please
provide the staff with the location and dimensions of any
alternate flow paths through which water could reach the core
under these circumstances. Provide the height of flow holes
above the bottom of the core as well as their radial distribution
about the core periphery.
RESPONSE This information was also provided to NRC by Westinghouse in a separate audit meeting on June 8-9, 2006, on behalf of WBN. It
is TVA's understanding that Westinghouse considers this
information to be proprietary.
QUESTION 7 Pages 18 and 19 show the depletion of fibrous material in the
recirculating water for hot and cold leg breaks. A range of 97
percent to 95 percent depletion on the sump screens and a range
of 95 percent to 50 percent depletion on the fuel screens is
assumed. The depletion fraction is assumed to remain constant ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-16 with time for each cycle as the recirculating water passes the screens. Please explain whether a fiber so short or a particle
so small that it can pass through the sump screen and the fuel
inlet screens once, will also pass through the sump screens and
fuel inlet screens for sequent recirculation passes. Please
justify your assumptions.
RESPONSE As previously noted, in the NRC's safety evaluation for NEI 04-07, debris that is larger than the sump screen holes may
pass through the sump screens. This may be due to orientation of
the debris as it passes through the screen, or due to deformation
of debris enabling it to pass through the screen. Thus, not all
debris that passes through the sump screen is "too small" to be
collected on subsequent passes. Therefore, this debris may be
filtered in subsequent passes. This approach provides for a
conservative estimation of the collection of fibrous debris.
QUESTION 8 Pages 36 and 37 state that the fuel assembly support grids
typically have flow dimensions of 0.04 to 0.115 inches. How do
these dimensions compare with those of the Watts Bar fuel? Page
37 further states that the support grids may cause a fiber bed to
form across a given elevation to resemble a bed forming across a
flat plate. Please explain how the trapping of debris within the
support grids and the resulting effect on core heat transfer has
been evaluated for Watts Bar. In particular, consider the
possibility that a layer of debris and steam forms between a fuel
rod and the adjacent support grid so as to prevent water from
contacting the fuel rod surface within the support grid. Please
explain whether excessive local temperatures would be encountered
in this scenario.
RESPONSE With a 95 percent capture efficiency at the fuel bottom nozzle, this location became limiting with respect to
potential blockage. An assumed capture efficiency of 95
percent for the sump screen allows only 5 percent of the
debris reaching the screen to pass through it. If fuel
nozzle is taken to be 95 percent efficient at capturing
debris, then only 0.25 percent of the debris that reaches
the sump screen is available to the fuel. With the 50 ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-17 percent to 95 percent capture of fibrous debris on the fuel bottom nozzle, only 0.125 percent to 0.0125 percent of the
fiber reaching the sump screen was available to the first
support grid of the fuel. This amount of fibrous debris
available for collection along the active length of the fuel
is very small compared to the amount of fibrous debris that
might be collected on the fuel bottom nozzle. With the
large depletion of fibrous debris at the sump screen and the
bottom nozzle, there was insufficient remaining fibrous
debris to evaluate blockage higher in the fuel. As noted
above, the effects of potential debris collection behind
spacer grids is currently being considered generically under
a PWR Owners Group program.
QUESTION 9 Pages 43 through 47 evaluate the potential of particulate material such as reflective metal fragments, concrete, latent
containment debris and paint chips to flow into the core. It is
generally concluded that this material will not reach the core, but will settle out in the lower plenum of the reactor vessel.
Please provide an evaluation of the potential to clog the core
inlet due to filling the lower reactor vessel with a volume of
debris.
RESPONSE The debris ingestion evaluation determined that the total volume of particulate and coatings debris that may pass into the
containment sump following switchover of the ECCS is
approximately 16 cubic feet (ft 3). Based on a preliminary evaluation, the volume of the reactor vessel lower plenum below
the core support plate is calculated to be approximately 612 ft 3 for Watts Bar. If all of the approximately 16 ft 3 of the debris entering the ECCS downstream of the screen is conservatively
assumed to settle out in the lower plenum and assuming that the
debris bed is twice the theoretical volume of the debris (i.e.,
32 ft 3), the free volume of the lower plenum would not be challenged by blockage and a sufficient area would remain open to
provide for continued flow into the core.
In addition, the velocity in the lower vessel plenum is higher
than the velocity in the sump pool. Any material that will drop
out in the reactor vessel will settle out in the sump pool and
not make it to the reactor vessel.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-18 QUESTION 10 Page 43 refers to recent internal studies using disk-like particulates of various shapes with a specific gravity of 1.6.
These studies were reported to have shown that particulates
having a characteristic length of about 70 mils and thickness of
5 mils or greater would settle out in a reactor vessel lower
plenum. Please provide documentation for this study describing
the test apparatus and procedures. What vertical velocities were
used?
RESPONSE TVA will provide the documentation for the recent internal
Westinghouse studies using disk-like particulates of various
shapes in the supplemental response.
QUESTION 11 Page 47 states that coating debris no larger than 0.02 inch are
expected to be transported through the fuel. Although this
statement may be true for hot leg breaks, it would not be true
for large cold leg breaks where the boiling process would cause
this material to congregate in the core. Please provide the
results of an evaluation of the effect of paint debris on core
boiling heat transfer, including the effect of reaction products
from the mix of chemicals which would be concentrated in the core
by the boiling process following a cold leg break. The effect of
the high-radiation field within the core on the chemical and
physical nature of the mixture within the core needs to be
considered. The potential for heat transfer loss from a chemical
film that might form or be plated out by the boiling process
needs to be evaluated. Please justify that adequate heat
transfer will be maintained during the long-term cooling period.
RESPONSE There is very little material in the core at the initiation of hot leg recirculation besides boron and sodium tetraborate.
There are less than 1.5 pounds of chemical compounds formed at
three hours into the accident. This is negligible compared to
the amount of boron. This 1.5 pounds is present in the entire
sump pool not just the core. Even assuming all 1.5 pounds was in ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-19 the core at time zero and dissolved in the water, it would have no effect on heat transfer. As a point of comparison, there is
over 19,000 pounds of boron in the sump water. The boron is
dissolved in the water not suspended solids. Boron plates out on
the fuel when the water boils at the clad surface. The effect of
boron was reviewed by the NRC for WBN very recently during the
license amendment (Amendment 40) associated with the inclusion of
the tritium production rods. The coatings are particulates in
suspension not dissolved which would stay in the liquid. The
inorganic zinc coating fails as a small particulate. The
phenolic fails as chips. The chips will either settle out on the
floor of the containment or be trapped on the sump screen. There
will be virtually no phenolics in the core. The silicon coating
on the steam generators is being eliminated with the installation
of the replacement generators. The new generators are being
installed during the Fall 2006 outage and thus this coating
material is not an issue for the core. The remaining coating is
an alkyd which is present in very small quantities. There is a
total of 44 pounds of this material. Therefore, it is TVA's
conclusion that coatings in the core in the first 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> are not
of a concern.
Also, the following conservatisms are not directly credited but
should be considered before pursuing this issue further:
- 1. It was assumed that all coatings failed and were available for transport to the sump strainers for evaluation of head
loss. Most coatings either inside or outside of the ZOI of
the break, will not fail whether qualified or not.
- 2. The high energy jet from a large break has a very short duration. In a fraction of a second, the RCS pressure will
drop 500 pounds per square inch (psi). By ten seconds, the
pressure is well below 1000 psi and by 30 seconds the RCS
pressure is about 100 psi.
- 3. All of the RCS piping has been rigorously analyzed. While it has been assumed that a break can occur at any location, the reality is that if a break were to occur, it would be
where stresses are high. These are known and stresses are
going to be at welds to the reactor vessel, coolant pumps, or steam generator.
- 4. The piping is restrained in these locations such that if a break occurred the ends of the break will not fully offset.
Much of the jet energy will be dissipated on the pipe ends ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-20 and the reactor coolant pumps, steam generators, or reactor vessel.
For large breaks, there is no sustained jet to strip coatings.
In addition, the ends of the pipes will not offset to direct a
jet directly at floors or walls. Small breaks that can have a
sustained jet are small and impact a very small area. Such small
breaks may erode coatings but the affected area will be small and
the coatings debris generated is also small. Few of the coatings
are chemically reactive. Thus, the conservatism in the modeling
of coating debris is large.
QUESTION 12 Please provide an evaluation of the concentration of various materials that would occur following a large cold leg break under
the conditions that water enters the bottom of the core and is
boiled leaving all dissolved and suspended material behind.
Consider that hot leg injection begins at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the
accident. Consider all the constituents within the ECCS water
including boric acid, containment spray buffering agents, paint
and fibrous debris.
- a. Provide graphs showing the concentration of each constituent as a function of time.
- b. Concentration of material within the reactor core will depend on the water volume that is assumed to be available
for mixing. Since the core will be boiling at low pressure
it will be in a highly voided condition as will the upper
plenum. Please provide and justify the values used for core
void fraction and upper plenum void fraction used in the
concentration analysis. Provide justification for the
fraction of the lower plenum volume, which is included, as
well as for any other contribution to the total mixing
volume.
- c. Provide the flow rates into the reactor system as a function of time during cold leg recirculation and during hot leg
recirculation.
- d. Provide and justify the concentrations flowing into the reactor core as a function of time for each constituent in
the ECCS water for both cold leg and hot leg recirculation.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-21 Consider boric acid, containment spray buffering solution, paint debris, and fibrous debris.
RESPONSE The PWR Owners Group Program is looking generically at post-LOCA
core heat transfer, however based on the response to Question 4
above, TVA does not consider chemical effects on core heat
transfer to be an issue for WBN.
QUESTION 13 Following the initiation of hot leg recirculation, material which passes through the sump screen will be available to flow to the
reactor core from the top. Please provide a comparison of flow
restrictions at the top of the core including the fuel elements
to that of the sump screens.
RESPONSE Drawings of the fuel design are included in the UFSAR. Detailed drawings of the fuel are considered proprietary to the fuel
vendor and have been separately discussed with the NRC staff on
behalf of WBN on June 8 and 9, 2006. If WBN can facilitate
additional data transfer please let us know.
Head Loss Testing
QUESTION 1 Please provide the Sequoyah head loss test report that may provide validation that the paint chips would not have
transported in the Watts Bar tests had the flow velocities been
more prototypical.
RESPONSE The Sequoyah Nuclear Plant test report is enclosed on the CD for
your information.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-22 QUESTION 2 Please provide the paint chip specification parameters used in
the cell floor drain analyses, specifically the floor tumbling
velocity and the settling velocity for the turbulence model.
RESPONSE Test data on the transport metrics for paint chips is limited.
However, in general, the available test data shows that the
tumbling and settling velocities of paint chips are more
dependent on the paint chip thickness and density than the paint
chip size.
NUREG/CR-6772, "GSI-191: Separate-Effects Characterization of
Debris Transport in Water," dated August 31, 2002, (Section
3.3.2) discusses tests that were performed on epoxy-based paint
chips ranging in size from 1-inch x 1/2-inch to 1/8-inch x 1/8-
inch. The chips were approximately 15 mils thick. The results
of the tests showed that the chips first started tumbling at 0.4
feet per second (ft/s). At 0.45 ft/s bulk motion occurred and at
0.5 ft/s transport was almost instantaneous. The settling
velocity for these chips was also reported as a single value
(0.15 ft/s).
Testing performed by ALION (formerly ITSC) in 1999 for
Fitzpatrick Nuclear Power Plant showed similar results for the
settling of 5 mil thick epoxy paint chips. The chips ranged in
size, but had an approximate settling velocity of 0.08 ft/s.
QUESTION 3 Please provide an evaluation of the 3M fiber glass insulation to
justify why other fiber surrogate material can be used to
represent the 3M fiber glass in the head loss test.
RESPONSE No surrogate material was used for 3M-M20C fiber glass insulation
in the head loss test performed for the WBN strainers. The test
used 3M. Therefore, an evaluation of surrogate material is not
required.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-23 Downstream Effects (Component)
QUESTION 1 Please provide the downstream component hardware change plan, design and completion report.
RESPONSE A hardware change that resulted from the downstream effects
evaluation is to replace the orifice in 1-FE-63-170 in the common
header downstream of the centrifugal charging pumps but upstream
of the four boron injection line throttle valves. This will add
head loss upstream of the individual branch line orifices and the
branch line throttle valves to reduce the pressure drop required
across the throttle valves and allow the valves to opened such
that potential blockage of the valves does not occur.
Please see the Westinghouse report, LTR-SEE-06-118, Revision 1, Watts Bar ECCS Analysis Report, dated June 21, 2006, on the
enclosed CD. Please note this document is considered "Proprietary Information" as classified by Westinghouse Electric company, LLC for which withholding is being requested under
10 CFR 2.390(b)(4). Enclosure 2 to this letter provides the
required Affidavit (CAW-06-2169).
QUESTION 2 C hemical Considerations
- a. During the ICET, in certain chemical environments such as sodium tetraborate, precipitates formed as the solution
cooled from the 140 o F test temperature. These products could interact with other downstream debris to cause
clogging in narrow passages of downstream components such as
valves and pump internals, or affect internal surfaces of
heat exchangers or the reactor vessel. Describe your
evaluation of potential downstream effects related to
interaction with chemical products and the criteria used to
determine that performance of downstream components is
acceptable for your plant-specific chemical products and
debris combination.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-24 RESPONSE The chemical constituents are considered as part of the downstream effects. TVA notes that the amount of chemical
constituents is small compared to the other debris.
The total amount of chemical precipitants is less than 45 pounds. The latent dirt weight is 170 pounds. The design basis coating debris is over 2000 pounds.
As a comparison, the amount of boron in the sump water is 19,000
pounds. The chemical precipitants quantity is so low that heat
transfer in heat exchangers will not be degraded. Similarly the
surface area of the reactor vessel and main loop RCS piping is
sufficiently large that any film layer will be so thin that it
will not have any affect. Also, the reactor vessel is stainless
steel. If the chemicals will plate out on the vessel, it will
also plate out on the RMI insulation that has a huge surface area
as well as the outside of RCS piping, and the sump strainer
surface. This means that the film layer if it exists, will be so
thin as to have no impact on the operation of equipment
downstream of the sump screen.
- b. Explain how the interaction of downstream chemical effects combined with debris will be evaluated.
RESPONSE The chemicals are considered as one of several particulate debris types. There are no materials in the downstream that are not
present in the sump pool at large. The chemical precipitants are
such a small fraction of the total debris source term that it
will not result in a different behavior. There is no indication
that chemical precipitants bind to other debris present in WBN
containment. However, should chemical precipitants bind to other
debris, it would occur in the sump pool. This would result in
somewhat larger particles. This would make the particles more
likely to settle out or not make it through the strainer to have
a downstream effect. Particles that are small enough to go
through the strainer are too small to be captured downstream.
The chemical precipitants are being considered as a downstream
particulate with other particles that will go through the
strainer.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-25 QUESTION 3 Throttle Valves
- a. The TVA response to NRC GL 2004-02 dated September 1, 2005, indicated that an updated evaluation will be performed
following final selection of strainer design and that the
conclusions will be provided in a supplemental response.
Describe the approach, including testing program, and
schedule to finalize throttle valve positions/openings.
RESPONSE Based on a walkdown of a sample of Unit 1 throttle valves, a position or number of turns open was determined by correlating
the valve stem position measured, to the valve stem position
measurements on the valve drawing.
As a result of this walkdown, it could not be concluded that the
boron injection valves were open such that the valve blockage
would not occur. Steps have been taken to reposition these
valves such that adequate clearance is available for potential
debris passing through the sump strainers as discussed in the
response to Question 1 of this section. The valve position will
be recorded during the performance of 1-SI-63-905, "Boron
Injection Check Valve Flow Test During Refueling Outages," with
the current acceptance criterion and an additional acceptance
criteria that the valves be open at least 1.5 turns. This
approach and applicable procedures will be finalized during the
Design Change process for the installation of the orifice.
The remaining valves could be positively confirmed to have an
adequate clearance such that blockage of these valves due to
debris passing through the sump strainers would not occur. The
final calculation has not been issued since it is impacted by the
revised debris generation results. As previously committed, the
final results will be provided in a supplemental response.
The strainer design includes a hole size of 0.085-inch diameter.
An additional orifice is being added into the centrifugal
charging pump injection line to permit the throttle valves to be
opened wider. The new throttle valve position will provide an
opening that is more than 1.15 times greater than the strainer
hole size. Thus, the throttle valve will not be subject to
potential blockage from strainer bypass.
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-26 b. Explain how NRC Information Notice 96-27, and the recent NRC Throttle Valve Testing (NUREG/CR-6902), when available, will
be considered in the throttle valve evaluation.
RESPONSE With the modification discussed above, all valves will have
adequate clearance such that the valves will not be susceptible
to debris blockage. If the debris is small enough to pass
through the 0.085-inch nominal diameter hole in the sump
strainers, the debris will also pass through the throttle valves
given that an additional acceptance criteria for valve position
is required. As a result, NRC Information Notice 96-27, "Potential Clogging of High Pressure Safety Injection Throttle
Valves During Recirculation" and NUREG/CR-6902, "Effects of
Insulation Debris on Throttle-Valve Flow Performance," have been
addressed since no credit is being given for ECCS pumps
pulverizing the material that passes through the sump strainers.
QUESTION 4 Methodology
- a. The TVA response to GL 2004-02 dated September 1, 2005, indicated that the evaluation of downstream effects is
consistent with the Westinghouse Commercial Atomic Power (WCAP) Report, WCAP-16406-P, and during the audit the
licensee confirmed that they are not taking any exceptions
to the WCAP-16406-P methodology. The NRC staff has
outstanding questions (NRC letter dated October 27, 2005) on
the WCAP-16406-P methodology, and has recently been
requested by the Westinghouse Owners Group to formally
review WCAP-16406-P as a topical report. Explain how you
plan to address comments that result in a revision or
addendum to the methodology for topics such as:
- Validation of potential non-conservative assumptions,
- Conservatism to account for uncertainties,
- Wear rates correlated to testing data,
- Debris adhesion to solid surfaces, and
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1-27
- Downstream matting effect.
RESPONSE Westinghouse has responded to the questions asked the NRC's
October 27, 2006 letter. These responses were discussed with NRC
staff members during a technical exchange meeting held in May, 2006. It is Westinghouse's opinion that the responses to the
questions poised by NRC do not affect or change the evaluation
methods described in the WCAP. Rather, those responses are
considered clarificiation of the methods presented in
WCAP 16406-P
Following that meeting, a draft Revision 1 of WCAP 16406-P that
contains line-in/line-out edits was prepared and submitted to NRC
in June 2006. This submittal was made to support NRC's review
and issuance of a Safety Evaluation Report on WCAP 16406-P. NRC
is currently reviewing this document.
Sump Structure QUESTION 1 Please provide the strainer final design and structure analyses
report. If it is not available now, please indicate when it will
be available.
RESPONSE The final design and structural analysis report has received TVA
review and is being held open by the vendor pending shop
fabrication of the strainer packages. This fabrication is
currently underway. Any required design changes will be reviewed
for incorporation during this shop fabrication and assembly
process. Assembly, drawing issuance, and final analysis are
expected by the end of July. The analysis will be submitted for
your use following final issuance in the supplemental response.
ENCLOSURE 1 ATTACHMENT WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS E1A-1 LIST OF DOCUMENTS ON COMPACT DISC (CD)
Debris Generation
Question 1 TVAW001-RPT-001 - "Report on Watts Bar Unit 1 Containment Building
Walkdowns for Emergency Sump Strainer Issues," Revision 0.
Net Positive suction Head/Loss-of-Coolant Accident
Question 1 FSDA-C-597 "RHR Pump NPSH," dated November 16, 1994 (PROPRIETARY INFORMATION)
Question 3 N3-72-4001 - Containment Spray System Description, Revision 16.
N3-74-4001 - Residual Heat Removal System Description, Revision
- 12.
Watts Bar Calculation EPM-RCP-120291, "Containment Spray Pump Net
Positive Head (NPSH)," Revision 2.
Westinghouse Calculation FSDA-C-597 - See Question 1 above.
Head Loss Testing
Question 1 Sequoyah Nuclear Plant head loss test report.
Downstream Effects
Question 1 LTR-SEE-06-118, Revision 1, Watts Bar ECCS Analysis Report, dated June 21, 2006. (PROPRIETARY INFORMATION)
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS WESTINGHOUSE APPLICATIONS FOR WITHHOLDING PROPRIETARY INFORMATION CAW-06-2163 and CAW-06-2169
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 GENERIC LETTER 2004-02 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC AUDIT OF CONTAINMENT SUMP MODIFICATIONS LIST OF OPEN ITEMS E3-1 The following items remain open in the NRC Audit scope for the containment sump:
- 1. The final debris calculation will be provided as part of an update of the remaining open items.
[Break Selection and Zone of Influence Analysis, Question 1]
- 2. A hardware change that resulted from the downstream effects evaluation, is to replace the orifice in 1-FE-63-170 in the
common header downstream of the centrifugal charging pumps
but upstream of the four boron injection line throttle
valves. [Downstream Effects (Component), Question 1]
- 3. TVA will provide the documentation for the recent internal studies using disk-like particulates of various shapes in
the supplemental response. [Downstream Effects (Core), Question 10]
- 4. The final results of the Downstream Effects Calculation will be provided in a supplemental response. [Downstream Effects (Component), Question 3a]
- 5. The design and structural analysis The analysis will be submitted for your use following final issuance in the
supplemental response. [Sump Structure, Question 1]